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| issue date = 08/05/2017 | | 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 | | 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 | | author name = Haass C | ||
| author affiliation = Northwest Medical Isotopes, LLC | | author affiliation = Northwest Medical Isotopes, LLC | ||
| addressee name = | | addressee name = | ||
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. NORTHWEST MEDICAL ISOTOPES Chapter 13.0 - Accident Analysis Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 2 August2017 Prepared by: | |||
Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis , Oregon 97330 | |||
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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: | *:~*:h NWMI | ||
August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 2 | ~e *~ 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: c ~~ e.. ' ~ | |||
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis This page intentionally left blank. | |||
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....... | * NOtliTifWEST MEDfCAl ISOTOPES NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not requ ired 1 6/26/2017 Incorporate changes based on responses to NRC C. Haass Requests for Additional Information 2 8/5/2017 Mod ifications based on comments from NRC staff C. Haass | ||
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....... 13- | NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis CONTENTS 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 Radioisotope Production Facility Integrated Safety Analysis Process .... ..... ... ....... ..... ........ .......... ................... ... .......... ..... ... I 3-3 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and Risk Matrix .............. ...... .. .... ..... ...... ... ................... I 3-5 13 .1.1.2 Accident Consequence Analysis ... ........ .... .......................... .... ... ..... 13-7 13.1.1 .3 What-If and Structured What-If................... ....... ...... ... ... .............. .. 13-7 13.1.1.4 Hazards and Operability Study Method ........... ...... .......... .... ........... 13-8 13.1.1.5 Event Tree Analysis ...... .... ..... ...... ................. .. .... ..... ....... ............... 13-8 13.1 .1.6 Fault Tree Analys is ......... ........ .................................. ....... ... ........... 13-8 13.1.1.7 Failure Modes and Effects Analysis ...... .. ..... .................... .... .......... 13-8 13.1.2 Accident-Initiating Events .. ........ ..... ......... ....... .................... .. ..... ................ .... 13-8 13.1.3 Preliminary Hazards Analysis Results ............ .. ..... ....... .. ......... ...... .. ....... ... .. . 13-12 13.1.3.1 Hazard Criteria .......... .... .............. ......................... .. ..................... 13-12 13 .1.3.2 Radioisotope Production Facility Accident Sequence Evaluation ................... .......... ... ..... ....... .... .................... ........... .. .. 13-1 3 13.2 Analysis of Accidents with Radiological and Criticality Safety Consequences ............ 13-38 13.2. l Reserved ......................................... .................. ..... ........ .... .......... ...... .......... 13-39 13.2.2 Liquid Spills and Sprays with Radiological and Criticality Safety Consequences ... ........... ..... .. ......................................................................... 13-39 13 .2.2.1 Initial Conditions .... ..................................................................... 13-39 I 3 .2.2.2 Identification of Event Initiating Conditions ............. ............ .. .. ... . 13-44 I 3.2.2.3 Description of Accident Sequences ..... .... ........ ...... ......... .............. I 3-44 13.2.2.4 Function of Components or Barriers ....... ..... ................................. 13-44 13 .2.2.5 Unmitigated Likelihood ... ... ...... ............... .... .... ..... ............... ..... ... 13-45 I 3.2.2.6 Radiation Source Term .... ... ..... ....... .. ........ ...... ...... ..... .......... .... .... 13-45 13.2.2.7 Evaluation of Potential Radiological Consequences ..... ........ .. ....... 13-47 13 .2.2.8 Identification ofltems Relied on for Safety and Associated Functions ...................... .... ............. ..... .............. .............. ............. 13-50 13 .2.2.9 Mitigated Estimates ....... ...... ........... ...................... .. ..................... 13-54 13.2.3 Target Dissolver Offgas Accidents with Radiological Consequences .... ........ 13-54 13 .2.3. 1 Initial Conditions ............................................. ........... .. ........ ....... I 3-55 13 .2.3.2 Identification of Event Initiating Conditions ............ .. ......... .. ........ 13-56 13 .2.3.3 Description of Accident Sequences ............... ....... .... ...... ...... ........ 13-56 13.2.3.4 Function of Components or Barriers ............... ..... .................. ... .... 13 -56 13 .2.3.5 Unmitigated Likelihood ... ............ ........................ .... .................... 13-56 I 3.2.3.6 Radiation Source Term ................................. .. ................... ...... .... I 3-57 13.2.3.7 Evaluation of Potential Radiological Consequences ...................... 13-57 13 .2.3.8 Identification ofltems Relied on for Safety and Associated Functions ....... ............ ................................... .... ........................... 13-58 13.2.3.9 Mitigated Estimates ... .... ... ............ ............... ......... .. ... .......... .. .. .... 13-59 13-i | ||
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' ~* * ~ NORTHWUT MEDtCAl ISOTOPES 13.2.4 Leaks into Auxiliary Services or Systems with Radiological and Criticality Safety Consequences .... .............................................................. . 13-59 13.2.4.1 Initial Conditions ............................................................. ... ......... 13-59 13.2.4.2 Identification of Event Initiating Conditions ................................. 13-63 13 .2.4.3 Description of Accident Sequences .................... ........ ... , ........ ...... 13-64 13 .2.4.4 Function of Components or Barriers .. ... ............. ........................... 13-64 13 .2.4.5 Unmitigated Likelihood ..... ............. ..... ........ ........................... ..... 13-64 13.2.4.6 Radiation Source Term ......... ... ... .................... ............................. 13-65 13 .2.4.7 Evaluation of Potential Radiological Consequences ................. ..... 13-65 13.2.4.8 Identification ofltems Relied on for Safety and Associated Functions ......... ..... ...... ...... ........ ..................... ........ .. ............... ..... 13-65 13.2.4.9 Mitigated Estimates ..... .. .. ... ... ......... ............. .................. .............. 13-69 13 .2.5 Loss of Power .. ............................................. ...... .............. .. ............. ... ......... 13-69 13.2.5.1 Initial Conditions ......................................................................... 13-69 13 .2.5.2 Identification of Event Initiating Conditions .................... .... ......... 13-69 13.2.5.3 Description of Accident Sequences ........... .......... .. ... ... .... ............. 13-69 13.2.5.4 Function of Components or Barriers ............................................. 13-70 13.2.5.5 Unmitigated Likelihood .. ........ .................. ... ................................ 13-70 13.2.5.6 Radiation Source Term ................................................................ 13-70 13.2.5.7 Evaluation of Potential Radiological Consequences ............. ......... 13-70 13.2.5.8 Identification ofltems Relied on for Safety and Associated Functions ................................ .... ................. ........ ................. ... .... 13-70 13.2.6 Natural Phenomena Events .............. ..... ... .............................. ....................... 13-71 13.2.6. l Tornado Impact on Facility and Structures, Systems, and Components ........ ......... .............................................. ................ .. 13-71 13.2.6.2 High Straight-Line Winds Impact the Facility and Structures, Systems, and Components ....... .... ............ ..................................... 13-72 13 .2.6.3 Heavy Rain Impact on Facility and Structures, Systems, and Components ..... .. ................................. ........... .... .... ...................... 13-72 13.2.6.4 Flooding Impact to the Facility and Structures, Systems, and Components ............................. ............ ............................. ....... ... . 13-73 13.2.6.5 Seismic Impact to the Facility and Structures, Systems, and Components ............................................ ....... .. .. ...... .. .................. 13-73 13.2.6.6 Heavy Snow Fall or Ice Buildup on Facility and Structures, Systems, and Components ............................. .............. ...... ........... 13-74 13.2.7 Other Accidents Analyzed ....... ....... .... ........... .................................. ...... ....... 13-75 13.2.7.1 Items Relied on for Safety for Radiological Accident Sequences (S .R.) ..... ............................ ....... .................................. 13-85 13.2. 7.2 Items Relied on for Safety for Criticality Accident Sequences (S.C.) ........................... ....................... ...... .......... ..... .................... 13-87 13.2.7.3 Items Relied on for Safety for Fire or Explosion Accident Sequences (S.F.) ................. ... .......... ........... ................. ................ 13-93 13.2.7.4 Items Relied on for Safety for Natural Phenomena Accident Sequences (S.N.) ............ ......... .......... .. ...................................... ... 13-93 13.2.7.5 Items Relied on for Safety for Man-Made Accident Sequences (S.M.) .. ...... .... .. .................................... ........ ................................ 13-94 13.2.7.6 Items Relied on for Safety for Chemical Accident Sequences (S.CS.) ................ .. ..... ........................ ...... ......................... ...... .... . 13-94 13-ii | |||
NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis 13.3 Analysis of Accidents with Hazardous Chemicals ......... .................................... .... ..... 13-95 13.3 .1 Chemical Bum s from Contaminated Solutions During Sample Ana lysis .. ..... 13-95 13.3.1.1 Chemical Accident Description ........................................ ......... ... 13-95 13.3. 1.2 Chemi cal Accident Consequences ....... ... ...... ........ .. ..... ................. 13-95 13.3. 1.3 Chemi cal Process Controls ... .. .... ... .. .... ............................ ..... .. ...... 13-95 13.3 .1.4 Chemi cal Process Survei llance Requirements ............................... 13-95 13.3.2 Nitric Acid Fume Release .................... .... ... ......................................... .... ..... 13-96 13.3.2.1 Chemi cal Accident Description ..... .. ......... ....... ....... .. .......... .. .... .... 13-96 13.3.2.2 Chemi cal Acc ident Consequences .......................... ...................... 13-96 13.3.2.3 Chemi cal Process Controls .. ......................... ..... .. ............ ............. 13-96 13.3.2.4 Chemi cal Process Surveillance Requirements ....... .... .................... 13-96 13.4 References ............................................................................... .... .............. .. .............. 13-97 13-iii | |||
NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis FIGURES Figure 13-1. Integrated Safety Analysis Process Flow Diagram ..................................................... 13-4 Figure 13-2. Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident .................... 13-49 TABLES Table 13-1. Likelihood Categories ............................................................................................... 13-5 Table 13-2. Qualitative Likelihood Category Guidelines .............................................................. 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 ................. ..... .................. .... ................ 13-6 Table 13-5. Radioisotope Production Facility Preliminary Hazard Analysis Accident Sequence Category Designator Definitions ................................................................ 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 ......... ........ ....... ........................ ... .... ... ... ..... 13-9 Table 13-7. Radioisotope Production Facility Preliminary Hazards Analysis Primary Process Nodes and Subprocesses (2 pages) ............... .. .......................................................... 13-10 Table 13-8. Crosswalk of Radioisotope Production Facility Preliminary Hazards Analysis Process Nodes and Top-Level Accident Sequence Categories ..... ..... ... ............... ...... 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) ..................... ..... ...... ..................... 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 ............................ ..... ................................... 13-32 Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) ......... .... .... ............... .......... .. ....... ... ................ 13-33 Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) ..... ..... ................... 13-40 Table 13-18. Source Term Parameters ............................ ............................ ............ ........... ........... 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 Release Consequence Summary at 100 Meters ..... 13-49 Table 13-21. Maximum Bounding Inventory ofRadioiodine [Proprietary Information] ................ 13-55 Table 13-22. Target Dissolver Offgas Accident Total Effective Dose Equivalent.. .... ............ .. ...... 13-58 Table 13-23 . Bounding Radionuclide Liquid Stream Concentrations (4 pages) .. ... ................. .... ... 13-60 Table 13-24. Analyzed Accidents Sequences (9 pages) ..... ... ....... ........ ... ... ...... ....................... ....... 13-75 Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (2 pages) ............ ........ ... ..... ............ ..... ..................................... .... .................... .... ... 13-84 Table 13-26. Accident Sequence Category Definitions ............................. .................................... 13-85 13-v | |||
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 99 mTc technetium-99m 235 U uranium-235 241 Am americium-241 AAC augmented administrative control AC administrative control ACI American Concrete Institute AEC active engineered control 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 release fraction ASCE American Society of Civil Engineers CDE committed dose equivalent CEDE committed effective dose equivalent CFR Code of Federal 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 irradiation ET A event tree ana lysis FEMA Federal Emergency Management Agency FMEA fai lure modes and effects analysis FT A fault tree analysis HAZOP hazards and operability HEGA high-efficiency gas adsorption HEPA high-efficiency particulate air HIC high-integrity canister HN03 nitric acid HV AC heating, venti lation, and air conditioning IBC International Building Code IROFS items relied on for safety IRU iodine removal unit ISA integrated safety analysis ISG Interim Staff Guidance IX ion exchange LEU low enriched uranium LPF leak path factor MAR material at risk Mo molybdenum MURR University of Missouri Research Reactor NaOH sodium hydroxide NDA nondestructive assay NIOSH National Institute for Occupational Safety and Health NOx nitrogen oxide 13-vi | |||
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis NOAA National Oceanic and Atmospheric Administration NRC U.S . Nuclear Regulatory Commission NWMI Northwest Medical Isotopes, LLC NWS National Weather Service OSTR Oregon State University TRIGA Reactor osu Oregon State University P&ID piping and instrumentation drawing PEC passive engineered control PFD process flow diagram PHA preliminary hazards analysis PMP probable maximum precipitation QRA quantitative risk assessment RASCAL Radiological Assessment System for Consequence Analysis RF respirable fraction RPF Radioisotope Production Facility RSAC Radiological Safety Analysis Code SNM special nuclear material SSC structures, systems, and components ST source term TCE trichl oroethy lene TEDE total effective dose equivalent u uranium U.S. United States UN uranyl nitrate 13-vii | |||
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...... NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis | |||
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* NORTHWEST MlDtCAL ISOTOPES Units oc degrees Celsius OF degrees Fahrenheit Ci cune Cm centimeter ft feet ft3 cubic feet g gram hr hour in. 2 square inch kg kilogram km kilometer km2 square kilometer L liter lb pound m meter M molar m3 cubic meter mg milligram m1 mile mi2 square mile mil thousandth of an inch mm minute mrem millirem oz ounce ppm parts per million rem roentgen equivalent man sec second Sv sievert wk week wt% weight percent yr year 13-viii | |||
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 (99Mo) 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 packaged and transported to medical industry users where the radioactive decay product, technetium-99m 99 | |||
( mTc), 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 transportation activities are not analyzed in this 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 Packaging and shipping LEU targets to the university reactor network for irradiation Returning irradiated LEU targets for dissolution, recovery, and purification of 99Mo Recovering and recycling LEU to minimize radioactive, mixed, and hazardous waste generation Treating/packaging wastes generated by RPF process steps to enable transport to a disposal site Chapter Organization Section 13.1 describes hazard and accident 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 preliminary hazards analysis (PHA) (NWMI-2015-SAFETY-001, NWMI Radioisotope Production Facility Preliminary Hazards Analysis). The PHA discussion in Section 13.1.3 identifies the accident scenarios that required further evaluation. | |||
Section 13.2 presents analyses of radiological and criticality accidents, including: | |||
Section 13.2. l (Reserved) | 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 | 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 electrical power Section 13.2.6 discusses natural phenomena accidents Section 13.2. 7 identifies the additional accident sequences evaluated and associated items relied on for safety (IROFS) 13-1 | ||
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. | 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. | |||
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 | 13-2 | ||
Each event with an adverse consequence that involves licensed material or its byproducts is evaluated for | |||
13-3 | 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. | ||
develop IROFS and basis for each in complete QRA Develop PSAR, ISA summary, technical specifications | The 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 latory guidelines. | ||
Each event with an adverse consequence that involves licensed material or its byproducts is evaluated for risk 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. | |||
+---------1* license submit to NRC application | Fault trees and failure mode and effects analysis can be used to (I) provide quantitative fai lure analysis data (failure frequencies) for use in the event tree ana lysis 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. | |||
* Radiological dose* | 13-3 | ||
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* | * NWMI | ||
* 10-min exposure limit) | ~ * *! NORTHWEST MmtCAL ISOTOf'ES NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis Design and Design and Safety ISATeam Engineering NRCReview Functions Functions Deve lop process Initiate ISA process descript ions, PFDs, by collecting P& IDs preliminary data Perform PHA on Identify preliminary facility operations hazards and consequences (radiological, Categorize events criticality, chemical, for likelihood, fire, extern al) using consequence, regulato ry guides and risk where applicable l | ||
Develop CSAs, FHA, Indeter-minate, Document and other support high, or ~ identified low-risk documents intermediate events (no IROFS) risk? | |||
* Airborne, contaminated | Yes + | ||
," | Perform QRA to quantitatively evaluate risk and identify IROFS High or No intermediate risk event? | ||
* As total effective dose | Yes Design function Identify "accident Start Phase 1 development of sequence" and development of IROFS 1------++ develop IROFS and ...--. IROFS boundary specifications/ | ||
basis for each in definition packages conceptual complete QRA for each IROFS drawings Complete Phase 1 Develop PSAR, ISA development of | |||
to model bounding facility releases that affect the public) Reference to nationally recognized safety organizations (e.g., | ~----++ summary, technical IROFS boundary specifications definition packages I ISA team review and recommendation I | ||
Approved methods for evaluation of natural and man-made phenomenon and comparison to the design | for approval Management approval of ISA basis NRC review of document r----------+ - - - - - - - --1* license submit to NRC application 1cm_r Figure 13-1. Integrated Safety Analysis Process Flow Diagram 13-4 | ||
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 | |||
; 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 | NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and Risk Matrix Table 13-1 shows the accident likelihood Table 13-1. Likelihood Categories categories applied to the RPF ISA process. | ||
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. | Table 13-2 shows qualitative guidelines for Event frequency limit applying the likelihood categories from Not unlikely 3 More than I 0- 3 events per year Table 13-1. Table 13-3 shows accident consequence severity categories from Unlikely 2 Between I 0-3 and I 0- 5 events 10 CFR 70.61, "Performance Requirements." per year Table 13-4 shows the RPF risk matrix, which Highly unlikely Less than 10-5 per events per is a product of the likelihood and consequence year severity categories from Table 13-1 and Table 13-3, respectively. | ||
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 | Table 13-2. Qualitative Likelihood Category Guidelines 11.* 3 An event initiated by a human error Initiator 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 combustibles 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 process variable monitored or unmonitored by a control system 3 A valve out of position or a valve that fai ls to seat and isolate 3 Most standard industrial component failures (valves, sensors, safety devices, gauges, etc.) | ||
ETA is also used in the QRA process to demonstrate that the IROFS, | 3 An adverse chemical reaction caused by improper quantities ofreactants, out-of-date reactants, out-of-specification reaction environment, or the wrong reactants are used 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 external chall enges applied (e.g., spontaneous rupture of an all-welded dry nitrogen system pipe operating at or below design pressure in a clean, vibration-free environment) 1-2 An adverse chemical reaction when proper quantities of in-date chemicals are reacted in the proper environment Natural phenomenon such as tsunami, volcanos, and asteroids for the Missouri facility site 13-5 | ||
13.1.1.6 Fault Tree Analysis Fault tree analysis (FT A) is a top-down , deductive failure | |||
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 | 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 | ||
Any other event potentially related to unique faci | *iii High Consequence category 3 | ||
Workers | |||
Any other event potentially related to | * Radiological dose* > I Sv Off-site public | ||
--- | * Radiological dose* | ||
* | Environment consequence (I 00 rem) > 0.25 Sv (25 rem) | ||
* Airborne, radiologically | |||
* Toxic intake > 30 mg contaminated nitric acid soluble U | |||
> 170 ppm nitric acid (AEGL-3, | |||
* Airborne, contaminated 10-min exposure limit) nitric acid > 24 ppm | |||
* Unshieldedb nuclear criticality nitric acid (AEGL-2, 60-min exposure limit) | |||
Intermediate 2 . Radiological dose* between . Radiological dose* 24-hr radioactive consequence 0.25 Sv (25 rem) and I Sv between 0.05 Sv (5 rem) release > 5,000 x and 0.25 Sv (25 rem) | |||
. (100 rem) | |||
Airborne, radiologically | |||
* Airborne, contaminated nitric acid > 0.16 ppm Table 2 of 10 CFR 20, 0 Appendix B contaminated nitric acid | |||
> 43 ppm nitric acid (AEGL-2, nitric acid (AEGL-1, 10-min exposure limit) 60-min exposure limit) | |||
Low Accidents with lower Accidents with lower Radiological consequence radiological, chemical, and/or radiological, chemical, releases producing toxicological exposures than those and/or toxicological lower effects than above from licensed material and exposures than those above those listed above byproducts of licensed material from li censed material and from licensed byproducts of licensed material material Source: I 0 CFR 70.61 , " Performance Requirements," Code of Federal Regulations, Office of the Federal Register, as amended. | |||
* As total effective dose equivalent. | |||
b A shielded criticality accident is also cons idered a high-consequence event. | |||
c IO CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended. | |||
AEGL Acute Exposure Guideline Level. u = uranium. | |||
Table 13-4. Radioisotope Production Facility Risk Matrix Likelihood of occurrence Severity of Highly unlikely Unlikely Not unlikely consequences (Likelihood category 1) (Likelihood category 2) (Likelihood Category 3) | |||
High consequence Risk index = 3 Risk index = 6 Risk index = 9 | |||
~ | |||
(Consequence category 3) Acceptable risk Unacceptable ri sk Unacceptabl e risk Intermediate consequence Risk index = 2 Risk index= 4 Risk index = 6 (Consequence Acceptable risk Acceptable risk { | |||
Unacceptable risk category 2) | |||
Low consequence Risk index = 1 Risk index = 2 Risk index = 3 (Consequence category 1) Acceptable risk Acceptable risk Acceptable risk 13-6 | |||
:**:*;.** NWMI | |||
*:*~~!~* | |||
0 0 NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis NORTHWEST MEDICAL ISOlWES 13.1.1.2 Accident Consequence Analysis The 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 of 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., use 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 basis (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 processing High radiation dose due to accidental nuclear criticality Toxic uptake of licensed material by workers or the public during 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 | |||
NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis 13.1.1.4 Hazards and Operability Study Method 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 adverse 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 frequency given the initiator. ETA is also used in the QRA process to demonstrate that the IROFS, selected 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 analysis 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-level 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 is included in this category) | |||
Any other event potentially related to unique faci lity operations 13-8 | |||
NWM l-2013-02 1, Rev . 2 Chapter 13.0 - Accident Ana lysis The PHA (NWMl-2015-SAFETY-OOI) identifies Table 13-5. Radio isotope Production Facility and categorizes accident sequences that require Preliminary Hazard Analys is Accident further evaluation. Table 13-5 defines the top- Seq uence Category Designator Defi nitions level accident sequence notation used in the RPF PHA top-level accident PHA. sequence categorya Definition Table 13-6 provides a crosswalk between the PHA S.C. Criticality top-level accident sequence categories and the S.F. Fire or explosion NUREG-1537, Guidelines for Preparing and S.R. Radiological Reviewing Applications for the Licensing of Non-Power Reactors - Format and Content, Part 1 S.M. Man-made Interim Staff Guidance (ISG) accident initiating S.N. Natural phenomena events listed above. As noted at the bottom of S.CS. Chemical safety Table 13-6, PHA accident sequences involve one or more of the NUREG-1537 Part 1 ISG accident | |||
* The alpha category designator is fol lowed in the PHA by a two-digit number "XX" that refers to the specific accident initiating event categories, as noted by ./ in the sequence (e.g., S.C.01 , S.F.07). Specific accident sequences corresponding table cell, but the PHA accident are di scussed in Sections 13.1.3 and 13.3 . | |||
sequences themselves are not necessari ly initiated PHA = prelimi nary hazard analysis. | |||
by the ISG accident initiating event. Table 13-6 shows how PHA accident sequences correspond with ISG accident initiating events, and demonstrates that the PHA considers the full range of accident events identified in the ISG. | |||
Table 13-6. Crosswalk ofNUREG-1537 Part I Interim Staff Guidance Accident Initiating Events versus Radioisotope Production Facility Preliminary Hazards Analysis Top-Level Accident Sequence Categories NUREG-1537a Part 1 ISG accident initiating event category Criticality accident Loss of electrical power External events (meteorological, | |||
------ ,/ | |||
,/ | |||
PHA Top-Level Accident Sequence Categoryb | |||
,/ | |||
,/ | |||
,/ | |||
,/ | |||
,/ | |||
,/ | |||
,/ ,/ | |||
seismic, fire, flood) | |||
Critical equipment malfunction ,/ ,/ ,/ ,/ ,/ | |||
Operator error ,/ ,/ ,/ ,/ | |||
Facility fire (explosion is included in ,/ | |||
this category) | |||
Any other event potentially related to ,/ ,/ | |||
unique faci lity operations | |||
* NURE0-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of No n-Power Reactors - Format and Content, Part I, U.S. Nuclear Regulatory Commission, Office ofNuclear Reactor Regulation, Washington, D.C., | |||
February 1996. | |||
h PHA accident sequences involve one or more of the NURE0 -1 537 Part I ISO accident initiating event categori es, as noted by an ./ in the corresponding tabl e cell, but the PHA sequences themselves are not necessarily initiated by the ISO accident initiating event. | |||
ISO = lnterim StaffOuidance. PHA = preliminary hazard analysis. | |||
13-9 | |||
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...... NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis | |||
~* * ~ NOmfWlST MEOtC.Al ISOTOP£S 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 Target fabrication | |||
* Fresh uranium receipt and storage process | |||
* Fresh uranium dissolution | * Fresh uranium dissolution | ||
* Uranyl nitrate blending and feed preparation | * Uranyl nitrate blending and feed preparation | ||
* Nitrate extraction | * Nitrate extraction | ||
* Recycled uranyl nitrate concentration | * Recycled uranyl nitrate concentration | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* Uranium | * Uranium scrap recovery | ||
* Target assembly, | * Target assembly, loading, inspection, quality checking, verification, packaging and storage 2.0.0 Target dissolution * [Proprietary Information] | ||
* [Proprietary Information] | process * [Proprietary Information] | ||
* Primary process offgas treatment | * Primary process offgas treatment | ||
* Fission gas retention | * Fission gas retention 3.0.0 Molybdenum recovery | ||
* Feed preparation | * Feed preparation and purification process | ||
* First stage recovery | * First stage recovery | ||
* First stage purification preparation | * First stage purification preparation | ||
| Line 335: | Line 208: | ||
* Second stage purification preparation | * Second stage purification preparation | ||
* Second stage purification | * Second stage purification | ||
* Final purification adjustment | * Final purification adjustment 99 | ||
* | * Mo preparation for shipping 4.0.0 Uranium recovery and | ||
* Impure uranium lag storage | * Impure uranium lag storage recycle process | ||
* First-cycle uranium recovery | * First-cycle uranium recovery | ||
* Second-cycle uranium purification | * Second-cycle uranium purification | ||
* Product uranium lag storage | * Product uranium lag storage | ||
* Other support (storage | * Other support (storage vessels, transfer lines, solid waste handling for resin bed replacement) 13-10 | ||
* 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 | |||
* Liquid waste | *:~*;~:* NWM I | ||
* High dose liquid waste volume | ~* *~ 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 Waste handling system | ||
* Liquid waste storage process | |||
* High dose liquid waste volume reduction | |||
* Condensate storage and recycling | |||
* Concentrated high dose liquid waste storage/preparation | * Concentrated high dose liquid waste storage/preparation | ||
* Low dose liquid waste | * Low dose liquid waste volume reduction and storage | ||
* Liquid waste solidification | * Liquid waste solidification | ||
* | * Solid waste hand! ing | ||
* Waste | * Waste encapsulation | ||
* TCE solvent | * TCE solvent reclamation | ||
* Mixed waste accumulation | * Mixed waste accumulation 6.0.0 Target receipt and | ||
* Cask receipt and target unloading | * Cask receipt and target unloading disassembly process | ||
* Target Inspection | * Target Inspection | ||
* Target disassembly | * Target disassembly | ||
* [Proprietary Information] | * [Proprietary Information] | ||
* Target disassembly stations | * Target disassembly stations | ||
* Gaseous fission product control * [Proprietary Information] | * Gaseous fission product control | ||
* Empty target hardware handling * (No subprocesses identified in PHA. Ventilation system | * [Proprietary Information] | ||
* Natural phenomena | * Empty target hardware handling 7.0.0 Ventilation system * (No subprocesses identified in PHA. Ventilation system provides cascading pressure zones, a common air supply system with makeup air as necessary, heat recovery for preconditioning incoming air, and HEPA filtration.) | ||
* Man-made external events | 8.0.0 Natural phenomena, | ||
* Chemical storage and preparation areas | * Natural phenomena man-made external | ||
* Man-made external events events, and other facility | |||
* Chemical storage and preparation areas operations | |||
* On-site vehicle operation | * On-site vehicle operation | ||
* General storage, utilities , and maintenance activities | * General storage, utilities, and maintenance activities | ||
* Laboratory operations | * Laboratory operations | ||
* Hot cell support activities | |||
* Waste storage operations including packaging and shipment 99 Mo molybdenum-99 PHA preliminary hazards anal ysis. | |||
HEPA high-efficiency particulate air. TCE = trichloroethylene. | |||
Table 13-8 shows a crosswalk that identifies the | |||
[Proprietary Information] | [Proprietary Information] | ||
(eventually handled by the worker) 1.8.3 .7 Loading limits are not adhered Hi gh-dose to workers or S.R.28, Target or waste to by the operators or the the public from shipping cask not loaded closure requirements are not improperly shielded or secured according to satisfied, and the cask does cask procedure, leading to not provide | |||
[Proprietary Information] | [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 dose 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 shipment receipt streaming radiation inspection and cask preparation activities due to damaged irradiated target cask 6.1.1.10 Cask involved in in-transit High dose to workers during S.R.29, High dose to accident or targets failed during receipt inspection and workers from release of irradiation, leading to excessive opening activities gaseous radionuclides offgassing from damaged targets during cask receipt inspection | |||
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* | ...... NWMl-2013-021, Rev . 2 Chapter 13.0 -Accident Analysis | ||
' !e* ~ ' NORTHWtn MEDICAL ISOTOPU 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 double-wall barrier to contain any spills that may occur from the primary confinement piping. IROFS CS -09 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 safe-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. | |||
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. | 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 | Continuous air monitoring will be provided to alert operators of high airborne radiation levels that exceed derived air concentration (DAC) limits. | ||
13.2.2.9 Mitigated Estimates The controls selected will mitigate both the frequency and consequences of this accident. | HEPA filters on hot cell outlets are not credited and will reduce the impact of spills or sprays to the public. | ||
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 [ | 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. | ||
IR Us Pressure-relief vessel Primary adsorbers (carbon media beds for 6 days noble gas holdup) 13-54 NWM I ...... | 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. | |||
NWMJ expects the availability and operation ofIRUs will become part of the technical specification to meet annual release limits. The iodine released from | 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 [NOx] scrubber) or equipment fai lure (e.g., loss of the IRU heater) during the dissolution of irradiated targets. The primary components of the dissolver offgas include: | ||
(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 | NOx scrubbers (caustic and absorbers) | ||
[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 | IR Us Pressure-relief vessel Primary adsorbers (carbon media beds for 6 days noble gas holdup) 13-54 | ||
13.2.3.1 Initial Conditions The target dissolver and associated | |||
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...... NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis | |||
~* * ~ NOfITTfWHT MfDtCAL ISOTOP£S Iodine guard beds (remove any iodine not trapped in the IRUs) | |||
Filters Vacuum receiver tanks Vacuum pumps (draw a downstream vacuum on the target dissolver offgas treatment train) | |||
Secondary adsorbers (additional carbon media beds to hold up noble gases for an additional 60 days) | |||
The IR Us nominall y removes about 99.9 percent of the iodine in the offgas stream after the NOx scrubbers. NWMJ expects the availability and operation ofIRUs will become part of the technical specification to meet annual release limits. The iodine released from dissolution of the irradiated targets 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 NOx treatment absorbers) and end up in the high dose liquid waste tanks, and (3) the remainder of the iodine will be captured in the IRUs. | |||
These 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 ed 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 adsorbers 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 offgas Table 13-21. Maximum Bounding Inventory of treatment train are assumed to be operational Radio iodine [Proprietary Information] | |||
and in service prior to the occurrence of any Isotope accident sequence that affects the IR Us. | |||
The JR Us are assumed to be loaded with the [Proprietary Information] | |||
conservative bounding holdup inventory of [Proprietary Information] | |||
iodine, as determined in NWMI-2013-CALC- [Proprietary Information] | |||
01 I. [Proprietary In form ati on] | |||
1J 2m No credible event has been identified where 1 [Proprietary lnformati on] | |||
the total captured inventory on the IR Us would [Proprietary Information] | |||
be released. This accident evaluation is for the 133mJ [Proprietary Information] | |||
release of the iodine generated from a single | |||
[Proprietary Information] | [Proprietary Information] | ||
dissolution of [Proprietary Information]. The maximum amount of iodine [Proprietary [Proprietary Information] | |||
Information] is shown in Table 13-21. The Total I Ci [Proprietary Information] | |||
mass balance projects about 20 percent of the = iodine. | |||
iodine will stay in the dissolver solution and nearly 50 percent of the elemental iodine (h) that does volatize will be captured in the NOx 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 offgas stream going to the IR Us. | |||
13-55 | |||
is | |||
.-.;;..NWMI NWMl-2013-021, Rev. 2 | |||
* ~* * ~ NORTHWEST MEDK:Al ISOTIN'ES Chapter 13.0 - Accident Analysis Therefore, this evaluation focuses on accident sequences where the inventory at risk is that generated directly 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 offgas treatment train. The three most likely 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 NOx 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. | |||
The | 13.2.3.3 Description of Accident Sequences The accident sequences for loss ofIRU efficiency include the following. | ||
[Proprietary Information] is being dissolved. | |||
A process upset occurs that reduces the IRU efficiency by an unspecified amount. | |||
The event is 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 opens the valve to the pressure-relief tank in the dissolver offgas system to capture the offgas stream. | |||
If the initiator for the event is loss of power or the event creates a condition where vacuum in the dissolver offgas system is lost, the pressure-relief tank valve would automatically open to capture the offgas 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 IRUs will be the primary iodine capture devices; however, there will be iodine guard beds downstream of each of the primary noble gas adsorbers. The vent system piping will direct the dissolver offgas 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 offgas 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 IRUs. 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 iodine radioisotopes. Therefore, the iodine source term is the focus of this accident sequence evaluation. | |||
No credit is taken for any iodine removal in the 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. | |||
The accident | 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 that were designed to bound the credible offgas scenarios. These assumptions include: | |||
[Proprietary Information] | [Proprietary Information] | ||
Dissolver roduct Uranium se aration feed 24 1Am [Proprietary Information] | All the iodine in the targets released into the offgas system, and no iodine or noble gases captured in the NOx scrubbers or retained in the dissolver solution Iodine removal efficiency of the dissol ver offgas 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 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 will 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. | |||
13.2.3. 7 Evaluation of Potential Radiological Consequences Radiological 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 performance of the target dissolver offgas treatment train IRUs. Additional detailed information describing validation of the models, 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 inventory would 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 llowing parameters were used for model runs: | |||
Mixing depth: 400 m (1 ,312 ft) (default) | |||
Air density: 1,240 g/m 3 (1.24 ozlft 3) (sea level) | |||
Pasquill-Gifford o (NRC Regulatory Guide 1.145) | |||
No plume rise (i.e., buoyancy or stack momentum effects) 13-57 | |||
..::*...*. NWMI NWM l-2013-021, Rev. 2 Chapter 13.0 - Acci dent Analysis | |||
~* * ~ NOffTlfW'En MEDICAi. lSOTOPfS No plume depletion (wet or dry Table 13-22. Target Dissolver Offgas Accident deposition) Total Effective Dose Equivalent 2-hr release (constant release of all TEDE (rem) activity) 2-hr exposure Distance (m) ' Total ICRP-30 inhalation model 100 2.05E-01 Finite cloud immersion model 200 l.98E-OI Breathing rate: 3.42E-4 m3/sec 300 2.21E-01 (l .2E-2 ft 3/sec) (ICRP-30 heavy activity) 400 6.41 E-OI Respiratory fraction: 1.0 500 l.76E+OO Table 13-22 shows the distance-dependent total 600 3.18E+OO receptor accident doses versus distance from the 700 4.50E+OO RPF stack for 2-hr exposure. This table was 800 5.47E+OO developed using the results from the Section 19.4 1,000 6.50E+OO dose consequences and dividing by a ratio of the 1,100 6.65E+OO accident source terms. The maximum public dose 1,200 6.62E+OO is 6.65 rem at 1, 100 m. | |||
1,300 6.50E+OO RSAC 6.2 calculates inhalation doses using the 1,400 6.29E+OO ICRP-30 model with Federal Guidance Report 1,500 6.06E+OO No. 11 dose conversion factors 1,600 5.82E+OO (EPA 520/1-88-020, Limiting Values of 1,700 2.05E-OI Radionuclide Intake and Air Concentration and Peak total dose is balded and italicized. | |||
Dose Conversion Factors for Inhalation, Submersion, and Ingestion) . The committed dose TEDE = total effective dose equivalent. | |||
equivalent (CDE) is calculated for individual organs and tissues over a 50-year period after inhalation. | |||
The CDE for each organ or tissue is multiplied by the appropriate ICRP-26, Recommendations of the International Commission on Radiological Protection, weighting factor and then summed to calculate 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 (~6.65 rem) are within intermediate consequences severity categories (<25 rem). | |||
13.2.3.8 Identification of Items Relied on 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 offgas 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 | |||
*:i*~h- NWMI | |||
~ * *!' NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis IROFS RS-09, Primary Offgas Relief System As an AEC, a relief device will be provided that relieves pressure 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 The following defense-in-depth features preventing target dissolver 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 primary 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 controls 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 leaks into secondary containment as an event that could lead to an accidental nuclear criticality. The accidents covered by this analysis bound the family of accidents where highly radioactive or fissile solution leaves the hot cell or other shielded areas via auxiliary 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 primary 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 needs to be condensed. The bounding source term is expected to be the dissolvers or the feed tanks in the Mo recovery and purification system. | |||
Table 13-23 lists 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] | |||
139Ba [Proprietary Information] [Proprietary Information] | |||
i4oBa [Proprietary Information] [Proprietary Information] | |||
141ce [Proprietary Information] [Proprietary Information] | |||
143Ce [Proprietary lnformation] [Proprietary Information] | |||
144Ce [Proprietary Information] [Proprietary Information] | |||
242cm [Proprietary Information] [Proprietary Information] | |||
243Cm [Proprietary Information] [Proprietary Information] | |||
244Cm [Proprietary Information] [Proprietary Information] | |||
134Cs [Proprietary Information] [Proprietary Information] | |||
134m Cs [Proprietary Information] [Proprietary Information] | |||
136Cs [Proprietary Information] [Proprietary Information] | |||
137 [Proprietary Information] | |||
Cs [Proprietary Information] | |||
1ssEu [Proprietary Information] [Proprietary Information] | |||
1s6Eu [Proprietary Information] [Proprietary Information] | |||
1s1Eu [Proprietary Information] [Proprietary Information] | |||
1291 [Proprietary Information] [Proprietary Information] | |||
130J [Proprietary Information] [Proprietary Information] | |||
13 IJ [Proprietary Information] [Proprietary Information] | |||
1321 [Proprietary Information] [Proprietary Information] | |||
132m I [Proprietary Information] [Proprietary Information] | |||
1331 [Proprietary Information] [Proprietary Information] | |||
133mI [Proprietary Information] [Proprietary Information] | |||
1341 [Proprietary Information] [Proprietary Information] | |||
135J [Propri etary Information] [Proprietary Information] | |||
83m Kr [Proprietary Information] [Proprietary Information] | |||
85Kr [Proprietary Information] [Proprietary Information] | |||
85m Kr [Proprietary Information] [Proprietary Information] | |||
87Kr [Proprietary Information] [Proprietary Information] | |||
88Kr [Proprietary Information] [Proprietary Information] | |||
140La [Proprietary Information] [Proprietary Information] | |||
141La [Proprietary Information] [Proprietary Information] | |||
142La [Proprietary Information] [Proprietary Information] | |||
99Mo [Proprietary Information] [Proprietary Information] | |||
95Nb [Proprietary Information] [Proprietary Information] | |||
95mNb [Proprietary Information] [Proprietary Information] | |||
96Nb [Proprietary Information] [Proprietary Information] | |||
13-60 | |||
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Acci dent 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 97Nb [Proprietary Information] [Proprietary Information] | |||
97mNb [Proprietary Information] [Proprietary Information] | |||
141Nd [Proprietary Information] [Proprietary Information] | |||
236mNp [Proprietary Information] [Proprietary Information] | |||
231Np [Proprietary Information] [Proprietary Information] | |||
23sNp [Proprietary Information] [Proprietary Information] | |||
239Np [Proprietary Information] [Proprietary Information] | |||
233Pa [Proprietary Information] [Proprietary Information] | |||
234pa [Proprietary Information] [Proprietary Information] | |||
234mPa [Proprietary Information] [Proprietary Information] | |||
11 2pd [Proprietary Information] [Proprietary Information] | |||
141Pm [Proprietary Information] [Proprietary In formation] | |||
148Pm [Proprietary Information] [Proprietary Information] | |||
I48mpm [Proprietary Information] [Proprietary Information] | |||
149Pm [Proprietary Information] [Proprietary Information] | |||
ISOpm [Proprietary Information] [Proprietary Information] | |||
ISIPm [Proprietary Information] [Proprietary Information] | |||
142Pr [Proprietary Information] [Proprietary Information] | |||
143Pr [Proprietary Information] [Proprietary Information] | |||
I44pr [Proprietary Information] [Proprietary Information] | |||
144mpr [Proprietary Information] [Proprietary Information] | |||
I4Spr [Proprietary In formation] [Proprietary Information] | |||
2Jspu [Proprietary Information] [Proprietary Information] | |||
239Pu [Proprietary Information] [Proprietary Information] | |||
240pu [Proprietary Information] [Proprietary Information] | |||
24 1Pu [Proprietary Informati on] [Proprietary Information] | |||
103mRh [Proprietary Information] [Proprietary Information] | |||
IOSRh [Proprietary Informati on] [Proprietary Information] | |||
106Rh [Proprietary Information] [Proprietary Information] | |||
J06mRh [Proprietary Information] [Proprietary Information] | |||
103Ru [Proprietary Information] [Proprietary Information] | |||
1osRu [Proprietary Information] [Proprietary Information] | |||
106Ru [Proprietary Information] [Proprietary Information] | |||
122 sb [Proprietary Information] [Proprietary Information] | |||
124 Sb [Proprietary Information] [Proprietary Information] | |||
125 Sb [Proprietary Information] [Proprietary Information] | |||
126Sb [Proprietary Information] [Proprietary Information] | |||
127 Sb [Proprietary Information] [Proprietary Information] | |||
128 Sb [Proprietary Information] [Proprietary Information] | |||
13-61 | |||
NWMl-201 3-021, Rev. 2 Chapter 13.0 -Accident An alysis Table 13-23. Bounding Radionuclide Liq uid Stream Concentrations (4 pages) | |||
Unit operation Target dissolution Uranium recovery and recycle | |||
[Proprietary Information] | [Proprietary Information] | ||
Dissolver roduct 12smsb [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 129Sb [Proprietary Information] [Proprietary Information] | ||
1s1sm [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 1s3sm [Proprietary Information] [Proprietary Information] | ||
1s6sm [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 89Sr [Proprietary Information] [Proprietary Information] | ||
9osr [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 91sr [Proprietary Information] [Proprietary Information] | ||
92 [Proprietary Information] | |||
[Proprietary Information] | Sr [Proprietary Information] | ||
99Tc [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 99mTc [Proprietary Information] [Proprietary Information] | ||
12smTe [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 121Te [Proprietary Information] [Proprietary Information] | ||
121mTe [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 129Te [Proprietary Information] [Proprietary Information] | ||
129mTe [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 131Te [Proprietary Information] [Proprietary Information] | ||
131mTe [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 132Te [Proprietary Information] [Proprietary Information] | ||
133Te [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 133mTe [Proprietary Information] [Proprietary Information] | ||
134Te [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 23 1Th [Proprietary Information] [Proprietary Information] | ||
234Th [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 232u [Proprietary Information] [Proprietary Information] | ||
234U [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 23su [Proprietary Information] [Proprietary Information] | ||
236u [Proprietary In formation] [Proprietary Information] | |||
[Proprietary Information] | 231u [Proprietary Information] [Proprietary Information] | ||
mu [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 1J1mxe [Proprietary Information] [Proprietary Information] | ||
133 Xe [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 1JJmxe [Proprietary Information] [Proprietary Information] | ||
135 Xe [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 1Jsmxe [Proprietary Information] [Proprietary Information] | ||
89my [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 90y [Proprietary Information] [Proprietary Information] | ||
90my [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 9J y [Proprietary Information] [Proprietary Information] | ||
13-62 | |||
[Proprietary Information] | |||
NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) | |||
[Proprietary Information] | Unit operation Target dissolution Uranium recovery and recycle | ||
[Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | Dissolver roduct Uranium se aration feed 9Imy [Proprietary Information] [Proprietary Information] | ||
92y [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 93y [Proprietary Information] [Proprietary Information] | ||
93zr [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | 9szr [Proprietary Information] [Proprietary Information] | ||
97 Zr [Proprietary Information] [Proprietary Information] | |||
[Proprietary Information] | Totals [Proprietary Information] [Proprietary Information] | ||
Source: Table 2-1 ofNWM l-2013-CALC-0 11 , Source Term Calculations, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, February 2015. | |||
[Proprietary Information] | EOI = end of irradiation. | ||
In each case, a jacketed vessel is assumed to be filled with process solution appropriate to the process location, with the process offgas ventilation system operating. A level monitoring system will be available to monitor tank transfers and stagnant store vo lumes on all tanks processing LEU or fission product solutions. | |||
[Proprietary Information] | 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 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. | ||
13.2.4.2 Identification of Event Initiating Conditions The accident initiating event is generally described as a process equipment failure. The PHA identified simi lar accident sequences in four nodes associated with leaks of enriched uranium solution into heating and/or cooling coils surrounding safe-geometry tanks or vessels. The PHA identified predominately corrosive degradation of the tank or overpressure of the tank as potential causes that might damage this interface and allow enriched uranium solution to leak into the cooling system media or into the steam condensate for the heating system. | |||
[Proprietary Information] | The primary containment fails , which allows radioactive or fissi le solutions to enter an auxiliary system. | ||
Radioactive or fissile solution leaks across the mechanical boundary between a process vessel and associated heating/cooling jacket into the heating/cooling media . Where heating/cooling jackets or heat exchangers are used to heat or cool a fissile and/or high-dose process solution, the potential exists for the ba1Tier between the two to fail and allow fissile and/or hi gh-dose process solution 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. | |||
[Proprietary Information] | 13-63 | ||
[Proprietary Information] | 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. | |||
[Proprietary Information] | 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 also occur. Either of these potential outcomes can exceed the performance criteria of one process upset, leading to accidental nuclear criticality 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) system includes the following. | |||
[Proprietary Information] | 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. | |||
[Proprietary Information] | 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. | |||
[Proprietary | Maintenance activities to identify, repair, or replace the cause of the leak are initiated after achieving the final stable condition. | ||
[Proprietary Information] | 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-geometry 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. | |||
[Proprietary Information] | 13.2.4.5 Unmitigated Likelihood Leaks into auxiliary services can be initiated by mechanical failure of equipment boundaries between the process solutions 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." | |||
[Proprietary Information] | 13-64 | ||
13- | |||
[Proprietary Information] | * :~*:~*:* NWM I | ||
..*... NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis | |||
[Proprietary Information] | ~ * *! NOfllTHWEn MEDCAl. ISOTOPfS 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 in the process system. The high-dose uranium solution source term bounds this analysis. Solution leaks into 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 uranyl 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. | ||
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. | |||
[Proprietary Information] | |||
[ | |||
[Proprietary Information] | |||
[Proprietary Information] | |||
[Proprietary Information] | |||
11 | |||
2 | |||
13.2.4.7 Evaluation of Potential Radiological Consequences The following evaluations are based on information developed for the Construction Permit Application. | 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. | Additional detailed information describing radiological consequences will be developed for the Operating License Application. | ||
13.2.4.7.1 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, have similar exposure consequences due to the criticality 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. | |||
Even the low-dose uranium solutions , while generally contact-handled , | Based on the analysis of several accidental nuclear criticalities in industry, LA-13638 identifies that a uranium solution criticality can yield between 10 16 to 10 17 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. | ||
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. 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 | 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 | ||
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 | ;... NWMI | ||
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. | .~ NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis | ||
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 | . ~* * ~ NOllTHWUT MEDICAL tSOTOPES 13.2.4.8.1 IROFS RS-04, Hot Cell Shielding Boundary 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 systems. | ||
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. | 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 shielding for workers and the public during process 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 boundary 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 reduce the exposure from postulated upsets within the hot cell shielding boundary to less 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 operable when the equipment in that area is in the operating or standby modes. | ||
Each shielded protected area will be | 13.2.4.8.2 IROFS CS-06, Pencil Tank and Vessel Spacing Control using the Diameter of the Tanks, Vessels, or Piping All tanks, vessels, or piping systems involved 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 | 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 cooling loop with monitoring for uranium process solution or high-dose process solution will be provided to safely contain fissile process solution that leaks across this boundary, if the primary boundary fails. The dual-purpose safety function of this closed loop is to prevent fissile process solution from causing accidental nuclear criticality and to prevent high-dose 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 | 1 Some operations may have higher doses during short periods of the operation . The average worker expos ure rate is designed to be 0.5 mrem/hr, or less. Areas not normall y accessible by the worker may have higher dose rates (e.g., streaming radiation around normally inaccessible remote manipulator penetrations well above the worker's head). | ||
Sampling of the | 2 The shielding is not credited for mitigat ing dose rates during an accidental nuclear criticality; instead, additional IROFS are identified to provide double-contingency protection to prevent (reduce the likelihood of) an accidental nuclear criticality. | ||
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 | 13-66 | ||
The closed loop will also | |||
The heat | 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 heating or cooling media (e.g., steam condensate conductivity, cooling water radiological activity, uranium concentration, etc.) wi ll be conducted to alert the operator that a breach has occurred and that additional corrective actions are required to identify and isolate the failed component and restore the closed loop integrity. Discharged solutions from this system wi ll be handled as potentially fissile and 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 closed cooling loop with monitoring for breakthrough of process solution will be provided to contain process solution that leaks across this boundary, if the boundary fails. IROFS CS-27 is applied to those high-heat capacity cooling jackets (requiring very large loop heat exchangers) servicing condensers where the leakage is always from the cooling loop to the condenser, reducing back-leakage, and the risk of product solutions entering the condenser is very low by evaporator or concentrator design. | |||
Closed loop | The purpose of this safety function is to monitor the condition of the condenser cooling jacket to ensure that in the unlikely event that a condenser 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 isolate any high-dose fissile product solids (from the same event) from penetrating the hot cell shielding boundary, and any high-dose fission gases from penetrating the hot cell shielding boundary during normal operations. 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 cooling media (e.g., cooling water radiological activity, uranium concentration, etc.) will be conducted to alert the operator that a breach has occurred and that additional corrective actions are required to identify and isolate the failed component and to restore the closed loop integrity. Closed loop pressure will also be monitored to identify a leak from the closed loop to the process system. Discharged solutions from this system will be handled as potentially fissile and sampled according to IROFS CS-16 and CS- 17 prior to discharge to a non-safe geometry. | ||
,...----- | 13.2.4.8.5 IROFS CS-20, Evaporator or Concentrator Condensate Monitoring As an AEC, the condensate tanks 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 system is to (1) detect an anomaly in the evaporator or concentrator indicating high uranium content in the condenser (due to flooding or excess ive foaming) , and (2) prevent high concentration uranium solution from being available in the condensate tank for discharge to a non-favorable geometry system or in the condenser for leaking to the non-safe geometry cooling loop. The safety function of this IROFS is to prevent an accidental nuclear criticality. The detection system will work by continuously monitoring condensate uranium content and detecting high uranium concentration, and then shutting down the evaporator to isolate the condensate from the condenser and condensate tank. At a limiting setpoint, the uranium monitor-detecting devi ce will close an isolation valve in the inlet to the evaporator (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 designed to produce a valve-open permissive signal that fails to an open state, closing the valve on loss of electrical power. The isolation valve is des igned to fail-closed on loss of instrument air, and the solenoid is designed to fai l-closed on loss of signal. The locations where this IROFS is used will be determined during final design. | |||
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The | NWMl-20 13-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.2.4.8.6 IROFS CS-18, Backflow Prevention Device 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 solutions from backflowing from the tank into systems that are not designed for fissile solutions that could lead to accidental nuclear criticality or to locations 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 lity and availability for that location. | ||
The backflow prevention device features for high-dose product solutions will be located inside the hot cell shielding and confinement boundaries ofIROFS RS-04 and RS-01 , respectively. The feature is designed such that spills from overflow are directed to a safe geometry confinement berm controll ed by IROFS CS-08 (described in NWMI-2015-SAFETY-004, Quantitative Risk Analysis 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 is to prevent accidental nuclear criticality by providing a safe-geometry tank if a fissile solution backs-up into an auxiliary chemical addition system. | |||
IROFS CS-19 will be used where conventional backflow prevention in pressurized 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 vessel that is filled with chemical reagent using the conventional backflow prevention devices, and then provides a pump to add the reagents to the respective process system under pressure. Safe-geometry day tanks servicing high-dose product solutions systems will be located in the hot cell shielding 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. | 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 | All tanks will be vented and unpressurized under normal use. | ||
The heating and cooling 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 remotely. Operating procedures will identify an operational high-level fill operating limit for each tank. As part of the level detector, a high-level audible alarm and light will be provided to indicate a high level above this operating limit so that the operator can take action to correct conditions leading to fai lure 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). | |||
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*:~*~h NWMI | |||
* NCMO'HWHT MEOfC.Al ISOTOPU NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis Tank and | ~* * ~ NCMO'HWHT MEOfC.Al ISOTOPU NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis Tank and vessel 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. | 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. | ||
The controls | 13.2.4.9 Mitigated Estimates The controls selected wil l mitigate both the frequency and consequences of this accident. The controls selected and described above will prevent a criticality associated with SNM leaks into auxiliary systems. | ||
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. | The selected IROFS have reduced the potential worker safety consequences to acceptable levels. | ||
: 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 | Additional 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 by 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 are 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 electrical heaters cease operation as part of the passive operation mode. | |||
Pressure-relief confinement system for the target dissolver offgas 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). | |||
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NWMI | *:~*:~":" NWMI NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis | ||
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Pumps performing liquid | * NOflTtfWUTMEDtCALISOTOPES Process vessel emergency purge system is activated for hydrogen concentration control in tank vapor spaces (IROFS FS-03). | ||
: 5. The facility is now in a stable condition. | 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). | ||
13.2.5.4 Function of Components or Barriers All facility structural components of the hot cell | All equipment providing a motive force for process activities cease, including: | ||
The engineered safety features requiring power will activate or go to their fail-safe configuration. | Pumps performing liquid transfers of process solutions Pumps supporting operation of the steam and cooling utility heat transfer fluids Equipment supporting physical transfer of items (primarily cranes) | ||
13.2.5.5 Unmitigated Likelihood Loss of power can be initiated by off-site events or mechanical failures of equipment. | : 4. Operators follow alarm response procedures. | ||
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. | : 5. The facility is now in a stable condition. | ||
13.2.5.6 Radiation Source Term The loss of power evaluation is | 13.2.5.4 Function of Components or Barriers All facility structural components of the hot cell secondary 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 based 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. | 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. | 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. | 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. | 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. | ||
A standby diesel generator will be available at the RPF. 13-70 | Defensive-in-Depth The following defense-in-depth feature, minimizing the impact of a loss of power event, was identified by the accident evaluations. | ||
' 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 | A standby diesel generator will be available at the RPF. | ||
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NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis 13.2.6 Natural Phenomena Events Chapter 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 sequences 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. | 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 / | This evaluation addresses the facility design as impacted by the maximum-sized tornado with a return frequency of 10-5/year (yr). | ||
* High winds can lead to significant damage to the facility structure. | * 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 prevents impact to SS Cs in the building. | ||
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 | The local area impact may result in loss of utilities (e.g., electrical power) and reduced access by 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. | ||
This evaluation | High winds may directly impact SSCs important to safety (e.g., components 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 loss of electrical power) that can be initiated concurrently with the tornado (either as an indirect result or as an additional random failure). This evaluation analyzes the impact of tornado winds on these SSCs. | ||
-High wind | Tornado impact on the facility structure - High wind pressures could cause a partial or complete collapse of the facility structure, which may cause damage to SS Cs important to safety or impact the availability and reliability of those SSCs. A partial or complete structural collapse may also lead directly to a radiological or chemical release or a potential nuclear criticality, if damage caused by the collapse creates a violation of criticality spacing requirements. Tornado wind-driven missiles could penetrate the facility building envelope (walls and roof), impacting the availabi lity and reliability of SSCs important to safety, or may lead directly to a radiological or chemical release. | ||
Tornado wind-driven missiles could penetrate the | Tornado impact on SSCs important to safety located outside the main facility - High wind pressures and tornado wind-driven missiles could damage SSCs important to safety located outside the RPF building envelope. The damage sustained may impact the availability and reliability of the SSCs important to safety. Loss of site power may affect the ability of SSCs important to safety located within the facility building envelope to respond to additional events. | ||
The damage sustained | A partial or complete collapse of the facility structure could also lead directly to an accident with adverse intermediate or high consequences. The only IROFS located outside the RPF building envelope is the exhaust stack. Buckling or toppling of the exhaust stack would affect its ability and availabi lity to mitigate other events with intermediate consequences. The return frequency of the design basis tornado is 10-5/yr, making the initiating event highly unlikel y. | ||
The | No additional IROFS are required. | ||
The return | 13-71 | ||
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 | |||
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 | 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 safety. This evaluation demonstrates how the facility design addressed straight-line winds with a return interval of 100 years or more, as required by building codes. | ||
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. | 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 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 | 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 Design Loads for Buildings and Other Structures. 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 , | ||
For impact on the facility, the PMP for 25.9 square kilometers (km 2) (10 square miles [mi 2]) is | 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 Code Requirements for Structural Concrete, 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 . | ||
From the | 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. | ||
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 | No additional IROFS are required. | ||
* NORTMWUT MEOtC.Al ISOTOl'£S | 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 empirical methodology with no defined annual exceedance probability. | ||
Direct damage or impairment of SSCs could also be caused by flooding in the facility. | For impact on the facility, the PMP for 25.9 square kilometers (km 2) (10 square miles [mi 2]) is evaluated. | ||
The site will be graded to direct the stormwater from localized | Large amounts of rain water accumulating on the roof could lead to collapse of the roof. A partial or complete collapse of the faci lity 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. | ||
However , the site | From the National Weather Service (NWS)/National Oceanic and Atmospheric Administration (NOAA) | ||
3 No additional IROFS are required. | Hydrometeorological Report No. 51 , Probable Maximum Precipitation Estimates, United States East of the 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 empirical 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 Generalized Estimates ofProbable Maximum Precipitation 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. | ||
13.2.6.5 Seismic Impact to the Facility and | 13-72 | ||
This event would result in potentially lethal doses to workers in a short period of time. | |||
3 The | *:i*;~*:* NWMI | ||
~* *~ NORTMWUT MEOtC.Al ISOTOl'£S NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis No additional IROFS are required. | |||
* NORTHWtsT MEOtCAL ISOTOPfS NWMl-2013-021, Rev. 2 Chapter 13.0 - | 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 safely 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. | ||
NWMI-2015-SAFETY-004 , Section 3.1 , identifies IROFS to prevent and mitigate this accident scenario. | 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 rivers 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 downpours 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 standard industrial design. Rainwater that falls on the waste management truck ramp and accumulates in the trench drain has low 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 2x 10*3 year return frequency flood, which can be considered an unlikely event according to performance criteria. However, the site is located at an elevation of 248.4 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 dry 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 Structures, 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. During 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 docking 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 shield plug in the presence of personnel. This event would result in potentially lethal doses to workers in a short period of time. | |||
Seismic ground shaking can directly damage SSCs relied on for safety or lead to damage of the facility, including partial or complete 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 intermediate consequence. | |||
3 The recommended standard for determi ning the probably maximum flood, ANS 2.8, Determining Design Basis Flooding at Power Reactor Sites, has been withdrawn. | |||
13-73 | |||
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~* * ~ NORTHWtsT MEOtCAL ISOTOPfS NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 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. | 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 , | 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 , NEHRP Recommended Seismic Provisions for New Buildings and Other Structures. The MCfa for this site is governed by the probabilistic maximum-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. | ||
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. | 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 . | 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 maximum-considered earthquake ground-shaking, or a probability of failure of l .2x 1o-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, / | 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, /earthquake = 4x 10-4. | ||
The facility protects the SSCs , and an extreme snow- | 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. | ||
NRC DC/COL ISG-07 , | 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- loading event may cause failure of the roof, impacting the SSCs' ability to perform associated safety functions. NRC DC/COL ISG-07, Interim Staff Guidance on Assessment ofNormal 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 bounds 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. | ||
The probable maximum winter precipitation is based on the seasonal variation of the PMP , given in NWS/NOAA Hydrometeorological Report 53, | 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 primary confinement boundary damaged), or may prevent an SSC from being available to perform its function. | ||
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 | 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, Seasonal Variation of 10-Square Mile Probable Maximum Precipitation Estimates, United States East of the 1051h 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. | |||
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~ * *! N<HllTHWE.ST MEDJCAL ISOTOP£S 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 1o-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. | 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. | 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. | ||
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. | 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. | ||
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. | ||
The preliminary IROFS for each sequence are listed in the far right column of Table 13-24. The IROFS number and title are provided. | Table 13-24. Analyzed Accidents Sequences (9 pages) | ||
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 from PHA Descriptor Preliminary IROFS Identified S.R.01 High-dose solution or enriched | ||
* IROFS RS-01, Hot Cell Liquid Confinement Boundary | * IROFS RS-01, Hot Cell Liquid Confinement Boundary uranium solution spill causing a | ||
* IROFS RS-03, Hot Cell Secondary Confinement Boundary | * IROFS RS-03, Hot Cell Secondary Confinement Boundary radiological exposure hazard | ||
* IROFS RS-04 , Hot Cell Shielding 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-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-08, Floor and Sum Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Spill Containment Berms | ||
* IROFS CS-09, Double-Wall Piping | * IROFS CS-09, Double-Wall Piping | ||
* See Section 13.2.2.8 S.R.02 Spray release of solutions spilled | * 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 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 | * 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 Table 13-24. Analyzed Accidents Sequences (9 pages) | |||
Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.04 Liquid enters process vessel | |||
* IROFS RS-09, Primary Offgas Relief System ventilation system damaging | |||
* IROFS RS-03, Hot Cell Secondary Confinement Boundary IRU or retention beds, releasing | |||
* See Section 13.2.3.8 retained radionuclides S.R.05 High-dose solution enters the | |||
* Not credible or low consequence UN blending and storage tank S.R.06 High flow through IRU causing | |||
* Bounded by S.R.04 premature release of high-dose iodine gas S.R.07 Loss of temperature control on | |||
* Bounded by S.R.04 the IRU leading to release of high-dose iodine S.R.08 Loss of vacuum pumps | |||
* Bounded by S.R.04 S.R.09 Loss ofIRU or carbon bed | |||
* IROFS RS-04, Hot Cell Shielding Boundary | * Bounded by S.R.04 media to downstream part of the system S.R.10 Wrong retention media added to | ||
* IROFS CS-06, Pencil Tank and Vessel Spacing Control using the Diameter of the Tanks, Vessels, or Piping | * Event unlikely with intermediate consequence bed or saturated retention media S.R.12 Mo product cask removed from | ||
* Event unlikely with intermediate consequence the hot cell boundary with improper shield plug installation S.R.13 High-dose containing solution | |||
* IROFS RS-04, Hot Cell Shielding Boundary leaks to chilled water or steam | |||
* IROFS CS-06, Pencil Tank and Vessel Spacing Control using condensate system the Diameter of the Tanks, Vessels, or Piping | |||
* IROFS CS-10, Closed Safe-Geometry Heating or Cooling Loop with Monitoring and Alarm | * 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-27, Closed Heating or Cooling Loop with Monitoring and Alarm | ||
| Line 1,246: | Line 1,183: | ||
* IROFS CS-18, Backflow Prevention Device | * IROFS CS-18, Backflow Prevention Device | ||
* IROFS CS-19, Safe-Geometry Day Tanks | * IROFS CS-19, Safe-Geometry Day Tanks | ||
* See Section 13.2.4.8 | * See Section 13.2.4.8 S.R.14 IX resin failure due to wrong | ||
* Bounded by S.R.01 | * Bounded by S.R.01 reagent or high temperature S.R.16 Backflow of high-dose | ||
* Bounded by S.R.13 13-76 | * Bounded by S.R.13 radiological and/or fissile solution into auxiliary system (purge air, chemical addition line, water addition line, etc.) | ||
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* IROFS RS-08, Sample and Analysis of Low Dose Waste Tank Dose Rate Prior to | ***~**: | ||
* IROFS RS-10 , Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer | ** * | ||
* See Section 13.2.7.1 | * NORTtfWf.$T MEDICAL ISOTOf'ES NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) | ||
* Low consequence event that does not challenge IROFS RS-04 S.R.19 High target basket retrieval dose | Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.17 Carryover of high-dose solution | ||
* Design evolved after PHA , accident sequence eliminated S.R.20 S.R. | * IROFS RS-08, Sample and Analysis of Low Dose Waste Tank into condensate (a low-dose Dose Rate Prior to Transfer Outside the Hot Cell Shielded waste stream) Boundary | ||
* IROFS RS-10, Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer | |||
* See Section 13.2.7.1 S.R.18 High-dose solution flows into | |||
* Low consequence event that does not challenge IROFS RS-04 the solidification media hopper S.R.19 High target basket retrieval dose | |||
* Design evolved after PHA, accident sequence eliminated rate S.R.20 Radiological spill of irradiated | |||
* Bounded by S.R.01 LEU target material in the hot cell area S.R.21 Damage to the hot cell wall | |||
* Low consequence event that does not damage shielding providing shielding function ofIROFS RS-04 S.R.22 Decay heat buildup in | |||
* Low consequence event unprocessed LEU target material removed from targets leads to higher-dose radionuclide offgassing S.R.23 Offgassing from irradiated target | |||
* IROFS RS-03 , Hot Cell Secondary Confinement Boundary dissolution tank occurs when the | * 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 | * See Section 13.2.2.8 upper valve is opened S.R.24 Bagless transport door failure | ||
* IROFS RS-03, Hot Cell Secondary Confinement Boundary | * IROFS RS-03, Hot Cell Secondary Confinement Boundary | ||
* IROFS RS-04, Hot Cell Shielding Boundary | * IROFS RS-04, Hot Cell Shielding Boundary | ||
* See Section 13.2.2.8 | * See Section 13.2.2.8 S.R.25 HEPA filter failure | ||
* IROFS RS-03 , Hot Cell Secondary Confinement Boundary | * IROFS RS-03 , Hot Cell Secondary Confinement Boundary | ||
* See Section 13.2.2.8 | * See Section 13.2.2.8 S.R.26 Failed negative air balance from | ||
* IROFS RS-03, Hot Cell Secondary Confinement Boundary | * IROFS RS-03, Hot Cell Secondary Confinement Boundary zone-to-zone or failure to | ||
* See Section 13.2.2.8 | * See Section 13.2.2.8 exhaust a radionuclide buildup in an area S.R.27 Extended outage of heat leading | ||
* Highly unlikely | * Highly unlikely event for process solutions containing fission to freezing, pipe failure, and products release ofradionuclides from Bounded by S.C.04 for target fabrication systems liquid process systems S.R.28 Target or waste shipping cask or | ||
* Information will be provided in the Operating License Application 13-77 | * Information will be provided in the Operating License container not loaded or secured Application according to procedure, leading to personnel exposure 13-77 | ||
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* IROFS RS-12 , Cask Containment Sampling Prior to Closure Lid Removal | *::**::* NWMl-2013-021, Rev . 2 Chapter 13.0 -Accident Analysis | ||
* IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations | ~* * ~ NORTHWEST MEDtcAL ISOTOP£S Table 13-24. Analyzed Accidents Sequences (9 pages) | ||
* See Section 13.2.7.1 | Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.29 High dose to worker from | ||
* IROFS RS-04, Hot Cell Shielding Boundary | * IROFS RS-12, Cask Containment Sampling Prior to Closure release of gaseous radionuclides Lid Removal during cask receipt inspection | ||
* IROFS RS-15, Cask Docking Port Enabling Sensor | * IROFS RS-13, Cask Local Ventilation During Closure Lid and preparation for target basket Removal and Docking Preparations removal | ||
* See Sections 13.2.2.8 and 13.2.7.l | * See Section 13.2.7.1 S.R.30 Cask docking port failures lead | ||
* Judged unlikely event with intermediate | * IROFS RS-04, Hot Cell Shielding Boundary to high-dose to worker due to | ||
* IROFS RS-15, Cask Docking Port Enabling Sensor streaming radiation and/or high | |||
* See Sections 13.2.2.8 and 13.2.7.l airborne radioactivity S.R.31 Chemical burns from | |||
* Judged unlikely event with intermediate consequence contaminated solutions during sample analysis S.R.32 Crane load drop accidents | |||
* IROFS FS-01, Enhanced Lift Procedure | * IROFS FS-01, Enhanced Lift Procedure | ||
* IROFS FS-02, Overhead Cranes | * IROFS FS-02, Overhead Cranes | ||
* See Section 13.2. 7.1 | * See Section 13.2. 7.1 S.C.01 Failure of facility enrichment | ||
* Judged highly unlikely based on supplier's checks and balances | * Judged highly unlikely based on supplier's checks and balances limit S.C.02 Failure of administrative control | ||
* IROFS CS-02, Mass and Batch Handling Limits for Uranium Metal, [Proprietary Information], Targets, and Laboratory Sample Outside Process Systems | * IROFS CS-02, Mass and Batch Handling Limits for Uranium on mass (batch limit) during Metal, [Proprietary Information], Targets, and Laboratory handling of fresh U, scrap U, Sample Outside Process Systems LEU target material, targets, and | ||
* IROFS CS-03, Interaction Control Spacing Provided by samples Administrative Control | |||
* IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement | |||
* See Section 13.2. 7.2 S.C.03 Failure of interaction limit | |||
* IROFS CS-02, Mass and Batch Handling Limits for Uranium during handling of fresh U, scrap Metal, [Proprietary Information], Targets, and Laboratory U, LEU target material, targets, Sample Outside Process Systems containers, and samples | |||
* IROFS CS-03, Interaction Control Spacing Provided by Administrative Control | * 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-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement | ||
* See Section 13.2. 7.2 | * See Section 13.2.7.2 S.C.04 Spill of process solution from a | ||
* IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry tank or process vessel leading to Confinement using the Diameter of Tanks, Vessels, or Piping accidental criticality | |||
* 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-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-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-09, Double-Wall Piping | ||
* IROFS CS-26, Processing Component Safe Volume Confinement | * IROFS CS-26, Processing Component Safe Volume Confinement | ||
* See Section 13.2.7.2 13-78 | * See Section 13.2.7.2 13-78 | ||
.. ;. NWMI *::**::* | |||
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* | I | ||
* IROFS CS-11, Simple Overflow to Normally | ~ ..;. NWMI | ||
* IROFS CS-12 , | *::**::* NWMl-2013-021 , Rev . 2 | ||
* IROFS CS-13, Simple Overflow to Normally | **~~!'!* . NO~ST MEDtcAL ISOTOPES Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) | ||
* See Section 13.2.7.2 | Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.05 Leak of fi ssile solution into the | ||
* | * Bounded by S.R 13 heating or cooling jacket on the tank or vessel S.C.06 System overflow to process | ||
* Bounded by S.R.13 | * IROFS CS-11, Simple Overflow to Normally Empty Safe ventilation involving fissile Geometry Tank with Level Alarm material | ||
* IROFS CS-06 , Pencil | * IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line | ||
* IROFS CS-07 , | * IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary | ||
* | * See Section 13.2.7.2 S.C. 07 Fissile solution leaks across | ||
* See | * Bounded by S.R.13 mechanical boundary between process vessels and heating/cooling jackets into heating/cooling media S.C.08 Backflow of high-dose | ||
* IROFS CS-14, Active Discharge | * Bounded by S.R.13 radiological and/or fissile solution into auxiliary system (purge air, chemical addition line, water addition line, etc.) | ||
* IROFS CS-15 , Independent Active | S.C.09 High concentrations of uranium | ||
* IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry enter the concentrator or Confin ement using the Diameter of Tanks, Vessels, or Piping evaporator condensates | |||
* IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels | |||
* IROFS CS-26, Processing Component Safe Volume Confinement | |||
* See Section 13.2.7.2 S.C.10 High concentrations of uranium | |||
* IROFS CS-14, Active Discharge Monitoring and Isolation enter the low-dose or high-dose | |||
* IROFS CS-15, Independent Active Discharge Monitoring and waste collection tanks Isolation | |||
* IROFS CS-16, Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal | * 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-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal | ||
* See Section 13.2. 7 .2 13-79 | * See Section 13.2.7 .2 13-79 | ||
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* Bounded by S.C.04 and S.C. l 0 | . ~;~;; | ||
* IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement | * ~- * ~ . 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 High concentrations of uranium | |||
* Bounded by S.C.04 and S.C. l 0 in contactor solvent regeneration aqueous waste S.C.12 High concentrations of uranium | |||
* IROFS CS-04, Interaction Control Spacing Provided by in the LEU target material wash Passively Designed Fixtures and Workstation Placement solution | |||
* IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confinement using the Diameter of Tanks, Vessels, or Piping | * 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-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels | ||
* See Section 13.2.7.2 | * See Section 13.2.7.2 S.C.13 High concentrations of uranium | ||
* IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry | * IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry in the nitrous oxide scrubber Confinement using the Diameter of Tanks, Vessels, or Piping | ||
* IROFS CS-16, Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal | * 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-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal | ||
* See Section 13.2.7.2 | * See Section 13.2.7.2 S.C.14 High concentrations of uranium | ||
* IROFS CS-16, Sampling and Analysis of Uranium Mass or in the IX waste collection tanks Concentration Prior to Discharge or Disposal effluent | |||
* IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal | |||
* IROFS CS-16 , | * See Section 13.2.7.2 S.C.15 High concentrations of uranium | ||
* IROFS CS-17 , Independent | * IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry in the IX resin waste Confinement using the Diameter of Tanks, Vessels, or Piping | ||
* See Section 13.2.7.2 | * 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 Disposal | * 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 S.C.17 High concentrations of uranium | |||
* IROFS CS-16, Sampling and Analysis of Uranium Mass or in the solid waste encapsulation Concentration Prior to Discharge or Disposal process | |||
* IROFS CS-17, Independent Sampling and Analysis of Uranium 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-21, Visual Inspection of Accessible Surfaces for Foreign Debris | ||
| Line 1,343: | Line 1,296: | ||
* IROFS CS-24, Independent 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 | * IROFS CS-25, Target Housing Weighing Prior to Disposal | ||
* See Section 13.2. 7.2 13-80 | * See Section 13.2. 7.2 13-80 | ||
* IROFS CS-06, Pencil Tank, | |||
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.19 Failure of PEC - Component | |||
* IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry safe geometry dimension or safe Confinement using the Diameter of Tanks, Vessels, or Piping volume | |||
* IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels | * IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels | ||
* IROFS CS- | * IROFS CS-26, Processing Component Safe Volume Confinement | ||
* See Section 13.2.7.2 S.C.20 Failure of concentration limits | |||
* See Section 13.2. 7 .2 | * No credible path leading to criticality identified or not credible by design S.C.21 Target basket passive design | ||
* | * IROFS CS-02, Mass and Batch Handling Limits for Uranium control failure on fixed Metal, [Proprietary Information], Targets, and Laboratory interaction spacing Sample Outside Process Systems | ||
* IROFS CS- | |||
* IROFS CS-03, Interaction Control Spacing Provided by Administrative Control | * 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 S.C.22 High concentration of uranium | ||
* IROFS CS-04, Interaction Control Spacing Provided by in the TCE evaporator residue 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 | |||
* 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 S.C.23 High concentration in the spent | |||
* IROFS CS-04, Interaction Control Spacing Provided by si licone oi l waste Passively Designed Fixtures and Workstation Placement | |||
* IROFS CS-05, Container Batch Volume Limit | * IROFS CS-05, Container Batch Volume Limit | ||
* See Section 13.2.7.2 | * IROFS CS-06, Penci l Tank, Vessel, or Piping Safe Geometry Confi nement 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-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 Secti on 13.2. 7 .2 S.C.24 High uranium content on HEPA | |||
* Bounded by S.C.17 filters and subsequent failure 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 | |||
* IROFS CS-03, Interaction Control Spacing Provided by controlled container volume Administrative Control limits | |||
* 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 S.C.28 Crane load drop accidents | |||
* IROFS FS-01, Enhanced Lift Procedure | * IROFS FS-01, Enhanced Lift Procedure | ||
* IROFS FS-02, Overhead Cranes | * IROFS FS-02, Overhead Cranes | ||
* See Section 13.2.7.2 | * See Section 13.2.7.2 S.F.01 Pyrophoric fire in uranium metal | ||
* Event highly unlikely based on credible physical conditions | * Event highly unlikely based on credible physical conditions S.F.02 Accumulation and ignition of | ||
* IROFS FS-03, Process Vessel Emergency Purge System | * IROFS FS-03, Process Vessel Emergency Purge System flammable gas in tanks or | ||
* See Section 13.2.7.3 | * See Section 13.2.7.3 systems S.F.03 Hydrogen detonation in | ||
* Judged highly unlikely based on credible physical conditions | * Judged highly unlikely based on credible physical conditions reduction furnace S.F.04 Fire in reduction furnace | ||
* Judged unlikely based on event frequency | * Judged unlikely based on event frequency S.F.05 Fire in a carbon retention bed | ||
* IROFS FS-05, Exhaust Stack Height | * IROFS FS-05, Exhaust Stack Height | ||
* See Section 13.2.7.3 S.F.06 Accumulation of flammable gas | |||
* Bounded by S.F.02 in ventilation system components S.F.07 Fire in nitrate extraction system - | |||
* Event unlikely with intermediate or low consequences combustible solvent with uranium S.F.08 General facility fire | |||
* Information will be provided in the Operating License Application S.F.09 Hydrogen exp losion in the | |||
* Information will be provided in the Operating License facility due to a leak from the Application hydrogen storage or distribution system S.F.10 Combustible fire occurs in hot | |||
* Information will be provided in the Operating License cell area Application S.F.11 Detonation or deflagration of | |||
* Information will be provided in the Operating License natural gas leak in steam Application generator room S.N.01 Tornado impact on facility and | |||
* Judged highly unlikely event based on return frequency SSCs important to safety S.N.02 High | |||
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~ * *! NOmfMST MEDICAL lSOTDPU NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis | |||
==13.4 REFERENCES== | ==13.4 REFERENCES== | ||
10 CFR 20, "Standards for Protection Against Radiation," Code a/Federal Regulations, Office of the Federal Register, as amended. | |||
Code | 10 CFR 30, "Rules of General App licability to Domestic Licensing of Byproduct Material," Code of Federal Regulations, Office of the Federal Register, as amended. | ||
.... ; NWMI | 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register, as amended. | ||
* MEDICAL ISOTOP£S | 10 CFR 70, "Domestic Licensing of Special Nuclear Material," Code of Federal Regulations, Office of the Federal Register, as amended. | ||
* * * * * * * * | 10 CFR 70.61 , "Performance Requirements," Code a/Federal Regulations, Office of the Federal Register, as amended. | ||
10 CFR 71, "Packaging and Transportation of Radioactive Material,'' Code ofFederal Regulations, Office of the Federal Register, as amended. | |||
NWMl-2013-021 , Rev. 1 Chapter 14 | ACI 318, Building Code Requirements for Structural Concrete, American Concrete Institute, Farmington Hills, Michigan, 2014. | ||
AISC 360, Specification for Structural Steel Buildings, American Institute of Steel Construction, Chicago, Illinois, 2010. | |||
ANS 2.8, Determining Design Basis Flooding at Power Reactor Sites, American Nuclear Society, La Grange Park, Illinois, 1992, W2002 . | |||
ANSI N 14.6, Standard for Special Lifting Devices for Shipping Containers Weighing 10, 000 Pounds (4,500 kg) or More for Nuclear Materials, American Nuclear Society, La Grange Park, Illinois, 1993 . | |||
ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, American Nuclear Society, La Grange Park, lllinois, 1998 (Reaffirmed 2007). | |||
ASCE 7, Minimum Design Loads for Buildings and Other Structures, American Society of Civil Engineers, Reston, Virginia, 2010. | |||
ASME B30.2, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist) , American Society of Mechanical Engineers, New York, New York, 2005. | |||
CDC, 2010, NIOSH Pocket Guide to Chemical Hazards, 2010- 168c, Centers for Disease Control and Prevention, http://www.cdc.gov/niosh/npg/, down loaded February 27, 2015. | |||
DC/COL ISG-07, Interim Staff Guidance on Assessment ofNormal and Extreme Winter Precipitation Loads on the Roofs ofSeismic Category I Structures, U.S . Nuclear Regulatory Commission, Washington, D.C., 2008. | |||
DOE-HDBK-30 I 0, DOE Handbook-Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, Change Notice No. 1, U.S. Department of Energy, Washington, D.C., December 1994 (R2013). | |||
DOE-STD-1090-2011, Hoisting and Rigging, U.S. Department of Energy, Washington, D.C., | |||
September 30, 20 11. | |||
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** * * ~OflTHWUT MEDICAL ISOTOP£S EPA 52011 020, Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 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 , NEHRP Recommended Seismic Provisions for New Buildings and Other Structures, Federal Emergency Management Agency, Washington, D.C., 2009. | |||
Hydrometeorological Report No. 51 , Probable Maximum Precipitation Estimates, United States East of the 105th Meridian, U.S. Department of Commerce, National Oceanic and Atmospheric Administration, Washington, D.C., 1978. | |||
Hydrometeorological Report No. 53 (NUREG/CR-1486), Seasonal Variation of JO-Square Mile Probable Maximum Precipitation Estimates, United States East of the 1051h Meridian, U.S. Department of Commerce, National Oceanic and Atmospheric Administration, U.S . Nuclear Regulatory Commission, Office of Hydrology National Weather Service, Washington, D.C. , April 1980. | |||
IBC, 2012, International Building Code, as amended, International Code Council, Inc. , Washington, D.C., February 2012. | |||
ICRP-26, Recommendations of the International Commission on Radiological Protection , International Commission on Radiological Protection, Ottawa, Canada, 1977. | |||
ICRP-30, Limits for Intakes of Radionuclides by Workers , International Commission on Radiological Protection, Ottawa, Canada, 1979. | |||
ICRP-72, Age-Dep endent Doses to the Members of the Public from Intake of Radionuclides - Part 5 Compilation of Ingestion and Inhalation Coefficients , International Commission on Radiological Protection, Ottawa, Canada, 1995 . | |||
LA-13638, A Review of Criticality Accidents, Los Alamos National Laboratory, Los Alamos, New Mexico, 2000. | |||
NAP 1994, Estimating Bounds on Extreme Precipitation Events, National Academy Press, National Research Council, Washington, D.C., I 994. | |||
NOAA Technical Report NWS 25 , Comparison of Generalized Estimates of Probable Maximum Precipitation with Greatest Observed Rainfalls, National Oceanic and Atmospheric Administration, Washington, D.C., 1980. | |||
NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors - Format and Content, Part 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February I 996. | |||
NUREG-1940, RASCAL 4: Description of Models and Methods, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C., December 2012. | |||
NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C. , March 1998. | |||
NWMI-2013-CALC-006, Overall Summary Material Balance - MURR Target Batch, Rev. D, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. | |||
NWMI-2013-CALC-011 , Source Term Calculations, 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 Facility, Rev. A, Northwest Medical Isotopes, Corvallis, Oregon, 2014. | |||
NWMI-2014-CALC-014, Selection ofDominant Target i sotopes for NWMJ Material Balances, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2014. | |||
NWMI-2015-RPT-009, Fission Product Release Evaluation, Rev. B, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. | |||
WMI-2015-SAFETY-OO 1, NWMJ Radioisotope Production Facility Preliminary Hazards Analysis, Rev. A, Northwest Medical Isotopes, Corvallis, Oregon, 2015. | |||
NWMI-2015-SAFETY-004, Quantitative Risk Analysis of Process Upsets Associated with Passive Engineering Controls Leading to Criticality Accident Sequences, Rev. A, Northwest Medical Isotopes, Corvallis, Oregon, 2015. | |||
Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C., February 1983. | |||
WSRC-TR-93-262, Savannah River Site Generic Data Base Development, Rev. I, Westinghouse Savannah River Company, Savannah Ri ver Site, Aiken, South Carolina, May 1988. | |||
13-99 | |||
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. NORTHWEST MEDICAL ISOTOPES Chapter 14.0 - Technical Specifications Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 1 August 2017 Prepared by: | |||
Northwest Medical Isotopes, LLC 815 NW g th Ave , Suite 256 Corvallis, OR 97330 | |||
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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 | |||
==Title:== | |||
Chapter 14.0 - Technical Specifications Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Sianature: | |||
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I NWMl-2013-021 , Rev. 1 Chapter 14.0 - Technical Specifications REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not required Incorporate changes based on responses to 1 8/5/2017 C. Haass NRC Requests for Add itional Information | |||
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~* * ~ NOflTHWlST MEDK:Al ISOT~S CONTENTS 14.0 TECHNICAL SPECIFICATIONS ............. .... ......................... .... ................................................ 14-1 14.1 Outline ... ....................................................................................... ..... ............ ................. .. 14-2 14.1. l Introduction ............................... ........................ ...... .......... ........ ................... ...... 14-2 14.1 .2 Safety Limit and Limiting Safety System Setting ... ....... ... ................... .............. 14-3 14.1.3 Limiting Condition of Operation ...... ...... ...... ........................... ......... ........... ...... 14-3 14.1 .4 Surveillance Requirements ................................... .............................................. 14-4 14.1.5 Design Features .................... ...... ................. ......... ........................... ................... 14-4 14.1 .6 Administrative Controls ......... ........................ ...... .... ............ ................ ...... ........ 14-4 14.2 References ................................................................................................. ....................... 14-5 TABLES Table 14-1. Potential Technical Specifications ........ ........................... ... ........ ..... .. ............... ............. 14-1 14-i | |||
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. ~* * ~ . NORTHWUT MEOfCAL ISOTOPlS TERMS Acronyms and Abbreviations AC administrative control ANS American Nuclear Society 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 Radioisotope 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 This 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, samples, and Criticality control approved containers* | |||
Spacing requirements on targets and containers Criticality control with SNM" Floor and sump designs* Criticality control Hot cell liquid confinement* Criticality control Process tank size and spacing* Criticality control Evaporator condensate monitor Criticality control Criticality monitoring system Criticality control In-line uranium content monitoring Criticality control Air pressure differential between zones* Control of airborne RAM Ventilation system filtration* Control of airborne RAM Process offgas subsystem Control of airborne RAM Primary offgas relief system Control of airborne RAM Hot cell shield thickness and integrity" Occupation and general public dose reduction Hot eel 1 secondary confinement boundary" Control of airborne RAM Double-wall piping Control of liquid RAM/criticality control Process closed heating and cooling loops Control of both airborne and liquid RAM System backflow prevention devices Control of liquid RAM/criticality control Stack height" Control of airborne RAM Area radiation monitoring system Occupation and general public dose reduction a Items that will significantly influence the final design. | |||
RAM = radioactive material. SNM special nuclear material. | |||
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* NORTHWEST Mf.DtCAl ISOTOPES 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 Development of Technical Specifications for Research Reactors; NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing 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 Federal 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 seems less likely than translating it into a LCO. | |||
The outline for the technical specifications that will be prepared during development 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 ' Definition Actions Actions are that part of a limiting condition for operation that prescribes Required Actions to be taken under designated conditions within specified completion times . | |||
Administrative .. . (described in Section 14.1.6) control (AC) | |||
Design features . .. (described in Section 14.1.5) | |||
Limiting condition ... (described in Section 14.1.3) for operation (LCO) | |||
Limiting safety .. .(described in Section 14.1.2) system setting (LSSS) 14-2 | |||
NWMl-2013-021 , Rev. 1 Chapter 14.0 - Technical Specifications Term Definition Modes 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. | |||
Operable/ A system, subsystem, component, or device shall be operable or have operability 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) . | |||
Safety limit (SL) ... (described in Section 14.1.2) | |||
Shall Denotes a mandatory requirement that must be complied with to maintain the requirements, assumptions, or conditions of the facility safety basis . | |||
Surveillance ... (described in Section 14.1.4) requirements Verify/verification 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 process 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 nor 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. | |||
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~ .* NORTHWEST MEDICAL ISOTOPES 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 arrangements. These design functions , if altered or modified, are implied to significantly affect safety and will not be identified in other sections. Anticipated areas covered in this section include the site and facility description, and fissionable material storage. Design features that will be provided in the technical specifications 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. | |||
* | |||
Anticipated areas covered in | |||
The requirements of 10 CFR 50.36(c)(4) are specified here as they pertain to the above referenced processes. | 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 | 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 lines for NWMI management (e.g., Levels 1 through 4). Other requirements include: | ||
* Identifying minimum staffing and supervisory functions | |||
* Identifying minimum staffing and supervisory functions Preparing and maintaining call lists Selecting and training | * Preparing and maintaining call lists | ||
14-4 NWM I ...... * | * Selecting and training personnel | ||
* NOITHWEST MEDtCAL ISOTDPH | * 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 conditions 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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==14.2 REFERENCES== | ==14.2 REFERENCES== | ||
10 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations , Office of the Federal Register, as amended. | |||
10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register, as amended. | |||
ANSI/ANS 15.1 , The Development ofTechnical Specifications for Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park Illinois, 2013. | |||
NRC, 20 I 2, Final Interim Staff Guidance Augmenting NUREG-153 7, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors," Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors, Docket ID: | |||
NRC-2011-0135 , U.S. Nuclear Regulatory Commission, Washington, D.C., October 30, 2012 . | |||
NUREG-153 7 (Part 1), Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content, U.S. Nuclear Regulatory Commission, Washington, D.C., | |||
February 1996. | |||
14-5 | |||
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Latest revision as of 10:36, 24 February 2020
| ML17221A203 | |
| Person / Time | |
|---|---|
| Site: | Northwest Medical Isotopes |
| Issue date: | 08/05/2017 |
| From: | Haass C Northwest Medical Isotopes |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17221A370 | List: |
| References | |
| NWMI-LTR-2017-011 NWMI-2013-021, Rev. 2 | |
| Download: ML17221A203 (127) | |
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~e *~ 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: c ~~ e.. ' ~
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- NOtliTifWEST MEDfCAl ISOTOPES NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not requ ired 1 6/26/2017 Incorporate changes based on responses to NRC C. Haass Requests for Additional Information 2 8/5/2017 Mod ifications based on comments from NRC staff C. Haass
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NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis CONTENTS 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 Radioisotope Production Facility Integrated Safety Analysis Process .... ..... ... ....... ..... ........ .......... ................... ... .......... ..... ... I 3-3 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and Risk Matrix .............. ...... .. .... ..... ...... ... ................... I 3-5 13 .1.1.2 Accident Consequence Analysis ... ........ .... .......................... .... ... ..... 13-7 13.1.1 .3 What-If and Structured What-If................... ....... ...... ... ... .............. .. 13-7 13.1.1.4 Hazards and Operability Study Method ........... ...... .......... .... ........... 13-8 13.1.1.5 Event Tree Analysis ...... .... ..... ...... ................. .. .... ..... ....... ............... 13-8 13.1 .1.6 Fault Tree Analys is ......... ........ .................................. ....... ... ........... 13-8 13.1.1.7 Failure Modes and Effects Analysis ...... .. ..... .................... .... .......... 13-8 13.1.2 Accident-Initiating Events .. ........ ..... ......... ....... .................... .. ..... ................ .... 13-8 13.1.3 Preliminary Hazards Analysis Results ............ .. ..... ....... .. ......... ...... .. ....... ... .. . 13-12 13.1.3.1 Hazard Criteria .......... .... .............. ......................... .. ..................... 13-12 13 .1.3.2 Radioisotope Production Facility Accident Sequence Evaluation ................... .......... ... ..... ....... .... .................... ........... .. .. 13-1 3 13.2 Analysis of Accidents with Radiological and Criticality Safety Consequences ............ 13-38 13.2. l Reserved ......................................... .................. ..... ........ .... .......... ...... .......... 13-39 13.2.2 Liquid Spills and Sprays with Radiological and Criticality Safety Consequences ... ........... ..... .. ......................................................................... 13-39 13 .2.2.1 Initial Conditions .... ..................................................................... 13-39 I 3 .2.2.2 Identification of Event Initiating Conditions ............. ............ .. .. ... . 13-44 I 3.2.2.3 Description of Accident Sequences ..... .... ........ ...... ......... .............. I 3-44 13.2.2.4 Function of Components or Barriers ....... ..... ................................. 13-44 13 .2.2.5 Unmitigated Likelihood ... ... ...... ............... .... .... ..... ............... ..... ... 13-45 I 3.2.2.6 Radiation Source Term .... ... ..... ....... .. ........ ...... ...... ..... .......... .... .... 13-45 13.2.2.7 Evaluation of Potential Radiological Consequences ..... ........ .. ....... 13-47 13 .2.2.8 Identification ofltems Relied on for Safety and Associated Functions ...................... .... ............. ..... .............. .............. ............. 13-50 13 .2.2.9 Mitigated Estimates ....... ...... ........... ...................... .. ..................... 13-54 13.2.3 Target Dissolver Offgas Accidents with Radiological Consequences .... ........ 13-54 13 .2.3. 1 Initial Conditions ............................................. ........... .. ........ ....... I 3-55 13 .2.3.2 Identification of Event Initiating Conditions ............ .. ......... .. ........ 13-56 13 .2.3.3 Description of Accident Sequences ............... ....... .... ...... ...... ........ 13-56 13.2.3.4 Function of Components or Barriers ............... ..... .................. ... .... 13 -56 13 .2.3.5 Unmitigated Likelihood ... ............ ........................ .... .................... 13-56 I 3.2.3.6 Radiation Source Term ................................. .. ................... ...... .... I 3-57 13.2.3.7 Evaluation of Potential Radiological Consequences ...................... 13-57 13 .2.3.8 Identification ofltems Relied on for Safety and Associated Functions ....... ............ ................................... .... ........................... 13-58 13.2.3.9 Mitigated Estimates ... .... ... ............ ............... ......... .. ... .......... .. .. .... 13-59 13-i
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' ~* * ~ NORTHWUT MEDtCAl ISOTOPES 13.2.4 Leaks into Auxiliary Services or Systems with Radiological and Criticality Safety Consequences .... .............................................................. . 13-59 13.2.4.1 Initial Conditions ............................................................. ... ......... 13-59 13.2.4.2 Identification of Event Initiating Conditions ................................. 13-63 13 .2.4.3 Description of Accident Sequences .................... ........ ... , ........ ...... 13-64 13 .2.4.4 Function of Components or Barriers .. ... ............. ........................... 13-64 13 .2.4.5 Unmitigated Likelihood ..... ............. ..... ........ ........................... ..... 13-64 13.2.4.6 Radiation Source Term ......... ... ... .................... ............................. 13-65 13 .2.4.7 Evaluation of Potential Radiological Consequences ................. ..... 13-65 13.2.4.8 Identification ofltems Relied on for Safety and Associated Functions ......... ..... ...... ...... ........ ..................... ........ .. ............... ..... 13-65 13.2.4.9 Mitigated Estimates ..... .. .. ... ... ......... ............. .................. .............. 13-69 13 .2.5 Loss of Power .. ............................................. ...... .............. .. ............. ... ......... 13-69 13.2.5.1 Initial Conditions ......................................................................... 13-69 13 .2.5.2 Identification of Event Initiating Conditions .................... .... ......... 13-69 13.2.5.3 Description of Accident Sequences ........... .......... .. ... ... .... ............. 13-69 13.2.5.4 Function of Components or Barriers ............................................. 13-70 13.2.5.5 Unmitigated Likelihood .. ........ .................. ... ................................ 13-70 13.2.5.6 Radiation Source Term ................................................................ 13-70 13.2.5.7 Evaluation of Potential Radiological Consequences ............. ......... 13-70 13.2.5.8 Identification ofltems Relied on for Safety and Associated Functions ................................ .... ................. ........ ................. ... .... 13-70 13.2.6 Natural Phenomena Events .............. ..... ... .............................. ....................... 13-71 13.2.6. l Tornado Impact on Facility and Structures, Systems, and Components ........ ......... .............................................. ................ .. 13-71 13.2.6.2 High Straight-Line Winds Impact the Facility and Structures, Systems, and Components ....... .... ............ ..................................... 13-72 13 .2.6.3 Heavy Rain Impact on Facility and Structures, Systems, and Components ..... .. ................................. ........... .... .... ...................... 13-72 13.2.6.4 Flooding Impact to the Facility and Structures, Systems, and Components ............................. ............ ............................. ....... ... . 13-73 13.2.6.5 Seismic Impact to the Facility and Structures, Systems, and Components ............................................ ....... .. .. ...... .. .................. 13-73 13.2.6.6 Heavy Snow Fall or Ice Buildup on Facility and Structures, Systems, and Components ............................. .............. ...... ........... 13-74 13.2.7 Other Accidents Analyzed ....... ....... .... ........... .................................. ...... ....... 13-75 13.2.7.1 Items Relied on for Safety for Radiological Accident Sequences (S .R.) ..... ............................ ....... .................................. 13-85 13.2. 7.2 Items Relied on for Safety for Criticality Accident Sequences (S.C.) ........................... ....................... ...... .......... ..... .................... 13-87 13.2.7.3 Items Relied on for Safety for Fire or Explosion Accident Sequences (S.F.) ................. ... .......... ........... ................. ................ 13-93 13.2.7.4 Items Relied on for Safety for Natural Phenomena Accident Sequences (S.N.) ............ ......... .......... .. ...................................... ... 13-93 13.2.7.5 Items Relied on for Safety for Man-Made Accident Sequences (S.M.) .. ...... .... .. .................................... ........ ................................ 13-94 13.2.7.6 Items Relied on for Safety for Chemical Accident Sequences (S.CS.) ................ .. ..... ........................ ...... ......................... ...... .... . 13-94 13-ii
NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis 13.3 Analysis of Accidents with Hazardous Chemicals ......... .................................... .... ..... 13-95 13.3 .1 Chemical Bum s from Contaminated Solutions During Sample Ana lysis .. ..... 13-95 13.3.1.1 Chemical Accident Description ........................................ ......... ... 13-95 13.3. 1.2 Chemi cal Accident Consequences ....... ... ...... ........ .. ..... ................. 13-95 13.3. 1.3 Chemi cal Process Controls ... .. .... ... .. .... ............................ ..... .. ...... 13-95 13.3 .1.4 Chemi cal Process Survei llance Requirements ............................... 13-95 13.3.2 Nitric Acid Fume Release .................... .... ... ......................................... .... ..... 13-96 13.3.2.1 Chemi cal Accident Description ..... .. ......... ....... ....... .. .......... .. .... .... 13-96 13.3.2.2 Chemi cal Acc ident Consequences .......................... ...................... 13-96 13.3.2.3 Chemi cal Process Controls .. ......................... ..... .. ............ ............. 13-96 13.3.2.4 Chemi cal Process Surveillance Requirements ....... .... .................... 13-96 13.4 References ............................................................................... .... .............. .. .............. 13-97 13-iii
NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis FIGURES Figure 13-1. Integrated Safety Analysis Process Flow Diagram ..................................................... 13-4 Figure 13-2. Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident .................... 13-49 TABLES Table 13-1. Likelihood Categories ............................................................................................... 13-5 Table 13-2. Qualitative Likelihood Category Guidelines .............................................................. 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 ................. ..... .................. .... ................ 13-6 Table 13-5. Radioisotope Production Facility Preliminary Hazard Analysis Accident Sequence Category Designator Definitions ................................................................ 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 ......... ........ ....... ........................ ... .... ... ... ..... 13-9 Table 13-7. Radioisotope Production Facility Preliminary Hazards Analysis Primary Process Nodes and Subprocesses (2 pages) ............... .. .......................................................... 13-10 Table 13-8. Crosswalk of Radioisotope Production Facility Preliminary Hazards Analysis Process Nodes and Top-Level Accident Sequence Categories ..... ..... ... ............... ...... 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) ..................... ..... ...... ..................... 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 ............................ ..... ................................... 13-32 Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) ......... .... .... ............... .......... .. ....... ... ................ 13-33 Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) ..... ..... ................... 13-40 Table 13-18. Source Term Parameters ............................ ............................ ............ ........... ........... 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 Release Consequence Summary at 100 Meters ..... 13-49 Table 13-21. Maximum Bounding Inventory ofRadioiodine [Proprietary Information] ................ 13-55 Table 13-22. Target Dissolver Offgas Accident Total Effective Dose Equivalent.. .... ............ .. ...... 13-58 Table 13-23 . Bounding Radionuclide Liquid Stream Concentrations (4 pages) .. ... ................. .... ... 13-60 Table 13-24. Analyzed Accidents Sequences (9 pages) ..... ... ....... ........ ... ... ...... ....................... ....... 13-75 Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (2 pages) ............ ........ ... ..... ............ ..... ..................................... .... .................... .... ... 13-84 Table 13-26. Accident Sequence Category Definitions ............................. .................................... 13-85 13-v
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 99 mTc technetium-99m 235 U uranium-235 241 Am americium-241 AAC augmented administrative control AC administrative control ACI American Concrete Institute AEC active engineered control 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 release fraction ASCE American Society of Civil Engineers CDE committed dose equivalent CEDE committed effective dose equivalent CFR Code of Federal 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 irradiation ET A event tree ana lysis FEMA Federal Emergency Management Agency FMEA fai lure modes and effects analysis FT A fault tree analysis HAZOP hazards and operability HEGA high-efficiency gas adsorption HEPA high-efficiency particulate air HIC high-integrity canister HN03 nitric acid HV AC heating, venti lation, and air conditioning IBC International Building Code IROFS items relied on for safety IRU iodine removal unit ISA integrated safety analysis ISG Interim Staff Guidance IX ion exchange LEU low enriched uranium LPF leak path factor MAR material at risk Mo molybdenum MURR University of Missouri Research Reactor NaOH sodium hydroxide NDA nondestructive assay NIOSH National Institute for Occupational Safety and Health NOx nitrogen oxide 13-vi
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis NOAA National Oceanic and Atmospheric Administration NRC U.S . Nuclear Regulatory Commission NWMI Northwest Medical Isotopes, LLC NWS National Weather Service OSTR Oregon State University TRIGA Reactor osu Oregon State University P&ID piping and instrumentation drawing PEC passive engineered control PFD process flow diagram PHA preliminary hazards analysis PMP probable maximum precipitation QRA quantitative risk assessment RASCAL Radiological Assessment System for Consequence Analysis RF respirable fraction RPF Radioisotope Production Facility RSAC Radiological Safety Analysis Code SNM special nuclear material SSC structures, systems, and components ST source term TCE trichl oroethy lene TEDE total effective dose equivalent u uranium U.S. United States UN uranyl nitrate 13-vii
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- NORTHWEST MlDtCAL ISOTOPES Units oc degrees Celsius OF degrees Fahrenheit Ci cune Cm centimeter ft feet ft3 cubic feet g gram hr hour in. 2 square inch kg kilogram km kilometer km2 square kilometer L liter lb pound m meter M molar m3 cubic meter mg milligram m1 mile mi2 square mile mil thousandth of an inch mm minute mrem millirem oz ounce ppm parts per million rem roentgen equivalent man sec second Sv sievert wk week wt% weight percent yr year 13-viii
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 (99Mo) 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 packaged and transported to medical industry users where the radioactive decay product, technetium-99m 99
( mTc), 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 transportation activities are not analyzed in this 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 Packaging and shipping LEU targets to the university reactor network for irradiation Returning irradiated LEU targets for dissolution, recovery, and purification of 99Mo Recovering and recycling LEU to minimize radioactive, mixed, and hazardous waste generation Treating/packaging wastes generated by RPF process steps to enable transport to a disposal site Chapter Organization Section 13.1 describes hazard and accident 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 preliminary hazards analysis (PHA) (NWMI-2015-SAFETY-001, NWMI Radioisotope Production Facility Preliminary Hazards Analysis). The PHA discussion in Section 13.1.3 identifies the accident scenarios that required further evaluation.
Section 13.2 presents 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 electrical power Section 13.2.6 discusses natural phenomena accidents Section 13.2. 7 identifies 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.
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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.
The 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 latory guidelines.
Each event with an adverse consequence that involves licensed material or its byproducts is evaluated for risk 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 lure analysis data (failure frequencies) for use in the event tree ana lysis 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.
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~ * *! NORTHWEST MmtCAL ISOTOf'ES NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis Design and Design and Safety ISATeam Engineering NRCReview Functions Functions Deve lop process Initiate ISA process descript ions, PFDs, by collecting P& IDs preliminary data Perform PHA on Identify preliminary facility operations hazards and consequences (radiological, Categorize events criticality, chemical, for likelihood, fire, extern al) using consequence, regulato ry guides and risk where applicable l
Develop CSAs, FHA, Indeter-minate, Document and other support high, or ~ identified low-risk documents intermediate events (no IROFS) risk?
Yes +
Perform QRA to quantitatively evaluate risk and identify IROFS High or No intermediate risk event?
Yes Design function Identify "accident Start Phase 1 development of sequence" and development of IROFS 1------++ develop IROFS and ...--. IROFS boundary specifications/
basis for each in definition packages conceptual complete QRA for each IROFS drawings Complete Phase 1 Develop PSAR, ISA development of
~----++ summary, technical IROFS boundary specifications definition packages I ISA team review and recommendation I
for approval Management approval of ISA basis NRC review of document r----------+ - - - - - - - --1* license submit to NRC application 1cm_r Figure 13-1. Integrated Safety Analysis Process Flow Diagram 13-4
NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and Risk Matrix Table 13-1 shows the accident likelihood Table 13-1. Likelihood Categories categories applied to the RPF ISA process.
Table 13-2 shows qualitative guidelines for Event frequency limit applying the likelihood categories from Not unlikely 3 More than I 0- 3 events per year Table 13-1. Table 13-3 shows accident consequence severity categories from Unlikely 2 Between I 0-3 and I 0- 5 events 10 CFR 70.61, "Performance Requirements." per year Table 13-4 shows the RPF risk matrix, which Highly unlikely Less than 10-5 per events per is a product of the likelihood and consequence year severity categories from Table 13-1 and Table 13-3, respectively.
Table 13-2. Qualitative Likelihood Category Guidelines 11.* 3 An event initiated by a human error Initiator 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 combustibles 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 process variable monitored or unmonitored by a control system 3 A valve out of position or a valve that fai ls to seat and isolate 3 Most standard industrial component failures (valves, sensors, safety devices, gauges, etc.)
3 An adverse chemical reaction caused by improper quantities ofreactants, out-of-date reactants, out-of-specification reaction environment, or the wrong reactants are used 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 external chall enges applied (e.g., spontaneous rupture of an all-welded dry nitrogen system pipe operating at or below design pressure in a clean, vibration-free environment) 1-2 An adverse chemical reaction when proper quantities of in-date chemicals are reacted in the proper environment Natural phenomenon such 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 High Consequence category 3
Workers
- Radiological dose* > I Sv Off-site public
- Radiological dose*
Environment consequence (I 00 rem) > 0.25 Sv (25 rem)
- Airborne, radiologically
- Toxic intake > 30 mg contaminated nitric acid soluble U
> 170 ppm nitric acid (AEGL-3,
- Airborne, contaminated 10-min exposure limit) nitric acid > 24 ppm
- Unshieldedb nuclear criticality nitric acid (AEGL-2, 60-min exposure limit)
Intermediate 2 . Radiological dose* between . Radiological dose* 24-hr radioactive consequence 0.25 Sv (25 rem) and I Sv between 0.05 Sv (5 rem) release > 5,000 x and 0.25 Sv (25 rem)
. (100 rem)
Airborne, radiologically
- Airborne, contaminated nitric acid > 0.16 ppm Table 2 of 10 CFR 20, 0 Appendix B contaminated nitric acid
> 43 ppm nitric acid (AEGL-2, nitric acid (AEGL-1, 10-min exposure limit) 60-min exposure limit)
Low Accidents with lower Accidents with lower Radiological consequence radiological, chemical, and/or radiological, chemical, releases producing toxicological exposures than those and/or toxicological lower effects than above from licensed material and exposures than those above those listed above byproducts of licensed material from li censed material and from licensed byproducts of licensed material material Source: I 0 CFR 70.61 , " Performance Requirements," Code of Federal Regulations, Office of the Federal Register, as amended.
- As total effective dose equivalent.
b A shielded criticality accident is also cons idered a high-consequence event.
c IO CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended.
AEGL Acute Exposure Guideline Level. u = uranium.
Table 13-4. Radioisotope Production Facility Risk Matrix Likelihood of occurrence Severity of Highly unlikely Unlikely Not unlikely consequences (Likelihood category 1) (Likelihood category 2) (Likelihood Category 3)
High consequence Risk index = 3 Risk index = 6 Risk index = 9
~
(Consequence category 3) Acceptable risk Unacceptable ri sk Unacceptabl e risk Intermediate consequence Risk index = 2 Risk index= 4 Risk index = 6 (Consequence Acceptable risk Acceptable risk {
Unacceptable risk category 2)
Low consequence Risk index = 1 Risk index = 2 Risk index = 3 (Consequence category 1) Acceptable risk Acceptable risk Acceptable risk 13-6
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0 0 NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis NORTHWEST MEDICAL ISOlWES 13.1.1.2 Accident Consequence Analysis The 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 of 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., use 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 basis (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 processing High radiation dose due to accidental nuclear criticality Toxic uptake of licensed material by workers or the public during 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.
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NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis 13.1.1.4 Hazards and Operability Study Method 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 adverse 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 frequency given the initiator. ETA is also used in the QRA process to demonstrate that the IROFS, selected 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 analysis 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-level 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 is included in this category)
Any other event potentially related to unique faci lity operations 13-8
NWM l-2013-02 1, Rev . 2 Chapter 13.0 - Accident Ana lysis The PHA (NWMl-2015-SAFETY-OOI) identifies Table 13-5. Radio isotope Production Facility and categorizes accident sequences that require Preliminary Hazard Analys is Accident further evaluation. Table 13-5 defines the top- Seq uence Category Designator Defi nitions level accident sequence notation used in the RPF PHA top-level accident PHA. sequence categorya Definition Table 13-6 provides a crosswalk between the PHA S.C. Criticality top-level accident sequence categories and the S.F. Fire or explosion NUREG-1537, Guidelines for Preparing and S.R. Radiological Reviewing Applications for the Licensing of Non-Power Reactors - Format and Content, Part 1 S.M. Man-made Interim Staff Guidance (ISG) accident initiating S.N. Natural phenomena events listed above. As noted at the bottom of S.CS. Chemical safety Table 13-6, PHA accident sequences involve one or more of the NUREG-1537 Part 1 ISG accident
- The alpha category designator is fol lowed in the PHA by a two-digit number "XX" that refers to the specific accident initiating event categories, as noted by ./ in the sequence (e.g., S.C.01 , S.F.07). Specific accident sequences corresponding table cell, but the PHA accident are di scussed in Sections 13.1.3 and 13.3 .
sequences themselves are not necessari ly initiated PHA = prelimi nary hazard analysis.
by the ISG accident initiating event. Table 13-6 shows how PHA accident sequences correspond with ISG accident initiating events, and demonstrates that the PHA considers the full range of accident events identified in the ISG.
Table 13-6. Crosswalk ofNUREG-1537 Part I Interim Staff Guidance Accident Initiating Events versus Radioisotope Production Facility Preliminary Hazards Analysis Top-Level Accident Sequence Categories NUREG-1537a Part 1 ISG accident initiating event category Criticality accident Loss of electrical power External events (meteorological,
,/
,/
PHA Top-Level Accident Sequence Categoryb
,/
,/
,/
,/
,/
,/
,/ ,/
seismic, fire, flood)
Critical equipment malfunction ,/ ,/ ,/ ,/ ,/
Operator error ,/ ,/ ,/ ,/
Facility fire (explosion is included in ,/
this category)
Any other event potentially related to ,/ ,/
unique faci lity operations
- NURE0-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of No n-Power Reactors - Format and Content, Part I, U.S. Nuclear Regulatory Commission, Office ofNuclear Reactor Regulation, Washington, D.C.,
February 1996.
h PHA accident sequences involve one or more of the NURE0 -1 537 Part I ISO accident initiating event categori es, as noted by an ./ in the corresponding tabl e cell, but the PHA sequences themselves are not necessarily initiated by the ISO accident initiating event.
ISO = lnterim StaffOuidance. PHA = preliminary hazard analysis.
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~* * ~ NOmfWlST MEOtC.Al ISOTOP£S 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 Target fabrication
- Fresh uranium receipt and storage process
- 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 scrap recovery
- Target assembly, loading, inspection, quality checking, verification, packaging and storage 2.0.0 Target dissolution * [Proprietary Information]
process * [Proprietary Information]
- Primary process offgas treatment
- Fission gas retention 3.0.0 Molybdenum recovery
- Feed preparation and purification process
- First stage recovery
- First stage purification preparation
- First stage purification
- Second stage purification preparation
- Second stage purification
- Final purification adjustment 99
- Mo preparation for shipping 4.0.0 Uranium recovery and
- Impure uranium lag storage recycle process
- First-cycle uranium recovery
- Second-cycle uranium purification
- Product uranium lag storage
- Other support (storage vessels, transfer lines, solid waste handling for resin bed replacement) 13-10
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~* *~ 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 Waste handling system
- Liquid waste storage process
- High dose liquid waste volume reduction
- Condensate storage and recycling
- Concentrated high dose liquid waste storage/preparation
- Low dose liquid waste volume reduction and storage
- Liquid waste solidification
- Solid waste hand! ing
- Waste encapsulation
- TCE solvent reclamation
- Mixed waste accumulation 6.0.0 Target receipt and
- Cask receipt and target unloading disassembly process
- Target Inspection
- Target disassembly
- [Proprietary Information]
- Target disassembly stations
- Gaseous fission product control
- [Proprietary Information]
- Empty target hardware handling 7.0.0 Ventilation system * (No subprocesses identified in PHA. Ventilation system provides cascading pressure zones, a common air supply system with makeup air as necessary, heat recovery for preconditioning incoming air, and HEPA filtration.)
8.0.0 Natural phenomena,
- Natural phenomena man-made external
- Man-made external events events, and other facility
- Chemical storage and preparation areas operations
- On-site vehicle operation
- General storage, utilities, and maintenance activities
- Laboratory operations
- Hot cell support activities
- Waste storage operations including packaging and shipment 99 Mo molybdenum-99 PHA preliminary hazards anal ysis.
HEPA high-efficiency particulate air. TCE = trichloroethylene.
Table 13-8 shows a crosswalk that identifies the applicability of RPF PHA top-level accident sequence categories to the primary process nodes. The information in this table is referenceable to Table 13-6 and ultimately shows the relationship between the PHA process nodes and the NUREG-1537 Part 1 ISG accident initiating event categories via the PHA top-level accident scenario categories.
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NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis Table 13-8. Crosswalk of Radioisotope Production Facility Preliminary Hazards Analysis Process Nodes and Top-Level Accident Sequence Categories PHA Top-Level Accident Sequence Category
- - ,, ,, ,, m1111*
Target 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)
Ventilation 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 that the accident sequence category applies to the process node. If it does not, the cell is blank.
PHA = preliminary haza rds analys is.
13.1.3 Preliminary Hazards Analysis Resu lts 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. Numerous 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 35U). 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 , Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors. The RPF separates 99Mo 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 separate from and mixed with licensed material, that present chemical hazards to workers. The National 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 licensed material.
The facility can also be impacted by various internal and external man-made and natural phenomena events that have the potential to damage structures, systems, and components (SSC) that control the licensed material, thereby leading to intermediate- and high-consequence events.
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NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis Known and credited safety features for normal operations include:
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 process system nodes were analyzed using a HAZOP approach with special emphasis on criticality, radiological, and chemical safety hazards. Fire safety issues are addressed in every node and addressed generally in Node 8.0.0. Fire safety issues include the explosive hazard associated with hydrogen gas generation via radiolytic decomposition of water in process solutions and due to certain chemical reactions encountered during dissolution processes. Most hot cell processing areas contain very 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 being low consequence and/or low frequency Unacceptable accident sequences that require further evaluation via the QRA process An accident sequence number was assigned to each accident initiator that results in the same, or similar, bounding accident sequence result and consequence. The same accident sequence designator can appear in multiple nodes. (Table 13-5 provides definitions of accident sequence category designators.)
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' ~* * ~ 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% mu 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 issue - Too much administrative control on 1.8. 1.6, 1.8.2.2, 1.8.2.3, batches of uranium uranium mass m one interaction limit during 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 batches closer scrap U, [Proprietary together than the approved Information], and targets interaction control di stance 1.2.1.1, 1.2.1.11, 1.2.1.14, Failure of safe geometry Accidental criticality S.C.04, Spill of fissile 1.2.1.25, 1.3.1.1, 1.3.1.6, confinement from fissi le 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 solution Accidental criticality S.C.05, Leak of fissile leaks out of safe geometry from fissile solution not solution into heating/
confinement into the confined in safe cooling jacket on vessel heating/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 13-14
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 Accidental criticality S.C.19, Failure 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 (installation, 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 solution or component causing 1.9.1.8 entering a system not fissile solution 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 Accidental criticality S.C.20, Fai lure of 1.9.1.20 high uranium solids from fissile solution not concentration limits -
accumu late in safe geometry confined to safe Precipitation of uranium vessel geometry and in safe geometry tank interaction controls within all 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 system (water line, purge line, entering a system not auxiliary system at a fill chemical addition line) due to necessarily 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 Accidental criticality S.C.11, Fissi le material confinement due to from fiss ile solution not contamination of inadvertent transfer to confined in safe contactor regeneration U-bearing solution across a geometry aqueous waste stream -
boundary into non-favorable boundary to unsafe 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 confined in safe evaporator condensate -
U-bearing solution across a geometry boundary to unsafe boundary into non-favorable geometry system geometry 1.6.1.3 Failure of safe geometry Accidental criticality S.C.12, Wash of confinement due to from fissile solution not [Proprietary Information]
inadvertent transfer to confined in safe with wrong reagent U-bearing solution across a geometry contamin ating wash boundary into non-favorable solution with fissile U; geometry boundary to unsafe geometry system 1.1.1.11 Dusty surface generated Potential exposure to S.F.01, Pyrophoric fire during shipping on uranium workers due to airborne in uranium metal pieces spontaneously ignites uranium generation due to pyrophoric nature of uranium 13-15
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!* *~ 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 Hydrogen buildup in tanks or Explosion leading to S.F.02, Accumulation of
- 1. 7 .1.11 system, leading to explosive radiological and flammable gas in tanks concentrations criticality concern s or systems 1.4.1.17, 1.4.1.21, and Fire in process system Radiological and S.F.07, Fire in nitrate 1.4.1 .23 containing high concentration criticality issue - extraction system -
uranium spreads the uranium Radiological airborne flammable solvent with release of uranium and uranium uncontrolled spread of uranium outside safe geometry confinement 1.6.1.6, 1.6.1.9, and 1.6. 1.1 2 Air inleakage into the Accidental criticality S.F.03, Hydrogen reduction furn ace during Hz issue - Uncontroll ed detonation in reduction purge cycle or Hz in leakage spread of uranium furnace into reducti on furnace before outside safe geometry inerting with nitrogen can lead confi nement to an explosive mixture in the presence of an igniti on source 1.6.1.8 Loss of cooling of exhaust or Radiological issue - S.F.04, High temperature fire in the reduction furnace Potential accelerated damage to process leads to high temperatures in release of high-dose ventilation system due to downstream ventilation radionuclides to the loss of cooling in component and accelerated stack (worker and reduction furnace release of adsorb public exposure) exhaust or fire in radionuclides reduction furnace 1.2.1.l I, 1.2.1.14, 1.4.1.1 7, High concentrati on uranium Radiologica l release of S.R.03, Solution spray 1.4.1.1 9, 1.4.1.20, 1.4.1.2 I, solution is sprayed from the uranium solution spray release potentially 1.4.1.23, 1.4.2.6, 1.4.3. 14, system, causing high airborne th at remains suspended creating airborne 1.4.3.26, 1.4.3.3 I, 1.4.3.32, radioactivity in the air, exposing uran ium above DAC 1.7.1.11 , 1.7.1.14, an d 1.9.1.6 workers or the public limits 1.2.1.1 I , 1.2.1.12, 1.2.1.14, High concentration uranium Potential radiological S.R.01 , Uranium-1.2.1.25, 1.3.1.1 , 1.3.1.6, solution is spilled from the exposure to workers contaminated solution 1.3.1.11 , 1.3.1.17, 1.4.1.17, system from uranium- spill 1.4.1.18, 1.4.1.19, 1.4.1.21, contaminated solution 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 13-16
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~* * ~ NOWTtfWUT 11£DtCAl tSOTOPU 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, Boiling or carryover of steam Radiological release S.R.04, Liquid enters 1.7.1.21, and 1.7.1.22 or high concentration water from retention beds process vessel ventilation vapor into the primary system damaging IRU or ventilation system, affecting retention beds releasing retention beds from partial or retained radionuclides complete loss of cooling system capabilities 1.3.1.16 and 1.4.1.24 High-dose solution (fai lure of Potentially high S.R.05, High-dose the uranium recovery process) radiological exposure to solution enters the UN results in high-dose workers blending and storage radionuclides entering the first tank stage of processing uranium
[Proprietary Information]
(eventually handled by the worker) 1.8.3 .7 Loading limits are not adhered Hi gh-dose to workers or S.R.28, Target or waste to by the operators or the the public from shipping cask not loaded closure requirements are not improperly shielded or secured according to satisfied, and the cask does cask procedure, leading to not provide the containment or personnel exposure shielding function that it is designed to perform mu uranium-235. PHA process hazards analys is.
DAC derived air concentration . u uranium.
H2 hydrogen gas. UN uranyl nitrate.
!RU iodine removal unit.
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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 2.1.1.1, 2.1.1.11 , Fai lure of safe geometry Accidental criticality from S.C.04, Failure of 2.1.1.13, 2.1.1.17' confinement fissile solution 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 solution leaks out of safe fissile solution not confined in solution in to geometry confinement into safe geometry heating/cooling jacket the heating/cooling jacketed on vessel space 2.1.1.3 Uranium solution is Accidental criticality from S.C.07, Leak of fissile transferred via a leak fissile solution not confined in solution 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 dimension fissile solution not confined in passive design feature; safe geometry component safe-geometry dimension 2.1.1.12, 2.1.1.15, and Fai lure of safe-geometry Accidental criticality from S.C.13, Fissile solution 2.3.1.4 confinement fissile solution not confined in enters the NOx scrubber safe geometry 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 ventilation system Fissile solution entering a overflow to process system not necessarily designed ventilation involving for fissile solutions fissile material 2.3.4.11 Uranium enters carbon Accidental criticality from S.C.24, Build-up of high retention bed dryer where it fissi le material or solution not uranium particulate in can mix with condensate to confined in safe geometry the carbon retention bed form a fissile solution dryer system
- 2. 1.1.33 and 2.1.1.34 Uranium solution backflows Accidental criticality and high S.C.08, System into an auxiliary support radiological dose - High-dose backflow into auxiliary system (water line, purge and fissile solution entering a support system line, chemical addition line) system not necessarily designed due to various causes for fissile solutions that exist outside of hot cell walls 13-18
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...... NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis
~* * ~ NOITHWEST MEDM:.Al tsOTOP£S 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 explosive concentrations concerns 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-dose radionuclides 2.1 .1.1, 2.1 .1.2, High-dose and/or high- Potential radiological exposure 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, solution is spilled from the and/or high uranium- liquid spi ll of high-dose 2.2.1.12, 2.2.1 .15, system contaminated solution and/or high uranium 2.3.6.5, 2.3.6.12, and concentration soluti 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, Spill leading to spray-type Radiological dose from S.R.03, Spray of product 2.2.1.1 5, and 2.3. 6.13 release, causing airborn e airborne spray of product solution in hot cell area radioactivity above DAC solution from systems limits 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 release, affecting workers and heavy vapor or solution 2.3.4.12, and 2.3.4.17 solutions 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, release radionuclides releasing radionuclides 2.3.1.17, 2.3 .1.22, A spi II of low-dose Potential radiological dose to S.R.02, Spill oflow-2.3. 1.24, 2.3.2.17, condensate occurs for a workers and the public from dose condensate 2.3.2.22, 2.3.2.24, variety of reasons from the spilled liquid 2.3.3.8, 2.3.3.20, confinement 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 Low temperatures in the Potential radiological dose to S.R.07, Loss of IRU inlet gas stream drives workers and the public from temperature control on release of iodine from the iodine above regulatory limits the IRU leads to unit premature release of high-dose iodine 2.3.3.22 and 2.3.5.8 Liquid and water vapor in Potential radiological dose to S.R.04, Liquid/high the IRU inlet gas stream workers and the public from vapor in the IRU leads drives release of iodine from iodine above regulatory limits to premature release of the unit high-dose iodine 2.3.4.4, 2.3 .4.5, and Loss of vacuum pumps in Potential radiological dose to S.R.08, Loss of vacuum 2.3.4.6 the dissolver offgas workers and the public from pumps treatment system leads to spilled liquid pressure buildup inside the process and potential release of radionuclides from the system upstream 2.3.4.11 Uncontrolled loss of media Potential radiological dose to S.R.09, Loss ofIRU and contact with a liquid workers and the public from media to downstream with potential for premature iodine above regulatory limits dryer release of the adsorbed iodine 2.3.3.28, 2.3.4.19, Using the wrong retention Potential radiological dose to S.R.10, Wrong retention 2.3. 5.9, 2.3.4.15, and media (IRU or carbon beds) workers and the public from media added to bed or 2.3.5.11 or using saturated media radionuclides above regulatory saturated retention with potential for ineffective limits media adsorption of high-dose gaseous radionuclides 2.3.4.16, 2.3.5.5, and An event causes damage to Potential radiological dose to S.R.09, Breach of an 2.3.5.12 the structure holding the workers and the public from IRU or retention bed retention media, and radionuclides above regulatory resulting in release of retention media is released limits the media to an uncontrolled environment 2.1.1.33 and 2.1.1.34 High-dose process solution High radiological dose - High S.R.11, System backflows into an auxiliary dose process solution enters a backflow of high-dose support system (water line, system that exits outside of the solution into an purge line, chemical hot cell walls auxiliary support system addition line) due to various and outside the hot cell causes boundary DAC derived air concentration. NOx nitrogen oxide.
IRU iodine remova l unit. PHA process hazards analysis.
13-20
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 3.3. 1.24 Higher radiation dose due to Higher localized dose in NIA hold-up accumulation or hot cell boundary transient batch differences (unoccupied by workers) 3.2.3 .7, 3.2.4.7, 3.4.3.7, 3.4.4.7, Chemical spills of Standard industrial NIA 3.6.3.7, and 3.6.4.7 nonradiological ly accident - Chemical contaminated bulk exposure (not involving chemicals licensed material) to workers
- 3. 7.4.5 and 3. 7.4.6 Dropped cask or cask Standard industrial NIA component during loading accident - Worker injury or handling 3.7.4.2, 3.7.5.2, and 3.7.5.3 Mo product is exposed with Potential dose to the NIA - Addressed by no shielding as the result of public and/or environment DOT packaging and an accident, shipment due to release or transportation mishap, or shipment mishandling of Mo regulations mishandling after leaving product during 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 criticality from S.C.04, Fai lure of 3.1.2.7, 3.1.2.13, 3.1.2.16, confinement fissi le soluti on not confinement in safe 3.1.2.17, 3.2.1.6, 3.2.1.10, confined in safe geometry geometry; spi ll of 3.2.1.20, 3.2.1.22, 3.2.1.23, fi ssile 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 process Accidental criticality issue S.C.06, System and 3.8.1.4 ventilation system - Fissile solution entering overflow to process a system not necessarily ventilation involving designed for fissile fissile material solutions 3.1.1.23, 3.2.1.23, 3.2.5.23, and Uran ium solution is Accidental criticality from S.C.07, Leak of 3.8. 1.22 transferred via a leak fissi le solution not fissile solution between the process system confined in safe geometry across auxiliary and the heater/cooling system boundary jackets or coils on a tank or (chi lled water or in an exchanger steam) 3.2.1.4, 3.2.1.5, 3.2.2.3, 3.2.2.4, Fissile product solution Criticality safety issue - S.C.10, Inadvertent 3.2.2.5, 3.2.3.6, and 3.2.4.6 transferred to a system not Fissile solution directed to transfer of solution designed for safe-geometry a system not intended for to a system not confinement fissile solution designed for fissile solutions 13-21
- .*.NWMI NWMl-2013-021, Rev . 2 Chapter 13.0 -Accident Analysis
. ~ * *! NORTIIW'En MEOtCAL ISOTOP£S 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, Fai lure of safe-geometry Accidental criticality from S.C. 19, Failure of 3.2.5.13, and 3.8.1.12 dimension fissile solution not passive design confined in safe geometry feature; component safe-geometry dimension 3.1.1.25, 3.2.5.25, 3.3.1.25, Hydrogen buildup in tanks Explosion leading to S.F.02, 3.5.1.25, and 3.8.1.24 or system, leading to radiological and criticality Accumulation of explosive concentrations concerns flammable gas in tanks or systems 3.7.1.1 , 3.7.1.2, 3.7.2.1 , 3.7.3. 1, Operator spi ll s Mo product Radiological spill of high- S.R.O1, Radiological 3.7.3.2, and 3.7.4.l solution during remote dose Mo solution spill of Mo product handling operations during remote handling 3.1.1.9, 3.1.1.14, 3.1.1.23, 3.1.2.7, Spill of product solution in Radiological dose from S.R.01, Spill of 3.1.2.13, 3.1.2.16, 3.1.2.17, the hot cell area spill of product solution product solution in 3.2.1.6, 3.2.1.20, 3.2.1.22, from systems hot cell area 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 13-22
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. 1.1.9, 3.2. 1.10, 3.2.1.22, 3.2.2.7, Spill leading to spray-type Radiological dose from S.R.03 , Spray of 3.2.2.9, 3.2.3.8, 3.2.3.10, 3.2.4.10, release, causing airborne airborne spray of product product solution in 3.2.5.9, 3.3.1.9, 3.3.1.18, 3.3.1.22, radioactivity above DAC soluti on from systems hot cell area 3.3.2.7, 3.4.1.10, 3.4.1.22, 3.4.2.7, limits for exposure 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, Boiling or carryover of Radiological release from S.R.04, Loss of 3.3.1.4, 3.3.1.7, 3.3.1.16, 3.5.1.4, steam or high-concentration retention beds cooling, leading to 3.5.1. 7, 3.5.1.16, 3.5.1.22, 3.8.1.7, water vapor into the primary liquid or steam and 3.8.1.13 process offgas ventilation carryover into the system affecting retention primary offgas beds with partial or treatment train complete loss of cooling system capabilities 3.7.4.3 A Mo product cask is Potential dose to workers, S.R.12, Mo product removed from the hot cell the public, and/or is released during boundary with improper environment due to sh ipment shield plug installation release or mishandling of Mo product during transit 3.3.1.23, 3.3.2.16, 3.4.1.22, High-dose radionuclide High-dose radionuclide S.R.13, High dose 3.5.1.23, and 3.6.1.23 solution leaks through an solution that leaks to the radionuclide interface between the environment through containing solution process system and a another system to expose leaks to chilled heating/cooling jacket coil workers or the public water or steam into a secondary system condensate system (e.g., chilled water or steam condensate) releasing radionuclides to workers, the public, and environment
- 10 CFR 71, "Packagi ng and Transportation of Radioactive Materi al," Code of Federal Regulations, Office of the Federal Register, as amended.
DAC derived ai r concentration . NI A not applicable.
DOT U.S. Department of Transportation. PHA process hazards analysis.
Mo mol ybdenum .
13-23
..;....:..;.... NWMI
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- NORTHWlST MEDICAL ISOTOPES 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, Tank overflow into Accidental criticality S.C.06, System overflow 4.2. I . I 7, 4.2. I .18, 4.2.3.6, 4.2.8.4, process ventilation system issue - Fissile solution to process ventilation 4.2.8.18, 4.2.10.4, 4.3.1.4, 4.3.1.6, enters a system not involving fissi le material 4.3. I .18, 4.3 . I .19, 4.3 .3 .6, 4.3.8 .4, necessarily designed 4.3.8. I 8, 4.3 . I 0.4, 4.4. I .4, for fissile solutions 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, Uranium solution Accidental criticality S.C.08, System backflow 4.2.3.7, 4.2.3.8, 4.2.8.7, 4.3.1.7, backflows into an issue - Fissile solution into auxiliary support 4.3.2.4, 4.3.3.4, 4.3.3.7, 4.3.3.8, auxiliary support system enters a system not system 4.3.8. 7, 4.4.1.6, 4.5.2.6, and (water line, purge line, necessarily designed 4.5.3.6 chemical addition line) for fissile solutions due to various causes 4.1.1.14, 4.2.1.14, 4.2.3. 16, Failure of safe geometry Accidental criticality S.C.19, Failure of 4.2.8. 15, 4.3.1.15, 4.3 .3. 16, dimension caused by from fissile soluti on passive design feature; 4.3.8.15, 4.3.9.20, 4.4.1.14, configuration management not confined in safe component safe-4.5.1. I 4, 4.5.2. I 4, and 4.5.3.11 (installation, maintenance) geometry geometry dimension or external event 4.1.1.8, 4.1.1.9, 4.1.1.12, 4.1.1.13, Uranium precipitate or Accidental criticality S.C.20, Failure of 4.1.1.16, 4.2.1.9, 4.2.1.13, other high uranium solids from fissile solution concentration limits 4.2.5.11, 4.2.8.10, 4.2.8.13, accumulate in safe- not confined to safe 4.2.8.14, 4.2.8.17, 4.2.9.18, geometry vessel geometry and 4.3.1.10, 4.3.1.11, 4.3.1.14, interaction controls 4.3.1.17, 4.3.1.18, 4.3.5.11, within allowable 4.2.8.10, 4.3.8.13, 4.3.8.14, concentrations 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, Failure of safe-geometry Accidental criticality S.C.04, Failure of 4.2.1.11 , 4.2.1.15, 4.2.1.24, 4.2.2.1 , confinement due to spill from fissile solution confinement in safe 4.2.3.11, 4.2.3.13, 4.2.3 .18, of uranium solution from not confined in safe geometry; spill of fissile 4.2.3.22, 4.2.3.23, 4.2.3.24, the system geometry material solution 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, 4.3.3.22, 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 13-24
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 solution confinement in safe inadvertent transfer to not confined in safe geometry; transfer of U-bearing resin to the U geometry U-bearing resin to U IX IX waste collection tanks waste collection tanks through a broken retention element 4.2.5.5, 4.3.1.9, 4.3.5.5, and Fai lure of safe-geometry Accidental criticality S.C.14, Failure of 4.5.1.5 confinement due to from fissile solution confinement in safe inadvertent transfer to not confined in safe geometry; transfer of U-bearing solution to the geometry U-bearing solution to U IX waste collection U IX waste collection tanks tanks 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 solution or resins to spent mass in waste stream resin waste stream resin tanks 4.2.9.10, 4.2.9.19, 4.2.9.21 , Uranium is inadvertently Accidental criticality S.C.09, Carryover of 4.2.9.23, 4.2.10.10, 4.2.10.1 2, carried over from the from fissile solution uranium to the condenser 4.3.9.10, 4.3 .9.19, 4.3.9.21, concentrator (I or 2) to the not confined in safe or condensate tanks 4.3.9.23, 4.3.10.10, and 4.3.10.12 condenser and geometry subsequently, the condenser condensate collection tanks 4.2.9.36 and 4.3.9.36 Uranium solution is Accidental criticality S.C.07, Uranium-transferred via a leak from fissile solution containing solution leaks between the process not confined in safe to chilled water or steam system and heater/cooling geometry condensate system 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 radionuclide release, heavy vapor or solution 4.2.9.17, 4.2.9.23, 4.2.9.30, of solutions into the affecting workers and into the process 4.2.9.32, 4.2.10.14, 4.3.1.10, process ventilation header the public ventilation header causes 4.3.1.18, 4.3.1.24, 4.3.9.11 , can cause poor downstream fai lure of 4.3.9.14, 4.3.9.17, 4.3.9.23, performance of the retention 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 radionuclides and 4.5.2.8 13-25
.*:~*:~*....:* NWM I
~* * ~ NOmtMST MEDtcAL lSOTOPlS 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, High-dose radionuclide Radiological release of S.R.01, Spill of product 4.2.1.11, 4.2.1.15, 4.2.1.24, 4.2.2.1, solution is spilled from the high-dose solution solution in hot cell area 4.2.2.4, 4.2.3.11, 4.2.3.13, 4.2.3.18, system with potential to 4.2.3.22, 4.2.3.23, 4.2.3.24, impact workers, the 4.2.4.10, 4.2.5.10, 4.2.6.11, 4.2.7.8, public, or environment 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 , High -dose radionuclide Radiological release of S.R.03, Spray of product 4.2.3.13, 4.2.3.18, 4.2.3.22, solution is sprayed from high-dose spray that solution in hot cell area 4.2.3.23, 4.2.4.10, 4.2.5. I 0, the system, causing high remains suspended in 4.2.6.11, 4.2.8.11, 4.2.8.1 6, airborne radioactivity the air, giving high 4.2.8.23, 4.2.9.16, 4.2.9.28, dose to workers or the 4.2.9.34, 4.2.9.35, 4.3. 1.12, public 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 High-dose radionuclide High-dose S.R.13, High-dose, 4.3.9.37 solution leaks through an radionuclide solution radionuclide-containing interface between the that leaks to the solution leaks to chilled process system and a environment through water or steam heating/cooling jacket coil another system to condensate system into a secondary system expose workers or the (e.g., chilled water or public steam condensate),
releasing radionuclides to workers, the public, and environment 13-26
~**; :* NWMI
' ~* * ~ NOATitWUT MEDtCAL ISOTOPES NWM 1-2013-021 , 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 .1.25, 4.2.1 .26, 4.2.8.25, Hydrogen buildup in tanks Expl osion leading to S.F.02, Accumulation of 4.3. 1.27, 4.3. 8.25, 4.4.1.25, or system, leading to radiological and flammable gas in tanks 4.5 .1.25, 4.5.2.25, and 4.5.3.21 explosive concentrations criticality concerns or systems 4.1 .1.24, 4.2. 1.25, 4.2.8.24, Higher dose than normal Radiation dose is Hot cell shielding is 4.2.10.18, 4.3. 1.26, 4.3 .8.24, due to double-batching an elevated over normal credited as the normal 4.3. 10.18, 4.4.1.24, 4.5.1.24, activity or due to buildup operational levels, but condition, mitigating 4.5.2.24, and 4.5.3.20 of radionuclides in the does not exceed low safety feature for this system over time consequence values hazard (adverse condition for exposure to does not represent failure workers due to of the safety function of shielding the IROFS) 4.2.4.8 and 4.3.4.8 High temperature Consequence is not Tentatively S.R.14 pre-elution or regeneration fu lly understood reagent causes unknown impact on IX resin 4.2.10.6 and 4.3.10.6 Same as S.C.08 except Low consequence NIA with low-dose solution resulting in from condenser condensate contaminated system 4.2.10.8, 4.2.10.11 , 4.2.10.1 7, Spi ll or spray of low-dose Low consequence NIA 4.3.10.8, 4.3.10.11, and 4.3.10.17 condensate resulting in contaminated surfaces and dose to worker below intermediate consequence dose levels IROFS items relied on for safety. PHA process hazards analysis.
TX ion exchange. u = uranium.
NIA not applicab le.
Uranium Recovery Open Item The following adverse event needs to be further researched.
PHA items 4.2.4.8 and 4.3.4.8 postulate high-temperature 2 molar (M) nitric acid (HN03) solution being used on the uranium purification ion-exchange (IX) media as a pre-elution rinse. The consequence of the bounding accident was not full y understood and needs to be further researched. The likelihood was identified as low, as there are no good causes of the high temperature from the supply tank other than an improper m1xmg sequence. This upset would not cause extremely elevated temperatures nor go undetected.
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....~. NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis
' ~* * ~ Nomtw£ST MEDICAL ISOTOPES 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. 1.1.21, 5.1.2.28, Several tanks or Radiological release may cause a S.R.04, High-dose and 5.1.4.20 components vented to th e high-dose exposure to workers and radi onuclide release due to process vessel venti lation the public high vapor content in system evolve high liquid exhaust vapor concentrations, resulting in accelerated high-dose radionuclide release to the stack from wetted retention beds 5.1.1.22, 5.1.2.26, Catastrophic failure of a Radiological release may cause a S.R.03, High-dose solution 5.1.2.31 , 5.1.2.32, component (high pressure high-dose exposure to workers and spray events from 5.1.4.10, and or detonation) leads to the public equipment upsets may cause 5.1.4.21 rapid release of solution high airborne radioactivity and higher airborne levels 5.1.2.9, 5.1.2.18, Adverse events in the Radiological exposure levels on the S.R.17, Carryover ofhigh-5.1.2. I 9, and concentrator or evaporator low-dose encapsulated waste may dose solution into 5.1.2.21 systems lead to carryover exceed intermediate or high condensate (a low-dose of high-dose solution into consequence levels waste stream) the condenser, resulting in high-dose radionuclides in the low-dose waste collection tanks 5.1.2.33 Normally low-dose vapor Radiological release may cause a S.R.13, Process vapor from in the condenser leaks high-dose exposure to workers and the evaporator leaks across through the boundary into the public the condenser cooling coils the chilled water system into the chilled water system 5.1.5.8 High-dose solution is Radiological release may cause a S.R.18, High-dose solution inadvertently misfed into high-dose exposure to workers and flows into the solidification the solidification hopper the public hopper 5.5.1.1 Due to several potential Radiological issue - Depending on S.R.32, Container or cask initiators, the payload damage from the drop, workers dropped during transfer container or the shipping could receive high-dose radiation cask of high-dose exposure. Unshielded package may encapsulated waste is impact dose rates at the controlled dropped during transfer area boundary.
from the storage location to the conveyance PHA process hazards analysis. TCE trich 1oroethylene.
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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 the target Accidental nuclear criticality S.C.21, Target basket 6.1.2.11 , 6.1.2. 14, and basket fixed-interaction passive leads to high dose to workers passive design control 6.1.2.15 design feature leads to accidenta l and potential dose to the fa ilure 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 handling 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 accumulates more Accidental nuclear criticality S.C.03, Failure 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 specifi c room than allowed and public during handling of violates 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 dose 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 shipment receipt streaming radiation inspection and cask preparation activities due to damaged irradiated target cask 6.1.1.10 Cask involved in in-transit High dose to workers during S.R.29, High dose to accident or targets failed during receipt inspection and workers from release 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, Seal between cask and hot cell 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 lures lead to 6.1.2.16 of causes high airborne radioactivity high dose to workers due to streaming radiation and/or high airborne radioactivity 13-30
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 Improper handling activities High external dose to S.R.19, High target result in high external dose rates workers basket retrieval dose through the hot cell wall when rate removing the target basket and setting it in the target basket carousel shielded 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 heavy degraded shielding hot cell wall providing sh ielding housing) with a hoist shielding leads to striking the wall and damaging the hot cell wall shi elding 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 unprocessed Information] results in long-term [Proprietary heating outside of target housing Information] removed from targets leads 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 premature opening of the high airborne radioactivity from irradiated target valve during processing of a dissolution tank occurs previously added batch results in when the upper 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, Bagless 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 process hazards analysis.
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- .**.* NWMl-2013-021, Rev . 2 Chapter 13.0 -Accident Analysis
' ~* * ~ NomtW£ST MEDICAL ISOTOf'fS Table 13-15. Adverse Event Summary for Ventilation System and Identification of Accident Sequences Needing Further Evaluation PHA item numbers Bounding accident description Consequence Accident sequence 7.1.1.7 and Too much uranium accumulated Accidental nuclear criticality S.C.24, High uranium 7.1.1.8 on the HEP A filter allows an leads to high dose to workers content on HEPA filters accidental criticality when left in and potential dose to the th e wrong configuration publi c 7.1.1.2, 7.1.1.3, Hydrogen buildup in the A detonation or deflagration S.F.06, Accumulation of and 7.1.1 .6 ventilation system, due to event in the ventilation flammable gas in ventilation insufficient flow to sweep it system rapidly releases system components away, leads to fire in the HEPA retained high-dose filters or carbon beds radionuclides, causing high airborne radioactivity 7.1.1.1 0 and Ignition source causes fire in the Fire event in the ventil ation S.F.05, Fire in the carbon 7.2.1.19 carbon bed system rapidly releases bed retained high-dose radionuclides, causing high airborne radioactivity 7.1.1.11 and Overloading of HEPA filter leads High dose to workers from S.R.25, HEPA filter failure 7.2.1 .20 to failure and release of high airborne radioactivity accumulated radionuclide particulate 7.1.1.12, 7.1.1.14, The accumulated high-dose (and High dose to workers from S.R.04, Carbon bed and 7.2.1.21 low-dose) radionuclides retained high airborne radioactivity radi onuclide retention failure in the carbon bed are released through a flow, heat, or chemical reaction from the media (or the media is released) 7.2.1.4, 7.2.1.7, Loss of the negative air balance High dose to workers from S.R.26, Failed negative air 7.2.1.8, 7.2.1.9, between zones (a confinement high airborne radioactivity balance from zone to zone or 7.2.1.13, 7.2.1.14, feature that prevents migration of failure to exhaust a 7.2.1.17, and radionuclides from areas of high radionuclide buildup in an 7.2.1.22 dose and high concentration to area areas oflow concentration) 7.2.1.12 and During an extended power High dose to workers from S.R.27, Extended outage of 7.2.1.1 7 outage, some soluti on systems high airborne radioactivity heat, leading to freezing, freeze and cause failure of the pipe failure, and release of piping system, leading to radionuclides from liquid radiological spills process systems HEPA high-efficiency particulate air. PHA process hazards analysis.
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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 8.2.1.5 Large leak leads to localized low Standard industrial hazard - Localized Nitrogen storage or oxygen levels that adversely asphyxiant distribution system leak impact worker performance and may lead to death 8.5.1.1 and Operator double-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 Accidental criticality issue - Too much S.C.03, Failure of AC 8.5.1.5 containers of uranium or batches uranium mass in one location on interaction limit of uranium components brings during handling of two containers or batches closer fresh U, scrap U, together than the approved [Proprietary interaction control distance Information], and targets 8.6.1.7 A liquid spill ofrecycle uranium Criticality issue - Fissi le solution may S.C.04, A liquid spill or target di ssolution solution collect in unsafe geometry of fissile solution occurs within the hot cell occurs boundary 8.6.1.9 Process solutions backflow Criticality issue - Fissi le solution may S.C.08, Fissi le process through chemical addition lines to collect in unsafe geometry solutions backflow locations outside the hot cell through chemical boundary addition lines 8.6.1.13 Improper instalJation of HEPA Accidental nuclear criticality leads to S.C.24, High uranium fi lters (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 enriched Criticality hazard - Too much uranium S.C.27, Fai lure of AC 8.5.1 .5 solutions pours solution into an mass in one place can lead to accidental on volume limit during unapproved container nuclear criti cality sampling 8.4.1.8 and Drop of a hot cell cover block or Criticality issue - 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
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 8.1.2.7 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, resulting in fire of the facility, wildfire, chemical, radiological, or criticality combustible fire in non-industrial hazards 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 systems 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 criticality leads 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 personnel action) personnel action) operation 13-34
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. 1.2.13 Flooding from external events and Criticality issue - Water accumulation S.M.03, Flooding internal events compromises the under safe geometry storage vessels or occurs in building due safe geometry slab area under in safe interaction storage arrays, to internal system leak certain tanks. Depending on the causing interspersed moderation . or fire suppression liquid level, interspersed Flooding could compromise safe- system activation moderation of components may geometry storage capacity for (likely) be impacted. Floor storage arrays subsequent spills of fissile solution.
are subject to stored containers Either event could compromise floating (loss of interaction control). criticality 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 building).
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 adversely damage SSCs relied on for facility and SSCs safety. Facility could lose all electrical distribution. Facility could lose chi ll ed water system function (cooling tower outside of building).
8.1.1.3 A 48-hr probable maximum Radiological, chemical, 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 issue - Minor structural S.N.04, Flooding excess of 500-year return damage is not anticipated to impact impact on facility and frequency SSCs relied on for safety except that the SS Cs facility could lose 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 lose all electrical damage impacts SSCs inside and distribution. Facility could 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.
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- ~*:~:* NWMI NWMl-2013-021, Rev . 2
- *~~~~*
- NOtmfWUTM£0tcAllSOTOP£S 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, Heavy snowfall or ice buildup Radiological, chemical, and criticality S.N.06, Heavy
- 8. 1.1.10 exceeds design loading of the issue - Structural damage from roof snowfall or ice buildup roof, resulting in collapse of th e collapse could adversel y damage SSCs on faci lity and SSCs roof and damage to SSCs (e.g., reli ed on for safety. Loss of site those outside of the hot cell s) electrical power is highly likely in heavy ice storm event.
8.6.1.8 Any stored high-dose product Radiological issue - High-dose solution S.R.01 , A liquid spill solution spills within the hot cell is unconfined or uncontrolled and can of high-dose fission boundary cause exposures to workers, the public, product solution occurs and environment
- 8. 5.1 .5 Operator spills diluted sample Radiological issue - Potential spray or S.R.01 , Spill of product outside of th e hot cell area vaporization of radionuclide containing solution in laboratory vapor-causing adverse worker exposure (based on typi cal low quantiti es handled in the laboratory, thi s is postulated to be an intermedi ate consequence event) 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 cell areas, 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 Process solutions backflow Radiological issue - High radiation may S.R.16, High-dose through chemical addition lines to occur in non-hot cell areas, impacting process solutions locati ons outside the hot cell workers with higher than normal backflow through boundary external doses chemical 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, some streaming of hot cell wall transfers) offer large opening high radiation may occur, impacting penetration, potentials for radiation streaming workers with higher than normal compromising external doses shielding 8.6. 1.1 The seal on the bagless transport Radiological issue - Degraded or loss of S.R. 24, Bagless door fails and leads to high-dose cascading negative air pressure between tran sport door failure radionuclides escaping th e hot zones may allow high radiological cell confi nement boundary airborne contamination to release with out proper filtration and adsorpti on ,
leading to higher than allowed exposure rates to workers and the public 8.6.1.13 Following process upsets and Radiological and criticality issue - 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 13-36
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~* * ~ NOfliTHWHT MEDICAL ISOTOPfS 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 8.6.1.2, An improperly sealed cover block 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 zones 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 workers 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 administrative control. SSC structures, systems, and components.
HEPA high efficiency particulate air. TBD to be determined.
IROFS items relied on for safety. u uranium.
PHA process hazards analys is. UN uranyl 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, Integrated Safety Analysis Plan for the Radioisotope Production Facility.
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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 the 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 recycle, 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 quantitative 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 Operating License Application.
Sections 13.2.2 through 13.2.5 present key representative sequences for radiological and criticality accidents.
Section 13.2.2 discusses spills and spray accidents with both radiological and criticality safety consequences Section 13.2.3 discusses dissolver offgas accidents with radiological consequences Section 13.2.4 discusses leaks into auxiliary system accidents with both radiological and criticality safety consequences
- Section 13 .2.5 discusses loss of electrical power These accidents cover fa ilure of primary vessels and piping in the processing areas, loss of fission product gas removal efficiency, leaks into auxiliary systems, and loss of power to the RPF.
Section 13.2.6 briefly presents evaluations of natural phenomena events. The stringent design 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 unlikely per Table 13-1 .
The remainder of the accident sequences, identifi ed in the PHA as requiring further evaluation, are summarized in Section 13 .2. 7. Each sequence is identified and the associated IROFS (if any) listed. The IROFS not discussed 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 the preceding 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 this point in the RPF project development. The additional required information will be provided in the Operating License Application.
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...... NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis
~* * ~ NORTHWEST MEDICAL lSOTOP£S 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 systems 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 the process offgas ventilation system operating. A level monitoring system is available to monitor tank transfers and stagnant storage volumes on all tanks processing LEU or fission product solutions.
Bounding radionuclide concentrations in liquid streams were developed for five regions of the process in NWMJ-2013-CALC-Ol l , Source 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 quid 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 as 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
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...... NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis
~* * ~ frlO<<THW£ST MUMCAl tSOTOl'U 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, Overall Summary Material Balance - 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 has 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 by 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 separation 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 1Am]) have increased due to parent decay.
Table 13-17. Bounding Radionuclide Liq uid Stream Concentrations (4 pages)
Unit operation Target dissolution Uranium recovery and recycle Decay, hours after EOI [Proprietary Information] [Proprietary Information]
Stream description Dissolver product Uran iwn separation feed Isotope Bounding concentration (Ci/L) Bounding concentration (Ci/L) 24 1Am [Proprietary Information] [Proprietary Information]
136mBa [Proprietary Information] [Proprietary Information]
137mBa [Proprietary Information] [Proprietary Information]
139Ba [Proprietary Information] [Proprietary Information]
140Ba [Proprietary Information] [Proprietary Information]
141ce [Proprietary Information] [Proprietary Information]
143Ce [Proprietary Information] [Proprietary Information]
144Ce [Proprietary Information] [Proprietary Information]
242Cm [Proprietary Information] [Proprietary Information]
z43Cm [Proprietary Information] [Proprietary Information]
244Cm [Proprietary Information] [Proprietary Information]
134Cs [Proprietary Information] [Proprietary Information]
I34mCs [Proprietary Information] [Proprietary Information]
136Cs [Proprietary Information] [Proprietary Information]
137 Cs [Proprietary Information] [Proprietary Information]
1ssEu [Proprietary Information] [Proprietary Information]
1s6Eu [Proprietary Information] [Proprietary Information]
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...... NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis
' ~* *~
- NOmtWtST MEO.CAL lSOTDPfS Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages)
Unit operation Target dissolution ' Uranium recovery and recycle Decay, hours after EOI [Proprietary Information] [Propri etary Information]
Stream description Dissolver product Uranium separation feed Isotope Bounding concentration (Cill) Bounding concentration (Ci/L) 1s1Eu [Proprietary Information] [Proprietary Information]
1291 [Proprietary Information] [Proprietary Information]
1301 [Proprietary Information] [Proprietary Information]
1311 [Proprietary Information] [Proprietary Information]
132J [Proprietary Information] [Proprietary Information]
132mI [Proprietary Information] [Proprietary Information]
1331 [Proprietary Information] [Proprietary Information]
133mI [Proprietary Information] [Proprietary Information]
134J [Proprietary Information] [Proprietary Information]
13SI [Proprietary Information] [Proprietary Information]
83mKr [Proprietary Information] [Proprietary Information]
85Kr [Proprietary Information] [Proprietary In formation]
8Sm Kr [Proprietary Information] [Proprietary Information]
87Kr [Proprietary Information] [Proprietary Information]
88Kr [Proprietary Information] [Proprietary Information]
140La [Proprietary Information] [Proprietary Information]
141La [Proprietary Information] [Proprietary Information]
142La [Proprietary Information] [Proprietary Information]
99Mo [Proprietary Information] [Proprietary Information]
95Nb [Proprietary lnfonnation] [Proprietary Information]
95mNb [Proprietary Information] [Proprietary Information]
96Nb [Proprietary Information] [Proprietary Information]
97Nb [Proprietary Information] [Proprietary Information]
97mNb [Proprietary Information] [Proprietary Information]
141Nd [Proprietary Information] [Proprietary Information]
236mNp [Proprietary Information] [Proprietary Information]
231Np [Proprietary Information] [Proprietary Information]
23gNp [Proprietary Information] [Proprietary Information]
239Np [Proprietary Information] [Proprietary Information]
233pa [Proprietary Information] [Proprietary Information]
234Pa [Proprietary Information] [Proprietary Information]
234m Pa [Proprietary Information] [Proprietary Information]
112pd [Proprietary Information] [Proprietary Information]
147Pm [Proprietary Information] [Proprietary Information]
148Pm [Proprietary Information] [Proprietary Information]
148mpm [Proprietary Information] [Proprietary Information]
13-41
NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages)
Unit operation Target dissolution Uranium recovery and recycle Decay, hours after EOI [Proprietary Information] [Proprietary Information]
Stream description Dissolver product Uranium separation feed Isotope Bounding concentration (Ci/L) Bounding concentration (Ci/L) 149Pm [Proprietary Information] [Proprietary Information]
ISOPm [Proprietary Information] [Proprietary Information]
ISIPm [Proprietary Information] [Proprietary Information]
142Pr [Proprietary Information] [Proprietary Information]
143pr [Proprietary Information] [Proprietary Information]
144Pr [Proprietary Inform ati on] [Proprietary Information]
J44mpr [Proprietary Information] [Proprietary Information]
145Pr [Proprietary Information] [Proprietary Information]
23 8pu [Proprietary Information] [Proprietary Information]
239pu [Proprietary Information] [Proprietary Information]
240pu [Proprietary Information] [Proprietary Information]
24 1Pu [Proprietary Information] [Proprietary Information]
103mRh [Proprietary Information] [Proprietary Information]
IOSRh [Proprietary Information] [Proprietary Information]
106Rh [Proprietary Information] [Proprietary Information]
106mRh [Proprietary Information] [Proprietary Information]
103Ru [Proprietary Information] [Proprietary Information]
1osRu [Proprietary Information] [Proprietary Information]
106Ru [Proprietary Information] [Proprietary Information]
122 Sb [Proprietary Information] [Proprietary Information]
124 Sb [Proprietary Information] [Proprietary Information]
125 Sb [Proprietary Information] [Proprietary Information]
126Sb [Proprietary Information] [Proprietary Information]
121sb [Proprietary Information] [Proprietary Information]
12ssb [Proprietary Information] [Proprietary Information]
12smsb [Proprietary Information] [Proprietary Information]
129Sb [Proprietary Information] [Proprietary Information]
1s1 sm [Proprietary Information] [Proprietary Information]
153 Sm [Proprietary Information] [Proprietary Information]
156 [Proprietary Information]
Sm [Proprietary Information]
s9sr [Proprietary Information] [Proprietary Information]
9osr [Proprietary Information] [Proprietary Information]
91sr [Proprietary Information] [Proprietary Information]
92 Sr [Proprietary Information] [Proprietary Information]
99Tc [Proprietary Information] [Proprietary Information]
99mrc [Proprietary Information] [Proprietary Information]
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- ~~!~*
- NORTKWtU MEDtcAL ISOTOPES Table 13-17. Bounding Radionuclide Liq uid Stream Concentrations (4 pages)
Unit operation Target dissolution Uranium recovery and recycle Decay, hours after EOI [Proprietary Information] [Proprietary Information]
Stream description Dissolver product Uranium separation feed Isotope Bounding concentration (Ci/L) Bounding concentration (Ci/L) 12smre [Proprietary Information] [Proprietary Information]
127 Te [Proprietary Information] [Proprietary Information]
127mre [Proprietary Information] [Proprietary Information]
129Te [Proprietary Information] [Proprietary Information]
129mre [Proprietary lnformati on] [Proprietary Information]
n1re [Proprietary Information] [Proprietary Information]
131mre [Proprietary Information] [Proprietary Information]
132Te [Proprietary Information] [Proprietary Information]
133 Te [Proprietary Information] [Proprietary Information]
133mre [Proprietary Information] [Proprietary Information]
134Te [Proprietary Information] [Proprietary Information]
23 1Th [Proprietary Information] [Proprietary Information]
234Th [Proprietary Information] [Proprietary Information]
232u [Proprietary Information] [Proprietary Information]
234u [Proprietary Information] [Proprietary Information]
23su [Proprietary Information] [Proprietary Information]
236u [Proprietary Information] [Proprietary Information]
237u [Proprietary Information] [Proprietary Information]
23su [Proprietary Information] [Proprietary Information]
131mxe [Proprietary Information] [Proprietary Information]
133 Xe [Proprietary Information] [Proprietary Information]
133mxe [Proprietary lnformati on] [Proprietary Information]
135 Xe [Proprietary Information] [Proprietary Information]
I3smxe [Proprietary Information] [Proprietary Information]
89my [Proprietary Information] [Proprietary Information]
90y [Proprietary Information] [Proprietary Information]
90my [Proprietary Information] [Proprietary Information]
9Iy [Proprietary Information] [Proprietary Information]
9Jmy [Proprietary Information] [Proprietary Information]
92y [Proprietary Information] [Proprietary Information]
93y [Proprietary Information] [Proprietary Information]
93Zr [Proprietary Information] [Proprietary Information]
9szr [Proprietary Information] [Proprietary Information]
97Zr [Proprietary Information] [Proprietary Information]
Totals [Proprietary Information] [Proprietary Information]
Source: Table 2-1 ofNWMI-2013-CALC-O 11 , Source Term Calculations, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, February 2015.
EOI = end of irradiation.
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- ~;:::* NWMI NWMl-2013-021, Rev. 2
- *
- NORTHWUT MEDtCAl ISOTDr£S Chapter 13.0 -Accident Analysis 13.2.2.2 Identification of Event Initiating Conditions 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 le 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 as follows.
- 1. The process line, containing pressurized liquid, ruptures or develops a leak during a transfer, spraying solution into the source 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 sump 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 stable condition.
13.2.2.4 Function of Components or Barriers The process vessel enclosure floor , walls, and ceiling will provide 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 secondary confinement system is to function throughout the accident until a stable 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-contingency analysis) for preventing the occurrence of a solution-type accidental nuclear criticality due to spills or sprays of fissile material. The sump system is to function throughout the accident until a stable condition has been achieved.
13.2.2.5 Unmitigated Likelihood A spill or spray 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 River 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 unlikely."
Additional detailed information describing a quantitative evaluation, including assumptions, methodology, uncertainties, and other data, will be developed for the Operating License Application.
13.2.2.6 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.
13.2.2.6.1 Direct Expos ure Source Terms Liquid spill source terms are dependent on the vessel location in the process 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, shows that a pencil tank of this material would be less than 1 millirem (mrem)/hr; therefore, no radiological IROFS are required for this stream.
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 uranyl 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 small solution volume (less than 1 L) containing the weekly inventory of product from processing MURR targets. The product is an aqueous solution containing - 0.2 M sodium hydroxide (NaOH) with a total inventory of 1.3 x10 4 curies (Ci) 99 Mo.
13.2.2.6.2 Confinement Release Source Terms Confinement release source terms are based on the five-factor algebraic formula for calculating source terms for airborne release accidents from NUREG/CR-6410, as shown by Equation 13-1 .
ST= MARxDRxARFxRFxLPF Equation 13-1
- where, ST Source term (activity)
MAR Material at risk (activity) 13-45
NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis DR Damage ratio (dimensionless)
ARF Airborne release fraction (dimensionless)
RF Respirable fraction (dimensionless)
LPF Leak path factor (dimensionless)
Confinement release source terms for spray used the source term parameters listed in Table 13-18. Four source term cases were developed for evaluation based on the two bounding liquid concentrations shown in Table 13-17 and the source term parameter alternatives.
Table 13-18. Source Term Parameters Parameter 3 Unmitigated spray release Mitigated spray release Material at risk (MAR) Table 13-17 Table 13-17 Damage ratio (DR) 1.0 1.0 Airborne release fraction (ARF) 0.0001 0.0001 (I .0 for Kr, Xe, and iodine)b ( 1.0 for Kr, Xe, and iodine)h Respirable fraction (RF) 1.0 1.0 Leak path factor (LPF) 1.0 0.0005 (1.0 for Kr, Xe; 0.1 for iodine)
Source: Table 2- 1 ofNWMI-2015-RPT-009, Fission Product Release Evaluation, Rev. B, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 20 15.
a Parameter definitions derived from NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, U.S. Nuclear Regulatory Commiss ion, Office of Nuclear Material Safety and Safeguards, Washington, D.C., March 1998.
b Accident dose conseq uences were found to be sensitive to iodine source term parameters. Further work may allow for a lower iodine ARF.
The DR was set to 1.0 for all cases. The assumed volume was 100 L of solution contained by a vessel being affected by the spill or spray release.
The ARF and RF values are functions of the release mechanism and do not enter 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 Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities. 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 level. Both NUREG/CR-6410 and DOE-HDBK-3010 report an ARF of 1x10-4 and a RF of 1.0 for a spray leak involving 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 mechanism is also bounded by a droplet size distribution produced from commercial spray nozzles. This approach is conservative, as the effective nozzle created by a pipe failure is unlikely to be optimized to the extent of a manufactured spray 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 scenario credits no confinement measures (i.e., no credit was taken for any aspect of the facility design or equipment performance). The gravitational settling associated with flow throughout the faci lity and the removal action of high-efficiency particulate air (HEPA) filtration may be lumped into an effective value for LPF.
The performance of different filtration systems is presented in Appendix F ofDOE-HDBK-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.
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~* * ~ 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 removal 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 slightly 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 99Mo. 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 pre!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 x103 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 99Mo 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 Review of Criticality Accidents, identifies that a uranium solution criticality can yield between 10 16 to 10 17 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 inputs 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 Description Primary tool STDose - Source term to dose option selected as the primary tool in RASCAL for all cases.
Event type 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.
Facility locationa Columbia, Missouri County Boone Time zone Central Latitude/longitude 38.9520° N/92.3290° W Elevation 231 m Plume rise None - For scoping purposes, the enthalpy and momentum of the RPF stack exhaust was assumed negligible.
Meteorology Summer-night-calm - Selected for scoping purposes and features wind speed of 6.4 km/hr (4 mi/hr), Pasquill Class F stability, no precipitation, relative humidity of 80%, and ambient temperature of l 2.8°C (55 °F). Low wind speed and stable conditions selected to provide maximum dose to near-field receptors.
Receptor distance 100 m - Selected to approximate site boundary. Input represents minimum value for RASCAL input.
Dose conversion factors ICRP- 72b - Selected as the most current and authoritative set of dose conversion factors available.
Source: Table 2- 1 ofNWMl-20 l 5-RPT-009, Fission Product Release Evaluation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, February 20 15.
a Location information obtained from Wikipedia.
b ICRP-72, Age-Dependent Doses to the Members of the Public from Intake of Radionuclides - Part 5 Compilation of Ingestion and Inhalation Coefficients , International Commi ssion on Radi ological Protecti on, Ottawa, Canada, 1995 .
RASCAL = Radiological Assessment System for RPF = Radioi sotope Production Facility.
Consequence 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 density: 1,240 g/cubic meter [m 3] (1.24 ounce [oz] /cubic feet [ft3]) (sea level)
Pasquill-Gifford a (NRC Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power 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 Intakes 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 NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis
- * ~~~~*
- NOM'HW£ST MEDICAL ISOTOKt Consequence evaluation results are shown in Figure 13-2 and Table 13-20 for a 100 L (26.4 gal) spray release event. NWMI is considering the unmitigated spray release of dissolver product solution as an off-site public intermediate consequence event (pending completion of the final safety analysis). The nearest permanent resident, at 432 m (0.27 mil es [mi]), dissolver product spray unmitigated dose estimate is 300 mrem, while the maximum receptor location (1 , 100 m [0.68 mi]) has a total effective dose equivalent (TEDE) of 1.8 rem. The mitigated consequences are an order of magnitude lower due to the credited IROFS in the Zone I exhaust system. Therefore, the nearest permanent res ident (432 m
[0.27 mi]) dissolver product spray mitigated dose estimate is 30 mrem, whil e the maximum receptor location (1, I 00 m [0.68 mi]) has a TEDE of 0.18 rem.
2.0 1.8 1.6 1.4 E 1.2 Q) l-e) 1.0 _._ Inhalation CEDE t/l 0
0 0.8
_._ External EDE 0.6 0.4 0.2 100 200 300 400 500 600 700 800 900 1000110012001300140015001600 Distance, meters Figure 13-2. Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident Table 13-20 shows that the uranium separation feed solution spray release unmitigated dose is below the immediate consequences thresholds of I 0 CFR 70.61. Even though this receptor dose is at 100 m, the uranium feed modeling will be rerun using RS AC 6.2 as part of the Operation License Application.
Table 13-20. Uranium Separations Feed Spray Release Consequence Summary at 100 Meters Process stream Uranium separations feed Case Mitigation Unmitigated Mitigated Receptor dose, total EDE 0.078 rem 0.006 rem Stack height I 0 m (33 ft)" 23 m (75 ft)
Release mechanism Spray leak, 100 L Release duration 1 hr Source: Table 2- 1 and Table 2-7 ofNWMI-20 15-RPT-009, Fission Product Release Evaluation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, February 20 15.
a Lowest value for plume height avai lable as input to RASCAL and recommended by help tile as input modeling a ground-level release.
EDE = effective dose equi va lent. RAS CAL = Radi ological Assessment System for Consequence Anal ys is.
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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 spray releases 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 spray-type release also carries the extra hazard of potential chemical burns to eyes and skin, with the complication of radiological contamination. Consequently, 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 levels for workers and the public during the event.
Three IROFS are identified to control liquid spill and spray accidents from process vessels.
- IROFS RS-04, "Hot Cell Shielding Boundary" Liquid spill and spray events involving solutions containing fissile material have the potential for producing liquid nuclear criticalities that must be prevented. The following 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, "Double-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 solutions 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 processing 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.
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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 secondary function of the liquid confinement boundary is to prevent contact chemical exposure to workers fro m acidic or caustic sol utions contaminated with licensed material that exceeds the performance criteria established by NWMI for the RPF.
As a PEC to contain spills and sprays of high-dose product solution, the hot cell liquid confi nement boundary will consist of sealed flooring with multiple layers of protection from release to the environment. Various areas will be diked to contain specific releases, and sumps of appropriate design will be provided with remote-operated pumps to mitigate liquid spills by capturing the liquid in appropriate safe-geometry tanks. Additional IROFS apply to the flooring and sumps for criticality safety double-contingency controls in some areas. In the 99 Mo purification product and sample hot cell, smaller confinement catch basins will be provided under points of credible spill potential in addition to use of a sealed floor. Entryway doors into a designated liquid confinement area will be sealed against credible liquid leaks to outside the boundary. This continuous barrier is also credited to prevent spills or sprays of hi gh-dose product solutions that are acidic or caustic from causing adverse exposure to personnel through direct contact with skin, eyes, and mucus membranes, where the combination of the chemical exposure and the radiological contamination would lead to serious injury and long-lasting effects or even death.
Specific design features of the liquid confinement barrier, a liquid barrier to uncontrolled areas and worker radiation exposure from leaked solution, include:
- Continuous, impervious floor with an acid- or caustic-resistant surface finish
- Hot cell walls and ceiling designed to control worker dose from liquids accumulated in sumps
- Monitors with alarms to indicate a liquid release has occurred
- Sealed penetrations designed to prevent liquid leaks through th e barrier to uncontrolled areas
- Sump solution collection vessels for accumulating leaked process solution 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 vessels in the hot cells. As a system of PECs and AECs, the hot cell secondary confinement boundary safety feature is engineered to provide backup to credible upsets in the primary confinement system using the following safety functions:
Provide negative air pressure in the hot cell (Zone I) relative to lower zones outside the hot cell using exhaust fans equipped with HEP A filters and HEGA modul es to remove the release of radionuclides (both particulate and gaseous) to outside the primary confinement boundary to below 10 CFR 20 release limits during normal and abnormal operations.
Components credited include:
Zone I Inlet HEPA filters to provide an efficiency of 99 .97 percent for removal of radiological particulates from the air that may reverse flow from Zone I to Zone 11 Zone I ducting to ensure that negative air pressure can be maintained by conveying exhaust air to the stack Zone I exhaust train HEPA filters to provide 99.97 percent removal of radiological particulates from the air that flows to the stack Zone I exhaust train HEGA modules to provide 90 percent removal of iodine gas from the air that flows to the stack Zone I exhaust stack to provide dispersion of radionuclides in normal and abnormal releases at a discharge point of 22.9 m (75 ft) above the building ground level 13-51
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~* * ~ 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-tight, 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 appropriately sized ventilation 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 offgas 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 99Mo 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.
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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 boundary 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, steel-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 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 reduce the exposure from postulated upsets within the hot cell shielding boundary to less 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 operable when the equipment in that area is in the operating or standby modes .
13.2.2.8.4 IROFS CS-07, Pencil Tank and Vessel 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 prevent a nuclear criticality event. As a PEC, pencil tanks and other standalone vessels (controlled with safe geometry or volume 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 spacing 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 process upsets, the systems will remain subcritical. The fixed interaction control of tanks, vessels, or components containing fissile solutions will prevent accidental nuclear criticality, a high consequence event. The fixed interaction control distance from the safe slab depth spill containment berm is specified 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 by geometry if filled with liquid from a spill 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 safe-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 safety function of this spill containment berm is to safely contain spilled fissile solution from systems overhead and prevent an accidental nuclear criticality 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 have a monitoring system to alert the operator that the IROFS has been used and may not be available for a follow-on event. A spill containment berm will be operable if it contains reserve volume for the largest single credible spill. Spill containment berm sizes and locations will be determined by the final design.
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...... NWMl-2013-021, Rev . 2 Chapter 13.0 -Accident Analysis
' !e* ~ ' NORTHWtn MEDICAL ISOTOPU 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 double-wall barrier to contain any spills that may occur from the primary confinement piping. IROFS CS -09 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 safe-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 alert 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 [NOx] scrubber) or equipment fai lure (e.g., loss of the IRU heater) during the dissolution of irradiated targets. The primary components of the dissolver offgas include:
NOx scrubbers (caustic and absorbers)
IR Us Pressure-relief vessel Primary adsorbers (carbon media beds for 6 days noble gas holdup) 13-54
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...... NWMl-2013-021 , Rev. 2 Chapter 13.0 -Accident Analysis
~* * ~ NOfITTfWHT MfDtCAL ISOTOP£S Iodine guard beds (remove any iodine not trapped in the IRUs)
Filters Vacuum receiver tanks Vacuum pumps (draw a downstream vacuum on the target dissolver offgas treatment train)
Secondary adsorbers (additional carbon media beds to hold up noble gases for an additional 60 days)
The IR Us nominall y removes about 99.9 percent of the iodine in the offgas stream after the NOx scrubbers. NWMJ expects the availability and operation ofIRUs will become part of the technical specification to meet annual release limits. The iodine released from dissolution of the irradiated targets 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 NOx treatment absorbers) and end up in the high dose liquid waste tanks, and (3) the remainder of the iodine will be captured in the IRUs.
These 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 ed 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 adsorbers 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 offgas Table 13-21. Maximum Bounding Inventory of treatment train are assumed to be operational Radio iodine [Proprietary Information]
and in service prior to the occurrence of any Isotope accident sequence that affects the IR Us.
The JR Us are assumed to be loaded with the [Proprietary Information]
conservative bounding holdup inventory of [Proprietary Information]
iodine, as determined in NWMI-2013-CALC- [Proprietary Information]
01 I. [Proprietary In form ati on]
1J 2m No credible event has been identified where 1 [Proprietary lnformati on]
the total captured inventory on the IR Us would [Proprietary Information]
be released. This accident evaluation is for the 133mJ [Proprietary Information]
release of the iodine generated from a single
[Proprietary Information]
dissolution of [Proprietary Information]. The maximum amount of iodine [Proprietary [Proprietary Information]
Information] is shown in Table 13-21. The Total I Ci [Proprietary Information]
mass balance projects about 20 percent of the = iodine.
iodine will stay in the dissolver solution and nearly 50 percent of the elemental iodine (h) that does volatize will be captured in the NOx 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 offgas stream going to the IR Us.
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- ~* * ~ NORTHWEST MEDK:Al ISOTIN'ES Chapter 13.0 - Accident Analysis Therefore, this evaluation focuses on accident sequences where the inventory at risk is that generated directly 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 offgas treatment train. The three most likely 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 NOx 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 sequences for loss ofIRU efficiency include the following.
[Proprietary Information] is being dissolved.
A process upset occurs that reduces the IRU efficiency by an unspecified amount.
The event is 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 opens the valve to the pressure-relief tank in the dissolver offgas system to capture the offgas stream.
If the initiator for the event is loss of power or the event creates a condition where vacuum in the dissolver offgas system is lost, the pressure-relief tank valve would automatically open to capture the offgas 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 IRUs will be the primary iodine capture devices; however, there will be iodine guard beds downstream of each of the primary noble gas adsorbers. The vent system piping will direct the dissolver offgas 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 offgas 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 IRUs. 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."
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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 iodine radioisotopes. Therefore, the iodine source term is the focus of this accident sequence evaluation.
No credit is taken for any iodine removal in the 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 that were designed to bound the credible offgas scenarios. These assumptions include:
[Proprietary Information]
All the iodine in the targets released into the offgas system, and no iodine or noble gases captured in the NOx scrubbers or retained in the dissolver solution Iodine removal efficiency of the dissol ver offgas 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 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 will 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.
13.2.3. 7 Evaluation of Potential Radiological Consequences Radiological 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 performance of the target dissolver offgas treatment train IRUs. Additional detailed information describing validation of the models, 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 inventory would 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 llowing parameters were used for model runs:
Mixing depth: 400 m (1 ,312 ft) (default)
Air density: 1,240 g/m 3 (1.24 ozlft 3) (sea level)
Pasquill-Gifford o (NRC Regulatory Guide 1.145)
No plume rise (i.e., buoyancy or stack momentum effects) 13-57
..::*...*. NWMI NWM l-2013-021, Rev. 2 Chapter 13.0 - Acci dent Analysis
~* * ~ NOffTlfW'En MEDICAi. lSOTOPfS No plume depletion (wet or dry Table 13-22. Target Dissolver Offgas Accident deposition) Total Effective Dose Equivalent 2-hr release (constant release of all TEDE (rem) activity) 2-hr exposure Distance (m) ' Total ICRP-30 inhalation model 100 2.05E-01 Finite cloud immersion model 200 l.98E-OI Breathing rate: 3.42E-4 m3/sec 300 2.21E-01 (l .2E-2 ft 3/sec) (ICRP-30 heavy activity) 400 6.41 E-OI Respiratory fraction: 1.0 500 l.76E+OO Table 13-22 shows the distance-dependent total 600 3.18E+OO receptor accident doses versus distance from the 700 4.50E+OO RPF stack for 2-hr exposure. This table was 800 5.47E+OO developed using the results from the Section 19.4 1,000 6.50E+OO dose consequences and dividing by a ratio of the 1,100 6.65E+OO accident source terms. The maximum public dose 1,200 6.62E+OO is 6.65 rem at 1, 100 m.
1,300 6.50E+OO RSAC 6.2 calculates inhalation doses using the 1,400 6.29E+OO ICRP-30 model with Federal Guidance Report 1,500 6.06E+OO No. 11 dose conversion factors 1,600 5.82E+OO (EPA 520/1-88-020, Limiting Values of 1,700 2.05E-OI Radionuclide Intake and Air Concentration and Peak total dose is balded and italicized.
Dose Conversion Factors for Inhalation, Submersion, and Ingestion) . The committed dose TEDE = total effective dose equivalent.
equivalent (CDE) is calculated for individual organs and tissues over a 50-year period after inhalation.
The CDE for each organ or tissue is multiplied by the appropriate ICRP-26, Recommendations of the International Commission on Radiological Protection, weighting factor and then summed to calculate 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 (~6.65 rem) are within intermediate consequences severity categories (<25 rem).
13.2.3.8 Identification of Items Relied on 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 offgas 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.
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~ * *!' NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis IROFS RS-09, Primary Offgas Relief System As an AEC, a relief device will be provided that relieves pressure 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 The following defense-in-depth features preventing target dissolver 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 primary 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 controls 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 leaks into secondary containment as an event that could lead to an accidental nuclear criticality. The accidents covered by this analysis bound the family of accidents where highly radioactive or fissile solution leaves the hot cell or other shielded areas via auxiliary 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 primary 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 needs to be condensed. The bounding source term is expected to be the dissolvers or the feed tanks in the Mo recovery and purification system.
Table 13-23 lists 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.
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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]
139Ba [Proprietary Information] [Proprietary Information]
i4oBa [Proprietary Information] [Proprietary Information]
141ce [Proprietary Information] [Proprietary Information]
143Ce [Proprietary lnformation] [Proprietary Information]
144Ce [Proprietary Information] [Proprietary Information]
242cm [Proprietary Information] [Proprietary Information]
243Cm [Proprietary Information] [Proprietary Information]
244Cm [Proprietary Information] [Proprietary Information]
134Cs [Proprietary Information] [Proprietary Information]
134m Cs [Proprietary Information] [Proprietary Information]
136Cs [Proprietary Information] [Proprietary Information]
137 [Proprietary Information]
Cs [Proprietary Information]
1ssEu [Proprietary Information] [Proprietary Information]
1s6Eu [Proprietary Information] [Proprietary Information]
1s1Eu [Proprietary Information] [Proprietary Information]
1291 [Proprietary Information] [Proprietary Information]
130J [Proprietary Information] [Proprietary Information]
13 IJ [Proprietary Information] [Proprietary Information]
1321 [Proprietary Information] [Proprietary Information]
132m I [Proprietary Information] [Proprietary Information]
1331 [Proprietary Information] [Proprietary Information]
133mI [Proprietary Information] [Proprietary Information]
1341 [Proprietary Information] [Proprietary Information]
135J [Propri etary Information] [Proprietary Information]
83m Kr [Proprietary Information] [Proprietary Information]
85Kr [Proprietary Information] [Proprietary Information]
85m Kr [Proprietary Information] [Proprietary Information]
87Kr [Proprietary Information] [Proprietary Information]
88Kr [Proprietary Information] [Proprietary Information]
140La [Proprietary Information] [Proprietary Information]
141La [Proprietary Information] [Proprietary Information]
142La [Proprietary Information] [Proprietary Information]
99Mo [Proprietary Information] [Proprietary Information]
95Nb [Proprietary Information] [Proprietary Information]
95mNb [Proprietary Information] [Proprietary Information]
96Nb [Proprietary Information] [Proprietary Information]
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NWMl-2013-021 , Rev. 2 Chapter 13.0 - Acci dent 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 97Nb [Proprietary Information] [Proprietary Information]
97mNb [Proprietary Information] [Proprietary Information]
141Nd [Proprietary Information] [Proprietary Information]
236mNp [Proprietary Information] [Proprietary Information]
231Np [Proprietary Information] [Proprietary Information]
23sNp [Proprietary Information] [Proprietary Information]
239Np [Proprietary Information] [Proprietary Information]
233Pa [Proprietary Information] [Proprietary Information]
234pa [Proprietary Information] [Proprietary Information]
234mPa [Proprietary Information] [Proprietary Information]
11 2pd [Proprietary Information] [Proprietary Information]
141Pm [Proprietary Information] [Proprietary In formation]
148Pm [Proprietary Information] [Proprietary Information]
I48mpm [Proprietary Information] [Proprietary Information]
149Pm [Proprietary Information] [Proprietary Information]
ISOpm [Proprietary Information] [Proprietary Information]
ISIPm [Proprietary Information] [Proprietary Information]
142Pr [Proprietary Information] [Proprietary Information]
143Pr [Proprietary Information] [Proprietary Information]
I44pr [Proprietary Information] [Proprietary Information]
144mpr [Proprietary Information] [Proprietary Information]
I4Spr [Proprietary In formation] [Proprietary Information]
2Jspu [Proprietary Information] [Proprietary Information]
239Pu [Proprietary Information] [Proprietary Information]
240pu [Proprietary Information] [Proprietary Information]
24 1Pu [Proprietary Informati on] [Proprietary Information]
103mRh [Proprietary Information] [Proprietary Information]
IOSRh [Proprietary Informati on] [Proprietary Information]
106Rh [Proprietary Information] [Proprietary Information]
J06mRh [Proprietary Information] [Proprietary Information]
103Ru [Proprietary Information] [Proprietary Information]
1osRu [Proprietary Information] [Proprietary Information]
106Ru [Proprietary Information] [Proprietary Information]
122 sb [Proprietary Information] [Proprietary Information]
124 Sb [Proprietary Information] [Proprietary Information]
125 Sb [Proprietary Information] [Proprietary Information]
126Sb [Proprietary Information] [Proprietary Information]
127 Sb [Proprietary Information] [Proprietary Information]
128 Sb [Proprietary Information] [Proprietary Information]
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NWMl-201 3-021, Rev. 2 Chapter 13.0 -Accident An alysis Table 13-23. Bounding Radionuclide Liq uid Stream Concentrations (4 pages)
Unit operation Target dissolution Uranium recovery and recycle
[Proprietary Information]
Dissolver roduct 12smsb [Proprietary Information] [Proprietary Information]
129Sb [Proprietary Information] [Proprietary Information]
1s1sm [Proprietary Information] [Proprietary Information]
1s3sm [Proprietary Information] [Proprietary Information]
1s6sm [Proprietary Information] [Proprietary Information]
89Sr [Proprietary Information] [Proprietary Information]
9osr [Proprietary Information] [Proprietary Information]
91sr [Proprietary Information] [Proprietary Information]
92 [Proprietary Information]
Sr [Proprietary Information]
99Tc [Proprietary Information] [Proprietary Information]
99mTc [Proprietary Information] [Proprietary Information]
12smTe [Proprietary Information] [Proprietary Information]
121Te [Proprietary Information] [Proprietary Information]
121mTe [Proprietary Information] [Proprietary Information]
129Te [Proprietary Information] [Proprietary Information]
129mTe [Proprietary Information] [Proprietary Information]
131Te [Proprietary Information] [Proprietary Information]
131mTe [Proprietary Information] [Proprietary Information]
132Te [Proprietary Information] [Proprietary Information]
133Te [Proprietary Information] [Proprietary Information]
133mTe [Proprietary Information] [Proprietary Information]
134Te [Proprietary Information] [Proprietary Information]
23 1Th [Proprietary Information] [Proprietary Information]
234Th [Proprietary Information] [Proprietary Information]
232u [Proprietary Information] [Proprietary Information]
234U [Proprietary Information] [Proprietary Information]
23su [Proprietary Information] [Proprietary Information]
236u [Proprietary In formation] [Proprietary Information]
231u [Proprietary Information] [Proprietary Information]
mu [Proprietary Information] [Proprietary Information]
1J1mxe [Proprietary Information] [Proprietary Information]
133 Xe [Proprietary Information] [Proprietary Information]
1JJmxe [Proprietary Information] [Proprietary Information]
135 Xe [Proprietary Information] [Proprietary Information]
1Jsmxe [Proprietary Information] [Proprietary Information]
89my [Proprietary Information] [Proprietary Information]
90y [Proprietary Information] [Proprietary Information]
90my [Proprietary Information] [Proprietary Information]
9J y [Proprietary Information] [Proprietary Information]
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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 9Imy [Proprietary Information] [Proprietary Information]
92y [Proprietary Information] [Proprietary Information]
93y [Proprietary Information] [Proprietary Information]
93zr [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 , Source Term Calculations, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, February 2015.
EOI = end of irradiation.
In each case, a jacketed vessel is assumed to be filled with process solution appropriate to the process location, with the process offgas ventilation system operating. A level monitoring system will be available to monitor tank transfers and stagnant store vo lumes on all tanks processing LEU or fission product solutions.
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 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.
13.2.4.2 Identification of Event Initiating Conditions The accident initiating event is generally described as a process equipment failure. The PHA identified simi lar accident sequences in four nodes associated with leaks of enriched uranium solution into heating and/or cooling coils surrounding safe-geometry tanks or vessels. The PHA identified predominately corrosive degradation of the tank or overpressure of the tank as potential causes that might damage this interface and allow enriched uranium solution to leak into the cooling system media or into the steam condensate for the heating system.
The primary containment fails , which allows radioactive or fissi le solutions to enter an auxiliary system.
Radioactive or fissile solution leaks across the mechanical boundary between a process vessel and associated heating/cooling jacket into the heating/cooling media . Where heating/cooling jackets or heat exchangers are used to heat or cool a fissile and/or high-dose process solution, the potential exists for the ba1Tier between the two to fail and allow fissile and/or hi gh-dose process solution 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.
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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 also occur. Either of these potential outcomes can exceed the performance criteria of one process upset, leading to accidental nuclear criticality 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) system 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-geometry 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 solutions 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."
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- :~*:~*:* NWM I
..*... NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis
~ * *! NOfllTHWEn MEDCAl. ISOTOPfS 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 in the process system. The high-dose uranium solution source term bounds this analysis. Solution leaks into 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 uranyl 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.
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.
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.
13.2.4.7.1 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, have similar exposure consequences due to the criticality 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.
Based on the analysis of several accidental nuclear criticalities in industry, LA-13638 identifies that a uranium solution criticality can yield between 10 16 to 10 17 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.
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- ... NWMI
.~ NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis
. ~* * ~ NOllTHWUT MEDICAL tSOTOPES 13.2.4.8.1 IROFS RS-04, Hot Cell Shielding Boundary 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 systems.
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 shielding for workers and the public during process 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 boundary 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 reduce the exposure from postulated upsets within the hot cell shielding boundary to less 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 operable when the equipment in that area is in the operating or standby modes.
13.2.4.8.2 IROFS CS-06, Pencil Tank and Vessel Spacing Control using the Diameter of the Tanks, Vessels, or Piping All tanks, vessels, or piping systems involved 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 cooling loop with monitoring for uranium process solution or high-dose process solution will be provided to safely contain fissile process solution that leaks across this boundary, if the primary boundary fails. The dual-purpose safety function of this closed loop is to prevent fissile process solution from causing accidental nuclear criticality and to prevent high-dose 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 Some operations may have higher doses during short periods of the operation . The average worker expos ure rate is designed to be 0.5 mrem/hr, or less. Areas not normall y accessible by the worker may have higher dose rates (e.g., streaming radiation around normally inaccessible remote manipulator penetrations well above the worker's head).
2 The shielding is not credited for mitigat ing dose rates during an accidental nuclear criticality; instead, additional IROFS are identified to provide double-contingency protection to prevent (reduce the likelihood of) an accidental nuclear criticality.
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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 heating or cooling media (e.g., steam condensate conductivity, cooling water radiological activity, uranium concentration, etc.) wi ll be conducted to alert the operator that a breach has occurred and that additional corrective actions are required to identify and isolate the failed component and restore the closed loop integrity. Discharged solutions from this system wi ll be handled as potentially fissile and 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 closed cooling loop with monitoring for breakthrough of process solution will be provided to contain process solution that leaks across this boundary, if the boundary fails. IROFS CS-27 is applied to those high-heat capacity cooling jackets (requiring very large loop heat exchangers) servicing condensers where the leakage is always from the cooling loop to the condenser, reducing back-leakage, and the risk of product solutions entering the condenser is very low by evaporator or concentrator design.
The purpose of this safety function is to monitor the condition of the condenser cooling jacket to ensure that in the unlikely event that a condenser 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 isolate any high-dose fissile product solids (from the same event) from penetrating the hot cell shielding boundary, and any high-dose fission gases from penetrating the hot cell shielding boundary during normal operations. 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 cooling media (e.g., cooling water radiological activity, uranium concentration, etc.) will be conducted to alert the operator that a breach has occurred and that additional corrective actions are required to identify and isolate the failed component and to restore the closed loop integrity. Closed loop pressure will also be monitored to identify a leak from the closed loop to the process system. Discharged solutions from this system will be handled as potentially fissile and sampled according to IROFS CS-16 and CS- 17 prior to discharge to a non-safe geometry.
13.2.4.8.5 IROFS CS-20, Evaporator or Concentrator Condensate Monitoring As an AEC, the condensate tanks 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 system is to (1) detect an anomaly in the evaporator or concentrator indicating high uranium content in the condenser (due to flooding or excess ive foaming) , and (2) prevent high concentration uranium solution from being available in the condensate tank for discharge to a non-favorable geometry system or in the condenser for leaking to the non-safe geometry cooling loop. The safety function of this IROFS is to prevent an accidental nuclear criticality. The detection system will work by continuously monitoring condensate uranium content and detecting high uranium concentration, and then shutting down the evaporator to isolate the condensate from the condenser and condensate tank. At a limiting setpoint, the uranium monitor-detecting devi ce will close an isolation valve in the inlet to the evaporator (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 designed to produce a valve-open permissive signal that fails to an open state, closing the valve on loss of electrical power. The isolation valve is des igned to fail-closed on loss of instrument air, and the solenoid is designed to fai l-closed on loss of signal. The locations where this IROFS is used will be determined during final design.
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NWMl-20 13-021 , Rev. 2 Chapter 13.0 -Accident Analysis 13.2.4.8.6 IROFS CS-18, Backflow Prevention Device 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 solutions from backflowing from the tank into systems that are not designed for fissile solutions that could lead to accidental nuclear criticality or to locations 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 lity and availability for that location.
The backflow prevention device features for high-dose product solutions will be located inside the hot cell shielding and confinement boundaries ofIROFS RS-04 and RS-01 , respectively. The feature is designed such that spills from overflow are directed to a safe geometry confinement berm controll ed by IROFS CS-08 (described in NWMI-2015-SAFETY-004, Quantitative Risk Analysis 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 is to prevent accidental nuclear criticality by providing a safe-geometry tank if a fissile solution backs-up into an auxiliary chemical addition system.
IROFS CS-19 will be used where conventional backflow prevention in pressurized 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 vessel that is filled with chemical reagent using the conventional backflow prevention devices, and then provides a pump to add the reagents to the respective process system under pressure. Safe-geometry day tanks servicing high-dose product solutions systems will be located in the hot cell shielding 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 cooling 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 remotely. Operating procedures will identify an operational high-level fill operating limit for each tank. As part of the level detector, a high-level audible alarm and light will be provided to indicate a high level above this operating limit so that the operator can take action to correct conditions leading to fai lure 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).
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- ~*~h NWMI
~* * ~ NCMO'HWHT MEOfC.Al ISOTOPU NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis Tank and vessel 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 selected 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.
Additional 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 by 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 are 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 electrical heaters cease operation as part of the passive operation mode.
Pressure-relief confinement system for the target dissolver offgas 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).
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- ~*:~":" NWMI NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis
- *~:!~*
- NOflTtfWUTMEDtCALISOTOPES 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 transfers 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 secondary 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 based 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.
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NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis 13.2.6 Natural Phenomena Events Chapter 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 sequences 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/year (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 prevents impact to SS Cs in the building.
The local area impact may result in loss of utilities (e.g., electrical power) and reduced access by 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 directly impact SSCs important to safety (e.g., components 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 loss of electrical power) that can be initiated concurrently with the tornado (either as an indirect result or as an additional random failure). This evaluation analyzes the impact of tornado winds on these SSCs.
Tornado impact on the facility structure - High wind pressures could cause a partial or complete collapse of the facility structure, which may cause damage to SS Cs important to safety or impact the availability and reliability of those SSCs. A partial or complete structural collapse may also lead directly to a radiological or chemical release or a potential nuclear criticality, if damage caused by the collapse creates a violation of criticality spacing requirements. Tornado wind-driven missiles could penetrate the facility building envelope (walls and roof), impacting the availabi lity 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 pressures and tornado wind-driven missiles could damage SSCs important to safety located outside the RPF building envelope. The damage sustained may impact the availability and reliability of the SSCs important to safety. Loss of site power may affect the ability of SSCs important to safety located within the facility building envelope to respond to additional events.
A partial or complete collapse of the facility structure could also lead directly to an accident with adverse intermediate or high consequences. The only IROFS located outside the RPF building envelope is the exhaust stack. Buckling or toppling of the exhaust stack would affect its ability and availabi lity to mitigate other events with intermediate consequences. The return frequency of the design basis tornado is 10-5/yr, making the initiating event highly unlikel y.
No additional IROFS are required.
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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 safety. This evaluation demonstrates how the facility 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 Design Loads for Buildings and Other Structures. 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 Code Requirements for Structural Concrete, 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 empirical methodology with no defined annual exceedance probability.
For impact on the facility, the PMP for 25.9 square kilometers (km 2) (10 square miles [mi 2]) is evaluated.
Large amounts of rain water accumulating on the roof could lead to collapse of the roof. A partial or complete collapse of the faci lity 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 National Weather Service (NWS)/National Oceanic and Atmospheric Administration (NOAA)
Hydrometeorological Report No. 51 , Probable Maximum Precipitation Estimates, United States East of the 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 empirical 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 Generalized Estimates ofProbable Maximum Precipitation 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.
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~* *~ NORTMWUT MEOtC.Al ISOTOl'£S NWMl-2013-021, Rev . 2 Chapter 13.0 - Accident Analysis No additional IROFS are required.
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 safely 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 rivers 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 downpours 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 standard industrial design. Rainwater that falls on the waste management truck ramp and accumulates in the trench drain has low 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 2x 10*3 year return frequency flood, which can be considered an unlikely event according to performance criteria. However, the site is located at an elevation of 248.4 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 dry 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 Structures, 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. During 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 docking 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 shield plug in the presence of personnel. This event would result in potentially lethal doses to workers in a short period of time.
Seismic ground shaking can directly damage SSCs relied on for safety or lead to damage of the facility, including partial or complete 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 intermediate consequence.
3 The recommended standard for determi ning the probably maximum flood, ANS 2.8, Determining Design Basis Flooding at Power Reactor Sites, has been withdrawn.
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~* * ~ NORTHWtsT MEOtCAL ISOTOPfS NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 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 , NEHRP Recommended Seismic Provisions for New Buildings and Other Structures. The MCfa for this site is governed by the probabilistic maximum-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 maximum-considered earthquake ground-shaking, or a probability of failure of l .2x 1o-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, /earthquake = 4x 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- loading event may cause failure of the roof, impacting the SSCs' ability to perform associated safety functions. NRC DC/COL ISG-07, Interim Staff Guidance on Assessment ofNormal 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 bounds 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 primary 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, Seasonal Variation of 10-Square Mile Probable Maximum Precipitation Estimates, United States East of the 1051h 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.
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~ * *! N<HllTHWE.ST MEDJCAL ISOTOP£S 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 1o-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.
Table 13-24. Analyzed Accidents Sequences (9 pages)
Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.01 High-dose solution or enriched
- 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$0TOP£S Table 13-24. Analyzed Accidents Sequences (9 pages)
Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.04 Liquid enters process vessel
- See Section 13.2.3.8 retained radionuclides S.R.05 High-dose solution enters the
- Not credible or low consequence UN blending and storage tank S.R.06 High flow through IRU causing
- Bounded by S.R.04 premature release of high-dose iodine gas S.R.07 Loss of temperature control on
- Bounded by S.R.04 the IRU leading to release of high-dose iodine S.R.08 Loss of vacuum pumps
- Bounded by S.R.04 S.R.09 Loss ofIRU or carbon bed
- Bounded by S.R.04 media to downstream part of the system S.R.10 Wrong retention media added to
- Event unlikely with intermediate consequence bed or saturated retention media S.R.12 Mo product cask removed from
- Event unlikely with intermediate consequence the hot cell boundary with improper shield plug installation S.R.13 High-dose containing solution
- IROFS CS-06, Pencil Tank and Vessel Spacing Control using condensate system the Diameter of the Tanks, Vessels, or Piping
- IROFS CS-10, Closed Safe-Geometry Heating or Cooling Loop with Monitoring and Alarm
- IROFS CS-27, Closed Heating or Cooling Loop with Monitoring and Alarm
- IROFS CS-20, Evaporator or Concentrator Condensate Monitoring
- IROFS CS-18, Backflow Prevention Device
- IROFS CS-19, Safe-Geometry Day Tanks
- See Section 13.2.4.8 S.R.14 IX resin failure due to wrong
- Bounded by S.R.01 reagent or high temperature S.R.16 Backflow of high-dose
- Bounded by S.R.13 radiological and/or fissile solution into auxiliary system (purge air, chemical addition line, water addition line, etc.)
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- NORTtfWf.$T MEDICAL ISOTOf'ES 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.17 Carryover of high-dose solution
- IROFS RS-08, Sample and Analysis of Low Dose Waste Tank into condensate (a low-dose Dose Rate Prior to Transfer Outside the Hot Cell Shielded waste stream) Boundary
- See Section 13.2.7.1 S.R.18 High-dose solution flows into
- Low consequence event that does not challenge IROFS RS-04 the solidification media hopper S.R.19 High target basket retrieval dose
- Design evolved after PHA, accident sequence eliminated rate S.R.20 Radiological spill of irradiated
- Bounded by S.R.01 LEU target material in the hot cell area S.R.21 Damage to the hot cell wall
- Low consequence event that does not damage shielding providing shielding function ofIROFS RS-04 S.R.22 Decay heat buildup in
- Low consequence event unprocessed LEU target material removed from targets leads to higher-dose radionuclide offgassing S.R.23 Offgassing from irradiated target
- See Section 13.2.2.8 upper valve is opened S.R.24 Bagless transport door failure
- See Section 13.2.2.8 S.R.25 HEPA filter failure
- See Section 13.2.2.8 S.R.26 Failed negative air balance from
- See Section 13.2.2.8 exhaust a radionuclide buildup in an area S.R.27 Extended outage of heat leading
- Highly unlikely event for process solutions containing fission to freezing, pipe failure, and products release ofradionuclides from Bounded by S.C.04 for target fabrication systems liquid process systems S.R.28 Target or waste shipping cask or
- Information will be provided in the Operating License container not loaded or secured Application according to procedure, leading to personnel exposure 13-77
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~* * ~ NORTHWEST MEDtcAL ISOTOP£S Table 13-24. Analyzed Accidents Sequences (9 pages)
Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.29 High dose to worker from
- IROFS RS-12, Cask Containment Sampling Prior to Closure release of gaseous radionuclides Lid Removal during cask receipt inspection
- IROFS RS-13, Cask Local Ventilation During Closure Lid and preparation for target basket Removal and Docking Preparations removal
- See Section 13.2.7.1 S.R.30 Cask docking port failures lead
- See Sections 13.2.2.8 and 13.2.7.l airborne radioactivity S.R.31 Chemical burns from
- Judged unlikely event with intermediate consequence contaminated solutions during sample analysis S.R.32 Crane load drop accidents
- IROFS FS-01, Enhanced Lift Procedure
- IROFS FS-02, Overhead Cranes
- See Section 13.2. 7.1 S.C.01 Failure of facility enrichment
- Judged highly unlikely based on supplier's checks and balances limit S.C.02 Failure of administrative control
- IROFS CS-02, Mass and Batch Handling Limits for Uranium on mass (batch limit) during Metal, [Proprietary Information], Targets, and Laboratory handling of fresh U, scrap U, Sample Outside Process Systems LEU target material, targets, and
- IROFS CS-03, Interaction Control Spacing Provided by samples Administrative Control
- IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement
- See Section 13.2. 7.2 S.C.03 Failure of interaction limit
- IROFS CS-02, Mass and Batch Handling Limits for Uranium during handling of fresh U, scrap Metal, [Proprietary Information], Targets, and Laboratory U, LEU target material, targets, Sample Outside Process Systems containers, and samples
- 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 S.C.04 Spill of process solution from a
- IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry tank or process vessel leading to Confinement using the Diameter of Tanks, Vessels, or Piping accidental criticality
- 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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- ~~!'!* . NO~ST MEDtcAL ISOTOPES 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 ssile solution into the
- Bounded by S.R 13 heating or cooling jacket on the tank or vessel S.C.06 System overflow to process
- IROFS CS-11, Simple Overflow to Normally Empty Safe ventilation involving fissile Geometry Tank with Level Alarm material
- IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line
- IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary
- See Section 13.2.7.2 S.C. 07 Fissile solution leaks across
- Bounded by S.R.13 mechanical boundary between process vessels and heating/cooling jackets into heating/cooling media S.C.08 Backflow of high-dose
- Bounded by S.R.13 radiological and/or fissile solution into auxiliary system (purge air, chemical addition line, water addition line, etc.)
S.C.09 High concentrations of uranium
- IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry enter the concentrator or Confin ement using the Diameter of Tanks, Vessels, or Piping evaporator condensates
- IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels
- IROFS CS-26, Processing Component Safe Volume Confinement
- See Section 13.2.7.2 S.C.10 High concentrations of uranium
- IROFS CS-14, Active Discharge Monitoring and Isolation enter the low-dose or high-dose
- IROFS CS-15, Independent Active Discharge Monitoring and waste collection tanks Isolation
- IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal
- See Section 13.2.7 .2 13-79
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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 High concentrations of uranium
- Bounded by S.C.04 and S.C. l 0 in contactor solvent regeneration aqueous waste S.C.12 High concentrations of uranium
- IROFS CS-04, Interaction Control Spacing Provided by in the LEU target material wash Passively Designed Fixtures and Workstation Placement solution
- 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 S.C.13 High concentrations of uranium
- IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry in the nitrous oxide scrubber Confinement using the Diameter of Tanks, Vessels, or Piping
- IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal
- See Section 13.2.7.2 S.C.14 High concentrations of uranium
- IROFS CS-16, Sampling and Analysis of Uranium Mass or in the IX waste collection tanks Concentration Prior to Discharge or Disposal effluent
- IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal
- See Section 13.2.7.2 S.C.15 High concentrations of uranium
- IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry in the IX resin waste 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-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal
- See Section 13.2.7.2 S.C.17 High concentrations of uranium
- IROFS CS-16, Sampling and Analysis of Uranium Mass or in the solid waste encapsulation Concentration Prior to Discharge or Disposal process
- 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
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.19 Failure of PEC - Component
- IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry safe geometry dimension or safe Confinement using the Diameter of Tanks, Vessels, or Piping volume
- IROFS CS-07, Pencil Tank and Vessel Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels
- IROFS CS-26, Processing Component Safe Volume Confinement
- See Section 13.2.7.2 S.C.20 Failure of concentration limits
- No credible path leading to criticality identified or not credible by design S.C.21 Target basket passive design
- IROFS CS-02, Mass and Batch Handling Limits for Uranium control failure on fixed Metal, [Proprietary Information], Targets, and Laboratory interaction spacing Sample Outside Process Systems
- IROFS CS-03, Interaction Control Spacing Provided by Administrative Control
- See Section 13.2. 7.2 S.C.22 High concentration of uranium
- IROFS CS-04, Interaction Control Spacing Provided by in the TCE evaporator residue 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
- IROFS CS-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal
- See Section 13.2. 7.2 S.C.23 High concentration in the spent
- IROFS CS-04, Interaction Control Spacing Provided by si licone oi l waste Passively Designed Fixtures and Workstation Placement
- IROFS CS-05, Container Batch Volume Limit
- IROFS CS-06, Penci l Tank, Vessel, or Piping Safe Geometry Confi nement 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-17, Independent Sampling and Analysis of Uranium Concentration Prior to Discharge or Disposal
- Bounded by S.C.17 filters and subsequent failure 13-81
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Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.27 Failure of administratively
- IROFS CS-03, Interaction Control Spacing Provided by controlled container volume Administrative Control limits
- 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 S.C.28 Crane load drop accidents
- IROFS FS-01, Enhanced Lift Procedure
- IROFS FS-02, Overhead Cranes
- See Section 13.2.7.2 S.F.01 Pyrophoric fire in uranium metal
- Event highly unlikely based on credible physical conditions S.F.02 Accumulation and ignition of
- IROFS FS-03, Process Vessel Emergency Purge System flammable gas in tanks or
- See Section 13.2.7.3 systems S.F.03 Hydrogen detonation in
- Judged highly unlikely based on credible physical conditions reduction furnace S.F.04 Fire in reduction furnace
- Judged unlikely based on event frequency S.F.05 Fire in a carbon retention bed
- IROFS FS-05, Exhaust Stack Height
- See Section 13.2.7.3 S.F.06 Accumulation of flammable gas
- Bounded by S.F.02 in ventilation system components S.F.07 Fire in nitrate extraction system -
- Event unlikely with intermediate or low consequences combustible solvent with uranium S.F.08 General facility fire
- Information will be provided in the Operating License Application S.F.09 Hydrogen exp losion in the
- Information will be provided in the Operating License facility due to a leak from the Application hydrogen storage or distribution system S.F.10 Combustible fire occurs in hot
- Information will be provided in the Operating License cell area Application S.F.11 Detonation or deflagration of
- Information will be provided in the Operating License natural gas leak in steam Application generator room S.N.01 Tornado impact on facility and
- Judged highly unlikely event based on return frequency SSCs important to safety S.N.02 High straight-line winds impact
- Judged highly unlikely to result in structure failure the facility and SSCs important to safety S.N.03 Heavy rain impact on facility
- Bounded by S.N.06 and SSCs important to safety 13-82
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Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S. .04 Flooding impact to the facility
- Judged highly unlikely event based on facility location above and SSCs important to safety the 500-year flood plain S.N.05 Seismic impact to the facility
- Judged highly unlikely to result in structure failure and SSCs important to safety
- IROFS FS-04, Irradiated Target Cask Lifting Fixture
- See Section 13.2.6.5 S.N.06 Heavy snowfall or ice buildup on
- Judged highly unlikely to result 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 leads to an accident sequence of intermediate or high consequence S.M.02 Facility evacuation impacts on
- Judged likely event with low consequence operations S.M.03 Localized flooding due to
- IROFS CS-08, Floor and Sump Geometry Control of Slab internal system leakage or fire Depth, Sump Diameter or Depth for Floor Spill Containment suppression sprinkler activation Berms
- See Section 13.2. 7.2 S.CS.01 Nitric acid fume release
IROFS items relied on for safety. PHA preliminary hazards analysis.
IRU iodine removal unit. SSC structures, systems, and components.
IX ion exchange. TCE trichloroethylene LEU low-enriched uranium. U uranium.
Mo mol ybdenum . UN uranyl nitrate.
Table 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 detail developed for the Operating License Application.
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- NORTHWEST MfDICAL lSOTOP£S 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 cell secondary confinement boundary ./
RS-04 Hot cell shielding boundary ./
RS-05 Reserved RS-06 Reserved RS-07 Reserved RS-08 Sample 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 lation during closure lid removal and docking preparations 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, vessel, 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, sump diameter or ./
depth for floor spill containment berms CS-09 Double-wall piping ./
CS-10 Closed 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 Condensing pot or seal pot in ventilation vent line ./
CS-13 Simple overflow to normally empty safe geometry floor with level ./
alarm in the hot cell containment boundary 13-84
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IROFS Engineered Administrative designator Descriptor safety feature control CS-14 Active discharge monitoring and isolation ./
CS-15 Independent active discharge monitoring and isolation ./
CS-16 Sampling and analysis of uranium mass or concentration prior to ./
discharge or disposal CS-17 Independent sampling and analysis of uranium concentration prior ./
to discharge or disposal CS-18 Backflow prevention device ./
CS-19 Safe-geometry day tanks ./
CS-20 Evaporator or concentrator condensate monitoring ./
CS-21 Visual inspection of accessible surfaces for foreign debris ./
CS-22 Gram estimator survey of accessible surfaces for gamma activity ./
CS-23 Nondestructive assay of items with inaccessible surfaces ./
CS-24 Independent nondestructive assay of items with inaccessible surfaces ./
CS-25 Target housing weighing prior to disposal ./
CS-26 Processing component safe volume confinement ./
CS-27 Closed heating or cooling loop with monitoring and alarm ./
FS-01 Enhanced lift procedure ./
FS-02 Overhead cranes ./
FS-03 Process vessel emergency purge system ./
FS-04 Irradiated target cask lifting fixture ./
FS-05 Exhaust stack height ./
IROFS items relied on for safety.
Table 13-26. Accident Sequence Category The following subsections describe the IROFS that Definitions are not previously discussed elsewhere in this Section containing chapter. The IROFS are grouped according to . I related IROFS their respective accident sequence categories, as
- Definition description shown in Table 13-26. S.R. Radiological 13.2.7.1 S.C. Criticality 13.2.7.2 13.2.7.1 Items Relied on for Safety for S.F. Fire or explosion 13.2.7.3 Radiological Accident Sequences S.N. Natural phenomena 13.2.7.4 (S.R.)
S.M. Man-made 13.2.7.5 The following IROFS fall under the radiological s.cs. Chemical safety 13.2.7.6 accident sequence category and are not discussed IROFS items relied on for safety.
elsewhere in this chapter.
13.2.7.1.1 IROFS RS-08, Sample and Analysis of Low Dose Waste Tank Dose Rate Prior to Transfer O utside the Hot Cell Shielded Boundary As an augmented administrative control (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, sampled, and the sample analyzed for high radiation.
Batches that satisfy the sample criteria can be transferred to the low-dose waste encapsulation system.
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~ *.*! . NOffTHWl:ST MEOtCAl ISOTOPfS The safety function of 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 outside 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 limits to be determined). The safety function ofIROFS RS-12 is to prevent personnel exposure to high-dose gaseous radionuclides.
The system wi ll identify 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 alann. 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 lugs. 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.
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~* * ~ NOflTHWtST MEDtcAL ISOTDitf.S 13.2.7.1.5 IROFS RS-15, Cask Docking Port Enabling Sensor 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 close.
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 pre-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 elsewhere 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 activities 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 interaction 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.
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~ * *! NORTHWEST MEDICAL ISOTOPES 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 safe-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 reasons. 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 all 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 installed 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 non-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 alarm 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.
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~* * ~ 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 safe-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 system. 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 potentially fissile process solution for which IROFS CS-13 is assigned, a simple overflow line will be installed above the high alarm setpoint. The overflow will be directed to one or more safe-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 uranium 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 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 of IROFS CS-14 is to prevent an accidental nuclear criticality. The closed-loop system is designed to isolate the discharge points listed below by actively monitoring the solution stream for uranium concentration using a suitable uranium monitor. At a limiting setpoint, the uranium monitor will close an isolation valve in the discharge line to stop the discharge. The uranium monitor is designed to produce a valve-open permissive signal 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.
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- ~* '~ NOflTtfWEST MEDICAL ISOTOPES 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-loop system is designed to isolate the discharge points listed below by actively monitoring the solution stream for uranium concentration using a suitable 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 isolation 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 electrical power.
The isolation valve is designed to fail-c losed 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 discharge from the safe-geometry container, tanks, or vessels assigned IROFS CS-16 to non-favorable geometry systems, the container, tank, or vessel will be isolated and placed under admjnistrative control, recirculated or otherwise uniformly mixed, 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 volumes 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 solution concentration or under a safe mass for the disposal container. Systems, tanks, or vessels 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, sampled, 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 analysis equipment, checked with independent standards) of that performed in IROFS CS-16.
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NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis The discharge or disposal will only be approved following independent review of the sample results to confirm the uranium content is below the limiting setpoint for uranium concentration or batch mass for the contents and under the independent oversight 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 mass or concentration limit for the container is not exceeded.
The safety function of IROFS CS-17 is to prevent accidental nuclear criticality caused by discharging 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 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 include:
TCE recycle 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 visual inspection will be performed to identify foreign matter on accessible surfaces of equipment and waste materials approved for this method prior to disposal. All visible foreign material is assumed to be uranium. All surfaces must be non-porous. Materials involved must be solids (no solutions or liquids present). All surfaces must be visually accessible either directly or through approved inspection devices. The inspection criterion is for no foreign material of discernible thickness to be visible (transparent films allowed). The safety function of this AC is to ensure that no significant 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 items approved by the Criticality Safety Manager. The waste will not be consolidated 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 initially approved include disassembled irradiated or scrap target housi ng parts or pieces.
13.2.7.2.13 IROFS CS-22, Gram Estimator Survey of Accessible Surfaces for Gamma Activity As an AAC, a gram estimator survey will be performed on all accessible surfaces of equipment and waste materials approved for this method prior to disposal. The survey will be performed on low-risk waste streams that have surfaces that are I 00 percent accessib le with the measurement instrument. The measurement setpoint is designed to detect activity from 15 g of 235 U uniformly spread over 30 kilograms (kg) of 4-mil (thousandth of an inch) thick polyethylene sheeting (both sides) 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 non-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 prevent accidental nuclear criticality. All surfaces will need to be accessible to the instrument used. The waste stream must 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 applicable to radiological waste generated outside the hot cell boundary that has had a low risk for direct contact with uranium-bearing materials.
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' ~* * ~ NOfllffWtrT MEDICAi. ISOTIM'ES NWMl-2013-021 , Rev. 2 Chapter 13.0 - Accident Analysis 13.2.7.2.14 IROFS CS-23, Non-Destructive Assay of Items with Inaccessible Surfaces As an AAC, a nondestructive 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 Criticality 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 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. The control will be used as a backup to IROFS CS-16, IROFS CS-21 or IROFS CS-23, as approved by the Criticality Safety Manager for each waste stream. 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 Criticality Safety Manager. Waste consolidation will be conducted after independent verification of the two methods of quantifying uranium mass 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 housings, target housing pieces 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 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. 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 will be exercised on the waste items weighed on approved scales and at mass or concentration setpoint(s) approved by the Criticality Safety Manager.
Waste consolidation will be conducted after independent verification of the two methods 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 As 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 leak) 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 elsewhere 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, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist). Lifting devices for shipping containers will be designed, operated, and tested according to ANSI N14.6, Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More for Nuclear Materials.
The safety function ofIROFS FS-02 is to prevent (by reducing the likelihood) mechanical fai lure 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 As 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 water 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.
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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.
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NWMl-2013-021, Rev. 2 Chapter 13.0 - Accident Analysis 13.3 ANALYSIS OF ACCIDENTS WITH HAZARDOUS CHEMICALS This 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 analysis 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 sequence 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, analysis procedures, reagents, analytical equipment requirements, and sample material disposal protocols will all be specified per procedures and/or work plans prepared and discussed prior to sampling or analytical activities. Operators and technicians will be required to wear personal protective equipment, specifically for eye and skin protection.
Radiologically contaminated acidic and caustic solution 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 provi de an additional layer of protection for eyes and skin (e.g., protective eyewear/face shield, laboratory coat or apron, anti-contamination chemical resistant 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.
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. ' ~ *.*! NOKTifWHT MEDfCAl ISOTOPES NWMl-2013-021, Rev. 2 Chapter 13.0 -Accident Analysis 13.3.2 Nitric Acid Fume Release 13.3.2.1 Chemical Accident Description 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 identifies hazardous chemical release 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 dispersion conditions of moderate wind. Unmitigated exposure to a nearby 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 170 ppm for a high consequence exposure to the worker. These determinations were made using the ALOHA (Areal Locations of Hazardous 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 release on RPF operations would require 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. Environmental 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.
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13.4 REFERENCES
10 CFR 20, "Standards for Protection Against Radiation," Code a/Federal Regulations, Office of the Federal Register, as amended.
10 CFR 30, "Rules of General App licability to Domestic Licensing of Byproduct Material," Code of Federal Regulations, Office of the Federal Register, as amended.
10 CFR 50, "Domestic Licensing of Production and 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 Regulations, Office of the Federal Register, as amended.
10 CFR 70.61 , "Performance Requirements," Code a/Federal Regulations, Office of the Federal Register, as amended.
10 CFR 71, "Packaging and Transportation of Radioactive Material, Code ofFederal Regulations, Office of the Federal Register, as amended.
ACI 318, Building Code Requirements for Structural Concrete, American Concrete Institute, Farmington Hills, Michigan, 2014.
AISC 360, Specification for Structural Steel Buildings, American Institute of Steel Construction, Chicago, Illinois, 2010.
ANS 2.8, Determining Design Basis Flooding at Power Reactor Sites, American Nuclear Society, La Grange Park, Illinois, 1992, W2002 .
ANSI N 14.6, Standard for Special Lifting Devices for Shipping Containers Weighing 10, 000 Pounds (4,500 kg) or More for Nuclear Materials, American Nuclear Society, La Grange Park, Illinois, 1993 .
ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, American Nuclear Society, La Grange Park, lllinois, 1998 (Reaffirmed 2007).
ASCE 7, Minimum Design Loads for Buildings and Other Structures, American Society of Civil Engineers, Reston, Virginia, 2010.
ASME B30.2, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist) , American Society of Mechanical Engineers, New York, New York, 2005.
CDC, 2010, NIOSH Pocket Guide to Chemical Hazards, 2010- 168c, Centers for Disease Control and Prevention, http://www.cdc.gov/niosh/npg/, down loaded February 27, 2015.
DC/COL ISG-07, Interim Staff Guidance on Assessment ofNormal and Extreme Winter Precipitation Loads on the Roofs ofSeismic Category I Structures, U.S . Nuclear Regulatory Commission, Washington, D.C., 2008.
DOE-HDBK-30 I 0, DOE Handbook-Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, Change Notice No. 1, U.S. Department of Energy, Washington, D.C., December 1994 (R2013).
DOE-STD-1090-2011, Hoisting and Rigging, U.S. Department of Energy, Washington, D.C.,
September 30, 20 11.
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- * * ~OflTHWUT MEDICAL ISOTOP£S EPA 52011 020, Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 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 , NEHRP Recommended Seismic Provisions for New Buildings and Other Structures, Federal Emergency Management Agency, Washington, D.C., 2009.
Hydrometeorological Report No. 51 , Probable Maximum Precipitation Estimates, United States East of the 105th Meridian, U.S. Department of Commerce, National Oceanic and Atmospheric Administration, Washington, D.C., 1978.
Hydrometeorological Report No. 53 (NUREG/CR-1486), Seasonal Variation of JO-Square Mile Probable Maximum Precipitation Estimates, United States East of the 1051h Meridian, U.S. Department of Commerce, National Oceanic and Atmospheric Administration, U.S . Nuclear Regulatory Commission, Office of Hydrology National Weather Service, Washington, D.C. , April 1980.
IBC, 2012, International Building Code, as amended, International Code Council, Inc. , Washington, D.C., February 2012.
ICRP-26, Recommendations of the International Commission on Radiological Protection , International Commission on Radiological Protection, Ottawa, Canada, 1977.
ICRP-30, Limits for Intakes of Radionuclides by Workers , International Commission on Radiological Protection, Ottawa, Canada, 1979.
ICRP-72, Age-Dep endent Doses to the Members of the Public from Intake of Radionuclides - Part 5 Compilation of Ingestion and Inhalation Coefficients , International Commission on Radiological Protection, Ottawa, Canada, 1995 .
LA-13638, A Review of Criticality Accidents, Los Alamos National Laboratory, Los Alamos, New Mexico, 2000.
NAP 1994, Estimating Bounds on Extreme Precipitation Events, National Academy Press, National Research Council, Washington, D.C., I 994.
NOAA Technical Report NWS 25 , Comparison of Generalized Estimates of Probable Maximum Precipitation with Greatest Observed Rainfalls, National Oceanic and Atmospheric Administration, Washington, D.C., 1980.
NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors - Format and Content, Part 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February I 996.
NUREG-1940, RASCAL 4: Description of Models and Methods, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C., December 2012.
NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C. , March 1998.
NWMI-2013-CALC-006, Overall Summary Material Balance - MURR Target Batch, Rev. D, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.
NWMI-2013-CALC-011 , Source Term Calculations, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.
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NWMl-2013-021 , Rev . 2 Chapter 13.0 - Accident Analysis NWMI-2014-051 , integrated Safety Analysis Plan for the Radioisotope Production Facility, Rev. A, Northwest Medical Isotopes, Corvallis, Oregon, 2014.
NWMI-2014-CALC-014, Selection ofDominant Target i sotopes for NWMJ Material Balances, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2014.
NWMI-2015-RPT-009, Fission Product Release Evaluation, Rev. B, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.
WMI-2015-SAFETY-OO 1, NWMJ Radioisotope Production Facility Preliminary Hazards Analysis, Rev. A, Northwest Medical Isotopes, Corvallis, Oregon, 2015.
NWMI-2015-SAFETY-004, Quantitative Risk Analysis of Process Upsets Associated with Passive Engineering Controls Leading to Criticality Accident Sequences, Rev. A, Northwest Medical Isotopes, Corvallis, Oregon, 2015.
Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C., February 1983.
WSRC-TR-93-262, Savannah River Site Generic Data Base Development, Rev. I, Westinghouse Savannah River Company, Savannah Ri ver Site, Aiken, South Carolina, May 1988.
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. NORTHWEST MEDICAL ISOTOPES Chapter 14.0 - Technical Specifications Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 1 August 2017 Prepared by:
Northwest Medical Isotopes, LLC 815 NW g th Ave , Suite 256 Corvallis, OR 97330
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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
Title:
Chapter 14.0 - Technical Specifications Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Sianature:
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I NWMl-2013-021 , Rev. 1 Chapter 14.0 - Technical Specifications REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not required Incorporate changes based on responses to 1 8/5/2017 C. Haass NRC Requests for Add itional Information
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~* * ~ NOflTHWlST MEDK:Al ISOT~S CONTENTS 14.0 TECHNICAL SPECIFICATIONS ............. .... ......................... .... ................................................ 14-1 14.1 Outline ... ....................................................................................... ..... ............ ................. .. 14-2 14.1. l Introduction ............................... ........................ ...... .......... ........ ................... ...... 14-2 14.1 .2 Safety Limit and Limiting Safety System Setting ... ....... ... ................... .............. 14-3 14.1.3 Limiting Condition of Operation ...... ...... ...... ........................... ......... ........... ...... 14-3 14.1 .4 Surveillance Requirements ................................... .............................................. 14-4 14.1.5 Design Features .................... ...... ................. ......... ........................... ................... 14-4 14.1 .6 Administrative Controls ......... ........................ ...... .... ............ ................ ...... ........ 14-4 14.2 References ................................................................................................. ....................... 14-5 TABLES Table 14-1. Potential Technical Specifications ........ ........................... ... ........ ..... .. ............... ............. 14-1 14-i
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. ~* * ~ . NORTHWUT MEOfCAL ISOTOPlS TERMS Acronyms and Abbreviations AC administrative control ANS American Nuclear Society 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 Radioisotope 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 This 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, samples, and Criticality control approved containers*
Spacing requirements on targets and containers Criticality control with SNM" Floor and sump designs* Criticality control Hot cell liquid confinement* Criticality control Process tank size and spacing* Criticality control Evaporator condensate monitor Criticality control Criticality monitoring system Criticality control In-line uranium content monitoring Criticality control Air pressure differential between zones* Control of airborne RAM Ventilation system filtration* Control of airborne RAM Process offgas subsystem Control of airborne RAM Primary offgas relief system Control of airborne RAM Hot cell shield thickness and integrity" Occupation and general public dose reduction Hot eel 1 secondary confinement boundary" Control of airborne RAM Double-wall piping Control of liquid RAM/criticality control Process closed heating and cooling loops Control of both airborne and liquid RAM System backflow prevention devices Control of liquid RAM/criticality control Stack height" Control of airborne RAM Area radiation monitoring system Occupation and general public dose reduction a Items that will significantly influence the final design.
RAM = radioactive material. SNM special nuclear material.
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- NORTHWEST Mf.DtCAl ISOTOPES 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 Development of Technical Specifications for Research Reactors; NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing 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 Federal 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 seems less likely than translating it into a LCO.
The outline for the technical specifications that will be prepared during development 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 ' Definition Actions Actions are that part of a limiting condition for operation that prescribes Required Actions to be taken under designated conditions within specified completion times .
Administrative .. . (described in Section 14.1.6) control (AC)
Design features . .. (described in Section 14.1.5)
Limiting condition ... (described in Section 14.1.3) for operation (LCO)
Limiting safety .. .(described in Section 14.1.2) system setting (LSSS) 14-2
NWMl-2013-021 , Rev. 1 Chapter 14.0 - Technical Specifications Term Definition Modes 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.
Operable/ A system, subsystem, component, or device shall be operable or have operability 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) .
Safety limit (SL) ... (described in Section 14.1.2)
Shall Denotes a mandatory requirement that must be complied with to maintain the requirements, assumptions, or conditions of the facility safety basis .
Surveillance ... (described in Section 14.1.4) requirements Verify/verification 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 process 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 nor 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.
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~ .* NORTHWEST MEDICAL ISOTOPES 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 arrangements. These design functions , if altered or modified, are implied to significantly affect safety and will not be identified in other sections. Anticipated areas covered in this section include the site and facility description, and fissionable material storage. Design features that will be provided in the technical specifications 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 lines 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 personnel
- 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 conditions 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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14.2 REFERENCES
10 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations , Office of the Federal Register, as amended.
10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register, as amended.
ANSI/ANS 15.1 , The Development ofTechnical Specifications for Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park Illinois, 2013.
NRC, 20 I 2, Final Interim Staff Guidance Augmenting NUREG-153 7, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors," Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors, Docket ID:
NRC-2011-0135 , U.S. Nuclear Regulatory Commission, Washington, D.C., October 30, 2012 .
NUREG-153 7 (Part 1), Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content, U.S. Nuclear Regulatory Commission, Washington, D.C.,
February 1996.
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