ML17193A428

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NWMI-2013-021, Revision 1, Construction Permit Application for Radioisotope Production Facility, Chapters 3.0, 6.0, 7.0, 8.0, 9.0 and 13.0, Attachment 3
ML17193A428
Person / Time
Site: Northwest Medical Isotopes
Issue date: 06/30/2017
From:
Northwest Medical Isotopes
To:
Office of Nuclear Reactor Regulation
Shared Package
ML17193A418 List:
References
NWMI-LTR-2017-007 NWMI-2013-021, Rev. 1
Download: ML17193A428 (243)
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!* * ~ NORTHWEST MEDICAL ISOTOPES ATTACHMENT 3 Northwest Medical Isotopes, LLC Docket No. 50-609 Construction Permit Application for Radioisotope Production Facility Chapters 3.0, 6.0, 7.0, 8.0, 9.0, and 13.0 (Document No. NWMl-2013-021, Rev. 1, June 2017)

Public Version Information is being provided via hard copy

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. *. * . * . NORTHWEST MEDICAL ISOTOPES Chapter 3.0 - Design of Structures, Systems, and Components Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 June 2017 Prepared by:

Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, Oregon 97330

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  • NORTHWEST MEDtcAL ISOTOPES Chapter 3.0 - Design of Structures, Systems, and Components Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1

Title:

Chapter 3.0 - Design of Structures, Systems and Components Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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  • NOITHWUTMEDICAllSOTOHS REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not required 1 6/26/2017 Incorporate changes based on responses to C. Haass NRC Requests for Additional Information

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  • ~ ~.* ~ * . NOITHWEST MlDICAl ISOTOPCS Chapter 3.0 - Design of Structures, Systems and Components NWMl-2013-021 , Rev. 1 CONTENTS 3.0 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ........ ........... ........ ......... .......... 3-1 3.1 Design Criteria ... ... ........... ...... ...... .......... ..... .... ....... .... ...... .... ... ......... ... ...... ......... ....... .... ... .. 3-4
3. l.1 Radioisotope Production Facility Structures, Systems, and Components ............ 3-4 3.1 .2 Code of Federal Regulations ................................................................................ 3-8 3.1.3 U.S. Nuclear Regulatory Commission ...... ... ......... ...... ... ... ... ...... ... .......... ............. 3-8 3.1.4 Other Federal Regulations, Guidelines, and Standards ...................................... 3-10 3.1.5 Local Government Documents .................................................. ... ..................... 3-10 3.1.6 Discovery Ridge/University of Missouri ........ ... ........ .. ......... ... ... ............. ... ... .. .. 3-11 3.1. 7 Codes and Standards .. .. .................. ............ ......................................... ............ ... 3-12 3.2 Meteorological Damage ............................ ......... ............................................ .................. 3-24 3.2.1 Combinations of Loads ................ ........................................... ........................... 3-25 3.2.1.1 Nuclear Safety-Related Structures, Systems, and Components ...... .. 3-26 3 .2.1 .2 Commercial and Nuclear Non-Safety-Related Structures, Systems, and Components ............. .................... ................... ............ 3-26 3 .2.2 Combinations for Serviceability Based Acceptance Criteria ....... ...................... 3-27 3.2.3 Normal Loads ....................................... .. .............................................. ...... ..... ... 3-27 3.2.4 Wind Loading ....... ....... .................. ....... ........................................................ ..... 3-30 3.2.4.1 Wind Load ......... ......... ... ...... ......... ...... .......... .. ...... ............................. 3-30 3.2.4.2 Tornado Loading ................. ... .......... .. .... ......... .... ......... ...... ........... .... 3-30 3.2.4.3 Effect of Failure of Structures, Systems, or Components Not Designed for Tornado Loads ... ......... ..... .............................................. 3-32 3.2.5 Rain, Snow, and Ice Loading ... ..... ....................................................... ... ........... 3-32 3.2.5.1 Rain Loads .............................. ............... ... ... .... .... .... .. ....... ................ 3-32 3.2.5.2 Snow Load .............................. .......... .... .......... ........ .. ..................... ... 3-33 3.2.5.3 Atmospheric Ice Load ...................... ..... ... ..... ............ ... ......... ......... .. . 3-34 3.2.6 Operating Thermal/Self-Straining Loads ..... ......................................... ............. 3-34 3.2.7 Operating Pipe Reaction Loads ........... ....... ..................... ... ...... .............. ... ......... 3-34 3.2.8 External Hazards ................... ............... ......... ...... ......... ............................ .......... 3-34 3.3 Water Damage ......... .. ............ ........................... ....... ......... ... ... ................. .. ... ...... .............. 3-35 3.3.1 Flood Protection ........................................................ ... ... .. .... ............ .......... .. ..... 3-35 3.3.1.1 Flood Protection Measures for Structures, Systems, and Components ......................... ..... ..................... ...... ..... ....................... .. 3-35 3.3.1.2 Flood Protection from External Sources .... ... ............. .. ........ ........ ...... 3-36 3.3. 1.3 Compartment Flooding from Fire Protection Discharge ... ..... ...... ...... 3-37
3. 3.1 .4 Compartment Flooding from Postulated Component Failures .......... . 3-3 7 3.3.1.5 Permanent Dewatering System .. ......... .. ............... .... ............. .... .... ..... 3-37 3.3.1.6 Structural Design for Flooding .................................. ... ........ ............. 3-37 3.4 Seismic Damage ........ ..... ........................................ .... ... .... ..... ..... ........ ........... ............. ..... 3-38 3.4.1 Seismic Input. .......... ........................ ... ...... ..... ...................... ............... .. .. ....... ..... 3-38 3.4.1.1 Design Response Spectra .......................................... ...................... .. 3-38 3.4.1.2 Method of Analysis .......... ................................... ... ... ....................... . 3-39 3 .4.2 Seismic Qualification of Subsystems and Equipment... ......... ........................ .... 3-40 3.4.2. 1 Qualification by Analysis .. ..... .................. ....................... .. ................ 3-40 3.4.2.2 Qualification by Testing ........... ....................... ........... ............... ........ 3-41 3.4.3 Seismic Instrumentation .................... ................... .... ............................ .... ......... . 3-41 3.4.3.1 Location and Description ... ................. ... ... ... .......... ...... ... ............ ...... 3-42 3.4.3.2 Operability and Characteristics ....... ... ................................. .............. 3-42 3-i

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' ~ ~-~~ ; . NORTKWUlMEDIW ISOTOPES Chapter 3.0 - Design of Structures, Systems and Components NWMl-2013-021 , Rev. 1 3.5 Systems and Components ............. ... ........................................................... .... .................. 3-43 3.5.1 General Design Basis Information .. ............. .......... ............ .. ... ... ... ... ... ..... ..... ...... 3-43 3.5.1.1 Classification of Systems and Components Important to Safety .... .... 3-43 3.5.1.2 Classification Definitions ......... ............. ......... ... ..... .. ... ......... .... ......... 3-43 3.5.1.3 Nuclear Safety Classifications for Structures, Systems, and Components ............................. ................................... ... ........... .. ....... 3-44 3.5.2 Radioisotope Production Facility ............. ................. ........ ........ ..... .... ................ 3-47 3.5.2. 1 System Classification ........... ............... ........ ........... ........................... 3-52 3.5.2.2 Classification of Systems and Components Important to Safety .... .. 3-52 3.5.2.3 Design Basis Functions, Values, and Criteria ........ ............. ..... ....... .. 3-54 3.5.2.4 System Functions/Safety Functions ....... .......................... ....... ....... ... 3-54 3.5.2.5 Systems and Components ....... ...... .... .............................. .... ..... ......... 3-54 3.5.2.6 Qualification Methods ..... ....... ...................................... .. ................... 3-55 3.5.2. 7 Radioisotope Production Facility Specific System Design Basis Functions and Values ......... ........ ....... ...... .. ........... ..... ... ............. ........ 3-55 3.6 References .............................................................. ................................ .... ...................... 3-66 3-ii

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  • NOITNWEST llEDICAl tSGTOPCS Chapter 3.0 - Design of Structures, Systems and Components NWMl-2013-021 , Rev. 1 TABLES Table 3-1. List of System and Associated Systems and Construction Permit Application Crosswalk (2 pages) ..... .... ............. ..... ... ........... .. ..... .. .. ......... .. ........... .. ....... ........ .............. 3-4 Table 3-2. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages) ... ....... ...... .......... .......................... ....... .. ..... .. ..... .... ..... ... .... .... ..... ....... ... ... ...... ...... 3-5 Table 3-3 . Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) ... .. ... ... .... .............. . 3-8 Table 3-4. Other Federal Regulations, Guidelines, and Standards ..... ... ... ........ .. ........ .. ....... ....... ..... 3-10 Table 3-5. Local Government Documents (2 pages) ....... ....... ......... .. ...... ........................ .... ............ 3-1 0 Table 3-6. Discovery Ridge/University of Missouri Requirements .. .... .. ... .. .... .... ... ........ ..... ........... 3-11 Table 3-7. Design Codes and Standards (12 pages) ... .... ....... ......... .. .. .. ...... ..... ... ..... ..... ... ............ .... 3-12 Table 3-8. Load Symbol Definitions (2 pages) ........ ...... ....... ....... ..... ...... ........ .. ............. ............... .. 3-24 Table 3-9. Load Combinations for Strength Based Acceptance Criteria, Nuclear Safety-Related ............ ..... ............. ..... ... ..... ......... .................. .......... ... ....... ... ... .......... ...... .... ....... 3-26 Table 3-10. Load Combinations for Strength Base Acceptance Criteria, Commercial ... ...... .... .... ... 3-27 Table 3-11. Load Combinations for Serviceability Based Acceptance Criteria .......... ..... .... .. .. .... ..... 3-27 Table 3-12. Lateral Earth Pressure Loads .. ........ .... ........ ......... ....... ..... .... ...... ............... .... .. ............... 3-28 Table 3-13. Floor Live Loads .................... .. ........... ............... ....... ..... ...... ......... ...... ... .. .... .. ........... ..... 3-29 Table 3-14. Crane Load Criteria ............................. ...................... .... .... .. ... ...... .... ..... ..... .............. ..... 3-29 Table 3-15. Wind Loading Criteria ....... .. ............ .... .... ........ ........... ... ...... .................... ...... ................ 3-30 Table 3-16. Design-Basis Tornado Field Characteri stics .. .... ............ ...... .. ... .... .... ... ..... .. ....... ....... .. ... 3-31 Table 3-17. Design-Basis Tornado Missile Spectrum ..... .. ... .. ...... ... ..... ...... .... ...... ... .. ...... ........ .......... 3-32 Table 3-18. Rain Load Criteria ... .......... ................ ...... .... .............. .. ... .. ...... .... ..... ..... ..... ........ .... ....... . 3-33 Table 3-19. Snow Load Criteria ..... ............... ....................... ........... ...... ..... ........ .. .... ...... .......... ..... .... 3-33 Table 3-20. Extreme Winter Precipitation Load Criteria ..... ..... ....... ........ ..... ...................... .............. 3-34 Table 3-21. Atmospheric Ice Load Criteria ..... .... .. ...... .... ..... ....... ...... .... ..... ... ..... ........ ..... .................. 3-34 Table 3-22. Design Criteria Requirements (4 pages) ........... ............................ ..... ... .................. ....... 3-4 7 Table 3-23 . System Classifications .... ........ ...... ............ .... ..... ......... .. .... .... .. ....... ... .. ........................... 3-52 Table 3-24. System Safety and Seismic Classification and Associated Quality Level Group (2 pages) ....... ... ... .. .. ............ .. .. ........ .... ........ ... .. ... ........ ... ..... .... .. ..... .. .. ...... ..... ....... ........ ... 3-52 Table 3-25. Likelihood Index Limit Guidelines ......... .. ........ ...... ... ... .. ..... ... ......... ..... ... ...... ......... ....... 3-53 3-iii

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~~~.:!** NOATHWfSTMEDtCAllSOTOP£S TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 AASHTO American Association of State Highway and Transportation Officials ACGIH American Conference on Governmental Industrial Hygienists ACI American Concrete Institute AHRI Air Conditioning, Heating and Refrigeration Institute AISC American Institute of Steel Construction ALARA as low as reasonably achievable AMCA Air Movement and Control Association ANS American Nuclear Society ANSI American National Standards Institute ASCE American Society of Civil Engineers ASHRAE American Society of Heating, Refrigeration, and Air-Conditioning Engineers ASME American Society of Mechanical Engineers ASNT American Society for Nondestructive Testing ASTM American Society for Testing and Materials AWS American Welding Society BMS building management system CDC Centers for Disease Control and Prevention CFR Code of Federal Regulations CRR Collected Rules and Regulations CSR Missouri Code of State Regulations Discovery Ridge Discovery Ridge Research Park DBE design basis event DBEQ design basis earthquake DOE U.S. Department of Energy EIA Electronic Industries Alliance ESF engineered safety feature FEMA Federal Emergency Management Agency FPC facility process control FSAR final safety analysis report H2 hydrogen gas HR hydrometeorological report HY AC heating, ventilation, and air conditioning I&C instrumentation and control IAEA International Atomic Energy Agency IBC International Building Code ICC International Code Council ICC-ES International Code Council Evaluation Service IEEE Institute of Electrical and Electronics Engineers IES Illuminating Engineering Society IFC International Fire Code IROFS items relied on for safety ISA International Society of Automation ISG Interim Staff Guidance IX ion exchange LEU low enriched uranium MDNR Missouri Department of Natural Resources Mo molybdenum 3-iv

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0 llOITNW£ST ~ tsOTOPU NWMl-2013-021 , Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components MO DOT Missouri Department of Transportation MRI mean recurrence interval MU University of Missouri NECA National Electrical Contractors Association NEMA National Electrical Manufacturers Association NEP normal electrical power NESHAP National Emissions Standards for Hazardous Air Pollutants NETA InterNational Electrical Testing Association NFPA National Fire Protection Association NIOSH National Institute for Occupational Safety and Health NOAA National Oceanic and Atmospheric Administration NRC U.S. Nuclear Regulatory Commission NS non-seismic NSR non-safety-related NWMI Northwest Medical Isotopes, LLC NWS National Weather Service PMF probable maximum flood PMP probable maximum precipitation PMWP probable maximum winter precipitation QA quality assurance QA PP quality assurance program plan RCA radiologically controlled area RPF Radioisotope Production Facility SEP standby electrical power SMACNA Sheet Metal and Air Conditioning Contractors National Association SNM special nuclear material SR safety related SSC structures, systems and components TIA Telecommunications Industry Association U.S . United States UL Underwriters Laboratory UPS uninterruptible power supply USGS U.S. Geological Survey Units oc degrees Celsius Of degrees Fahrenheit

µ micron cm centimeter cm 2 square centimeters ft feet ft2 square feet ft3 cubic feet g acceleration of gravity gal gallon hp horsepower hr hour

m. inch in.2 square inch kg kilogram kip thousand pounds-force 3-v
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. *~ ~**.! : NORTHWEST MEDfCAl ISOTOP(S NWMl -2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.0 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS This chapter identifies and describes the principal architectural and engineering design criteria for the facility structures, systems and components (SSC) for the Northwest Medical Isotopes, LLC (NWMI)

Radioisotope Production Facility (RPF). The information presented emphasizes the safety and protective functions and related design features that help provide defense-in-depth against the uncontrolled release of radioactive material to the environment. The bases for the design criteria for some of the systems discussed in this chapter are developed in other chapters of the Construction Permit Application and are appropriately cross-referenced, when required.

NWMI's RPF design is based on applicable standards, guides, codes, and criteria and provides reasonable assurance that the RPF SSCs, including electromechanical systems, are :

  • Built and will function as designed and required by the analyses in Chapter 13. 0, "Accident Analysis"
  • Built to have acceptable protection of the public health and safety and environment from radiological risks (e.g., radioactive materials, exposure) resulting from operations
  • Protected against potential meteorological damage
  • Protected against potential hydrological (water) damage
  • Protected against seismic damage
  • Provided surveillance activities and technical specifications required to respond to or mitigate consequences of seismic damage
  • Based on technical specifications developed to ensure that safety-related functions of electromechanical systems and components will be operable and protect the health and safety of workers, the public, and environment The design of the RPF and SSCs are based on defense-in-depth practices.

The NRC defines design-in-depth as the following:

An approach to designing and operating nuclear facilities that prevents and mitigates accidents that release radiation or hazardous materials. The key is creating multiple independent and redundant layers of def ense to compensate for potential human and mechanical failures so that no single layer, no matter how robust, is exclusively relied upon.

Defense in depth includes the use of access controls, physical barriers, redundant and diverse key safety functions, and emergency response measures.

Defense-in-depth is a design philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety are not wholly dependent on any single element of the design, construction, maintenance, or operation of the facility.

The net effect of incorporating defense-in-depth practices is a conservatively designed faci lity and systems that exhibit higher tolerances to failures and external challenges. The risk insights obtained through performance of accident analysis can then be used to supplement the final design by focusing attention on the prevention and mitigation of the higher risk potential accidents.

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  • ~ *.~ ~ * . NORTMWEST MUHCAL lSOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components This application to the U.S. Nuclear Regulatory Commission (NRC) seeks to obtain a license for a production facility under Title l 0, Code of Federal Regulations (CFR), Part 50 (10 CFR 50), "Domestic Licensing of Production and Utilization Facilities." Embedded in the 10 CFR 50-licensed facility will be several activities subject to 10 CFR 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, use, and transfer special nuclear material (SNM) and 10 CFR 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material ," to process and transport molybdenum-99 (99 Mo) for medical applications.

This 10 CFR 50 license application for the RPF follows the guidance in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non Power Reactors - Format and Content, that encompasses activities regulated under different NRC requirements (e.g., 10 CFR 70 and I 0 CFR 30), in accordance with 10 CFR 50.31 , "Combining Applications," and 10 CFR 50.32, "Elimination of Repetition."

The NRC has determined that a radioisotope separation and processing facility, which also conducts separation of SNM, will be considered a production facility and as such, will be subject to licensing under 10 CFR 50. The operation of the NWMI RPF will primarily be focused on the disassembly of irradiated low-enriched uranium (LEU) targets, separation and purification of fi ssion product 99 Mo, and the recycle of LEU that is licensed under I 0 CFR 50.

RPF operations will also include the fabrication of LEU targets, which will be licensed under 10 CFR 70.

These targets will be shipped to NWMI's network of research or test reactors for irradiation (considered a connected action) and returned to the RPF for processing. The LEU used for the production of LEU target materials will be obtained from the U.S. Department of Energy (DOE) and from LEU reclaimed from processing the irradiated targets.

NWMI's licensing approach for the RPF defines the following unit processes and areas that fall under the following NRC regulations:

  • 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities" Target receipt and disassembly system Target dissolution system Molybdenum (Mo) recovery and purification system Uranium recovery and recycle system Waste management system Associated laboratory and support areas
  • 10 CFR 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material" Any byproduct materials produced or extracted in the RPF Design information for the complete range of normal operating conditions for various facility systems is provided throughout the Construction Permit Application, and includes the following.
  • RPF-specific design criteria (e.g., codes and standards, NRC guidelines) for SSCs are provided in Sections 3.1.
  • NRC general design criteria and associated applicability to the RPF SSCs are addressed in Section 3.5.
  • RPF description is presented in Chapter 4.0, "Radioisotope Production Facility Description."

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  • Postulated initiating events and credible accidents that form the design basis for the SSCs are discussed in Chapter 13 .0.
  • Potential hazards and credible accidents that could be encountered in the RPF during operations involving SNM, irradiated and unirrradiated, Mo recovery and purification, uranium recovery and recycle, waste management, and/or the use of hazardous chemicals relative to these radiochemical processes that form the bases for the SSCs located in the RPF, are discussed in Chapter 13.0.
  • Design redundancy of SSCs to protect against unsafe conditions with respect to single failures of engineered safety features (ESF) and control systems are described in Chapter 6.0, "Engineered Safety Features," and Chapter 7.0, "Instrumentation and Control System," respectively.
  • ESFs are described in Chapter 6.0, and the administrative controls are discussed in Chapter 14.0, "Technical Specifications."
  • Quality standards commensurate with the safety functions and potential risks that were used in the design of the SSCs are described in Table 3-7 (Section 3.1.7).
  • Hydrological design bases describing the most severe predicted hydrological events during the life of the facility are provided in Chapter 2.0, "Site Characteristics, Section 2.4.
  • Design criteria for facility SSCs to withstand the most severe predicted hydrological events during the lifetime of the facility are provided in Section 3.3 .
  • Seismic design bases for the facility are provided in Chapter 2.0, Section 2.5. Seismic design criteria for the facility SSCs are provided in Section 3.4.
  • Analyses concerning function, reliability, and maintainability of SSCs are described throughout the Construction Permit Application .
  • Meteorological design bases describing the most severe weather extremes predicted to occur during the life of the faci lity are provided in Chapter 2.0, Section 2.3 . Design criteria for faci lity SSCs to withstand the most severe weather extremes predicted to occur during the life of the facility are provided in Section 3.2.
  • Potential conditions or other items that wi ll be probable subjects of technical specifications associated with the RPF structures and design features are discussed in Chapter 14.0.

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  • NOkTNWlST lllDICAl tsOTDIO 3.1 DESIGN CRITERIA Section 3.1 describes the design criteria applied to the RPF and SSCs within the facility. The principal design criteria for a production facility establish the necessary design, fabrication , construction, testing, and performance requirements for SSCs important to safety (i.e., those that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of workers and the public).

The systems associated with the RPF are identified. Those items relied on for safety (IROFS) are identified in Chapters 6.0 and 13.0. Requirements are derived from:

  • Code of Federal Regulations
  • U.S. Nuclear Regulatory Commission
  • Federal regulations, guidelines, and standards
  • Local government regulations and requirements
  • Discovery Ridge Research Park (Discovery Ridge) covenants
  • University of Missouri System (MU) requirements
  • Other codes and standards 3.1.1 Radioisotope Production Facility Structures, Systems, and Components Table 3-1 lists the RPF systems and identifies the RPF material accountability area and the Construction Permit Application reference chapter that provides the associated detailed system descriptions.

Table 3-1. List of System and Associated Systems and Construction Permit Application Crosswalk (2 pages)

Construction Permit Application reference Primary structure and associated systems . (primary references)

Radioisotope Production Facility (RPF - primary structure) 10 CFR 70" Target fabrication Chapter 4.0, Sections 4.1.3.1 and 4.4 IO CFR sob Target receipt and disassembly Chapter 4.0, Section 4.1.3.2, 4.3.2, and 4.3.3 Target dissolution Chapter 4.0, Sections 4.1 .3.3 and 4.3.4 Molybdenum recovery and purification Chapter 4.0, Sections 4.1.3.4 and 4.3.5 Uranium recovery and recycle Chapter 4.0, Sections 4.1.3.5 and 4.3.6 Waste handling Chapter 4.0, Section 4.1.3.6; Chapter 9.0, Section 9. 7.2 Criticality accident alarm Chapter 6.0, Section 6.3 .3.1; Chapter 7.0, Section 7.3 .7 Radiation monitoring Chapter 7 .0, Section 7 .6; Chapter 11.0, Section 11.1.4 Normal electrical power Chapter 8.0, Section 8.1 Standby electrical power Chapter 8.0, Section 8.2 Process vessel ventilation Chapter 9.0, Section 9.1 Facility ventilation Chapter 9.0, Section 9.1 Fire protection Chapter 9.0, Section 9.3 Plant and instrument air Chapter 9.0, Section 9.7.1 Emergency purge gas Chapter 6.0, Section 6.2.1.7 .5 Gas supply Chapter 9.0, Section 9. 7.1 3-4

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  • NOR'THWHT MEDtcAL ISOTOPES Chapter 3.0 - Design of Structures, Systems and Components Table 3-1. List of System and Associated Systems and Construction Permit Application Crosswalk (2 pages)

Construction Permit Application reference Primary structure and associated systems (primary references)

Process chilled water Chapter 9.0, Section 9.7.1 Facility chilled water Chapter 9.0, Section 9.7. l Facility heated water Chapter 9 .0, Section 9. 7. I Process stream Chapter 9.0, Section 9.7.l Demineralized water Chapter9 .0, Section 9.7.1 Chemical supply Chapter 9.0, Section 9.7.4 Biological shield Chapter 4.0, Section 4.2 Facility process control Chapter 7.0, Section 7.2.3

b 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register, as amended.

In addition to Table 3-2, NWMI-2015-LIST-003, NWMI Radioisotop e Production Facility Master Equipment List, provides a summary of the RPF systems, components, and equipment used in the RPF design.

Table 3-2 provides a summary of the IROFSs identified by the accident analyses in Chapter 13 .0, and a crosswalk to where the IROFSs are described in the Construction Permit Application. Chapter 13.0 also provides the associated detailed descriptions. Table 3-2 also identifies whether the IROFS are considered ESFs or administrative controls. Additional IROFS may be identified (or the current IROFS modified) during the RPF final design and development of the Operating License Application.

Table 3-2. Summary ofltems Relied on for Safety Identified by Accident Analyses (3 pages)

IROFS Construction Permit Application designator Descriptor ESF AC crosswalk (primary references)

RS-01 Hot cell liquid confinement boundary Chapter 6.0, Sections 6.2.1.1 - 6.2.1 .6 Chapter 13.0, Section 13 .2.2.8 RS-02 Reserved*

RS-03 Hot cell secondary confinement boundary Chapter 6.0, Sections 6.2.1.1 - 6.2.1.6 Chapter 13.0, Sections 13.2.2.8, 13.2.3.8 RS-04 Hot cell shielding boundary Chapter 6.0, Sections 6.2.1.1 - 6.2.1.6 Chapter 13 .0, Sections 13 .2.2.8, 13 .2.4.8 RS-05 Reserved*

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RS-06 Reserved*

RS-07 Reserved*

RS-08 Sample and analysis of low-dose waste tank Chapter 13.0, Section 13.2.7.1 dose rate prior to transfer outside the hot cell shielded boundary RS-09 Primary offgas relief system Chapter 6.0, Section 6.2.1.7 Chapter 13 .0, Section 13 .2.3.8 3-5

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  • NORTHWEST MEOM:Al tsOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components Table 3-2. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages)

IROFS Construction Permit Application designator Descriptor ESF AC crosswalk (primary references)

RS-10 Active radiation monitoring and isolation of ./ Chapter 6.0, Section 6.2. I. 7 low-dose waste transfer Chapter 13.0, Section 13.2.7.1 RS-1 I Reserved" RS-12 Cask containment sampling prior to closure ./ Chapter 13.0, Section 13.2.7.l lid removal RS-13 Cask local ventilation during closure lid ./ Chapter 6.0, Section 6.2.1.7 removal and docking preparations Chapter 13 .0, Section 13 .2.7.1 RS-14 Reserved" RS-15 Cask docking port enabling sensor Chapter 6.0, Section 6.2.1.7 Chapter 13.0, Section 13 .2.7.1 CS-01 Reserved" CS-02 Mass and batch handling limits for uranium Chapter 13.0, Section 13.2.7.2 metal , uranium oxides, targets, and laboratory sample outside process systems CS-03 Interaction control spacing provided by ./ Chapter 13.0, Section 13.2.7.2 administrative control CS-04 Interaction control spacing provided by ./ Chapter 6.0, Section 6.3.1.2 passively designed fixture s and workstation Chapter 13 .0, Section 13.2.7 .2 placement CS-05 Container batch volume limit ./ Chapter 13.0, Section 13.2.7.2 CS-06 Pencil tank, vessel, or piping safe geometry ./ Chapter 6.0, Section 6.3 .1.2 confinement using the diameter of tanks, Chapter 13 .0, Section 13 .2.4.8 vessels, or piping CS-07 Pencil tank and vessel spacing control using ./ Chapter 6.0, Section 6.3. I .2 fixed interaction spacing of individual tanks Chapter 13.0, Section 13.2.2.8 or vessels CS-08 Floor and sump geometry control of slab ./ Chapter 6.0, Section 6.3.1.2 depth, sump diameter or depth for floor spill Chapter 13 .0, Section 13.2.2.8 containment berms CS-09 Double-wall piping ./ Chapter 6.0, Section 6.2. l. 7 Chapter 13 .0, Section 13.2.2.8 CS-10 Closed safe geometry heating or cooling loop ./ Chapter 6.0, Section 6.3.1.2 with monitoring and alarm Chapter 13 .0, Section 13.2.4.8 CS-I I Simple overflow to normally empty safe ./ Chapter 6.0, Section 6.3. I .2 geometry tank with level alarm Chapter 13.0, Section 13.2.7.2 CS-12 Condensing pot or seal pot in ventil ation vent ./ Chapter 6.0, Section 6.3.1 .2 line Chapter 13.0, Section 13 .2.7.2 CS-13 Simple overflow to normally empty safe ./ Chapter 6.0, Section 6.3 . l.2 geometry floor with level alarm in the hot cell Chapter 13.0, Section 13.2.7.2 containment boundary CS -14 Active di scharge monitoring and isolation ./ Chapter 6.0, Section 6.3.1.2 Chapter 13.0, Section 13.2.7.2 3-6

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  • NORTlfWESl MEDICAi. lSOTOP£S Table 3-2. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages)

IROFS Construction Permit Application designator Descriptor ESF AC crosswalk (primary references)

CS-1 5 Independent active discharge monitoring and ,/ Chapter 6.0, Section 6.3.1.2 isolation Chapter 13.0, Section 13.2.7.2 CS-1 6 Sampling and analysis of uranium mass or ,/ Chapter 13.0, Section 13.2.7.2 concentration prior to di scharge or disposal CS- 17 Independent sampling and analysis of ,/ Chapter 13.0, Section 13.2.7.2 uranium concentration prior to discharge or disposal CS- 18 Backflow preventi on device ,/ Chapter 6.0, Sections 6.2. 1.7 and 6.3.1.2 Chapter 13.0, Section 13.2.4.8 CS-1 9 Safe-geometry day tanks ,/ Chapter 6.0, Section 6.3.1.2 Chapter 13.0, Section 13.2.4.8 CS-20 Evaporator or concentrator condensate ,/ Chapter 6.0, Section 6.3 .1.2 monitoring Chapter 13.0, Section 13.2.4.8 CS-2 1 Visual inspection of accessible surfaces for ,/ Chapter 13 .0, Section 13.2.7.2 foreign debris CS-22 Gram estimator survey of accessible surfaces ,/ Chapter 13.0, Section 13.2.7.2 for gamma activity CS -23 Nondestructive assay of items with ,/ Chapter 13.0, Section 13 .2.7.2 inaccessible surfaces CS -24 Independent nondestructive assay of items ,/ Chapter 13.0, Section 13.2.7.2 with inaccessible surfaces CS -25 Target housing weighi ng prior to disposal ,/ Chapter 13.0, Section 13.2.7.2 CS -26 Processing component safe volume ,/ Chapter 6.0, Section 6.3 .1.2 confinement Chapter 13.0, Section 13.2. 7.2 CS -27 Closed beating or cooling loop with ,/ Chapter 6.0, Section 6.3 .1.2 monitoring and alarm Chapter 13.0, Section 13 .2.4.8 FS-0 1 Enhanced lift procedure ,/ Chapter 13.0, Section 13 .2.2.8 and 13.2.7.1 FS-02 Overhead cranes ,/ Chapter 13.0, Section 13 .2.7.3 FS-03 Process vessel emergency purge system ,/ Chapter 6. 0, Section 6. 2. 1.7 Chapter 13.0, Section 13.2.7.3 FS-04 Irradiated target cask lifting fixture ,/ Chapter 6.0, Section 6.2.1.7 Chapter 13.0, Section 13.2.6.5 FS-05 Exhaust stack height ,/ Chapter 6.0, Section 6.2.1. 7 Chapter 13.0, Section 13.2.7.3

  • Reserved - IROFS designator currently unassigned.

AC administrative control. IROFS items relied on for safety.

ESF = engineered safety feature.

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  • NDmlWESTMEDICALISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.1.2 Code of Federal Regulations NWMI-DRD-2013-030, NWMI Radioisotope Production Facility Design Requirements Document, summarizes the CFR design inputs (in whole or in part) for the RPF, which include the following:
  • 10 CFR 20, "Standards for Protection Against Radiation"
  • 10 CFR 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material"
  • 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities"
  • 10 CFR 71 , "Energy: Packaging and Transportation of Radioactive Material "
  • 21 CFR 210, "Current Good Manufacturing Practice in Manufacturing, Processing, Packaging, or Holding of Drugs'
  • 21 CFR 211 , "Current Good Manufacturing Practice for Finished Pharmaceuticals"
  • 29 CFR 1910, "Occupational Safety and Health Standards"
  • 40 CFR 61 , "National Emissions Standards for Hazardous Air Pollutants (NESHAP)"
  • 40 CFR 63 , "NESHAP for Source Categories"
  • 40 CFR 141, "National Primary Drinking Water Regulations" 3.1.3 U.S. Nuclear Regulatory Commission Table 3-3 lists the NRC design inputs for the RPF identified in NWMI-DRD-2013-030. The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference.

Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages)

CF Ra Title Docket Number: Final Interim Staff Guidance Augmenting NUREG-153 7, "Guidelines for Preparing and NRC-2011-0135 Reviewing Applications for the Licensing ofNon-Power Reactors," Parts 1 and 2, for (NRC, 2012) Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors NRC Regulatory Guides - Power Reactors (Division 1)

Regulatory Guide I .53 Application of the Single-Failure Criterion to Safety Systems , 2003 (R201 I)

Regulatory Guide 1.60 Design Response Spectra for Seismic Design of Nuclear Power Plants, 2014 Regulatory Guide 1.76 Design Basis Tornado and Tornado Missiles fo r Nuclear Power Plants , 2007 Regulatory Guide 1.97 Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, 2006 (R2013)

Regulatory Guide I. I 00 Seismic Qualification ofElectrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants, 2009 Regulatory Guide 1.152 Criteria for Use of Computers in Safety Systems of Nuclear Power Plants , 201 l Regulatory Guide 1.166 Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post Earthquake Actions, 1997 Regulatory Guide 1.167 Restart of a Nuclear Power Plant Shut down by a Seismic Event, 1997 Regulatory Guide I .208 Performance Based Approach to Defin e the Site-Specific Earthquake Ground Motion , 2007 NRC Regulatory Guides - Fuels And Materials Facilities (Division 3)

Regulatory Guide 3 .3 Quality Assurance Program Requirements for Fuel Reprocessing Plants and for Plutonium Processing and Fuel Fabrication Plants , 1974 (R2013) 3-8

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  • ~~.~~ : . NORTKW(STMEDtcAl.ISOTOPES Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages)

Title Regulatory Guide 3.6 Content of Technical Specification for Fuel Reprocessing Plants, l 973 (R2013)

Regulatory Guide 3. l 0 Liquid Waste Treatment System Design Guide for Plutonium Processing and Fuel Fabrication Plants, l 973 (R2013)

Regulatory Guide 3.18 Confinement Barriers and Systems for Fuel Reprocessing Plants, 1974 (R2013)

Regulatory Guide 3.20 Process Offgas Systems for Fuel Reprocessing Plants, 1974 (R20l3)

Regulatory Guide 3.71 Nuclear Criticality Safety Standards for Fuels and Materials Facilities, 2010 NRC Regulatory Guides - Materials and Plant Protection (Division 5)

Regulatory Guide 5.7 Entry/Exit Control for Protected Areas, Vital Areas, and Material Access Areas, May 1980 (R20l0)

Regulatory Guide 5.12 General Use of Locks in the Protection and Control of Facilities and Special Nuclear Materials, 1973 (R20l0)

Regulatory Guide 5.27 Special Nuclear Material Doorway Monitors, 1974 Regulatory Guide 5.44 Perimeter Intrusion Alarm Systems, 1997 (R2010)

Regulatory Guide 5.57 Shipping and Receiving Control of Strategic Special Nuclear Material, 1980 Regulatory Guide 5.65 Vital Area Access Control, Protection of Physical Security Equipment, and Key and Lock Controls, 1986 (R2010)

Regulatory Guide 5.71 Cyber Security Programs for Nuclear Facilities, 2010 NUREG-0700, Human-System Interface Design Review Guidelines NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWREdition Section 2.3.1 "Regional Climatology," Rev. 3, March 2007 Section 2.3 .2 "Local Climatology," Rev. 3, March 2007 Section 3.3. l "Wind Loading," Rev. 3, March 2007 Section 3.3 .2 "Tornado Loading," Rev. 3, March 2007 Section 3.7.1 "Seismic Design Parameters," March 2007 Section 3.7.2 "Seismic System Analysis," Rev. 4, September 2013 Section 3.7 .3 "Seismic Subsystem Analysis," Rev . 4, September 2013 NUREG-1513, Integrated Safety Analysis Guidance Document NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility Part 3, Appendix D "Natural Hazard Phenomena" NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors

- Format and Content, Part 1 NUREG/CR-4604, Statistical Methods for Nuclear Material Management NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook Process hazard analysis "Development of Quantitative Risk Analyses" NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems -

Final Report NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology a Complete references are provided in Section 3.6.

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  • * ~ ~.* ~ : . NCMTNWUTM£DICAL ISOTOf'lS NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.1.4 Other Federal Regulations, Guidelines, and Standards Table 3-4 lists other Federal design inputs for the RPF (NWMl-DRD-2013-030). The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference.

Table 3-4. Other Federal Regulations, Guidelines, and Standards Referencea Title Federal Emergency Management Agency (FEMA)

NI A "National Flood Insurance Program, Flood Insurance Rate Map , Boone County, Missouri and Incorporated Areas" National Oceanic and Atmospheric Administration (NOAA)

Hydrometeorological Probable Maximum Precipitation Estimates, United States East of the 105th Meridian Report No. 51 Hydrometeorological Application of Probable Maximum Precipitation Estimates, United States East of the 105th Report No. 52 Meridian Hydrometeorological Seasonal Variation of JO-Square-Mile Probable Maximum Precipitation Estimates, United Report No . 53 States East of the 105th Meridian U.S. Geological Survey (USGS)

NIA "2008 U.S. Geological Survey National Seismic Hazard Maps" Open-File Report Documentation for the 2008 Update of the United States National Seismic Hazard Maps 2008-1128 Centers for Disease Control and Prevention (CDC)

NIOSH 2003-136 Guidance for Filtration and Air-Cleaning Systems to Protect Building Environments from Airborne Chemical, Biological, and Radiological Attacks

NIOSH National Institute for Occupational Safety and USGS U.S. Geological Survey.

Health.

3.1.5 Local Government Documents Table 3-5 lists the design inputs for the RPF from the State of Missouri, City of Columbia, and Boone County government sources (NWMI-DRD-2013 -030). The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference.

Table 3-5. Local Government Documents (2 pages)

Referencea Title Missouri Code of State Regulations (CSR), Title 10 10 CSR 10-6.01 Ambient Air Quality Standards Missouri CSR, Title 20 20 CSR 2030-2 .040(1) Evaluation Criteria for Building Design Missouri Department of Transportation (MODOT) Standards and Specifications Missouri Department of Natural Resources (MDNR)

Missouri State Adopted International Code Council (ICC) Building Code Set 2012 Boone County Building Code City of Columbia, Missouri, Code of Ordinances Article II - Building and Fire Codes Section 6-16, Adopted Building Code 3-10

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NOmlWUT llllEOtCAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components Table 3-5. Local Government Documents (2 pages)

Referencea Title Section 6-17, Amendments Building Code Section 9-2 1 Fire Code Section 9-22 Fire Code a Complete references are provided in Section 3.6 CSR Code of State Regulations. MDNR Missouri Department of Natural Resources.

ICC = International Code Council. MO DOT Missouri Department of Transportation.

3.1.6 Discovery Ridge/University of Missouri Table 3-6 lists the MU system requirements and Di scovery Ridge covenants design inputs for the RPF identified in NWMI-DRD-2013-030. The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference.

Table 3-6. Discovery Ridge/University of Missouri Requirements Requirements Reference section/requirementa Civil Design and construction of the civil system is regulated by the NRC as requi red by Discovery Ridge/MU.

Collected Rules and Regulations (CRR)

Structural CRR Section 70.060.I, "Codes and Standards" - Adopts ICC codes University of Missouri, Consultant Procedures and Design Guidelines Electrical Section 2.4.2, "Building Codes and Standards fo r University Faciliti es" HV AC CPDG Division 23 , "Heating, Ventilating, and Air-Conditioning (HV AC)"

Instrumentation Section 2.4.2, "Building Codes and Standards fo r University Faciliti es" and Controls Planning CPDG Section 2.4, "Planning, Design and Contract Document Development Guidelines for Master Construction Delivery Method" Plumbing CPDG Division 22 , "Plumbing" Process Section 2.4.2, "Building Codes and Standards for University Facilities" University of Missouri, Facilities Management Policy and Procedures Manual Electrical Chapter 2, "Design and Construction Policy" Instrumentati on Chapter 2, "Design and Construction Policy" and Controls Structural Section 3.A, Refers to CRR 70.060 for the Basic Building Code Section 3.0, Refers to the University Building Adopted Codes for currently adopted codes University Building Adopted Codes IMC-2012 International Mechanical Code Structural Adopts IBC 20 12 a Complete references are provided in Section 3.6 CRR Collected Rules and Regulations. MU University of Missouri.

IBC International Building Code. NRC U.S. Nuclear Regulatory Commission.

ICC International Code Counci l.

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. * ~ ~* * ~ : . NORTHWUT MEDttAl tsOTOPES 3.1.7 Codes and Standards Table 3-7 lists design inputs for the RPF identified in NWMI-DRD-2013-030. The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference.

The Construction Pennit Application and associated preliminary design documents identify codes, standards, and other referenced documents that may be applicable to the RPF. The specific RPF design codes, standards, and other referenced documents, including exceptions or exemptions to the identified requirements, will be finalized in the RPF final design and provided to the NRC. In addition, the codes, standards, and referenced documents for the RPF safety SSCs that are needed to demonstrate compliance with regulatory requirements will be identified and committed to in the Operating License Application.

Table 3-7. Design Codes and Standards (12 pages)

Document number* Document title American Concrete Institute (ACI)

ACI 349 Code Requirements.for Nuclear Safety-Related Concrete Structures and Co mmentc11y, 20 13 American Institute of Steel Construction (AISC)

ANSI/ AISC N690 Specification for Safety-Related Steel Structures for Nuclear Facilities, 20 12 Air Movement and Control Association (AMCA)

AMCA Publication 201 Fans and Systems, 2002 (R2011)

AMCA Publication 203 Field Performance Measurement of Fan Systems, 1990 (R2011)

ANSI/AMCA 210 Laboratory Methods for Testing Fans for Aerodynamic Performance Rating, 2007 AMCA Publication 211 Certified Ratings Program - Product Rating Manual f or Fan Air Performance, 2013 AMCA Publication 311 Certified Ratings Program - Product Rating Manualfor Fan Sound Performance, 2006 (R2010)

American Conference on Governmental Industrial Hygienists (ACGIH)

ACGIH 2097 Industrial Ventilation: A Manual of Recommended Practice for Design, 2013 American National Standards Institute (ANSI)

ANSl/ITSDF B56.1 Safety Standard for Low Lift and High Lift Trucks ANSI/IEEE C2 2012 National Electrical Safety Code (NESC), 2012 ANSI C84.1 American National Standard for Electric Power Systems and Equipment - Voltage Ratings (60 Hertz), 2011 ANSI N 13 series Addresses radiation monitoring equipment ANSI N13 .1 Sampling and Monitoring Releases ofAirborne Radioactive Substances from the Stacks and Ducts ofNuclear Facilities 2011 ANSI N323D American National Standard.for Installed Radiation Protection Instrumentation ,

2002 ANSl/AIHA/ASSE Z9.5 Laboratory Ventilation, 2012 ANSI/NEMA Z535.1 Safety Colors, 2006 (R2011)

ANSI/NEMA Z535.2 Environmental and Facility Safety Signs, 2011 ANSI/NEMA Z535.3 Criteria f or Safety Symbols, 2011 ANSl/NEMA Z535.4 Product Safety Signs and Labels, 2011 3-12

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Document number* Document title ANSI/AMCA 204 Balance Quality and Vibration Levels for Fans , 2005 (R2012)

ANSI/AMCA 210 Laboratory Methods of Testing Fans for Aerodynamic Performance Rating, 2007 ANSI/AHRJ Standard 390 Performance Rating of Single Package Vertical Air-Conditioners and Heat Pumps, 2003 ANSI/ AHRJ Standard 410 Forced-Circulation Air-Cooling and Air-Heating Coils, 2001 ANSI/AHRJ Standard 430 Performance Rating of Central Station Air-Handling Units , 2009 ANSI/ AHRJ Standard 850 Performance Rating of Commercial and Industrial Air Filter Equipment, 2013 ANSI/HI 3. 1-3.5 Rotary Pumps, 2008 ANSI N42. l 7B American National Standard Performance Specifications for Health Physics Instrumentation - Occupational Airborne Radioactivity Monitoring Instrumentation, 1989 ANSI N42 . l 8 Specification and Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents, 2004 ANSI/IEEE N320 American National Standard Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation, 1979 American Nuclear Society (ANS)

ANSI/ ANS-2.3 Estimating Tornado, Hurricane, and Extreme Straight Line Wind Characteristics at Nuclear Facility Sites, 2011 ANSI/ ANS-2 .26 Categorization ofNuclear Facility Structures, Systems, and Components for Seismic Design, 2004 (R2010)

ANSI/ANS-2.27 Criteria for Investigations ofNuclear Facility Sites for Seismic Hazard Assessments, 2008 ANSI/ ANS-2 .29 Probabilistic Seismic Hazard Analysis, 2008 ANSI/ ANS-6.4 Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants, 2006 ANSI/ ANS-6.4 .2 Specification for Radiation Shielding Materials , 2006 ANSI/ ANS-8 .1 Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, 1998 (R2007) (W2014)

ANSI/ANS-8 .3 Critically Accident Alarm System, I 997 (R2012)

ANSI/ANS-8.7 Nuclear Criticality Safety in the Storage of Fissile Materials, 1998 (R2007)

ANSI/ANS-8.10 Criteria for Nuclear Criticality Control in Operations with Shielding and Confinement, 1983 (R2005)

ANSI/ ANS-8 . 19 Administrative Practices for Nuclear Criticality Safety, 1996 (R2014)

ANSI/ ANS-8 .20 Nuclear Criticality Safety Training, 1991 (R2005)

ANSI/ ANS-8 .21 Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors, 1995 (R201 l)

ANSI/ ANS-8 .24 Validation ofNeutron Transport Methods for Nuclear Criticality Safety Calculations, 2007 (R20 12)

ANSI/ANS-10.4 Verification and Validation ofNon-Safety-Related Scientific and Engineering Computer Programs for the Nuclear Industry, 2008 3-13

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Document number3 Document title ANSI/ANS -10.5 Accommodating User Needs in Computer Program Development, 2006 (R2011)

ANSI/ ANS-15 .17 Fire Protection Program Criteria for Research Reactors, 1981 (R2000) (W20 10)

ANSI/ANS-40.37 Mobile Low-Level Radioactive Waste Processing Systems, 2009 ANSI/ANS-55 .l Solid Radioactive Waste Processing System for Light Water Cooled Reactor Plants, 1992 (R2009)

ANSI/ANS-55.4 Gaseous Radioactive Waste Processing Systems for Light Water Reactor Plants ,

1993 (R2007)

ANSI/ ANS-55.6 Liquid Radioactive Waste Processing System for Light Water Reactor Plants, 1993 (R2007)

ANSI/ ANS-58.3 Physical Protection for Nuclear Safety-Related Systems and Components, 1992 (R2008)

ANSI/ ANS-58.8 Time Response Design Criteria for Safety-Related Operator Actions, 1994 (R2008)

ANSI/ ANS-59.3 Nuclear Safety Criteria for Control Air Systems, 1992 (R2002) (W2012)

Design Guides for Radioactive Material Handling Facilities and Equipment, Remote Systems Technology Division, 1988, Air Conditioning, Heating and Refrigeration Institute (ARRI)

ANSI/ARRI Standard 365 Performance Rating of Commercial and Industrial Unitary Air-Conditioning Condensing Units, 2009 ANSI/ARRI Standard 410 Forced-Circulation Air-Conditioning and Air-Heating Coils, 2001 American Society of Civil Engineers (ASCE)

ASCE4 Seismic Analysis of Safety-Related Nuclear Structures and Commentary, 2000 ASCE 7 Minimum Design Loads for Buildings and Other Structures, 2005 (R2010)

ASCE43 Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, 2005 ASCE Manual of Practice Design and Construction of Sanitary and Storm Sewers, 1969 37 American Society of Heating, Refrigeration and Air-Conditioning Engineers (ASHRAE)

ANSI/ASHRAE Standard Safety Standard for Refrigeration Systems, 20 13 15 ANSI/ASHRAE 51-07 Laboratory Methods of Testing Fans for Certified Aerodynamic Performance Rating, 2007 ANSI/ ASHRAE Standard Method for Testing General Ventilation Air Cleaning Devices for Removal 52.2 Efficiency by Particle Size, 2007 ANSI/ASHRAE Standard Thermal Environmental Conditions for Human Occupancy, 2013 55 ANSI/ ASHRAE Standard Ventilation for Acceptable Indoor Air Quality, 2010 62.l ASHRAE Standard 70 Method of Testing the Performance ofAir Outlets and Air Inlets, 2011 ANSI/ ASHRAE/IES Energy Standard for Buildings Except Low-Rise Residential Buildings, 20 10 Standard 90.1 ANSI/ASHRAE 110 Method of Testing Performance of Laboratory Fume Hoods, 1995 3-14

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Document number3 Document title ANSI/ ASHRAE 111 Measurement, Testing, Adjusting and Balancing of Building Heating, Ventilation, Air-Conditioning and Refrigeration Systems, 2008 American Society of Mechanical Engineers (ASME)

ASME A I 7. I Safety Code for Elevators and Escalators, 201 3 ASME AG-1 Code on Nuclear Air and Gas Treatment, 2012 ASME Bl6.5 Pipe Flanges and Flanged Fittings: NPW !Ii through 24, 2003 ASME B20. l Safety Standard for Conveyors and Related Equipment, 2012 ASME B30. l 7 Overheard and Gantry Cranes (Top Running Bridge, Single Girder, Underhung Hoist), 2006 ASME B30.20 Below-the-Hook Lifting Devices, 2013 ASME B31 .3 Process Piping, 20 14 ASME B3 l .9 Building Services Piping, 2011 /2014 ASME B3 l . l 2 Hydrogen Piping and Pipelines, 20 14 ASME B40.100 Pressure Gauges and Gauge Attachments, 2013 ASME B40.200 Th ermometers, Direct Reading and Remote Reading, 2013 ASME Boiler and Pressure Section VIII Division 1, 20I0/2013 Vessel Code Section IX ASME HST-1 Performance Standard for Electric Chain Hoists, 2012 ASME N509 Nuclear Power Plant Air-Cleaning Units and Components, 2002 (R2008)

ASME 510 Testing of Nuclear Air-Treatment Systems, 2007 ASME NQA-1 Quality Assurance Requirements for Nuclear Facility Applications, 2008 with NQA-1a-2009 addenda ASME QME-1 Qualification ofActive Mechanical Equipment Used in Nuclear Power Plants , 2012 American Society for Nondestructive Testing (ASNT)

SNT-TC- I A Recommended Practice No. SNT- TC- JA : Personnel Qualification and Certification in Nondestructive Testing, 2011 American Society for Testing and Materials (ASTM)

ASTM C I 055 Standard Guide for Heated System Surface Conditions that Produce Contact Burn Injuries, 2003 (20 14)

ASTM Cl217 Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials, 2000 ASTM C l5 33 Standard Guide for General Design Considerations for Hot Cell Equipment, 20 15 ASTM Cl554 Standard Guide for Materials Handling Equipment for Hot Cells, 2011 ASTM C1572 Standard Guide for Dry Lead Glass and Oil-Filled Lead Glass Radiation Shielding Window Components for Remotely Operated Facilities, 2010 ASTM Cl615 Standard Guide for Mechanical Drive Systems for Remote Operation in Hot Cell Facilities, 2010 ASTM Cl661 Standard Guide for Viewing Systems for Remotely Operated Facilities, 2013 3-1 5

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Document numbera Document title ASTM E493 Standard Practice for Leaks Using the Mass Spectrometer Leak Detector in the Inside-Out Testing Mode, 2011 ASTM F 14 71 Standard Test Method for Air Cleaning Performance ofHigh-Efficiency Particulate Air-Filter System, 2009 American Welding Society (A WS)

AWS B2. l/B2.1M Specification for Welding Procedure and Performance Qualification , 2009 AWS Dl.l / Dl.IM Structural Welding Code - Steel, 2010 AWS Dl .3/Dl .3M Structural Welding Code - Sheet Steel, 2008 AWS Dl.6/Dl.6M Structural Welding Code - Stainless Steel, 2007 AWS D9. l / D9.1M Sheet Metal Welding Code, 2006 AWS QCl Standard/or AWS Certification of Welding Inspectors , 2007 Centers for Disease Control and Prevention (CDC) - National Institute for Occupational Safety and Health (NIOSH)

DHHS (NIOSH) Publication Guidance for Filtration and Air Cleaning Systems to Protect Building Environments No. 2003-136 from Airborne Chemical, Biological, and Radiological Attacks, 2003 Electronic Industries Alliance (EIA)/Telecommunications Industry Association (TIA)

ANSI/TIA-568-C. l Commercial Building Telecommunications Cabling Standard, 2012 ANSI/TIA-568-C.2 Balanced Twisted-Pair Telecommunications Cabling and Components Standards ,

2014 ANSl/TIA-568-C.3 Optical Fiber Cabling and Components Standard, 2011 ANSI/TIA-5 69 Telecommunications Pathways and Spaces, 20 13 ANSI!fIA-606 Administration Standard for Commercial Telecommunications Infrastructure, 2012 ANSI/TIA-607 Commercial Building Grounding (Earthing) and Bonding Requirements for Telecommunications, 2013 ANSI/TIA-758-A Customer-Owned Outside Plant Telecommunications Infrastructure Standard, 2004 International Code Council ICC All7.l Accessible and Usable Buildings and Facilities Standard, 2009 IECC 2012 International Energy Conservation Code, May 2011

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IMC 2012 International Mechanical Code, June 2011 IPC international Plumbing Code, April 20 11 Institute of Electrical and Electronics Engineers (IEEE)

IEEE 7-4.3 .2 Standard Criteria fo r Digital Computers in Safety Systems of Nuclear Power Generating Stations , 2003 IEEE 141 Recommended Practice for Electric Power Distribution for Industrial Plants (Red Book), 1993 (R1999)

IEEE 142 Recommended Practice for Grounding ofIndustrial and Commercial Power Systems (Green Book), 2007 3-16

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  • NOltntWEST MEDtCAL ISOTOPES Chapter 3.0 - Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages)

Document number3 Document title IEEE 241 Recommended Practice for Electric Power Systems in Commercial Buildings (Gray Book), 1990 (Rl997)

IEEE 242 Recommended Practice for Protection and Coordination of Industrial and Commercial Power Systems (Buff Book) , 2001 IEEE 308 Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations, 2012 IEEE 315 Graphic Symbols for Electrical and Electronics Diagrams, 1975 (Rl 993)

IEEE 323 Standard for QualifYing Class IE Equipment for Nuclear Power Generating Stations, 2003 IEEE 336 Recommended Practice for Installation, Inspection, and Testing for Class IE Power, Instrumentation, and Control Equipment at Nuclear Facilities, 20 10 IEEE 338 Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, 2012 IEEE 344 Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations, 2013 IEEE 379 Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, 2014 IEEE 384 Standard Criteria/or Independence of Class IE Equipment and Circuits , 2008 IEEE 399 Recommended Practice for Power Systems Analysis (Brown Book), 1997 IEEE 446 Recommended Practice for Emergency and Standby Power Systems for Industrial and Commercial Applications (Orange Book) , 1995 (R2000)

IEEE 493 Recommended Practice for the Design of Reliable Industrial and Commercial Power Systems (Gold Book), 2007 IEEE 497 Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations , 2010 IEEE 519 Recommended Practice and Requirements for Harmonic Control in Electrical Power Systems, 2014 IEEE 535 Standard for Qualification of Class IE Lead Storage Batteries for Nuclear Power Generating Stations, 20 13 IEEE 577 Standard Requirements for Reliability Analysis in the Design and Operation of Safety Systems for Nuclear Facilities, 2012 IEEE 603 Standard Criteria for Safety Systems for Nuclear Power Generating Stations, 2009 IEEE 650 Standard for Qualification of Class IE Static Battery Chargers and Inverters for Nuclear Power Generating Stations, 2006 IEEE 739 Recommended Practice for Energy Management in Industrial and Commercial Facilities (Bronze Book), 1995 (R2000)

IEEE 828 Standard for Configuration Management in Systems and Software Engineering, 2012 IEEE 829 Standard for Software and System Test Documentation , 2008 3-17

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Document number3 Document title IEEE 902 Guide for Maintenance, Operation, and Safety of Industrial and Commercial Power Systems (Yellow Book), 1998 IEEE 946 Generating Stations, 2004 IEEE 1012 Standard Criteria for Software Verification and Validation, 2012 IEEE 1015 Recommended Practice Applying Low- Vo ltage Circuit Breakers Used in Industrial and Commercial Power Systems (Blue Book), 2006 (C2007)

IEEE 1023 Guide for the Application ofHuman Factors Engineering to Systems, Equipment, and Facilities ofNuclear Power Generating Stations, 2004 (R2010)

IEEE 1028 Standard f or Software Reviews and Audits, 2008 IEEE 1046 Application Guide for Distributed Digital Control and Monitoring for Power Plants, 1991 (Rl996)

IEEE 1050 Guide f or Instrumentation and Control Equipment Grounding in Generating Stations , 2004 IEEE 1100 Recommended Practice for Powering and Grounding Electronic Equipment (Emerald Book), 2005 IEEE 1289 Guide for the Application of Human Factors Eng ineering in the Design of Computer-Based Monitoring and Control Displays for Nuclear Power Generating Stations, I 998 (R2004)

IEEE 1584 IEEE Guide for Performing Arc-Flash Hazard Calculations , 2002 ANSI/IEEE C2 201 2 National Electrical Safety Code (NESC), 20 I 2 Illuminating Engineering Society of North America (IES)

IES-20 I I The Lighting Handbook, 20 I I ANSl/IES RP-1-12 American National Standard Practice for Office Lighting, 20 I 2 IES RP-7 American National Standard Practice fo r Lighting Industrial Facilities, 1991 (W2001)

International Society of Automation (ISA)

ANSl/ISA-5 . 1-2009 Instrumentation Symbols and Identification, 2009 ISA-5.3-1983 Graphic Symbols for Distributed Control/Shared Display Instrumentation, Logic, and Computer Systems, 1983 ISA-5.4-199 I Instrument Loop Diagrams, 1991 ISA-5.5-1985 Graphic Symbols for Process Displays, 1985 ANSl/ISA-5 .06.01-2007 Fun ctional Requirements Documentation for Control Sof tware Applications, 2007 ANSI/ISA 7.0.01-1996 Quality Standard for Instrument Air ANSI/ISA-12 .0 I .01-2013 Definitions and Information Pertaining to Electrical Equip ment in Hazardous (Classified) Locations, 2013 ISA-18.1-1979 Annunciator Sequences and Specifications, 1979 (R2004)

ISA-TR20.00.01-2007 Specification Forms f or Process Measurement and Control Instruments Part 1:

General Considerations Updated with 27 new sp ecification forms in 2004-2006 and updated with 11 new specification form s in 200 7 3-18

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Document numbera Document title ISA-RP60.1-1990 Control Center Facilities, 1990 ISA-67.01 .01-2002 Transducer and Transmitter Installation for Nuclear Safety Applications, 2002 (R2007)

ANSUISA-67.04.01-2006 Setpoints for Nuclear Safety-Related Instrumentation, 2006 (R2011)

ISA-RP67 .04.02 -20 10 Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation, 20 10 ANSl/ISA-75 .05 .01-2000 Control Valve Terminology, 2000 (R2005)

ANSl/ISA-82 .03-1988 Safety Standard for Electrical and Electronic Test, Measuring, Controlling, and Related Equipment, 1988 ISA-TR84.00.04-2011 Part 1 Guideline for the Implementation ofANSJIISA-84.00.01-2004 (!EC 61511),

2011 ISA-TR84.00.09-2013 Security Countermeasures Related to Safety Instrumented Systems (SIS), 2013 ISA-TR9 l .00.02-2003 Criticality Classification Guideline for Instrumentation , 2003 ANSUISA-TR99.00.0l- Security Technologies for Industrial Automation and Control Systems, 2007 2007 International Atomic Energy Agency (IAEA)

IAEA-TECDOC-1250 Seismic Design Considerations of Nuclear Fuel Cycle Facilities, 2001 IAEA-TECDOC-134 7 Consideration of External Events in the Design of Nuclear Facilities Other Than Nuclear Power Plants, With Emphasis on Earthquakes, 2003 IAEA-TECDOC-1430 Radioisotope Handling Facilities and Automation of Radioisotope Production , 2004 International Code Council (ICC)

IBC 2012 International Building Code, 20 12 IFC 2012 International Fire Code, 2012 IMC 201 2 International Mechanical Code, 201 2 International Code Council Evaluation Service (ICC-ES)

ICC-ES AC 156 "Acceptance Criteria for Seism ic Certification by Shake-Table Testing of Nonstructural Components," 20 I 0 National Electrical Contractors Association (NECA)

NECA 1 Standard Practice of Good Workma nship in Electrical Construction , 20 10 NECA90 Recommended Practice for Commissioning Building Electrical Systems (ANSI),

2009 NECA 100 Symbols fo r Electrical Construction Drawings (ANSI), 2013 NECA 101 Standard for Installing Steel Conduits (Rigid, IMC, EMT) (ANSI), 2013 NECA/AA 104 Standard for Installing Aluminum Building Wire and Cable (ANSI), 2012 NECA/NEMA 105 Standard for Installing Metal Cable Tray Systems (ANSI), 2007 NECA 111 Standard for Installing Nonmetallic Raceways (RNC, ENT, LFNC) (ANSI), 2003 3-19

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  • ..NWMI NOllTHWllT lllOtCAl ISDTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages)

Document number3 Document title NECA 120 Standard for Installing Armored Cable (Type AC) and Metal-Clad Cable (Type MC)

(ANSI) , 2013 NECA 202 Standard for Installing and Maintaining Industrial Heat Tracing Systems (ANSI),

2013 NECA 230 Standard for Selecting, Installing, and Maintaining Electric Motors and Motor Controllers (ANSI), 2010 NECA/FOA 301 Standard for Installing and Testing Fiber Optics, 2009 NECA331 Standard for Building and Service Entrance Grounding and Bonding, 2009 NECA400 Standard for Installing and Maintaining Switchboards (ANSI) , 2007 NECA402 Standard for Installing and Maintaining Motor Control Centers (ANSI), 2007 NECA/EGSA 404 Standard for Installing Generator Sets (ANSI), 2014 NECA407 Recommended Practice for Installing and Maintaining Pane/boards (ANSI), 2009 NECA408 Standard for Installing and Maintaining Busways (ANSI), 2009 NECA409 Standard for Installing and Maintaining Dry-Type Transformers (ANSI), 2009 NECA 410 Standard for Installing and Maintaining Liquid-Filled Transformers (ANSI), 2013 NECA 411 Standard for Installing and Maintaining Uninterruptible Power Supplies (UPS)

(ANSI), 2006 NECA420 Standard for Fuse Applications (ANSI) , 2014 NECA430 Standard for Installing Medium-Voltage Metal-Clad Switchgear (ANSI), 2006 NECA/IESNA 500 Recommended Practice for Installing Indoor Lighting Systems (ANSI) , 2006 NECA/IESNA 501 Recommended Practice for Installing Exterior Lighting Systems (ANSI), 2006 NECA/IESNA 502 Recommended Practice for Installing Industrial Lighting Systems (ANSI), 2006 NECA/BICSI 568 Standard for Installing Building Telecommunications Cabling (ANSI), 2006 NECA/NCSCB 600 Recommended Practice for Installing and Maintaining Medium-Voltage Cable (ANSI) , 2014 NECA/NEMA 605 Installing Underground Nonmetallic Utility Duct (ANSI), 2005 National Electrical Manufacturers Association (NEMA)

NEMAMG-1 Motors and Generators, 2009 InterNational Electrical Testing Association (NET A)

ANSI/NETA ATS-2013 Standard for Acceptance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2013 ANSI/NETA ETT-2010 Standard for Certification of Electrical Testing Technicians, 2010 ANSI/NETA MTS-2011 Maintenance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2011 National Fire Protection Association (NFPA)

NFPA 1 Fire Code, 2015 FPA2 Hydrogen Technologies Code, 20 11 3-20

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Document numbera Document title NFPA4 Standard for Integrated Fire Protection and Life Safety System Testing, 2015 NFPA 10 Standard for Portable Fire Extinguishers , 2013 NFPA 13 Standard for the Installation of Sprinkler Systems, 2013 NFPA 14 Standard for the Installation of Standpipe and Hose Systems , 2013 NFPA20 Standard for the Installation of Stationary Pumps for Fire Protection, 2013 NFPA 22 Standard for Water Tanks for Private Fire Protection, 2013 NFPA 24 Standard for the Installation ofPrivate Fire Service Mains and Their Appurtenances, 2013 NFPA 25 Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems, 2014 NFPA 30 Flammable and Combustible Liquids Code, 2015 NFPA 37 Standard for the Installation and Use of Stationary Combustion Engines and Gas Turbines, 2015 NFPA45 Standard on Fire Protection for Laboratories Using Chemicals, 2015 NFPA 55 Compressed Gases and Cryogenic Fluids Code, 2013 NFPA68 Standard on Explosion Protection by Deflagration Venting, 2013 NFPA 69 Standard on Explosion Prevention Systems , 2014 NFPA 70 National Electrical Code (NEC), 2014 NFPA 70B Recommended Practice for Electrical Equipment Maintenance, 2013 NFPA 70E Standard for Electrical Safety in the Workplace, 2015 NFPA 72 National Fire Alarm and Signaling Code, 2013 NFPA 75 Standard for the Fire Protection ofInformation Technology Equipment, 2013 NFPA 79 Electrical Standard for Industrial Machinery, 2015 NFPA 80 Standard for Fire Doors and Other Opening Protectives, 2013 NFPA 80A Recommended Practice for Protection of Buildings from Exterior Fire Exposures, 2012 NFPA 86 Standard for Ovens and Furnaces, 2015 NFPA 86C Standard for Industrial Furnaces Using a Special Processing Atmosphere, 1999 NFPA90A Standard for the Installation ofAir-Conditioning and Ventilating System, 2015 NFPA 90B Standard for the Installation of Warm Air Heating and Air-Conditioning Systems, 2015 NFPA 91 Standard for Exhaust Systems for Air Conveying of Vapors, Gases, Mists, and Noncombustible Particulate Solids, 2015 NFPA 92 Standard for Smoke Control Systems, 2012 NFPA92A Standard for Smoke-Control Systems Utilizing Barriers and Pressure Differences, 2009 NFPA 92B Standard for Smoke Management Systems in Malls, Atria, and Large Spaces, 2009 3-21

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Document numbera Document title NFPA IOIB Code for Means ofEgress for Buildings and Structures, 2002 (W-Next Edition)

NFPA 105 Standard for the Installation of Smoke Door Assemblies and Other Opening Protectives, 20 13 NFPA 110 Standard for Emergency and Standby Power Systems, 2013 NFPA Ill Standard on Stored Electrical Energy Emergency and Standby Power Systems, 2013 NFPA 170 Standard for Fire Safety and Emergency Symbols, 2012 NFPA 204 Standard for Smoke and Heat Venting, 2012 NFPA 220 Standard on Types of Building Construction, 2015 NFPA 221 Standard for High Challenge Fire Walls, Fire Walls, and Fire Barrier Walls, 2015 NFPA 262 Standard Method of Test for Flame Travel and Smoke of Wires and Cables for Use in Air-Handling Spaces, 2015 NFPA 297 Guide on Principles and Practices for Communications Systems, 1995 NFPA 329 Recommended Practice for Handling Releases of Flammable and Combustible Liquids and Gases, 2015 NFPA 400 Hazardous Materials Code, 2013 NFPA496 Standard for Purged and Pressurized Enclosures for Electrical Equipment, 2013 NFPA 497 Recommended Practice for the Classification of Flammable liquids, Gases, or Vapors and of Hazardous (Classified) locations for Electrical Installations in Chemical Process Areas, 2012 NFPA 704 Standard System for the Identification of the Hazards ofMaterials for Emergency Response, 2012 NFPA 730 Guide for Premises Security, 20 14 NFPA 731 Standard for the Installation ofElectronic Premises Security Systems, 2015 NFPA 780 Standard for the Installation ofLightning Protection Systems, 2014 NFPA 791 Recommended Practice and Procedures for Unlabeled Electrical Equipment Evaluation, 201 NFPA 801 Standard for Fire Protection for Facilities Handling Radioactive Materials, 2014 Sheet Metal and Air Conditioning Contractors National Association (SMACNA)

National Oceanic and Atmospheric Administration (NOAA)

NOAA Atlas 14 Precipitation-Frequency Atlas of the United States, Vol. 8 Version 2.0, 2013 SMACNA 1143 HVAC Air Duct Leakage Test, 1985 SMACNA 1520 Round Industrial Duct Construction Standard, 1999 SMACNA 1922 Rectangular Industrial Duct Co nstruction Standard, 2004 SMACNA 1966 HVAC Duct Construction Standard - Metal and Flexible, 2006 SMACNA-2006 HVAC Systems Duct Design, 2006 3-22

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Document number3 Document title ANSl/SMACNA 00 1-2008 Seismic Restraint Manual: Guidelines for Mechanical Systems, 2008 U.S. Weather Bureau Technical Paper No. 40 Rainfall Frequency Atlas of the Un ited States for Durations from 30 Minutes to 24 Hours and Return Periods from 1 to 100 Years, 1963 Underwriters Laboratory, Inc. (UL) Feder al Specifications UL 18 1 Standard for Factory-Made Air Ducts and Connectors, 2013 UL 499 Standard for Electric Heating Appliances, 20 14 UL 55 5 Standard for Fire Dampers, 2006 UL 586 Standard f or High Efficiency, Particulate, Air Filter Units, 2009 UL900 Standard for Air Filter Units, 2004 UL 1995 Heating and Cooling Equipment, 2011

  • Complete references are provided in Section 3.6 ACG JH Ameri can Conference on Govern mental IAEA International Atomic Energy Age ncy.

Industrial Hygienists. ICC International Code Council.

AC I American Concrete Institute. ICC-ES International Code Council Evaluati on Servi ce.

AHRI Air Conditioning, Heating and Refrigeration IEEE Institute of Electrical and Electronics Engineers.

Institute. IES Illuminating Engineering Society.

AISC American Institute of Steel Construction. ISA In ternati onal Society of Automation.

AMCA Air Movement and Control Association. NECA National Electrical Contractors Association.

ANS Ameri can Nuclear Society. NEMA Nati onal Electrical Manu factu rers Associati on.

ANSI Ameri can Nati onal Standard s Institute. NETA InterNational Electrical Testing Association .

ASCE Ameri can Society of Civil Engineers. NF PA National Fire Protection Association.

ASHRAE Ameri can Society of Heating, Refrigerati on NIOSH National Institute for Occupati onal Safety and and Air-Conditi oning Engineers. Health.

ASME Ameri can Society of Mechanical Engineers. NOAA National Oceanic and Atmospheri c ASNT American Society fo r Nondestructive Administration Testing. SMACNA Sheet Meta l and Air Conditionin g Contractors ASTM American Society for Testing and Materials. National Association .

AWS American Welding Society. TIA Telecommunicati ons Industry Association.

CDC Centers fo r Disease Contro l and Preventi on . UL Und erwriters Laboratory.

EIA E lectroni c Industries All iance.

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' ~ ~.*-~ : . NOmlWESTMUHCAl.ISOTOP£1 NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.2 METEOROLOGICAL DAMAGE RPF meteorological accidents with radiological consequences are evaluated in NWMI-2015-SAFETY-01 l ,

Evaluation of Natural Phenomenon and Man-Made Events on Safety Features and Items Relied on for Safety. The basis for the structural design of the RPF is described in NWMI-2013-043 , NWM!

Radioisotope Production Facility Structural Design Basis.

Updates and development of technical specifications associated with the meteorological design of the RPF SSCs will be provided in Chapter 14.0 as part of the Operating License Application.

The demands on structural elements due to applied loads are evaluated using the criteria and methodology discussed below. The effect of each load case is determined separately, and total demand is determined by combining the load effects using the load combinations for evaluating strength and evaluating the serviceability criteria given below.

Four categories of load cases are used: normal, severe environmental, extreme environmental, and abnormal loads. The definition of each load is the following :

  • Normal loads are loads that are expected to be encountered during normal plant operations and shutdown, and load due to natural hazard phenomena likely to be encountered during the service life of the facility.
  • Severe environmental loads are loads that may be encountered infrequently during the service life of the facility.
  • Extreme environmental loads are loads that are credible but are highly improbable to occur during the service life of the facility.
  • Abnormal loads are loads generated by a postulated high-energy pipe break accident used as a design basis.

Definitions of load case symbols are provided in Table 3-8.

Table 3-8. Load Symbol Definitions (2 pages)

Symbol Definition Normal Load Cases D Dead loads due to the weight of the structural elements, fixed-position equipment, and other permanent appurtenant items; weight of crane trolley and bridge F Load due to fluids with well-defined pressures and maximum heights H Load due to lateral earth pressure, groundwater pressure, or pressure of bulk materials L Live load due to occupancy and moveabl e equipment, including impact Lr Roof live load Ccr Rated capacity of crane (will include the maximum wheel loads of the crane and the vertical, lateral ,

and longitudinal forces induced by the moving crane)

S Snow load as stipulated in ASCE 7" for risk category IV facilities R Rain load T0 Self-staining load, thermal effects, and loads during normal operating, startup, or shutdown conditions, based on the most critical transient or steady-state condition Ro Pipe reactions during normal operating, startup , or shutdown conditions, based on the most critical transient or steady-state condition 3-24

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  • NOfllTHWEST MlDfCAl ISOTOPU Table 3-8. Load Symbol Definitions (2 pages)

Symbol Definition Severe Environmental Load Cases D; Weight of ice Fa Flood load W Load due to wind pressure Wa Load based on serviceability wind speed W; Wind-on-ice Eo Where required as part of the design basis, loads generated by the operating basis earthquake, as defined in I 0 CFR 50,b Appendix S, "Earthquake Engineering Criteria for Nuclear Power Plants," or as specified by the authority having jurisdiction.

Extreme Environmental Load Cases S, Weight oftbe 48-hour probable maximum winter precipitation superimposed on S W1 Loads generated by the specified design basis tornado, including wind pressures, pressure differentials, and tornado-borne mi ssiles, as defi ned in NUREG-0800,c or as specified by the authority having jurisdiction Ess Loads generated by the safe shutdown, or design basis earthquake, as defined in IO CFR 50,b Appendix S, or as specified by the authority having jurisdiction Abnormal Load Cases Pa Maximum differential pressure load generated by the postulated accident Ra Pipe and equipment reactions generated by the postulated accident, including Ro Ta Thermal loads generated by the postulated accident, including To Yj Jet impingement load generated by the postulated accident Ym Missile impact load, such as pipe whip generated by or during the postulated accident Y, Loads on the structure generated by the reaction of the broken hi gh-energy pipe during the postulated accident a ASCE 7, Minimum Design loads for Buildings and Other Structures, American Society of Civi l Engineers, Reston, Virgi nia, 2005 (R2010).

b I 0 CF R 50, " Domesti c Licensing of Prod uction and Utili zation Faci lities," Code of Federal Regulations, Office of th e Federal Register, as amended.

c NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, L WR Edition, U.S. Nuclear Regulatory Commission, Office ofNuclear Material Safety and Safeguards, Washington, D.C., 1987.

3.2.1 Combinations of Loads Load combinations used for evaluating strength and serviceability are given in the following subsections.

Combinations for strength-based acceptance criteria are given for both nuclear safety-related SSCs and for commercial SSCs.

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  • . . NO<<THWEST MEOK:AL &SOTOfl£S NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.2.1.1 Nuclear Safety-Related Structures, Systems, and Components For nuclear safety-related SSCs, the loading combinations from ACI 349, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, are used. The load combinations from ACI 349 are essentially identical to the combination from ANSVAISC N690, Specification for Safety-Related Steel Structures for Nuclear Facilities. Table 3-9 presents nuclear safety-related SSC loads. In addition, the load combination for extreme winter precipitation load (S,) takes DC/COL-ISG-007, Interim Staff Guidance on Assessment ofNormal and Extreme Winter Precipitation Loads on the Roofs of Seismic Category I Structures, guidance into account.

Table 3-9. Load Combinations for Strength Based Acceptance Criteria, Nuclear Safety-Related Normal Load Combinations Combination ea1u* ANSl/AISC N690b l .4(D + F + Ro) + To (9-1) (NB2-l)

I .2(D + F + T0 + R0 ) + I .6(L + H) + 1.4Ccr + 0.5(L, or S or R) (9-2) (NB2-2) l .2(D + F + Ro) + 0.8(L + H) + l .4Ccr + I .6(Lr or S or R) (9-3) (NB2-3)

Severe Environmental Load Combinations 1.2(D + F + Ro) + l.6(L + H + Eo) (9-4) (NB2-4) 1.2(D + F + Ro) + l.6(L + H + W) (9-5) (NB2-5)

Extreme Environmental and Abnormal Load Combinations D + F + 0.8L + Ccr + H + T0 + Ro + Ess (9-6) (NB2-6)

D + F + 0.8L + H + To+ Ro + W1 (9-7) (NB2-7)

D + F + 0.8L + Ccr + H + Ta + Ra + l.2Pa (9-8) (NB2-8)

D + F + 0.8L + H + Ta + Ra + Pa+ Y, + Yj + Ym + Ess (9-9) (NB2-9)

D + F + 0.8L + Ccr + H + T0 + Ro + s,

  • ACI 349, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, American Concrete Institute, Farmington Hills, Michigan, 201 3.

b ANSUAISC N690, Sp ecification for Safety-Related Steel Structures fo r Nu clear Facilities, American Institute of Steel Construction, Chicago, Illinois, January 31 , 201 2.

3.2.1.2 Commercial and Nuclear Non-Safety-Related Structures, Systems, and Components For commercial and nuclear non-safety-related SSCs, the loading combinations from American Society of Civil Engineers (ASCE) 7, Chapter 2 are used. When the loading includes earthquake effects, the special seismic load combinations are taken from ASCE 7, Minimum Design Loads for Buildings and Other Structures, Chapter 12. The basic load combinations for the strength design of commercial type and non-safety-related nuclear SSCs are given in Table 3-10. The combinations listed are obtained from the 2012 International Building Code (IBC) and ASCE 7. The crane live load case (Ccr) is separated from other live loads in the combinations for design purposes.

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  • NOllTHWEST MEDICAL ISOTOPH Chapter 3.0 - Design of Structures, Systems and Components Table 3-10. Load Combinations for Strength Base Acceptance Criteria, Commercial Combination 1sca ASCE 7b Basic Load Combinations 1.4(D +F) (16-I) I I .2(D + F) + I .6(L + Ccr + H) + 0.5(Lr or S or R) (16-2) 2 1.2(D + F) + 1.6(Lr or Sor R) + l.6H + l/1(L + Ccr) or 0.5W] (16-3) 3 I .2(D + F) + I .OW + f 1(L + Ccr) + I .6H + 0.5(Lr or Sor R) (16-4) 4 l.2(D + F) + I.OE + f 1(L + Ccr) + l.6H + fiS (16-5) 5 0.9D+ I .OW+ I .6H (16-6) 6 0.9(D + F) + I.OE+ 1.6H (16-7) 7 Load Combinations, including Flood Load I .2D + (0.5W + I .OF.) + L + 0.5(Lr or Sor R) §1605.2.1 §2.3.3.2 0.9D + (0.5W + I .OFa) §1605.2. l §2.3.3.2 Load Combinations, including Atmospheric Ice I .2D + I .6L +(0 .2Di + 0.5S) §I605.2.l §2.3.4.1 l .2D + L + (Di + Wi + 0.5S) §1605.2.1 §2.3.4.2 0.9D + (Di + Wi) § 1605.2.I §2.3.4.3 Where:

fl = 0.5 for other li ve loads.

f2 = 0.7 for flat roof configurations, which do not shed snow, and 0.2 for other roof configurations a IBC 2012, International Building Code, International Code Council, Inc., Washington D.C.

b ASCE 7, Minimum Design Loads for Buildings and Other Structures , American Society of Civil Engi neers, Reston, Virginia, 2010.

3.2.2 Combinations for Serviceability Based Acceptance Criteria Based on ASCE 7, Appendix C Commentary, Table 3-11. Load Combinations for Serviceability the load combinations given in Table 3-11 are Based Acceptance Criteria used when evaluating serviceability based acceptance criteria. Combination ASCE7 Short-Term Effects 3.2.3 Normal Loads D+L (CC-la)

The RPF is required to resist loads due to: D + 0.5S (CC-lb)

Creep, Settlement and Similar Long-Term of Permanent

  • Operating conditions of the systems Effects and components within the RPF D + 0.5L (CC-2)
  • Normal and severe natural phenomena Drift of Walls and Frames hazards, remaining operational to D + 0.5L + Wa (CC-3) maintain life-safety and safety-related Seismic Drift SSCs Per ASCE 7, Section I2 .8.6
  • Extreme natural phenomena hazards, a Appendix C, Commentary, of ASCE 7, Minimum Design maintaining life-safety and safety- Loads for Buildings and Other Structures, American Society of related SSCs Civi l Engineers, Reston, Virginia, 2013 .

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0 NOITifMST MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components Structural loads are due to the following :

  • Self-weight of building materials and SSCs
  • Occupancy and normal use of the RPF
  • Off-normal conditions and accidents
  • Natural phenomena hazards Section 3.1 describes the structural discipline source requirements for these criteria. Structural load criteria are summarized below. Site-specific natural phenomena hazard criteria are based on the physical location of the RPF given in Chapter 2.0, Sections 2.3 and 2.5.

3.2.3.1.1 Dead Loads Dead loads consist of the weight of all materials of construction comprising the building, including walls, floors , roofs, ceilings, confinement doors, stairways, built-in partitions, wall and floor finishes , and cladding. Dead loads also consist of the weight of fixed equipment, including the weight of cranes. The density of all interconnections (e.g., heating, ventilation, and air conditioning [HV AC] ductwork, conduits, cable trays, and piping) between equipment will be conservatively estimated and included in the final design for dead load for fixtures attached to ceilings or anchored to floors in the RPF.

3.2.3.1.2 Lateral Earth and Ground Water Pressure Loads Lateral earth and groundwater pressure loads are lateral pressures due to the weight of adjacent soil and groundwater, respectively. The design lateral earth load is a function of the composition of the soil. The Discovery Ridge Phase I Environmental Assessment (Terracon, 2011 a) indicates that the soils present are clayey gravels consistent with the Unified Soil Classification "GC." In addition, the assessment indicates that expansive soils are present. Chapter 2.0, Section 2.5.3 presents additional on-site soil information.

The design lateral earth pressure load for the RPF is based on ASCE 7, Table 3.2.1 , and has been augmented to account for the expansive soils (e.g., surcharge load is applied to account for the weight of the facility above the soils adjacent to the tank hot cell).

The design groundwater depth is estimated to be Table 3-12. Lateral Earth Pressure Loads approximately 5.5 meters (m) (18 feet [ft]) below-ground surface and will be verified pending final Element Value geotechnical investigation. Additional information Base design lateral soil load 45 lb/ft2 per ft is presented in Chapter 2.0, Section 2.4.2.

Design lateral load (expansive increase) 60 lb/ft 2 per ft The lateral earth pressure loads for the RPF are

Reference:

Table 3.2-1 of ASCE 7, Minimum Design Loads presented in Table 3-12. for Buildings and Other Structures, Ameri can Society of Civil Engineers, Reston, Virginia, 20 13.

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  • NORTHWEn MEDJCAL lSOTOPH NWMl-2013-02 1, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.2.3.1.3 Live Loads Floor Live Load Table 3-13. Floor Live Loads Live loads are produced by the use and occupancy Description Uniform Concentrated of the RPF, and as such, different live load 2

magnitudes are appropriate for different areas of Production area 250 lb/ft 3,000 lb the facility. Design floor loads provided in Hot cell roof TBD TBD Table 3-13 are based on ASCE 7, Sections 4.3 and Cover block laydown TBD TBD 4.4, and Section C4.3 Commentary. 2 Mechanical rooms 200 lb/ft 2,000 lb During the structural analysis, unknown loads Laboratory 100 lb/ft 2 2,000 lb (e.g., hot cell roof in Table 3-13) will have a conservative value assumed and marked with Office 50 lb/ft 2 2,000 lb

"(HOLD)." As the design matures, the actual 2 Office partitions 20 lb/ft values will be inserted in the analysis and the Corridors 100 lb/ft 2 HOLDs removed. Final design media cannot be Truck bay Per AASHTO issued if there are HOLDs identified. The facility live loads will be established during the Based on Sections 4.3 , 4.4, and C4.3 Commentary of completion of the final facility design and ASCE 7, Minimum Design loads for Buildings and Other Structures, American Society of Civil Engineers, Reston, provided as part of the Operating License Virgi nia, 2013.

App lication.

AASHTO American Association of State Highway and Roof Live Load Transportation Officials.

TBD to be determined.

The minimum roof live load (Lr) prescribed by the City of Columbia is 20 pounds (lb )/square foot (ft2) , non-reducible (Ordnance No. 21804, Section 6-17).

Snow loads (e.g., normal and extreme rain-on-snow) are discussed separately in Section 3.2.5.2.

Crane Loads The design basis crane load criteria are given in Table 3-14. Crane Load Criteria Table 3-14 and are based on a preliminary quote provided in NWMI-2015-SDD-001 , RPF Facility Element Value SDD. The crane design is to run a top-running Crane capacity 75 ton (150 kip) bridge crane with a remotely operated, powered Crane weight (with hoists) 69,990 !bf bridge and hoist.

Bridge weight 62,330 !bf The crane design basis will be refined in the final design and Operating License Application to Hoist and trolley weight 7,660 !bf account for the following: Wheel load (static) 54.3 kip

  • ASCE 7, Chapter 3 - Include weights of crane and runway beams in dead loads
  • ASCE 7, Chapter 4 - Increase wheel load by 25 percent to account for vertical impact
  • ASCE 7, Chapter 4 - Determine lateral force by multiplying sum of hoist and trolley weight and rated capacity of crane by 20 percent
  • ASCE 7, Chapter 4 - Determine longitudinal force by multiplying the wheel load by 10 percent 3-29

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~. NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components

  • ~ * . * ~ ' . NOITNWUT MUHCAl lSOTWU 3.2.4 Wind Loading 3.2.4.1 Wind Load PerNUREG-1537, Section 2.3 .1 , "General and Local Climate, wind loads will be based on the 100-year return period wind speed. In addition, based on NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 3.3.1 , the wind speed will be transformed to equivalent pressure per ASCE 7-05. For RPF SSCs per current applicable 2012 IBC guidance, ASCE 7-10 is used for this transformation of wind speed to equivalent pressure. From Table 1.5-1 of ASCE 7-10 and based on use and occupancy of the RPF, a Risk Category IV is assigned to RPF SSCs.

Figure 26.5- I B for a Risk Category IV building of ASCE 7-10 is used to obtain the basic wind speed for the RPF site.

The mean recurrence interval (MRI) of the basic wind speed for Risk Category IV buildings is 1,700 years. Since the MRI stipulated in ASCE 7-10 is more stringent than NUREG-1537 100-year wind speeds, wind loads will be determined in accordance with ASCE 7-10, Chapters 26 through 30, as applicable, for a Risk Category IV building.

The surface roughness surrounding RFP SSCs Table 3-15. Wind Loading Criteria is currently Surface Category C, which in turn Element Value indicates Exposure Category C for the RFP per Basic wind speed, V 193 .1 km/hr (120 mi/hr)

ASCE 7-10. The RPF main building is an Exposure category c enclosed building. The wind loading criteria Enclosure classification Enclosed are provided in Table 3-15 . The basic wind Risk category IV speed given in Table 3-15 is a 3-second (sec) gust wind speed at 10 m (33 ft) aboveground Source: ASCE 7- 10, Minimum Design Loads f or Buildings and Other Structures, American Society of Civil Engineers, Reston, for Exposure Category C and Risk Category IV.

Virginia, 2010.

The wind loading criteria will be updated in the Operating License Application.

3.2.4.2 Tornado Loading Tornado loads are a combination of tornado wind effects, atmospheric pressure change, and tornado-generated missile impact effects and are discussed separately in the following sections. NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility, Part 3, Appendix D, states that an annual exceedance probability of 10-5 may need to be considered. The maximum tornado wind speed from NRC Regulatory Guide 1.76, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants , for Region I, has an annual exceedance probability of 10-7 that is significantly lower than the target probability stated in NUREG-1520.

For the RPF preliminary safety analysis report, the maximum tornado wind speed from NRC Regulatory Guide 1.76 for Region I will be used. The tornado load criteria will be updated by using tornado loading in accordance with 10-5 annual probability of exceedance in the Operating License Application.

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.. NORTHWUT MEDICAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.2.4.2.1 Maximum Tornado Wind Speed Tornado wind field characteristics used to Table 3-16. Design-Basis Tornado Field Characteristics calculate tornado wind pressures on the RPF are provided in Table 3-16 per NRC Description Value Regulatory Guide 1.76. The maximum Tornado region Region I tornado wind speed has two components:

Maximum wind speed 370.1 km/hr (230 mi/hr) translational and rotational. The maximum total tornado wind speed is the Translational speed 74.0 km/hr (46 mi/hr) sum of these two components and is Radius of maximum rotational speed 45.7 m (150 ft) applied to the RPF building from each Pressure drop, Af> (1.2 lb/in.2) direction separately. Based on Source: NRC Regulatory Guide I. 76, Design-Basis Tornado and NUREG-0800, Section 3.3 .2, ASCE 7-05 Tornado Missiles for Nuclear Power Plants , Rev. I, U.S. Nuclear may be used to transform maximum Regulatory Commission, Washington, D.C., March 2007.

tornado wind speed to equivalent pressure.

For RPF SSCs per current applicable 2012 IBC guidance, Chapters 26 and 27 of ASCE 7-10 is used for this transformation of tornado wind speed to equivalent pressure. From Table 1.5-1 of ASCE 7-10 and based on use and occupancy of the RPF, a Risk Category IV is assigned to RPF SSCs. Per NUREG-800, Section 3.3.2, tornado wind speed is assumed not to vary with the height aboveground. Additional information is provided in Chapter 2.0, Section 2.3 .1.5, and Chapter 13 .0, Section 13 .2.6. l.

3.2.4.2.2 Atmospheric Pressure Change NRC Regulatory Guide 1.76 provides guidance for determining the pressure drop and the rate of pressure drop caused by the passing of a tornado. Depending on the final design of the RPF building and whether it is enclosed (unvented) or partially enclosed (vented structure), the procedures outlined in NUREG-800 Section 3.3.2 will be used to account for atmospheric pressure change effects. At the preliminary stage of the design, the RPF building is known not to be open. The value for atmospheric pressure drop, corresponding to the design-basis tornado is given in Table 3-16.

3.2.4.2.3 High Straight-Line Winds 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.

The RPF is designed as a Risk Category IV structure, a standard industrial facility with equivalent chemical hazards, in accordance with ASCE 7. The return frequency of the basic (design) wind speed for Risk Category IV structures is 5.88 x J0-4/year (MRI = 1,700 year). The provisions of ASCE 7, 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 targeted for Risk Category IV structures is 5.0 x l0-6 .

3.2.4.2.4 Tornado-Generated Missile Impact Effects Tornado-generated missile impact effects are based on the standard design missile spectrum from NRC Regulatory Guide 1. 76 and are presented in Table 3-17. These requirements are considered more severe than the characteristics from DOE-STD-1020, Natural Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities, that are cited in NUREG-1520, Section 3. The recommended RPF roof and wall system design criteria are also taken from DOE-STD-1020, Table 3-4.

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  • ~ * .* ~ . NORTHWEST MEDtCAl JSOTOftl:S NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures , Systems and Components Table 3-17. Design-Basis Tornado Missile Spectrum Description Automobile
  • uwn* 4,000 lb Dimensions I 6.4 ft x 6.6 ft x 4.3 ft Horizontal velocity 92 mi/hr Vertical velocity 62 mi/hr Pipe 287 lb 6.625 in. diameter x 15 ft long 92 mi/hr 62 mi/hr Steel Sphere 0.147 lb 1.0 in. diameter 18 mi/hr 12 mi/hr Source: NRC Regulatory Guide 1.76, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants , U.S.

uclear Regulatory Commission, Washington, D.C. , March 2007 .

The impact-type missile, an automobile is limited to a hei ght of no more than 9.1 m (30 ft) above-grade.

Structural wall openings are subjected to the impact of a 0.25 centimeters (cm) (I-inch [in.]) diameter steel sphere. The vertical velocities are taken as 0.67 of the horizontal velocity. For an automobile and pipe missile, a normal impact is assumed. The tornado load criteria will be updated by using tornado loading in accordance with 10-5 annual probability of exceedance in the Operating License Application and accordingly, the design-basis tornado missile spectrum will also be updated.

3.2.4.2.5 Combined Tornado Load Effects After tornado-generated wind pressure effects, atmospheric pressure change effects and missile impact effects are determined; the combination thereof will be established in accordance with procedures outlined n NUREG-800, Section 3.3.2. The effect of atmospheric pressure drop by itself will be considered, and the total effects of wind pressure and missile impact effects with one-half of the atmospheric pressure drop effects will be considered jointly.

3.2.4.3 Effect of Failure of Structures, Systems, or Components Not Designed for Tornado Loads SSCs, in which failure during a tornado event could affect the safety-related portions of the RPF, are either designed to:

  • Resist the tornado loading or the effect on the safety-related structures from the failure of these SS Cs
  • Be bounded by the tornado missile or aircraft impact evaluations The effects and mitigations of failure of SSCs not designed for tornado loads will be developed during final design and the Operating License Application.

3.2.5 Rain, Snow, and Ice Loading 3.2.5.1 Rain Loads 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 probable maximum precipitation (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."

Per NUREG-1537, Section 2.3 .1, "General and Local Climate," rain loads will be based on the estimate of the weight of the 48-hour (hr) probable maximum precipitation, as specified by the U.S. Geological Survey. This rain load estimate is compared with the local building code rain load (i.e., ASCE 7-10), and the greater value is used in design of the RPF roof.

The roof of the RPF is designed to prevent rainwater from accumulating on the roof. In accordance with 2012 IBC and ASCE 7-10, 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 48-hr probable maximum precipitation.

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  • ~ * .~ ~
  • NOmfWEU MEDtCAl tsOTOPl:S NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures , Systems and Components Rain loads are determined by the amount Table 3-18. Rain Load Criteria of water that can accumulate on the undeflected building roof if the primary Element Value drainage system becomes blocked (static Static head 5 cm (2-in) head), plus a uniform depth of water above Hydraulic head TBD the inlet of the secondary drainage system Rainfall intensity 3 .14 in./hr" at its design flow (hydraulic head). The
  • NOAA Atlas 14, Precipitation-Frequency Atlas of th e United rain load criteria are determined per States, Volume 8, Version 2.0: Midwestern States, Nati onal Oceanic and ASCE 7-10, Chapter 8, and are provided Atmospheri c Administration, Silver Spring, Maryland, 201 3.

in Table 3-1 8. TBD = to be determined.

The hydraulic head is dependent on the roof drain size, roof area drained, and the rainfall intensity. The rainfall intensity used to determine the hydraulic head is taken from NOAA Atlas 14, Precipitation-Frequency Atlas of the United States, web tool for the 100-year storm, 1-hr duration.

The rain load criteria will be updated in the Operating License Application .

3.2.5.2 Snow Load Per the guidance in DC/COL-ISG-007, two types of snow load are considered: normal snow load and the extreme winter precipitation load. The normal snow load will be included in normal load combinations given below. Per the guidance in the DC/COL-ISG-007, the extreme winter precipitation load is included in the extreme environmental load combinations.

The snow load criteria will be updated in the Operating License Application.

3.2.5.2.1 Normal Snow Load Per NUREG-1537, Section 2.3.1 and DC/COL-ISG-007, the normal snow load is the 100-year ground snow, modified using the procedures of ASCE 7 to determine the roof snow load, including snow drifting.

The 100-year ground snow load is calculated by factoring the ground snow load stipulated in the City of Columbia Code of Ordinances amendments (City of Columbia, 2014) and IBC 20 I 2 and is equivalent to the mapped ground snow load from Figure 7-1 of ASCE 7. This information is determined using the conversion factor provided in ASCE 7, Table C7-3 .

The exposure factor provided in ASCE 7, Table 3-19. Snow Load Criteria Table 7-2, for partially exposed roof in terrain category C is similar with the exposure used Element Value for determining wind loads. Since the RPF Mapped ground snow load (50-year) *20 lb/ft 2 does not fall into any of the special cases Conversion factor, I 00-year to 50-year b0.82 indicated in ASCE 7, Table 7-3 , the thermal Design ground snow load, pg (100-year) 24.4 lb/ft 2 factor is assumed to be 1.0. Exposure factor (Ce)

The importance factor is taken to be unity Thermal factor (Ci) from ASCE 7-10, Table 1.5-2, for the RPF, Importance factor which is designated Risk Category N. Snow

  • City of Columbia, "City of Columbia Code of Ordinances,"

load criteria are summarized in Table 3-19. www.gocolumbiamo.com/Council/Code_of_ Ordinances_PDF/,

accessed September 8, 2014.

b ASCE 7, Minimum Design Loads fo r Buildings and Other 3.2.5.2.2 Extreme Winter Precipitation Structures, American Society of Civil Engineers, Reston, Virginia, 2013 .

Load Per NUREG-1537, Section 2.3.1 and DC/COL-ISG-007, the extreme winter precipitation load is the normal snow load as presented in Section 3.2.5.2.1 , plus the liquid weight of the 48-hr probable maximum winter precipitation (PMWP) .

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, *! ~.* ~ : . NORTHWEST MEDtcAllSOTOrES The 48-hr PMWP is determined from the Table 3-20. Extreme Winter Precipitation Load NOAA/NWS Hydrometeorological Report Criteria (HR) 53, Seasonal Variation of 10-Square-Mile Probable Maximum Precipitation Element Value Estimates, United States East of the J05th 2 24-hr, I O-mi PMWP 46.7 cm (18 .2 in.)"

Meridian, for a 10-mi2 area. HR 53 gives mid- 2 72-hr, I O-mi PMWP 56.9 cm (22.5 in.)"

month PMP estimates for six 24- and 72-hr 2

durations. Using the NOAA web tool for 48-hr, I O-mi PMWP (interpolated) 22.2 cm (8.73 in.)

Columbia (NOAA, 2017), a two-day (48-hr) Weight of 48-hr PMWP 106 lb/ft 2 average 100-year rain is 8. 73 in. precipitation.

  • NWS/NOAA HR 53, Seasonal Variation of JO-Square-Mile To determine the PMWP, the months of Probable Maximum Precipitation Estimates, United States East of December, January, February, and March are the 105th Meridian , National Oceanic and Atmospheric considered. Using HR 53, Figures 26 through Administration, Silver Spring, Maryland, 1980.

45, the PMWP was determined to occur in the PMWP probab le max imum winter precipitation.

month of March. The PMWP criteria are given in Table 3-20.

Winter weather events since 1996 in Boone County, Missouri, are provided in Chapter 2.0, Table 2-36.

3.2.5.3 Atmospheric Ice Load Table 3-21. Atmospheric Ice Load Criteria Element Value 3 For SSCs to be considered sensitive to ice, the ice thickness and concurrent wind loads are Ice thickness (50-year) 2.54 cm (I in. )

determined using the procedures in ASCE 7, Concurrent wind speed 64.4 km/hr (40 mi/hr)

Chapter 10. Consistent with the requirements Ice thickness MRI multiplier 1.25 for snow and wind loads, the mapped values Wind speed MRI multiplier 1.00 are converted to 100-year values using the Importance factor 1.00 MRI multipliers given in ASCE 7, Table C 10-1.

Table 3-21. MRI = mean recurrence interval.

3.2.6 Operating Thermal/Self-Straining Loads The operating thermal/self-straining loads will be evaluated in the Operating License Application. These loads will be consistent with the requirements of ACI 349 or ANSI/AISC N690, as applicable to the material of construction.

3.2. 7 Operating Pipe Reaction Loads The operating pipe reaction loads will be evaluated in the Operating License Application. These loads will be consistent with the requirements of applicable American Society of Mechanical Engineers (ASME) B3 l , Standards of Pressure Piping, codes.

3.2.8 External Hazards External hazards include aircraft impact, external explosions, and external fire. The RPF is a production facility, as opposed to a nuclear power reactor, as such 10 CFR 50.150(a)(3) is interpreted to mean that the requirement for the aircraft impact assessment is not applicable to this facility. Sources of accidental external explosions have been considered and were found to not be an accident of concern. The RPF is constructed of robust, noncombustible materials, and adequate setbacks from transportation routes and landscaping consisting of fire fuels are provided such that externals fires are not an accident of concern.

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NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.3 WATER DAMAGE This section identifies the requirements and guidance for the water damage design of the RPF SSCs.

NUREG-1520 and ASCE 7, Chapter 5, provide guidance on flood protection of nuclear safety-related SSCs. Updates and development of technical specifications associated with the water damage design of the RPF SSCs will be provided in Chapter 14.0 as part of the Operating License Application.

3.3.1 Flood Protection This subsection discusses the flood protection measures that are applicable to safety-related SSCs for both external flooding and postulated flooding from failures of facility components containing liquid. A compliance review will be conducted of the as-built design against the assumptions and requirements that are the basis of the flood evaluation presented below.

Additional information is presented in Chapter 2.0, Section 2.4.3 and Chapter 13 .0, Section 13.2.6.4.

This as-built evaluation will be documented in a flood analysis report and be part of the Operating License Application.

3.3.1.1 Flood Protection Measures for Structures, Systems, and Components 3.3.1.1.1 Flooding from Precipitation Events Regional flooding from large precipitation events raising the water levels of local streams and rivers to above the 500-year flood level can have an adverse impact on the structure and SSCs 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 100-year storm for a I-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-year 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 231.6 m (760 ft) , or 16.8 m (55 ft) below the site. Since the site, 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-year flood plain, and is considered a dry site, the probable maximum flood from regional flooding is considered highly unlikely, without further evaluation. 1 1

The recommended standard for determining the probably maximum flood, ANS 2. 8, Determining Design Basis Flooding at Power Reactor Sites, has been withdrawn.

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' ~ -.. ~

  • HOITHWEST MEDtCAL tsOTOPH Per NUREG-1520, Section 3.2.3.4(1)(c), and ASCE 7, Chapter 5, flood loads will be based on the water level of the 100-year flood (one percent probability of exceedance per year). The facility has been determined to be above both the 100-year and the 500-year flood plain. Chapter 2, Section 2.4.3, provides additional detail for flood protection measures.

Postulated flooding from component failures in the building compartments will be prevented from adversely affecting plant safety or posing any hazard to the public. Exterior or access openings and penetrations into the RPF will be above the maximum postulated flooding level. Therefore, flood loads are considered highly unlikely and are not considered design loads.

3.3.1.1.2 Flooding from Inadvertent Discharge of Fire Protection System Water Design of fire suppression systems using water (e.g. , automatic sprinklers, hose stations) includes elements such as the grading and channeling of floors , raising of equipment mounts above floors ,

shelving and floor drains, and other passive means. These features will ensure sufficient capacity for gravity-driven collection and drainage of the maximum water discharge rate and duration to avoid localized flooding and resulting water damage to equipment within the area. In addition, particularly sensitive systems and components, whether electrical, optical, mechanical and/or chemical, are typically protected within enclosures designed for the anticipated adverse environmental conditions resulting from these types of water discharges. If critical for safety, these water-sensitive systems and components will be installed within the appropriate severe environment-rated enclosures in accordance with the relevant industry standard(s) (e.g. , National Electrical Manufacturers Association [NEMA] enclosure standards).

Selection of specific fire suppression systems for facility locations will be guided by recommendations in relevant industry standards (e.g., NFPA 801, Standard for Fire Protection for Facilities Handling Radioactive Materials) and will depend on the level of fire hazards at those locations, as determined from the final facility and process systems designs. These final detailed designs will include any facility design elements and sensitive equipment protection measures deemed necessary for addressing the maximum inadvertent rate and duration of water discharges from the fire protection systems. The final comprehensive facility design, along with commitments to design codes, standards, and other referenced documents (including any exceptions or exemptions to the identified requirements), will be identified and provided as part of the Operating License Application.

3.3.1.2 Flood Protection from External Sources Safety-related components located below-grade will be protected using the hardened protection approach.

The safety-related systems and components will be protected from external water damage by being enclosed in a reinforced concrete safety-related structure. The RPF will have the following characteristics:

  • Exterior safety-related walls below-grade will be 0.6 I m (2-ft) thick minimum
  • Water stops will be provided in all construction joints below-grade
  • Waterproof coating will be applied to external surfaces below-grade and as required above-grade
  • Roofs will be designed to prevent pooling of large amounts of water in accordance with Regulatory Guide 1.102, Flood Protection for Nuclear Power Plants Waterproofing of foundations and walls of safety-related structures below-grade will be accomplished primarily by the use of water stops at expansion and construction joints. In addition to water stops, waterproofing of the RPF will be provided to protect the external surfaces from exposure to water. The level above the RPF first level where waterproofing is to be used will be determined in the Operating License Application.

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  • . * *
  • NOllTIIWEST MEDICAi. ISOTOHS Chapter 3.0 - Design of Structures, Systems and Components The flood protection measures that are described above will also guard against flooding from the rupture of the on-site fire protection water storage tank (if future design development determines that a fire protection storage tank is necessary). Any flash flooding that may result from tank rupture will drain away from the RPF and thereby cause no damage to facility equipment.

3.3.1.3 Compartment Flooding from Fire Protection Discharge The total discharge from the failure of fire protection piping consists of the combined volume from any sprinkler and hose systems. The sprinkler system, if used, is capable of delivering a water density of 20 gallons per minute (gal/min) (76 liters per minute [L/min]) over a 139 m2 (1 ,500 ft 2) design area; therefore, the sprinkler system is calculated to have a flow rate of 1, 136 L/min (300 gal/min). The hose stream will be a manually operated fire hose capable of delivering up to 946 Umin (250 gal/min). In accordance with NFP A 801 , Section 5 .10, the credible volume of discharge is sized for the suppression system operating for a duration of 30 min. The design of water-sensitive, safety-related equipment will ensure that potential flooding from sprinkler discharge will not adversely affect the safety features. For example, equipment may be raised from the floor sufficiently such that the potential flooding due to sprinkler discharge will not impact the criticality analyses.

Outside of the radiologically controlled area (RCA), as defined in Chapter 11.0, "Radiation Protection and Waste Management, there is limited water discharge from fire protection systems. Any water-sensitive, safety-related equipment will be installed above the floor slab at-grade to ensure that the equipment remains above the flooded floor during sprinkler discharge.

3.3.1.4 Compartment Flooding from Postulated Component Failures Piping, vessels, and tanks with flooding potential in the safety-related portions of the RPF will be seismically qualified. Water-sensitive, safety-related equipment will be raised above the floor. The depth of water on the floor will be minimized by using available floor space to spread the flood water and limiting the water volumes. Analyses of the worst flooding due to pipe and tank failures and their consequences will be developed in the Operating License Application.

3.3.1.4.1 Potential Failure of Fire Protection Piping The total discharge from the operation of the fire protection system bounds the potential water collection due to the potential failure of the fire protection piping.

3.3.1.5 Permanent Dewatering System There is no permanent dewatering system provided for the flood design.

3.3.1.6 Structural Design for Flooding Since the design PMP elevation is at the finished plant-grade and the probable maximum flood (PMF) elevation is approximately 6.1 m (20 ft) below-grade, there is no dynamic force due to precipitation or flooding. The lateral surcharge pressure on the structures due to the design PMP water level is calculated and does not govern the design of the below-grade walls. The load from buildup of water due to discharge of the fire protection system in the RCA is supported by slabs-on-grade, with the exception of the mezzanine floor. Drainage is provided for the second level in the RCA to ensure that the second level slab is not significantly loaded. The second level slab is designed to a live load of 610 kilograms (kg)/m2 (125 lb/ft 2); therefore, the slab is capable of withstanding any temporary water collection that may occur while water is draining from that floor.

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.* ~ ~-~ ~ : NOmfWUT MEotc:Al ISOTOHS 3.4 SEISMIC DAMAGE Seismic analysis criteria used for the RPF will conform to IAEA-TECDOC-134 7, Consideration of External Events in the Design of Nuclear Facilities Other Than Nuclear Power Plants, with Emphasis on Earthquakes. This report provides requirements and guidance for the seismic design of nuclear facilities other than nuclear power plants. NUREG-0800 and other NRC Regulatory Guides provide additional detailed guidance for the seismic analysis and design of the RPF. Additional information is provided in Chapter 2.0, Section 2.5.4, and Chapter 13 .0, Section 13.2.6.5. Updates and development of technical specifications associated with the seismic damage design of the RPF SSCs will be provided in Chapter 14.0 as part of the Operating License Application.

3.4.1 Seismic Input 3.4.1.1 Design Response Spectra Safe-Shutdown Earthquake The NRC has recommended using Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, for radioisotopes production facilities (e.g. , IO CFR 50). NWMI will use a spectrum anchored to 0.20 g peak ground acceleration for the RPF design basis. Regulatory Guide 1.60 is not indexed to any specific soil type, with its frequency content sufficiently broad to cover all soil types .

Therefore, soil type for the RPF will not be a parameter used to determine the RPF ' s design response spectra. The composition of soil in which the RPF is embedded will be included in the soil-structure-interaction analysis as part of the building response analysis. This information will be provided in the final safety analysis report (FSAR) as part of Operating License Application.

This peak ground acceleration matches that of the University of Missouri Research Reactor and the Calloway Nuclear Generating Station, which both are within 80.5 km (50 mi) of the RPF, as suggested by the NRC staff during the November I 0, 2016 Public Meeting. The analysis procedure develops ground motion acceleration time histories that match or exceed the Regulatory Guide 1.60 spectrum as input to the building finite element model. Structural damping will follow the recommendations of Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, which range from about 3 to 7 percent.

Response spectra corresponding to the recommended damping values of Regulatory Guide 1.61 will be used to derive seismic loads. Damping varies depending on the type of SSC. Structural damping will follow Regulatory Guide 1.61 guidance (ranging from about 3 to 7 percent). Plotting response spectra at 5 percent damping for purposes of illustration is a convention within the nuclear industry, but for analysis loads, damping will vary depending on the earthquake level (operating basis earthquake or safe-shutdown earthquake) and the type of SSC.

Soil-Structure Interaction and Dynamic Soil Pressures The structure is supported on a shallow foundation system on stiff competent soils. The Phase 1 Assessment (Terracon, 201 la/b) stated the site is classified as Site Class C. Prescribed in ASCE 7, Table 20.3-1 , the typical shear wave velocities for the soils present at the site are 1,200 to 2,500 ft/sec.

Typical practice is to define competent soil as having a shear wave velocity greater than 1,000 ft/sec. The analysis of the RPF building structure to the safe shutdown earthquake will include the effects of a soil-structure interaction. Dynamic soil pressures were determined using ASCE 4, Seismic Analysis of Safety-Related Nuclear Structures and Commentary, Section 3.5.3.2, and applied to the earth retaining walls in the hot cell area.

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~. NOITKW£ST MEDICAL ISOTOPO NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components Operating Basis Earthquake For preliminary design, the operating basis earthquake was selected to be one-third the safe-shutdown earthquake defined previously (based on Regulatory Guide 1.6 1). Since this option was selected, explicit design and analysis of the facility structure for the operating basis earthquake ground-motion is not required.

3.4.1.2 Method of Analysis The effect of loads other than earthquake-induced (seismic) loads is determined by static analysis methods in accordance with ASCE 7 and the fundamental principles of engineering. Seismic analysis of SSCs will be performed by either equivalent-static methods or dynamic analysis methods in accordance wi th ASCE 4 and ASCE 43, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities. The equivalent-static and dynamic seismic analysis methods are discussed below.

3.4.1.2.1 Equivalent-Static Analysis Equivalent-static seismic analysis of commercial type structure will be performed in accordance with ASCE 7, Section 12.8.

Direction of Seismic Loading Design ofIROFS will consider seismic loads in all three directions using a combination of square-root-of-the-sum-of-squared or 10/40/40 methodologies per Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis. The I 0/40/40 methodology will be used in the development of the final RPF design and included as part of the Operating License Application .

3.4.1.2.2 Dynamic and Static Analysis Dynamic analyses are only used for the evaluation of RPF structural components. A static analysis will be completed during final design by using a combination of static load computations to ensure the SSCs remain in place and intact, and a combination of existing shake table test data and existing earthquake experience to ensure that the equipment functions following the earthquake. The analysis of safety-related structures may be either completed by the:

  • Linear-elastic response spectra method performed in accordance with ASCE 4, Section 3.2.3.1, and ASCE 43 , Section 3.2.2
  • Linear-elastic time history method performed in accordance with ASCE 4, Section 3.2.2, and ASCE 43 , Section 3.2.2 Damping - The damping values used for dynamic analysis for the structural system considered will be taken from Regulatory Guide 1.6 1. Inelastic energy adsorption factors and damping values used for the analysis of nuclear safety-related structures will be selected from ASCE 43 , Table 5-1.

Modeling - Finite element models will only be used for the RPF building structures. The mesh for plate elements and member nodes will be selected to provide adequate discretization and distribution of the mass. Further, the aspect ratio of plate elements will be limited to no greater than 4: l to ensure accurate analysis results.

Direction of seismic loading - Three orthogonal directions of seismic loading are used in the RPF design, two horizontal and one vertical. The modal components of the dynamic analysis and the spatial components ofresponse analysis are combined as described below.

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  • Modal combinations -The structure of the RPF is designed to be relatively stiff, and components are combined using the complete quadratic combination method.
  • Spatial component combinations - Spatial components are calculated separately and combined using the square-root-sum-of-the-squares method to determine the combined earthquake effect and resulting demands.

3.4.2 Seismic Qualification of Subsystems and Equipment This subsection discusses the methods by which the RPF systems and components are qualified to ensure functional integrity. Based on the characteristics and complexities of the subsystem or equipment, seismic qualification will be done by a combination of static load computations to ensure that the SSCs remain in place and intact, and a combination of existing shake table test data and existing earthquake experience to ensure that the equipment functions following the earthquake.

3.4.2.1 Qualification by Analysis NWMI will define specific acceptable qualification methods in the procurement packages to demonstrate seismic qualifications. Seismic qualification of IROFS will include three options of: (1) calculations and verification that the main structural components of the SSC can withstand the seismic loads derived from the in-structure floor response spectra at the damping value derived from Regulatory Guide 1.61 ,

(2) reference to available shake table testing that demonstrates the seismic capacity of the SSC or of multiple similar items, and (3) demonstration of the seismic capacity through the performance of the type of SSC in actual earthquakes.

3.4.2.1.1 Equivalent Static Analysis The equivalent static analysis of nuclear safety-related subsystems and equipment is performed in accordance ASCE 43, Section 8.2.1.1. The equivalent static analysis of subsystems and equipment that are not relied on for nuclear safety but are designated as a component of a seismic system per IBC 2012, Chapter 17, is performed in accordance with ASCE 7, Chapter 13 .

3.4.2.1.2 Static Analysis The static analysis of non-structural, safety-related subsystems and equipment is performed in accordance ASCE 4, Section 3.2.3 .1, and ASCE 43 , Section 8.2.1.2. A portion of the seismic qualification process will involve simple static analysis of the main structural elements (anchorage and primary framing) of IROFS components, using seismic loads from in-structure response spectra derived from the RPF building structure dynamic response analysis. In-structure response spectra are determined using ASCE 4, Section 3 .4.2, and NRC Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components. In-structure floor response spectra will be developed through a finite element model of the RPF building using an artificial time history that matches or envelops the Regulatory Guide 1.60 spectrum at a peak ground acceleration = 0.20 g.

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  • NORTHWEST MEDtCAL ISOTOPES 3.4.2.2 Qualification by Testing NWMI will define specific acceptable qualification methods in the procurement packages to demonstrate seismic qualifications. Seismic qualification of IROFS will include three options of: (1) calculations and verification that the main structural components of the SSC can withstand the seismic loads derived from the in-structure floor response spectra at the damping value derived from Regulatory Guide 1.61, (2) reference to available shake table testing that demonstrates the seismic capacity of the SSC or of multiple similar items, and (3) demonstration of the seismic capacity through the performance of the type of SSC in actual earthquakes.

Per NRC Regulatory Guide 1.100, Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification ofActive Mechanical Equipment for Nuclear Power Plants:

  • Active mechanical equipment relied on for or important to nuclear safety will be required to be seismically qualified in accordance with Regulatory Guide 1.100.
  • Active electrical equipment important to or relied on for nuclear safety will be required to be seismically qualified in accordance with IEEE 344, IEEE Standard for Seismic Qualification of Equipment for Nuclear Power Generating Stations.

Subsystems and equipment not relied on for nuclear safety but designated as a component of a seismic system per IBC 2012, Chapter 17, will be required. Existing databases of past shake table tests will be used, such as the Office of Statewide Health Planning and Development database provided by the state of California. These tests have typically been done based on the ICC-ES AC156, "Acceptance Criteria for Seismic Certification by Shake-Table Testing of Nonstructural Components," spectrum.

The capacity of the standard support design for overhead fixtures mounted above RPF IROFS will be checked to ensure that the supports can withstand the seismic loads derived from the floor spectra (e.g.,

remain stable during and after postulated earthquake effects) of the attachment floor slab. This information will be provided in the FSAR as part of the Operating License Application.

The RPF seismic design will also include a check to ensure that pounding or sway impact will not occur between adjacent fixtures (e.g., rattle space). Estimates of the maximum displacement of any fixture can be derived from the appropriate floor response spectrum and an estimate of the fixture 's lowest response frequency. This information will be provided as part of the Operating License Application.

3.4.3 Seismic Instrumentation Seismic recording instrumentation will be triaxial digital systems that record accelerations versus time accurately for periods between 0 and 10 sec. Recorders will have rechargeable batteries such that if there is a loss of power, recording will still occur. All instrumentation will be housed in appropriate weather and creature-proofed enclosures. As a minimum, one recorder should be located in the free-field mounted on rock or competent ground generally representative of the site. In addition, at sites classified as Seismic Design Category D, E, or Fin accordance with ASCE 7, Chapter 11 , using Occupancy Category IV, recorders will be located and attached to the foundations and roofs of the RPF and in the control room.

The systems will have the capability to produce motion time histories. Response spectra will be computed separately.

The purpose of the instrumentation is to (1) permit a comparison of measured responses of C-1 structures and selected components with predetermined results of analyses that predict when damage might occur, (2) permit facility operators to understand the possible extent of damage within the facility immediately following an earthquake, and (3) be able to determine when an safe-shutdown earthquake event has occurred that would require the emptying of the tank(s) for inspection as specified in NFPA 59A, Standard for the Production, Storage, and Handling of Liquefied Natural Gas , Section 4.1.3.6(c ).

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. *~ ~.~! :. NORTHW(n 11(01CA1.ISOTOP£S Chapter 3.0 - Design of Structures, Systems and Components Seismic instrumentation for the RPF site is not an IROFS; it provides no safety function and is therefore not "safety-related." Although the seismic recorders have no safety function, they must be designed to withstand any credible level of shaking to ensure that the ground motion would be recorded in the highly unlikely event of an earthquake. This capability requires verification of adequate capacity from the manufacturer (e.g., prior shake table tests of their product line), provision of adequate anchorage (e.g.,

manufacturer-provided anchor specifications to ensure accurate recordings), and a check for seismic interaction hazards such as water spray or falling fixtures. With these design features , the instrumentation would be treated as if it were safety-related QL-2. Additional information on seismic instruction will be provided as part of the Operating License Application.

3.4.3.1 Location and Description Seismic instrumentation is installed for structural monitoring. The seismic instrumentation consists of solid-state digital, tri-axial strong motion recorders located in the free-field, at the structure base, and at the roof of the RPF.

3.4.3.2 Operability and Characteristics The seismic instrumentation operates during all modes of RPF operations. The maintenance and repair procedures provide for keeping the maximum number of instruments in service during RPF operations.

The instrumentation installation design includes provisions for in-service testing. The instruments selected are capable of in-place functional testing and periodic channel checks during normal facility operation.

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  • NORTHWEST llEDtCAl ISOTGnS 3.5 SYSTEMS AND COMPONENTS Certain systems and components of the RPF are considered important to safety because they perform safety functions during normal operations or are required to prevent or mitigate the consequences of abnormal operational transients or accidents. This section summarizes the design basis for design, construction, and operating characteristics of safety-related SS Cs of the RPF.

3.5.1 General Design Basis Information 3.5.1.1 Classification of Systems and Components Important to Safety The RPF systems and components will be classified according to their importance to safety, quality levels, and seismic class. The guidance used in developing these classifications during preliminary design with the support ofregulatory guidance reviews, hazards and operability analysis, accident analysis, integrated safety analysis, and national consensus code requirements is presented below.

The RPF systems identified in Table 3-1 and their associated subsystems and components are discussed in the subsections that follow.

3.5.1.2 Classification Definitions The definitions used in the classification of SSCs include the following.

In accordance with 10 CFR 50.2, "Definitions," design basis refers to information that identifies the specific functions to be performed by an SSC of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be:

  • Restraints derived from generally accepted state-of-the-art practices for achieving functional goals
  • Requirements derived from analysis (e.g. , calculation, experiments) of the effects of a postulated accident for which a SSC must meet its functional goals These reference bounds are to include the bounding conditions under which SSCs must perform design basis functions and may be derived from normal operation or any accident or events for which SSCs are required to function , including anticipated operational occurrences, design basis accidents, external events, natural phenomena, and other events specifically addressed in the regulations.

Safety-related is a classification applied to items relied on to remain functional during or following a design basis event (DBE) to ensure the:

  • Integrity of the facility infrastructure
  • Capability to shut down the facility and maintain it in a safe-shutdown condition
  • Capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the applicable guideline exposures set forth in 10 CFR 70.61 ,

"Performance Requirements," as applicable Design basis accident is a postulated accident that a nuclear facility must be designed and built to withstand, without loss to the SSCs necessary to ensure public health and safety.

Design basis event (DBE) is an event that is a condition of normal operation (including anticipated operational occurrences), a design basis accident, an external event, or natural phenomena for which the facility must be designed so that the safety-related functions are achievable.

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' ~ *.*! . NOkTHWEST MEOK:Al lSOTOl'lS Design basis accidents and transients are those DBEs that are accidents and transients and are postulated in the safety analyses. The design basis accidents and transients are used in the design of the facility to establish acceptable performance requirements for SSCs.

Single failure is considered a random failure and can include an initiating event (e.g. ,

component failure, natural phenomenon, external man-made hazard) or consequential failures .

Mechanical, instrumentation, and electrical systems and components required to perform their intended safety function in the event of a single failure are designed to include sufficient redundancy and independence. This type of design verifies that a single failure of any active component does not result in a loss of the capability of the system to perform its safety functions.

Mechanical, instrumentation, and electrical systems and components are designed to ensure that a single failure, in conjunction with an initiating event, does not result in the loss of the RPF's ability to perform its intended safety function. Design techniques such as physical separation, functional diversity, diversity in component design, and principles of operation, will be used to the extent necessary to protect against a single failure.

An initiating event is a single occurrence, including its consequential effects, that places the RPF (or some portion) in an abnormal condition. An initiating event and its resulting consequences are not considered a single failure .

Active components are devices characterized by an expected significant change of state or discernible mechanical motion in response to an imposed demand on the system or operation requirements (e.g., switches, circuit breakers, relays, valves, pressure switches, motors, dampers, pumps, and analog meters). An active component failure is a failure of the component to complete its intended safety function(s) on demand.

Passive components are devices characterized by an expected negligible change of state or negligible mechanical motion in response to an imposed design basis load demand on the system.

Defense-in-depth is an approach to designing and operating nuclear facilities that prevents and mitigates accidents that release radiation or hazardous material through the creation of multiple independent and redundant layers of defense to compensate for potential human and mechanical failures so that no single layer, no matter how robust, is exclusively relied on. Defense-in-depth includes the use of access controls, physical barriers, redundant and diverse key safety functions , and emergency response measures.

The RPF structure and system designs are based on defense-in-depth practices. The RPF design incorporates:

  • Preference for engineered controls over administrative controls
  • Independence to avoid common mode failures
  • Other features that enhance safety by reducing challenges to safety-related components and systems Safety-related systems and components identified in this section are described in Chapters 4.0; 5.0, "Coolant Systems;" 6.0; 7.0; 8.0, "Electrical Power Systems;" and 9.0, "Auxiliary Systems," as appropriate.

3.5.1.3 Nuclear Safety Classifications for Structures, Systems, and Components SSCs in the RPF are classified as safety-related and non-safety-related. The safety-related SSCs include IROFS to meet the performance requirement of 10 CFR 70.61 and other safety-related SSCs to meet the requirements of 10 CFR 20. The purpose of this section is to classify SSCs according to the safety function being performed.

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0 HOllTHWEST MEOtCAl tsOlWH Chapter 3.0 - Design of Structures, Systems and Components In addition, design requirements wil I be placed on SS Cs to ensure the proper performance of their safety function , when required.

  • Safety-related is a classification applied to items relied on to remain functional during or following a postulated DBE to ensure the:

Integrity of the facility infrastructure (e.g. , water, sewer, electricity)

Capability to shut down the facility and maintain it in a safe shutdown condition Capability to prevent or mitigate the consequences of postulated accidents identified through accident analyses that could result in potential offsite and worker exposures comparable to the applicable guideline exposures set forth in 10 CFR 70.6l(b), 10 CFR 70.6l(c), and 10 CFR 70.61 (d)

Operation of the facility without undue risk to the health and safety of workers, the public, and the environment to meet 10 CFR 20 normal release or exposure limits for radiation doses and applicable limits for chemical exposures

  • Safety-related Non-IROFS - SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of workers, the public, and environment, and includes SSCs to meet 10 CFR 20 normal release or exposure limits.
  • Non-safety-related - SSCs related to the production and delivery of products or services that are not in the above safety classifications 3.5.1.3.1 Quality Group Classifications for Structures, Systems, and Components The assignment of safety-related classification and use of codes and standards conforms to the requirements NWMI ' s Quality Assurance Program Plan (QAPP) for the development of a Quality Group classification and the use of codes and standards. The classification system provides a recognizable means of identifying the extent to which SSCs are related to safety-related and seismic requirements, including ANS nuclear safety classifications, NRC quality groups, ASME Code Section III classifications, seismic categories, and other applicable industry standards, as shown in Table 3-7.

Quality assurance (QA) requirements are defined in the NWMI QAPP (Chapter 12.0, "Conduct of Operations," Appendix C) . The definitions of QA Levels 1, 2, and 3 are provided below.

QA Level 1 will implement the full measure of the QAPP and will be applied to IROFS. IROFS are QA Level 1 items in which failure or malfunction could directly result in a condition that adversely affects workers, the public, and/or environment, as described in 10 CFR 70.61. The failure of a single QA Level 1 item could result in a high or intermediate consequence. The failure of a QA Level 2 item, in conjunction with the failure of an additional item, could result in a high or intermediate consequence. All building and structural IROFS associated with credible external events are QA Level 1. QA Level 1 items also include those attributes of items that could interact with IROFS due to a seismic event and result in high or intermediate consequences, as described in 10 CFR 70.61. Examples include:

  • Items to prevent nuclear criticality accidents (e.g., preventive controls and measures to ensure that under normal and credible abnormal conditions, all nuclear processes are subcritical)
  • Items credited to withstand credible design-bases external events (e.g. , seismic, wind)
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. * ~ ~-~! ; NomtWfSTMEOtCAl.tSOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components QA Level 2 will be applied to non-QA Level 1 safety SSCs. The QA program is important to the acceptability and suitability of the item or service to perform as specified. Acceptance methods shall be specified (including acceptance and other applicable performance criteria), documented, and verified before use of the item or service. Some of the required characteristics may be examined less rigorously than for QA Level 1. Examples of QA Level 2 items include:

  • Fire protection systems
  • Safeguards and security systems
  • Material control and accountability systems QA Level 3 will include non-safety-related quality activities performed by NWMI that are deemed necessary to ensure the manufacture and delivery of highly reliable products and services to meet or exceed customer expectations and requirements. QA Level 3 items include those items that are not classified as QA Level 1 or QA Level 2. QA Level 3 items are controlled in accordance with standard commercial practices.

These quality activities are embodied in NWMI's QAPP and will be further specified in the Operating License Application, and when necessary.

3.5.1.3.2 Seismic Classification for Structures, Systems, and Components SSCs identified as IROFS will be designed to satisfy the general seismic criteria to withstand the effects of natural phenomena (e.g., earthquakes, tomados, hurricanes, floods) without loss of capability to perform their safety functions. ASCE 7, Chapter 11 , sets forth the criteria to which the plant design bases demonstrate the capability to function during and after vibratory ground-motion associated with the safe-shutdown earthquake conditions.

The seismic classification methodology used for the RPF complies with the preceding criteria, and with the recommendations stated in Regulatory Guide 1.29, Seismic Design Classification. The methodology classifies SSCs into three categories: seismic Category I (C-I), seismic Category II (C-11), and non-seismic (NS).

Seismic C-1 applies to both functionality and integrity, while C-11 applies only to integrity. SSCs located in the proximity of IROFS, the failure of which during a safe-shutdown earthquake could result in loss of function of IROFS, are designated as C-11. Specifically:

  • C-I applies to IROFS. C-I also applies to those SSCs required to support shutdown of the RPF and maintain the facility in a safe shutdown condition
  • C-11 applies to SSCs designed to prevent collapse under the safe-shutdown earthquake. SSCs are classified as C-11 to preclude structural failure during a safe-shutdown earthquake, or where interaction with C-1 items could degrade the functioning of a safety-related SSC to an unacceptable level or could result in an incapacitating injury to occupants of the main control room.
  • NS SSCs are those that are not classified seismic C-1 or C-11.

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  • NORTHWEST llBHCAl ISOTOP'El NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.5.2 Radioisotope Production Facility Systems and components within the RPF are presented in Section 3.5.1. The RPF design basis evaluated the general design criteria from 10 CFR 70.64, "Requirements for New Facilities or New Processes at Existing Facilities." This evaluation is presented in Table 3-22. These general design criteria provide a rational basis from which to initiate design but are not mandatory. There are some cases where conformance to a particular criterion is not directly measurable. For each of the criteria, a specific assessment of the RPF design is made, and a complete list of references is included to identify where detailed design information pertinent to each criterion is treated. The Chapter 13 .0 accident sequences for credible events define the DBE. The safety-related parameter limits ensure that the associated design basis is met for the events presented in Chapter 13 .0.

Table 3-22. Design Criteria Requirements (4 pages)

Design criteria and description Application and compliance 10 CFR 70.64, "Requirements for New Facilities or New Processes at Existing Facilities""

Quality standards and records

  • SSCs important to safety will be designed, fabricated, erected, tested, operated,
  • Develop and implement design in and maintained to quality standards commensurate with the importance of the accordance with management safety functions to be performed. Where generally recognized codes and standards are used, they will be identified and evaluated to determine their measures to ensure that IROFS are applicability, adequacy, and sufficiency and will be supplemented or modified as available and reliable to perform their function when needed. necessary to ensure a quality product in keeping with the required safety function.
  • Maintain appropriate records of these items by or under the control of the
  • NWMl's QAPP will be established and implemented to provide adequate licensee throughout the life of the assurance that SSCs satisfactori ly perform their safety functions.

facility.

  • Appropriate records of design, fabricatio n, erection, and testing of SSCs important to safety will be main tained by or under control ofNWMJ for the life of RPF.
  • NWMI will use a graduated QAPP that links quality classification and associated documentation to safety classification and to the manufacturing and delivery of highly reliable products and equipment.
  • The WMI QAPP will provide details of the procedures to be applied, including quality and safety level classifications.

Natural phenomena hazards

  • SSCs important to safety will be designed, fabricated, erected, tested, operated, Provide for adequate protecti on against and maintained to quality standards commensurate with the importance of the natural phenomena, with consideration safety functions to be performed. Where generall y recogn ized codes and of the most severe documented standards are used, they will be identified and evaluated to determine their historical events for the site. applicabi li ty, adeq uacy, and sufficiency and will be supplemented or modified as necessary to ensure a quality product in keeping with the required safety function .
  • The design basis for these SSCs wi II include :

- Appropri ate consideration of the most severe natural phenomena that have been historically reported for the RPF site and surro undin g area , including sufficient margin for limited accuracy, quantity, and period of time for which historical data has been accumulated

- Appropriate combinations of natural phenomena effects during normal and accident operating conditions

- Importance of the safety functions to be performed

  • Specific RPF des ign criteria and NRC general design criteria are discussed in Sections 3.1 and 3.5 , respectively.

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Design criteria and description Application and compliance Fire protection

  • SSCs important to safety will be designed and located throughout the RPF to Provide for adequate protection against minimize, consistent with other safety requirements, the probability and effect of fires and explosions fires and explosions.
  • Noncombustible and heat resistant materials will be used wherever practical throughout the RPF, particularly in locations such as confinement and the control room.
  • Fire detection and suppression systems of appropriate capacity and capability will be provided and designed to minimize the adverse effects of fires on SSCs important to safety.
  • Firefighting systems will be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these SSCs.
  • Where necessary, within zoned areas or where criticality and access are an issue, required systems will be manually initiated by operations after review ofa detection signal.
  • RPF fire protection system will be designed such that a failure of any component will not impair the ability of safety-related SSCs to safely shut down and isolate the RPF or limit the release of radioactivity to provide reasonable assurance that the public will be protected from radiological risks resulting from RPF operations
  • RPF fire protection system will be designed to provide reasonable assurance that the public will be protected from radiological risks resulting from RPF operations (e.g., failure of any component will not impair the ability of safety-related SSCs to safely shutdown and isolate the RPF or limit the release of radioactivity).
  • Chapters 6.0 and 9.0 provide additional information.

Environmental and dynamic effects

  • SSCs important to safety are designed to accommodate effects of, and to be Provide for adequate protection from compatible with, the environmental conditions associated with normal operation, environmental conditions and dynamic maintenance, testing, and postulated accidents. Due to low temperature and effects associated with normal pressure RPF processes, dynamic effects due to pipe rupture and discharging operations, maintenance, testing, and flu ids are not applicable to the RPF.

postulated accidents that could lead to loss of safety functions Chemical protection

  • Chemical protection in the RPF will be provided by confinement isolation Provide for adequate protection against systems, liquid retention features, and use of appropriate personal protective chemical risks produced from licensed equipment.

material, facility conditions that affect

  • Chapter 6.0, Section 6.2.1 , provides additional information.

the safety of licensed material, and hazardous chemicals produced from licensed material Emergency capability

  • Emergency procedures will be developed and maintai ned for the RPF to control Provide fo r emergency capability to SNM and hazardous chemicals produced from the SNM.

maintain control of:

  • Licensed material and hazardous chemi cals produced from licensed material
  • Evacuation of on-site personnel
  • On-site emergency fac ilities and services that fac ilitate the use of available off-site services 3-48

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Design criteria and description Application and compliance Utility services

  • The RPF is designed for passive, safe shutdown and to prevent uncontrolled Provide for continued operation of release of radioactive material if normal electric power is interrupted or lost.

essential utility services

  • A standby diesel generator will be provided for asset protection of selected RPF systems.
  • Uninterruptable power supplies will automatically provide power to systems that support the safety functions protecting workers and the public.
  • A combination ofuninterruptable power supplies and a standby electrical power system will provide emergency electrical power to the RPF. A 1,000 kW (1,34 l hp) diesel generator will provide facility electric power.
  • Chapter 8.0, Section 8.2 provides additional information.

Inspection, testing, and maintenance

  • The RPF is designed to provide access and controls fo r testing, maintenance, and Provide for adequate inspecti on, testin g, inspection of safety-related SSCs, as needed, throughout the RPF.

and maintenance ofIROFS to ensure

  • Chapters 4.0, 6.0, 7.0, and 9.0 provide additional information .

ava ilability and reliability to perform their function when needed Criticality control

  • The RPF design will provide adequate protection against criticality hazards related Provide for criticality control, including to the storage, handling, and processing of SNM , which will be accomplished by:

adherence to the double-contingency - Including equipment, facilities, and procedures to protect worker and public principle health and to minimize danger to life or property Ensuring that the design provides for criticality control, including adherence to the double-contingency principle Incorporating a criticality monitoring and alarm system into the facility design

  • Compliance with the requirements of criticality control, including adherence to the double-contingency principle, are described in detail in Chapter 6.0, Section 6.3 .

Instrumentation and control

  • RPF SNM processes will be enclosed predominately by hot cells and glovebox The design must provide for inclus ion of designs except for the target fabrication area.

l&C systems to monitor and control the

  • The FPC system will provide mon itoring and control of safety-related components behavior of items relied on for safety. and process systems within the RPF.
  • The BMS (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems.
  • ESF systems will operate independently from the FPC system or BMS . Each ESF safety function will use hard-wired analog controls/ interlocks to protect workers, the public, and environment. The ESF parameters and alarm function s wi ll be integrated into and monitored by th e FPC system or BMS.
  • RPF designs are based on defense-in-depth practices and incorporate a preference for engineered controls over admini strative controls, independence to avoid common mode fai lures, and incorporate other features that enhance safety by reducing challenges to safety-related components and systems.
  • The FPC system will provide the capabi li ty to monitor and control the behavior of safety-related SSCs. These systems ensure adeq uate safety of process and utili ty service operations in connection with their safety function. Controls are provided to maintain th ese variables and systems with in the prescribed operating ranges under all normal condi tions.
  • The FPC system is designed to fail to a safe-state or to assume a state demonstrated to be acceptable if conditions such as loss of signal, loss of energy or motive power, or adverse environments are experienced.
  • Chapter 7.0 provides additional I&C system information . Safety-related SSCs are described in Section 3.5 and Chapters 4.0, 5.0, 6.0, 7.0, and 8.0.

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  • NORTHWEST liEMCAL. tsOTOPU Chapter 3.0 - Design of Structures, Systems and Components Table 3-22. Design Criteria Requirements (4 pages)

Design criteria and description Application and compliance Defense-in-depthb

  • Defense-in-depth is a design philosophy that NWMI has applied from the Base facility and system design and beginning of the project and will continue through completion of a design that is facility layout on defense-in-depth based on providing successive levels of protection such that health and safety are practices. The design must incorporate, not wholly dependent on any single element of the design, construction, to the extent practicable: maintenance, or operation of the RPF.
  • Preference for the selection of
  • NWMJ's risk insights obtained through performance of the accident analysis will engineered controls over be used to supplement the final design by focusing attention on the prevention and administrative controls to increase mitigation of the higher risk potential accidents.

overall system reliability

  • Chapter 6.0 and 13.0 provide additional information.
  • Features that enhance safety by reducing challenges to IROFS
  • I 0 CFR 70.64, "Requirements for New Facilities or New Processes at Existing Facilities," Code of Federal Regulations, Office of the Federal Register, as amended.

b As used in I 0 CFR 70.64, requirements fo r new faciliti es or new processes at existing fa cili ties, defense-in-depth practices means a design philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety will not be wholl y dependent on any single element of the design, construction, maintenance, or operation of the fac ility. The net effect of incorporating defense-in-depth practices is a conservati vely designed faci li ty and system that wi ll exhibit greater to lerance to fa ilures and external challenges.

BMS building management system. NRC U.S. Nuclear Regulatory Commission.

CFR Code of Federal Regulations. NWMI Northwest Medical Isotopes, LLC.

ESF engineered safety fea ture. QAPP quality assurance program plan.

FPC fac ility process control. RPF Radioisotope Production Facility.

l&C instrumentation and control. SNM special nuclear material.

IROFS items relied on for safety. SSC structures, systems, and components.

The criteria are generic in nature and subject to a variety of interpretations; however, they also establish a proven basis from which to provide for and assess the safety of the RPF and develop principal design criteria. The general design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for SSCs important to safety (i .e. , SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of workers, the public, and environment).

Safety-related SSCs that are determined to have safety significance for the RPF will be designed, fabricated, erected, and tested as required by the NWMI QAPP, described in Chapter 12.0, Appendix C. In addition, appropriate records of the design, fabrication , erection, procurement, testing, and operations of SSCs will be maintained throughout the life of the plant.

The RPF design addresses the following:

  • Radiological and chemical protection
  • Natural phenomena hazards
  • Fire protection
  • Environmental and dynamic effects
  • Emergency capability (e.g., licensed material, hazardous chemicals, evacuation of on-site personnel, on-site emergency facilities/off-site emergency facilities)
  • Utility services
  • Inspection, testing, and maintenance
  • Criticality safety 3-50

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  • Instrumentation and controls
  • Defense-in-depth Safety-related systems and components will be qualified using the applicable guidance in the Institute of Electrical and Electronics Engineers (IEEE) Standard IEEE 323, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations. The qualification of each safety-related system or component needs to demonstrate the ability perform the associated safety function :
  • Under environmental and dynamic service conditions in which they are required to function
  • For the length of time the function is required Additionally, non-safety-related components and systems will be qualified to withstand environmental stress caused by environmental and dynamic service conditions under which their failure could prevent satisfactory accomplishment of the safety-related functions.

The RPF instrumentation and control (l&C) system (also known as the facility process control [FPC]

system) will provide monitoring and control of the process systems within the RPF that are significant to safety over anticipated ranges for normal operations and abnormal operations. The FPC system will perform as the overall production process controller. This system will monitor and control the process instrumented functions within the RPF, including monitoring of process fluid transfers and controlled inter-equipment pump transfers of process fluids.

The FPC system will also ensure that process and utility systems operate in accordance with their safety function. Controls will be provided to maintain variables and systems within the prescribed operating ranges under all normal conditions. In addition, the FPC system is designed to fail into a safe state or to assume a state demonstrated to be acceptable if conditions such as loss of signal, loss of energy or motive power, or adverse environments are experienced.

The building management system (BMS) (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (turn on and off) the mechanical utility systems.

ESF systems will operate independently from the FPC system or BMS. Each ESF safety function will use hard-wired analog controls/interlocks to protect workers, the public, and environment. The ESF parameters and alarm functions will be integrated into and monitored by the FPC system or BMS.

The fire protection system will have its own central alarm panel. The fire protection system will report the status of the fire protection equipment to the central alarm station and the RPF control room.

This integrated control system will be isolated from safety-related components consistent with IEEE 279, Criteria for Protection Systems for Nuclear Power Generating Stations. In addition, the RPF is designed to meet IEEE 603, Standard Criteria for Safety Systems for Nuclear Power Generating Stations, for separation and isolation of safety-related systems and components. Chapter 7 .0 provides additional details on the integrated control system.

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  • ! ~. ~~ ' NOmfWEST MEDICAl 1.SOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.5.2.1 System Classification Table 3-23. System Classifications The RPF is classified as a non-reactor Classification nuclear production faci lity per 10 CFR 50. description Classification Source In addition, a portion of the RPF will Hazard category Intermediate hazard NRC fabricate LEU targets, similar to fuel Occupancy type Mixed, A-2 , B, F-1, me 2012*

fabrication per 10 CFR 70. Due to the H-3 and H-4 nature of the work performed within faci lity, a hazardous occupancy applies. Construction type II-B IBC 2012*

Table 3-23 provides the RPF classification Risk category IV ASCE 7b for hazards occupancy, construction, risk, Seismic design category c ASCE 7b and seismic design categories.

  • IBC 20 12, " International Bui lding Code," as amended, International Code Council, Inc., Washington, D.C. , February 20 12.

3.5.2.2 Classification of Systems and b ASCE 7, Minimum Design Loads for Buildings and Other Components Important to Structures, American Society of Civil Engineers, Reston, Virginia, 20 13.

Safety NRC = U.S. Nuclear Regulatory Commission .

RPF SSCs, including their foundations and supports, designed to remain functional in the event of a DBE are designated as C-1. SSCs designated IROFS are also classified as C-1. SSCs co-located with C-1 systems are reviewed and supported in accordance with II over I criteria. This avoids any unacceptable interactions between SSCs.

C-1 structures should be designed using dynamic analysis procedures, or when justified, equivalent static procedures using both horizontal and vertical input ground motions. For dynamic analyses, either response spectra or time history analyses approaches may be used. Dynamic analysis should be performed in accordance with the procedures of ASCE 4, with the exception of the damping limitations presented in Section 3 .4.1.

Table 3-24 lists the RPF SSCs and associated safety and seismic classifications and quality level group for the top-level systems. Subsystems within these systems may be identified with lower safety classifications. For example, the day tanks of the chemical supply system are IROFS, while the rest of the chemical supply system is classified as safety-related or not-safety-related.

Table 3-24. System Safety and Seismic Classification and Associated Quality Level Group (2 pages)

Highest safety Seismic Quality level System name (code) classification* classificationb group Facility structure (RPF) IROFS C-1 QL-1 Target fabrication (TF) IROFS C-1 QL-1 Target receipt and disassembly (TD) IROFS C-I QL-1 Target dissolution (DS) IROFS C-I QL-1 Mo recovery and purification (MR) IROFS C-1 QL-1 Uranium recovery and recycle (UR) IROFS C-1 QL-1 Waste handling (WH) IROFS C-1 QL-1 Criticality accident alarm (CA) IROFS C-1 QL-1 Radiation monitoring (RM) IROFS C-1 QL- 1 Standby electrical power (SEP) IROFS C-1 QL-1 Normal electrical power (NEP) SR C-1 QL-1 3-52

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  • 0 ffOlfTlfWfSTllEDtCALISOTOPU Table 3-24. System Safety and Seismic Classification and Associated Quality Level Group (2 pages)

I Highest safety Seismic Quality level System name (code) classificationa classificationb group Process vessel ventilation (PVV) IROFS C-1 QL-1 Facility ventilation (FV)c IROFS C-1/11 QL-1 /2 Fire protection (FP) SR C-II QL-2 Plant and instrument air (PA) NSR C-11 QL-2 Emergency Purge gas (PG) IROFS C-II QL-1 Gas supply (GS) NSR C-II QL-2 Process chilled water (PCW) IROFS C-1 QL-1 Facility chilled water (FCW) NSR C-11 QL-2 Facility heated water (HW) NSR C-11 QL-2 Process steam IROFS C-1 QL-1 Demineralized water (DW) NSR C-II QL-2 Chemical supply (CS) IROFS C-I QL-1 Biological shield (BS) IROFS C-1 QL-1 Facility process control (FPC) SR C-II QL-2

  • Safety classification accounts for highest classification in the system. Systems that are classified as safety-related may include both safety-related and non-safety-related components. Only safety-related components will be used to satisfy the safety fun ctions of the system, whereas non-safety-related components can be used to perform non-safety fun ctions. For examp le, there are non-safety-related components, such as fa ns, within the safety-related ventilatio n systems that perform non-safety-related functions.

b Seismic category may be locally revised to account for II over I design criteria and to eliminate potential system degradation due to seismic interactions.

c Ventilation zone classifications vary - Venti lation Zone I and II are considered safety-related, C-1 and QL-1 ; Ventilation Zone III and IV are considered non-safety-related, C-11 and QL-2.

IROFS = items relied on for safety. RPF = Radioisotope Production Facility.

NSR = non-safety related. SR = safety-related (not IROFS).

SSCs that must maintain structural integrity post-DBE, but are not required to remain functional are C-11.

All other SSCs that have no specific NRC-regulated requirements are designed to local jurisdictional requirements for structural integrity and are C-III. All C-1 SSCs are analyzed under the loading conditions of the DBE and consider margins of safety appropriate for that earthquake. The margin of safety provided for safety-class SSCs for the DBE are sufficient to ensure that their design functions are not put at risk. Table 3-25 presents the likelihood index limit guidelines and associated event frequency and risk index limits.

Table 3-25. Likelihood Index Limit Guidelines Likely normal faci lity process condition Not unlikely (frequent facility process condition)

Unlikely (infrequent faci lity process condition)

      • 4 3

2 Event frequency limits Multiple events per year 4

Between I 0 and 4

More than 10 per event, per year l 0-5 per event,

> or = 0

>-4 <O

-4 to 5 per year Highly unlikely (limiting facility process condition) Less than 10-5 per event, per year < -5 3-53

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NOmfWEST MEDtCAl tsOTOl'£1 NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components 3.5.2.3 Design Basis Functions, Values, and Criteria The design basis for systems and components required for safe operation and shutdown of the RPF are established in three categories, which are described below. The preliminary design basis functions and values for each major system are provided in the following subsections.

Design Basis Functions

  • License conditions, orders, or technical specifications
  • Functions credited in the safety analysis to ensure safe shutdown of the facility is achieved and maintained, prevent potential accidents, or mitigate the potential consequences of accidents that could result in consequences greater than applicable NRC exposure guidelines Design Basis Values
  • Values or ranges of values of controlling parameters established as reference bounds for RPF design to meet design basis function requirements
  • Values may be established by an NRC requirement, derived from or confirmed by the safety analysis, or selected by the designer from an applicable code, standard, or guidance document Design Basis Criteria
  • Code-driven requirements established for the RPF fall into seven categories, including fabrication, construction, operations, testing, inspection, performance, and quality
  • Codes include national consensus codes, national standards, and national guidance documents
  • Design of safety-related systems (including protection systems) is consistent with IEEE 379, Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, and Regulatory Guide 1.53 , Application ofthe Single-Failure Criterion to Nuclear Power Plant Protection Systems
  • Protection system is designed to provide two or three channels for each protective systems and functions and two logic train circuits:

Redundant channels and trains will be electrically isolated and physically separated in areas outside of the RPF control room Redundant design will not prevent protective action at the system level 3.5.2.4 System Functions/Safety Functions The NWMI RPF will provide protection against natural phenomena hazards for the personnel, SNM, and systems within the facility. The facility will also provide protection against operational and accident hazards to personnel and the public. Table 3-2 lists the IROFS defined by the preliminary hazards analysis.

3.5.2.5 Systems and Components 3.5.2.5.1 Mechanical RPF C-1 mechanical equipment and components (identified in Table 3-24) will be qualified for operation under the design basis earthquake (DBEQ) seismic conditions by prototype testing, operating experience, or appropriate analysis. The C-1 mechanical equipment is also designed to withstand loadings due to the DBEQ, vibrational loadings transmitted through piping, and operational vibratory loading, such as floor vibration due to other operating equipment, without loss of function or fluid boundary. This analysis considers the natural frequency of the operating equipment, the floor response spectra at the equipment location, and loadings transmitted to the equipment and the equipment anchorage.

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  • NOmfWHTMfDICALISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components The qualification documents and all supporting analysis and test reports will be maintained as part of the permanent plant record in accordance with the requirements of the NWMI QAPP.

The safety-related equipment and components within the RPF will be required to function during normal operations and during and following DBEs. This equipment will be capable of functioning in the RPF environmental conditions associated with normal operations and design basis accidents. Certain systems and components used in the ESF systems will be located in a controlled environment. This controlled environment is considered an integral part of the ESF systems.

3.5.2.5.2 Instrumentation and Electrical C-1 instrumentation and electrical equipment (identified in Table 3-24) is designed to resist and withstand the effects of the postulated DBEQ without functional impairment. The equipment will remain operable during and after a DBEQ. The magnitude and frequency of the DBEQ loadings that each component experiences will be determined by its location within the RPF. In-structure response curves at various building elevations will be developed to support design. The equipment (e.g., batteries and instrument racks, control consoles) has test data, operating experience, and/or calculations to substantiate the ability of the components and systems to not suffer loss of function during or after seismic loadings due to the DBEQ. This information will be completed during final design of the RPF and provided in the Operating License Application.

This certification of compliance with the specified seismic requirements, including compliance with the requirements of IEEE 344, is maintained as part of the permanent plant record in accordance with the NWMIQAPP.

3.5.2.6 Qualification Methods Environmental qualification of safety-related mechanical, instrumentation, and electrical systems and components is demonstrated by tests, analysis, or reliance on operating experience. Qualification method testing will be accomplished either by tests on the particular equipment or by type tests performed on similar equipment under environmental conditions at least as severe as the specified conditions. The equipment will be qualified for normal and accident environments. Qualification data will be maintained as part of the permanent plant record in accordance with the NWMI QAPP.

3.5.2.7 Radioisotope Production Facility Specific System Design Basis Functions and Values The design basis functions and values for each system identified in Table 3-1 are discussed in the following subsections. Additional details for each system described below will be updated during the development of the Operating License Application.

3.5.2.7.1 Target Fabrication System An overview and detailed description of the target fabrication system are provided in Chapter 4.0, Sections 4.1.3.1and4.4, respectively.

Design Basis Functions

  • Store fresh LEU, LEU target material , and new LEU targets
  • Produce LEU target material from fresh and recycled LEU material
  • Assemble, load, and fabricate LEU targets
  • Reduce or eliminate the buildup of static electricity
  • Minimize uranium losses through target fabrication
  • Safety-related functions :

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  • NOml'WEST M£tNCAl lSOTOffl Maintain subcriticality conditions within target fabrication system Prevent flammable gas composition within target fabrication system Limit personnel exposure to hazardous chemicals and offgases Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps)
  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs 3.5.2.7.2 Target Receipt and Disassembly System An overview and detailed description of the target receipt and disassembly system are provided in Chapter 4.0, Section 4.1 .3.2, and Sections 4.3.2/4.3.3 , respectively.

Design Basis Functions

  • Handle irradiated target shipping cask, including all opening, closing, and lifting operations
  • Retrieve irradiated targets from a shipping cask
  • Disassemble targets and retrieving irradiated target material from targets
  • Reduce or eliminate the buildup of static electricity
  • Safety-related functions:

Provide radiological shielding during receipt and disassembly activities Maintain subcriticality conditions within target receipt and disassembly system Prevent radiological materials from being released during target receipt and disassembly operations to limit the exposure of workers, the public, and environment to radioactive material Maintain positive control of radiological materials (LEU target material and radiological waste)

Protect personnel and equipment from industrial hazards associated with system equipment (e.g. , moving parts)

Design Basis Values

  • 30-year design life
  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs
  • Crane designed for anticipated load (e.g. , hot cell cover block) of approximately 68 metric tons (MT) (75 ton) 3.5.2.7.3 Replace Target Dissolution (DS)

An overview and detailed description of the target dissolution system are provided in Chapter 4.0, Sections 4.1 .3.3 and 4.3.4, respectively.

Design Basis Functions

  • Fill the dissolver basket with the LEU target material
  • Dissolve the LEU target material within dissolver basket
  • Treat the offgas from the target dissolution system
  • Handle and package solid waste created by normal operational activities 3-56

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  • Safety-related functions :

Provide radiological shielding during target dissolution activities Control and prevent flammable gas from reaching lower flammability limit conditions Maintain subcriticality conditions through inherently safe design of target dissolution equipment Maintain positive control of radiological materials (LEU target material and radiological waste)

Design Basis Values

  • 30-year design life with the exception of common replaceable parts (e.g., pumps)
  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs
  • Prevent radiological materials from being released during target dissolution operations to limit the exposure of workers, the public, and environment to radioactive material per I 0 CFR 20 3.5.2.7.4 Molybdenum Recovery and Purification (MR)

An overview and detailed description of the Mo recovery and purification system are provided in Chapter 4.0, Sections 4.1.3.4 and 4.3.5, respectively.

Design Basis Functions

  • Recovery of Mo product from a nitric acid solution created from dissolved irradiated uranium targets
  • Purification of the recovered Mo product to reach specified purity requirements, followed by shipment of the Mo product
  • Safety-related functions:

Maintain subcriticality conditions through inherently safe design of components that could handle high-uranium content fluid Prevent radiological materials from being released by containing fluids in appropriate tubing, valves, and other components Control and prevent flammable gas from reaching lower flammability limit conditions Maintain positive control of radiological materials (99Mo product, intermediate streams, and radiological waste)

Provide appropriate containers and handling systems to protect personnel from industrial hazards such as chemical exposure (e.g., nitric acid, caustic, etc.)

Design Basis Values

  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs
  • 30-year design life with the exception of common replaceable parts (e.g. , pumps)
  • Replace consumables after each batch 3.5.2.7.5 Uranium Recovery and Recycle (UR)

An overview and detailed description of the uranium recovery and recycle system are provided in Chapter 4.0, Sections 4. l.3 .5 and 4.3.6, respectively.

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NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components

. * ~ ~:! ; . NORTHWESTMEDICAi. ISOTOPfS Design Basis Functions

  • Receive and decay impure LEU solution
  • Recover and purify impure LEU solution
  • Decay and recycle LEU solution
  • Transfer process waste
  • Safety-related functions :

Provide radiological shielding during uranium recovery and recycle system activities Prevent radiological release during uranium recovery and recycle system activities Maintain subcriticality conditions through inherently safe design of the uranium recovery and recycle equipment Control and preventing flammable gas from reaching lower flammability limit conditions Maintain positive control of radiological materials Protect personnel and equipment from industrial hazards associated with the system equipment, such as moving parts, high temperatures, and electric shock Design Basis Values

  • 30-year design life with the exception of common replaceable parts (e.g. , pumps)
  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs 3.5.2.7.6 Waste Handling An overview and detailed description of the waste handling system are provided in Chapter 4.0, Section 4.1.3.6 and Chapter 9.0, Section 9.7.2, respectively.

Design Basis Functions

  • Receive liquid waste that is divided into high-dose source terms and low-dose source terms to lag storage
  • Transfer remotely loaded drums with high-activity solid waste via a solid waste drum transit system to a waste encapsulation cell
  • Encapsulate solid waste drums
  • Load drums with solidification agent and low-dose liquid waste
  • Load high-integrity containers with solidification agent and high-dose liquid waste
  • Handle and load a waste shipping cask with radiological waste drums/containers
  • Safety-related functions :

Maintain subcriticality conditions through mass limits Prevent spread of contamination to manned areas of the facility that could result in personnel exposure to radioactive materials or toxic chemicals Provide shielding, distance, or other means to minimize personnel exposure to penetrating radiation Design Basis Values

  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3-58

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  • NCMmlWESTMEDK:AllSOTDPH Chapter 3.0 - Design of Structures , Systems and Components 3.5.2.7.7 Criticality Accident Alarm System Chapter 6.0, Section 6.3.3.1 , and Chapter 7.0, Section 7.3 .7, provide descriptions of the criticality accident alarm system.

Design Basis Functions

  • Provide analysis for criticality accident alarm system coverage in all areas where SNM is handled, processed, or stored
  • Provide for continuous monitoring, indication, and recording of neutron or gamma radiation levels in areas where personnel may be present and wherever an accidental criticality event could result from operational processes.
  • Provide both local and remote annunciation of a criticality excursion
  • Remain operational during DBEs Design Basis Values
  • 30-year design life
  • Capable of detecting a criticality accident that produces an absorbed dose in soft tissue of 20 absorbed radiation dose (rad) of combined neutron or gamma radiation at an unshielded distance of 2 m from reacting material within one minute 3.5.2.7.8 Continuous Air Monitoring System Chapter 7.0, Section 7.6, and Chapter 11.0, Section 11.1.4, provide detailed descriptions of the RPF continuous air monitoring system.

Design Basis Functions

  • Provide real-time local and remote annunciation of airborne contamination in excess of preset limits
  • Provide real-time local and remote annunciation of radiological dose of excess of preset limits
  • Provide environmental monitoring of nuclear radioactive stack releases
  • Provide the capability to collect continuous samples
  • Remain operational during DBEs Design Basis Values
  • Activate when airborne radioactivity levels exceed predetermined limits
  • Activate when radiological dose levels exceed predetermined limits
  • Adjust volume of air sampled to ensure adequate sensitivity with minimum sampling time 3.5.2.7.9 Standby Electrical Power Chapter 8.0, Section 8.2 provides a detailed description of the RPF standby electrical power (SEP) system.

Design Basis Functions SEP includes two types of components: uninterruptible power supplies (UPS) and a standby diesel generator:

  • UPS - Provides power when normal power supplies are absent
  • Standby diesel generator - Provides power when normal power supplies are absent to allow continued RPF processing 3-59

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  • NOmlWESTMEDICAllSOTOPH Design Basis Values
  • 30-year design life
  • Maintain power availability for a minimum of 120 min post-accident (UPS)
  • Maintain power availability for 12 hr (diesel generator) 3.5.2.7.10 Normal Electrical Power Chapter 8.0, Section 8.1 provides a detailed description of the RPF normal electrical power (NEP) system.

Design Basis Functions

  • Provide facility power during normal operations Design Basis Values
  • 30-year design life 3.5.2.7.ll Process Vessel Ventilation System Chapter 9.0, Section 9.1 provides a detailed description of the process vessel ventilation system.

Design Basis Functions

  • Provide primary system functions to protect on-site and off-site personnel from radiological and other industrial related hazards
  • Collect air in-leakage sweep from each of the numerous vessels and other components in main RPF processes and maintain hydrogen concentration process tanks and piping below lower flammability limit
  • Minimize reliance on administrative or complex active engineering controls to provide a confinement system as simple and fail-safe as reasonably possible Design Basis Values
  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs
  • 30-year design life
  • Contain and store noble gases generated in the RPF to meet 10 CFR 20 requirements 3.5.2.7.12 Facility Ventilation System Chapter 9.0, Section 9.1 provides a detailed description of the facility ventilation system.

Design Basis Functions

  • Provide confinement of hazardous chemical fumes and airborne radiological materials and conditioning of RPF environment for facility personnel and equipment
  • Prevent release and dispersal of airborne radioactive materials (e.g., maintain pressure gradients to ensure proper flow of air from least potentially contaminated areas to most potentially contaminated areas) to protect health and minimize danger to life or property
  • Maintain dose uptake through ingestion to levels as low as reasonably achievable (ALARA)
  • Provide makeup air and condition the RPF environment for process and electrical equipment
  • Process exhaust flow from the process vessel ventilation system 3-60

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  • Provide confinement of airborne radioactive materials by providing for the rapid, automatic closure of isolation dampers within confinement zones for various accident conditions
  • Provide conditioned air to ensure suitable environmental conditions for personnel and equipment inRPF Design Basis Values
  • Maintain primary fission product boundary during and after normal operations, shutdown conditions, and DBEs
  • Provide an integrated leak rate for confinement boundaries that meets the requirements of accident analyses that complies with 10 CFR 10.61
  • 30-year design life
  • Maintain occupied space at 24 degrees Celsius (0 C) (75 degrees Fahrenheit [°F]) (summer) and 22°C (72°F) (winter), with active ventilation to support workers and equipment
  • Maintain air quality that complies with 10 CFR 20 dose limits for normal operations and shutdown 3.5.2.7.13 Fire Protection System Chapter 9.0, Section 9.3 provides a detailed description of the RPF fire protection system.

Design Basis Functions

  • Provide detection and suppression of fires
  • Generate alarm signals indicating presence and location of fire
  • Execute commands appropriate for the particular location of the fire (e.g., provide varying levels of notification of a fire event and transmitting notification to RPF central alarm station and RPF control room)
  • Provide fire detection in RPF and initiate fire-rated damper closures
  • Remain functional during DBEs Design Basis Values
  • 30-year design life
  • Provide a constant flow of water to an area experiencing a fire for a minimum of 120 min based on the size of the area per International Fire Code (IFC, 2012)
  • Provide sprinkler systems, when necessary, per National Fire Protection Association (NFPA) 13, Standard for the Installation of Sprinkler Systems 3.5.2.7.14 Plant and Instrument Air System Chapter 9.0, Section 9.7.1 provides a detailed description of the RPF plant and instrument air system.

Design Basis Functions

  • Provide small , advective flows of plant air for several RPF activities (e.g., tool operation, pump power, purge gas in tanks, valve actuation, and bubbler tank level measurement)
  • Provide plant air receiver buffer capacity to make up difference between peak demand and compressor capacity 3-61
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  • Provide plant air to instrument air subsystem for bubblers and valve actuation
  • Provide instrument air receiver buffer capacity to make up difference between peak demand and compressor capacity Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g. , pumps)
  • Provide instrument air dried in regenerable desiccant beds to a dew point of no greater than -40°C

(-40°F) and filtered to a maximum 40 micron (µ) particle size 3.5.2.7.15 Emergency Purge Gas System Chapter 6.0, Section 6.2.1.7 .5 provides a detailed description of the emergency purge gas system.

Design Basis Functions

  • Provide > 12 hr of nitrogen to the emergency purge gas system
  • Emergency purge gas system to provide nitrogen to the required process tanks
  • Remain functional during DBEs Design Basis Values
  • 30-year design life with the exception of common replaceable parts
  • Maintain hydrogen gas (H2) concentrations less than 25% of the lower flammability limit 3.5.2.7.16 Gas Supply System Chapter 9.0, Section 9.7.1 provides a detailed description of the gas supply system.

Design Basis Functions

  • Provide nitrogen from a tube truck to the chemical supply room where manifold piping will be used to distribute the gas
  • Provide adequate flow to ensure that the accumulation of combustible gases is below hazardous concentrations and reduces radiological hazards due to accumulation of gaseous fission products Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g. , pumps)
  • Provide standard gas bottles, with capacity of approximately 8,495 L (300 cubic feet [ft3])

3.5.2.7.17 Process Chilled Water System Chapter 9.0, Section 9.7.1 provides a detailed description of the RPF chilled water system.

Design Basis Functions

  • Provide process chilled water loop for three secondary loops heat exchangers One large geometry secondary loop in hot cell One criticality-safe geometry secondary loop in hot cell One criticality-safe geometry secondary loop in target fabrication area

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  • Provide cover gas to prevent flammable conditions in secondary loops Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps)
  • Chilled water to various process equipment at no greater than 10°C (50°F) during normal operations
  • Maintain the hydrogen concentration in the coolant system at less than 25 percent of the lower flammability limit of 5 percent H2 3.5.2.7.18 Facility Chilled Water System Chapter 9.0, Section 9.7.1.2.2 provides a detailed description of the RPF facility chilled water system.

Design Basis Functions

  • Provide cooling media to heating, ventilation, and air conditioning (HVAC) system
  • Supply HVAC system with cooling water that is circulated through the chilled water coils in air-handling units Design Basis Values
  • Provide cooling water at a temperature of9°C (48°F) to the HV AC air-handling unit cooling coils
  • 30-yeardesign life with the exception of common replaceable parts (e.g., pumps) 3.5.2.7.19 Facility Heated Water System Chapter 9.0, Section 9.7.1.2.2 provides a detailed description of the RPF heated water system.

Design Basis Functions

  • Provide heated media to HV AC system
  • Supply the HV AC system with heated water that is circulated through the heated water coils in the air-handling units Design Basis Values
  • Provide heated water at a temperature of 82°C (180°F) to HV AC air-handling unit heating coils and reheat coil
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3.5.2.7.20 Process Steam System - Boiler Chapter 9.0, Section 9. 7.1 provides a detailed description of the RPF process steam system for the boiler.

Design Basis Functions

  • Generate low- and medium-pressure steam using a natural gas-fired package boiler
  • Provide a closed loop steam system for the hot cell secondary loops that meets criticality control requirements
  • Limit sludge or dissolved solids content with automatic and makeup water streams in the boiler 3-63

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  • 30-year design life with the exception of common replaceable parts (e.g., pumps)
  • Provide saturated steam at 1.7 kg/square centimeters (cm 2) (25 lb/square inch [in. 2]) and 4.2 kg/cm 2 (60 lb/in. 2) gauge to various process equipment 3.5.2.7.21 Process Steam System - Hot Cell Secondary Loops Chapter 9.0, Section 9.7. l provides a detailed description of the RPF process steam system for the hot cell secondary loops.

Design Basis Functions

  • Provide a closed loop steam system for the hot cell secondary loops
  • Generate low-pressure steam using a vertical shell-and-tube heat exchanger
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3.5.2.7.22 Demineralized Water System Chapter 9.0, Section 9.7.1 provides a detailed description of the RPF demineralized water system.

Design Basis Functions

  • Provide demineralized water to RPF except for administration and truck bay areas
  • Remove mineral ions from municipal water through an ion exchange (IX) process and accumulate in a storage tank
  • Provide regenerable IX media using a strong acid and a strong base
  • Feed acids and bases from local chemical drums by toe pumps Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps)
  • Provide the water at 4.2 kg/cm 2 (60 lb/in. 2) gauge 3.5.2.7.23 Supply Air System Chapter 9.0, Section 9.1.2 provides a detailed description of the supply air system. The design basis functions and values are identified in Section 3.5.2.7.12.

3.5.2.7.24 Chemical Supply System Chapter 9.0, Section 9.7.4 provides a detailed description of the chemical supply system.

Design Basis Functions

  • Provide storage capability for nitric acid, sodium hydroxide, reductant, and nitrogen oxide absorber solutions, hydrogen peroxide, and fresh uranium IX resin
  • Segregate incompatible chemicals (e.g., acids from bases)
  • Provide transfer capability for chemical solutions mixed to required concentrations and used in target fabrication, target dissolution, Mo recovery and purification, and waste management systems 3-64

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  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3.5.2. 7.25 Biological Shielding System Chapter 4.0, Section 4.2, provides a detailed description of the RPF biological shielding.

Design Basis Functions

  • Provide biological shielding from radiation sources in the hot cells for workers in occupied areas of the RPF
  • Limit physical access to hot cells
  • Remain functional through DB Es without loss of structural integrity Design Basis Values
  • 30-year design life
  • Provide dose rates consistent with ALARA goals for normally occupied areas 3.5.2.7.26 Facility Process Control System Chapter 7.0, Section 7.2.3 provides a description of the FPC system.

Design Basis Functions

  • Perform as overall production process controller
  • Monitor and control process instrumented functions within the RPF (e.g., process fluid transfers, controlled inter-equipment pump transfers of process fluids)
  • Provide monitoring of safety-related components while BMS (a subset of the FPC system) monitors ventilation system and mechanical utility systems
  • Ensure ESF systems operate independently from FPC system or BMS
  • Use hard-wired analog controls/interlocks for each ESF safety function to protect workers, public, and environment
  • Integrate into and monitor ESF parameters and alarm functions by FPC system or BMS
  • Initiate actuation of isolation dampers for hot cell area or analytical area on receipt of signals from fire protection system Design Basis Values
  • 30-yeardesign life with the exception of common replaceable parts (e.g., controllers) 3-65
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3.6 REFERENCES

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I 0 CFR 50.31 , "Combining Applications," Code of Federal Regulations, Office of the Federal Register, as amended.

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40 CFR 63 , "NESHAP for Source Categories," Code of Federal Regulations, Office of the Federal Register, as amended.

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  • !*.*!
  • NORTHWEST MmlCAL tsOTOPES 40 CFR 141 , "National Primary Drinking Water Regulations," Code of Federal Regulations, Office of the Federal Register, as amended.

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ANSI/AHR! Standard 430, Performance Rating of Central Station Air-Handling Units, Air-Conditioning, Heating, and Refrigeration Institute, Arlington, Virginia, 2009.

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' !*.*! ." NO<<TKWEST MEDtcAl '5GTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components ANSl/AHRI Standard 850, Performance Rating of Commercial and Industrial Air Filter Equipment, Air-Conditioning, Heating, and Refrigeration Institute, Arlington, Virginia, 2013 .

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  • !~.~~ : NCMITifWEST MEOtcAl tsOTOP(S NWMl-201 3-021, Rev. 1 Chapter 3. 0 - Design of Structures , Systems and Components ANSVANS-10.5, Accommodating User Needs in Computer Program Development, American Nuclear Society, La Grange Park, Illinois, 2006 (R201 l) .

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ANSVANS-59.3, Nuclear Safety Criteria f or Control Air Systems, American Nuclear Society, La Grange Park, Illinois, 1992 (R2002) (W2012).

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' !~.~~ ; . NOmfWEST MEDtCAllSOTOPfS NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components ANSI/IEEE N320, American National Standard Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 1979.

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. ! *.* ~ . NOmlWlST .OICAl tsOTOPfl NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components ANSI/SMACNA 001-2008, Seismic Restraint Manual: Guidelines for Mechanical Systems, Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia, 2008.

ANSI/TIA-568-C.1, Commercial Building Telecommunications Cabling Standard, Telecommunications Industry Association, Arlington, Virginia, 2012.

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ANSI/TIA-606, Administration Standard for Commercial Telecommunications Infrastructure, Telecommunications Industry Association, Arlington, 2012.

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  • NOmfWEST MtDetAl tsOTOPU NWMl-2013-021 , Rev. 1 Chapter 3.0 - Design of Structures , Systems and Components ASME B31 , Standards of Pressure Piping, American Society of Mechanical Engineers, New York, New York, 2014.

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' ~ * . ~ ~ : . NORTHWEST MEDtcAl. lSOTOPU AWS B2.1 /B2.1M, Specification for Welding Procedure and Performance Qualification , American Welding Society, Miami, Florida, 2009.

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  • NORTHWEST MO>>CAl ISOTOPU NWMl-2013-021 , Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components IEEE 338, Standard/or Criteria/or the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2012.

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~ *.* ~ ' NO<<TNWEST MEOtcAl. ISOTOP'lS NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components ISA-RP60. l-1990, Control Center Facilities, The International Society of Automation, Research Triangle Park, North Carolina, 1990.

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NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components NECA 408, Standard/or Installing and Maintaining Busways (ANSI) , National Electrical Contractors Association, Bethesda, Maryland, 2009.

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  • ~ ~* ~ ! ." NORTHWEST MEOtcAI. ISOTOPES NWMl-201 3-02 1, Rev. 1 Chapter 3. 0 - Design of Structures , Systems and Components NFPA 10, Standard for Portable Fire Extinguishers, National Fire Protection Association, Quincy, Massachusetts, 2013 .

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  • NOllT1fWUT MEDtCAl ISOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components NFPA 80, Standard for Fire Doors and Other Opening Protectives, National Fire Protection Association, Quincy, Massachusetts, 2013.

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  • .*.NWMI NOfllltWEST 11£111CAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 - Design of Structures, Systems and Components NFPA 297, Guide on Principles and Practices for Communications Systems, National Fire Protection Association, Quincy, Massachusetts, 1995 .

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Regulatory Guide 1.167, Restart of a Nuclear Power Plant Shut down by a Seismic Event, U.S. Nuclear Regulatory Commission, Washington, D.C., March 1997.

Regulatory Guide 1.208, Performance Based Approach to Define the Site-Specific Earthquake Ground Motion , U.S . Nuclear Regulatory Commission, Washington, D.C., March 2007.

Regulatory Guide 3.3, Quality Assurance Program Requirements for Fuel Reprocessing Plants and for Plutonium Processing and Fuel Fabrication Plants, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C. , March 1974 (R2013).

Regulatory Guide 3.6, Content of Technical Specification for Fuel Reprocessing Plants, U.S. Nuclear Regulatory Commission, Washington, D.C. , April 1973 (R2013).

Regulatory Guide 3.10, Liquid Waste Treatment System Design Guide for Plutonium Processing and Fuel Fabrication Plants, U.S. Nuclear Regulatory Commission, Washington, D.C., June 1973 (R2013).

Regulatory Guide 3 .18, Confinement Barriers and Systems for Fuel Reprocessing Plants, U. S Nuclear Regulatory Commission, Washington, D.C. , February 1974 (R2013) .

Regulatory Guide 3.20, Process Offgas Systems for Fuel Reprocessing Plants , U.S. Nuclear Regulatory Commission, Washington, D.C., February 1974 (R2013).

Regulatory Guide 3.71 , Nuclear Criticality Safety Standards for Fuels and Materials Facilities, Rev. 2, U.S. Nuclear Regulatory Commission, Washington, D.C., December 2010.

Regulatory Guide 5.7, Entry/Exit Control for Protected Areas, Vital Areas, and Material Access Areas ,

Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1980 (R2010).

Regulatory Guide 5 .12, General Use of Locks in the Protection and Control of Facilities and Special Nuclear Materials, U.S. Nuclear Regulatory Commission, Washington, D.C. , November 1973 (R2010).

Regulatory Guide 5.27, Special Nuclear Material Doorway Monitors, U.S. Nuclear Regulatory Commission, Washington, D.C., June 1974.

Regulatory Guide 5.44, Perimeter Intrusion Alarm Systems, Rev. 3, U.S. Nuclear Regulatory Commission, Washington, D.C. , October 1997 (R2010).

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Nuclear Regulatory Commission, Washington, D.C., June 1980.

Regulatory Guide 5.65, Vital Area Access Control, Protection of Physical Security Equipment, and Key and Lock Controls , U.S. Nuclear Regulatory Commission, Washington, D.C., September 1986 (R2010).

Regulatory Guide 5.71, Cyber Security Programs for Nuclear Facilities, U.S. Nuclear Regulatory Commission, Washington, D.C., 2010.

SMACNA 1143, HVAC Air Duct Leakage Test, Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia, 1985.

SMACNA 1520, Round Industrial Duct Construction Standard, Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia, 1999.

SMACNA 1922, Rectangular Industrial Duct Construction Standard, Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia, 2004.

SMACNA 1966, HVA C Duct Construction Standard - Metal and Flexible, Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia, 2006.

SMACNA-2006, HVAC Systems Duct Design , Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia, 2006.

SNT-TC-1 A, Recommended Practice No. SNT-TC-1 A: Personnel Qualification and Certification in Nondestructive Testing, American Society for Nondestructive Testing, Columbus, Ohio, 20 11 .

Technical Paper No. 40, Rainfall Frequency Atlas of the United States for Durations from 30 Minutes to 24 Hours and Return Periods from I to JOO Years, Weather Bureau, U.S. Department of Commerce, Washington, D.C. 1963 .

Terracon, 2011 a, Phase I Environmental Site Assessment Discovery Ridge Lots 2, 5, 6, 7, 8, 9, I 0, I!, I 2, 13, I 4, 15, I 6, I 7, and I 8, Terracon Consultants, Inc., prepared for University of Missouri and Trabue, Hansen & Hinshaw, Inc., Terracon Project No. 09117701 , March 23, 2011 .

Terracon, 2011 b, Preliminary Geotechnical Engineering Report Discovery Ridge-Certified Site Program Lots 2, 5, 6, 7, 8, 9, I 0, !!, 12, 13, 14, I 5, 16, I 7, and 18, Terracon Consultants, Inc., prepared for University of Missouri and Trabue, Hansen & Hinshaw, Inc., Terracon Project No. 09105094.1 ,

February 11 , 2011.

UL 181, Standard for Factory-Made Air Ducts and Connectors, Underwriters Laboratories, Washington, D.C., 2013.

UL 499, Standard for Electric Heating Appliances, Underwriters Laboratories, Washington, D.C., 2014.

UL 555, Standard for Fire Dampers, Underwriters Laboratories, Washington, D.C., 2006.

UL 586, Standard for High Efficiency, Particulate, Air Filter Units , Underwriters Laboratories, Washington, D.C., 2009.

UL 900, Standard for Air Filter Units, Underwriters Laboratories, Washington, D.C., 2004.

UL 1995, Heating and Cooling Equipment, Underwriters Laboratories, Washington, D.C., 2011 .

USGS, "2008 U.S. Geological Survey National Seismic Hazard Maps," U.S. Geological Survey, Rolla, Missouri, 2008 .

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. NORTHWEST MEDICAL ISOTOPES Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 June 2017 Prepared by:

Northwest Medical Isotopes, LLC 815 NW gth Ave , Suite 256 Corvallis, OR 97330

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........ NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features

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  • NORTHWEST MEDtCAl ISOTOHS Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1

Title:

Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

C w.J~c.f/~

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  • NDlrTHWHTMBHCAllSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not required 1 6/26/2017 Incorporate changes based on responses to C. Haass NRC Requests for Additional Information

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  • NOITHWBT IWMCAl lSOTOPU CONTENTS 6.0 ENGINEERED SAFETY FEATURES ...................... ..... ......... ........... .. ..... ..... ... ....... ...... ....... ...... . 6-1 6.1 Summary Description .. .. .............................. .... ..... ...... .... ....... ............ ..... ... .................. ... .... 6-1 6.2 Detailed Descriptions ...... ........................... ...... ... ....... .... .... ..... ... ...... .... ...... .... .................... 6-5 6.2.1 Confinement .... ...... ........ ............... ........................................ ..... .......... ................. 6-5 6.2.1.1 Confinement System ............ ................ ........... ............ .................... .... 6-7 6.2. I .2 Accidents Mitigated .... .................. .. ...... ... ..... ..... ........................... .... 6-1 I 6.2. I .3 Functional Requirements ... .... ........ ...... ... ... .......... ................ .. .... ... .... 6-1 I 6.2. 1.4 Confinement Components ...... ........ .... ... ..... ......... .. ... ..................... .... 6-1 I 6.2.1.5 Test Requirements ............ ..... ...... ... ..... ... .... ...... ..... ..... ..... ...... ......... ... 6-12 6.2. 1.6 Design Basis .. ....... ... ....... ..... ............... ...... ... ....... .......... ..................... 6-13 6.2.1. 7 Derived Confinement Items Relied on for Safety .... ....... ............. ..... 6-13 6.2. I .8 Dissolver Offgas Systems .... .. ... ... ... ... .. .... ..... ..... .. .... ......... .............. .. 6-23 6.2.1.9 Exhaust System ................................................................... ..... .... ..... 6-26 6.2. 1.10 Effluent Monitoring System ... ....... ....... .... ... ............ ....... .............. ..... 6-26 6.2.1.11 Radioactive Release Monitoring ... .. ...... ..... .................. .... ................. 6-26 6.2.1 .12 Confinement System Mitigation Effects ... ... ..................................... 6-26 6.2 .2 Containment .................................. .......... .. .. ........ ..... .. .. ............ ..... ..................... 6-27 6.2.3 Emergency Cooling System ......... .... ........... ..... ........ ....... ... .... ..... ..... .............. .... 6-27 6.3 Nuclear Criticality Safety in the Radioisotope Production Facility .. ............................... 6-28 6.3.1 Criticality Safety Controls ...... ...... ......... ......... .. ... ..... ... ..... .... ........ ... ................... 6-36 6.3 .1 .1 Preliminary Criticality Safety Evaluations .............. ........ .................. 6-36 6.3.1.2 Derived Nuclear Criticality Safety Items Relied on for Safety ....... .. 6-59 6.3.2 Surveillance Requirements ... ........ ........ ......... .. ........ .. ................ ........ .... .... ....... .. 6-71 6.3.3 Technical Specifications ......................... ...... .. ...... ................. .... ........... .... ......... 6-71 6.4 References ...... ...... ...... ... ....................................... ...................... .. ... .......... ....................... 6-72 6-i

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  • ~ *.. ~ ." . NOUHWf.ST M£DfCAI. ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features FIGURES Figure 6-1. Simplified Zone I Ventilation Schematic .. ........ .. .. ...... ............. ..... ..... ..... ... ... ... ... ... ..... .. ... 6-6 Figure 6-2. Ground Level Confinement Boundary ..... .......... ... ............ ..... .... ... .. ... ....... ... .... .... .... ... .. ... 6-8 Figure 6-3 . Mechanical Level Confinement Boundary ........... ...... ...... .... .. ....... ... ........ .... ....... ............ 6-9 Figure 6-4. Lower Level Confinement Boundary ........... ............................... .............. ..... .. .... .... ..... 6-10 Figure 6-5 . Dissolver Offgas System Engineered Safety Features ....... .... .. ............. .... .. ................... 6-14 Figure 6-6. Dissolver Offgas Hot Cell Equipment Location .. .. .......... ...... ............... .... .. ........ .... ..... .. 6-15 Figure 6-7. Proposed Location of Double-Wall Piping (Example) ......... ............. ............ .......... .. .... 6-21 TABLES Table 6-1. Summary of Confinement Engineered Safety Features (2 pages) ... ....... ...... ................... 6-2 Table 6-2. Summary of Criticality Engineered Safety Features (2 pages) ...... ......... ... ....... .. ... .... ... ... 6-3 Table 6-3 . Confinement System Safety Functions .... ... ...... .... ..... .... ..... ..... ............. .. ................... .. .... 6-7 Table 6-4. Area of Applicability Summary ... ............... .... .... ..... ...... .......... ... ....... ..... ... ... ... .. ........ .... 6-37 Table 6-5. Controlled Nuclear Criticality Safety Parameters ........ .......... .. ... .. ... ...... .... ... .............. ... 6-38 Table 6-6. [Proprietary Information] Double-Contingency Controls .......... ... ....... .. .... ...... ... ..... .. .... 6-39 Table 6-7. [Proprietary Information] Double-Contingency Controls (2 pages) .. ...... ... .... .. .. .. ..... .... 6-40 Table 6-8 . [Proprietary Information] Double-Contingency Controls (2 pages) ..... ........... ........... ... 6-41 Table 6-9 . [Proprietary Information] Double-Contingency Controls (8 pages) ....... .... .. .......... ...... . 6-43 Table 6-10. [Proprietary Information] Double-Contingency Controls (2 pages) .. .. ....... ..... .... ... ...... . 6-5 I Table 6-11 . [Proprietary Information] Double-Contingency Controls (3 pages) ..... .... ..... ........ .. ... ... 6-53 Table 6-12. [Proprietary Information] Double-Contingency Controls (2 pages) ...... ... ... ...... ..... ... .... 6-56 Table 6-13. [Proprietary Information] Double-Contingency Controls (2 pages) .. .. .. ...... ... .... ..... .... .. 6-57 6-ii

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  • * ** NCMn"tfWHT Mf.DtCAl ISOTOPU NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features TERMS Acronyms and Abbreviations 99Mo molybdenum-99 23su uranium-235 ADUN acid-deficient uranium nitrate AEC active engineered control ANECF average neutron energy causing fission ANS American Nuclear Society ANSI American National Standards Institute CAAS criticality accident alarm system CFR Code of Federal Regulations CSE criticality safety evaluation DBE design basis earthquake HEGA high-efficiency gas adsorber HEPA high-efficiency particulate air HVAC heating, ventilation, and air conditioning IEU intermediate-enriched uranium IX ion exchange IROFS item relied on for safety Kr krypton LEU low-enriched uranium MCNP Monte-Carlo N-Particle Mo molybdenum N02 nitrogen dioxide NOx nitrogen oxide NRC U.S. Nuclear Regulatory Commission NWMI Northwest Medical Isotopes, LLC PEC passive engineered control PHA preliminary hazards analysis RPF radioisotope production facility SSC structures, systems, and components SPL single parameter limit UN uranium nitride

[Proprietary Information] [Proprietary Information]

USL upper subcritical limits Xe xenon 6-iii

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. ' ~ ~**.! ; . NCMITHW'EST MEDtCAl ISOTOfl(S NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features Units oc degrees Celsius op degrees Fahrenheit atm atmosphere cm centimeter cm3 cubic centimeter ft feet ft2 square feet ft3 cubic feet g gram hr hour

m. inch L liter m meter m2 square meter mm minute mL milliliter mol mole rad radiation absorbed dose wt% weight percent yr year 6-iv

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' ~ * . * ~ 0 N0<<11IWEST MEDICAL lSOTOPO NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.0 ENGINEERED SAFETY FEATURES 6.1

SUMMARY

DESCRIPTION Engineered safety features are active or passive features designed to mitigate the consequences of accidents and to keep radiological exposures to workers, the public, and environment within acceptable values. The engineered safety features associated with confinement of the process radionuclides and hazardous chemicals for the Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF) are summarized in Table 6-1 , including the accidents mitigated; structures, systems, and components (SSC) used to provide the engineered safety features ; and references to subsequent sections providing a more detailed engineered safety feature description.

Confinement is a general engineered safety feature that is credited as being in place as part of the preliminary hazards analysis (PHA) described in Chapter 13.0, "Accident Analysis." Additional items relied on for safety (IROFS) associated with the confinement system were derived from the accident analyses in Chapter 13 .0. The derived IROFS are also listed in Table 6-1 , with reference to more detailed descriptions in Section 6.2.1.

The current design approach does not anticipate requiring containment or an emergency cooling system as engineered safety features , as discussed in Sections 6.2.2 and 6.2.3 .

Nuclear criticality safety is discussed in Section 6.3 . Criticality safety controls are described in Section 6.3.1 . The currently defined criticality safety controls are derived from a combination of preliminary criticality safety evaluations (CSE) and accident analyses, which are described in Chapter 13.0. The criticality safety analyses produce a set offeatures needed to satisfy the double-contingency requirements for nuclear criticality control. These features are evaluated by major systems within the RPF and listed by major system in Section 6.3.1.1 , Table 6-6 through Table 6-13. The accident analyses in Chapter 13.0 identify IROFS for the prevention of nuclear criticality, which are summarized in Table 6-2, with reference to more detailed descriptions in Section 6.3.1.2.

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' ~ * ,* ~ ' NOllTifWHT MEDtCAl. tsOTOPES NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features Table 6-1. Summary of Confinement Engineered Safety Features (2 pages)

Detailed Engineered safety SSCs providing engineered description feature IROFS Accident(s) mitigated safety features section Confinement . Equipment . Confinement enclosures 6.2.1.1

.includes: malfunction and/or

. Zone including penetration seals through Hot cell liquid confinement RS-01

. mai ntenance Hazardous chemical I exhaust ventilati on system, including ducting, 6.2. 1.6

. boundary Hot cell RS-03 spills

. Zone filters, and exhaust stack I inlet ventil ati on system, secondary including ducting, filters, and confinement bubble-ti ght isolati on dampers Ventilation control system

. boundary Hot cell shielding RS-04 Secondary iodine removal bed Berms boundary Confinement IROFS Derived from Accident Analyses and Potential Technical Specifications Primary offgas relief RS-09 Dissol ver offgas failure .. Pressure reli ef device 6.2 .1.7.1 system during dissolution Pressure reli ef tank operation Active radiation RS-JO Transfer of high-dose Radiation monitoring and isolation 6.2.1.7.2 monitoring and process liquid outside the system for low-dose liquid isolation of low- hot cell shielding transfers dose waste transfer boundary Cask local RS-1 3 Target cladding leakage Local capture ventilation system 6.2.1.7.3 ventil ation during during shipment over closure lid during lid removal closure lid removal and docking preparations Cask docking port RS-15 Cask not engaged in cask Sensor system controlling cask 6.2.1 .7.4 enabler docking port prior to docking port door operation opening docking port door Process vessel FS-03 SSC damage due to Backup bottled nitrogen gas 6.2. 1.7.5 emergency purge hydrogen detlagration or supply system detonation Irradiated target FS-04 Dislodging the target . Cask lifting fixture design that 6.2.1.7.6 cask lifting fixture cask shield plug while workers present during . prevents cask tipping Cask lifting fixture design that target unloading prevents lift from toppling activities during a seismic event 6-2

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  • * ~ * . ~ ~ ; . NORTHWEn MEDICAL ISOTOP£S Table 6-1. Summary of Confinement Engineered Safety Features (2 pages)

Detailed Engineered safety SSCs providing engineered description feature IROFS Accident(s) mitigated safety features section Exhaust stack height FS-05 . Equipment . Zone I exhaust stack 6.2. 1.7.7 malfunction resulting

. in liquid spill or spray Carbon bed fire Double-wall piping CS-09 Solution spill in facility Double-wall piping for selected 6.2.1.7.7 area where spill transfer lines containment berm is neither practical nor desirable for personnel chemical protection purposes Backflow CS -1 8 High worker exposure Backflow prevention devices 6.2. 1.7.9 preventi on devices from backflow of high- located on process lines crossing Safe geometry day CS-1 9 dose solution the hot cell shi elding boundary tanks Dissolver offgas . Potential limiting Dissolver offgas iodine removal 6.2.1.8 iodine removal unit"

. control for operation Primary iodine control units (DS-SB-600A/B/C) system during normal operation Dissol ver offgas . Potential limiting Dissolver offgas primary adsorber 6.2 .1 .8 .2 primary adsorber"

. control for operati on Pri mary noble gas units (DS-SB-620A/B/C) control system during normal operation Dissolver offgas . Potential limiting . Dissolver offgas vacuum 6.2.1.8.3 vacuum receiver or vacuum pump" . control for operation Motive force for . receiver tanks (DS-TK-700A/B)

Dissolver offgas vacuum pumps dissolver offgas (DS-P-710A/B)

  • Examples of candidate technical specification rather than engineered safety feature.

IROFS item relied on for safety. SSC = structures, systems, and components.

Table 6-2. Summary of Criticality Engineered Safety Features (2 pages)

Engineered safety feature Interaction control spacing provided by passively designed fi xtures and workstation placement Pencil tank, vessel , or piping safe CS-04 CS-06 SSC features providing engineered safety features Defines spacing between SSC components using geometry to prevent nuclear criticality Defines dimensions of SSCs using geometry to 6.3.1.2. 1 6.3.1.2.2 geometry confinement using the prevent nuclear criticality diameter of tanks, vessels, or piping Pencil tank geometry control on fixed CS-07 Defines spacing between different SSCs using 6.3. 1.2.3 interaction spacing of individual tanks geometry to prevent nuclear criticality 6-3

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' ~ * . * ~ _- NORTHWEST llfDICAL ISOTOH:S NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features Table 6-2. Summary of Criticality Engineered Safety Features (2 pages)

Engineered safety feature Floor and sump geometry control on slab depth, and sump diameter or depth for floor dikes Double-wall piping CS-08 CS-09 SSC features providing engineered safety features Defines sump geometry and dimensions for SSCs using geometry to prevent nuclear criticality Defines transfer line leak confinement in 6.3.1.2.4 6.3. 1.2.5 locations where sumps under piping are neither feasible nor desirable Closed safe-geometry heating or CS-10 Closed-loop heat transfer fluid systems to 6.3.1.2.6 cooling loop with monitoring and prevent nuclear criticality or transfer ofhigh-alarm dose material across shielding boundary in the event of a leak into the heat transfer fluid Simple overflow to normally empty CS -11 Overflow to prevent nuclear criticality fro m 6.3 .1.2.7 safe-geometry tank with level alarm fi ssil e soluti on entering non-geometricall y favorable ventil ati on equipment Condensing pot or seal pot in CS-12 Seal pots to prevent nuclear criticality from 6.3.1.2.8 ventilation vent line fissile solution entering non-geometrically favorable ventilation equipment Si mple overflow to normally empty CS-1 3 Overflow to prevent nuclear criticality fro m 6.3. 1.2.9 safe geometry floor with level alarm fi ssile solution entering non-geometrically in the hot cell containment boundary fa vorable ventilation equipment Active discharge monitoring and CS- I 4 Information to be provided in the Operating 6.3 .1.2. l 0 isolation License Application Independent active discharge CS- 15 Info rmation will be provided in the Operating 6.3 .1.2. 11 monitoring and isolation License Appli cation Backflow prevention device CS-18 Backflow prevention to preclude fissile or high 6.3.1.2.12 dose solution from crossing shielding boundary to non-geometrically favorable chemical supply tanks and prevent nuclear criticality Safe geometry day tanks CS -1 9 Alternate backflow prevention device 6.3. 1.2. 13 Evaporator or concentrator CS-20 Prevent nuclear criticality from high-volume 6.3 .1.2.14 condensate monitoring transfer to non-geometrically favorable vessels in solutions with normally low fissile component concentrations Processing component safe volume CS -26 Defines volume of SSCs to prevent nuclear 6.3. 1.2. 15 confinement cri ti cali ty Closed heating or cooling loop with CS-27 Closed-loop, high-volume heat transfer fluid 6.3.1.2.16 monitoring and alarm systems to prevent nuclear criticality or transfer of high-dose material across shielding boundary in the event of a leak into the heat transfer fluid with normally low fissile component concentrations IROFS item relied on fo r safety. SSC = structures, systems, and components.

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.. Chapter 6.0 - Engineered Safety Features

  • ~* *~ * ~ UEDICAl lSOTOPH 6.2 DETAILED DESCRIPTIONS The PHA used to identify accidents in Chapter 13 .0, Section 13.1.3, assumed the following known and credited safety features , or IROFS, are in place for normal operations:
  • Hot cell shielding boundary, credited for shielding workers and the public from direct exposure to radiation (a normal hazard of the operation)
  • Hot cell confinement boundaries, credited for confining the fissile and high-dose solids, liquids, and gases, and controlling gaseous releases to the environment
  • Administrative and passive design features on uranium batch, volume, geometry, and interaction controls on the activities, credited for maintaining normal operations involving the handling of fissile material subcritical (the PHA identified initiators for abnormal operations that require further evaluation for IROFS satisfying the double-contingency principle)

This section provides detailed descriptions of the engineered safety features identified by the accident analyses shown in Chapter 13.0.

6.2.1 Confinement The PHA was based on a definition for confinement, as follows :

Confinement - An enclosure of the facility (e.g. , the hot cell area in the RPF) that is designed to limit the exchange of effluents between the enclosure and its external environment to controlled or defined pathways. A confinement should include the capability to maintain sufficient internal negative pressure to ensure inleakage (i .e., prevent uncontrolled leakage outside the confined area), but need not be capable of supporting positive internal pressure or significantly shielding the external environment from internal sources of direct radiation. Air movement in a confinement area could be integrated into the heating, ventilation, and air conditioning (HVAC) systems, including exhaust stacks or vents to the external environment, filters , blowers, and dampers (ANSl/ANS-15.1 , The Development of Technical Specifications for Research Reactors).

Confinement describes the low-leakage boundary surrounding radioactive or hazardous chemical materials released during an accident to facility regions surrounding the physical process equipment containing process materials. The confinement systems localize releases of radioactive or hazardous materials to controlled areas and mitigate the consequences of accidents.

The principal design and safety objective of the confinement system is to protect on -site workers, the public, and environment. Personnel protection control features (e.g., adequate shielding and ventilation control) will minimize hazards normally associated with radioactive or chemical materials.

The second design objective is to minimize the reliance on administrative or complex active engineering controls and provide a confinement system that is as simple and fail-safe as reasonably possible.

This subsection describes the confinement systems for the RPF. The RPF confinement areas will consist of hot cell and glovebox enclosures housing process operations, tanks, and piping. Confinement will be provided by a combination of the enclosure boundaries (e.g. , walls, floor, and ceiling), enclosure ventilation, and ventilation control system. The enclosure boundaries will restrict bulk quantities of process materials, potentially present in solid or liquid forms , to the confinement and limit in-leakage of gaseous components controlled by the ventilation system. The ventilation and ventilation control systems will restrict the gaseous components (including gas phase components and solid/liquid dispersions) to the confinement. Figure 6-1 provides a simplified schematic of the confinement ventilation system, which is described in more detail as the Zone I ventilation system in Chapter 9.0, "Auxiliary Systems."

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0 NOmtWtn MEDICAL ISOTOPfS NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

Source: Figure 2-5 ofNWMI-20 I 5-SDD-013, System Design Description for Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, March 2015.

Figure 6-1. Simplified Zone I Ventilation Schematic 6-6

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  • ~ *.* ~ NOmlWEST llEIHCAl tSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features The enclosure boundary of the hot cells will also function as biological shielding for operating personnel.

Shielding functions of the hot cells are discussed in Chapter 4.0, "Radioisotope Production Facility Description."

Hazardous chemical confinement will be provided by berms located within the RPF to confine spilled material to the vicinity where a spill may originate.

6.2.1.1 Confinement System Confinement system enclosure structures, ventilation ducting, isolation dampers, and Zone I exhaust filter trains are designated as IROFS . Table 6-3 provides a description of the system component safety functions . Figure 6-2, Figure 6-3 , and Figure 6-4 indicate the general location of confinement structure boundaries to the facility ground level, mechanical level, and lower level layouts, respectively. The confinement system is an engineered safety feature that performs the functions identified by IROFS RS-01, RS-03 , and RS-04 in Chapter 13.0.

Table 6-3. Confinement System Safety Functions System, structure, component Description Classification Zone I enclosure inlet isolation dampers and Provide confi nement isolation at Zone I/Zone II IROFS ducting leading from isolation dampers to enclosure boundaries enclosures Zone I enclosure exhaust ducting leading from Provides confinement to the confinement exhaust IROFS enclosures to the exhaust stack, filters, and boundary exhaust stack Process vessel vent exhaust ducting leading Provides confinement to the confinement exhaust IROFS from process vessels to Zone I exhaust plenum boundary Ventilation control system Provides stack monitoring and interlocks to IROFS monitor discharge and signal changing on service tilter trains during normal and abnormal operation Secondary iodine removal bed Mitigates a release of the iodine inventory in the IROFS dissolver offgas treatment system Hot cells, tank vaults, and glovebox enclosure Provide solid, liquid, gas confinement IROFS structures IROFS = item re lied on for safety.

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[Proprietary Information]

Source: Figure 2- 1 of NWMJ-2015-SDD-01 3, System Design Description for Ventilation , Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, March 20 15.

Figure 6-2. Ground Level Confinement Boundary 6-8

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[Proprietary Information]

Source: Figure 2-2 ofNWMl-2015-SDD-O 13, System Design Description for Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, March 2015.

Figure 6-3. Mechanical Level Confinement Boundary 6-9

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[Proprietary Information]

Source: Figure 2-3 ofNWMI-2015-SDD-013, System Design Description for Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, March 2015.

Figure 6-4. Lower Level Confinement Boundary During normal operation, passive confinement is provided by the contiguous boundary between the hazardous materials and the surrounding environment and is credited with confining the hazards generated as a result of accident scenarios. The boundary includes the enclosure structures and extension of the structures through the Zone I ventilation components. The intent of the passive boundary is to confine hazardous materials while also preventing disturbance of the hazardous material inventory by external energy sources. This passive confinement boundary extends from the isolation valve downstream of the intake high-efficiency particulate air (HEPA) filter to the exhaust stack.

An event that results in a release of process material to a confinement enclosure will be confined by the enclosure structural components. Each process line that connects with vessels located outside of a confinement boundary with vessels located inside a confinement boundary will be provided with backflow prevention devices to prevent releases of gaseous or liquid material. The backflow prevention devices on piping penetrating the confinement boundary are designed as passive devices and will be located as near as practical to the confinement boundary or take a position that provides greater safety on loss of actuating power.

The consequences of an uncontrolled release within a confinement enclosure, and the off-site consequences of releasing fission products through the ventilation system, will be mitigated by use of an active component in the form of bubble-tight isolation dampers as IROFS on the inlet ventilation ducting to each enclosure.

This engineered safety feature reduces the ducting to the confinement volume that needs to remain intact to achieve enclosure confinement. The dampers will close automatically (fail-closed) on loss of power, and the ventilation system will automatically be placed into the passive ventilation operating mode .

Overall performance assurance of the active confinement components will be achieved through factory testing and in-place testing. Duct and housing leak tests will be performed in accordance with minimum acceptance criteria, as specified in ASME AG- I, Code on Nuclear Air and Gas Treatment. Specific owner requirements with respect to acceptable leak rates will be based on the safety analysis.

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.**.*.*. NWMl-2013-021, Rev. 1

~ *.*! *

. **NOmlWEST MEDICAL lSOTOrU Chapter 6.0 - Engineered Safety Features Berms will employ a passive confinement methodology. Passive confinement will be achieved through a continuous boundary between the hazardous materials and the surrounding area. In the event of an accidental release, the hazardous liquid will be confined to limit the exposed surface area of the liquid.

6.2.1.2 Accidents Mitigated The hot cell confinement system and shielding boundary are credited as being in place by the accident analysis in Chapter 13.0, Section 13.1.3. l. Accidents mitigated consist of equipment malfunction events that result in the release of radioactive material or hazardous chemicals to a confinement enclosure. The confinement system is also credited with mitigating the impact of a non-specific initiating event resulting in release of the iodine inventory in the dissolver offgas treatment system.

6.2.1.3 Functional Requirements Functional requirements of the confinement structural components include:

  • Capturing and containing liquid or solid releases to prevent the material from exiting the boundary and causing high dose to a worker or member of the public or producing significant environment contamination
  • Preventing spills or sprays of radioactive solution that are acidic or caustic from causing adverse exposure to personnel through direct contact with skin, eyes, and mucus membranes where the combination of chemical exposure and radiological contamination would lead to serious injury and long-lasting effects Functional requirements of the confinement ventilation components include:
  • Providing negative air pressure in the hot cell (Zone I) relative to lower zones outside of the hot cell using exhaust fans equipped with HEPA filters and high-efficiency gas adsorbers (HEGA) to reduce the release of radionuclides (both particulate and gaseous) outside the primary confinement boundary to below Title I 0, Code of Federal Regulations, Part 20, "Standards for Protection Against Radiation" (10 CFR 20) release limits during normal and abnormal operations.
  • Mitigating high-dose radionuclide releases to maintain exposure to acceptable levels to workers and the public in a highly reliable and available manner. The hot cell secondary confinement boundary will perform this function using a system of passive and active engineered features to ensure a high level of reliabi lity and availability.
  • Removing iodine isotopes present in the process vessel vent under accident conditions to comply with 10 CFR 70.61, "Performance Requirements," for an intermediate consequence release.

Berms confining potential hazardous chemical spills are designed to hold the entire contents of the container in the event the container fails .

6.2.1.4 Confinement Components The following components are associated with the confinement barriers of the hot cells, tank vaults, and gloveboxes. The specific materials, construction, installation, and operating requirements of these components are evaluated based on the safety analysis.

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~.

  • ~ * *!'
  • NMTNWEn MElHCAl tSOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Confinement structural components include the following.
  • Sealed flooring will provide multiple layers of protection from release to the environment.
  • Diked areas will contain specific releases. Sumps of appropriate design will be provided with remote operated pumps to mitigate liquid spills by capturing the liquid in appropriate safe-geometry tanks.
  • In the molybdenum-99 (99 Mo) purification clean room, smaller confinement catch basins will be provided under points of credible spill potential in addition to the sealed floor.
  • Entryway doors into a designated liquid confinement area will be sealed against credible liquid leaks to outside the boundary.
  • Piping penetrations and air ducts will be located to minimize the potential for liquid leaks across the confinement boundary.

Ventilation system components that are credited include the following.

  • Zone I inlet HEPA filters will provide an efficiency of greater than 99.9 percent for removal of radiological particulates from the air that may reverse flow from Zone I to Zone IL
  • Zone I ducting will ensure that negative air pressure can be maintained by conveying exhaust air to the stack.
  • Bubble-tight dampers will be provided to comply with the requirements of ASME AG-1 ,

Section DA-5141. Ventilation ductwork and ductwork support materials will meet the requirements of ASME AG-1. Supports will be designed and fabricated in accordance with the requirements of ASME AG-1.

  • Zone I exhaust train HEPA filters will provide an efficiency of greater than 99.95 percent for removal of radiological particulates from the air that flows to the stack.
  • Zone I exhaust train HEGA filters will provide an efficiency of greater than 90% for removal of iodine.
  • The Zone I exhaust stack will provide dispersion of radionuclides in normal and abnormal releases at a discharge point of 23 meters (m) (75 feet [ft]) above the building ground level.
  • Stack monitoring and interlocks will monitor discharge and signal changing of service filter trains during normal and abnormal operations.

Secondary process offgas treatment iodine removal beds (VV-SB-520) will mitigate an iodine release.

6.2.1.5 Test Requirements Engineered safety features will be tested to ensure that components maintain operability and can provide adequate confidence that the safety system performs satisfactorily during postulated events. The confinement engineered safety features that initiate the system interlocks are designed to permit testing during plant operation.

The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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  • ~ ~.* ! : NOtlTKWUTMEOK:AltSOTOP£S NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.2.1.6 Design Basis Codes and standards are discussed in Chapter 3 .0, "Design of Structures, Systems, and Components."

The design bases for Zone I and Zone II ventilation systems are described in Chapter 9.0. The design basis of confinement enclosure structures is described in Chapter 4.0. Chapter 7.0, "Instrumentation and Control Systems," identifies the engineered safety feature-related design basis of the ventilation control system.

The following information was developed for the Construction Permit Application to describe the process offgas secondary iodine removal bed:

  • Sorbent bed of [Proprietary Information]
  • Iodine removal efficiency greater than [Proprietary Information]
  • Nominal superficial gas flow velocity of [Proprietary Information]
  • Nominal sorbent bed operating temperature of less than [Proprietary Information]
  • Nominal sorbent bed depth of [Proprietary Information]
  • Nominal gas relative humjdity less than [Proprietary Information]

Additional detailed information on the process offgas iodine retention bed design basis will be developed for the Operating License Application.

Potential variables, conditions, or other items that will be probable subjects of a technical specification associated with the RPF confinement systems and components are discussed in Chapter 14.0, "Technical Specifications."

6.2.1.7 Derived Confinement Items Relied on for Safety The following subsections describe additional engineered safety features that are derived from the accident analyses described in Chapter 13.0 and are projected technical specifications defining limited conditions for operation.

6.2.1.7.1 IROFS RS-09, Primary Offgas Relief System IROFS RS-09, "Primary Offgas Relief System," is identified by the accident analysis in Chapter 13 .0. As an active engineered control (AEC), the primary offgas relief system will be a component included in the offgas train for the two irradiated target dissolvers. The dissolver offgas system is intended to operate at a pressure that is less than the confinement enclosures to maintain gaseous components generated during dissolution within the vessels and route the gaseous components through the offgas treatment unit operations. The primary offgas relief system, or pressure relief tank, will be used to confine gases to the dissolver and a portion of the dissolver offgas equipment, if the offgas motive force (vacuum pumps) ceases operation during dissolution of a dissolver batch.

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. * ~ *.~ ! : . NOlmlWEn MEOM:Al tsoTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Figure 6-5 is a diagram of the dissolver offgas system process, which shows the pressure relief tank position in the offgas treatment equipment train. Figure 6-6 shows the location of the pressure relief tank within the RPF hot cell (identified as "pressure relief').

[Proprietary Information]

Figure 6-5. Dissolver Offgas System Engineered Safety Features 6-14

.-.~ *.

~ *.- !

  • ." .*NWMI NORTifWUT MEOtcAl lSOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

Figure 6-6. Dissolver Offgas Hot Cell Equipment Location The pressure relief tank will be evacuated to a specified, subatmospheric pressure prior to initiating dissolution of a target batch and selected valves (indicated as 2, 3, and 4 on Figure 6-5) closed. Valve I will be open during normal dissolver operation. An upset during the dissolver operation (e.g., loss of vacuum pump operation) will result in closing Valve 1 and opening Valve 2 to contain dissolver offgas within the dissolver and offgas vessels. Due to the short duration of dissolver operation, dissolution is assumed to go to completion independent of an offgas system upset. The pressure relief tank will contain the offgas as dissolution is completed.

Valves 3, 4, and 5 are provided for upset recovery. After correction of the upset cause, gases collected in the pressure relief tank will be routed to the downstream treatment unit operations via Valve 3 or returned to a caustic scrubber via Valve 4. Liquid condensed in the pressure relief tank as a result of activation will be routed to the dissolver offgas liquid waste collection tank via Valve 5 for disposal.

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. ' !~.~! : NOATHWESTMEOICAUSOTOPES Accident Mitigated

  • Irradiated target dissolver offgas system malfunctions, including loss of power during target dissolution operations System Components Pressure relief valves Pressure relief tank (DS-TK-500)

Functional Requirements

  • As an AEC, use relief device to relieve 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 dissolver
  • Prevent a failure of the primary confinement system by capturing gaseous effluents in a vacuum receiver tank Design Basis The following information was developed for the Construction Permit Application describing the pressure relief tank.
  • Pressure-relief tank sizing is based on a maximum dissolver batch of [Proprietary Information]

that has just started dissolution when the pressure relief event is initiated.

  • The non-condensable gas volume to the pressure relief tank is equivalent to all nitrogen oxide (NOx) generated by dissolution, plus the sweep gas flow for flammable hydrogen gas mitigation.
  • Worst-case reaction stoichiometry of [Proprietary Information] dissolved is used .
  • No credit is taken for reaction of N02 with water to produce nitric acid .
  • Dissolver gas additions, other than the minimum sweep gas flow for hydrogen mitigation, are terminated by the pressure relief event.
  • Gas contained by the pressure relief tank and associated dissolver offgas piping is saturated with water vapor.
  • The pressure change from [Proprietary Information], absolute activates the pressure relief tank .

Additional detailed information on the pressure relief tank design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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  • NOmlWEST llB>tCAl lSOTOf'H 6.2.1.7.2 IROFS RS-10, Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer IROFS RS-10, "Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer," is identified by the accident analyses described in Chapter 13.0. As an AEC, the recirculating stream and the 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 instrument will monitor the transfer lines to provide an open permissive signal to dedicated isolation valves.

Accident Mitigated Transfer of high-dose process liquid solutions outside the hot cell shielding boundary System Components Additional detailed information of the radiation monitor and isolation of low-dose waste transfers will be developed for the Operating License Application.

Functional Requirement Maintain worker and public exposure rates within approved limits Design Basis Additional detailed information of the radiation monitor and isolation of low-dose waste transfers will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.3 IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations IROFS RS-13 , "Cask Local Ventilation During Closure Lid Removal and Docking Preparations," is identified by the accident analyses described in Chapter 13 .0. As an AEC, a local capture ventilation system will be used over the irradiated target cask closure lid to remove any escaped gases from the worker breathing zone during removal of the closure lid, removal of the shielding block bolts, and installation of the lifting lugs.

Accident Mitigated

  • Irradiated target cladding fails during transportation, releasing gaseous radionuclides within the cask containment boundary System Components
  • Use a dedicated evacuation hood over the top of the cask during containment closure lid removal
  • Remove gases to the Zone I secondary confinement system for processing 6-17
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NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

. * ~ ~. *! ." . NOflTHWEn MmtcAL ISOTOPES Functional Requirement

  • Prevent exposure to workers by evacuating any high-dose gaseous radionuclides from the worker breathing zone and preventing immersion of the worker in a high-dose environment Design Basis The following information was developed for the Construction Permit Application describing the cask local ventilation system:

Use the local capture ventilation 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 Additional detailed information on the cask local ventilation system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.4 IROFS RS-15, Cask Docking Port Enabling Sensor IROFS RS-15 , "Cask Docking Port Enabling Sensor, is identified by the accident analyses described in Chapter 13 .0. 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.

Accident Mitigated

  • Cask lift failure occurs after shield plug removal (but before target basket removal) with targets inside the cask System Components Enabling contact signal and positive closure signal when the sensor does not sense a cask mated to the cask docking port, causing the cask docking port door to close Functional Requirement
  • Prevent the cask docking port door from being opened and allowing a streaming radiation path to areas accessible by workers Design Basis Detailed information on the system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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  • ~ ~**.~ : NOATHWUTMEDICA.LtsOTOHI 6.2.1.7.5 IROFS FS-03, Process Vessel Emergency Purge System IROFS FS-03 , "Process Vessel Emergency Purge System," is identified by the accident analyses described in Chapter 13 .0. Hydrogen gas will be evolved from process solutions through radiolytic decomposition of water in the high radiation fields . An air purge to the vapor space of selected tanks will be provided by the facility air compressors to control the hydrogen concentration from radiolysis in vessel vapor space to below the flammability limit for hydrogen. As an AEC, an emergency backup set of bottled nitrogen gas will be provided for all tanks that have the potential to evolve significant volumes of hydrogen gas through the radiolytic decomposition of water (in both a short- and long-term storage condition).

Accident Mitigated Hydrogen deflagration or detonation in a process vessel System Components Information will be provided in the Operating License Application.

Functional Requirement

  • Prevent development of an explosive hydrogen-air mixture in the tank vapor spaces to prevent the deflagration or detonation hazard Design Basis The following information was developed for the Construction Permit Application describing the process vessel emergency purge system:
  • Monitor the purge pressure going into the individual tanks and open an isolation valve on low pressure (setpoint to be determined) to restore the continuous sweep of the system using nitrogen
  • Provide sweep gas sufficient for the facility to allow repair of a compressed gas system outage
  • Activate by sensing low pressure on the normal sweep air system, introducing a continuous purge of nitrogen from a reliable emergency backup station of bottled nitrogen into each affected vessel near the bottom (e.g., through a liquid level detection leg) of the vessel
  • Dilute hydrogen as it rises to the top of the vessel and is vented to the respective vent system Additional detailed information on the process vessel emergency purge system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.6 IROFS FS-04, Irradiated Target Cask Lifting Fixture IROFS FS-04, "Irradiated Target Cask Lifting Fixture," is identified by the accident analyses described in Chapter 13.0. As a passive engineered control (PEC), the irradiated target cask lifting fixture will be designed to prevent the cask from tipping within the fixture and the fixture itself from toppling during a seismic event.

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' * ! *,~ ! ." . NOtmfWfST MEDtCAL ISOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Accident Mitigated

  • Dislodged irradiated target shipping cask shield plug in the presence of workers during target unloading activities System Components Detailed information on the system components will be developed for the Operating License Application.

Functional Requirements Detailed information on the system functional requirements will be developed for the Operating License Application.

Design Basis Detailed information on the system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.7 IROFS FS-05, Exhaust Stack Height IROFS FS-05, "Exhaust Stack Height," is identified by the accident analyses described in Chapter 13.0.

Accidents Mitigated Process solution spills and sprays Carbon bed fire System Component Zone I exhaust stack Functional Requirement

  • Provide an offgas release height for ventilation gases consistent with the stack height used as input to mitigated dose consequence evaluations.

Design Basis The Zone I exhaust stack height is 23 m (75 ft).

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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  • ~ *.*! 0 NOltTifWEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.2.1.7.8 IROFS CS-09, Double Wall Piping IROFS CS-09, "Double Wall Piping," is identified by the accident analyses in Chapter 13 .0. This IROFS has both a [Proprietary Information]

confinement and nuclear criticality prevention function. As a PEC, the piping system conveying fissile solution between credited confinement locations will be provided with a Figure 6-7. Proposed Location of Double-Wall Piping double-wall barrier to contain any spills that (Example) may occur from the primary confinement piping. This IROFS will be used at those locations that pass through the facility , where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes. Figure 6-7 provides an example location where IROFS CS-09 will be applied (e.g. , the transfer line between the recycle uranium decay tanks and the [Proprietary Information]).

Accident Mitigated Leak in piping that passes between confinement enclosures System Components The following double-wall piping segments are identified at this time:

  • Transfer piping containing fissile solutions traversing between hot cell walls
  • Transfer piping connecting the uranium product transfer send tank (UR-TK-720) and uranyl nitrate storage tank (TF-TK-200)
  • Other locations to be identified in final design Functional Requirements
  • Double-wall piping prevents personnel injury from exposure to acidic or caustic licensed material solutions conveyed in the piping that runs outside a confinement enclosure
  • Double-wall piping routes pipe leaks to a critically-safe leak collection tank or berm as a nuclear criticality control feature Design Basis The double-wall piping arrangement is designed to gravity drain to a safe-geometry set of tanks or to a safe geometry berm.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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  • NOmfWEST MUHCAl lSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.2.1.7.9 IROFS CS-18, Backflow Prevention Devices, and IROFS CS-19, Safe-Geometry Day Tanks IROFS CS-18, "Backflow Prevention Devices," and IROFS CS-19, "Safe-Geometry Day Tanks," are identified by the accident analyses in Chapter 13.0. As a PEC or AEC, chemical and gas addition ports to fissile process solution systems will enter a confinement enclosure through a backflow prevention device.

Backflow prevention devices and safe-geometry day tanks will provide alternatives for preventing process addition backflow across confinement boundaries. The 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. Therefore, these IROFSs have both a confinement and a nuclear criticality prevention function.

Accident Mitigated Backflow of process material located inside a confinement boundary to vessel located outside confinement via connected piping due to process upset.

System Components System component information will be provided in the Operating License Application.

Functional Requirements

  • Prevent fissile solutions and/or high dose solutions from backflowing from the tank into systems outside the confinement boundaries that may lead to accidental nuclear criticality or high exposures to workers
  • Provide each hazardous location with an engineered backflow prevention device that provides high reliability and availability for that location
  • Locate the backflow prevention device features for high-dose product solutions inside the confinement boundaries
  • Support the backflow prevention devices with safe-geometry day tanks located inside the confinement boundary
  • Direct spills from the backflow prevention device to a safe-geometry confinement berm Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

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~ *.* ~

  • NOmfWEST lllEDICAl lSOTOPll NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.2.1.8 Dissolver Offgas Systems 6.2.1.8.1 Dissolver Offgas Iodine Removal Unit A significant fraction of iodine entering the RPF in targets is projected to be released to dissolver offgas during target dissolution. The dissolver offgas iodine removal units will be included in the RPF as the primary SSCs for controlling the release of iodine isotopes to the environment or facility areas occupied by workers. Components of the dissolver offgas system, beginning with the iodine removal unit, will also be used to treat vent gas from the target disassembly system. Target disassembly vent gas is treated by dissolver offgas components for the Construction Application Permit configuration as a measure to mitigate the unverified potential for a release of fi ssion gas radionuclides during target transportation.

Figure 6-5 (Section 6.2 .1.7.1) shows the iodine removal unit position in the offgas treatment equipment train . The dissolver offgas iodine removal unit location in the facility is shown in Figure 6-6 (identified as "primary fission gas treatment").

Accidents Mitigated Projected limiting control for operation Required for normal operation and not for accident mitigation System Components Iodine removal unit A (DS-SB-600A)

Iodine removal unit B (DS-SB-600B)

Iodine removal unit C (DS-SB-600C)

Functional Requirement Remove iodine isotopes from the dissolver offgas during normal operations such that the dose to workers complies with 10 CFR 20.120 I , "Occupational Dose Limits for Adults," and the dose to the public complies with 10 CFR 20.1301 , "Dose Limits for Individual Members of the Public."

Design Basis The following information was developed for the Construction Permit Application describing each individual iodine removal unit:

Sorbent bed of [Proprietary Information]

Iodine removal efficiency greater than [Proprietary Information]

Nominal superficial gas flow velocity of [Proprietary Information]

Nominal sorbent bed operating temperature of [Proprietary Information]

Nominal sorbent bed depth of [Proprietary Information] , provi ding iodine removal capacity of greater than l year (yr).

Additional detailed information on the iodine removal unit design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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~; *

. ' ~ * .* ~ ." . '"*11fWDT MEDICAL lSOTOPf.S NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.2.1.8.2 Dissolver Offgas Primary Adsorber Noble gases (krypton [Kr] and xenon [Xe]) entering the RPF in targets are projected to be released to dissolver offgas during target dissolution. The dissolver offgas primary adsorber units will be included in the RPF as the primary SSCs for controlling the release of noble gas isotopes to the environment or facility areas occupied by workers. Components of the dissolver offgas system will also be used to treat vent gas from the target disassembly system. Target disassembly vent gas is treated by dissolver offgas components for the Construction Application Permit configuration as a measure to mitigate the unverified potential for a release of fission gas radionuclides during target transportation.

Figure 6-5 (Section 6.2.1. 7.1) shows the primary adsorber position in the offgas treatment equipment train. The dissolver offgas primary adsorber location in the facility is shown in Figure 6-6 (identified as "primary fission gas treatment").

Accidents Mitigated Projected limiting control for operation Required for normal operation and not for accident mitigation System Components Primary adsorber A (DS-SB-620A)

Primary adsorber B (DS-SB-620B)

Primary adsorber C (DS-SB-620C)

Functional Requirement Delay the release of noble gas isotopes via the dissolver offgas during normal operations such that the dose to workers complies with 10 CFR 20.1201 and the dose to the public complies with 10 CFR 20.1301.

Design Basis The following information was developed for the Construction Permit Application describing each individual primary adsorber unit:

  • Sorbent bed of [Proprietary Information]
  • Nominal sorbent bed operating temperature of [Proprietary Information]
  • Nominal gas relative humidity less than [Proprietary Information]
  • Average gas flow rate of [Proprietary Information]
  • Nominal superficial gas flow velocity of [Proprietary Information]
  • Delay time for release of Xe isotopes of 10 days and Kr isotopes of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (hr) (additional delay time is provided by the secondary adsorber)

Additional detailed information on the primary adsorber unit design basis will be developed for the Operating License Application.

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  • ~ *.* ~
  • NOmfWEIT MEDICAL ISOTOPES Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.8.3 Dissolver Offgas Vacuum Receiver/Vacuum Pump The dissolver offgas vacuum pump will provide the motive force for transferring offgas, generated in the dissolvers and disassembly equipment during operation, through the dissolver offgas equipment train while maintaining dissolver vessels at a pressure less than the equipment enclosure pressure. Vacuum receiver tanks will be provided as part of the motive force system to allow the vacuum pumps to cycle on and off less frequently and accommodate the wide variations in gas flow rate associated with a target dissolution cycle.

Figure 6-5 (Section 6.2 .1.7.1) shows the vacuum receiver tank and vacuum pump positions in the offgas treatment equipment train. The vacuum receiver tank and vacuum pump location in the facility is shown in Figure 6-3 in the vicinity of equipment identified for the process offgas secondary iodine removal bed.

Accidents Mitigated Projected limiting control for operation Required for normal operation and not for accident mitigation System Components Vacuum receiver tank A (DS-TK-700A)

Vacuum receiver tank B (DS-TK-700B)

Vacuum pump A (DS-P-710A)

Vacuum pump B (DS-P-710B)

Functional Requirements

  • Maintain the dissolver vessel gas space at a pressure less than the dissolver vessel enclosure pressure throughout the target dissolution cycle
  • Accommodate pressure drops associated with dissolver offgas unit operations over the range of gas flow rates generated in both dissolvers and the target disassembly equipment vent throughout a target dissolution cycle Design Basis The following information was developed for the Construction Permit Application describing the vacuum receiver tanks and vacuum pump:
  • Minimum inlet setpoint pressure of [Proprietary Information]
  • Maximum inlet setpoint pressure of [Proprietary Information]
  • Outlet pressure of [Proprietary Information]
  • Maximum sustained gas flow into [Proprietary Information]
  • Receiver tank provides a [Proprietary Information] with the vacuum pump off and inlet at the maximum sustained gas flow 6-25

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~~

  • .
  • NORTNWEn MEDIC.Al lSOTOrEI Additional detailed information on the vacuum receiver tank and vacuum pump design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.9 Exhaust System The ventilation exhaust system is described in Chapter 9.0, Section 9. I .2. Additional detailed information will be developed for the Operating License Application, including:

  • Describing changes in operating conditions in response to potential accidents and the mitigation of accident radiological consequences
  • Demonstrating how dispersion or distribution of contaminated air to the environment or occupied spaces is controlled
  • Identifying the design bases for location and operating characteristics of the exhaust stacks 6.2.1.10 Effluent Monitoring System Each RPF exhaust stack will include an effluent monitoring system. The monitoring system sample lines are designed to comply with ANSI NB.I , Sampling and Monitoring Releases ofAirborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities. Additional detailed information on the effluent monitoring systems will be developed for the Operating License Application.

6.2.1.11 Radioactive Release Monitoring The effluent monitoring system will provide flow rate, temperature, and composition inputs for dispersion modeling of releases from the exhaust stacks. These inputs will provide the capability for calculating potential exposures as a basis for actions to ensure that the public is protected during both normal operation and accident conditions. Additional detailed information on radioactive release monitoring will be developed for the Operating License Application.

6.2.1.12 Confinement System Mitigation Effects Detailed information describing the confinement system mitigation effects will be developed for the Operating License Application. This information will compare the radiological exposures to the facility staff and the public with and without the confinement system engineered safety feature. The comparison will be based on analyses showing airflow rates, reduction in quantities of airborne radioactive material by filter systems, system isolation, and other parameters that demonstrate the effectiveness of the system.

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  • NOmlWUT MEDtCAl tSOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features 6.2.2 Containment Containment for the RPF is defined based on NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing of Non -Power Reactors - Format and Content, Part 1 interim staff guidance.

Containments are required as an engineered safety feature on the basis of the radioisotope production facility design, operating characteristics, accidents scenarios, and location.

A potential scenario for such a release could be a significant loss of integrity of the radioisotope extraction system or the irradiated fu el processing system. The containment is designed to control the release to the environment of airborne radioactive material that is released in the facility even if the accident is accompanied by a pressure surge or steam release.

The NUREG-153 7 Part 1 interim staff guidance has been applied to the RPF target processing systems.

The current accident analysis described in Chapter 13 .0 has not identified a need for a containment system as an engineered safety feature.

6.2.3 Emergency Cooling System An emergency cooling system for the RPF is defined by NUREG-1537 Part 1 interim staff guidance.

In the event of the loss of any required primary or normal cooling system, an emergency cooling system may be required to remove decay heat from the fuel to prevent the failure or degradation of the gas management system, the isotope extraction sy stem, or the irradiated fu el processing system.

An evaluation of RPF cooling requirements provided in Chapter 5.0, "Coolant Systems," indicates that an emergency cooling system will not be required to avoid rupture of the primary process vessels. In addition, the current accident analysis described in Chapter 13 .0 has not identified a need for an emergency cooling system as an engineered safety feature.

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0 NOITHWEST MmtCAl lSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.3 NUCLEAR CRITICALITY SAFETY IN THE RADIOISOTOPE PRODUCTION FACILITY The RPF design will provide adequate protection against criticality hazards related to the storage, handling, and processing of SNM outside a reactor. This is accomplished by:

  • Including equipment, facilities, and procedures to protect health and minimize danger to life or property
  • Ensuring that the design provides for criticality control, including adherence to the double-contingency principle
  • Incorporating a criticality monitoring and alarm system into the facility design For the Construction Permit Application, the design has assumed that a nuclear criticality accident is a high-consequence event independent of whether shielding or other isolation is available between the source of radiation and facility personnel. While not considered likely at this time, justification for considering criticality events as other than a high-consequence event will be provided in the Operating License Application, if this assumption is changed for specific locations by future design activities.

The nuclear criticality safety program defines the programmatic elements that work in concert to maintain criticality controls throughout the operating life of the RPF. The nuclear criticality safety program and facility design are developed based on the following American National Standards Institute/ American Nuclear Society (ANSI/ ANS) standards, with exceptions described in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 3. 71 , Nuclear Criticality Safety Standards for Fuels and Material Facilities.

  • ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors
  • ANSI/ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materials
  • ANSI/ ANS-8. l 0, Criteria for Nuclear Criticality Safety Controls in Operations With Shielding and Confinement
  • ANSI/ ANS-8.19, Administrative Practices for Nuclear Criticality Safety
  • ANSI/ ANS-8.20, Nuclear Criticality Safety Training
  • ANSl/ANS-8.22, Nuclear Criticality Safety Based on Limiting and Controlling Moderators
  • ANSI/ ANS-8.23 , Nuclear Criticality Accident Emergency Planning and Response
  • ANSl/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations
  • ANSl/ANS-8.26, Criticality Safety Engineer Training and Qualification Program For the Construction Permit Application, no deviations from standards or requirements have been identified that would require development of equivalent requirements for the RPF.

NWMI commits to the following standards and guides during design and construction:

  • ANSI/ANS-8. l - Nuclear criticality safety practices, including administrative practices, technical practices, and validation of a calculational method 6-28

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! *.* ! . NOmlWUT MEJHCAL ISOTOP'(S Chapter 6.0 - Engineered Safety Features

  • ANSVANS-8.3 - Criticality accident alarm system (CAAS) placement analysis and procedure development; the standard is used as modified by NRC Regulatory Guide 3.71
  • ANSVANS-8 .19 - NWMI nuclear criticality safety program development as it applies to organization, administration, roles, and responsibilities
  • ANSVANS-8 .20 - Nuclear criticality safety staff and contractor qualification and training procedure development
  • ANSVANS-8 .24 - Validation of a calculational method
  • NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility - Guidance for meeting 10 CFR 70.61
  • NUREG/CR-4604, Statistical Methods for Nuclear Material Management - Guidance for normality testing of the data from critical experiment calculations
  • NUREG/CR-6698, Guide for Validation ofNuclear Criticality Safety Calculational Methodology -

Guidance for validation of a calculational method The nuclear criticality safety program includes the following elements:

  • Responsibilities Criticality safety evaluations Criticality safety control implementation Nuclear criticality safety training Criticality safety assessments Criticality prevention specifications Operating procedures and maintenance work Criticality safety postings Fissile material container labeling, storage, and transport Criticality safety nonconformance response Criticality safety configuration control Criticality detector and alarm system Criticality safety guidelines for firefighting Emergency preparedness plan and procedures Components of the nuclear criticality safety program specifically implemented during the design and construction phases of the RPF will include:

Nuclear criticality safety program policy Nuclear criticality safety program procedure Nuclear criticality safety evaluation procedure Nuclear criticality safety technical/peer review procedure

  • Nuclear criticality safety engineer training and qualification procedure Nuclear criticality safety validation procedure Preliminary descriptions of the nuclear criticality safety program elements developed for the Construction Permit Application are summarized below. Modifications to the nuclear criticality safety program elements are anticipated as the design matures and will be included in the Operating License Application.

Responsibilities This element describes the responsibilities of management and staff in implementing the nuclear criticality safety program.

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  • General facility management will ensure that the nuclear safety function is as independent as practical from the facility operating functions .
  • A Nuclear Criticality Safety Manager will be assigned and responsible for overall coordination, maintenance, and management of the nuclear criticality safety program.
  • A Criticality Safety Representative will be assigned who is qualified to interpret criticality safety requirements and serve as a liaison between custodians of fissionable material and other operations, advising operating personnel and supervisors on questions concerning conformance to criticality safety requirements.
  • Qualified Criticality Safety Engineers will responsible for performing criticality analyses and evaluations of systems, maintaining current verified and validated criticality computer codes, advising staff on technical aspects of criticality controls, and supporting/participating in inspections and management assessments.
  • Operations management will be responsible for establishing the responsibility for criticality safety throughout the operations organization, communicating criticality safety responsibilities for each individual involved in operations, ensuring that controls identified by CSEs are implemented, ensuring each worker has necessary training and qualifications, and ensuring that procedures that include controls significant to criticality safety are prepared before operations commence.
  • Supervisors and workers will be responsible for completing training before performing fissile material operations, understanding and ensuring compliance with all applicable criticality safety controls, and reporting any proposed change in fissile material operations to the Criticality Safety Representative for evaluation and approval before the operation commences.

Criticality Safety Evaluations This element describes the process for preparing CSEs that demonstrate fissile material operation will be subcritical under both normal and credible abnormal conditions.

  • CSEs will determine, identify, and document the controlled parameters and associated limits on which criticality safety depends .
  • CSEs will be required to evaluate normal operations, and contingent and upset conditions .
  • Preliminary CSEs prepared for the Construction Permit Application, including verification and validation of supporting computer codes, are described in Section 6.3.1.1 and provide examples of the CSEs.
  • Design changes impacting criticality will be reviewed by the Criticality Safety Representative .
  • CSEs will be independently reviewed to confirm the technical adequacy of the evaluation prior to commencing new or modified fissile material operations.

Nuclear criticality safety limits established for controlled parameters in the NWMI facility processes will ensure that all nuclear processes are subcritical, including an adequate margin of subcriticality for safety in accordance with the Interim Staff Guidance augmenting NUREG-15 37, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria, Part 2, Section 6.b.3 (NRC, 2012). Monte-Carlo N Particle (MCNP) calculation results used to set limits on parameters are compared to the upper subcritical limit (USL) established in the NWMI MCNP code validation report ([Proprietary Information]), after applying a 2cr calculation uncertainty.

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  • NOfllllfW(n MEDICAl tsOTOPf:I NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features The USL includes the method bias and uncertainty established in [Proprietary Information] and a 0.05 ~k margin of subcriticality. In addition, the area of applicability, also established in [Proprietary Information] , is checked to ensure that the NWMI RPF process model physics and materials are within the bands of applicability. If either the physics or materials are outside the bands of applicability, an additional margin of subcriticality will be applied.

Criticality Safety Control Implementation This element describes the process for implementing criticality safety controls defined by the CSEs .

  • Implementation includes confirming that:

All required engineered criticality safety controls are maintained by a configuration management system.

Equipment dimensions, volumes, or other features relied on for controls are with limits documented in the CSEs.

Administrative criticality safety controls from CSEs are implemented in written operating and maintenance procedures.

  • Fissile material inventories will be monitored and incorporated into implementation of criticality safety controls.
  • Access to fissionable material will be controlled .

Nuclear Criticality Safety Train ing This element describes the training program for nuclear criticality safety based on the worker' s duties and responsibilities.

  • This training program is developed and implemented with input from the nuclear criticality safety staff, training staff, and management, with a focus on:

Knowledge of the physics associated with nuclear criticality safety Analysis of jobs and tasks to determine the knowledge a worker must have to perform tasks efficiently Design and development of learning objectives based on the analysis of jobs and tasks that reflect the knowledge, skills, and abi lities needed by the worker Implementation of revised or temporary operating procedures Testing methods to demonstrate competence in training materials dependent on an individual's responsibility Training records maintenance

  • General training on criticality hazards and alarm responses will be provided to all RPF personnel and visitors.
  • Operators responsible for some aspect of nuclear criticality safety will :

Satisfy defined minimum initial qualifications Complete an initial criticality safety training course designed for operators Perform periodic requalification training

  • Management, operations supervisor, and technical staff responsible for some aspect of nuclear criticality safety will:

Satisfy defined minimum initial qualifications Complete an initial criticality safety training course designed for managers and engineers 6-31

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  • NOflTHWEST UO>>CAl ISOfOP'ES Perform periodic requalification training The Criticality Safety Representative will:

Satisfy defined minimum initial qualifications Complete an initial criticality safety program designed for the Criticality Safety Representative Demonstrate competence in understanding facility nuclear criticality controls and procedures Perform periodic requalification training

  • Criticality Safety Engineers will be trained and qualified in accordance with ANSI/ANS-8 .26 .

Nuclear criticality safety staff members and contract support will meet the qualification and training requirements specified in the NWMI nuclear criticality safety qualification and training program. The NWMI nuclear criticality safety qualification and training program is compliant with ANSI/ ANS 8.26.

Criticality Safety Assessments This element describes the periodic criticality safety inspections and assessments conducted to ensure that the criticality safety program is maintained at an adequate level for the RPF .

  • Annual criticality safety inspections will be conducted to satisfy the requirement of ANSI/ ANS-8.1 and 8.19 for operational reviews to be conducted at least annually.
  • Procedures will be developed for performing periodic criticality safety inspections. The facility Criticality Safety Representative and inspection team will comprise individuals (typically from Engineering) who are knowledgeable of criticality safety, and who, to the extent practicable, are not immediately responsible for the operation being inspected.
  • Facility inspections are conducted to verify that the facility configuration and activities comply with the nuclear criticality safety program. Facility inspections generally consist of observation of task preparation and verification of field procedures and training.
  • Management assessments will be conducted of the nuclear criticality safety program. These assessments will be led by the Nuclear Criticality Safety Manager, with assistance from other members of the criticality safety staff. The criticality safety staff is independent of the operating organization and not directly responsible for the operations.
  • Records generated during performance of criticality safety inspections and assessments will be included in a criticality safety inspection report or specialty assessment report.

An audit to assess the overall effectiveness of the nuclear criticality safety program will be performed at least once every three years. The audit will be led by a qualified senior criticality safety engineer from outside the NWMI organization. The senior nuclear criticality safety engineer conducting the audit will be independent of the NWMI program and will not have participated in any nuclear criticality safety evaluation that will be a subject of the audit. In addition to the triennial audit from an outside organization, NWMI senior management will perform periodic audits of the NWMI nuclear criticality safety program. The senior manager will be chosen from an NWMI organization other than the nuclear criticality safety group. The NWMI Quality Assurance Manager will select and assign auditors who are independent of the NWMI nuclear criticality safety program.

Criticality Prevention Specifications This element describes the requirements for the criticality prevention specifications used to implement limits and controls established in the CS Es for safe handling of fissionable material and implement the ANSI/ANS-8 series requirement for clear communication of criticality safety limits and controls.

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  • Each criticality prevention specification will:

Be based on an approved CSE and refer to the CSE used as a specification source Be prepared by either the Criticality Safety Representative of a qualified Criticality Safety Engineer Emphasize limits controllable by the operator Have clear and unambiguous meaning and be written, to the extent practical, using operations terminology with common units of measure Operating Procedures and Maintenance Work This element describes the requirements for implementing nuclear criticality controls in written procedures for operations and maintenance work.

  • Procedures will meet the intent of ANSVANS-8 .19 .
  • Procedures for operations and maintenance work will be prepared according to approved procedure control programs, developed and maintained to reflect changes in operations, and written so that no single inadvertent failure to follow a procedure can cause a criticality accident.
  • Operating procedures will include:

Controls and limits significant to nuclear criticality safety of the operation Periodic revisions, as necessary Periodic review of active procedures by supervisors

  • Operating procedures will be supplemented by criticality safety postings on equipment or incorporated in operating checklists.
  • Maintenance work procedures associated with SSCs affecting nuclear criticality safety will be reviewed by the Criticality Safety Representative or a Criticality Safety Engineer for compliance with nuclear criticality safety limits based on current RPF conditions present prior to initiating each maintenance evolution.

Criticality Safety Postings

  • Criticality safety postings will be developed for the Operating License Application .

Fissile Material Container Labeling, Storage, and Transport

  • Fissile material container labeling, storage, and transport will be developed for the Operating License Application.

Criticality Safety Nonconformance Response This element describes the response to deviations from defined nuclear criticality safety controls .

  • Deviations from procedures and unforeseen alterations in process conditions that affect criticality safety will be immediately reported to management and the Criticality Safety Representative or a Criticality Safety Engineer.
  • NWMI management will provide the required notifications of the deviation to the U.S. Nuclear Regulatory Commission Operations Center.

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  • The Criticality Safety Representative or a Criticality Safety Engineer will support an investigative team comprising, at a minimum, the Operations Manager and operations personnel familiar with the operation in question during the development of a recovery plan for safely returning to compliance with the procedures.
  • The deviation will be corrected per the recovery plan and the incident documented .
  • Action is to be taken to ensure that a similar situation does not exist in another part of the facility and to prevent recurrence of the nonconformance.

Criticality Safety Configuration Control This element describes the criticality safety configuration controls.

  • The primary criticality safety control , performed at the start of a proposed activity or equipment change, is for the Criticality Safety Representative to confirm if an existing active CSE is applicable.
  • All dimensions, nuclear properties, and other features on which reliance is placed will be documented and verified prior to beginning operations, and control will be exercised to maintain them.
  • The nuclear criticality safety staff will provide technical guidance for the design of equipment and processes and for the development of operating procedures.
  • All proposed criticality safety-related changes to design or process configuration will be reviewed by a Criticality Safety Representative or Criticality Safety Engineer to ensure that the change can be performed under an approved CSE.
  • All operational changes that impact criticality safety will be documented and include proper approval designation.
  • The project manager will request a CSE applicability review at the earliest practical stage of a project to determine if there could be criticality safety impacts. If the potential exists for the physical configuration or operating parameters for new or revised equipment to affect criticality safety, the drawings and process control plans will be reviewed and approved by a Criticality Safety Representative or Criticality Safety Engineer, in compliance with standard engineering practices and procedures.
  • Facility and process change control will include the following .

The change management process will be in accordance with ANSVANS-8.19.

All dimensions, nuclear properties, and other features on which reliance is placed will be documented and verified prior to beginning operations, and control will be exercised to maintain them.

Changes that involve or could affect nuclear criticality controls will be evaluated under 10 CFR 50.59, "Changes, Tests, and Experiments."

Changes include new designs, operation, or modification to existing SSCs, computer programs, processes, operating procedures, or management measures.

Changes that involve or could affect nuclear criticality controls will be reviewed and approved by the Criticality Safety Representative.

Prior to implementing the change, the process will be determined to be subcritical (with an approved margin for safety) under both normal and credible accident scenarios .

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  • NOITIIW£ST MBMCAl. tsOTOPU Testing and Calibration of Active Engineered Controls
  • Testing and calibration of AECs will be developed for the Operating License Application .

Criticality Safety Guidelines for Firefighting

  • Criticality safety guidelines for firefighting will be developed for the Operating License Application.

Emergency Preparedness Plan and Procedures This element describes the response to criticality accidents.

  • The CAAS will be used as described in Section 6.3.1.1 and provides for detection and annunciation of criticality accidents.
  • Emergency procedures will be prepared and approved by management.
  • Facility and off-site organizations expected to respond to emergencies will be informed of conditions that might be encountered.
  • Procedures will:

Designate evacuation routes that are clearly identified and follow the quickest, most direct routes practical Include assessment of exposure to individuals Designate personnel assembly stations outside the areas to be evacuated.

  • A method to account for personnel will be established and arrangements made in advance for the care and treatment of injured and exposed personnel.
  • The possibility of personnel contamination by radioactive material will be considered .
  • Personnel will be trained in evaluation methods, informed of routes and assembly stations, and drills performed at least annually.
  • Instrumentation and procedures will be provided for determining radiation in an evacuated area following a criticality accident and information collected in a central location.
  • Emergency procedures will be maintained for each area in which special nuclear material is handled, used, or stored to ensure that all personnel withdraw to an area of safety on sounding the alarm.
  • Emergency procedures will include conducting drills to familiarize personnel with the evacuation plan, designation ofresponsible individuals to determine the cause of the alarm, and placement of radiation survey instruments in accessible locations for use in such an emergency.
  • The current emergency procedures for each area will be retained as a record for as long as licensed special nuclear material is handled, used, or stored in the area.
  • Superseded sections of emergency procedures will be retained for three years after the section is superseded.
  • Fixed and personnel accident dosimeters will be provided in areas that require a CAAS .
  • Dosimeters will be readily available to personnel responding to an emergency and a method provided for prompt on-site dosimeter readouts.

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NCMllTIIWEST MfOICAL tsOTOPU NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features 6.3.1 Criticality Safety Controls The following sections describe criticality safety controls based on information developed for the Construction Permit Application . Section 6.3 .1.1 summarizes the results of preliminary CSEs that define PECs and AECs credited to satisfy the double-contingency control principle. Section 6.3.1 .2 summarizes IROFS related to preventing a nuclear criticality identified by the accident analyses described in Chapter I 3.0.

6.3.1.1 Preliminary Criticality Safety Evaluations A series of calculations were performed to support the Construction Permit Application investigating parameters associated with prevention of nuclear criticality in the current equipment configuration of major process systems. The calculations are described in the following documents:

  • NWMI-2015-CRITCALC-001 , Single Parameter Subcritical Limits for 20 wt% 235 U - Uranium Metal, Uranium Oxide, and Homogenous Water Mixtures
  • NWMI-2015-CRITCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution
  • NWMI-2015-CRITCALC-003 , 55-Gallon Drum Arrays
  • NWMI-2015-CRITCALC-005, Target Fabrication Tanks, Wet Processes, and Storage
  • NWMI-2015-CRITCALC-006, Tank Hot Cell Calculations were performed using the MCNP 6.1 code (LA-CP-13-00634, MCNP6 User Manual) .

Validation of the MCNP 6.1 code used in the calculations is described in [Proprietary Information]. The validation report documents the methodology and results for the bias and bias uncertainty values calculated for homogeneous and heterogeneous uranium systems for the MCNP 6.1 code system. The bias is expressed as USLs calculated using a facility-specific [Proprietary Information]. The primary focus of the validation was to determine the bias and bias uncertainty for intermediate-enriched uranium (IEU) systems. However, sufficient experiments for low-enriched uranium (LEU) and high-enriched uranium were included to demonstrate that there is no variation in the USL with varying enrichment.

Similarly, the primary focus of the validation was on thermal neutron energy systems. Sufficient experiments for intermediate and fast energy experiments were also included to demonstrate that there is no variation in the USL with increasing neutron energy.

The purpose of the computer code validation is to determine values of ketrthat are demonstrated to be subcritical (at or below the USL) for areas of applicability similar to systems or operations being analyzed. The USL is defined by Equation 6-1.

USL = 1.0 - Bias - Bias Uncertainty - Margin of Subcriticality Equation 6-1

[Proprietary Information] rearranges Equation 6-1 to produce a criterion for model cases that are considered acceptable as subcritical, as shown by Equation 6-2, and incorporates the margin of subcriticality in the USL as required by ANSI/ ANS-8 .1.

k eff + (2 X O"calc) ~ USL Equation 6-2 where keff is the MCNP calculated k-effective and G caJc is the MCNP calculation uncertainty.

[Proprietary Information]

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[Proprietary Information] indicates the validation is appropriate for homogeneous and heterogeneous IEU systems. A summary of the area of applicability is provided in Table 6-4. For systems outside the validation area of applicability, an increased margin of subcriticality value may be warranted, depending on the specific problem being analyzed. The analyst must document any extrapolation beyond the validation area of applicability, and justification must be documented for no adjustments to the margin of subcriticality when extrapolating.

Table 6-4. Area of Applicability Summary Parameter Area of Applicability Fissile material [Proprietary Information]

Fissile material form [Proprietary Information]

H/235 U ratio [Proprietary Information]

Average neutron energy causing fission [Proprietary Information]

Enrichment [Propri etary Information]

Moderating materials [Proprietary Information]

Reflecting materials [Proprietary Information]

Absorber materials [Proprietary Information]

Geometry [Proprietary Information]

  • Source: [Proprietary In format ion] .

ANECF = average neutron energy causing fission.

The RPF was divided into 13 activity groups for development of preliminary CS Es of the activities and associated equipment. Controlled nuclear criticality safety parameters vary with the activity group and are summarized in Table 6-5. A minimum of two nuclear criticality safety parameters are controlled to satisfy the double-contingency principle.

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Nuclear parameter Mass y y y y y y y N y y yb y y Geometry y y y y y ye ye y N y y y y Moderation y N N N N N N N N N N N N Interaction y y y y y y y y N y y y y Volume y y y y y y y N N N y N y Concentration/ N yd yd yd yd N N N ye ye ye N N density Reflection N N N N N N N N N N N N N Absorbers N N N N N N N N N N N N N Enrichmentf N N N N N N N N N N N N N

  • Deri ved from the indicated CSE reference document.

b Limited by nature of process in the a ir filtrati on.

c Limited by target design.

d Controll ed through input fiss ile mass.

e Limited by total uranium mass allowed in the system.

r Fac ili ty license limited to ::::20 wt% mu.

mu uranium-235. NWMI Northwest M edi cal Isotopes, LLC.

CSE = criticality safety evalu ation . y yes.

N = no.

The preliminary CSEs define a series of PECs, AECs, and administrative controls that are credited to satisfy the double-contingency control principle for prevention of nuclear criticality events such that at least two changes in process conditions must occur before criticality is possible. PECs, AECs, and administrative controls are described for the 13 activity groups in the following referenced tables:

  • NWMI-2015-CSE-01 , Irradiated Target Handling and Disassembly (Table 6-6)
  • NWMI-2015-CSE-02, Irradiated Low-Enriched Uranium Target Material Dissolution (Table 6-7)
  • NWMI-2015-CSE-03 , Moly bdenum-99 Recovery (Table 6-8)
  • NWMI-2015-CSE-04, Low-Enriched Uranium Target Material Production (Table 6-9)
  • NWMI-2015-CSE-05 , Target Fabrication Uranium Solution Processes (Table 6-9)
  • NWMI-20 l 5-CSE-06, Target Finishing (Table 6-9)
  • NWMI-2015-CSE-07, Target and Can Storage and Carts (Table 6-9)
  • NWMI-20 l 5-CSE-08, Hot Cell Uranium Purification (Table 6-10)
  • NWMI-2015-CSE-09, Waste Liquid Processing (Table 6-11)
  • NWMI-2015-CSE-10, Solid Waste Collection, Encapsulation, and Staging (Table 6-11)
  • NWMI-2015-CSE-l l , Offg as and Ventilation (Table 6-12)
  • NWMI-20 l 5-CSE-12, Target Transport Cask or Drum Handling - The shipping packages dictate design features that must be properly implemented for legal over-the-road transport. This CSE does not impose or credit additional passive controls other than those already incorporated in the respective shipping packages.

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  • NWMI-2015-CSE-13 , Analytical Laboratory (Table 6-13)

The CSEs will be updated for final design and the Operating License Application.

Criticality controls are selected based on the following order of preference:

Passive engineered controls Active engineered controls Enhanced administrative controls Administrative controls Note that a number of features listed in the preliminary CSEs are duplicated in multiple activity groups (e.g., the floor of cells is verified to be flat , with no collection points deeper than 3.5 centimeters [cm]).

Duplications are included in the current listings to clearly identify minor dimension variations that may exist in the defined features for different activity groups.

Table 6-6. [Proprietary Information] Double-Contingency Controls Identifier" Feature description and basis CSE-0I-PDF1 [Proprietary Information]

CSE-0 l-PDF2 [Proprietary Information]

CSE-Ol-PDF3 [Proprietary Information]

CSE-01-ACI [Proprietary Information]

CSE-Ol-AC2 [Proprietary Information]

CSE-0 l -AC3 [Proprietary Information]

CSE-0 l-AC4 [Proprietary Information]

a [Proprietary Information).

HEPA = high-efficiency particu late air. SPL single parameter limit.

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Double-Contingency Controls (2 pages)

Identifier* Feature description and basis CSE-02-PDF l [Proprietary Information]

CSE-02-PDF2 [Proprietary Information]

CSE-02-PDF3 [Proprietary Information]

CSE-02-PDF4 [Proprietary Information]

CSE-02-PDF5 [Proprietary Information]

CSE-02-PDF6 [Proprietary Information]

CSE-02-PDF7 [Proprietary Information]

CSE-02-PDF8 [Proprietary Information]

CSE-02-AEF I [Proprietary Information]

CSE-02-ACl [Proprietary Information]

CSE-02-AC2 [Proprietary Information]

a [Proprietary Information]

[Proprietary Information] = [Proprietary Information]

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.**.*.* ... NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

' ~ * .* ~

  • NOITMWUT M£DICAl ISOTOPU

[Proprietary Information]

Table 6-8. [Proprietary Information] Double-Contingency Controls (2 pages)

Identifier" Feature description and basis CSE-03-PDFI [Propri etary Information]

CSE-03-PDF2 [Proprietary Information]

CSE-03-PDF3 [Proprietary Information]

CSE-03-PDF4 [Proprietary Information]

CSE-03-PDF5 [Propri etary Information]

CSE-03-PDF6 [Proprietary Information]

CSE-03-PDF7 [Propri etary Information]

CSE-03-PDF8 [Proprietary Information]

CSE-03-PDF9 [Propri etary Information]

CSE-03-PDFIO [Proprietary Information]

CSE-03-PDF 11 [Proprietary Info rmation]

CSE-03-PDF12 [Proprietary Information]

CSE-03-AEFI [Propri etary Information]

CSE-03-ACl [Proprietary Information]

a [Proprietary Info rmation].

IX ion exchange. [Proprietary Info rmation] [Proprietary Informati on].

Mo = molybdenum.

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NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

6-42

..... NWMI

.**.*.*~..
  • ~ ~.* ~
  • NOmlWEST MEDICAi. tSOTOPf:S NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Table 6-9. [Proprietary Information] Double-Contingency Controls (8 pages)

Identifier Feature description and basis CSE-04-PDFI * [Proprietary Information]

CSE-04-PDF2* [Proprietary Information]

CSE-04-PDF3* [Proprietary Information]

CSE-04-PDF4* [Proprietary Information]

CSE-04-PDFS" [Proprietary Information]

CSE-04-PDF6* [Proprietary Information]

CSE-04-PDF7* [Proprietary Info rmation]

CSE-04-PDFS* [Proprietary Information]

CSE-04-PDF9* [Proprietary Information]

CSE [Proprietary Information]

PDF IO*

CSE [Proprietary Info rmation]

PDFJ J*

CSE [Proprietary Information]

PDF12*

CSE [Proprietary Information]

PDF1 3*

CSE [Proprietary Information]

PDF14*

CSE [Proprietary Information]

PDF 15*

CSE [Proprietary Information]

PDF16" CSE-04-AEFI" [Proprietary Information]

CSE-04-ACl * [Proprietary Information]

CSE-04-AC2" [Proprietary Information]

CSE-04-AC3" [Proprietary Information]

CSE-04-AC4" [Propri etary Information]

CSE-04-AC5" [Proprietary Information]

CSE-04-AC6" [Proprietary Information]

CSE-04-AC7" [Proprietary Information]

CSE-05 -PDF 1h [Proprietary Information]

CSE-05-PDF2h [Proprietary Information]

CSE-05-PDF3b [Proprietary Information]

CSE-05-PDF4b [Proprietary Information]

CSE-05-PDF5b [Proprietary Information]

CSE-05-PDF6b [Proprietary Information]

CSE-05-PDF7 h [Proprietary Information]

CSE-05-PDF8b [Proprietary Information]

6-43

-.;**:*...*.NWMI NWMl-2013-021, Rev. 1

~ ~*. ! * . NOmtWEn MEOK:Al lSOTOHS Chapter 6.0 - Engineered Safety Features Table 6-9. [Proprietary Information] Double-Contingency Controls (8 pages)

Identifier Feature description and basis CSE-05-AEFlb [Proprietary Information]

CSE-05-AEF2b [Proprietary Information]

CSE-05-AEF3b [Proprietary Information]

CSE-05-AC 1b [Proprietary Information]

CSE-05-AC2b [Proprietary Information]

CSE-05-AC3b [Proprietary Information]

CSE-06-PDFI c [Proprietary Information]

CSE-06-PDF2c [Proprietary Information]

CSE-06-ACl c [Proprietary Information]

CSE-06-AC2c [Proprietary Information]

CSE-06-AC3c [Proprietary Information]

CSE-06-AC4c [Proprietary Information]

CSE-06-ACSC [Proprietary Information]

CSE-06-AC6c [Proprietary Information]

CSE-07-PDFI d [Proprietary Information]

CSE-07-PDF2d [Proprietary Information]

CSE-07-PDF3d [Proprietary Information]

CSE-07-PDF4d [Proprietary Information]

CSE-07-ACl d [Proprietary Information]

CSE-07-AC2d [Proprietary Information]

CSE-07-AC3 d [Proprietary Information]

CSE-07-AC4d [Proprietary Information]

CSE-07-AC5d [Proprietary Information]

CSE-07-AC6d [Proprietary Information]

CSE-07-AC7d [Proprietary Information]

  • [Proprietary Information]

b [Proprietary Information]

c [Proprietary Information]

d [Proprietary Information]

ADUN acid-deficient uranium nitrate. UN = urani um ni tride.

DBE design basis earthquake. [Proprietary Information] [Proprietary Information]

u uranium.

6-44

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

  • ~ * .* ~
  • NOmfWUT MEDICAi. lS01WEI NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

6-45

.~; : . NWMI NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

~. ~ NOllTHW'EST MlOtCAL lSOl'OP£1

[Proprietary Information]

6-46

.., ** .....;*....NWMI

.!..... NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

~. ~ :

  • . NOlllfWHT MlDtCAl tsOTOPES

[Proprietary Information]

6-47

.:. .-.~ :.*. . NWMI NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features 0

~ ~. ~ ! ." . NGmfWEST MEDICAl. ISOl'Of'H

[Proprietary Information]

6-48

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

..... NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features

' !*.* ~ : NORTNWEn lrffOK:Al lSOTOPlS

[Proprietary Information]

6-49

.......*..*...*. NWMI

. ~.-.;

  • NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features
  • ~ * .* ~
  • NOUHWEST MEDICAL ISOTDPD

[Proprietary Information]

6-50

.......*. NWMI

~

  • !*.* ~ : NOllTNWUT llEDtCAl tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features Table 6-10. [Proprietary Information] Double-Contingency Controls (2 pages) ldentifiera Feature description and basis CSE-08-PDFl [Propri etary Information]

CSE-08-PDF2 [Proprietary Information]

CSE-08-PDF3 [Propri etary Info rmation]

CSE-08-PDF4 [Proprietary Information]

CSE-08-PDF5 [Propri etary In fo rmati on]

CSE-08-PDF6 [Proprietary Information]

CSE-08-PDF7 [Propri etary Info rmation]

CSE-08-PDF8 [Proprietary Information]

CSE-08-PDF9 [Propri etary Information]

CSE [Proprietary Information]

PDFlO CSE [Proprietary Information]

PDF!!

CSE [Proprietary Information]

PDF12 CSE-08-AEF I [Proprietary Information]

CSE-08-ACI [Proprietary Information]

CSE-08-AC2 [Propri etary Information]

  • [Proprietary Information]

DBE = design bas is earthquake. IX ion exchange.

6-51

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~ :

' ~ *.*!

  • NOATIIWUT ll£OICAl tSOTOPD NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

6-52

.*.*NWMI NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features
  • ~ *. * ~
  • MOmlWEST MEDtCAl. ISOTOPU Table 6-11. [Proprietary Information] Double-Contingency Controls (3 pages)

Identifier Feature description and basis CSE [Proprietary Information]

AEF l "

CSE-09-AC 1* [Proprietary Information]

CSE-09-AC2" [Proprietary Information]

CSE-09-AC3" [Proprietary Information]

CSE [Proprietary Information]

PDF l b CSE [Proprietary Information]

AEflb CSE-10-AC 1b [Proprietary Information]

CSE-10-AC2b [Proprietary Information]

CSE-10-AC3b [Proprietary Information]

CSE-10-AC4b [Proprietary Information]

CSE-1 O-AC5b [Proprietary Information]

CSE-10-AC6b [Proprietary Information]

CSE- I 0-AC7b [Proprietary Information]

CSE-10-AC&b [Proprietary Information]

CSE-1 O-AC9b [Proprietary Information]

  • [Proprietary Information]

b [Proprietary Information]

mu urani um-235. SPL single parameter limit.

HTC high-integrity container. u uranium.

RPF Radioisotope Production Facility.

6-53

......~.-.~......:.*.... NWMI NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features 0

~ ~~

  • . ."
  • NORTHWEST MEDtcAl tsOTOP£S

[Proprietary Information]

6-54

~ ..NWMI NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

. *. ~ ~* * ~ ." NOmfWEST MEDICAi. ISOTOPES

[Proprietary Information]

6-55

..~ ~ . NWMI NWMl-2013-021 , Rev. 1 Chapter 6.0 - Engineered Safety Features

  • ' ~ *.~ ~ :
  • NOmfWEST MEOICAl tS01VU Table 6-12. [Proprietary Information] Double-Contingency Controls (2 pages) ldentifiera Feature description and basis CSE-11-PDF l [Propri etary Information]

CSE- l l-PDF2 [Proprietary Information]

CSE-l l-PDF3 [Proprietary Information]

CSE- l l-PDF4 [Proprietary Information]

CSE- 11 -PDFS (Proprietary Info rmation]

CSE-I l-PDF6 (Proprietary Information]

CSE-l l-PDF7 [Proprietary Info rmation]

CSE-l l-PDF8 [Proprietary Information]

CSE-I 1-AEF l (Proprietary Information]

CSE-11-ACl [Proprietary Information]

  • [Proprietary Informati on]

DBE des ign basis earthquake. Mo molybdenum.

HEPA = high-effi ciency particulate air. NOx nitrogen oxide.

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  • ! *.* !
  • NOmlWEST MEDICAL ISOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

Table 6-13. [Proprietary Information] Double-Contingency Controls (2 pages) ldentifiera Feature description and basis CSE-13-PDF I [Proprietary Information]

CSE- l 3-PDF2 [Proprietary Information]

CSE-J 3-PDF3 [Proprietary Information]

CSE-13-ACI [Proprietary Information]

CSE-1 3-AC2 [Proprietary Information]

CSE-13-AC3 [Proprietary Information]

CSE-13-AC4 [Proprietary Information]

CSE-13-AC5 [Proprietary Information]

CSE-1 3-AC6 [Proprietary Information]

a [Proprietary Information]

R&D research and development. SPL single parameter limit.

RPF = Radioisotope Production Facility. u uranium.

6-57

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~ *.* ~ ' . NORTHWEST MED.CAL tSOTOP£S NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Each of the preliminary CSEs indicates that the process areas evaluated will be within the detector and alarm coverage of the CAAS. Evaluation of the CAAS coverage will be performed after final design is complete and prior to facility startup. To ensure the CAAS coverage is adequate for the facility, NWMI will conduct a coverage analysis using the minimum accident of concern that produces a detector response when the dose rate at the detector is equivalent to 20 rad/min at 2 m (6.6 ft) from the reacting material. Using the source from the minimum accident of concern, NWMI will conduct one-dimensional deterministic computations, when practical, to evaluate CAAS coverage. For areas of the facility where the use of one-dimensional deterministic computations is not practical , NWMI will use 3D Monte Carlo analysis to determine adequate CAAS coverage.

The CAAS will be designed to meet the following.

  • The facility CAAS:

Will be capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 radiation dose absorbed (rad) of combined neutron and gamma radiation at an unshielded distance of 2 m from the reacting material within 1 minute; two detectors will cover each area needing CAAS coverage Will use gamma and neutron sensitive radiation detectors that energize clearly audible alarm signals if an accidental criticality occurs Will comply with ANSl/ANS-8 .3, as modified by NRC Regulatory Guide 3.71 Will be appropriate for the type of radiation detected, the intervening shielding, and the magnitude of the minimum accident of concern Will be designed to remain operational during design basis accidents Will be clearly audible in areas that must be evacuated or there will be alternative notification methods that are documented to be effective in notifying personnel that evaluation is necessary

  • Operations will be rendered safe, by shutdown and quarantine, if necessary, in any area where CAAS coverage has been lost and not restored within a specified number of hours. The number of hours will be determined on a process-by-process basis, because shutting down certain processes, even to make them safe, may carry a larger risk than being without a CAAS for a short time. Compensatory measures (e.g., limiting access, halting SNM movement, or restoring CAAS coverage with an alternate instrument) when the CAAS is not functional will be determined for inclusion in the Operating License Application.
  • Emergency power will be provided to the CAAS by the uninterruptable power supply system .

6.3.1.2 Derived Nuclear Criticality Safety Items Relied on for Safety The following subsections describe engineered safety features that are derived from the accident scenarios that could result in a nuclear criticality, as described in Chapter 13.0.

6.3.1.2.1 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," is identified by the accident analyses in Chapter 13.0. During handling of uranium solids and solutions outside of processing systems under normal conditions, the material will be handled in safe masses controlled by either physical measurement or batch limits on well characterized devices.

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........ ;**...**NWMI NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

. ' ~ *: ! ." NomtWfST MEDICAL ISOTOH:S Solid uranium will be handled outside of processing systems during:

  • Receipt and processing of fresh uranium (and presumably shipment of spent uranium back to the supplier)
  • [Proprietary Information]
  • Fabrication of targets using [Proprietary Information] LEU target material (including movement of LEU target material to and from the fabrication workstation and handling of the completed targets)
  • Disassembly of targets following irradiation
  • Laboratory sampling and analysis activities (albeit in smaller quantities) .

Each activity is assigned a mass or batch limit for safe handling.

Accident Mitigated The accident occurs when a safe mass or batch limit is exceeded beyond some bounding extent based on the management measures on the control. Note that this accident involves normal condition criticality controlled limits for safe handling, and the upset represents failure of an associated administrative control.

The most limiting activity would involve processing the LEU target material from [Proprietary Information]. If the IROFS fails , accidental nuclear criticality is possible without additional control.

System Components As a PEC, fixed interaction control fixtures or workstations will be provided for holding or processing approved containers containing approved quantities of uranium metal, [Proprietary Information], batches of targets, and batches of samples.

Functional Requirements The fixtures are designed to hold only the approved container or batch and are fixed with 2-ft edge-to-edge spacing from all other fissile material containers, workstations, or fissile solution tanks, vessels, and ion exchange (IX) columns. Where LEU target material is handled in open containers, the design will prevent spills from readily spreading to an adjacent workstation or storage location.

Design Basis Final workstation and fixture spacing will be determined in final design when all process upsets are evaluated. Workstations with interaction controls include the following (not an all-inclusive listing):

  • [Proprietary Information]
  • [Proprietary Information]
  • Target basket fixture that provides safe spacing of a batch of targets from one another in the target receipt cell Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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..... ~.

  • ~ *.*! . NOtffHWfn MEOtCAl ISOTOPO NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.3.1.2.2 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," is identified by the accident analyses in Chapter 13.0. The PHA in Chapter 13.0 identified a number of individual potential initiating events that could lead to a spill of fissile solution from the geometrically safe confinement tanks, vessels, or piping that provide the primary safety functions of the processes. Four processing systems will handle fissile solutions:

Target fabrication (from the [Proprietary Information])

Target dissolution system First stage of molybdenum recovery and purification Entire uranium recovery and recycle system Three of these systems will be at least partially located within the hot cell wall boundary due to the high-dose of the fission products. Initiating events include the general categories of tank, vessel , or piping failure due to operator error (valves out of position), valves leaking, equipment leaking (pumps, piping, vessels, etc.), high pressure events from various causes including high temperature solutions (locked in boundary valves), hydrogen detonation, and exothermic reactions with the wrong resins or reagents used in the respective systems. Some of the initiators result in small leaks that are identified and mitigated (e.g., pump seal and small valve leaks). Over the life of the facility, these types of leaks are to be expected, but do not challenge the overall safety of RPF operations.

Accident Mitigated The accident of concern involves fissile process solution in quantities necessary to sustain accidental nuclear criticality. Larger catastrophic leaks or ruptures of equipment must occur for enough material to be released. Such leaks would represent a failure of the safe-geometry confinement IROFS for the respective equipment. Thus, scenarios leading to this accident sequence involve the failure of these IROFS. Due to the nature of the process, the worst-case accident involves the tanks with the largest capacity and the highest normal case concentrations.

System Components As a PEC, pencil tanks and other standalone vessels are designed and will be fabricated with a safe-geometry diameter for safe storage and processing of fissile solutions. The safe diameters of various tanks, vessels, or components will be provided in the Operating License Application.

Functional Requirements The safety function of safe diameter vessels is also one of confinement of the contained solution. The safe-geometry confinement of fissile solutions will prevent accidental nuclear criticality, a high consequence event. The safe-geometry confinement diameter will conservatively include the outside diameter of the tank wall or out to the outside diameter of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the vessels. Where insulation is used on the outside wall of a vessel, the insulation will be closed foam or encapsulated type (so as not to soak up solution during a leak) and will be compatible with the chemical nature of the contained solution.

Design Basis The safe-geometry diameter of tanks, vessels, and piping will be determined in final design after finalizing the reference CSEs. Note that preliminary vessel sizes for activity groups are listed in the double-contingency parameters described in Section 6.3.1 .1.

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0

......... .~.....*.* NWMI

~ ~* * !:. NatlTHWEST MEDtw. lSOTDPU NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.3 IROFS CS-07, Pencil Tank Geometry Control on Fixed Interaction Spacing of Individual Tanks IROFS CS-07, "Pencil Tank Geometry Control on Fixed Interaction Spacing of Individual Tanks," is identified by the accident analyses in Chapter I 3.0 (see description in Section 6.3.1.2.2).

Accident Mitigated See description in Section 6.3.1.2.2.

System Components As a PEC, pencil tanks and other standalone vessels (controlled with safe geometry or volume constraints) are designed and will be fabricated with a fixed interaction spacing for safe storage and processing of the fissile solutions. Tanks, vessels, and components requiring fixed interaction control spacing of the barrels within each set of pencil tanks and between various tanks, vessels, or components will be provided in the Operating License Application.

Functional Requirements The safety function of fixed interaction spacing of individual tanks 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 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 spacing will be measured from the outside of the respective tanks, vessels, or component or from the outside of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the vessels or component. The fixed interaction control distance from the safe slab depth spill containment berm will also be specified where applicable.

Design Basis Actual interaction control parameters will be defined during final design. In addition, the following generic interaction control parameters apply during design.

  • Connecting piping between fissile material components will not exceed a cross-sectional density to be determined during final evaluation of systems.
  • Edge-to-edge spacing between fissile material-bearing vessels and components and the concrete reflector presented by the hot cell shielding walls will be fixed at a distance to be determined during final evaluation of all components.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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  • !* * ~ ' NOmlWEST 11EDtCA1 ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.3.1.2.4 IROFS CS-08, Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Dikes IROFS CS-08, "Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Dikes," is identified by the accident analyses described in Chapter 13.0 (see description in Section 6.3 .1.2.2).

Accident Mitigated See description in Section 6.3 .1.2.2.

System Components As a PEC, the floor under designated tanks, vessels, and workstations will be constructed with a spill containment berm using a safe-geometry slab depth, and one or more collection sumps with diameters or depths, to be determined in final design.

Functional Requirements The safety function of a spill containment berm is to 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 overhead systems. 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 is operable if it contains reserve volume for the largest single credible spill.

Spill containment berm sizes and locations will be determined during final design.

Design Basis The safe-geometry slab depth under designated tanks, vessels, and workstations will be determined during final design after finalizing the reference CSEs. Note that the preliminary slab depth for the activity groups are listed in the double-contingency parameters described in Section 6.3 .1. l.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.5 IROFS CS-09, Double-Wall Piping IROFS CS-09, "Double Wall Piping," is identified by the accident analyses described in Chapter 13.0.

As a PEC, a piping system for conveying fissile solution between confinement structures will be provided with a double-wall barrier to contain any spills that may occur from the primary piping.

Accident Mitigated

  • Leak in piping that passes between confinement enclosures 6-63

.;*... .~.. . NWMI

  • ~ *  :

~ 0 NORTHW'EIT MEDICAL ISOTOf'll NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features System Components IROFS CS-09 is used at the locations listed below that pass through the facility where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes . The following double-wall piping segments are identified for criticality safety:

  • Transfer piping containing fissile solutions traversing between hot cell walls
  • Transfer piping connecting the uranium product transfer send tank (UR-TK-720) and the uranyl nitrate storage tank (TF-TK-200)
  • Any other locations in final design where fissile solution piping exits a safe-slab spill containment berm and enters another Functional Requirements 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 double-wall piping will also function as a barrier to prevent fissile solution from soaking into the concrete from lines passing through concrete walls where required by the criticality safety analysis (e.g., see PDF2 of Table 6-9). The secondary safety function of double-wall piping is to prevent personnel injury from exposure to acidic or caustic licensed material solutions that are conveyed in the piping.

Design Basis The double-wall piping arrangement is designed to gravity-drain to a safe-geometry set of tanks or a safe-geometry containment berm.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.6 IROFS CS-10, Closed Safe Geometry Heating/Cooling Loop with Monitoring and Alarm IROFS CS-10, "Closed Safe Geometry Heating or Cooling Loop with Monitoring and Alarm," is identified by the accident analyses in Chapter 13.0. 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 the heat transfer fluid boundary if the primary boundary fails.

Accidents Mitigated The dual-purpose safety function of the closed-loop system is to prevent (I) fissile process solution from causing accidental nuclear criticality, and (2) high-dose process solution from exiting the hot cell containment, confinement, or shielded boundary (or to prevent low-dose solution from exiting the facility, for systems located outside of the hot cell containment, confinement, or shielded boundary), and causing excessive dose to workers and the public, and/or causing a release to the environment.

System Components The closed loop steam and cooling water loop design is described in Chapter 9.0.

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  • ! *.* ! : NOmfWEIT MEDICAL ISOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Functional Requirements 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, or uranium concentration) 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 restore the closed loop integrity. Discharged solutions from this system will be handled as potentially fissile and sampled prior to discharge to a non-safe geometry.

Design Basis The closed loop steam and cooling water loop design is described in Chapter 9.0.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements wi ll be developed for the Operating License Application.

6.3.1.2.7 IROFS CS-11, Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm IROFS CS-I I, "Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm," is identified by the accident analyses described in Chapter 13 .0. As a PEC, 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) for each vented tank containing fissile or potentially fissile process solution for which this IROFS is assigned.

Accident Mitigated The overflow drain will prevent the process solution from entering the respective non-geometrically favorable sections of the process ventilation system and any chemical addition ports (where chemical addition ports enter through anti-siphon devices).

System Components Locations of the overflow and overflow collection tanks will be provided with the final design.

Functional Requirements The safety function of this feature is to prevent accidental nuclear criticality in non-geometrically favorable sections of the process ventilation system. The overflow will be directed to a safe-geometry storage tank. The overflow storage tank will normally be maintained 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 can be taken to restore operability of the safety feature by emptying the tank.

Design Basis Design basis information will be provided in the Operating License Application.

6-65

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. ;.-.~ *

  • .~:**:.*.<

NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features

' ~ *.~ ! ." . NORTHWUT MEOtCAl ISOTOKS Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.8 IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line IROFS CS-12, "Condensing Pot or Seal Pot in Ventilation Vent Line," is identified by the accident analyses described in Chapter 13 .0. As a PEC, a safe-geometry condensing pot or seal pot will be installed downstream of each tank for which this IROFS is assigned 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.

Accident Mitigated 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 over-capacity condition from defeating both.

System Components Locations of the condensing pots or seal pots and associated drain points will be provided with the final design.

Functional Requirements The safety function of the condensing or seal pots is to prevent accidental nuclear criticality in non-geometrically favorable sections 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 overflow capacity 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 can 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.

Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

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~ * .* ~ : NORTHWEST MEDtcAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features 6.3.1.2.9 IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary IROFS CS-13 , "Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary," is identified by the accident analyses described in Chapter 13.0. As a PEC, a simple overflow line will be installed above the high alarm setpoint for each vented tank containing fissile or potentially fis sile process solution for which this IROFS is assigned. The overflow will be directed to one or more safe-geometry flooring configurations with safe-geometry sumps.

Accident Mitigated This IROFS prevents accidental criticality by ensuring that overflowing fi ssile solutions are captured in a safe-geometry slab configuration with safe-geometry sumps.

System Components System component information will be provided in the Operating License Application.

Functional Requirements The floor areas (separated as needed to support operations in different hot cell areas) will normally be maintained empty. The floor area(s) will be equipped with a sump level alarm to inform the operator when use of the IROFS has been initiated.

Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.10 IROFS CS-14, Active Discharge Monitoring and Isolation IROFS CS-14, "Active Discharge Monitoring and Isolation," is identified by the accident analyses described in Chapter 13.0. Additional detailed information describing active discharge monitoring and isolation will be developed for the Operating License Application.

System Components System component information will be provided in the Operating License Application.

Functional Requirements Functional requirements information will be provided in the Operating License Application.

Design Basis Design basis information will be provided in the Operating License Application.

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!*: ! 0 NORTHWEST MEDICAi.. ISOTOPlS NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.11 IROFS CS-15, Independent Active Discharge Monitoring and Isolation IROFS CS-15 , " Independent Active Discharge Monitoring and Isolation," is identified by the accident analyses described in Chapter 13.0. Additional detailed information describing independent active discharge monitoring and isolation will be developed for the Operating License Application.

System Components System component information will be provided in the Operating License Application .

Functional Requirements Functional requirements information will be provided in the Operating License Application.

Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.12 IROFS CS-18, Backflow Prevention Device IROFS CS-18, "Back.flow Preventions Device," is identified by the accident analyses described in Chapter 13.0.

See description in Section 6.2.1.7.9.

Accident Mitigated See description in Section 6.2.1.7.9.

System Components See description in Section 6.2.1. 7.9.

Functional Requirements See description in Section 6.2.1.7.9.

Design Basis See description in Section 6.2.1. 7.9.

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6.3.1.2.13 IROFS CS-19, Safe-Geometry Day Tanks IROFS CS-19, "Safe Geometry Day Tanks," is identified by the accident analyses described in Chapter 13 .0. See description in Section 6.2.1.7.9.

Accident Mitigated See description in Section 6.2.1.7.9.

System Components See description in Section 6.2. 1.7.9.

Functional Requirements See description in Section 6.2.1.7.9.

Design Basis See description in Section 6.2.1.7.9.

Test Requirements See description in Section 6.2.1.7.9.

6.3.1.2.14 IROFS CS-20, Evaporator/Concentrator Condensate Monitoring IROFS CS-20, "Evaporator/Concentrator Condensate Monitoring," is identified by the accident analyses described in Chapter 13 .0. As an AEC, the condensate tanks will use a continuous 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 (I) detect an anomaly in the evaporator or concentrator indicating high uranium content in the condenser (due to flooding or excessive foaming) , and (2) prevent high concentration uranium solution from being available in the condensate tank for discharged to a non-favorable geometry system or in the condenser for leaking to the non-safe geometry cooling loop.

Accident Mitigated The safety function of this IROFS is to prevent an accidental nuclear criticality because of excessive uranium in the condensate carryover to a non-geometrically favorable waste collection tank.

System Components System components consist of:

Condensate sample tank 1A (UR-TK-340)

Condensate delay tank 1 (UR-TK-360)

Condensate sample tank 1B (UR-TK-370)

Condensate sample tank 2A (UR-TK-540)

  • Condensate delay tank 2 (UR-TK-560)

Condensate sample tank 2B (UR-TK-570)

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0 NORTHWEn MEOK:Al tSOTOPEI NWMl-2013-021, Rev. 1 Chapter 6.0 - Engineered Safety Features Functional Requirements The detection system works 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 device will close an isolation valve in the inlet to the evaporator (or otherwise secures 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 designed to fail-closed on loss of instrument air, and the solenoid is designed to fail-closed on loss of signal. Locations where these IROFS are used will be determined during final design.

Design Basis Design basis information will be provided in the Operating License Application .

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.15 IROFS CS-26, Processing Component Safe Volume Confinement IROFS CS-26, "Processing Component Safe Volume Confinement," is identified by the accident analyses described in Chapter 13 .0 (see description in Section 6.3.1.2.2).

Accident Mitigated See description in Section 6.3.1 .2.2 .

System Components 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 the fissile solutions. Components that may be controlled to a safe volume will be described in the Operating License Application.

Functional Requirements The safety function of a 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 will conservatively include 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 will be compatible with the chemical nature of the contained solution.

Design Basis The safe-volume confinement components will be determined in final design after finalizing the referenced CSEs.

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' ~ *.* ~ ' NORTMWUT 110HCAL lSOTOPU Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements wi ll be developed for the Operating License Application.

6.3.1.2.16 IROFS CS-27, Closed Heating or Cooling Loop with Monitoring and Alarm IROFS CS-27, "Closed Heating or Cooling Loop with Monitoring and Alarm," is identified by the accident analyses in Chapter 13 .0. As a PEC, closed cooling water loops with monitoring for breakthrough of process solution will be provided on the evaporator or concentrator condensers to contain process solution that leaks across this boundary, if the boundary fails. This IROFS will be 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. The inherent characteristics of the leak path will reduce back-leakage into the closed loop system, and the risk of product solutions entering the condenser will be very low by evaporator and concentrator design.

System Components The purpose of this safety function is to monitor the health of the condenser cooling jacket to ensure that in the unlikely event that a condenser overflow occurs, fissile and/or high-dose process solution will not flow into this non-safe-geometry cooling loop and cause nuclear criticality. The closed loop will also isolate any high-dose fis sile 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.

Functional Requirements 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, or uranium concentration) 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 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 prior to discharge to a non-safe geometry.

Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements wi ll be developed for the Operating License Application.

6.3.2 Surveillance Requirements A review of surveillance requirements to ensure the availability and reliability of safety controls when required to perform safety functions will be included in the Operating License Application.

6.3.3 Technical Specifications The technical specifications will be provided in the Operating License Application.

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6.4 REFERENCES

I 0 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended.

I 0 CFR 20.1201, "Occupational Dose Limits for Adults," Code of Federal Regulations, Office of the Federal Register, as amended.

IO CFR 20.1301, "Dose Limits for Individual Members of the Public," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 50.59, "Changes, Tests, and Experiments," Code of Federal Regulations , Office of the Federal Register, as amended.

IO CFR 70.61, "Performance Requirements," Code of Federal Regulations, Office of the Federal Register, as amended.

ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Material Outside of Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2014.

ANSI/ ANS-8.3, Criticality Accident Alarm System, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, I 997 (Reaffirmed in 2012).

ANSI/ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materia ls, American National Standards Institute/American Nuclear Society, La Grange Park, Illinois, 1998 (Reaffirmed in 2007).

ANSI/ANS-8.10, Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement, American National Standards Institute/American Nuclear Society, La Grange Park, Illinois, 2015.

ANSI/ANS-8.19, Administrative Practices for Nuclear Criticality Safety, American National Standards Institute/American Nuclear Society, La Grange Park, Illinois, 2014.

ANSI/ ANS-8.20, Nuclear Criticality Safety Training, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 1991 (Reaffirmed in 2005).

ANSI/ANS-8.22, Nuclear Criticality Safety Based on Limiting and Controlling Moderators , American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 1997 (Reaffirmed in 2011 ).

ANSI/ ANS-8.23, Nuclear Criticality Accident Emergency Planning and Response, American National Standards Institute/American Nuclear Society, La Grange Park, Illinois, 2007 (Reaffirmed in 2012).

ANSI/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 2007 (Reaffirmed in 2012).

ANSI/ANS-8.26, Criticality Safety Engineer Training and Qualification Program, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 2007 (Reaffirmed in 2012).

ANSI/ ANS-15 .1 , The Development of Technical Specifications for Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2013.

ANSI Nl 3.1, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities, American Nuclear Society, La Grange Park, Illinois, 2011.

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  • NOITifWEn MEIUCAL. 1$0TOP£S Chapter 6.0 - Engineered Safety Features ASME AG-1, Code on Nuclear Air and Gas Treatment, American Society of Mechanical Engineers, New York, New York, 2003.

LA-CP-13-00634, MCNP6 User Manual, Rev. 0, Los Alamos National Laboratory, Los Alamos, New Mexico, May 2013.

NRC, 2012, Final Interim Staff Guidance Augmenting NUREG-1537, "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 Number:

NRC-201 1-0135, U.S. Nuclear Regulatory Commission, Washington, D.C. , October 30, 20 12.

NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility, Rev. 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C. , May 2010.

NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors - Format and Content, Part 1, U.S . Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February 1996.

NUREG/CR-4604 I PNL-5849, Statistical Methods for Nuclear Material Management, Pacific Northwest Laboratory, Richland, Washington, December, 1988.

NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology ,

U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C. , January 2001.

[Proprietary Information]

[Proprietary Information]

NWMI-2015-SDD-013, System Design Description for Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015 -CRITCALC-001, Single Parameter Subcritical Limits for 20 wt% 235 U - Uranium Metal, Uranium Oxide, and Homogenous Water Mixtures , Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015-CRITCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution , Rev. A Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015 -CRITCALC-003, 55-Gallon Drum Arrays, Rev. A Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015 -CRITCALC-005 , Target Fabrication Tanks, Wet Processes, and Storage , Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015-CRITCALC-006, Tank Hot Cell, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015 -CSE-001, NWMI Preliminary Criticality Safety Evaluation: Irradiated Target Handling and Disassembly, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015 -CSE-002, NWMI Preliminary Criticality Safety Evaluation: Irradiated Low-Enriched Uranium Target Material Dissolution, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 20 15.

NWMI-2015 -CSE-003, NWMI Preliminary Criticality Safety Evaluation: Mo ly bdenum-99 Recovery, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 20 15.

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. ' ~ *.*.! . ~THWEST MEOICAl. ISOTOPH NWMI-2015 -CSE-004, N WMI Preliminary Criticality Safety Evaluation: Low-Enriched Uranium Target Material Production, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015-CSE-005 , NWMI Preliminary Criticality Safety Evaluation: Target Fabrication Uranium Solution Processes, Rev. A, Northwest Medical Isotopes, LLC, Corvall is, Oregon, 2015 .

NWMI-2015-CSE-006, NWMI Preliminary Criticality Safety Evaluation: Target Finishing, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015-CSE-007, N WMI Preliminary Criticality Safety Evaluation: Target and Can Storage and Carts, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-20 15-CSE-008, N WMI Preliminary Criticality Saf ety Evaluation: Hot Cell Uranium Purification , Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-20 15-CSE-009, N WMI Preliminary Criticality Saf ety Evaluation: Liquid Waste Processing, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015 -CSE-010, NWMI Preliminary Criticality Safety Evaluation: Solid Waste Collection, Encapsulation, and Staging, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015 -CSE-Ol I , NWMI Preliminary Criticality Saf ety Evaluation: Offg as and Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 20 15.

NWMI-20 15-CSE-O 12, NWMI Preliminary Criticality Safety Evaluation: Target Transport Cask or Drum Handling, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 20 15.

NWMI-20 15-CSE-Ol 3, N WMI Preliminary Criticality Safety Evaluation: Analytical Laboratory, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 20 15.

Regulatory Guide 3.71 , Nuclear Criticality Safety Standards for Fuels and Material Facilities, Rev. 2, U.S . Nuclear Regulatory Commission, Washington, D.C., December 2010.

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  • _- . NORTHWEST MEDICAL ISOTOPES Chapter 7.0 - Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 June 2017 Prepared by:

Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, OR 97330

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~ *. NOITHWlST MEmtAl ISOTOP£S Chapter 7.0 - Instrumentation and Control Systems Chapter 7.0 - Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1

Title:

Chapter 7.0 - Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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  • .* . NWMI NOR11fWEST MEDtcAl lSOTOPH NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not required 1 6/26/2017 Incorporate changes based on responses to C. Haass NRC Requests for Additional Information

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...... Chapter 7.0 - Instrumentation and Control Systems NWMl-2013-021, Rev. 1

  • ~ ~-~~ : , NORTHWESTMHHCAllSOTOPES CONTENTS 7.0 INSTRUMENTATION AND CONTROL SYSTEMS ....... .... ........ ................. ..... ........ ......... ....... 7-1 7.1 Summary Description .............. .............. .. .. ..................... ............... ..... ... ................. ... ........ . 7-1 7 .2 Design of Instrumentation and Control Systems ............ ... ........... ..... .. ... ..... ............ .... ....... 7-4 7.2.1 Design Criteria ..... .............................. ....... .. ... ...... ..... ........ ... ...... .... .............. ..... .. . 7-4 7.2.2 Design Basis and Safety Requirements ..... ...... ...... ............ .......... ....... .. ... ..... ... ..... 7-4 7.2.3 System Description .. ... ......... .. ....... ... ................ ... ................ ..... ........... ....... ........ 7-1 3 7.2.3 .1 Facility Process Control System ...... .. .... ..... .... ............ ....... ...... ... ...... 7-14 7.2.3.2 Engineered Safety Feature Actuation Systems .. ..... ... ....................... 7-14 7.2.3 .3 Control Room/Human-Machine Interface Description ............. ........ 7-14 7.2.3.4 Building Management System ........... .... ....... ...... ..... .. ... .. .................. 7-15 7.2.3.5 Fire Protection System ........... ....... ..... .................... ..... ..... ............ ..... 7-1 5 7.2.3.6 Facility Communication Systems ............... ............. ...... ............ .. .... .. 7-15 7.2.3. 7 Analytical Laboratory System ....... ................ ........ ..... ........ .......... ..... 7-16 7.2.4 System Performance Analysis .. .... .. ... ....... .... ... ....... ..... ...... .. .. ... ... ................. .. .. .. 7-16 7.2.4.1 Facility Trip and Alarm Design Basis ... ..... .. ...... ..... .... ...................... 7-16 7.2.4.2 Anal ysis .... ................ ...... .... ............... .... ...... .......... ... ....... .................. 7-17 7.2.4.3 Conclusion ............................................... .. ......... ... .... ................... .... 7-1 8 7.3 Process Control Systems .... ...... .... ......................... ..... ......... .... .. .... .. .... ... .... ... .......... ....... .. 7-22 7.3.1 Uranium Recovery and Recycle System ..... .... .. .. ....... ....... .... ... ..... ........ .... ...... ... 7-22 7.3.1.1 Design Criteria ............ ........... ......... .... .... ......... ... ........................ ..... . 7-23 7.3 .1.2 Design Basis and Safety Requirements ....... ... ..... .. ......... .. ... .. .. .. .. ...... 7-23 7.3 .1.3 System Description ............. ............ ... .... .... ..... .. ... .... .... ......... ............ 7-23
7. 3. 1.4 System Performance Analysis and Conclusion .... ....... .... ....... ....... .... 7-28 7.3 .2 Target Fabrication System ........... .... ... ............. .. ...... .. ... .. ........ ....... .... ...... .......... 7-28 7.3.2.1 Design Criteria ............ .. .... .... ... .. .. ...... .. ... .. ............. .......... ...... .. ...... ... 7-29 7.3 .2.2 Design Basis and Safety Requirements ... .. ... .. .... .. ... ..... .. ... .. .... .......... 7-29 7.3.2.3 System Description ........... .......... ....... ...... ..... ... ......... ... ................. .... 7-29 7.3.2.4 System Performance Analysi s and Conclusion .............. .... ... ... .. .. .. ... 7-32 7.3.3 Target Receipt and Disassembly System .... ......... .. ... ...... .... ... ........ ...... ... .. ...... ... 7-3 2 7.3.3.1 Design Criteria ...... ........ ................. ... ..... ...... .... .. .. .. ...... ....... ...... .... .... 7-32 7.3.3.2 Design Basis and Safety Requirements ........... ........ ....... .. ..... ............ 7-32 7.3.3.3 System Description ...... ... ......... ....... ... ... .... ..... ... ................ .... .. ... ..... .. 7-32 7.3.3.4 System Performance Analysis and Conclusion ..... ..... ..... .... .............. 7-33 7.3.4 Target Dissolution System ........ .... ........ ............... .... ............. .... .. ..... ..... ...... .... ... 7-33 7.3.4.1 Design Criteria ..................... ........ ..... ....... ..... ... ........ ... ..... ................. 7-34
7. 3.4.2 Design Basis and Safety Requirements .... ....... ... .. .... .... ... ....... ........... 7-34 7.3.4.3 System Description ...... ...................... ..... .. .. .... .......... ..... ...... .... ...... ... 7-34 7.3 .4.4 System Performance Analysis and Conclusion ... ......... .................. .. . 7-37 7.3 .5 Molybdenum Recovery and Purification System .. .... .... ...... ... ... ... .... ..... ... ...... .... 7-37 7.3.5.1 Design Criteria .... .............. ... .... ..... ... ........ .......... ........ ................. ...... 7-37 7.3.5.2 Design Basis and Safety Requirements ........... ... .... ..... .... ..... ........ .... . 7-37 7.3.5.3 System Description ............ ........ ....... ..... ... ..... ....... ... .. ..... ............ ...... 7-38 7.3 .5.4 System Performance Analysi s and Conclusion ........ .. ..... ........... ... ... . 7-39 7-i

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  • NOllTHWEST MEDICAL ISOTOPH FIGURES Figure 7-1. Radioisotope Production Facility Instrumentation and Control System Configuration ............. .... .... ....... ...... .. ..... .............. ... ....... .. ... .... ..... ..... ............ ... .. .............. 7-2 TABLES Table 7-1. Instrumentation and Control System Design Criteria (10 pages) ..... ....... ............. ...... .... . 7-5 Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (5 pages) ..... ...... .... ........... .... ... .......... ... .. ..... .. ..... ........ ................ ... 7-18 Table 7-3. Uranium Recovery and Recycle Control and Monitoring Parameters (2 pages) ......... .. 7-24 Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages) ........ ............. ....... ................. ................. ... ..... ..... ... .... .......... ........ ..... ................ 7-25 Table 7-5 . Target Fabrication System Control and Monitoring Parameters (2 pages) ... ..... .. ........ .. 7-29 Table 7-6 . Target Fabrication System Interlocks and Permissive Signals (2 pages) ...... .. ........ ...... 7-30 Table 7-7. Target Dissolution System Control and Monitoring Parameters .... .. ..... .................... .. .. 7-35 Table 7-8. Target Dissolution System Interlocks and Permissive Signals (2 pages) ...................... 7-36 Table 7-9. Molybdenum Recovery and Purification System Control and Monitoring Parameters .. .. ............. ................. ............. ..... .. .................. .. ... .......... ....... .... ... ... .............. 7-38 Table 7-10. Molybdenum Recovery and Purification System Interlocks and Permissive Signals .. .. ............ ........ .............. ...... ............. ......... ....... .. ..... .... .. ..... .............. .. ..... ............ 7-38 Table 7-11. Waste Handling System Control and Monitoring Parameters .... ....... ........... ..... ...... ...... 7-41 Table 7-12. Waste Handling System Interlocks and Permissive Signals ... .......... ..... .... ............ .. ...... 7-42 Table 7-13 . Engineered Safety Feature Actuation or Monitoring Systems (2 pages) ....................... 7-44 7-iii

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!*.*! 0 NOmfWEST ME.DM:Al ISOTOf'lS NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 ADUN acid-deficient uranyl nitrate ALARA as low as reasonably achievable BMS building management system CAAS criticality accident alarm system CAM continuous air monitor CFR Code of Federal Regulations CGD commercial grade dedication COTS commercial off-the-shelf DCS digital control system ESF engineered safety feature FPC facility process control HMI human-machine interface I iodine I&C instrumentation and control IEEE Institute of Electrical and Electronics Engineers IROFS items relied on for safety ISA integrated safety analysis IX ion exchange Kr krypton LEU low-enriched uranium Mo molybdenum NAVLAP National Voluntary Laboratory Accreditation NOx nitrogen oxide NRC U .S. Nuclear Regulatory Commission NWMI Northwest Medical Isotopes, LLC PLC programmable logic controller RAM radiation area monitor RPF Radioisotope Production Facility SDOE secure development and operational environment SIF safety instrumented function.

SIL safety integrity level.

SIS safety instrumented system SNM special nuclear material SSC structures, systems, and components TCE trichloroethylene U.S. United States

[Proprietary Information]

UPS uninterruptible power supply V&V verification and validation Xe xenon Units m meter mm minute rad radiation absorbed dose 7-iv

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' ! * *~ . NORTHWEST MUMCAl lSOTOPU Chapter 7.0 - Instrumentation and Control Systems 7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.1

SUMMARY

DESCRIPTION The Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF) preliminary instrumentation and control (I&C) configuration includes the special nuclear material (SNM) preparation and handling processes (e.g., target fabrication , and uranium recovery and recycle), radioisotope extraction and purification processes (e.g., target receipt and disassembly, target dissolution, molybdenum

[Mo] recovery and purification, and waste handling), process utility systems, criticality accident alarm system (CAAS), and systems associated with radiation monitoring.

The SNM processes will be enclosed predominately by hot cells and glovebox designs except for the target fabrication area. The facility process control (FPC) system will provide monitoring and control of the process systems within the RPF. In addition, the FPC system will provide monitoring of safety-related components within the RPF. The process strategy for the RPF involves the use of batch or semi-batch processes with relatively simple control steps.

The building management system (BMS) (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (turn on and off) the mechanical utility systems.

Engineered safety feature (ESF) systems will operate on actuation of an alarm setpoint reached for a specific monitoring instrument/device. For redundancy, this will be in addition to the FPC system or BMS ability to actuate ESF as needed. Each ESF safety function will use hard-wired analog controls/interlocks to protect workers, the public, and environment. The ESF parameters and alarm functions will be integrated into and monitored by the FPC system or BMS.

The preliminary concept for the RPF l&C system configuration is shown in Figure 7-1. The green circles identify the FPC and the BMS distributed process control or programmable logic controller (PLC) systems. The solid lines and dashed lines show how the SNM processes, support systems, utilities, radiation and criticality systems, and building functions relate to the FPC and BMS and to local human-machine interface (HMI) stations. Solid lines indicate the control functions , and dashed lines indicate the monitoring functions .

The FPC system will perform as the overall production process controller. This system will monitor and control the process instrumented functions within the RPF, including monitoring of process fluid transfers and controlled inter-equipment pump transfers of process fluids. Process control systems are described further in Section 7.3.

The fire protection system will have its own central alarm panel (green circle). The fire protection system will report the status of the fire protection equipment to the central alarm station and the RPF control room. The fire protection system is discussed further in Section 7.2.3.5 .

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Process Systems In Target Fabrication Area Process Utility Systems Figure 7-1. Radioisotope Production Facility Instrumentation and Control System Configuration Special nuclear material preparation and handling processes - The FPC system will control and/or monitor the SNM preparation and handling processes, the following.

  • Target fabrication - Batch processes located in the target fabrication area will be controlled by operators at local HMls, with surveillance monitoring in the control room.
  • Uranium recovery and recycle - Batch processes located inside the hot cell area will be controlled by operators in the control room.

Radioisotope extraction and purification processes - The FPC system will control and/or monitor the radioisotope production processes, including the following.

  • Target receipt and disassembly - Hardware/target movement located in irradiated target basket receipt bay area, target cask preparation airlock, target receipt hot cell, and target disassembly hot cell will normally be controlled by operators at local HMis, with surveillance monitoring in the control room.
  • Target dissolution - Batch process located inside the dissolution hot cell will occur at local HMis in the operating gallery, and offgas operations in the tank hot cell will be controlled by operators in the control room, with surveillance monitoring at both locations.

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  • Mo recovery and purification - Batch processes located inside the Mo hot cells will be controlled by operators at a local HMI in the operating gallery, with surveillance monitoring in the control room.
  • Waste handling - This system includes liquid waste handling, liquid waste solidification, and solid waste handling. Operators in the control room will control liquid waste handling, while operators at local HMls in the low-dose liquid solidification room (Wl07) will monitor and control liquid waste solidification, and solid waste nondestructive examination and solidification.

Process utility and support systems - The FPC system will control and monitor the process utility and process support systems. Operators in the control room will control the following subsystems:

  • Process chilled water hot cell secondary loops
  • Process steam hot cell secondary loops
  • Process vessel ventilation system Operators at local HMis will control the following subsystems, with surveillance monitoring in the control room using the FPC system or BMS.
  • Plant air system
  • Gas supply system
  • Process chilled water chillers
  • Process steam boilers
  • Demineralized water system
  • Chemical supply system
  • Standby electrical power system Criticality accident alarm system - The CAAS will be provided as an integrated vendor package. The detectors and alarm response are integral to the individual units/locations. The FPC system will monitor the CAAS status in the control room. The CAAS is described further in Section 7.3.

Radiation monitoring system - The FPC system will monitor the various radiation monitoring systems, including continuous air monitors (CAM), air samplers, radiation area monitors (RAM), and exhaust stack monitors. The CAMs and RAMs will be strategically placed throughout the RPF to alert personnel of any potential radiation hazards. The CAMs and RAMs will alarm in the control room and locally at locations throughout the RPF. The radiation monitoring systems are described further in Section 7.6.

Facility ventilation system and mechanical utility systems - The control function for most of the RPF ventilation system and mechanical utility systems will be local HMls and hard-wired interlocks for the ESF function s. The BMS will monitor the systems and provide ventilation and mechanical utility system status as an input to the FPC process controls.

The following subsystems will be monitored by the BMS:

  • Facility ventilation Zones I, II, III, and IV
  • Supply air system
  • Facility chilled water system
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  • NOflTHWUT MlDtCAL asGTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Safety-Related Components and Engineering Safety Features The ESF safety functions will operate independently from the FPC systems as hard-wired analog controls or interlocks. The FPC system will be a digital control system (DCS) that monitors safety-related components within the RPF. The ESFs will be integrated into the FPC systems and provide a common point of HMI, monitoring, and alarming at the control room and, as necessary, local HMI workstations.

Control Console and Display Instruments The control room will be the primary interface location for the RPF support systems and provide centralized process controls, monitoring, alarms, and acknowledgement. Mechanical utility systems with vendor packages and integrated controls will be controlled at associated local HMis. The BMS will provide primarily on/off control and system monitoring from the control room.

The tank hot cell processes will be controlled primarily in the control room, with surveillance monitoring of the FPC subsystems. The FPC system will have annunciation, alarms, and HMI displays. From the consoles, operators will view and trend essential measurement values from the HMI display, and evaluate real-time data from the essential measurements used to control and monitor the RPF process. This system is further described in Section 7.5.

Process utility and support systems with vendor package and integrated controls will be operated at associated local HMis. These systems are discussed further in Section 7.5. Local HMis are anticipated in the following locations:

  • Irradiated target basket receipt bay A/B (Rl 02A/B)
  • Cask preparation airlock (R012)
  • Operating gallery (GlOl A/B/C)
  • Target fabrication (Tl 04 A/B)
  • Low-dose liquid waste solidification (Wl 07)
  • Chemical supply room (Ll02)
  • Local to equipment with integrated control systems 7.2 DESIGN OF INSTRUMENTATION AND CONTROL SYSTEMS The design criteria and the codes and standards for I&C systems are outlined in Chapter 3.0, "Design of Structures, Systems, and Components," and discussed below.

7.2.1 Design Criteria The applicable design criteria and guidelines that apply to the RPF I&C systems are summarized in column one of Table 7-1. Additional, design criteria for I&C systems are provided in Chapter 3.0. The detailed and specific design criteria for I&C systems will be confirmed in the Operating License Application.

7.2.2 Design Basis and Safety Requirements The design basis for I&C systems used in the RPF are presented in the second column of Table 7-1. The second column maps the criteria to l&C systems or components and how compliance will be ensured.

Note that the FPC system callouts may also apply to the BMS. The design basis requirements for facility and process systems are described in Chapter 4.0, "Radioisotope Production Facility Description," and Chapter 9.0, "Auxiliary Systems."

The I&C system will use hard-wired interlocks for actuated engineered safety functions . Section 7.4 summarizes the I&C ESFs.

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Design criteria descriptiona Design bases as applied to RPF IEEE 379-2014, IEEE Standard Application of the Application:

Single-Failure Criterion to Nuclear Power Generating

  • Design ofFPC system, ESFs, and other Station Safety Systems instrumentation SSCs that are identified as IROFS Description : Application of the single-failure criterion Compliance:

to electrical power, instrumentation, and control portions

  • Ensure FPC system is a DCS designed, rated, and of nuclear power generating safety systems. approved for use in safety instrumented systems, as Keywords: Actuator, cascaded failu re, common -cause determined by ANSI/ISA 84.00.01 fai lure, design basis event, detectable failure , effects
  • Use a safety PLC, as recognized by IEC 61508 , in the analysis, safety system, single-failure criterion, system FPC system with redundant power supplies, actuation, system logic processors, and input/output channels
  • Evaluate controls that are classified as IROFS in Chapters 6.0 and 13.0, or NWMI-20 l 5-SAFETY-002, against single-failure criteria Exception :
  • NUREG-1537 allows for sharing and combining of systems and components with justification
  • The RPF is not considered a nuclear power reactor but a production faci lity. The facility will not have all of the systems detailed in this standard and guidance will be app lied as appropriate.

IEEE 577-2012, IEEE Standard Requirements for Application:

Reliability Analysis in the Design and Operation of

  • Use for design ofFPC system, ESFs, and other Safety Systems for Nuclear Facilities instrumentation SSCs that are identified as IROFS

Description:

Sets minimum acceptable requirements for Compliance:

the performance of reliability analyses for safety

  • Perform a reliability analysis of the proposed design systems when used to address the reliability solution for IROFS functions, as identified in considerations discussed in industry standards and Chapters 6.0 and 13.0, or NWMI-2015-SAFETY-002.

guidelines. The requirement that a reliability analysis be The analysis can be qualitative or quantitative in performed does not originate with this standard. nature, as described in the standard However, when reliability analysis is used to demonstrate compliance with reliability requirements, this standard describes an acceptable response to the requirements.

Keywords: Nuclear facilities, reliability analysis, safety systems 7-5

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Design criteria description 3 Design bases as applied to RPF IEEE 603-2009, IEEE Standard Criteria for Safety Application:

Systems for Nuclear Power Generating Stations

  • Use for design ofFPC system, ESFs, and other

Description:

Establi shes minimum functional and instrumentation SSCs that are identified as IROFS design criteria for the power, instrumentation, and

  • Apply minimum functional and design criteria to control portions of nuclear power generating station safety systems safety systems. Criteria are to be applied to those Compliance:

systems required to protect public health and safety by

  • Ensure design conforms to the practices detailed in functioning to mitigate the consequences of design basis the standard for the IROFS functions identified in events. The intent is to promote appropriate practices Chapters 6.0 and 13.0, orNWMI-2015-SAFETY-002 for design and evaluation of safety system performance Exception:

and reliability. The standard is limited to safety systems; many of the principles may have applicabi lity

  • The RPF is not considered a nuclear power reactor to equipment provided for safe shutdown, post-accident but a production facility . The fac ility will not have monitoring display instrumentation, preventive interlock all of the systems detailed in this standard and features, or any other systems, structures, or equipment guidance will be appli ed as appropriate.

related to safety.

Keywords: Actuated equipment, associated circuits, Class IE, design, failure, maintenance bypass, operating bypass, safety functio n, sense and command features, sensor IEEE 384-2008, IEEE Standard Criteria for Application:

Independence of Class IE Equipment and Circuits

  • Use for design of FPC system, ESFs, and other

Description:

Describes independence requirements of instrumentation SSCs that are identified as IROFS circuits and equipment comprising or associated with

  • Apply minimum criteria for separation and Class IE systems. Identifies criteria for independence independence of systems in a physical way that can be achieved by physical separation, and Compliance:

electrical isolation of circuits and equipment that are

  • Ensure design conforms to the practices detailed in redundant. The determination of what is to be the standard for the IROFS functions identified in considered redundant is not addressed. Chapters 6.0 and 13.0, orNWMI-2015-SAFETY-002 Keywords: Associated circuit, barrier, Class IE, Exception:

independence, isolation, isolation device, raceway,

  • The RPF is not considered a nuclear power reactor separation but a production facility. The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

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Design criteria descriptiona Design bases as applied to RPF IEEE 323-2003, IEEE Standard for Qualifying Class Application:

JE Equipment for Nuclear Power Generating Stations

  • Use fo r equipment quali fication when needed to

Description:

Identifi es requirements fo r qualifying quali fy equipment for appl ications or environments to Class IE equipment and interfaces that are to be used in whi ch the equipment may be exposed nuclear power generating stati ons. The principl es,

  • Use for qualification of Class IE equipment located methods, and procedures are intended for use in in harsh environments and fo r certain post-accident qualifying equipment, maintaining and extending monitoring equipment; may also be used fo r the qualification, and updating qualificati on, as required, if qualifi cation of equipment in mild environments the equipment is modified. The qualification Compliance:

requirements of the standard demonstrate and document

  • Ensure design conforms to the practices detailed in the ability of equipment to perform safety function(s) the standard fo r those systems determined to be under applicable service conditi ons, including design Class IE and located in harsh environments for safety basis events, reducing the ri sk of common-cause functions identifi ed in Chapters 6. 0 and 13, or equipment fai lure. NWMI-20 J 5-SAFETY-002 Keywords : Age conditioning, aging, conditi on
  • Appl y to SSCs within the hot cell area; not all safety monitoring, design basis event, equipment qualification, components reside in the hot cell area qualification methods, harsh environment, margin, mild
  • Apply standard using a graded approach environment, qualified li fe, radiati on, safety-related Exception:

function, signi ficant aging mechani sm, test plan, test

  • The RPF is not considered a nucl ear power reactor sequence, type testing but a production facility. The facility will not have all of the systems detailed in thi s standard and guidance wi ll be appli ed as appropri ate.

IEEE 344-2004, IEEE Recommended Practice for Application:

Seismic Qualification of Class JE Equipment/or

  • Apply seismic design requirements for equipment Nuclear Power Generating Stations used in Class 1E systems

Description:

Identifies recommended practices for Compliance:

establishing procedures that will yield data to

  • Use in design ofFPC system, ESFs, and other demonstrate that the Class IE equipment can meet instrumentation SSCs that are identified as a Class IE performance requirements during and/or following one system safe shutdown earthquake event, preceded by a number Exception:

of operating basis earthquake events. This recommended practice may be used to establish tests,

  • The RPF is not considered a nuclear power reactor analyses, or experience-based evaluations that will yield but a production facility. The facility will not have data to demonstrate Class 1E equipment performance all of the systems detailed in this standard and claims or to evaluate and verify performance of devices guidance will be applied as appropriate.

and assemblies as part of an overall qualification effort.

Common methods currently in use for seismic qualification by test are presented. Two approaches to seismic analysis are described: one based on dynamic analysis, and the other on static coefficient analysis.

Two approaches to experience-based seismic evaluation are described, one based on earthquake experience and the other on test experience.

Keywords: Class 1E, earthquake, earthquake experience, equipment qualification, inclusion rules, nuclear, operating basis earthquake, prohibited features, qualification methods, required response spectrum, response spectra, safe shutdown earthquake, safety function, seismic, seismic analysis, test response spectrum, test experience 7-7

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Design criteria descriptions Design bases as applied to RPF IEEE 338-2012, IEEE Standard for Criteria for the Application:

Periodic Surveillance Testing of Nuclear Power

  • Use fo r design ofFPC system, ESFs, and other Generating Station Safety Systems instrumentation SSCs that are identifi ed as IROFS Description : Provi des cri teria fo r the performa nce of
  • Use methods and criteria to establish a periodic periodic surveillance testing of nuclear power generating surveillance program station safety systems. The scope of periodic Compliance:

surveillance testing consists of functio nal tests and

  • Ensure design conforms to the practices detailed in checks, calibration verification, and time response the standard for the IROFS functions identified in measurements, as required, to verify that the safety Chapters 6.0 and 13.0, or NWMI-20 l 5-SAFETY-002 system perfo rms its defined safety fu nction. Post-Exception:

maintenance and post-modification testing are not covered by thi s document. Thi s standard ampli fies the

  • The RPF is not considered a nuclear power reactor periodic surveillance testing requirements of other but a production facility. The facility will not have nuclear safety-related IEEE standards. all of the systems detailed in thi s standard and gu idance will be applied as appropriate .

Keywords: Functional tests, IEEE 338, periodic testing, risk-info rmed testi ng, survei ll ance testing IEEE 497-2010, IEEE Standard Criteria for Accident Application:

Monitoring Instrumentation for Nuclear Power

  • Use as selection, design, performance, qualification, Generating Stations and display criteria for accident monitoring

Description:

Establishes criteria for variable selection, instrumentation performance, design, and qualification of accident

  • Apply guidance on the use of portable monitoring instrumentation, and includes the instrumentation and for examples of accident requirements for display alternatives for accident monitoring display configurations monitoring instrumentation, documentation of design Compliance:

bases, and use of portable instrumentation.

  • Ensure design conforms to standard for the Keywords: Accident monitoring, display criteria, monitoring functions determined to be required for selection criteria, type variables health and safety of workers or the public during normal operation and design basis accidents Exception:
  • The RPF is not considered a nuclear power reactor but a production facility. The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

IEEE 7-4.3.2-2010, IEEE Standard Criteria for Digital Application:

Computers in Safety Systems of Nuclear Power

  • In conjunction with IEEE 603-2009, use to establish Generating Stations minimum functional and design requirements fo r Abstract: Specifies addi tional computer-specific computers that are components of a safety system requirements to supplement IEEE 603 -2009. The
  • Design FPC system as a DCS, and appl y this standard standard defi nes the term computer as a system that to system development, specifically software includes computer hardware, software, firmware , and deve lopment interfaces, and establi shes minimum functional and
  • Appl y standard to CGD and implement an approach design requirements fo r computers used as components Compliance:

of a safety system.

  • Develop FPC system software usi ng this standard Keywords: Commercial-grade item, diversity, safety Exception:

systems, software, software tools, software verificati on

  • The RPF is not considered a nuclear power reactor and validation but a production facility. The faci lity wi ll not have all of the systems detai led in this standard and guidance wi ll be appl ied as appropriate.

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Design criteria description 3 Design bases as applied to RPF IEEE 828-2012, IEEE Standard/or Configuration Application:

Management in Systems and Software Engineering

  • Use to establish configuration management processes,

Description:

Establishes minimum requirements for define how configuration management is to be configuration management in systems and software accomplished, and identify who is responsible for engineering. This standard applies to any form, class, or performing specific activities, when the activities are type of software or system, and explains configuration to happen, and what specifi c resources are required management, including identifying and acquiring

  • Design FPC system as a DCS, and apply standard configuration items, controlling changes, reporting the during the development of software for systems with status of configuration items, and performing software IROFS functions builds and release engineering. This standard addresses Compliance:

what configuration management activities are to be

  • Develop FPC system software using this standard for done, when they are to happen in the life-cycle, and safety function implementation what planning and resources are required. The content areas for a configuration management plan are also identified. The standard supports IEEE STD 12207 and ISO/IEC/IEEE 15288, and adheres to the terminology in ISO/IEC/IEEE STD 24765 and the information item requirements of IEEE STD 15939.

Keywords: Change control, configuration accounting, configuration audit, configuration item, IEEE 828, release engineering, software builds, software configuration management, system configuration management IEEE 1028-2008, IEEE Standard/or Software Reviews Application:

and Audits

  • Use to identify minimum acceptable requirements fo r

Description:

Identifies fi ve types of soft ware reviews systematic software reviews and audits, together with procedures required fo r the

  • Identi fy organizational means for conducting a review executi on of each type. Thi s standard is concerned only and documenting the findi ngs with reviews and audits; procedures fo r determining the
  • Design FPC system as a DCS, and apply standard necessity of a review or audit are not defi ned, and the during the deve lopment of software fo r systems with dispositi on of the results of the review or audit is not IROFS functions specified. Types included are management reviews, Compliance:

technical reviews, inspections, walk-throughs, and

  • Develop FPC system using thi s standard audits.

Keywords: Audit, inspection, review, walk-through ANS 10.4-2008, Verification and Validation of Non- Application:

Safety-Related Scientific and Engineering Computer

  • Perform software V & V to build quality into the Programs for the Nuclear Industry software during the software life-cycle

Description:

Provides guidelines for V& V of non-

  • Use to verify and validate software development for safety- related scienti fie and engineering computer non-safety-related systems programs developed for use by the nuclear industry.
  • Use for software development in the RPF that is not Scope is restricted to research and other non-safety- safety significant (e.g., not safety-related or IROFS) related, noncritical applications. Compliance:

Keywords: Software integrity level, software life-cycle,

  • Develop non-safety-related software using this validation, verification, V & V standard 7-9

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..*. * *. NCMl:THWUT MlD"-'l tsOTDPES Chapter 7.0 - Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages)

Design criteria descriptiona Design bases as applied to RPF ANSI/ISA 67.04.01-2006, Setpoints for N uclear Application:

Safety-Related Instrumentation

  • Use methods and criteria to establi sh setpoints fo r Description : Defines requirements fo r assessing, safety systems and to maintain the documentation establishing, and maintaining nuclear safety-related and
  • Apply to the design of the FPC system and other other important instrument setpoints associated with instrumentation SSCs that are identified as IROF S for nuclear power plants or nuclear reactor faci lities. the RPF Keywords: Setpoint, drift, analog channel, reliability Compliance:

analysis

  • Ensure design conforms to the practices detail ed in the standard fo r IROFS functions with inherent setpoints identified in Chapters 6.0 and 13 .0, or NWMI-2 0 l 5-SAFETY-002 ANSI/ISA 84.00.01-2004, Functional Safety: Safety Application:

Instrumented Systems for the Process Industry Sector

  • Apply to the design of safety systems (standard Part 1: Framework, Definitions, System, Hardware specifically designed for industrial processes) and Software Requirements"
  • Standard is made up of three parts:

Part 2: "Guidelines for the Application of ANSI/ISA- - Use Part I to lay the groundwork for the safety 84.00.01-2004 Part 1 (IEC 61511-1 Mod)- system life-cycle, overall structure of safety Informative" systems, definitions used, and to implement safety Part 3: "Guidance for the Determination of the system design engineering Required Safety Integrity Levels - Informative" - Use Part 2 guidance for the specification, design,

Description:

Provides requirements for the installation, operation, and maintenance of safety specification, design, installation, operation, and instrumented functions and related safety maintenance of a safety instrumented system, so the instrumented systems, as defined in Part 1 system can be confidently entrusted to place and/or - Use Part 3 to develop underlying concepts of risk maintain the process in a safe state. This standard has in relation to safety integrity, identify tolerable been developed as a process sector implementation of risk, and determine the safety integrity levels of the IEC 61508. safety functions Keywords: Safety instrumented system (SIS), safety

  • Design physical hardware of the FPC system based integrated level (SIL), safety instrumented function on this standard and IEC 61508 (SIF)
  • Evaluate the IROFS function s required to be implemented by the FPC system using Parts 1, 2, and 3 of this standard
  • Use to demonstrate reliability and risk reduction of the FPC system, while having similar or higher documented and tested ability to reduce risk as fulfillment through other channels Compliance:
  • Use for the design and implementation for IROFS functions that are required of the FPC system 7-10

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  • NOmfWlST MEDttAl lSOTO,H NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages)

Design criteria description 3 Design bases as applied to RPF NUREG-0700, Human-System Interface Design Application:

Review Guidelines

  • Use comprehensive design review guidance to

Description:

Provides guidance to the NRC on the develop information displayed in human-interface evaluation of human factors engineering aspects of systems nuclear power plants in accordance with NUREG-0800.

  • Develop informative and effective designs that will Detailed design review procedures are provided in assist operators in the performance of their duties NUREG-0711. As part of the review process, the Compliance:

interfaces between plant personnel and the plant systems

  • Design FPC system to provide information to and components are evaluated for conformance with operators in a display format human factors engineering guidelines.
  • Display development used in connection with the Keywords: Display, HMI , human-interface system, FPC system wi ll be provided in the Operating License human-system interface App li cation NUREG/CR-6463, Review Guidelines on Software Application:

Languages for Use in Nuclear Power Plant Safety

  • Use guidance to review high-integrity software in a Systems nuclear facility

Description:

Provides guidance to the NRC on auditing

  • Develop FPC system as a DCS, with associated programs for safety systems written in the following six programming development needs for the RPF high-level languages: Ada, C and C++, PLC Ladder
  • Use guideline as a means to review FPC system Logic, Sequential Function Charts, Pascal, and PL/M. programming code The guidance could also be used by those developing Compliance:

safety significant software as a basis for project-specific

  • Develop FPC system software programs using this programming guidelines. guidance Keywords: Pascal, C, Ladder Logic, PL/M, Ada, C++, Exception:

PLC, programming, sequential function charts

  • The RPF is not considered a nuclear power reactor but a production facility. The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

NUREG/CR-6090, The Programmable Logic Application:

Controller and Its Application in Nuclear Reactor

  • Use gu idance to implement PLCs for nuclear Systems application and as a forum for what constitutes good Abstract: Outlines recommendations for review of the practices of previously installed systems app lication of PLCs to the control, monitoring, and
  • Use guidance during selection process for hardware, protection of nuclear reactors. failure analysis, and product life-cycle within the Keywords: PLC, programming, protection systems faci lity Compliance:
  • Design FPC system to use a PLC-type DCS
  • Select design and implement PLCs based on this guide, as applicable Exception:
  • The RPF is not considered a nuclear power reactor but a production facility . The fac ility will not have all of the systems detailed in this standard and guidance will be app lied as appropriate.

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  • NOITHWEST MEDttAl tSOTOPU Table 7-1. Instrumentation and Control System Design Criteria (10 pages)

Design criteria descriptiona Design bases as applied to RPF EPRI TR-106439, Guideline on Evaluation and Application:

Acceptance of Commercial Grade Digital Equipment

  • Use to identify appropriate critical characteristics for Nuclear Safety Applications with subsequent verification through testing, analysis,

Description:

Provides a consistent, comprehensive vendor assessments, and careful review of operating approach for the evaluation and acceptance of expenence commercial digital equipment for nuclear safety systems.

  • Use guidance for digital upgrades to safety-related Keywords: Commercial off-the-shelf(COTS), systems and for non-safety-related applications that programming, software, commercial grade dedication require high reliability or are compatible with utility-specific change processes, including graded approaches for quality assurance Compliance:
  • Ensure that digital systems components that require CGD apply the guidance of this standard, as applicable Regulatory Guide 1.152, Criteria/or Use of Computers Application:

in Safety Systems of Nuclear Power Plants

  • Use fo r l&C system designs with computers in safety-

Description:

Describes a method that the NRC staff related systems that make extensive use of advanced deems acceptabl e for complying with NRC regulations technology fo r promoting hi gh functional reliabili ty, design quality,

  • Use fo r RPF designs (that are expected to be and a secure development and operati onal environment significantly and functionally different fro m current fo r the use of digital computers in the safety systems of day process designs) with microprocessors, di gital nuclear power plants. systems and di splays, fiber opti cs, multipl exing, and Keywords : Secure development and operational different isolati on techniques to achieve suffi cient environment (SDOE), computers independence and redundancy Compliance:
  • Develop FPC system and associated HMI using thi s guidance Exception:
  • The RPF is not considered a nuclear power reactor but a production facility. The fac ili ty will not have all of the systems detail ed in thi s standard and guidance will be appli ed as appropriate.

Regulatory Guide 1.53, Application of the Single- Application:

F ailure Criterion to Safety Systems

  • Apply single-failure criterion to safety-related I&C

Description:

Provides methods acceptable to the NRC systems staff for satisfying NRC regulations with respect to the

  • Apply to end-devices used by the FPC system that are application of the single-failure criterion to the electrical identified as IROFS power and I&C portions of nuclear power plant safety Compliance:

systems.

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Design criteria description 3 Design bases as applied to RPF Regulatory Guide 1.97, Criteria for Accident Application:

Monitoring Instrumentation for Nuclear Power Plants

  • Use this guidance for development of accident

Description:

Provides a method that the NRC staff monitoring for the RPF considers acceptable for use in complying with NRC Compliance:

regulations with respect to satisfying criteria for accident

  • Design FPC system, CAAS, CAMs, and RAMs using monitoring instrumentation in nuclear power pl ants. this guidance Keywords: IEEE 497-20 I0, accident monitoring Exception:
  • The RPF is not considered a nuclear power reactor but a production facility. The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

Regulatory Guide 5. 71, Cyber Security Programs for Application:

Nuclear Facilities

  • Use this guidance for development of cybersecurity

Description:

Provides an approach that the NRC staff protections deems acceptable for complying with NRC regulations Compliance:

regarding the protection of digital computers,

  • Design the FPC system and associated HMI based on communications systems, and networks from a this guidance cyberattack, as defined by I 0 CFR 73 . l.

Keywords: Cybersecurity, 10 CFR 73.54(a)(2), design basis threat a Full references provided in Section 7.7.

CAAS criticality accident alarm system. IROFS items relied on fo r safety.

CAM continuous air monitor. NRC U.S. Nuclear Regulatory Commission .

CFR Code of Federal Regulations. PLC programmab le logic controller.

CGD commercial grade dedication. RAM radiation alarm monitor.

COTS commercial off-the-shelf. RPF Radioisotope Product ion Facility.

DCS digital control system. SDOE secure development and operational ESF engineered safety feature . environment.

FPC fac ility process control. SIF safety instrumented fun ction .

HM! human-machine interface. SIL safety integrity level.

I&C instrumentation and control. SIS safety instrumented system.

IEEE Institute of Electrical and Electronics SSC structures, systems, and components.

Engineers . Y&Y verification and validation .

Specific requirements will be developed during the next stages of design for the Operating License Application. The I&C design wi ll be expanded and analyzed to document fulfillment of the design criteria and design basis requirements for the Operating License Application.

7.2.3 System Description As described in Section 7 .1 , the RPF I&C system basic components include the FPC system, ESF actuation systems, control console and HMI display instruments, and BMS. These systems provide an interface for the operator to monitor and control those systems. The FPC system will be a DCS that functions independently and electrically isolated from power systems. The items relied on for safety (IROFS)/ESF safety functions will be activated via hardwire interlocks.

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  • ~ * . ~ ~ : . NORTHWEST MEDUl ISOTOPH NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.2.3.1 Facility Process Control System The FPC system controls and monitors the target fabrication system, hot cell area (e.g., Mo recovery and purification, uranium recovery and recycle system), process utility and support systems, and waste handling activities. The FPC system functions also include radiation monitoring, CAAS, HMis, safe shutdown control and initiation, supervisory information, and alarms. The BMS is a subsystem to the FPC system and monitors the facility ventilation system.

The primary control location of the FPC system is in the control room. The control room FPC system operates with a synchronized hot standby redundant system structure. The hot standby workstations provide redundant hardware with identical PLC software systems as automatic backup control systems.

The primary and backup PLC systems monitor each other. On loss of synchronizing signal from one system, the other system continues with control and monitoring. This automatic backup control system minimizes the likelihood of downtime during Mo production processing.

7.2.3.2 Engineered Safety Feature Actuation Systems The operator will have direct visualization of critical values and the ability to observe status of the features described in Table 7-13 (Section 7.4 .1). The engineered safety feature actuation system dedicated displays will perform the following functions :

  • Static display - This display will show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive control functionality.
  • Alarm/event annunciator display panel - This panel will display any event or alarm that is defined for the process. The display will enable the operator to acknowledge current events and alarms, and will provide a historical record of events.
  • Dynamic interface display panel or HMI - This panel will enable the operator to perform tasks, change modes, enable/disable overrides, and other tasks that require operator input to allow, perform, or modify a task or event.

The set of displays will be arranged in a workstation. This workstation will also include a keyboard and mouse that will be used to interface with the system.

7.2.3.3 Control Room/Human-Machine Interface Description The operator will have direct visualization of critical values and the ability to input control functions into the FPC system. The FPC system dedicated displays will perform the following functions :

  • Static display - This display will show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive control functionality.
  • Alarm/event annunciator display panel - This panel will display any event or alarm that is defined for the process. The display will enable the operator to acknowledge current events and alarms, and will provide a historical record of events.
  • Dynamic interface display panel or HMI - This panel will enable the operator to perform tasks, change modes, enable/disable overrides, and other tasks that require operator input to allow, perform, or modify a task or event.

The set of displays will be arranged in a workstation. This workstation will also include a keyboard and mouse that will be used to interface with the system.

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  • !*.* ~ . NOITHWEST MEIMCAL &SOTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.2.3.4 Building Management System The BMS will control the facility ventilation system and receive indications from the fire protection, FPC, and process vessel ventilation systems. The primary purpose of the BMS is to control the air balance of the facility ventilation system and to shut down the facility ventilation system in the event of receiving an alarm from the fire protection system or off-normal conditions indicated by the FPC or engineered safety features.

The operator will have direct visualization of critical values and the ability to input control functions into the BMS. The BMS dedicated displays will perform the following functions in the control room:

  • Static display - This display will show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive control functionality.
  • Alarm/event annunciator display panel - This panel will display any event or alarm that is defined for the process. The display will enable the operator to acknowledge current events and alarms, and will provide a historical record of events.
  • Dynamic interface display panel or HMI - This panel will enable the operator to perform tasks, change modes, enable/disable overrides, and other tasks that require operator input to allow, perform, or modify a task or event.

The set of displays will be arranged in a workstation. This workstation will also include a keyboard and mouse that will be used to interface with the system.

7.2.3.5 Fire Protection System The fire protection system will report the status of the fire protection equipment to the central alarm station and the RPF control room with sufficient information to identify the general location and progress of a fire within the protected area boundaries. Initiating devices for the fire detection and alarm subsystem, including monitoring devices for the fire suppression subsystem, will indicate the presence of a fire within the facility.

Once an initiating device activates, signals will be sent to the fire alarm control panel . The fire alarm control panel will transmit signals to the central alarm station and perform any ancillary functions . As an example, signals from the fire control panel may initiate actions such as shutdown of the ventilation equipment or actuating the deluge valves. The fire protection system is described in Chapter 9.0, Section 9.3.

7.2.3.6 Facility Communication Systems The RPF communication systems will relay information within the facility during normal and emergency conditions. The systems are designed to enable the RPF operator on duty to be in communication with the supervisor on duty, health physics staff, and other personnel required by the technical specifications, and to enable the operator, or other staff, to announce the existence of an emergency in all areas of the RPF complex. Two-way communication will be provided between all operational areas and the control room. Facility communications system is described in Chapter 9.0, Section 9.4.

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  • ~ ~-~! : NDllTHWESTMEOtW.ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.2.3.7 Analytical Laboratory System The analytical laboratory will support the production of the Mo product and recycle of uranium. Samples from the process will be collected, transported to the laboratory, and prepared in the laboratory gloveboxes and hoods, depending on the analysis to be performed. The analytical laboratory equipment will be provided as vendor package units . Control room monitoring of the analytical laboratory will be limited to the facility systems, including ventilation and radiation monitoring systems. Analytical laboratory system is described in Chapter 9.0, Section 9.7 .3.

7.2.4 System Performance Analysis The RPF I&C system will monitor the processes and ESFs when required. The IROFS will be managed by the FPC system. The FPC system will provide the central decision-making processor that evaluates monitored parameters from the various plant instrumentation and from the radiation monitoring systems of the CAMs, CAAS, and RAMs . The analysis herein discusses safety as it relates to the IROFS design criteria and design basis. Potential variables, conditions, or other items that will be probable subjects of technical specifications associated with the RPF I&C systems are provided in Chapter 14.0, "Technical Specifications."

7.2.4.1 Facility Trip and Alarm Design Basis The design basis information for the FPC system trip functions is based on the following two requirements from Title 10, Code of Federal Regulations, Part 70 (10 CFR 70), "Domestic Licensing of Special Nuclear Material."

  • Double-contingency principle - Process designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible (baseline design criteria of 10 CFR 70.64, "Requirements for New Facilities or New Processes at Existing Facilities," paragraph [9]) .
  • The safety program will ensure that each IROFS will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section (10 CFR 70.61, "Performance Requirements," paragraph [e]).

The FPC system trip and alarm annunciation are protective functions and will be part of the overall protection and safety monitoring systems for the RPF. The specific equipment design basis for the instrumentation and equipment used for the FPC system trip and alarming functions is discussed in Section 7.2.2.

The following discussion relates to the design basis used for monitoring specific signal values for RPF trips and alarms, requirements for performance, requirements for specific modes of operation of the RPF and the FPC system, and the general design criteria noted in Table 7-1.

7.2.4.1.1 Safety Functions Corresponding Protective or Mitigative Actions for Design Basis Events IEEE 603-2009, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations (Sections 4a and 4b). The results of the integrated safety analysis (ISA) for the RPF structures, systems, and components (SSC) are discussed in Chapter 13 .0, "Accident Analysis." Conditions that require monitoring and the subsequent action to be taken are described in Chapter 13 .0.

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  • NCMITHWtST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.2.4.1.2 Variable Monitored to Control Protective or Mitigative Action IEEE 603-2009 (Section 4d). The list of variables to be monitored in the RPF to eliminate or reduce the exposure for the operator will be provided in the Operating License Application.

7.2.4.1.3 Functional Degradation of Safety System Performance IEEE 603-2009 (Section 4h). These design requirements will be factored in and will be evaluated in the Operating Licensing Application.

7.2.4.2 Analysis 7.2.4.2.1 Facility Process Control System Trip Function Conformance to Applicable Criteria The FPC system will perform a trip as a protective function as part of the RPF safety analysis. The associated design criteria are discussed in Sections 7.2.1 and 7.2.2. The following discussions relate to conformance to the criteria for the FPC system trip function .

7.2.4.2.2 General Functional Requirement Conformance IEEE 603-2009 (Section 5). The FPC system will initiate and control ESF activation and isolation when the system detects an off-normal event appropriate for activation. The FPC system trips are discussed in Section 7 .2.4.1. These monitored values and subsequent trips are a result of the preliminary accident analysis in Chapter 13 .0 and provide a means to mitigate or reduce the consequences from the design basis accident to acceptable levels.

7.2.4.2.3 Requirements on Bypassing Trip Functions Conformance IEEE 603-2009 (Sections 5.8, 5.9, 6.6, and 6.7). Trip override or bypass is recognized as a design requirement. Channel bypass will be allowed based on the nature of the signal. No channel bypass will be allowed without a visual indication on the FPC system display and recording the bypass event in the historical log.

7.2.4.2.4 Requirements on Setpoint Determination and Multiple Setpoint Conformance IEEE 603-2009 (Section 6.8). Table 7-1 discusses the criteria to be used for setpoint derivation.

Setpoints will be calculated in accordance with ISA-RP-67.04.02, Methodologies f or the Determination of Setpoints for Nuclear Safety-Related Instrum entation .

7.2.4.2.5 Requirements for Completion of Trip Conformance IEEE 603-2009 (Section 5.2). The ESF and the interaction of a mitigative action going to completion will be provided in the design. The FPC system will monitor for a complete trip of the ESF . This information will be available on the operator display for the FPC system and at the local HMI terminals near the hot cell. An alarm/event annunciation will be displayed to the operator. Section 7.4.1 describes the activation of the ESF, alarm/event strategy, and operator requirements to manually reset the system after a facility trip.

7.2.4.2.6 Requirements for Manual Control of Trip Conformance IEEE 603-2009 (Section 6.2). The FPC system will have the ability to perform a manual activation of the ESF. Section 7.4.1 describes the activation of the ESF, alarm/event strategy, and operator requirements to manually reset the system after a facility trip.

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  • NORTHWEST MEDtCAl ISOTDPll Chapter 7.0 - Instrumentation and Control Systems 7.2.4.3 Conclusion The I&C systems for the RPF will meet the stated design criteria and design basis requirements outlined in NUREG-1 53 7, Guidelines fo r Preparing and Reviewing Applications for the Licensing of Non -Power Reactors - Format and Content. A crosswalk of the I&C subsystems, along with a cross-reference to specific design criteria, is presented in Table 7-2.

Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (5 pages)

Criteriaa Design basis applicability Functional means IEEE 379

  • Safety DCS preapproved platfo rm Single failure criterion
  • FPC system di splay
  • Redundant independent isolation
  • Redundant operator interface workstations
  • ESFs manual isolation
  • Redundant sensors
  • Safety DCS pre-approved platform for an Reliability analysis
  • FPC system display SIS criterion
  • FPC system IROFS end devices Redundant independent isolation
  • ESFs manual isolation
  • Redundant operator interface workstations
  • Redundant sensors
  • FPC system See Section 7 .3 for detai ls.

Standard criteria safety

  • FPC system display system
  • FPC system display development of the Construction Permit Class 1E equipment
  • FPC system IROFS end devices Application.

and circuits

  • Additional details will be developed for the
  • ESFs manual isolation Operating License Application.

--~

IEEE 323

  • Standard supports selection and Qualifying Class IE
  • FPC system di splay quali fication of equipment to be Class 1E Equipment
  • FPC system IROFS end devices use qualifi ed.
  • This standard will be reevaluated in the
  • ESFs manual iso lation Operating License Appli cation for appli cability.

IEEE 344

  • Standard supports selection and Recommended practice
  • FPC system display qualification of equipment to be Class IE for seismic
  • FPC system IROFS end devices use qualified.

qualification

  • Standard will be reevaluated in the
  • ESFs manual isolation Operating License Application for applicability.

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Criteriaa Design basis applicability Functional means IEEE 338

  • Standard supports selection of equipment; Criteria for the periodic
  • FPC system display which resulted in the use of general design survei llance testing of
  • FPC system IROFS end devices criteri a (presented in Chapter 3.0) during safety systems
  • ESFs development of the Construction Permit
  • ESFs manual isolation Application.
  • Standard will be reevaluated in the Operating License Appli cation for app licability.

IEEE 497

  • Standard supports selection of accident Criteria for accident
  • FPC system display monitoring equipment (e.g., radiation monitoring instruments
  • FPC system JROFS end devices monitoring, annunciation), which resulted
  • ESFs in the use of general design criteria
  • CAAS (presented in Chapter 3.0) during
  • RAMs development of the Construction Permit
  • Standard will be reevaluated in the Operating License Application for applicability.

IEEE 7-4.3 .2

  • Programming software must comp ly with Criteri a fo r digital
  • FPC system display this criteria and with the NWMI Software computers in safety
  • HM! di spl ays Quality Assurance Plan (prepared during systems development of the Operating License Application), which will be developed per the design criteria outlined in Chapter 3.0 and thi s standard.
  • Software and hardware used for the displays fo r the FPC system and HMI must also fo ll ow guide lines set fo rth in this standard .
  • Standard wi ll be reevaluated in the Operating License Applicati on for applicability.

IEEE 828

  • Complies with IEEE 7-4.3 .2 and the NWMI Configuration
  • FPC system display Software Quality Assurance Plan management in systems
  • HMI displays Standard will be reevaluated in the and software Operating License Application for engineering applicability.

IEEE 829

  • Complies with IEEE 7-4.3.2 and the NWMI Software and system
  • FPC system di splay Software Quality Assurance Pl an test documentation
  • HMI di spl ays
  • Standard will be reevaluated in the Operating License Application for appli cabil ity.

IEEE 1012

  • Complies with IEEE 7-4.3.2 and the NWMI Criteria for software
  • FPC system display Software Quality Assurance Plan verification and
  • HMI displays
  • Standard will be reevaluated in the validation Operating License Application for applicability.

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. ' ! ~.~~ ;. NOllTNWESTMEDtW.JSOTOl"fS NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (5 pages)

Criteriaa Design basis applicability Functional means IEEE 1028

  • Complies with IEEE 7-4.3.2 and the NWMI Software reviews and
  • FPC system display Software Quality Assurance Plan audits
  • HMI displays
  • Standard wi ll be reevaluated in the Operating License Application for applicability.

ANS-10.4

  • Complies with IEEE 7-4.3.2 and the NWMI Verification and
  • FPC system display Software Quality Assurance Plan validation for non-
  • HMI displays
  • Standard will be reevaluated in the safety software Operating License Application for applicability.

ANSI/ISA 67.04.01

  • Incorporated into overall design and the Setpoints for nuclear
  • FPC system IROFS end devices Construction Permit Application.

safety-related

  • Standard will be reevaluated in the instruments Operating License Application for applicability.

ANSI/ISA 84.00.01,

  • Standard supports the design and Parts l , 2, and 3
  • FPC system display development of non-safety-related systems Functional safety:
  • HMI displays that rely on safety, reliability, and safety instrumented functionality and was used during systems for the process development of the Construction Permit industry sector Application.
  • Standard will be reevaluated in the Operating License Application for applicability.

NUREG-0700

  • Standard supports the design and Human-system
  • FPC system display development of non-safety-related systems interface design review
  • HMI displays that pertain to control room arrangement, guidelines screen developments, and operator interface, and was used during development of the Construction Permit Application.
  • Standard wi ll be reevaluated in the Operating License Application for applicability.

NUREG/CR-6463

  • Standard supports the design, development, Review guidelines on and review of safety-related software and software languages for was used during development of the use in nuclear power Construction Permit Application.

plant safety systems

  • Standard will be reevaluated in the Operating License Application for applicability.

NUREG/CR-6090

  • Standard supports the design, development, PLC and applications and review of safety-related and non-in nuclear reactor safety-related software and was used during systems development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.

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Criteriaa Design basis applicability Functional means EPRI TR-106439

  • FPC system display
  • Standard supports the design, development, Guideline on
  • HMI displays and review of safety-related systems that evaluation/acceptance pertain to obtaining software or hardware of commercial grade for the FPC system, HMI displays, and data digital equipment for acquisition systems, and was used during nuclear safety development of the Construction Permit applications Application.
  • Standard will be reevaluated in the Operating License Application for applicability.

Regulatory Guide

  • Standard supports the desi gn and 1.152
  • FPC system di splay development of redundant safety PLC Criteria for use of
  • HMI di splays pl atforms, FPC system redundant HMI computers in safety workstations, and operator interface systems workstations, and was used during development of the Construction Permit Application .
  • Standard will be reevaluated in the Operating License Application for applicabili ty.

Regulatory Guide 1.53

  • Standard supports the design and Single failure criterion
  • FPC system display development of high-integrity safety PLCs, evaluation for safety
  • FPC system IROFS end devices redundant channels for ESFs, redundant systems
  • ESFs operator interface workstations, redundant
  • ESFs manual isolation sensors, and alternative manual means for ESF initiation, and was used during development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.

Regulatory Guide 5.71

  • Criteria require the development of a design Cybersecurity
  • FPC system di splay approach and impl ementation fo r programs for nuclear
  • HMI di splay cybersecuri ty.

fac ilities

  • Standard will be reevaluated in the Operating License Application fo r app licability.

a Full references are provided in Section 7. 7.

CAAS criticality accident alarm system. IROFS items relied on fo r safety.

CAM continuous air monitor. NWM I orthwest Medical Isotopes, LLC.

DCS digital control system. PLC programmable logic controller.

ESF engineered safety feature. RAM radiation alarm monitor.

FPC facility process control. SIS safety instrumented system.

HM! human-machine interface.

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. NWMI NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems
  • ! * * ~ . ** NORTMWEST MEDICAi. tsOTOPlS 7.3 PROCESS CONTROL SYSTEMS The process control systems for the RPF will include SNM preparation and handling processes and radioisotope production processes. SNM preparation and handling processes include uranium recovery and recycle, and target fabrication . Radioisotope production processes include target receipt and disassembly, target dissolution, Mo recovery and purification, and waste handling.

The RPF process control will be administered by the FPC system and is described in Section 7.2.3 . The FPC system will perform the following high-level process functions.

  • Monitor the remote valve position for routing process fluid for inter-equipment process fluid transfers - For specific transfers identified by the operator, the FPC system will provide a permissive to allow for the active pump in that circuit to be energized once the operator has manually configured the routing.
  • Monitor and control inter-equipment process fluid transfers in the RPF - For transport requiring a pump, the FPC system will control the ability of the pump to be energized. For specific transfers, the FPC system will provide controlled fluid flow transfers based on a closed-loop flow control . The operator wil 1initialize the transfer of fluids .
  • Other process fluid transfers, including:

Dissolved low-enriched uranium (LEU) solution to the Mo recovery and purification system Uranium solution to the uranium recovery and recycle system Liquid wastes to the waste handling system The I&C system for process utilities and support systems and for the ventilation systems will be described in more detail in the Operating License Application . The process systems described below provide for reliable control of the SNM preparation and handling process and the radioisotope production processes, and include:

  • Range of operation of the sensor that is sufficient to cover the expected range of variation of the monitored variable during normal and transient process operation
  • Reliable information about the status and magnitude of the process variable necessary for the full operating range of the radioisotope production and SNM recovery and recycle processes
  • Reliable operation in the normal range of environmental conditions anticipated within the facility
  • Safe state during loss of electrical power Potential variables, conditions, or other items that will be probable subjects of technical specifications associated with the RPF process control systems are discussed in Chapter 14.0.

7.3.1 Uranium Recovery and Recycle System The uranium recovery and recycle system will process raffinate from the Mo recovery and purification system for recycle to the target fabrication system. Two cycles of uranium purification will be included to separate uranium from unwanted fission products using ion exchange. The first ion exchange cycle will separate the bulk of the fission product contaminant mass from the uranium product. Product will exit the ion exchange column as a dilute uranium stream that is concentrated to control the stored volume of process solutions. Uranium from the first cycle will then be purified by a nearly identical second cycle system to further reduce fission product contaminants to satisfy product criteria. Each ion exchange system feed tank will include the capability of adding a reductant and modifying the feed chemical composition such that adequate separations are achieved, while minimizing uranium losses.

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  • ~ *.* ~
  • NOmfMST MB>>CAl lSOTOPH NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Due to the variety of process activities performed during uranium recovery and recycle, the system description is divided into the following subsystems :
  • Primary ion exchange
  • Primary concentration
  • Secondary ion exchange
  • Secondary concentration
  • Spent ion exchange resin
  • Waste collection 7.3.1.1 Design Criteria Design criteria for the uranium recovery and recycle I&C systems are described in Section 7.2.

7.3.1.2 Design Basis and Safety Requirements The design basis and safety requirements for the uranium recovery and recycle I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0, "Engineered Safety Features."

7.3.1.3 System Description The uranium recovery and recycle I&C system will be defined in the Operating License Application. The strategy and associated parameters for the system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

Normal operating functions will be performed remotely using the FPC system in the control room.

Table 7-3 lists the anticipated control parameters, monitoring parameters, and primary control locations for each subsystem. In addition, the implementation of IROFS CS-14, CS-15 , CS-20, CS-27, and RS-10 interlocks for this system are under development. Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements, indication, alarm, and control features wi ll be developed for the Operating License Application.

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! *: ~ . NC*TNWEST MEOtcAL tsOTOfl'H NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-3. Uranium Recovery and Recycle Control and Monitoring Parameters (2 pages)

Subsystem Control parameters Primary control name (automatic/manual) Monitoring parameters location Impure . Flowrate (A) .. Density Control room urarnum

. Pump actuation (M)

.. Differential pressure Flowrate collection

. Temperature (A)

Pump motor speed (A)

Level Valve actuation (NM) Pressure Temperature Valve position Primary ion . Flowrate (A) .. Analyzer, uranium Control room exchange Pump actuation (NM) Density Pump motor speed (A) Differential pressure Temperature (A) Flowrate Valve actuation (NM)

.. Flowrate totalizer Level Pressure

.* Temperature Valve position Primary . Density (A) ... Analyzer, uranium Density Control room concentration

. Level Flowrate (A)

.. Differential pressure

. Pump (A) Flowrate

.. Pump actuation (NM) motor speed (A)

.. Level

. Temperature (A)

Valve actuation (AIM)

Pressure Temperature Valve position Secondary ion . Flowrate (A) .. Analyzer, uranium Control room exchange

.. Pump actuation (AIM)

Pump motor speed (A) .. Density Differential pressure

. Temperature (A)

Valve actuation (NM) .. Flowrate Flowrate totalizer

.. Level Pressure

. Temperature Valve position Secondary .. Density (A) .. Analyzer, uranium Control room concentration

.. LevelFlowrate (A)

(A) .. Density Differential pressure

. Pump motor speed (A)

Pump actuation (AIM)

.. Flowrate Level

. Temperature .. Pressure

. Valve actuation(A)(NM) Temperature Valve position Uranium .. Flowrate (A) . Density Control room recycle Pump actuation (AIM) Differential pressure Pump motor speed (A) Flowrate Valve actuation (NM)

.. Level Pressure

. Temperature Valve position 7-24

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    • .*- NWMI NOflJTHWEST M£DtCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-3. Uranium Recovery and Recycle Control and Monitoring Parameters (2 pages)

Subsystem Control parameters Primary control name (automatic/manual) Monitoring parameters location

. Flowrate (A) . Density Uranium decay

. Pump actuation (AIM) . Differential pressure Control room and

. Pump motor speed (A) . Flowrate accountability

. Temperature (A) . Level

. Valve actuation (AIM)

... Temperature Pressure Valve position Spent ion .. Flowrate (A) .. Analyzer, uranium Control room exchange resin

.. Pump actuation (AIM)

Pump motor speed (A) .. Differential pressure Flowrate Valve actuation (AIM)

.. Level Pressure Valve position Waste Flowrate (A)

. Density Control room coll ection

. Pump actuation (AIM)

. Differential pressure Flowrate

. Pump motor speed (A)

Level

. Temperature (A)

Valve actuation (AIM) Pressure Valve position Table 7-4 provides a preliminary listing of the interlocks and permissive signals that have been identified.

These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages)

Hard-wired or Interlock or permissive input PLC Safety Interlock Impure uranium collection tank (UR-TK-lOOA) low-level PLC NIA switch (typical of eight tanks)

Impure uranium collection tank (UR-TK-lOOA) high-level PLC NIA switch (typical of eight tanks)

Impure uranium collection tank (UR-TK-1 OOA) high- PLC NIA temperature switch (typical of eight tanks)

IX feed tank 1 (UR-TK-200) low-level switch PLC NIA IX feed tank 1 (UR-TK-200) high-l evel switch PLC NIA IX feed tank 1 (UR-TK-200) high-temperature switch PLC NIA IX column IA (UR-IX-240) high-uranium alarm (AAH-252) PLC NIA IX column IA U solution filter (UR-F-250) high-differential PLC NIA pressure alarm IX column IA waste filter (UR-F-255) high-differenti al PLC NIA pressure alarm IX column lB (UR-IX-260) high-uranium alarm (AAH-272) PLC NIA IX column lB U solution filter (UR-F-270) high-differential PLC NIA pressure alarm 7-25

0

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! *.*.~ * . NCNITNWESf MEDK:Al. ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages)

Hard-wired or Interlock or permissive input PLC Safety Interlock IX Column lB Waste Filter (UR-F-275) high-differential PLC NIA pressure alarm Concentrator 1 feed tank (UR-TK-300) low-level switch PLC NIA Concentrator 1 feed tank (UR-TK-300) high-level switch PLC NIA Concentrator 1 (UR-Z-320) low-liquid level alarm PLC NIA Concentrator 1 (UR-Z-320) high-liquid level alarm PLC NIA Concentrator 1 (UR-Z-320) demister high-differential pressure PLC NIA alarm Concentrator 1 (UR-Z-320) condenser high-differential PLC NIA pressure alarm Concentrator 1 (UR-Z-320) condenser high-offgas temperature PLC NIA alarm Condensate sample tank IA (UR-TK-340) high-liquid level PLC NIA alarm Condensate sample tank 1A (UR-TK-340) high-uranium Hard-wired Reroute condensate transfer to switch (AE-356) UR-TK-300 (position V-396, close V-397)

Close IX column eluent addition control valves (V-244 and V-264)

Condensate delay tank 1 (UR-TK-370) high-liquid level alarm PLC NIA Condensate sample tank lB (UR-TK-340) high-liquid level PLC NIA alarm Condensate sample tank 1B (UR-TK-370) high-uranium Hard-wired Permissive to route condensate to switch (AE-386) WH-TK-420 (position V-496, open V-397)

Permissive to open IX column eluent addition control valves (V-244 and V-264)

IX feed tank 2A (UR-TK-400) low-level switch PLC NIA IX feed tank 2A (UR-TK-400) high-level switch PLC NIA IX feed tank 2A (UR-TK-400) high-temperature switch PLC NIA IX feed tank 2B (UR-TK-420) low-level switch PLC NIA IX feed tank 2B (UR-TK-420) high-level switch PLC NIA IX feed tank 2B (UR-TK-420) high-temperature switch PLC NIA IX column 2A (UR-IX-460) high-uranium alarm (AAH-472) PLC NIA IX column 2A U solution filter (UR-F-470) high-differential PLC NIA pressure alarm IX column 2A waste filter (UR-F-475) high-differential PLC NIA pressure alarm IX column 2B (UR-IX-480) high-uranium alarm (AAH-492) PLC NIA 7-26

~ . NWMI

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..NOITifWUT liWMCAl ISOTOf'U NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages)

Hard-wired or Interlock or permissive input PLC Safety Interlock IX column 2B U solution filter (UR-F-490) high-differential PLC NIA pressure alarm IX column 2B waste filter (UR-F-495) high-differential PLC NIA pressure alarm Concentrator 2 feed tank (UR-TK-500) low-level switch PLC NIA Concentrator 2 feed tank (UR-TK-500) high-level switch PLC NIA Concentrator 2 (UR-Z-520) low-liquid level alarm PLC NIA Concentrator 2 (UR-Z-520) high-liquid level alarm PLC NIA Concentrator 2 (UR-Z-520) demister high-differential pressure PLC NIA alarm Concentrator 2 (UR-Z-520) condenser high-differential PLC NIA pressure alarm Concentrator 2 (UR-Z-520) condenser high-offgas temperature PLC NIA alarm Condensate sample tank 2A (UR-TK-540) high-liquid level PLC NIA alarm Condensate sample tank 2A (UR-TK-540) high-uranium Hard-wired Reroute condensate transfer to switch (AE-556) UR-TK-500 (position V-596, close V-5 97)

Close IX co lumn eluent addition control valves (V-464 and V-484)

Condensate delay tank 2 (UR-TK-560) high-liquid level alarm PLC NIA Condensate sample tank 2B (UR-TK-570) high-liquid level PLC NIA alarm Condensate sample tank 2B (UR-TK-570) high-uranium Hard-wired Permissive to route condensate to switch (AE-586) WH-TK-420 (position V-596, open V-597)

Permissive to open IX column eluent addition control valves (V-464 and V-484)

Concentrate receiver tank (UR-TK-600) high-liquid level PLC NIA alarm Concentrate receiver tank (UR-TK-600) high-temperature PLC NIA alarm Product sample tank (UR-TK-620) high-liquid level alarm PLC NIA Product sample tank (UR-TK-620) high-temperature alarm PLC NIA Urani um rework tank (UR-TK-660) high-liquid level alarm PLC NIA Uranium rework tank (UR-TK-660) high-temperature alarm PLC NIA Urani um decay tank (UR-TK-700A) high-liquid level alarm PLC NIA (typical of 17 tanks)

Uranium decay tank (UR-TK-700A) high-temperature alarm PLC NIA (typical of 17 tanks) 7-27

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. ;.-.~*.. NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 0

!*.*! 0 NOITH'W(ST MlDICAl tsOlWU Table 7-4. Urani um Recycle and Recovery System Interlocks and Permissive Signals (4 pages)

Hard-wired or Interlock or permissive input PLC Safety Interlock Uranium accountability tank (UR-TK-720) high-liquid level PLC NIA alarm Uranium accountability tank (UR-TK-720) high-temperature PLC NIA alarm Spent resin tank A (UR-TK-820A) high-liquid level alarm PLC NIA Spent resin tank A (UR-TK-820A) high-temperature alarm PLC NIA Spent resin tank B (UR-TK-820B) high-liquid level alarm PLC NIA Spent resin tank B (UR-TK-820B) high-temperature alarm PLC NIA Resin transfer liquid tank (UR-TK-850) high-liquid level PLC NIA alarm IX waste collection I tank (UR-TK-900) high-liquid level PLC NIA alarm IX waste collection I tank (UR-TK-900) high-temperature PLC NIA alarm IX waste collection 2 tank (UR-TK-920) high-liquid level PLC NIA alarm IX waste collection 2 tank (UR-TK-920) high-temperature PLC NIA alarm IX ion exchange. TBD to be determined .

PLC programmable logic controller.

7.3.1.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system wi ll be provided in the Operating License Application.

7.3.2 Target Fabrication System The target fabrication system will produce LEU targets from fresh LEU material and recycled uranyl nitrate. The system will commence with the receipt of fresh LEU from the U.S. Department of Energy, and end with packaging new targets for shipment to the university research reactor faci lities.

Due to the variety of process activities performed during target fabrication, the system description is divided into the following subsystems.

  • Fresh uranium receipt and dissolution
  • Nitrate extraction
  • Acid-deficient uranyl nitrate (ADUN) concentration
  • [Proprietary Information]
  • [Proprietary Information]
  • [Proprietary Information]
  • Target fabrication waste
  • Target assembly
  • [Proprietary Information]
  • New target handling 7-28

..;*. NWMI

  • ~ *.*! . . NOtmfWHT MEDICAi. tsOTDPH NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.3.2.1 Design Criteria Design criteria for the target fabrication I&C systems are described in Section 7.2.

7.3.2.2 Design Basis and Safety Requirements The design basis and safety requirements for the target fabrication I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0.

7.3.2.3 System Description The target fabrication I&C system will be defined in the Operating License Application. The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

Normal operating functions will be performed remotely using the FPC system HMI in the target fabrication area. Table 7-5 lists the anticipated control parameters, monitoring parameters, and primary control location for each subsystem. In addition, the implementation ofIROFS CS-14, CS-15, CS-20, CS-27, and RS-10 interlocks for this system are under development. Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements, indication, alarm, and control features will be developed for the Operating License Application.

Table 7-5. Target Fabrication System Control and Monitoring Parameters (2 pages)

I Control parameters Primary control Subsystem name (automatic/manual) Monitoring parameters location Fresh uranium receipt and .. Current (A) .. Conductivity Local

. Density dissolution Conductivity (A)

(I 00-series tag numbers) Flow totalizer (A)

Heater actuation (NM) .. Differential pressure Flowrate

.. Level (A)

Pump actuation (A/M) .. Level Pressure Temperature

. Temperature (A)

Nitrate extraction .. Valve actuation (A/M)

Analyzer, pH (A) .. Analyzer, pH Local (200-series tag numbers)

.. Contactor actuation (M)

Flow totalizer (A) . Density Differential pressure

.. Flowrate (A)

Level (A) .. Flowrate Level Pressure Pump actuation (NM)

Pump motor speed (A) Pump motor speed Temperature (A) Temperature Valve actuation (AIM)

. Conductivity ADUN concentration (300-series tag numbers) ... Conductivity (A)

Density (A)

... Density Flowrate Local

. Flowrate (A)

. Pressure Level

. Level (A)

. Temperature

. Pump actuation (A/M)

. Pump motor speed (A)

Valve actuation (A/M) 7-29

0

~

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  • . * ~:
  • NOATNWEST MEDICAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-5. Target Fabrication System Control and Monitoring Parameters (2 pages)

Control parameters Primary control (automatic/manual) Monitoring parameters location Subsystem name

[Proprietary Information] .. Level (A) .. Flowrate Local (400-series tag numbers)

. Pump actuation (NM)

Tank agitator actuation .. Level Pressure (AIM) Temperature Tank agitator speed (A)

Temperature (A)

Valve actuation (AIM)

[Proprietary Information] .. Flowrate (A)

... Density Differential pressure Local (500-series tag numbers)

.. Pump actuation (NM)

Pump motor speed (A)

. Pressure

.. Temperature (A)

. Level Valve actuation (AIM)

Vibration dispersion . Temperature Vibration

[Proprietary Information] ..

assembly actuation (M)

Analyzer, hydrogen (A) .. Analyzer, hydrogen Local (600-series tag numbers)

.. Analyzer, oxygen (A)

Flow totalizer (A) .. Analyzer, oxygen Flowrate

.. Level (A)

Tank agitator speed (M) . Level Pressure

  • Temperature

. Temperature (A)

Valve actuation (AIM)

.. . Density Target fabrication waste Flowrate (A)

. Flowrate Local (700-series tag numbers)

. Level (A)

.. Level

.. Pump actuation (AIM)

Pump motor speed (A)

. Pressure Valve actuation (AIM) Temperature Target assembly TBD TBD Local

[Proprietary Information] TBD TBD Local New target handling TBD TBD Local ADUN acid-deficient uranyl nitrate. TBD to be determined.

LEU = low-enriched uranium.

Table 7-6 provides a listing of the target fabrication l&C system interlocks and permissive signals that have been identified. These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages)

Hard-wired or Interlock or permissive input PLC Safety interlock Dissolver column (TF-D-100) high-temperature switch PLC NIA Uranium dissolution heat exchanger (TF-E-120) chilled Hard-wired Close chilled water return control water return high-conductivity switch valve (XV-122) on high conductivity Uranium dissolution heat exchanger (TF-E-120) low- PLC NIA differential pressure alarm Uranyl nitrate storage tank (TF-TK-200) level switch PLC NIA 7-30

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~.-.~ * .. NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems

. ' ! *.* ~ ." . NOmfWEST MEOtcAl ISOTOPES Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages)

I Hard-wired or Interlock or permissive input PLC Safety interlock ADUN evaporator condenser (TF-E-350) chilled water Hard-wired Close chilled water return control return high-conductivity switch valve (HV-352) on high conductivity ADUN product heat exchanger (TF-E-360) low- PLC NIA differential pressure alarm ADUN product heat exchanger (TF-E-360) chilled water Hard-wired Close chilled water return control return high-conductivity switch valve (HV-361) on high conductivity ADUN evaporator reboiler (TF-E-330) steam condensate Hard-wired Close steam condensate control valve high-conductivity switch (XV-333) on high conductivity ADUN storage tank (TF-TK-400) low-level switch PLC NIA ADUN storage tank (TF-TK-405) low-level switch PLC NIA ADUN storage tank (TF-TK-410) low-level switch PLC NIA ADUN storage tank (TF-TK-415) low-level switch PLC NIA ADUN storage tank (TF-TK-400) high-level switch PLC NIA ADUN storage tank (TF-TK-405) high-level switch PLC NIA ADUN storage tank (TF-TK-401) high-level switch PLC NIA ADUN storage tank (TF-TK-415) high-level switch PLC NIA

[Proprietary Information] (TF-TK-480) high-level switch PLC NIA

[Proprietary Information] (TF-C-500) high-temperature PLC NIA switch Silicone oil heater (TF-E-550) outlet high-temperature Hard-wired NIA switch

[Proprietary Information] (TF-Z-660) high-temperature Hard-wired NIA switch

[Proprietary Information] (TF-Z-661) high-temperature Hard-wired NIA switch

[Proprietary Information] (TF-Z-662) high-temperature Hard-wired NIA switch

[Proprietary Information] (TF-Z-663) high-temperature Hard-wired NIA switch

[Proprietary Information] (TF-Z-660) door closed switch PLC NIA

[Proprietary Information] (TF-Z-661) door closed switch PLC NIA

[Proprietary Information] (TF-Z-662) door closed switch PLC NIA

[Proprietary Information] (TF-Z-663) door closed switch PLC NIA Reduction furnace offgas heat exchanger (TF-E-670) PLC NIA outlet high-oxygen concentration Reduction furnace offgas heat exchanger (TF-E-670) PLC NIA outlet high-hydrogen concentration 7-31

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.**.*.* NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems

' ~~-~~ ," . NCMTHWESTMEDICAllSOTOPCS Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages)

Hard-wired or Interlock or permissive input PLC Safety interlock Aqueous waste pencil tank (TF-TK-700) high-level alarm PLC NIA Aqueous waste pencil tank (TF-TK-705) high-level alarm PLC NIA TCE tank (TF-TK-760) high-level switch PLC NIA Target fabrication overflow tank (TF-TK-770) high-high- PLC NIA level switch ADUN acid-deficient uranyl nitrate. TBD to be determined.

PLC programmable logic controller. TCE trichloroethylene.

7.3.2.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.3 Target Receipt and Disassembly System The target receipt and disassembly system will include the delivery and receipt of the irradiated target cask, introduction of the irradiated targets into the hot cell, disassembly of the targets, and retrieval and transfer of the irradiated target material for processing. This system will feed the target dissolution system by the transfer of recovered irradiated target material through the dissolver I hot cell (DS-EN-100) and dissolver 2 hot cell (DS-EN-200) isolation door interfaces.

Due to the variety of activities performed during target receipt and disassembly, the system description is divided into the following subsystems:

  • Cask receipt
  • Target receipt
  • Target disassembly 7.3.3.1 Design Criteria Design criteria for the target receipt and disassembly I&C systems are described in Section 7.2.

7.3.3.2 Design Basis and Safety Requirements The design basis and safety requirements for the target receipt and disassembly I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0.

7.3.3.3 System Description The target receipt and disassembly I&C system will be defined in the Operating License Application.

The strategy and associated parameters for the l&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

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  • .! ~* ~ ! ." NORTHWEST Mt:DacA.L ISOT01'£S Normal operating functions will be performed remotely using the FPC system RMI in the truck bay, cask preparation airlock, and the operating gallery. Redundant control functions will be provided in the control room. In addition, the implementation of IROFS CS-14, CS-15 , CS-20, CS-27, and RS-10 interlocks for this system are under development. Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements, indication, alarm, and control features will be developed for the Operating License Application.

Prior to the start of disassembly operations, the following process control permissive signals will be required.

  • Ventilation inside the hot cell is operable .
  • Fission gas capture hood is on and functional.
  • Irradiated target material collection container is in position under the target cutting assembly collection bin.
  • Waste drum transfer port is open and there is physical space to receive the waste target hardware after disassembly and irradiated target material recovery.

The control parameters and monitoring parameters will be defined during design development for the Operating License Application.

7.3.3.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application .

7.3.4 Target Dissolution System The target dissolution system process will receive the LEU target material from the target receipt and disassembly system and dissolve the uranium and molybdenum-99 (99 Mo) in the solid irradiated target material in hot nitric acid. The concentrated uranyl nitrate solution will then be transferred to the Mo recovery and purification system for further processing.

The target dissolution process will be operated in a [Proprietary Information] transferred to a collection container. The collection container will move through the pass-through to a dissolver basket positioned over a dissolver, the target material will then be dissolved and the resulting solution transferred to the Mo recovery and purification system.

Target dissolution of irradiated LEU will result in gaseous fission products (iodine [I], krypton [Kr], and xenon [Xe]) with very high radiation fields . A primary function of the process offgas systems will be to control release of these gases both internal and external to the facility . The dissolver offgas treatment system will include the nitrogen oxide (NOx) treatment and fission gas treatment subsystems.

Due to the variety of process activities performed during target dissolution, the system description is divided into the following subsystems:

  • Target dissolution 1 and target dissolution 2
  • NOx treatment 1 or NOx treatment 2
  • Pressure relief
  • Primary fission gas treatment
  • Secondary fission gas treatment
  • Waste collection 7-33

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. ' ! ~.* ! : NOllTHWfST MEDICAL ISOTOfJfS NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.3.4.1 Design Criteria Design criteria for the target dissolution I&C systems are described in Section 7.2 .

7.3.4.2 Design Basis and Safety Requirements The design basis and safety requirements for the target dissolution I&C systems are described in Section 7.2.

The ESFs for this system are listed in Chapter 6.0.

7.3.4.3 System Description The target dissolution I&C system will be defined in the Operating License Application. The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

Loading of [Proprietary Information] into the dissolver will involve mechanical handling of the transfer containers. Operators using remote in-cell cranes and manipulators will perform these functions. Other normal operating functions will be performed remotely using the FPC system HMI in the operating gallery. Redundant control functions will be provided in the control room. Table 7-7 lists the anticipated control parameters, monitoring parameters, and primary control locations for each subsystem. Details of the control system (e.g., interlocks and permissive signals), control logic, indication, alarm, and control features will be defined in the Operating License Application.

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  • . * .*
  • NORTHWEST ll£DICAl ISOTOl'U Table 7-7. Target Dissolution System Control and Monitoring Parameters Subsystem Control parameters Primary control name (automatic/manual) Monitoring parameters location

. Dissol ver agitator actuation . Dissol ver agitator speed Target

. Flowrate Operating gall ery

. (AIM) ... Level dissolution 1 Flowrate totalizer and 2

. Flowrate Dissol ver agitator speed (A)

. Pump actuation (A)

. Pressure

.. (AIM)

Pump motor speed (A)

. Temperature Radiation

. Temperature (A)

Valve actuation (AIM) . Valve position

.. . Differential pressure NOx treatment I Flowrate (A)

. Flowrate Operating gallery or 2

.. Pump actuation (AIM)

Pump motor speed (A) .. Flowrate totalizer

. Temperature (A)

Valve actuation (AIM) .. Level Pressure

.. Radiation Temperature Valve position

. Pump actuation (AIM) .

Pressure relief

. Pump motor speed (A) . Flowrate Operating gall ery

. Temperature (A) . Level

. . Pressure Valve position Primary fission .. Valve actuation (AIM)

Temperature (A) .. Differential pressure Operating gallery gas treatment Valve actuation (AIM)

.. Flowrate Pressure

.. Radiation Temperature Valve position Secondary Valve actuation (AIM)

. Differential pressure Operating gallery fi ssion gas treatment .. Flowrate Pressure

.. Radiation Temperature Valve position Waste collection ... Pump actuation (AIM)

Pump motor Speed (A)

Differential pressure Flowrate Operating gallery

. Temperature (A)

Valve actuation (AIM) .. Level Temperature

.. Pressure Radi ation Valve position NOx nitrogen oxide.

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~ ~. * ! ." , NOftTHWUT Mfl>tw ISOnftl Table 7-8 provides a preliminary listing of the target dissolution I&C system interlocks and permissive signals that have been identified. In addition, the implementation ofIROFS CS-14, CS-15, CS-20, CS-27, and RS-10 interlocks for this system are under development. These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-8. Target Dissolution System Interlocks and Permissive Signals (2 pages)

Hard-wired or Interlock or permissive input PLC Safety interlock Dissolver I (DS-D-100) high-liquid level alann PLC NIA Dissolver l (DS-D-100) low-liquid level alarm PLC NIA Dissolver I (DS-D-100) high liquid temperature alarm PLC NIA Dissolver I Condenser (DS-E-130) high gas temperature alann PLC NIA Dissolver 2 (DS-D-200) high-liquid level alarm PLC NIA Dissolver 2 (DS-D-200) low-liquid level alarm PLC NIA Dissolver 2 (DS-D-200) high liquid temperature alarm PLC NIA Dissolver 2 condenser (DS-E-230) high gas temperature alarm PLC NIA Primary caustic scrubber I (DS-C-3 10) high-liquid level alarm PLC NIA Caustic scrubber I (DS-C-310) high gas temperature PLC NIA NOx oxidizer I (DS-C-340) high-liquid level alann PLC NIA NOx oxidizer I (DS-C-340) high gas temperature PLC NIA NOx absorber I (DS-C-370) high-liquid level alarm PLC NIA NOx absorber I (DS-C-370) high gas temperature PLC NIA Primary caustic scrubber 2 (DS-C-4 10) high-liquid level alann PLC NIA Caustic scrubber 2 (DS-C-410) high gas temperature PLC NIA NOx oxidizer 2 (DS-C-440) high-liquid level alann PLC NIA NOx oxidizer 2 (DS-C-440) high gas temperature PLC NIA NOx absorber 2 (DS-C-470) high-liquid level alann PLC NIA NOx absorber 2 (DS-C-470) high gas temperature PLC NIA Pressure relief tank (DS-TK-500) high-pressure alarm Hard-wired Opens valve to capture dissolver gases Pressure relief tank (DS-TK-500) high-liquid level alarm PLC NIA Pressure relief tank (DS-TK-500) low-liquid level alann PLC NIA Dryer A (DS-E-610A) high gas temperature alann PLC NIA Primary adsorber A (DS-SB-620A) high gas temperature alann PLC NIA Filter A (DS-F-630A) high-pressure differential alarm PLC NIA Dryer B (DS-E-610B) high gas temperature alarm PLC NIA Primary adsorber B (DS-SB-620B) high gas temperature alann PLC NIA Filter B (DS-F-630B) high -pressure differential alann PLC NIA Dryer C (DS-E-610C) high gas temperature alann PLC NIA Primary adsorber C (DS-SB-620C) high gas temperature alann PLC NIA Filter C (DS-F-630C) high-pressure differential alarm PLC NIA 7-36

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  • ~ * ** ~
  • NOl1lfMST llEDtCAl tSOTOf'U Table 7-8. Target Dissolution System Interlocks and Permissive Signals (2 pages)

Hard-wired or Interlock or permissive input PLC Safety interlock Secondary adsorber A (DS-SB-730A) high gas temperature alarm PLC NIA Secondary adsorber B (DS-SB-730B) high gas temperature alarm PLC NIA Secondary adsorber C (DS-SB -730C) high gas temperature alarm PLC NIA Waste collection and sampling tank 1 (DS-TK-800) high-liquid level PLC NIA alarm Waste collection and sampling tank I (DS-TK-800) high-liquid PLC IA temperature alarm Waste collection and sampling tank 2 (DS-TK-820) high-liquid level PLC NIA alarm Waste co ll ection and sampling tank 2 (DS-TK-820) high-liquids PLC NIA temperature alarm NIA not app licable. PLC programmable logic contro ller.

NOx = nitrogen oxide.

7.3.4.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.5 Molybdenum Recovery and Purification System The Mo recovery and purification system will receive the impure Mo/uranium solution from the target dissolution system into feed tank lA and feed tank lB (MR-TK-100 and MR-TK-140) located in the tank hot cell . The Mo/uranium solution will then be transferred to process hot cells and processed through three separate ion exchange unit operations to achieve the desired product criteria. A collection container holding the separated and purified Mo product material wi ll be used for final chemical adjustment and sampling for verification of batch acceptance. The product wi ll be sampled and weighed, placed in stainless steel bottles with lids applied and tightened, loaded into shielded containers, and then shipped in an approved cask.

Due to the variety of activities performed during Mo recovery and purification, the system description is divided into the following subsystems:

  • Primary ion exchange
  • Secondary ion exchange
  • Tertiary ion exchange
  • Mo product 7.3.5.1 Design Criteria Design criteria for the Mo recovery and purification J&C systems are described in Section 7.2.

7.3.5.2 Design Basis and Safety Requirements The design basis and safety requirements for the Mo recovery and purification I&C systems are described in Section 7.2. The ESFs for thjs system are listed in Chapter 6.0.

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' ~ *.* ! . NOITtfWEST M£otCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.3.5.3 System Description The Mo recovery and purification I&C system wi ll be defined in the Operating License Application. The strategy and associated parameters for the l&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

Operators using remote in-cell manipulators wi ll perform the product transfer and packaging functions .

All other normal operating functions will be performed remotely using the FPC system HMI in the operating gallery. Redundant control functions will be provided in the control room. Table 7-9 lists the anticipated control parameters, monitoring parameters, and primary control locations for each subsystem.

In addition, the implementation of IROFS CS-14, CS-15, CS-20, CS-27, and RS-10 interlocks for this system are under development. Details of the control system (e.g., interlocks and permissive signals),

nuclear and process instruments, control logic and elements, indication, alarm, and control features will be developed for the Operating License Application .

Table 7-9. Molybdenum Recovery and Purification System Control and Monitoring Parameters Control parameters Subsystem name (automatic/manual) Monitoring parameters Primary control location Primary ion exchange . Temperature (A) .. Density Operating gallery Flowrate Valve actuation (AIM)

. Level

.. Temperature Pressure

. Radiation Valve position Secondary ion exchange . Pumps (M) . Temperature Operating gallery

. . Density Tertiary ion exchange Pumps (M)

.. Flowrate Operating gallery

.. Level Pressure Temperature Molybdenum product

  • Actuate capping unit (M) . Weight Operating gallery Table 7-10 provides a preliminary listing of the Mo recovery and purification system interlocks and permissive signals that have been identified. These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-10. Molybdenum Recovery and Purification System Interlocks and Permissive Signals Hard-wired or Safety Interlock or permissive input PLC Interlock Feed tank lA (MR-TK-100) high-liquid level alarm PLC NIA Feed tank IA (MR-TK- 100) low-liquid level alarm PLC NIA Feed tank lA (MR-TK-100) high-temperature alarm PLC NIA Feed tank lA (MR-TK-100) high-pressure alarm PLC NIA Feed tank lB (MR-TK-140) high-liquid level alarm PLC NIA Feed tank lB (MR-TK-140) low-liquid level alarm PLC NIA 7-38

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  • ~ ~-~ ~ : . ffOITHWHT MEDICAL ISOTOP'l:S Table 7-10. Molybdenum Recovery and Purification System Interlocks and Permissive Signals Interlock or permissive input Feed tank lB (MR-TK-140) high-temperature alarm Feed tank lB (MR-TK-140) high-pressure alarm Hard-wired or PLC PLC PLC

.. NIA NIA U solution collection tank (MR-TK-180) high-liquid level alarm PLC NIA U solution collection tank (MR-TK-180) low-liquid level alarm PLC NIA U solution collection tank (MR-TK-180) high-pressure alarm PLC NIA Waste collection tank (MR-TK-340) high-liquid level alarm PLC NIA Waste collection tank (MR-TK-340) low-liquid level alarm PLC NIA Waste collection tank (MR-TK-340) high-pressure alarm PLC NIA NIA not applicable. u uranium.

PLC = programmable logic controller.

7.3.5.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.6 Waste Handling System The waste handling system will consist of storage tanks for accumulating waste liquids and adjusting the waste composition, and the equipment needed for handling and encapsulating solid waste. Liquid waste will be split into high-dose and low-dose streams by concentration. The high-dose fraction will be further concentrated and adjusted. Liquid waste will then be mixed with an adsorbent material. The solid waste streams will be placed in a waste drum and encapsulated by adding a cement material to fill voids remaining within the drum. All high-dose waste streams will be held for decay and shipped to a disposal facility.

Due to the variety of activities performed during waste handling, the system description is divided into the following subsystems:

  • High-dose liquid waste collection
  • Low-dose liquid waste collection
  • Low-dose waste evaporation
  • High-dose liquid waste solidification
  • Low-dose liquid waste solidification
  • Spent resin dewatering
  • Solid waste encapsulation
  • High-dose waste decay
  • High-dose waste handling 7.3.6.1 Design Criteria Design criteria for the waste handling I&C systems are described in Section 7.2.

7.3.6.2 Design Basis and Safety Requirements The design basis and safety requirements for the waste handling I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0.

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~*

NORTHWESTWDICAllSOTOPES NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.3.6.3 System Description The waste handling I&C system will be defined in the Operating License Application. The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

All normal operating functions for low-dose liquid solidification will be controlled locally using FPC system HMis in the low-dose waste room (Room W 107). A local control room will be provided in this room for most waste handling operations. All normal operating functions for the high-dose liquid waste solidification, high-dose waste decay, spent resin dewatering, and solid waste handling hot cell operations will be controlled from the waste handling control room. Liquid waste collection and low-dose liquid waste evaporation operations will be controlled from the control room. Table 7-11 lists the anticipated control parameters, monitoring parameters, and primary control locations for each subsystem. In addition, the implementation of IROFS CS-14, CS-15 , CS-20, CS-27, and RS-10 interlocks for this system are under development. Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements, indication, alarm, and control features will be developed for the Operating License Application.

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  • ~ ~**!
  • NOlmfWHT MEDtCAl ISOTGm NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-11. Waste Handling System Control and Monitoring Parameters Control parameters Subsystem name (automatic/manual) Monitoring parameters Primary control location High-dose liquid waste
  • Valve position
  • Density Control room coll ection
  • Differenti al pressure
  • Flowrate
  • Flowrate totalizer
  • Level
  • Temperature
  • Pressure
  • Radiation
  • Valve position High-dose liquid waste
  • Valve position
  • Density Low dose solidification room solidification
  • Differential Pressure
  • Flowrate
  • Flowrate totalizer
  • Level
  • Temperature
  • Pressure
  • Radiation
  • Valve Position Low-dose liquid waste
  • Flowrate (A)
  • Density Control room coll ecti on
  • Pump actuation (A/M)
  • Di ffe rential pressure
  • Pump motor speed (A)
  • Fl owrate
  • Temperature (A)
  • Flowrate totalizer
  • Valve actuation (A/M)
  • Level
  • Temperature
  • Pressure
  • Valve position

--~

Low-dose liquid waste

  • Flowrate (A)
  • Differential pressure Control room evaporation
  • Pump actuation (AIM)
  • Flowrate
  • Pump motor speed (A)
  • Level
  • Temperature (A)
  • Temperature
  • Valve actuation (A/M)
  • Pressure
  • Valve position Low-dose liquid waste
  • Flowrate (A)
  • Density Low dose solidification room solidifi cation
  • Pump actuation (A/M)
  • Differential pressure
  • Pump motor speed (A)
  • Flowrate
  • Temperature (A)
  • Flowrate totalizer
  • Valve actuati on (A/M)
  • Level
  • Temperature
  • Pressure
  • Valve positi on Spent resin dewatering
  • Valve actuation (AIM)
  • Valve position Low dose solidification room Solid waste
  • Pressure Low dose solidification room encapsul ation High-dose waste decay TBD TBD Low dose solidification room High-dose waste TBD TBD Low dose solidification room handling TBD = to be detennined.

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' ~ *. ~ ! : . NOlllfWHT MEDtCAl ISOTOPf.S Table 7-12 provides a preliminary listing of the waste handling system interlocks and permissive signals that have been identified. These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-12. Waste Handling System Interlocks and Permissive Signals Hard-wired or Safety Interlock or permissive input PLC interlock High-dose waste collection tank (WH-TK-100) bjgh-liquid level alarm PLC NIA High-dose waste collection tank (WH-TK-100) low-liquid level alarm PLC NIA High-dose waste collection tank (WH-TK-100) low-pressure alarm PLC NIA High-dose waste concentrator (WH-Z-200) high-liquid level alarm PLC NIA High-dose waste concentrator (WH-Z-200) low-liquid level alarm PLC NIA High-dose waste concentrator (WH-Z-200) demister high-differential pressure PLC NIA alarm High-dose waste concentrator (WH-Z-200) condenser mgh-differential pressure PLC NIA alarm High-dose waste concentrator (WH-Z-200) condenser offgas high-temperature PLC NIA alarm Low-dose waste collection tank (WH-TK-240) mgh-liquid level alarm PLC NIA Low-dose waste collection tank (WH-TK-240) low-liquid level alarm PLC NIA Low-dose waste collection tank (WH-TK-240) low-pressure alarm PLC NIA High-dose waste container offgas filter (WH-F-330) high-pressure differential PLC NIA alarm Condensate collection tank (WH-TK-400) high-liquid level alarm PLC NIA Condensate collection tank (WH-TK-400) low-liquid level alarm PLC NIA Condensate collection tank (WH-TK-400) low-pressure alarm PLC NIA Low-dose waste collection tank (WH-TK-420) high-liquid level alarm PLC NIA Low-dose waste collection tank (WH-TK-420) low-liquid level alarm PLC NIA Low-dose waste collection tank (WH-TK-420) low-pressure alarm PLC NIA Low-dose waste evaporation tank 1 (WH-TK-500) high-liquid level alarm PLC NIA Low-dose waste evaporation tank 1 (WH-TK-500) low-liquid level alarm PLC NIA Low-dose waste evaporation tank 1 (WH-TK-500) low-pressure alarm PLC NIA Low-dose waste evaporation tank 2 (WH-TK-530) high-liquid level alarm PLC NIA Low-dose waste evaporation tank 2 (WH-TK-530) low-liquid level alarm PLC NIA Low-dose waste evaporation tank 2 (WH-TK-530) low-pressure alarm PLC NIA Low-dose waste container offgas filter (WH-F-630) high-pressure differential PLC NIA alarm PLC = programmable logic controller. TBD to be determined.

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' ~*-- ~

  • NORTIIWlSTMEDtcAllSOTOrfl 7.3.6.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.7 Criticality Accident Alarm System The RPF will use a CAAS to monitor for a criticality and provide emergency notifications for evacuation.

7.3.7.1 Design Criteria Design criteria for the CAAS I&C systems are described in Section 7.2.

7.3.7.2 Design Basis and Safety Requirements The design basis and safety requirements for the CAAS I&C systems are described in Section 7 .2.

7.3.7.3 System Description The CAAS will be provided as a vendor package with an integrated control system. The CAAS control HMI will be located in the control room and will provide local alarms at the detector locations and at the CAAS HMI. The FPC system will provide alarm and status monitoring in the control room. The facility-wide notification system configuration will be provided in the Operating License Application.

The surveillance requirements for the CAAS system are described in Chapter 6.0.

7.3. 7.4 System Performance Analysis and Conclusion The system performance analysis for each process system will be provided in the Operating License Application. The overall I&C system performance analysis is discussed in Section 7.2.

The CAAS will provide for continuous monitoring, indication, and recording of neutron or gamma radiation levels in areas where personnel may be present and wherever an accidental criticality event could result from operational processes. The CAAS will be capable of detecting a criticality accident that produces an absorbed dose in soft tissue of 20 radiation absorbed dose (rad) of combined neutron or gamma radiation at an unshielded distance of 2 meters (m) from the reacting material within 1 minute (min), except for events occurring in areas not normally accessed by personnel and where shielding provides protection against radiation generated from an accidental criticality. Two detectors will cover each area needing CAAS coverage.

The control unit electronjcs will actuate local and remote alarms. The locations of the detectors will be provided in the Operating License Application.

The CAAS detectors will provide local annunciation and remote annunciation in the control room to alarm when the radiation levels exceed established setpoints. Alarming CAAS monitors will communicate the location of the criticality accident alarm to the FPC system. Diagrams of the CAAS and associated systems will be provided in the Operating License Application.

The uninterruptible power supply (UPS) will provide emergency power to the CAAS during a loss of off-site power. The CAAS will meet the criteria of 10 CFR 20.1501, "General," and use the guidance provided by ANSI/ANS 8.3 , Criticality Accident Alarm System, and Regulatory Guide 3.71 , Nuclear Criticality Safety Standards for Fuels and Material Facilities. As a safety-related system, the CAAS will be designed to remain operational during design basis accidents, which are described in Chapter 13.0.

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' ~ *.*! . NOITNWEST MEOM:Al lSOTOP£1 NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS 7.4.1 System Description The ESFs are active or passive features designed to mitigate the consequences of accidents and to keep radiological exposures to workers, the public, and environment within acceptable values. Chapter 6.0 provides a description of the ESFs, including the accidents mitigated and SSCs used to provide the ESFs.

The ESF systems will operate independently from the FPC systems as hard-wired controls. However, the ESFs will integrate into the FPC systems and provide a common point of HMI, monitoring, and alarming at the control room and local HMI workstations.

Table 7-13 lists the ESFs that will require actuation by the I&C system. Monitoring systems that are credited in the safety anal ysis are also included in the table.

Table 7-13. Engineered Safety Feature Actuation or Monitoring Systems (2 pages) l&C SSCs providing Engineered safety feature IROFS Accident(s) mitigated engineered safety feature Primary offgas reli ef system RS-09 Dissolver offgas failure during Pressure relief devi ce, pressure di ssoluti on operation relief tank Active radiation monitoring RS-10 Transfer of high-dose process Radiation monitoring and and isolation of low-dose liquid outside the hot cell isolation system for low-dose waste transfer shielding boundary liquid transfers Cask local ventil ation during RS-1 3 Target cladding leakage during Local capture ventilation closure lid removal and shipment system over closure lid during docking preparations lid removal Cask docking port enabler RS-15 Cask not engaged in the cask Sensor system controlling cask docking port prior to opening the docking port door operation docking port door Process vessel emergency FS-03 Hydrogen defl agration or Backup bottled nitrogen gas purge system detonation suppl y Active discharge monitoring CS-14 Accidental criticality To be provided in the Operating and isolation License Application Independent active di scharge CS-15 Accidental criticality To be provided in the Operating monitoring and isolation License Application Evaporator or concentrator CS-20 Prevent nuclear criticality from Conductivity analyzer and condensate monitoring high-volume transfer to non- control valve geometrically favorable vessels in solutions with normally low fissile component concentrations 7-44

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' ~ *.* ~

  • NWM I NOllTHWEST MEDtCAl ISOTO'U NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Table 7-13. Engineered Safety Feature Actuation or Monitoring Systems (2 pages) l&C SSCs providing Engineered safety feature IROFS Accident(s) mitigated engineered safety feature Closed heating or cooling CS-27 Accidental criticality Closed-loop, hi gh-volume heat loop with monitoring and transfer fluid systems to prevent alarm nuclear criticality or transfer of high-dose material across shielding boundary in the event of a leak into the heat transfer fluid with normally low fissile component concentrations Dissolver offgas vacuum TBD Potential limiting control for Dissolver offgas vacuum receiver or vacuum pump operations; motive force for receiver tanks, dissolver offgas dissolver offgas vacuum pumps l&C instrumentation and control. SSC structures, systems, and components.

IROFS items relied on for safety . TBD to be determin ed.

7.4.2 Annunciation and Display The actuation of an ESF will be displayed on the FPC system HMI and locally at the affected system with an audible alarm. The alarm annunciator display panel and the alarm or event display will show the triggering event. Once actuated, the ESFs will require manual input from the operator to reset the ESF.

Clearing the triggering event wi ll be required.

7.4.3 System Performance Analysis Section 7 .2.4 provides additional details on the analysis of system performance. Potential variables, conditions, or other items that will be probable subjects of technical specifications associated with the FPC system are provided in Chapter 14.0.

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NORTHWEST MEDICAl &SOTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 - Instrumentation and Control Systems 7.5 CONTROL CONSOLE AND DISPLAY INSTRUMENTS 7.5.1 Design Criteria Design criteria for the control room I&C systems are described in Section 7.2.

7.5.2 Design Basis and Safety Requirements The design basis and safety requirements for the control room I&C systems are described in Section 7.2.

7.5.3 System Description The control room will provide the majority of interfaces for the facility and process control systems, with overall process controls, monitoring, alarms, and acknowledgement. The control room will consist of a properly sized and shaped control console with two or three operator interface stations or HMis (one being a dedicated engineering interface), a master PLC or distributed controller, and all related and necessary cabinetry and subcomponents (e.g., input/output boards, gateways, Ethernet switches, power supplies, and UPS). This control system will be supported by a data highway of sensing instrument signals in the facility process areas that will be gathered onto the highway throughout the facility by an Ethernet communication-based interface backbone and brought into the control room and onto the console displays.

Dedicated controllers and human-machine monitoring interfaces or stations for other equipment systems will also be in the control room. This equipment includes the facility crane, closed-circuit television system, CAAS, and radiation monitoring system. A control panel for all facility on-site and off-site (if required) communications (e.g., telephone, intercom) will likely also be located there. The control room door into the facility will be equipped with controlled access.

The BMS will be primarily controlled and monitored from the control room. Utility systems with vendor packages and integrated controls will provide surveillance monitoring to the control room.

The FPC system will operate with a synchronized hot standby redundant system structure for all hot cell processes. Each hot cell process will be an independent subsystem having a local HMI with monitoring and control functions from the control room. Workstations for each system within the control room will be hot standby redundant. The redundant stations will run software on identical PLC systems. The PLC systems will monitor each other. On loss of synchronizing signal from one system, the other system will continue with control and monitoring.

Process systems that will be primarily controlled in the control room include uranium recovery and recycle, target dissolution, and liquid waste handling. The target receipt system will be controlled with local HMis in the irradiated target basket receipt bay or target cask preparation airlock. Mo production process hot cell systems, including target disassembly and Mo recovery and purification, will be controlled with local HMis in the hot cell operating gallery. The hot cell processes will have monitoring and redundant control functions from the control room.

The FPC subsystem for target fabrication processes will be controlled with local HMis in the target fabrication area, with surveillance monitoring in the control room.

Local HMis will be provided in Room W107 , which houses equipment for low-dose waste solidification.

Low-dose liquid waste will be piped in from the holding tanks in the utility area above Room W107, and drums of solidified waste will be transported out by pallet jack. This local HMI will be the primary control location for the high-dose liquid waste solidification, high-dose waste decay, spent resin dewatering, and solid waste handling hot cell operations.

7.5.4 System Performance Analysis and Conclusion The system performance analysis for each process system will be provided in the Operating License Application. The overall I&C system performance analysis and conclusions are provided in Section 7.2.

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' ~ ~.~ ~ : NORTifWESTMEDtcAllSOTOP£S 7.6 RADIATION MONITORING SYSTEMS The radiation monitoring systems will include CAMs, continuous monitoring at the exhaust stacks, process control instruments, and personnel monitoring and dosimetry. Process control instruments used to analyze for uranium concentrations are described in each respective process system in Section 7.3.

The objective of the radiation monitoring system is to provide the RPF control room personnel with a continuous record and indication of radiation levels at selected locations where radioactive materials may be present, stored, handled, or inadvertently introduced. The system is also designed to ensure that there is accurate and reliable information concerning radiation safety as related to personnel safety. The design considerations for the radiation monitoring system include the following:

  • Provision of information to RPF operators so that in the event of an accident resulting in a release of radioactive material, decisions on deployment of personnel can be properly made.
  • Indication and recording in the control room of the gamma and airborne radiation levels in selected areas as a function of time, and, if necessary, alarming to indicate any abnormal radiation condition. These indicators aid in maintaining plant contamination levels as low as reasonably achievable (ALARA) and in minimizing personnel exposure to radiation.
  • Provision oflocal alarms and/or indicators positioned at key points throughout the RPF where a substantial increase in radiation levels might be of immediate importance to personnel frequenting or working in the area.

Radiation Monitoring Locations RAMs will be located in areas where personnel may be present and where radiation levels could become significant based on the following considerations:

  • Occupancy status of the area, including time requirements of personnel in the area, the proximity to primary and secondary radioactive sources, and shielding
  • Potential for increase in the background radioactivity level
  • Desirability of surveillance of infrequently visited areas CAMs will be located in work areas where there is a potential for airborne radioactivity. The CAMs will have the capability to detect derived air concentrations within a specified time.

7.6.1 Design Criteria Design criteria for the radiation monitoring I&C systems are described in Section 7.2 .

7.6.2 Design Basis and Safety Requirements The design basis and safety requirements for the radiation monitoring I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0.

7.6.3 System Description The radiation safety monitoring system will include CAMs, continuous monitors at the exhaust stacks, and personnel monitoring and dosimetry.

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. * ! ~.~! : . NO<<rHW£STMEOfCALISOTOPES NWMl-2013-021 , Rev. 1 Chapter 7.0 - Instrumentation and Control Systems Three basic types of personnel monitoring equipment will be used at the facility: count rate meters (friskers), hand/foot monitors, and portal monitors. All personnel whose duties require entry to restricted areas will wear individual external dosimetry devices (e.g., passive dosimeters such as thermoluminescent dosimeters that are sensitive to beta, gamma, and neutron radiation) from a National Voluntary Laboratory Accreditation (NAVLAP)-certified vendor. Personnel monitoring and dosimetry is described in Chapter 11.0, "Radiation Program and Waste Management."

7.6.3.1 Air Monitoring Continuous air monitors - CAM units will consist of a particulate measuring channel with a filter to capture particulate. Air will be drawn through the system by a pump assembly. The sample will be withdrawn from inside the appropriate area, room, or cell through an isokinetic nozzle with the sampling volume flow at a known fixed rate, so that the accumulation of radioactive particles can be interpreted as a quantitative sample. After passing through the nozzle, the sample will be drawn through tubing and through a fixed or moving filter tape before being discharged to the atmosphere. The samplers also have a purging system for flushing the volume cell surrounding the gas sample chamber with clean air for purposes of calibration and the removal of crust activity. Replaceable liners will be changed out periodically when contamination becomes excessive. Flow regulating will ensure that flow through the filters remains constant.

Each instrument channel will include a detector, preamplifier, count rate meter, and power supply. The detector may be a scintillation counter or similar device having a gamma sensitive crystal , and a photo multiplier whose output pulses are counted by the rate meter. Each readout module will be equipped with a light that illuminates when the radiation level exceeds preset limits. The setpoint will be adjustable over the entire detection range. Pressing a button will cause the meter to indicate the alarm setpoint. Visible alarms will be accompanied by a simultaneous local audible alarm with an alarm light in the control room. A normally energized light will deenergize when there is a detector signal failure , circuit failure ,

power failure , or failure due to a disconnected cable. Power for the monitors that initiates a safety signal will be provided from the UPS . Loss of power and signal failure will be monitored for each detector.

CAMs will be provided with a check source. This check source will simulate a radiation field and will be used as a convenient operational and gross calibration check of the detectors and readout equipment.

CAM calibration will include, where practical, exposures to the specific isotopes that the particular system monitors in the field. Instrument calibrations will be performed at prescribed frequencies . An electronic test signal and/or radioactive check source drift indication may also require CAM recalibration.

Radiation area monitors - The RAM detector unit will be housed in an environmentally suitable container that is mounted in a duct, on a wall, or other suitable surface. The sensitivity of each detector will be sufficient to have the alarm setpoint an order of magnitude higher than the detection threshold.

The detectors are designed to be operational over a wide range of temperatures. The design of the detectors will meet expected normal and abnormal environmental design conditions, as appropriate.

Saturation will not be expected to adversely affect operation of the detector within its calibrated range.

Sensors will be mounted as close as practical to the most probable radiation sources with no objects, persons, pillars, and piping serving as shielding. The sensors will also be mounted so as to minimize inaccuracies due to any directionality of the detector.

Audible and visual alarm devices - When the radiation exceeds predetermined levels, alarms will actuate in the control room and at selected detector locations.

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  • NORTHWEST MEDICAL ISOTOPES The alarms will consist of the following capabilities :
  • "Alert light" will illuminate when the radiation level exceeds preset limits with an adjustable setpoint
  • "High alarm red light" will illuminate when radiation levels exceed a predetermined alarm setpoint
  • "Failure alarm" will sound when either the power or the channel's electronics fail The visual alarms will be accompanied by a simultaneous audible alarm annunciator at the selected detector locations and in the control room. The annunciator windows for the monitors will be located in the control room. The alarm can be manually reset when the alarm conditions are corrected. The local alarm horns and warning lights will remain on until the radiation leve l is below the present level.

Additional CAM requirements and locations are described in Chapter 11 .0.

7.6.3.2 Stack Release Monitoring The exhaust stacks will be provided with continuous monitors for noble gases, particulate, and iodine.

The stack monitoring system design basis is to continuously monitor the radioactive stack releases.

Additional information will be provided in the Operating License Application. Airborne exposure pathway monitoring is described in Chapter 11 .0.

7.6.4 System Performance Analysis and Conclusions The system performance analysis and conclusions for each process system will be provided in the Operating License Application. The overall I&C system performance analysis is provided in Section 7.2.

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7.7 REFERENCES

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I 0 CFR 70.64, "Requirements for New Facilities or New Processes at Existing Facilities," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 73.1, "Purpose and Scope," Code of Federal Regulations, Office of the Federal Register, as amended.

I 0 CFR 73.54, "Protection of Digital Computer and Communication Systems and Networks," Code of Federal Regulations, Office of the Federal Register, as amended.

ANS 10.4-2008, Verification and Validation of Non-Safety-Related Scientific and Engineering Computer Programs for the Nuclear Industry, American National Standards Institute, New York, New York, 2008.

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  • NOlllfWESTMEDtcAllSOTOitES Chapter 7.0 - Instrumentation and Control Systems IEEE 338-2012, IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2012 .

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IEEE 577-2012, IEEE Standard Requirements for Reliability Analysis in the Design and Operation of Safety Systems for Nuclear Facilities , Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2012.

IEEE 603-2009, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations ,

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IEEE 828-2012, IEEE Standard for Configuration Management in Systems and Software Engineering, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2012.

IEEE 829-2008, IEEE Standard for Software and System Test Documentation , Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2008.

IEEE 1012-2012, IEEE Standard for System and Software Verification and Validation , Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2012.

IEEE 1028-2008, IEEE Standard for Software Reviews and Audits, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2008.

IEEE STD 12207, ISOIIECIIEEE Standard for Systems and Software Engineering - Software Life Cycle Processes, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2008 .

IEEE STD 15939, IEEE Standard Adoption of ISOllEC 15939:2007 Systems and Software Engineering Measurement Process, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2008.

ISA-RP-67 .04.02, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation, Instrument Society of America, Research Triangle Park, North Carolina, 2010.

ISO/IEC/IEEE 15288, Systems and Software Engineering - System Life Cycle Processes, International Organization for Standardization, Geneva, Switzerland, 2015.

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NUREG-0700, Human-System Interface Design Review Guidelines, Rev. 2, U.S . Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C. , May 2002.

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  • ! ~.~~ ; NORTHWUTMEDtw.ISOTOf'CS NUREG-0711, Human Factors Engineering Program Review Model, Rev. 3, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C.,

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NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C. , as amended.

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NUREG/CR-6090, The Programmable Logic Controller and Its Application in Nuclear Reactor Systems ,

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NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems, U.S. Nuclear Regulatory Commission, Washington, D.C., June 1996.

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