ML17221A201

Attachment 3 to NWMI-2013-021, Rev. 2, Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility to Chapter 11.0 - Radiation Protection and Waste Management

ML17221A201
Person / Time
Site:	Northwest Medical Isotopes
Issue date:	08/05/2017
From:	Haass C Northwest Medical Isotopes
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17221A370	List: ML17221A199 ML17221A200 ML17221A201 ML17221A202 ML17221A203
References
NWMI-LTR-2017-011 NWMI-2013-021, Rev. 2
Download: ML17221A201 (236)

Text

• NORTHWEST MEDICAL ISOTOPES Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 2 August 2017 Prepared by:

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

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...*.. NWMl-2013-021 , Rev . 2 Chapter 6.0 - Engineered Safety Features

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• NORTHWEST MEDICAL ISOTOPES Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 2 Date Published:

August 5, 2017 Document Number. NWMl-2013-021 Revision Number. 2

Title:

Chapter 6.0 - Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Si nature: CUAJ"rr'-- (_ -j/~

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NWMl-2013-021 , Rev . 2 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 2 8/5/2017 Modification based on ACRS comments C. Haass

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~. NWMl-2013-021 , Rev . 2 Chapter 6.0 - Engineered Safety Features

' ~* * ~ NORllfWHT M£DfCA.L ISOTOP£S 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 6.2.1.2 Accidents Mitigated .......... .................... .. ... .... .. ....... ..... .. ................ 6-11 6.2.1.3 Functional Requirements ........ ...... ...................... .......... ..... ........ ..... 6-11 6.2.1.4 Confinement Components ....... ................. .. .... ..... .... ....... ................ 6-11 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. 1.8 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. l 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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' ! * *! ' NORTHWEST MlOtCAJ.. ISOTOPES FIGURES Figure 6-1. Simplified Zone I Ventilation Schematic .............................. .... ..... ... ... ..... ................... 6-6 Figure 6-2. Ground Level Confinement Boundary ........................... ... ... ... ..................................... 6-8 Figure 6-3. Mecham cal 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) ............................ 6-51 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

NWMl-2013-021, Rev. 2 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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• NCMmfWEST MEDfCAl ISOTOPES NWMl-2013-021 , Rev. 2 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 in. inch L liter m meter m2 square meter mm minute mL milliliter mo) mole rad radiation absorbed dose wt% weight percent yr year 6-iv

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~* * ~ NORTHWEST MEOICAl lSOTOf'£S 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 of features 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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' ~ *.* ~ NORTHWHTMEIHCAllSOTOPtS NWMl-2013-021, Rev. 2 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 including penetration seals through

• Hot cell liquid RS-01 maintenance Zone I exhaust ventilation 6.2.1.6 confinement

• Hazardous chemical system, including ducting, boundary spi lls filters, and exhaust stack

• Hot cell RS-03

• Zone I inlet ventilation system, secondary including ducting, filters, and confinement bubble-tight isolation dampers boundary

• Ventilation control system

• Hot cell shielding RS-04

• Secondary iodine removal bed boundary

• Berms Confinement IROFS Derived from Accident Analyses and Potential Technical Specifications Primary offgas relief system RS-09 Dissolver offgas failure during dissolution . Pressure relief device Pressure relief tank 6.2.1. 7.1 operation Active radiation RS-10 Transfer ofhigh-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-13 Target cladding leakage Local capture ventilation system 6.2.1.7.3 ventilation 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 deflagration 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 prevents cask tipping workers present during

• Cask lifting fixture design that target unloading prevents lift from toppling activities during a seismic event 6-2

NWMl-2013-021 , Rev . 2 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 Exhaust stack height FS-05 . Equipment . Zone I exhaust stack 6.2.1.7.7 malfunction resulting

. in liquid spi ll 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 prevention devices from backflow of high- located on process lin es crossing Safe geometry day CS-1 9 dose solution the hot cell shielding 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 Dissolver offgas . Potential limiting Dissolver offgas primary adsorber 6.2.1.8.2 primary adsorber*

. control for operation Primary 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-71 OA/B)

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

lROFS 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 fixtures 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 ofSSCs using geometry to 6.3.1.2.1 6.3. l .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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• NOfOlfWUT M£D.cAl ISOTOl'(S NWMl-2013-021, Rev. 2 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 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 Double-wall piping 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 from 6.3.1.2.7 safe-geometry tank with level alarm fissile solution entering non-geometrically favorable ventilation 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 Simple overflow to normally empty CS-13 Overflow to prevent nuclear criticality from 6.3.1.2.9 safe geometry floor with level alarm fissile solution entering non-geometrically in the hot cell containment boundary favorable ventilation equipment Active discharge monitoring and CS-14 Information to be provided in the Operating 6.3.1.2.10 isolation License Application Independent active discharge CS-15 Information will be provided in the Operating 6.3.1.2.11 monitoring and isolation License Application 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-19 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 criticality 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 for safety. SSC = structures, systems, and components.

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~* * ~ . NORTKW£$T M£DtCAl ISOTOPES 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, venti lation, and air conditioning (HVAC) systems, including exhaust stacks or vents to the external environment, filters , blowers, and dampers (ANSI/ ANS -1 5.1, 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 venti lation 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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[Proprietary Information]

Source: Figure 2-5 ofNWMl-2015-SDD-O13, 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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• NOllTifWEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features A typical glovebox enclosure is shown in Figure 6-1, and the inlet does not have an automatic closure on the isolation damper. During development of the final safety analysis and Operating License Application, each glovebox will be evaluated based on inventory of concern (e.g., fission product gases) and hazards to determine if the inlet isolation damper is required to be an IROFS confinement control. Until the analysis is complete, the design of gloveboxes will include a bubble-tight isolation damper, as required, for the hot cells.

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 confinement 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 filter 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 relied on for safety.

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[Proprietary Information]

Source: Figure 2-1 ofNWMI-2015-SDD-O 13, 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 of NWMl-20 15-SDD-O 13, System Design Description for Ventilation, Rev. A, Northwest Medi cal 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-20 I 5-SDD-01 3, 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. Tills 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 ofreleasing 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.

Tills 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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~* * ~ NOflTHWEST MHHCAl ISOTDPU 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 13.1.3.1. 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 Jong-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 10, 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 avai lable 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 reliability 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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• ~ * * ~* NORTHWlST MEDICAL ISOTOPES NW Ml-2013-021, Rev. 2 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 II.

• 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- I, 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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• ~* * ~ NORTHWEST MEDICAL ISOTOPES 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 ofless than [Proprietary Information]

• Nominal sorbent bed depth of [Proprietary Information]

• Nominal gas relative humidity Jess 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, ifthe offgas motive force (vacuum pumps) ceases operation during dissolution of a dissolver batch.

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NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

• ~ * .*.~ ." NORTHW£ST MEDfCAl ISOTOPES 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

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~* * ~ NORTHWEST MEDICAL lSOTWU

[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. , Joss of vacuum pump operation) will result in closing Valve I 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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...... NWMl-2013-021, Rev . 2 Chapter 6.0 - Engineered Safety Features

. ~ * *! NORTifWEST MEDtCAL ISOTOPES 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 ofN02 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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!* * ~ NORTHWEST MEDICAL lSOTOPES NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features 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 \ow-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 oflow-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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• NOITlfW£ST MEDICAL ISOTOf'ES NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features 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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NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features 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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NWMl-2013-021, Rev. 2 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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NWMl-2013-021, Rev. 2 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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NWMl-2013-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features 6.2.1.7.9 IROFS CS-18, Back.flow 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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NWMl-2013-021 , Rev. 2 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 fission 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 20.1201, "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], providing iodine removal capacity of greater than 1 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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• NORTifWUT MEDtCAL ISOTOP£S NWMl-2013-021, Rev. 2 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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NWMl-2013-021, Rev. 2 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.

6.2.1.8.3 Dissolver Offgas Vacuum ReceiverNacuum 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

NWMl-2013-021, Rev . 2 Chapter 6.0 - Engineered Safety Features 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 9.1.2. 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 N13.1, Sampling and Monitoring Releases ofAirborne Radioactive Substances from the Stacks and Ducts ofNuclear 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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~* * ~ NORTHWEST MEOICAl ISOTOPES 6.2.2 Containment Containment for the RPF is defined based on NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-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 fuel 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-1537 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-153 7 Part 1 interim staff guidance.

Jn 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 system, or the irradiated fuel 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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~ * * ~ ." NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev . 2 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-8.3 , Criticality Accident Alarm System

• ANSI/ ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materials

• ANSI/ ANS-8.10, Criteria for Nuclear Criticality Safety Controls in Operations With Shielding and Corifinement

• ANSI/ ANS-8.19, Administrative Practices for Nuclear Criticality Safety

• ANSI/ ANS-8 .20, Nuclear Criticality Safety Training

• ANSI/ ANS-8.22, Nuclear Criticality Safety Based on Limiting and Controlling Moderators

• ANSI/ ANS-8.23 , Nuclear Criticality Accident Emergency Planning and Response

• ANSI/ ANS-8 .24, Validation ofNeutron Transport Methods for Nuclear Criticality Safety Calculations

• ANSI/ 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-8.1 - Nuclear criticality safety practices, including administrative practices, technical practices, and validation of a calculational method 6-28

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• ANSl/ANS-8.3 - Criticality accident alarm system (CAAS) placement analysis and procedure development; the standard is used as modified by NRC Regulatory Guide 3.71

• ANSI/ ANS-8.19 - NWMI nuclear criticality safety program development as it applies to organization, administration, roles, and responsibilities

• ANSI/ ANS-8.20 - Nuclear criticality safety staff and contractor qualification and training procedure development

• ANSI/ ANS-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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~* *~* NORTHWHT MEDCAL ISOTOPfS Chapter 6.0 - Engineered Safety Features

• 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 .

• CS Es 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-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-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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...;.....;. NWMI

...~**=***

• • *

• NORTHWtST MEDICAL ISOTOPES NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features The USL includes the method bias and uncertainty established in [Proprietary Information] and a 0.05 &

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 Training 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 abilities 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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~

.... NWMI

• *! ' NOftTifWUT MEDICAL tsOTOPES NWMl-2013-021, Rev . 2 Chapter 6.0 - Engineered Safety Features 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 . l 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 CSEs 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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• • • • • NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

~* *~ HORTlfWUT MEDICAl lSOTOf'ES

• 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 ANSI/ ANS-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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~. NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

~ *.*! ' NOlmfWEST MlDtCAL ISOTOPlS

• 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 ANSI/ ANS-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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NWMl-2013-021 , Rev . 2 Chapter 6.0 - Engineered Safety Features 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 mi ght be encountered.

• Procedures wiJJ :

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 wiJJ be readily available to personnel responding to an emergency and a method provided for prompt on-site dosimeter readouts.

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NWMl-2013-021, Rev. 2 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 CS Es 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 13.0.

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-CRlTCALC-001 , Single Parameter Subcritical Limits for 20 wt% 235 U- Uranium Metal, Uranium Oxide, and Homogenous Water Mixtures

• NWMI-2015-CRlTCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution

• NWMI-2015-CRlTCALC-003 , 55-Gallon Drum Arrays

• NWMI-2015-CRlTCALC-005, Target Fabrication Tanks, Wet Processes, and Storage

• NWMI-2015-CRlTCALC-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 ofkeff that 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 ANSl/ANS-8.1.

Equation 6-2 where keff is the MCNP calculated k-effective and CJcaJ c is the MCNP calculation uncertainty.

[Proprietary Information]

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~ * *! . NORTHWEST MEDICAL lSCJTDfl£S

[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]

35 H/2 U ratio [Proprietary Information]

Average neutron energy causing fission [Proprietary Information]

Enrichment [Proprietary Information]

Moderating materials [Proprietary Information]

Reflecting materials [Proprietary Information]

Absorber materials [Proprietary In formation]

Geometry [Proprietary Information]

a Source: [Proprietary Information].

ANECF = average neutron energy causing fission.

The RPF was divided into 13 activity groups for development of preliminary CSEs 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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~ * * ~ .' Nl>f'THWEST MEDtCAl. ISOTOPES NWMl-201 3-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features Table 6-5. Controlled Nuclear Criticality Safety Par ameters Nuclear NWMI criticality safety evaluation (NWMl-2015-CSP) 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 Enrichmentr N N N N N N N N N N N N N a Derived from the indicated CSE reference document.

b Limited by nature of process in the air filtration.

c Limited by target design.

ct Controlled through input fissile mass.

e Limited by total uran ium mass a llowed in the system .

r Facility license limited to ::;20 wt% mu.

mu uranium-235 . NWM I Northwest Medical Isotopes, LLC.

CSE = criticality safety evaluat ion. y yes.

N = no.

The preliminary CS Es 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-20 15-CSE-Ol, Irradiated Target Handling and Disassembly (Table 6-6)

• NWMI-20 15-CSE-02, Irradiated Low-Enriched Uranium Target Material Dissolution (Table 6-7)

• NWMl-20 15-CSE-03, Molybdenum-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-2015-CSE-06, Target Finishing (Table 6-9)

• NWMI-2015 -CSE-07, Target and Can Storage and Carts (Table 6-9)

• NWMI-2015-CSE-08, Hot Cell Uranium Purification (Table 6-10)

• NWMI-20 15-CSE-09, Waste Liquid Processing (Table 6- 11)

• NWMI-20 15-CSE-10, Solid Waste Collection, Encapsulation, and Staging (Table 6-11)

• NWMI-20 15-CSE-l l , Offgas and Ventilation (Table 6-12)

• NWMI-201 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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~ *.* ~ NOmfWUT MEDtCAL ISOTOP£S NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

• NWMI-20 l 5-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 ldentifiera Feature description and basis CSE-01 -PDFI [Proprietary Information]

CSE-0 1-PDF2 [Proprietary Information]

CSE-01-PDF3 [Proprietary Information]

CSE-01-ACl [Proprietary Information]

CSE-Ol-AC2 [Proprietary Information]

CSE-Ol-AC3 [Proprietary Information]

CSE-01 -AC4 [Proprietary Information]

a [Proprietary Information].

HEPA = high-efficiency particulate air. SPL single parameter limit.

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' ~ * *! ' NOflTHWUT IHDtcAl ISOTOPCS Table 6-7. [Proprietary Information]

Double-Contingency Controls (2 pages)

Identifier" Feature description and basis CSE-02-PDFl [Proprietary Information]

CSE-02-PDF2 [Proprietary Information]

CSE-02-PDF3 [Proprietary Information]

CSE-02-PDF4 [Proprietary Information]

CSE-02-PDFS [Proprietary Information]

CSE-02-PDF6 [Proprietary Information]

CSE-02-PDF7 [Proprietary Information]

CSE-02-PDF8 [Proprietary Information]

CSE-02-AEFl [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. 2

~* NO<<llfWUT MEDICAL ISOTOPES Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

Table 6-8. [Proprietary Information] Double-Contingency Controls (2 pages) ldentifier3 Feature description and basis CSE-03-PDFl [Proprietary Information]

CSE-03-PDF2 [Proprietary Information]

CSE-03-PDF3 [Proprietary Information]

CSE-03-PDF4 [Proprietary Information]

CSE-03 -PDF5 [Proprietary Information]

CSE-03-PDF6 [Proprietary Information]

CSE-03 -PDF7 [Proprietary Information]

CSE-03-PDF8 [Proprietary Information]

CSE-03-PDF9 [Proprietary Information]

CSE-03-PDFlO [Proprietary Information]

CSE-03-PDFl I [Proprietary Information]

CSE-03-PDF12 [Proprietary Information]

CSE-03-AEFI [Proprietary Information]

CSE-03-ACI [Proprietary Information]

a [Proprietary Information].

IX ion exchange. [Proprietary Information] [Proprietary Information] .

Mo = molybdenum .

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. ' ~ * .* ~ . NOmfWUT MEDM:Al ISOTOP(S NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

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' ~* *~ .

NORTHWEST MEOtCAl ISOTOPES NWM l-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features Table 6-9. [Proprietary Information] Double-Contingency Controls (8 pages)

Identifier Feature description and basis CSE-04-PDFl * [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 Information]

CSE-04-PDF8" [Proprietary Information]

CSE-04-PDF9* [Proprietary Information]

CSE [Proprietary Information]

PDFlO" CSE [Proprietary Information]

PDFll" CSE [Proprietary Information]

PDF12" CSE [Proprietary Information]

PDF13" CSE [Proprietary Information]

PDF14" CSE [Proprietary Information]

PDF15*

CSE [Proprietary Information]

PDF16" CSE-04-AEFl * [Proprietary Information]

CSE-04-AC 1* [Proprietary Information]

CSE-04-AC2* [Proprietary Information]

CSE-04-AC3" [Proprietary Information]

CSE-04-AC4" [Proprietary Information]

CSE-04-AC5" [Proprietary Information]

CSE-04-AC6* [Proprietary Information]

CSE-04-AC7" [Proprietary Information]

CSE-05-PDFl b [Proprietary Information]

CSE-05-PDF2b [Proprietary Information]

CSE-05-PDF3b [Proprietary Information]

CSE-05-PDF4b [Proprietary Information]

CSE-05-PDFSb [Proprietary Information]

CSE-05-PDF6b [Proprietary Information]

CSE-05-PDF7 b [Proprietary Information]

CSE-05-PDF8b [Proprietary Information]

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.~ ..**;. NWMI NWMl-201 3-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features

• ~ *.*! . NORTHWtST MEDICAL ISOTOPfS Table 6-9. [Proprietary I nformation] Double-Contingency Controls (8 pages)

Identifier Feature description and basis CSE-05 -AEFJ b [Proprietary Information]

CSE-05-AEF2b [Proprietary Information]

CSE-05 -AEF3b [Proprietary Information]

CSE-05-AClb [Proprietary Information]

CSE-05-AC2b [Proprietary Information]

CSE-05-AC3b [Proprietary Information]

CSE-06-PDFl c [Proprietary Information]

CSE-06-PDF2c [Proprietary Information]

CSE-06-AC l 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-PDF Id [Proprietary Information]

CSE-07-PDF2d [Proprietary Information]

CSE-07-PDF3d [Proprietary Information]

CSE-07-PDF4d [Proprietary Information]

CSE-07-ACJd [Proprietary Information]

CSE-07-AC2d [Proprietary Information]

CSE-07-AC3d [Proprietary Information]

CSE-07-AC4d [Proprietary Information]

CSE-07-AC5d [Proprietary Information]

CSE-07-AC6d [Proprietary Information]

CSE-07-AC7d [Proprietary Information]

a [Proprietary lnformation]

b [Proprietary lnformation]

c [Proprietary lnformation]

d [Proprietary lnformation]

ADUN acid-deficient uranium nitrate. UN = uranium nitride.

DBE design basis earthquake. [Proprietary lnformation] [Proprietary lnformation]

u uranium.

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NWMl-2013-021 , Rev . 2 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

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[Proprietary Information]

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[Proprietary Information]

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.~ . .;.NWMI NWMl-2013-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features

• ~ * .* ~ ." NORTHWEST MEDICAL ISOTOPES

[Proprietary Information]

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NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

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~* *~

.. NOflTHWUT MEtMCAl ISOTOPH NWMl-2013-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

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' ~* * ~

NORTHWEST MEOtCAl. lSOTDH.S NWMl-2013-021, Rev . 2 Chapter 6.0 - Engineered Safety Features Table 6-10. [Proprietary Information] Double-Contingency Controls (2 pages) ldentifiera Feature description and basis CSE-08-PDFl [Proprietary Information]

CSE-08-PDF2 [Proprietary Information]

CSE-08-PDF3 [Proprietary Information]

CSE-08-PDF4 [Proprietary Information]

CSE-08-PDF5 [Proprietary Information]

CSE-08-PDF6 [Proprietary Information]

CSE-08-PDF7 [Proprietary Information]

CSE-08-PDF8 [Proprietary Information]

CSE-08-PDF9 [Proprietary Information]

CSE [Proprietary Information]

PDFlO CSE [Proprietary Information]

PDFll CSE [Proprietary Information]

PDF12 CSE-08-AEFl [Proprietary Information]

CSE-08-ACl [Proprietary Information]

CSE-08-AC2 [Proprietary Information]

• [Proprietary Information]

DB E = design basis earthquake. lX ion exchange.

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' ~* * ~: NORTHWEST MEOJCAl ISOTOPES

[Proprietary Information]

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NWMl-2013-021 , Rev . 2 Chapter 6.0 - Engineered Safety Features Table 6-11. [Proprietary Information] Double-Contingency Controls (3 pages)

Identifier Feature description and basis CSE [Proprietary Information]

AEF I" CSE-09-AC 1a [Proprietary Information]

CSE-09-AC2" [Proprietary Information]

CSE-09-AC3" [Proprietary Information]

CSE [Proprietary Information]

PDFlb CSE- I 0- [Proprietary Information]

AEFib CSE- I 0-AC I b [Proprietary Information]

CSE-I O-AC2b [Proprietary Information]

CSE- I O-AC3b [Proprietary Information]

CSE- I 0-AC4b [Proprietary Information]

CSE- I O-AC5b [Proprietary Information]

CSE-10-AC6b [Proprietary Information]

CSE-I O-AC7b [Proprietary Information]

CSE-10-AC8b [Proprietary Information]

CSE- I 0-AC9b [Proprietary Information]

• [Proprietary Information]

b [Proprietary Information]

23su uranium-235. SPL single parameter limit.

HIC high-integrity container. u uranium.

RPF Radioisotope Production Facility.

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~.-.; . NWMI

' ~ *. ~ ~ ." NomtwtST MEDtCAl ISOTOPES NWMl-2013-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

6-54

• .-:;..NWMI

~**:***

NWMl-2013-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features

' *e *

• NOATHW£Sl MEOtCAl ISOTOf'ES

[Proprietary Information]

6-55

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

' ~ * . * ~ .* NOflTHWt:ST MEOtCAL tSOTOPES NWMl-2013-021, Rev . 2 Chapter 6.0 - Engineered Safety Features Table 6-12. [Proprietary Information] Double-Contingency Controls (2 pages)

Identifier* Feature description and basis CSE-11-PDFl [Proprietary Information]

CSE-l l-PDF2 [Proprietary Information]

CSE- l l-PDF3 [Proprietary Information]

CSE-l l-PDF4 [Proprietary Information]

CSE- 11-PDFS [Proprietary Information]

CSE-l l-PDF6 [Proprietary Information]

CSE-l l-PDF7 [Proprietary Information]

CSE-l l-PDF8 [Proprietary Information]

CSE-11-AEFl [Proprietary Information]

CSE-11-ACl [Proprietary Information]

a [Proprietary Information]

DBE design basis earthquake. Mo molybdenum .

HEPA = high-efficiency particulate air. NOx nitrogen oxide.

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~ * *! NORTHWl.ST lllEDecAl. lSOTOPH NWMl-2013-021 , Rev. 2 Chapter 6.0 - Engineered Safety Features

[Proprietary Information]

Table 6-13. [Proprietary Information] Double-Contingency Controls (2 pages) ldentifiera Feature description and basis CSE-1 3-PDFl [Proprietary Information]

CSE-13-PDF2 [Proprietary Information]

CSE- 13-PDF3 [Proprietary Information]

CSE-13-ACI [Proprietary Information]

CSE-13-AC2 [Proprietary Information]

CSE-13-AC3 [Proprietary Information]

CSE- 13-AC4 [Proprietary Information]

CSE-13-ACS [Proprietary Information]

CSE-13-AC6 [Proprietary Information]

a [Proprietary Information]

R&D research and development. SPL single parameter limit.

RPF = Radi oisotope Production Facility. u uranium.

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~ * *! NOmfWlST MEDtCAL ISOTOPH

[Proprietary Information]

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NWMl-2013-021, Rev. 2 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 ANSI/ 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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~* * ~ NORTifW(tT MEOtCAi. tsOTOPE.S NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features Solid uranium will be handled outside of processing systems during:

• Receipt and processing of fresh uranium (and presumab 1y 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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NWMl-2013-021, Rev. 2 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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• • • NWMI NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

' ~- * ~ NORTHWEST 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.

6.3.1.2.3 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 oflndividual Tanks," is identified by the accident analyses 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, 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 fi ssile 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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NWMl-2013-021, Rev. 2 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 6.3.1.1.

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

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...... NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

~* * ~ NOATHW£ST MEDICAL tsOTOPES 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 (1) 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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' ~ * *! . NOfmfWE.IT M£DfCAL LSOTDPES NWMl-2013-021, Rev. 2 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 will 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-11 , "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.

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~**; :** NWMI

. ~* ~!: NORTifW£tT MlDICAl ISOTOftlS NWMl-2013-021, Rev. 2 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.

6.3.1.2.8 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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NWMl-2013-021, Rev . 2 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 fissile 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 fissile 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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.*:i*;~*:*

....NWMI

~ *.*! NORTifWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 2 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, "Backflow 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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.* *~**;

~ * *!

NORTHWEST MlDfCAL tsOTOP£S NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features Test Requirements See description in Section 6.2 .1. 7.9.

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 (1) 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)

Condensate sampling systems Condensate monitors 6-69

......;~~... .NWMI

' ~ *.*! NORTHWfST MEDtcAL ISOTOPES NWMl-2013-021, Rev. 2 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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NWMl-2013-021, Rev. 2 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.

6.3.1.2.16 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 fissile product solids, from the same event, from penetrating the hot cell shielding boundary, and any high-dose fission gases from penetrating the hot cell shielding boundary during normal operations.

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 will 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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• • • • NWMl-2013-021, Rev. 2 Chapter 6.0 - Engineered Safety Features

~ * *! NCNITHWHT M£DtCAl ISOTOP£S

6.4 REFERENCES

10 CFR 20, "Standards for Protection Against Radiation," Code ofFederal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1201, "Occupational Dose Limits for Adults," Code of Federal Regulations, Office of the Federal Register, as amended.

10 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.

10 CFR 70.61, "Performance Requirements," Code ofFederal 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.

ANSl/ANS-8.3, Criticality Accident Alarm System, American National Standards Institute/American Nuclear Society, La Grange Park, Illinois, 1997 (Reaffirmed in 2012).

ANSI/ ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materials, 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, lllinois, 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 (Reaffrrmed in 2012).

ANSl/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 (Reaffrrmed in 2012).

ANSI/ ANS-8.26, Criticality Safety Engineer Training and Qualification Program , American National Standards Institute/American Nuclear Society, La Grange Park, lllinois, 2007 (Reaffirmed in 2012).

ANSI/ANS-15 .1, The Development ofTechnical Specifications/or Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2013.

ANSI N13.1, 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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NWMl-2013-021, Rev. 2 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-153 7, "Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors, " Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors , Docket Number:

NRC-2011-0135 , U.S. Nuclear Regulatory Commission, Washington, D.C., October 30, 2012.

NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility, Rev. I , U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C., May 2010.

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors - Format and Content, Part 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February 1996.

NUREG/CR-4604 1PNL-5849, Statistical Methods for Nuclear Material Management, Pacific Northwest Laboratory, Richland, Washington, December, 1988.

NUREG/CR-6698, Guide for Validation ofNuclear 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, 2015.

NWMI-2015-CSE-003, NWMI Preliminary Criticality Safety Evaluation: Molybdenum-99 Recovery, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

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~* * ~ NORTHWUT MEDtCAl. ISOTDPES NWMI-2015-CSE-004, NWMI 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, Corvallis, 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, NWMI Preliminary Criticality Safety Evaluation: Target and Can Storage and Carts, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-CSE-008, NWMI Preliminary Criticality Safety Evaluation: Hot Cell Uranium Purification, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015-CSE-009, N WMI Preliminary Criticality Safety 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-O 11 , N WMI Preliminary Criticality Safety Evaluation: O.ffgas and Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-CSE-012, NWMI Preliminary Criticality Safety Evaluation: Target Transport Cask or Drum Handling, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015 .

NWMI-2015-CSE-013 , N WMI Preliminary Criticality Safety Evaluation: Analytical Laboratory, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

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. 2 August 2017 Prepared by:

Northwest Medical Isotopes, LLC 815 NW 9th Ave, Suite 256 Corvallis, OR 97330

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. NWMI NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

~ * *! NORTNW£ST M£DfCAl ISOTOPES Chapter 7 .0 - Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 2 Date Published:

August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 2

Title:

Chapter 7.0 - Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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. NWMI NWMl-2013-021 , Rev . 2 Chapter 7.0 - Instrumentation and Control Systems

' !* *~ NORTHWEST MlOfCAl ISOTOP1S 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 2 8/5/2017 Modification based on ACRS comments C. Haass

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.**:.**.*.* Chapter 7.0 - Instrumentation and Control Systems NWMl-2013-021 , Rev. 2 NOflTHWtST ME.DtCAL ISOTOftES CONTENTS 7.0 INSTRUMENTATION AND CONTROL SYSTEMS ... .... ........ ... .... ........ .. .......... ..... ... ............ 7-1 7 .1 Summary Description ................ .... ....... ..... ... ...... .... .. .. ..... ..... .. .... .... .... ...... .... ... ... ......... .. 7-1 7.2 Design ofl nstrumentati on 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 Descripti on ..... ... .. ...... ..... ...... ..... ... ....... ...... ... .. ... ........ ... .... ....... .... ...... 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 ... .. ..... .... ..... ...... ... .... .. ........... .. ......... 7-1 5 7 .2.3.5 Fire Protection System ...... ....... ......... ... .. .. .... ... .... .... .. ... ..... .. .... .... ... 7-1 5 7 .2.3 .6 Facility Communication Systems .......... .... ......... .. ... .. ...... .. ............. 7-1 5 7 .2.3.7 Analytical Laboratory System ...... .... .... ... ...... .... ..... .... ...... ... .. .. ... .. .. 7-1 5 7.2.4 System Performance Analys is .... ... ... ..... .. .. .... ... .. .... .. ... ... .... ... ........ ...... ........... 7-16 7.2.4. l Facility Trip and Alarm Design Basis .... .... .. .... ....... .... ..... .. ......... .... 7-16 7.2.4.2 Analysis ...... ... ... ... ... ..... ........ ... ... ... ...... .. .... .. .. ....... .......... ... .. ........ ... 7-1 7 7.2.4.3 Conclusion ..... .. ... ... ... ........ ...... ... ........... ........ ..... .. .. ...... .. ... .. .... .... ... 7-1 7 7.3 Process Control Systems ...... .. .. ... ............... .... ....... ... ........... ...... ..... ..... .... ...... .. .......... .. . 7-22 7.3.1 Uranium Recovery and Recycle System .... .... .. ...... ... ..... ... ..... ........ .... ... .......... 7-23 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 Analys is 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 Analys is and Conclusion .. .... ...... ......... ... ....... 7-32 7.3 .3 Target Receipt and Disassembly System .. .... .... ..... .... .... ....... .. .... ... ..... .. ..... ..... . 7-32 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 .. .. .. ..... .... ........ .... ..... ...... ..... .. ...... .. ....... .. .. ... ..... 7-33 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 Requ irements .... ... ...... ....... ... .... ..... .. ......... 7-34 7.3 .4.3 System Description .... ... .. .. .... .... ... ... ... .... ..... .... ........ ....... .. .. ......... ... 7-34 7.3.4.4 System Performance Analys is 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 Analysis and Conclusion ....... ...... ... .. .. ... ... ..... 7-3 9 7.3.6 Waste Handling System ... .. ..... ....... .... .. .... .... .... ... ....... .... .. ............... .. .. ........... 7-39 7.3 .6. 1 Design Criteria ...... .................. .. .. ...... ........ .... ..... ....... ..... .... .... .. .. .... 7-40

7. 3.6.2 Design Basis and Safety Requirements .... .... .... ..... ... ........ .. .... ... .. .. .. 7-40 7.3.6.3 System Description ..... .... .... ..... ... ... ..... .. .. ........ ...... ...... .... ....... .. .... .. 7-40 7-i

NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 7.3.6.4 System Performance Analys is and Conclusion ...... ...... .... ...... .. ...... . 7-43 7.3 .7 Criticality Accident Alarm System ..... ...... ......... ...... .. .... .. ..... ..................... ... .. 7-43 7.3.7.1 Design Criteria ....................................... .. ....... .. ..... ........... ..... ..... .. . 7-43 7.3.7.2 Design Basis and Safety Requirements ...... .... ................................. 7-43 7.3 .7.3 System Description ....... ...................... .. ......................................... 7-43 7.3.7.4 System Performance Analysis and Conclusion .... ... ..... .... ...... ... ...... 7-43 7.4 Engineered Safety Features Actuation Systems .. ... ........... ... ....... .... .............................. 7-44 7.4.1 System Description ........ ..... ......................... ........ ........ ... .................. ... ...... .. .. 7-44 7.4.2 Annunciation and Display ...... .. ........................... ... ........ .... .... .......... ....... ..... ... 7-45 7.4.3 System Performance Analysis ........................... ..... ...... ...... ...... .. ... ..... ..... ....... 7-45 7.5 Control Console and Display Instruments ....... ................... .. .............. ....... ............ ...... .. 7-46 7.5 .1 Design Criteria ................................................... ... .... ................. .. .................. 7-46 7.5.2 Design Basis and Safety Requirements .......... ....... .... ........ .. ...................... ... ... 7-46 7.5 .3 System Description .......... ................................. .. ........ ..... ... ...... .. ................ ... 7-46 7.5.4 System Performance Analysis and Conclusion ... .... ..... ...... ....... ........ .............. 7-46 7.6 Radiation Monitoring Systems ........................... ......... ............................................... .. 7-47 7.6.1 Design Criteria ................ .... ...... ............................................. ... .... ... .. ... ...... ... 7-47 7.6.2 Design Basis and Safety Requirements .......... ...... ...... ..... ...... ..... .................... . 7-47 7.6.3 System Description ... ...... ...... ............................. .. ....... ...... ........... ...... ....... ..... 7-47 7.6.3 .1 Air Monitoring .. ..... .. ...... ... ............. ....... ..... ...... ....... ..... ..... ........ ..... 7-48 7.6.3.2 Stack Release Monitoring .............. ....... ........ ... .... ....... .... .. ......... ... . 7-49 7.6.4 System Performance Analysis and Conclusions ... .......... .. .......................... ..... 7-49 7.7 References .... .. ....... ....... ... ... ...... ... .................................... ....................................... ..... 7-50 7-ii

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• NOllTlfWHT MlDtcAl ISOTOPES NWMl-2013-021 , Rev . 2 Chapter 7.0 - Instrumentation and Control Systems 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) .. .. .............. ... 7-31 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 ........ ..... ... .................. ... ............... ... ...... ..... .............. ........ .. .............. ... ....... .... 7-39 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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~* *~ NORTMWEST llEDICAl ISOTDPU 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] [Proprietary Information]

UPS uninterruptible power supply V&V verification and validation Xe xenon Units m meter min minute rad radiation absorbed dose 7-iv

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~* * ~ . NORTHWEST MlDICAI.. ISOTDH:S 7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.1

SUMMARY

DESCRIPTION The Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF) preliminary instrumentation and control (l&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 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) 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 (tum 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 I&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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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems Process Support Systems Waste Handling System

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---e Fire Protection System I

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I Facility Ventilation System Process Utility Systems

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+ - - + Control and Monitoring Process Systems In Hot Cell Area Process Systems In Target Fabrication Area 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 HMis, with surveillance monitoring in the control room.

• Uranium recovery and recycle - Batch processes located inside the hot cell area will be monitored and 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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• NORTHWEST MEDICAL ISOTOPES

• 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 HMis in the low-dose liquid solidification room (WI 07) 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 HMis and hard-wired interlocks for the ESF functions. 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

• Energy recovery and heating water 7-3

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' ~* * ~ NotmfWEST M£1HCAL ISOTIWES 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 (RO 12)

• Operating gallery (G 10 I A/B/C)

• Target fabrication (Tl 04 A/B)

• Low-dose liquid waste solidification (Wl 07)

• Chemical supply room (L102)

• Local to equipment with integrated control systems 7.2 DESIGN OF INSTRUMENTATION AND CONTROL SYSTEMS The design criteria and the codes and standards for l&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 l&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 ofthe 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 failure, common-cause determined by ANSI/ISA 84.00.01 failure, 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-2015-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 facility. The facility will not have all of the systems detailed in this standard and guidance will be applied 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/or Safety Application:

Systems for Nuclear Power Generating Stations

• Use for design ofFPC system, ESFs, and other

Description:

Establishes minimwn functional and instrumentation SSCs that are identified as IROFS design criteria for the power, instrwnentation, 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, or NWMI-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 applicability

• The RPF is not considered a nuclear power reactor to equipment provided for safe shutdown, post-accident but a production facility. The facility will not have monitoring display instrwnentation, preventive interlock all of the systems detailed in this standard and features, or any other systems, structures, or equipment guidance will be applied as appropriate.

related to safety.

Keywords: Actuated equipment, associated circuits, Class 1E, design, failure, maintenance bypass, operating bypass, safety function, sense and command features, sensor IEEE 384-2008, IEEE Standard Criteria/or Application:

Independence of Class JE Equipment and Circuits

• Use for design ofFPC system, ESFs, and other

Description:

Describes independence requirements of instrwnentation SSCs that are identified as IROFS circuits and equipment comprising or associated with

• Apply minimwn criteria for separation and Class lE 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 lE, 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:

IE Equipment for Nuclear Power Generating Stations

• Use for equipment qualification when needed to

Description:

Identifies requirements for qualifying qualify equipment for applications or environments to Class IE equipment and interfaces that are to be used in which the equipment may be exposed nuclear power generating stations. The principles,

• Use for qualification of Class 1E equipment located methods, and procedures are intended for use in in harsh environments and for certain post-accident qualifying equipment, maintaining and extending monitoring equipment; may also be used for the qualification, and updating qualification, as required, if qualification 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 for those systems determined to be under applicable service conditions, including design Class IE and located in harsh environments for safety basis events, reducing the risk of common-cause functions identified in Chapters 6.0 and 13, or equipment failure. NWMI-2015-SAFETY-002 Keywords: Age conditioning, aging, condition

• Apply 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 life, radiation, safety-related Exception:

function, significant aging mechanism, test plan, test

• The RPF is not considered a nuclear power reactor sequence, type testing 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 344-2004, IEEE Recommended Practice for Application:

Seismic Qualification of Class IE Equipment for

• 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 1E equipment can meet instrumentation SSCs that are identified as a Class 1E 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 l E, 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 description 3 Design bases as applied to RPF IEEE 338-2012, IEEE Standard for Criteria for the Application:

Periodic Surveillance Testing of Nuclear Power

• Use for design ofFPC system, ESFs, and other Generating Station Safety Systems instrumentation SSCs that are identified as IROFS

==

Description:==

Provides criteria for the performance 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 functional 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-2015-SAFETY-002 system performs its defined safety function . Post-Exception:

maintenance and post-modification testing are not covered by this document. This standard amplifies 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 this standard and guidance will be applied as appropriate.

Keywords: Functional tests, IEEE 338, periodic testing, risk-informed testing, surveillance testing IEEE 497-201 O, 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 for Abstract: Specifies additional computer-specific computers that are components of a safety system requirements to supplement IEEE 603-2009. The

• Design FPC system as a DCS, and apply this standard standard defines the term computer as a system that to system development, specifically software includes computer hardware, software, firmware, and development interfaces, and establishes minimum functional and

• Apply standard to CGD and implement an approach design requirements for computers used as components Compliance:

of a safety system.

• Develop FPC system software using this standard Keywords: Commercial-grade item, diversity, safety Exception:

systems, software, software tools, software verification

• The RPF is not considered a nuclear power reactor and validation 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 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 specific 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/lEC/lEEE STD 24765 and the information item requirements ofIEEE 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 for Software Reviews Application:

and Audits

• Use to identify minimum acceptable requirements for Description : Identifies fi ve types of software reviews systematic software reviews and audits, togeth er with procedures required for the

• Identify organizational means for conducting a review execution of each type. This standard is concerned only and documenting the findin gs with reviews and audits; procedures for determining the

• Design FPC system as a DCS, and apply standard necessity of a review or audit are not defined, and the during the development of software for systems with disposition of the results of the review or audit is not IROFS fun ctions specified. Types included are management reviews, Compliance:

technical reviews, inspections, walk-throughs, and

• Develop FPC system using this stan dard audits.

Keywords : Audit, inspection, review, walk-through ANS 10.4-2008, Verification and Validation ofNon- Application:

Safety-Related Scientific and Engineering Computer

• Perform software V& V to build quality into the Programs/or 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 scientific 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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Design criteria description* 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 establish setpoints for Description : Defin es requirements for 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 IROFS for nuclear power plants or nuclear reactor facilities. the RPF Keywords : Setpoint, drift, analog channel, reliability Compliance:

analysis

• Ensure design conforms to the practices detailed in the standard for IROFS functions with inherent setpoints identified in Chapters 6.0 and 13.0, or NWMI-2015-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 ANSl/ISA- - Use Part 1 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 functions required to be implemented by the FPC system using Parts l, 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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Design criteria descriptiona 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 will be provided in the Operating License human-system interface Application 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 guidance 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 application 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 facility 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 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 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 experience 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 N uclear Power Plants

• Use for l&C system designs with computers in safety-Description : Describes a method that th e NRC staff related systems that make extensive use of advanced deems acceptable for complying with NRC regulations technology for promoting high fun ctional reliability, design quality,

• Use for RPF designs (that are expected to be and a secure development and operati onal environment sign ifi cantly and functionally different from current for the use of digital computers in th e safety systems of day process designs) with microprocessors, digital nuclear power plants. systems and di splays, fi ber opti cs, multiplexing, an d Keywords: Secure development and operational different isolation techniques to achieve suffi cient environment (SDOE), computers independence and redundancy Compliance:

• Develop FPC system and associated HMI using this guidance Exception:

• The RPF is not considered a nuclear power reactor but a producti on faci lity. The facility wi ll not have all of th e systems detailed in thi s standard and guidance will be applied as appropriate.

Regulatory Guide 1.53, Application ofthe Single- Application:

Failure 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 ofnuclear power plant safety Compliance:

systems.

• Evaluate FPC system, ESFs, and IROFS end-devices Keywords: IEEE 379-2014, single-failure criterion using this guidance 7-12

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Design criteria descriptiona 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 plants . this guidance Keywords: IEEE 497-2010, 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 wi ll 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 10 CFR 73.1.

Keywords: Cybersecurity, 10 CFR 73.54(a)(2), design basis threat

• Full references provided in Section 7.7.

CAAS criticality accident alarm system. IROFS items relied on for safety.

CAM continuous air monitor. NRC U.S. Nuclear Regulatory Commi ss ion.

CFR Code of Federal Regulati ons. PLC programmable logic controll er.

COD com mercial grade dedi cati on. RAM radiation alarm monitor.

COTS commercial off-the-shelf. RPF Radi oisotope Production Facility.

DCS digita l control system . S DOE secure development and operati onal ESF engineered safety feature. environment.

FPC facility process control. SIF safety instrum ented funct ion.

HMI human-machine interface. SIL safety integrity level.

I&C instrumentation and control. SIS safety instrum ented system.

IEEE Institute of Electrical and Electroni cs SSC structures, systems, and components.

Engineers. V&V verification and validation.

Specific requirements will be developed during the next stages of design for the Operating License Application. The l&C design will be expanded and analyzed to document fulfillment of the design criteria and design basis requirements for the Operating License Appl ication.

7.2.3 System Description As described in Section 7.1, the RPF l&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. The items relied on for safety (IROFS)/ESF safety functions will be activated via hardwire (analog) interlocks.

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~* *~ HORTHW£ST MEDtCAl ISOTOPH 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 standby redundant system structure. The standby workstations provide redundant hardware with identical PLC software systems as automatic backup control systems. The primary and backup PLC systems monitor each other. This 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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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.

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.

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' ~* *~ . NORTHWEST MEDfCAl ISOTOPES NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 7.2.4 System Performance Analysis The RPF l&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 l&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.

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.

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• ~* * ~ NORTHWEST MEDICAL ISOTDnS 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, in addition to the ability of the ESF systems to perform the same, 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 for the Determination of Setpoints for Nuclear Safety-Related Instrumentation.

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.

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-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-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.

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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

~ - ~~

• NotmfWESTMEDfCAllSOTOPU Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages)

Criteria* Design basis applicability Functional means IEEE 379

• FPC system

• Safety DCS preapproved platform Single failure criterion

• FPC system display

• Redundant independent isolation

• FPC system IROFS end devices components

• ESFs

• Redundant operator interface workstations

• ESFs manual isolation

• Redundant sensors

• Alternative manual means for ESF initiation IEEE 577

• FPC system

• Safety DCS pre-approved platform for an Reliability analysis

• FPC system display SIS criterion

• FPC system IROFS end devices Redundant independent isolation

• ESFs components

• ESFs manual isolation

• Redundant operator interface workstations

• Redundant sensors

• Alternative manual means for ESF initiation IEEE 603

• FPC system See Section 7.3 for details.

Standard criteria safety

• FPC system display system

• FPC system IROFS end devices

• ESFs

• ESFs manual isolation IEEE 384

• FPC system

• IEEE 603 and IEEE 379 were used during Independence of

• FPC system display development of the Construction Permit Class IE equipment

• FPC system IROFS end devices Application.

and circuits

• ESFs

• Additional details will be developed for the

• ESFs manual isolation Operating License Application.

IEEE 323

• FPC system

• Standard supports selection and Qualifying Class IE

• FPC system display qualification of equipment to be Class 1E Equipment

• FPC system IROFS end devices use qualified.

• ESFs

• This standard will be reevaluated in the

• ESFs manual isolation Operating License Application for applicability.

IEEE 344

• FPC system

• 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

• ESFs

• Standard will be reevaluated in the

• ESFs manual isolation Operating License Application for applicability.

IEEE 338

• FPC system

• Standard supports selection of equipment; Criteria for the periodic

• FPC system display which resulted in the use of general design surveillance testing of

• FPC system IROFS end devices criteria (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 Application for applicability.

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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages)

Criteriaa Design basis applicability Functional means IEEE 497

• FPC system

• Standard supports selection of accident Criteria for accident

• FPC system display monitoring equipment (e.g., radiation monitoring instruments

• FPC system IROFS 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

• CAMs Application.

• Standard will be reevaluated in the Operating License Application for applicability.

IEEE 7-4.3.2

• FPC system

• Programming software must comply with Criteria for digital

• FPC system display these criteria and with the NWMI Software computers in safety

• HMJ displays 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 this standard.

• Software and hardware used for the displays for the FPC system and HMI must also follow guidelines set forth in this standard.

• Standard will be reevaluated in the Operating License Application for applicability.

IEEE 828

• FPC system

• 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

• FPC system

• Complies with IEEE 7-4.3.2 and the NWMI Software and system

• FPC system display Software Quality Assurance Plan test documentation

• HMJ displays

• Standard will be reevaluated in the Operating License Application for applicability.

IEEE 1012

• FPC system

• 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.

IEEE 1028

• FPC system

• Complies with IEEE 7-4.3.2 and the NWMI Software reviews and

• FPC system display Software Quality Assurance Plan audits

• HMJ displays

• Standard will be reevaluated in the Operating License Application for applicability.

ANS-10.4

• FPC system

• 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.

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' ~* * ~ NOAllfWUT MEOICAl ISOTOPH NWMl-2013-021 , Rev . 2 Chapter 7.0 - Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages)

Criteriaa Design basis applicability Functional means ANSI/ISA 67.04.01

• FPC system

• Incorporated into overall design and th e Setpoints for nuclear

• FPC system IROFS end devices Construction Permit Application.

safety-related

• Standard wi ll be reevaluated in the instruments Operating License Application for applicability.

ANSI/ISA 84.00.01,

• FPC system

• 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

• FPC system

• 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 Con struction Permit Application.

• Standard will be reevaluated in the Operating License Application for applicability.

NUREG/CR-6463

• FPC system

• 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

• FPC system

• 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.

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.

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NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means ( 4 pages)

Criteriaa Design basis applicability Functional means Regulatory Guide

• FPC system

• Standard supports the design and 1.152

• FPC system display development of redundant safety PLC Criteria for use of

• HMI displays platforms, 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 applicability.

Regulatory Guide 1.53

• FPC system

• 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

• FPC system

• Criteria require th e development of a design Cybersecurity

• FPC system display approach and implementation for programs for nuclear

• HMI display cybersecurity.

faci lities

• Standard wi ll be reevaluated in the Operating License Application for applicability.

• Full references are provided in Section 7.7.

CAAS criticality accident alarm system. CROFS items relied on for safety.

CAM continuous air monitor. NWM I Northwest Medical Isotopes, LLC.

DCS di gital control system. PLC programmable logic controller.

ESF engineered safety feature. RAM radiation alarm monitor.

FPC facility process control. SIS safety instrumented system.

HMI human-machine interface.

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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 system includes interlocks (both hardwired [ESF] and computer logic) to implement an automatic action on a parameter approaching or being outside its setting. Interlocks are defined as specific set of conditions or parameters that need to be met for an activity to occur. An example of an interlock is the shutting down a pump on a tank high-level alarm signal or switching to a spare unit or process train based on a change in parameters (and corresponding alarm). In addition to interlocks, the RPF will also implement a permissive philosophy that allows HMI operations to be enabled once the control room has confirmed the prerequisites conditions have been completed.

Permissives differ from interlocks in that permissives require manual approval via a switch (or similar) that must be satisfied for an activity to occur. Interlocks are engineered features , and permissives are administrative features. The permissive and interlocks will be described in more detail in the Operating License Application.

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 will initialize 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.

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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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.

Due to the variety of process activities performed during uranium recovery and recycle, the system description is divided into the following subsystems:

• Impure uranium collection

• Primary ion exchange

• Primary concentration

• Secondary ion exchange

• Secondary concentration

• Uranium recycle

• Uranium decay and accountability

• 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 l&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 will be developed for the Operating License Application.

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' ~ *-* ~ .

NOllTHW£STMEOtCALISOTOPH NWMl-20 13-02 1, Rev . 2 Chapter 7. 0 - Instrumentation and Control Systems Table 7-3. Uranium Recovery and Recycle Control and Monitoring Par ameters (2 pages)

Subsystem Control parameters Primary control name (automatic/manual) Monitoring parameters location

. Flowrate (A) . Density Impure

.. Differential pressure Control room uranium

. Pump actuation (M)

. Level Flowrate collection

. Temperature (A)

Pump motor speed (A)

. Pressure

. Valve actuation (AIM)

.. Temperature Valve position Primary ion . Flowrate (A) .. Analyzer, uranium Control room exchange

.. Pump actuation (AIM)

. Density Pump motor speed (A) Differential pressure

. Temperature (A)

Valve actuation (AIM)

Flowrate Flowrate totalizer

.. Level Pressure

. Temperature Valve position Primary Density (A)

.. Analyzer, uranium Density Control room concentration

. Flowrate (A)

.. Differential pressure Level (A)

Pump actuation (AIM) Flowrate Pump motor speed (A)

. Level Pressure

. Temperature (A)

Valve actuation (AIM) Temperature Valve position

. . Analyzer, uranium Secondary ion Flowrate (A)

. Density Control room exchange

.. Pump actuation (AIM)

. Differential pressure Pump motor speed (A)

. Flowrate

. Temperature (A)

Valve actuation (AIM) .. Flowrate totalizer Level Pressure Temperature Valve position

. .. Analyzer, uranium Secondary

. Density (A)

Control room concentration

.. Level (A)

Flowrate (A) Density Differential pressure Flowrate

. Pump actuation (AIM)

Level

. Temperature (A)

Pump motor speed (A)

. Valve actuation (AIM)

Pressure Temperature Valve position Uranium .. Flowrate (A) .. Density Control room recycle Pump actuation (AIM) Differential pressure Pump motor speed (A) Flowrate Valve actuation (AIM) Level

.. Pressure Temperature Valve position 7-24

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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

' ~* *~* NORTHWEST MEIKCAL ISOTON:S Table 7-3. Uranium Recovery and Recycle Control and Monitoring Parameters (2 pages)

Subsystem Control parameters Primary control name (automatic/manual) Monitoring parameters location

.. . Density UraniWTI decay Flowrate (A)

. Differential pressure Control room

. ... Level and Pump actuation (AIM)

Flowrate accountability PWTip motor speed (A)

Temperature (A)

Valve actuation (AIM)

. Pressure

. Temperature Valve position

.. . Analyzer, uraniWTI Spent ion Flowrate (A)

.. Control room exchange resin

.. Pump actuation (AIM)

Pump motor speed (A)

Differential pressure Flowrate Valve actuation (AIM)

.. Level Pressure Valve position

.. . Density Waste Flowrate (A)

. Differential pressure Control room collection

. Pump actuation (AIM)

. Flowrate

. Pump motor speed (A)

. Level

. Temperature (A)

. Pressure Valve actuation (AIM)

. 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-IOOA) 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-level switch PLC NIA IX feed tank 1 (UR-TK-200) high-temperature switch PLC NIA IX column lA (UR-IX-240) high-uranium alarm (AAH-252) PLC NIA IX column lA U solution filter (UR-F-250) high-differential PLC NIA pressure alarm IX column IA waste filter (UR-F-255) high-differential 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

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.... NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

. ~ *.* ! . NORTHW£ST MtDfCAl ISOTOPES 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 IA (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 l (UR-TK-370) high-liquid level alarm PLC NIA Condensate sample tank IB (UR-TK-340) high-liquid level PLC NIA alarm Condensate sample tank IB (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 NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

~* * ~ NOITHWUT M£OtCAl ISOTOf'lS 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-597)

Close IX column 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 Uranium rework tank (UR-TK-660) high-liquid level alarm PLC NIA Uranium rework tank (UR-TK-660) high-temperature alarm PLC NIA Uranium 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

• • • NWMI

...... NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

~* * ~ . NORllfWUT MEDK:Al lSOTOPES Table 7-4. Uranium 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 detennined.

PLC programmab le logic controller.

7.3.1.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.2 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 facilities .

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

NWM l-2013-021, Rev. 2 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 oflROFS CS-14, CS-15 , CS-20, CS-27, and RS- I 0 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 Mo nitoring Parameters (2 pages)

Control parameters Primary control Subsystem name (automatic/manual) Monitoring parameters location Fresh uranium receipt and dissolution ... Current (A)

Conductivity (A)

Conductivity Density Local Differential pressure (I 00-series tag numbers)

. Flow totalizer (A)

Flowrate

. Heater actuation (AIM)

.. Level

.. Level (A)

Pump actuation (AIM) Pressure Temperature

. Temperature (A)

Valve actuation (AIM)

Nitrate extraction .. Analyzer, pH (A)

... Density Analyzer, pH Local (200-series tag numbers) Contactor actuation (M)

Flow totalizer (A) Differential pressure Flowrate (A)

. Level Flowrate

.. Level (A)

Pump actuation (AIM) .. Pressure

.. Pump motor speed (A)

Temperature (A) . Pump motor speed Temperature ADUN concentration .. Valve actuation (AIM)

Conductivity (A) .. Conductivity Local

. Density (300-series tag numbers) Density (A)

Flowrate (A)

. Flowrate Level (A)

. Level

. Pump actuation (AIM)

. Pressure Temperature

. Pump motor speed (A)

Valve actuation (AIM) 7-29

NWMl-2013-021 , Rev . 2 Chapter 7.0 - Instrumentation and Control Systems Table 7-5. Target Fabrication System Control and Monitoring Parameters (2 pages)

Control parameters Primary control Subsystem name (automatic/manual) Monitoring parameters location

[Proprietary Information] .. Level (A) .. Flowrate Local (400-series tag numbers) Pump actuation (AIM) Level Tank agitator actuation Pressure

.. (AIM)

Tank agitator speed (A)

Temperature

. Temperature (A)

Valve actuation (AIM)

. Density

[Proprietary Information]

. Flowrate (A)

Differential pressure Local (500-series tag numbers) Pump actuation (AIM)

Pump motor speed (A)

Temperature (A) .. Pressure Level

. Valve actuation (AIM)

Vibration dispersion . Temperature Vibration

.. assembly actuation (M)

. Analyzer, hydrogen

[Proprietary Information] Analyzer, hydrogen (A)

.. Local (600-series tag numbers) Analyzer, oxygen (A) Analyzer, oxygen Flow totalizer (A)

Level (A) .. Flowrate Level

.. Tank agitator speed (M)

Temperature (A) . Pressure Temperature Valve actuation (AIM)

Target fabrication waste .. Flowrate (A) Density Local Flowrate (700-series tag numbers)

.. Level (A)

Pump actuation (AIM) Level

. Pump motor speed (A)

Valve actuation (AIM)

Pressure Temperature Target assembly TBD TBD Local

[Proprietary Information] TBD TBD Local New target handling TBD TBD Local ADVN acid-deficient uranyl nitrate. TBD to be determined.

LEU = low-enriched uranium.

Table 7-6 provides a listing of the target fabrication I&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 App lication.

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• ~*:~*:* NWM I

!* * ~ NORTHWHT MEDICAL ISOTOPES NWM l-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 Uraniwn 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 Uraniwn dissolution heat exchanger (TF-E-120) low- PLC NIA differential pressure alarm Uranyl nitrate storage tank (TF-TK-200) level switch PLC NIA 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 7-31

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.....;......~. NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

~- * ~ NOtmf'WUTMl:DICAltSOTOl'U Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages)

Hard-wired or Interlock or permissive input PLC Safety interlock

[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 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 tri chi oroethylene.

7.3.2.4 System Performance Analysis and Conclusion The system performance ana lysis 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 wi ll feed the target dissolution system by the transfer of recovered irradiated target material through the dissolver 1 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.

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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 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 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 (9 9Mo) 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.

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~.

NWMI NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

~* ' ~ . NOflTHWUT MEDfCAl ISOTOPES 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 7.3.4.1 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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NWMI NWMl -2013-021 , Rev. 2 Chapter 7.0 - Instrumentati on and Control Systems

~ * .* ~ NORTHWEST MEDfCAl ISOTOPU Table 7-7. Target Dissolution System Control and Monitoring Parameters Subsystem Control parameters Primary control name (automatic/manual) Monitoring parameters location

. Dissolver agitator actuation . Dissolver agitator speed Target

. Flowrate Operating gallery dissolution 1 and2 .. (AIM)

Dissolver agitator speed (A) .. Flowrate totalizer

.. Flowrate (A)

. Level Pump actuation (AIM)

. Pressure

. Temperature Radiation Pump motor speed (A)

Temperature (A)

. Valve position NOx treatment 1 .. Valve actuation (AIM)

F lowrate (A) .. Differential pressure Operating gallery Flowrate or 2 Pump actuation (AIM)

Pump motor speed (A) Flowrate totalizer Temperature (A) Level Valve actuation (AIM)

.. Pressure Radiation

. Temperature Valve position

.. Flowrate Pressure relief

... Pump actuation (AIM)

Pump motor speed (A)

. Level Operating gallery

. Temperature (A)

. Pressure Valve position Valve actuation (AIM)

.. Differential Primary fission gas treatment .. Temperature (A)

Valve actuation (AIM)

.. Flowrate Pressure pressure Operating gallery

.. Radiation Temperature Valve position

. . Differential pressure Secondary fission gas Valve actuation (AIM)

.. Flowrate Operating gallery treatment

. Pressure

.. Temperature Radiation Valve position Waste .. Pump actuation (AIM) .. Differential pressure Operating gallery

.. Flowrate collection Pump motor Speed (A)

Temperature (A) Level Valve actuation (AIM)

.. Temperature Pressure

. Radiation Valve position NOx nitrogen oxide.

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

.. NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

• ~* *~ ' NORTHWfSTMEDtCAL lSOTOPES Table 7-8 provides a preliminary listing of the target dissolution l&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- I 0 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 1 (DS-D-100) high-liquid level alarm PLC NIA Dissolver 1 (DS-D-100) low-liquid level alarm PLC NIA Dissolver l (DS-D-100) high liquid temperature alarm PLC NIA Dissolver 1 Condenser (DS-E-130) high gas temperature alarm 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 1 (DS-C-310) high-liquid level alarm PLC NIA Caustic scrubber I (DS-C-310) high gas temperature PLC NIA NOx oxidizer l (DS-C-340) high -liquid level alarm PLC NIA NOx oxidizer I (DS-C-340) high gas temperature PLC NIA NOx absorber l (DS-C-370) high-liquid level alarm PLC NIA NOx absorber 1 (DS-C-370) high gas temperature PLC NIA Primary caustic scrubber 2 (DS-C-410) high-liquid level alarm PLC NIA Caustic scrubber 2 (DS-C-410) high gas temperature PLC NIA NOx oxidizer 2 (DS-C-440) high-liquid level alarm PLC NIA NOx oxidizer 2 (DS-C-440) high gas temperature PLC NIA NOx absorber 2 (DS-C-470) high-liquid level alarm 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 alarm PLC NIA Dryer A (DS-E-610A) high gas temperature alarm PLC NIA Primary adsorber A (DS-SB-620A) high gas temperature alarm 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 alarm PLC NIA Filter B (DS-F-630B) high-pressure differential alarm PLC NIA Dryer C (DS-E-61 OC:) high gas temperature alarm PLC NIA Primary adsorber C (DS-SB-620C) high gas temperature alarm PLC NIA Filter C (DS-F-630C) high-pressure differential alarm PLC NIA 7-36

• i*;~*:* NWMI

~ * *! NOmfWtST MHNCAL ISOTOPES NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 1 (DS-TK-800) high-liquid PLC NIA temperature alarm Waste collection and sampling tank 2 (DS-TK-820) high-liquid level PLC NIA alarm Waste collection and sampling tank 2 (DS-TK-820) high-liquids PLC NIA temperature alarm NIA not applicable. PLC programmable logic controller.

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 IA 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 will be used for final chemical adjustment and sampling for verification of batch acceptance. The product will 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 I&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 this system are listed in Chapter 6.0.

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NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 7.3.5.3 System Description The Mo recovery and purification 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.

Operators using remote in-cell manipulators will 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 oflROFS 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 Flowrate Operating gallery Valve actuation (AIM)

Level

.. Temperature

.. Pressure Radiation Valve position Secondary ion exchange . Pumps (M) . Temperature Operating gallery Tertiary ion exchange . Pumps (M) .. Density Operating gallery

.. Flowrate 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.

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~* * ~ . NORTHWEST MEDfCAl ISOT01'£S NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems Table 7-10. Molybdenum Recovery and Purification System Interlocks and Permissive Signals Hard-wired or Interlock or permissive input PLC Feed tank IA (MR-TK-IOO) high-liquid level alarm PLC NIA Feed tank IA (MR-TK-IOO) low-liquid level alarm PLC NIA Feed tank IA (MR-TK-100) high-temperature alarm PLC NIA Feed tank IA (MR-TK-100) high-pressure alarm PLC NIA Feed tank 1B (MR-TK-140) high-liquid level alarm PLC NIA Feed tank lB (MR-TK-I40) low-liquid level alarm PLC NIA Feed tank 1B (MR-TK-I40) high-temperature alarm PLC NIA Feed tank lB (MR-TK-I40) high-pressure alarm PLC NIA U solution collection tank (MR-TK-I80) high-liquid level alarm PLC NIA U solution collection tank (MR-TK-I 80) 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 7-39

......~**;...NWMI NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

• ~* * ~ NORTHWEST Mf.OtCAl ISOTOPES 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.

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 HMis in the low-dose waste room (Room W107). A local HMI display area 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 and/or monitored from the low-dose waste room (Room WI 07). Liquid waste collection and low-dose liquid waste evaporation operations will be controlled from the RPF control room. Table 7-11 lists the anticipated control parameters, monitoring parameters, and primary control locations 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.

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.;..; NWMI

...~**:***

• NWMl-2013-02 1, Rev. 2 Chapter 7.0 - Instrumentati on and Control Systems

• • *

• NOflTHWf.ST M£Dtc.Al ISOTOPES Table 7-11. Waste Handling System Control and Mo nitoring Parameters Control parameters Subsystem name (automatic/manual) Monitoring parameters Primary control location High-dose liquid waste collection

. Valve position

... Density Differential pressure Control room

. Flowrate

.. Flowrate Level totalizer

. 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 collection

. Pump actuation (AIM)

.. Differential Flowrate pressure

. Pump motor speed (A)

. Flowrate totalizer

. Temperature (A)

... Pressure Valve actuation (AIM) Level Temperature Valve position

.. . Differential pressure Low-dose liquid waste Flowrate (A)

.. Flowrate Control room evaporation Pump actuation (AIM)

Pump motor speed (A)

Temperature (A) . Level

.. Temperature Valve actuation (AIM) Pressure Valve position Low-dose liquid waste .. Flowrate (A) .. Density Low dose solidification room solidification

.. Pump Pump actuation (AIM) motor speed (A) .. Differential pressure Flowrate

. Valve actuation (AIM)

Temperature (A)

. Flowrate totalizer

. Level

.. Temperature Pressure Valve position Spent resin dewatering . Valve actuation (AIM) . Valve position Low dose solidification room Solid waste . Actuate grout mixer (M) . Pressure Low dose solidification room encapsulation High-dose waste decay TBD TBD Low dose solidification room High-dose waste TBD TBD Low dose solidification room handling TBD = to be determi ned.

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..*..***.* ...:*NWMI NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems

' ~ * .* ~

• NOlffifWUT MEDICAL ISOTOPES 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) high-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 high-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) high-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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NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 describ ed 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 provi ded 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 of20 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 electronics 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 provi ded 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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' ~ *.~ ! . NORTHWEST MEDICAL ISOTOPES 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 SS Cs 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 analysis 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 relief system RS-09 Dissolver offgas failure during Pressure relief device, pressure dissolution 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 ventilation during RS-13 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 deflagration or Backup bottled nitrogen gas purge system detonation supply Active discharge monitoring CS-14 Accidental criticality To be provided in the Operating and isolation License Application Independent active discharge 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

NWMl-2013-021 , Rev. 2 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 criticali ty Closed- loop, high-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 I&C instrumentation and control. SSC structures, systems, and components.

IROFS items relied on for safety. TBD to be determined.

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 wi ll 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 variabl es, 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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NWMl-2013-021, Rev . 2 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 Joss 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 Wl07, 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 Wl07, 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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~ * *! NO<<nfWEST MEDM:Al. ISOTOPES NWMl-2013-021 , Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 ofradioactive 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 reasonabl y achievable (ALARA) and in minimizing personnel exposure to radiation.

• Provision of local alarms and/or indicators positioned at key points throughout theRPF wherea substantial increase in radiation levels mi ght 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 l&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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~* * ~: NOfUHW£ST MEDICAL ISOTOPU NWMl-2013-021, Rev. 2 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 (NA VLAP)-certified vendor. Personnel monitoring and dosimetry is described in Chapter 11.0, "Radiation Program and Waste Management."

7.6.3.l 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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NWMl-2013-021, Rev. 2 Chapter 7.0 - Instrumentation and Control Systems 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 level 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

10 CFR 20.1501, "General," Code ofFederal Regulations, Office of the Federal Register, as amended.

10 CFR 70, "Domestic Licensing of Special Nuclear Material," Code ofFederal Regulations, Office of the Federal Register, as amended.

10 CFR 70.61 , "Performance Requirements," Code ofFederal Regulations , Office of the Federal Register, as amended.

10 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.

10 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 ofNon-Safety-Related Scientific and Engineering Computer Programs for the Nuclear Industry, American National Standards Institute, New York, New York, 2008.

ANSI/ ANS 8.3, Criticality Accident Alarm System, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 1997, R2003 , R2012.

ANSI/ISA 67.04.01-2006, Setpoints for Nuclear Safety-Related Instrumentation, American National Standards Institute/International Society of Automation, Research Triangle Park, North Carolina, 2006 (R2011 ).

ANSI/ISA 84.00.01-2004 Part 1, Functional Safety: Safety Instrumented Systems for the Process Industry Sector - Part 1: Framework, Definitions, System, Hardware and Software Requirements ,

American National Standards Institute/International Society of Automation, Research Triangle Park, North Carolina, September 2004.

ANSI/ISA 84.00.01-2004 Part 2, Functional Safety: Safety Instrumented Systems for the Process Industry Sector - Part 2: Guidelines for the Application ofANSlllSA-84.00.01-2004 Part 1 (IEC 61511-1 Mod) - Informative, American National Standards Institute/International Society of Automation, Research Triangle Park, North Carolina, September 2004.

ANSI/ISA 84.00.01-2004 Part 3, Functional Safety: Safety Instrumented Systems for the Process Industry Sector - Part 3: Guidance for the Determination of the Required Safety Integrity Levels -

Informative, American National Standards Institute/International Society of Automation, Research Triangle Park, North Carolina, September 2004.

EPRI TR-106439, Guideline on Evaluation and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications, Electric Power Research Institute, Palo Alto, California, November 1996.

IEC 61508, Functional Safety of Electrical/Electronic/Programmable Electronic Safety-Related Systems, Parts 1 - 7, International Electrotechnical Commission, Geneva, Switzerland, as amended.

IEEE 7-4.3.2-2010, IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2010.

IEEE 323-2003, IEEE Standard for QualifYing Class lE Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2003.

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NWMl-2013-021 , Rev . 2 Chapter 7.0 - Instrumentation and Control Systems IEEE 338-2012, IEEE Standard for Criteria for the Periodic Surveillance Testing ofNuclear Power Generating Station Safety Systems, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2012.

IEEE 344-2004, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2004.

IEEE 379-2014, IEEE Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 20 14.

IEEE 384-2008, IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2008.

IEEE 497-2010, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 20 10.

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 Criteriafor Safety Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2009.

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 1012-20 12, 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 ofISOIIEC 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 ofSetpoints 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 .

ISO/IEC/IEEE STD 24765, Systems and Software Engineering - Vocabulary, International Organization for Standardization, Geneva, Switzerland, 20 10.

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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' ~ *-* ~ . NOllTHWHT MCDtCAL ISOTOP£S 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. ,

November 2012.

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.

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors - Format and Content, Part 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February 1996.

NUREG/CR-6090, The Programmable Logic Controller and Its Application in Nuclear Reactor Systems ,

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.,

September 1993 .

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.

NWMI-2015-SAFETY-002, Radioisotope Production Facility Integrated Safety Analysis Summary, Rev. 0, Northwest Medical Isotopes, Corvallis, Oregon, 2015 .

Regulatory Guide 1.53, Application of the Single-Failure Criterion to Safety Systems, Rev. 2, U.S. Nuclear Regulatory Commission, Washington, D.C., June 2003 .

Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, Washington, D.C., 2006.

Regulatory Guide 1.152, Criteria for Use of Computers in Safety Systems of Nuclear Power Plants ,

Rev. 3, U.S. Nuclear Regulatory Commission, Washington, D.C., June 2011.

Regulatory Guide 3. 71 , Nuclear Criticality Safety Standards for Fuels and Material Facilities ,

U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington, D.C., 2010.

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

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. NORTHWEST MEDICAL ISOTOPES Chapter 8.0 - Electrical Power Systems Construction Permit Application for Radioisotope Production Facility NWM 1-2013-021 , Rev. 2 August2017 Prepared by:

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

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• NO<<JHWUT MEDICAL ISOTOP£S NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power Systems Chapter 8.0 - Electrical Power Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 2 Date Published :

August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 2

Title:

Chapter 8.0 - Electrical Power Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power 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 2 8/5/2017 Modifications based on ACRS comments C. Haass

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~* * ~ NOkTlfWEST MfDN:.Al ISOTOPES NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power Systems CONTENTS 8.0 ELECTRICAL POWER SYSTEMS .. .......... .... .............. .. .. .............................. ...... ......... .............. 8- 1

8. 1 Normal Electrical Power Systems ....... ........................................ .......... ........ ..... ................ 8-2 8.1.1 Design Basis of the Normal Electric Power System ... ......................................... 8-4 8.1.2 Design for Safe Shutdown .... ......... ... ....... ..................... .. ............... ...................... 8-5 8.1.3 Ranges of Electri cal Power Required............ .. ............ .. ....................................... 8-5 8.1.4 Use of Substations Devoted Exclusively to the Radioisotope Production Facility .... ...... ... .. ........... ....... ............ ......... .. ................... .... ... ... ....... .. ..... ............ .. 8-6 8.1.5 Special Processing of Electrical Service .. .............. ..... ... .. ..... ........ ... .......... ........ .. 8-6 8.1.6 Design and Performance Specification ....................... ......... .. .......... ......... .... ....... 8-6 8.1.7 Special Routing or Isolation ... .. .. .... ........... .... .............. ... .... ... ... .. .......... .... ............ 8-6 8.1.8 Deviations from National Codes ............ ......... ......... ........ ........ ............................ 8-6 8.1.9 Technical Specifications .. ........ ................ ..... .... .......... .... ........ ................ ...... .. ..... 8-6 8.2 Emergency Electrical Power Systems ........ .. .......... ..................... .. ... .......... ... ..... ....... ......... 8-7 8.2.1 Design Basis of the Emergency Electric Power System ...................................... 8-8 8.2.2 Ranges of Emergency Electrical Power Required ... ........ ............ ............ ............ 8-8 8.2.3 Power for Safety-Related Instruments .......... ........... ....................... ..... ................ 8-8 8.2.4 Power for Effluent, Process, and Area Radi ati on Monitors ............... ............. ..... 8-8 8.2.5 Power for Physical Security Control, Information, and Communication Systems .......................... ............ ...... .... ....... ........... ... ... ....... .. ............... ................ 8-8 8.2.6 Power to Maintain Experimental Equipment in Safe Condition ... ..... .................. 8-8 8.2.7 Power for Active Confinement/Containment Engineered Safety Feature Equipment and Control Systems ............... .. .. ....... .............. ....... ........... ....... ......... 8-8 8.2.8 Power for Coolant Pumps or Systems ... .... .... ..... ...... ........... ... ....... ... ........... ........ . 8-9 8.2.9 Power for Emergency Cooling .................. ............................ ..... ................... ....... 8-9 8.2.1 0 Power for Engineered Safety Feature Equipment ....... ...... ...................... .. .......... . 8-9 8.2. 11 Power for Emergency Lighting ... .... ........ .. ... .. .... ...................... ................. ........... 8-9 8.2.12 Power for Instrumentation and Control Systems to Monitor Shutdown .... .......... 8-9 8.2.13 Technical Specifications ... ...... .......... .................................. ... ............. .. ............... 8-9 8.3 References ........ .... ............. .. .... .......... .. ................. ..... ................ ... .. .................. ..... .. ......... 8-10 8-i

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~ ~ *! . NORTHWEST MEDtCAl ISOTGnS FIGURES Figure 8-1. Radioisotope Production Facility Electrical One Line Diagram ............ .. ......... ..... ..... .... 8-3 TABLES Table 8-1. Summary of Radioisotope Production Facility and Ancillary Facilities Electrical Loads (2 pages) ............... .... ........................... ............ ....... .............. .. ........ .. ..... ................ 8-1 8-ii

NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems TERMS Acronyms and Abbreviations AEC active engineering control A TS automatic transfer switch CAAS criticality accident alarm system HV AC heating, ventilation, and air conditioning IEEE Institute of Electrical and Electronics Engineers IROFS item relied on for safety MCC motor control center NEP normal electrical power NFPA National Fire Protection Association NOx nitrogen oxides NWMI Northwest Medical Isotopes, LLC RPF Radioisotope Production Facility SEP standby electrical power UPS uninterruptable power supply Units gal gallon hp horsepower hr hour Hz hertz km kilometer kV kilovolt kW kilowatt L liter m1 mile mm minute sec second v volt 8-iii

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NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems 8.0 ELECTRICAL POWER SYSTEMS This chapter provides a description of the normal electrical power (NEP) and emergency electrical power systems within the Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF).

The RPF design uses high-quality, commercially available components and wiring in accordance with applicable code. Electrical power circuits will be isolated sufficiently to avoid electromagnetic interference with safety-related instrumentation and control functions. The facility is designed for passive, safe shutdown and to prevent uncontrolled release of radioactive material ifNEP is interrupted or lost. Uninterruptable power supplies (UPS) automatically provide power to systems that support the safety functions protecting workers and the public.

The NEP system is designed to provide reasonable assurance that use or malfunction of electrical power systems will not damage the RPF or prevent safe RPF shutdown. In addition, the RPF has a non-safety standby electrical power (SEP) system to reduce or eliminate process downtime due to electrical outages.

A combination ofUPSs and the SEP system will provide emergency electrical power (defined in Section 8.2) to the RPF.

Table 8-1 lists the RPF electrical loads, including the NEP system peak loads, which systems have UPSs, and the loads for those systems supported by the SEP system.

Table 8-1. Summary of Radioisotope Production Facility and Ancillary Facilities Electrical Loads (2 pages)

Normal electrical peak Standby electrical power load peak power load Uninterruptable Demand power Target fabrication system 125 168 No 0 0 Target receipt and disassembly system 30 40 No 0 0 Target dissolution system 40 54 No 40 54 Molybdenum recovery and purification 30 40 No 25 34 system Uranium recovery and recycle system 10 13 No 10 13 Waste handling system 25 34 No 5 7 Radiation monitoring and CAAS systems 5 7 Yes* 5 7 Standby electrical power system NIA No NIA NIA General facility electrical power 173 232 Yes* 101 135 Process vessel ventilation system 40 54 No 40 54 Facility ventilation system Ventilation Zone I 67 90 No 67 90 Ventilation Zone II/III 215 288 No 215 288 Ventilation Zone IV 295 396 No 295 396 Laboratory ventilation 38 51 No 10 13 Supply air 49 66 No 49 66 Fire protection system 0.8 1 Yes* Ob ob Plant and instrument air system 60 83 No 60 83 Gas supply system 0.8 I No 0.8 I Process chilled water system 280 375 No 140 188 8-1

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NWMI NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems

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Table 8-1. Summary of Radioisotope Production Facility and Ancillary Facilities Electrical Loads (2 pages)

Standby electrical Normal electrical peak power load peak power load Uninterruptable Demand power Facility chilled water system 1,300 1,743 No 0 0 Facility heated water system 47 63 No 0 0 Process stream system 0.8 No 0.8 1 Demineralized water system 0.8 No 0 0 Supply air system Chemical supply system 49 66 No 49 66 Facility process control and 5 7 Yes 5 7 communications systems Energy recovery 5 7 No 0 0 Safeguards and security 40 54 Yes 40 54 Administrative building 90 121 No 18 24 Waste management building 11 15 No 3 4 a Only parts of the system are provided with uninterruptable power supplies.

b The fire detection and fire alarm subsystems will be provided by an uninterruptable power supply with a 24-hr capacity. Chapter 9.0 provides additional detail.

CAAS = criticality accident alarm system NIA = not applicable.

8.1 NORMAL ELECTRICAL POWER SYSTEMS The NEP system will connect to electric utility power from the off-site utility transmission and distribution system at a point of common coupling. This point of common coupling will be located near the property line on the NWMI site. The NEP distribution system will operate in a redundant electrical system topology from the utility transmission and distribution system to the 480 volt (V) service entrance switchgear that services the RPF electrical distribution system and the devices and equipment within the facility. The RPF electrical distribution system is designed to support the safety functions protecting workers, the public, special nuclear material activities, and radioisotope production operation processes, as described in Chapter 4.0, "Radioisotope Production Facility Description," and to minimize the number of points where a failure in the RPF is a single point of power conveyance.

Figure 8-1 provides a preliminary electrical one-line diagrams for the electrical distribution topology.

The electrical one-line diagrams will be updated after completion of the RPF final design and included in the Operating License Application.

Power will be provided to the NWMI site from an underground utility feed 0 to the pad-mounted switchgear located outside of the RPF building. Power will then be routed underground from the switchgear to the Administrative Building 6 and the RPF t.

The underground feeders t to the RPF will comprise two redundant full-capacity service laterals to the RPF. Each service lateral will support redundant full-capacity service transformers 0 that will normally carry half the RPF load. Either of the RPF feeders can be opened and the tie breaker closed, as needed, allowing the other feeder to carry the entire RPF load.

Any RPF loads requiring SEP will be provided power from the diesel generator when required 0 .

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NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power Systems

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Figure 8-1. Radioisotope Production Facility Electrical One Line Diagram 8-3

NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems The two underground feeders will be located on each side of the switchgear and will normally carry approximately half of the electrical load. However, each underground feeder will be capable of carrying the entire load of the facility. The designed NEP topology will provide the RPF with redundancy. In addition, each underground feeder can be maintained and inspected independently, due to redundancy, while the RPF and associated safety functions are serviced with electrical power.

The 480 V service entrance equipment will have a main-tie-main arrangement on the service entrance electrical bus, with a service main on either end of a common bus. The common bus will be segregated by a tie-breaker. In normal mode operation, the two main breakers will be closed and the tie-breaker open. In the event one feeder is unavailable, the other feeder will carry the entire RPF load by opening the unavailable feeder main breaker and closing the tie breaker.

Electrical distribution on the load side of the 480 V service entrance switchgear and the heating, ventilation, and air conditioning (HV AC) redundant loads will be serviced from opposite sides of the switchgear through electrical equipment and feeders , including motor control centers (MCC),

switchboards, and distribution panel boards. Equipment, systems, and devices designed with redundant or N+ 1 capability will be fed from opposite sides of the service entrance switchgear. The planned loads on the MCC will be evaluated in the RPF final design to ensure the equipment is appropriately balanced.

These loads will be provided in the Operating License Application.

Systems requiring emergency electrical power in the event of the loss ofNEP will be serviced by an on-site diesel generator through the SEP system. Section 8.2 provides additional information on the SEP system.

UPSs will be provided for selected systems for the RPF, as identified in Table 8-1. UPS systems include unit device, rack-mounted, and/or larger capacity cabinet units (a large battery room as part of the UPS system is not planned). These UPS systems will service loads requiring uninterruptable power on a short-term basis. The UPS systems will be backed up by the on-site diesel generator to extend the duration of power available to connected loads. The UPS systems locations on the electrical one-line diagram will be defined in the RPF final design and provided in the Operating License Application.

Internal to the RPF and Administration Building, the NEP distribution system will service end user equipment and devices. Feeders, busing, overcurrent protection, devices, and equipment will provide the conveyance and conductor protection throughout the building. Design of the electrical distribution system includes recommended practices from the Institute of Electrical and Electronics Engineers (IEEE) 493 , Recommended Practice for the Design ofReliable Industrial and Commercial Power Systems, and IEEE 379, Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems. The electrical distribution system topology will employ a redundant power conveyance system.

The distribution system will include overcurrent protective devices, surge arresters, fusing, relays, and similar safety-related protective devices. These safety devices will conform to the requirements of the National Fire Protection Association (NFPA) 70, National Electric Code, relevant IEEE standards and recommendations, and local codes and standards.

8.1.1 Design Basis of the Normal Electric Power System The NEP system design basis will provide sufficient and reliable electrical power to the RPF systems and components requiring electrical power for normal operations, including the electrical requirements of the system, equipment, instrumentation, control, communication, and devices related to the safety functions and devices.

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NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power Systems There are no items relied on for safety (IROFS) applicable to the NEP, per Chapter 13.0, "Accident Analysis," Section 13 .2.5 (loss of power accident analysis scenario). The NEP will provide power to the active engineered control (AEC) systems through the instrumentation, monitoring, alarm, and related control systems. The design basis is provided in Chapter 3.0, "Design of Structures, Systems, and Components."

8.1.2 Design for Safe Shutdown In the event of the loss ofNEP, UPSs automatically provide power to the RPF systems and components that support the safety functions protecting workers and the public. The following systems and components are supported with UPSs:

• Process and facility monitoring and control systems

• Facility communication and security systems

• Emergency lighting

• Fire alarms

• Radiation protection and criticality accident alarm system (CAAS)

The UPSs will be designed to operate for a period of up to 120 minutes (min) or longer if identified as needed beyond 120 min in the final safety analysis. The fire protection system will have a UPS that provides 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (hr) of uninterrupted power. IfNEP service is reestablished within a determined timeframe (to be provided in the Operating License Application), the RPF will resume normal operation.

Upon loss of normal power:

• Inlet bubble-tight isolation dampers within the Zone I ventilation system will close, and the HVAC system will automatically be placed into the passive ventilation mode of operation

• The process vessel vent system will automatically be placed into the passive ventilation mode of operation, and all electrical heaters will cease operation as part of the passive operation mode

• Pressure-relief confinement system for the target dissolver offgas system will be activated on reaching the system relief setpoint, and dissolver offgas will be confined in the offgas piping, vessels, and pressure-relief tank

• Process vessel emergency purge system will be activated for hydrogen concentration control in tank vapor spaces

• Uranium concentrator condensate transfer line valves will be automatically configured to return condensate to the feed tank due to residual heating or cooling potential for transfer of process fluids to waste tanks

• Equipment providing a motive force for process activities will cease, including:

Pumps performing liquid transfers of process solutions Pumps supporting operation of the steam and cooling utility heat transfer fluids Equipment supporting physical transfer of items (primarily cranes) 8.1.3 Ranges of Electrical Power Required The RPF power service will be 480 V, 3-phase, 120 amp, 60 hertz (Hz). The total power required for the facility will be approximately 2,998 kilowatt (kW) (4,020 horsepower [hp]). Table 8-1 lists the loads for different locations and processes within the RPF.

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NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems 8.1.4 Use of Substations Devoted Exclusively to the Radioisotope Production Facility The RPF will receive power from Columbia Water and Light through the Grindstone Substation. This substation is approximately 2.4 kilometer (km) (1.5 miles [mi]) to the northwest of the RPF. The substation is 169 kilovolt (kV) that converts the current to 13,000 - 800 V for public distribution. The use of a shared substation will not affect the safe shutdown of the RPF.

8.1.5 Special Processing of Electrical Service Details on special processing of the electrical service, such as isolation, transformers, noise limiters, lightning arresters, or constant voltage transformers, will be provided in the Operating License Application.

8.1.6 Design and Performance Specification Design and performance specifications of principal and non-standard components will be provided in the Operating License Application.

8.1. 7 Special Routing or Isolation Special routing and isolation of wiring and circuits will be provided in the Operating License Application.

8.1.8 Deviations from National Codes The RPF electrical system will be designed to meet all required national codes and standards, as described in Chapter 3.0.

8.1.9 Technical Specifications As evaluated in Chapter 13.0, the RPF is designed to safely shut down without NEP for occupational safety and for protection of the public and environment. The NEP system will not require a technical specification per the guidelines in Chapter 14.0, "Technical Specifications."

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' ~* * ~ NOmfWlST MlOlCAL ISOTOPES NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems 8.2 EMERGENCY ELECTRICAL POWER SYSTEMS Emergency electrical power is defined by NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors - Format and Content, as any temporary substitute for normal electrical service. A combination of UPSs and the SEP system will provide emergency electrical power to the RPF, although only selected UPS systems will have a safety function. A 1,000 kW (1,341 hp) diesel generator will provide SEP.

Figure 8-1 also provides the electrical distribution topology for the SEP system. Power from this generator will service the RPF through an automatic transfer switch (ATS). The normal power side of the ATS will be connected to the RPF service entrance switchgear, with the load side of the ATS to be connected to the standby switchboard. The SEP system is designed to support the safety functions during RPF operations to protect workers, the public, and environment.

The SEP system design includes recommended practices from IEEE 446, Recommended Practice for Emergency and Standby Power Systems for Industrial and Commercial Applications, NFP A 110, Standard for Emergency and Standby Power Systems, IEEE 379, and IEEE 493.

The SEP system will include overcurrent protective devices, surge arresters, fusing, relays, and similar safety-related protective devices. These safety devices will conform to the requirements ofNFPA 70, relevant IEEE standards and recommendations, and local codes and standards.

SEP will be available to the exhaust system through a redundant electrical distribution topology.

Approximately half of the exhaust electrical load requiring standby will be connected to an MCC, with the other half connected to a redundant MCC.

The standby switchboard will service equipment and devices in the hot cell , control room, exhaust system ventilation system, and other loads requiring standby power. During switchover to the SEP, the loads will be sequenced to protect the generator and electrical equipment. Feeders, busing, overcurrent protection, devices, and equipment will provide the conveyance and conductor protection throughout the building.

During normal operations, loads connected to the standby switchboard will be serviced through the ATS with normal and facility electric power. In this way, any malfunctions of the SEP system during RPF operation with NEP will not interfere with normal RPF operations or prevent safe facility shutdown.

When the ATS senses a Joss of normal power, the switch will signal the on-site diesel generator to start up. When the diesel generator voltage and frequency are within acceptable limits, the ATS will switch from the normal power source to the diesel generator power source. Loads connected to the standby switchboard will continue to be serviced by the diesel generator until the normal power source returns.

The ATS will sense the normal power source voltage and frequency. Once the voltage and frequency are within acceptable limits and after a prescribed delay, the ATS will switch from the diesel generator power source to the normal power source.

UPSs wi ll be provided, as required. The function of the UPSs is to provide power to select loads while the diesel generator starts. The UPS systems will include unit devices, rack-mounted, and/or larger capacity cabinet units. The RPF loads requiring uninterruptable power on a short-term basis will be backed up by the on-site diesel generator to extend the duration of UPS power available to connected loads.

The 1,000 kW (1,341 hp) diesel generator will be serviced with a 3,785 liter (L) (1 ,000-gallon [gal])

diesel tank. This capacity will enable the generator to operate for 11-14 hr, depending on actual loads, without requiring additional fuel.

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• ..... . NWMl-2013-021, Rev. 2 Chapter 8.0 - Electrical Power Systems

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• NomtwESTMEl>>CALISOTOl(S 8.2.1 Design Basis of the Emergency Electric Power System The emergency electrical power system design basis is to provide uninterrupted power to instrumentation, control, communication systems, and devices required to support the safety functions protecting workers and the public, and to provide sufficient electrical power to the RPF to ensure safe shutdown in the event of loss ofNEP. The system design basis also provides SEP to operate select process-related equipment to limit the impacts of loss ofNEP on RPF production operations. Additional information on the design basis is provided in Chapter 3.0.

8.2.2 Ranges of Emergency Electrical Power Required The RPF power service is 480 V, 3-phase, 42 amp, 60 Hz. The total peak SEP for the RPF is 1, 178.6 kW (1,585 hp). Table 8-1 lists the backup peak electrical power loads for different locations and processes within the RPF .

8.2.3 Power for Safety-Related Instruments Safety-related instrumentation will be provided with UPSs. The UPSs will provide power to safety-related instruments while the diesel generator starts and will provide service loads requiring uninterruptable power on a short-term basis. The diesel generator will maintain power until the normal power system is operating within acceptable limits.

8.2.4 Power for Effluent, Process, and Area Radiation Monitors Effluent, process, and area radiation monitors will be provided with the UPSs. The UPSs will provide service loads requiring uninterruptable power for up to 120 min, while the diesel generator will maintain power until the normal power system is operating within acceptable limits.

8.2.5 Power for Physical Security Control, Information, and Communication Systems Physical security control, information, and communication systems will be provided with a UPS . The UPS provides service loads requiring uninterruptable power for up to 120 min, while the diesel generator will maintain power until the normal power system is operating within acceptable limits.

8.2.6 Power to Maintain Experimental Equipment in Safe Condition There are no experimental equipment or facilities in the RPF.

8.2. 7 Power for Active Confinement/Containment Engineered Safety Feature Equipment and Control Systems Based on the analysis in Chapter 13.0, the Zone I exhaust ventilation subsystems operations, equipment, and components ensures the confinement of hazardous materials during normal and abnormal conditions, including natural phenomena, fires , and explosions. After a loss ofNEP, the Zone I exhaust ventilation subsystem will automatically place itself into the passive mode, including inlet bubble-tight isolation dampers that close to provide passive confinement.

The system will remain in this configuration until the voltage and frequency of power from the diesel generator are within acceptable limits. At that point, the system can be manually started and operated in a reduced ventilation mode with one operating group ofHVAC fans and components. The Zone I exhaust ventilation subsystems are designed to function in a manner, whether operational or not, consistent with occupational safety and protection of workers, the public, and environment. Therefore, this system is not considered an IROFS .

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NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power Systems 8.2.8 Power for Coolant Pumps or Systems Based on the analysis provided in Chapter 5.0, "Coolant Systems," the coolant system is designed to function in a manner, whether operational or not, consistent with occupational safety and protection of the public and the environment. Therefore, power to coolant systems is not considered an IROFS .

8.2.9 Power for Emergency Cooling Based on the analysis provided in Chapter 5.0, an emergency cooling water system is not required.

8.2.10 Power for Engineered Safety Feature Equipment Engineered safety features requiring power will be provided with UPSs. The UPSs will provide service loads requiring uninterruptable power for up to 120 min. The di esel generator will maintain power until the normal power system is operating within acceptable limits. Additional information will be provided in the Operating License Application.

8.2.11 Power for Emergency Lighting Power required for emergency lighting will be provided by UPSs. The UPSs wi ll provide service loads requiring uninterruptable power for up to 120 min, while the diesel generator wi ll maintain power until the normal power system is operating within acceptable limits. Additional information will be provided in the Operating License Application.

8.2.12 Power for Instrumentation and Control Systems to Monitor Shutdown Power for instrumentation and control systems used to monitor safe shutdown will be provided with UPSs. The UPSs will provide service loads requiring uninterruptable power for up to 120 min, while the diesel generator will maintain power until the normal power system is operating within acceptable limits.

Additional information will be provided in the Operating License Application.

8.2.13 Technical Specifications As evaluated in Chapter 13.0, the RPF is designed to safely shut down without SEP consistent with occupational safety and protection of the public and the environment. The UPS systems, as required, are anticipated to be part of the technical specification for the system bei ng supported. The SEP system will not require a technical specification per the guidelines in Chapter 14.0.

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• • • • NWMl-2013-021 , Rev. 2 Chapter 8.0 - Electrical Power Systems

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8.3 REFERENCES

IEEE 379, Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2014.

IEEE 446, Recommended Practice for Emergency and Standby Power Systems for Industrial and Commercial Applications, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2014.

IEEE 493, Recommended Practice for the Design of Reliable Industrial and Commercial Power Systems (Gold Book), Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2007.

NFPA 70, National Electrical Code (NEC), National Fire Protection Association, Quincy, Massachusetts, 2014.

NFPA 110, Standard for Emergency and Standby Power Systems, Institute of Electrical and Electronics Engineers, Piscataway, New Jersey, 2014.

NUREG-1537, 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.

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. NORTHWEST MEDICAL ISOTOPES Chapter 10.0 - Experimental Facilities Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 August 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 10.0 - Experimental Facilities

! * *~ . NOITHWHT MEDtcAl lSOTOP(S Chapter 10.0 - Experimental Facilities Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 1

Title:

Chapter 10.0 - Experimental Facilities Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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' ~ * .' ~ NORTHWEST MEDtCAL tSOTOf'lS TERMS Acronyms and Abbreviations CFR Code of Federal Regulations NWMI Northwest Medical Isotopes, LLC RPF Radioisotope Production Facility

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NWMl-2013-021 , Rev. 1 Chapter 10.0 - Experimental Facilities

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NWMl-2013-021, Rev. 1 Chapter 10.0 - Experimental Facilities 10.0 EXPERIMENTAL FACILITIES This chapter of the Construction Permit Application addressing experimental facilities is not applicable to the Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF). Specifically, NWMI will not have any laboratory-scale facilities designed or used for experimental or analytical purposes that relate to the processing of irradiated materials containing special nuclear material per the definition of a production facility in Title 10, Code ofFederal Regulations (CFR), Part 50.2, "Definitions."

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~* * ~ NORTHWEST MEDICAl ISOTOttS NWM 1-2013-021 , Rev. 1 Chapter 10.0 - Experimental Facilities References 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register, as amended.

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. NORTHWEST MEDICAL ISOTOPES Chapter 11.0 - Radiation Protection and Waste Management Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 August2017 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 11 .0 - Radiation Protection and Waste Management Chapter 11.0 - Radiation Protection and Waste Management Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

August 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 1

Title:

Chapter 11.0 - Radiation Protection and Waste Management Construction Permit Application for Radioisotope Production Facility Signature:

Approved by: Carolyn Haass C wJ~('_f!/~

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NWMl-2013-021 , Rev . 1 Chapter 11 .0 - Radiation Protection and Waste Management REVISION HISTORY Rev Date Reason for Revision Revised By 0 6/29/2015 Initial Application Not required Incorporate changes based on responses to 1 8/5/2017 C. Haass NRC Requests for Additional Information

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• • • • Chapter 11 .0 - Radiation Protection and Waste Management NWMl-2013-021 , Rev. 1

~* * ~ NomtWEST MEDICAL ISOTOPES CONTENTS I 1.0 RADIATION PROTECTION AND WASTE MANAGEMENT ............................................. I I-I I I . I Radiation Protection ................... ...................... ............... .... .... .. .... .... ...... ...... .... ........... I I- I I I. I. I Radiation Sources ............................ ........ ............ ........ ....... .............. .... .. .. .. ... I I - I I 1.1.1.1 Airborne Radiation Sources ..................................................... ...... I I-2 I 1.1.1.2 Liquid Radioactive Sources .......... ............ .. ...................... ......... ..... I I -7 I 1.1.1.3 Solid Radioactive Sources ....... ............ .... .... .............. .. ................... I I-9 I 1.1.2 Radiation Protection Program ............... .... ............... ...... .................. .... ........ . I I- I 2 I 1.1.2. I Responsibilities of Key Program Personnel .................................. I I - I 3 I 1.1.2.2 Staffing of the Radiation Protection Program ............................... I I -I 5 I 1.1.2.3 Independence of the Radiation Protection Program ............... ...... .. I I- I 5 I 1.1.2.4 Radiation Safety Committee .... .................................... ..... ............ I I- I 5 I 1.1.2.5 Training Programs ............ .................. .......... ....... .... ...... .... .......... I I-I 5 I 1.1.2.6 Document Control... .. .......................... ............... .... ...... ...... .......... I I- I 7 I I. I .2.7 Audits .......... .................... .......... .................... .... ......... .... ............. I I-I 7 I 1.1.2.8 Radiation Work Control Procedures ................ .......... .... .... ........... 11- I 8 I I. I .2.9 Recordkeeping ......................... ............ .... .... ...... .. ........ .... ............ 11-18 l I . I .3 ALARA Program ................................. .. ......... ...... ..... ........... ..... ........ .......... I I- I 9 I I. I .3.I ALARA Policy .......................... ................. ................ .. ............... I I-I9 11.1.3.2 Approach to ALARA Program .... .................. .. ... ..... ...... ......... ...... I I- I 9 I I. I .4 Radiation Monitoring and Surveying ......................... .. ........... ........... ........... I I -22 l I. I .4.1 Monitoring Equipment ........ ........................ .... .... .... ...... ...... ......... 1 I-24 11.1.4.2 Technical Specifications ........ .................... .... .... .... .... ......... ...... ... I I-25 I 1.1.5 Radiation Exposure Control and Dosimeter .... ...... .. .... ............ ...................... I I-25 I 1.1.5.1 Process Design for ALARA .................... ........................... .......... I I-25 11.1.5.2 Facility Design for ALARA ............... ...... .... .. ...... .... ........ ............ I I-26 I l.1.5.3 Control of Entry .......................................... .. ..... ...... .... ................ 11-26 I 1.1.5.4 Protective Equipment and Materials ...... ..... .. ...... .... ...... ......... ....... 1 I-27 11.1.5.5 Radiological Areas ........................ ...... ........................ ............. .... 11-27 I 1.1.5.6 Personnel Monitoring and Assessment oflnternal and External Dose ... ......... ........ ....................... ............. ... ............ ...... ..... .. ........ 11-29

11. l.6 Contamination Control ...................................................................... .... ....... 11-30 11.1.6.1 Routine Monitoring to Detect Contamination ............................... I I-30 11.1.6.2 Access Control to Contaminated Areas .... .... .. ...... .... .................... 11-31 I 1.1.6.3 Anti-Contamination Techniques ................ ... ......... ..... .. ..... ........... I I-31 I I . I .6.4 Monitoring and Handling Contaminated Equipment and Components Outside Contaminated Areas .................. .... .............. 11-32 l I . I .6.5 Criteria for Classification of Contaminated Material, Equipment, and Working Areas .................. .. ... .......................... .. I I-32 I 1.1.6.6 Training Programs ................................... .... .... .... .............. .. ........ I I-32 11.1.6.7 Recordkeeping .................... .......... ........ ....................................... 11-32 11 .1.6.8 Technical Specifications ...... ............ ...... .......................... .... ........ l I-32 11.1 .7 Environmental Monitoring ...................... ................................. ...... ... ........ ... I 1-32 11.1.7.I Verification of Compliance ...... ...... ...... ........................................ I I-33 11.1.7.2 Identification of Potential Impacts .. ........ ................ ................ ...... I I-33 I 1.1. 7.3 Establishment of Baseline Environmental Quality ........................ l I-33 I 1.1.7.4 Environmental Surveillance Program ...... .... .... ...... ............... ........ I I-33 11-i

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........... Chapter 11 .0 - Radiation Protection and Waste Management NWMl-2013-021 , Rev. 1

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• NORTifWEST MEDtCAL lSOTOll'fS 11.2 Radioactive Waste Management ..................................... .... ........... .... .... .... ................ 11-36 11.2.1 Radioactive Waste Management Program .......... .. ..... ............ .. ....... ........ ..... .. 11-36 11.2.1.1 Waste Management Policy ..................... .. ....................... ...... .... ... 11-36 11.2.1.2 Waste Management Procedures ...... .. ............. ...... .. ... ......... .. ....... .. 11-36 11.2.1.3 Organizational Responsibilities ..... .. ....... .... ........... .. .... .... .... .... .. ... 11-36 11.2.1 .4 Training ................................................... ........... .. .... .... ...... .. ....... 11-37 11.2.1.5 Document Control and Record.keeping .. .... ................ ... ...... ... ....... 11-37 11.2.1.6 Reviews and Audits ..... .......... ............ ..... .... ............... ... ............. .. 11-37 11.2.1.7 Technical Specifications .................. ..... ..... ...... .. ..... ...... ....... ........ 11-37 11.2.2 Radioactive Waste Management Controls ........... .... ..... ....... ... .. ..................... 11-38 11.2.2.1 Waste Designation .. ................................ .............. ........ .... ... ........ 11-38 11.2.2.2 Waste Management Procedures .. ........................... .. ................ ..... 11-38 11 .2.2.3 Airborne Radioactive Waste Management.. ... ........... ..... ....... .... .... 11-38 11.2.3 Release of Radioactive Waste ................ ........ ........ ........ ....... ...... ....... ......... .. 11-39 11.2.3.1 Solid Radioactive Waste .................... ..... .... ..... .... .. .. .................... 11-39 11.2.3.2 Liquid Radioactive Waste .......... ...... .. ........ .... .... ...... ..... ...... .... ..... 11-40 11.2.3.3 Gaseous Radioactive Waste ........... ....... ..... .... ..... ......... ........ .. ...... I 1-40 11.3 Respiratory Protection Program ........... ........................ .. .................................... ..... ... 11-44 11.4 References .............................. .. .. ...... .... ... ... ..................... .... ....... ....... .... ... .. ....... ..... ... 11-4 7 11-ii

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! e*~ NORTHWUT ME.DM:Al ISOTOPES Chapter 11 .0 - Radiation Protection and Waste Management NWMl-2013-021 , Rev. 1 FIGURES Figure 11-1. Radioisotope Production Facility Airborne Radiation Source Areas ..... ..... ..... ........... . 11-2 Figure 11-2. Radioisotope Production Facility Radiation Zones (First Floor) .. ... ...... .. ........... ... .... 11-20 Figure 11-3 . Radioisotope Production Facility Radiation Zones (Second Floor) ........................... 11-21 Figure 11-4. Radioisotope Production Facility Radiation Zones (Basement) .................. .............. 11-21 Figure 11-5. Controlled and Unrestricted Areas ............................................... ...... ........... .... ... .... 11-29 Figure 11-6. Location of On-Site Environmental Thermo luminescent Dosimeters and Continuous Air Monitors ... ..................... ..... .. .......... .. ......... ..... ..... .. ..... .. ... ........ ....... 11-34 TABLES Table 11-1. Gaseous Radioactive Source (2 pages) ......... ... ........... ........... ... ....... .... ..... .................. 11-3 Table 11-2. Radionuclide Stack Release Source Term Input to COMPLY (2 pages) ..................... 11-6 Table 11-3. Liquid Radioactive Source (2 pages) ........................................................... ... .. .. ........ 11-8 Table 11-4. Solid Radioactive Source (2 pages) .. .... ................ .. ........... ..... ........ .... ............ .......... 11-10 Table 11-5. Estimated Radioisotope Production Facility Controlled and Restricted Area Dose Rates .......... ... ... ..... .... ........ ... ................... .... .. ........... ................. ... ...... .... .. ........ ....... 11-12 Table 11-6. Waste Produced in the Radioisotope Production Facility .... .. ........ .. ....... ... ................ 11 -39 Table 11-7. Low-Dose Radioactive Waste Sources ... .. .. ............... ......... .. ... .. .......... ............ ....... .. 11-41 Table 11-8. High-Dose Radioactive Waste Sources .. ... ... ... ......... ........................ ..... .......... ..... .... 11-42 Table 11-9. Encapsulated Solid Radioactive Waste Sources .... ....... ............. ... .... ............... .. ..... .. 11-43 11-iii

NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management TERMS Acronyms and Abbreviations 99Mo molybdenum-99 13 II iodine-131 ALARA as low as reasonably achievable ANS American Nuclear Society ANSI American National Standards Institute CAM continuous air monitor CFR Code of Federal Regulations CGA Compressed Gas Association coo Chief Operating Officer DAC derived air concentration DOT U.S . Department of Transportation EOI end of irradiation EPA U.S. Environment Protection Agency FSAR final safety analysis report G-M Geiger-Mueller HEGA high-efficiency gas adsorption HEPA high-efficiency particulate air HVAC heating, ventilation, and air conditioning ICRP International Commission on Radiological Protection Kr krypton LEU low-enriched uranium Mo molybdenum MURR University of Missouri Research Reactor Na OH sodium hydroxide NAVLAP National Voluntary Laboratory Accreditation NCRP National Council on Radiation Protection and Measurements NIOSH National Institute of Occupational Safety and Health NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission NWMI Northwest Medical Isotopes, LLC OSL optically stimulated luminescence OSTR Oregon State University TRIGA Reactor QAPP Quality Assurance Program Plan RPF Radioisotope Production Facility RWP radiation work permit SH&L safety, health, and licensing TEDE total effective dose equivalent TLD thermoluminescent dosimeter U.S. United States Xe xenon 11-iv

NWMl-2013-02 1, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management Units Bq becquerel Ci curie cm centimeter cm 2 square centimeter dpm disintegrations per minute ft feet ft3 cubic feet gal gallon hr hour

m. inch L liter m meter m3 cubic meter mg milligram mrem millirem mSv millisievert rem roentgen equivalent in man Sv sievert wk week yr year 11 -v

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...... NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management

~ * *! NOllT1fWEST MEDtCAl ISOTOP(S This page intentionally left blank.

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NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.0 RADIATION PROTECTION AND WASTE MANAGEMENT This chapter describes the Northwest Medical Isotopes, LLC (NWMI) radiation protection and waste management programs that are applied to the design of the Radioisotope Production Facility (RPF) and associated equipment and facility operations. The radiation protection program provides a complete list of all expected radiation and radioactive sources, including airborne, liquid, and solid sources. The radiation protection program also requires the development and implementation of procedures, identifies monitoring instrumentation and techniques, and specifies practices to be employed to verify compliance with the radiation dose limits and other applicable requirements . The basis and plans used to develop procedures for assessing and controlling radioactive wastes and the ALARA (as low as reasonable achievable) program are included.

This section also establishes the waste management program for radioactive wastes resulting from normal operations and maintenance of the RPF, including the required procedures to ensure that radiation exposures and releases ofradioactive materials are adequately assessed and controlled. The waste management program addresses the following elements:

Philosophy and approach to waste management Basis of procedures and technical specifications Organization, staffing, and associated training

• Document control and records management Review and audit committees for radioactive waste management activities Plans for shipping, disposal , and long-term waste storage 11.1 RADIATION PROTECTION The section identifies the sources of radiation within the RPF, including the physical and chemical form ,

type (e.g., neutron, gamma), curie strength or exposure rates, energy level, encapsulation (sealed or unsealed), use, storage conditions and locations, and planned program for disposal of the radioactive material.

11.1.1 Radiation Sources The RPF produces molybdenum-99 (99Mo) from low-enriched uranium (LEU) irradiated by a network of university research reactors. The primary RPF operations will include:

• Receiving LEU from the U.S . Department of Energy

• Producing LEU target materials and fabrication of targets

• Packaging and shipping LEU targets to the university reactor network for irradiation

• Returning irradiated LEU targets for dissolution, recovery, and purification of 99 Mo

• Recovering and recycling LEU to minimize radioactive, mixed, and hazardous waste generation

• Treating and packaging wastes generated by RPF processes to enable transport and disposal Along with the production of 99Mo, radioactive fission and activation products and actinides are also produced.

Chapter 4.0, "Radioisotope Production Facility Description," Table 4-41 , provides a summary of the estimated total radioactivity in curies (Ci) from fission products and actinides contained within each batch of irradiated LEU targets entering the RPF. The irradiated LEU targets entering the RPF provide the majority ofradiation sources within the RPF.

11-1

NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management This section identifies the source and nature of airborne, liquid, and solid radioactive materials within the RPF, including the types of radiation emitted (alpha, beta, gamma, and neutron).

11.1.1.1 Airborne Radiation Sources Airborne radioactive sources within the RPF will consist of radioactive gases produced during recovery and purification of 99Mo. Radioactive gases will originate from three areas in the RPF, which are shown in Figure 11-1:

• Target fabrication area

• Tank hot cell area, including:

Disassembly and dissolution of irradiated LEU targets Molybdenum (Mo) recovery and purification LEU recovery and recycle

• Waste management area A description of these areas and associated processes are provided in Chapter 4.0, Section 4.1.2.

Airborne radiation sources from process offgases are treated in two systems prior to discharge: the process vessel vent system, and the Zone 1 exhaust system.

Details of the process offgas system are provided in Chapter 9.0, Section 9.1.2. The process offgas system is designed to ensure that airborne radioactive releases are maintained at levels less than those defined in Table 2 of Title 10, Code of Federal Regu.lations, Part 20 (10 CFR 20), "Standards for Protection Against Radiation," Appendix B, "Annual Limits on Intake (ALis) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage."

Administration and support area 10 CFR 70 10 CFR 50 Figure 11-1. Radioisotope Production Facility Airborne Radiation Source Areas 11-2

NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.1.1.1.1 Gaseous Radioactive Source During normal operating conditions, airborne or gaseous radioactive materials will be contained within closed systems consisting of piping components and tanks. Table 11-1 provides source term information for the gaseous radioactive sources within the RPF (Barrington, 2015, and [Proprietary Information]).

This source is based on the gaseous effluent prior to entering the process vessel vent system.

Table 11-1. Gaseous Radioactive Source (2 pages)

Isotope 241 Am 136mBa 137mBa

' Principal radiations a

y y

Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

- 240pu 241 Pu 103mRh Principal radiations a

P-y Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

[Proprietary lnformation]

139Ba 105Rh P- [Proprietary Information] P- [Proprietary Information]

14oBa 106Rh P- [Proprietary Information] P- [Proprietary Information]

141ce 106mRh P- [Proprietary Information] P- [Proprietary Information]

143Ce 103 Ru P- [Proprietary Information] P- [Proprietary Information]

144Ce 105Ru P- [Proprietary Information] P- [Proprietary Information]

134Cs 106Ru P- [Proprietary Information] P- [Proprietary Information]

134mcs 122sb y [Proprietary Information] P- [Proprietary Information]

136Cs 124Sb P- [Proprietary Information] P- [Proprietary Information]

137 Cs P- [Proprietary Information] 125 Sb P- [Proprietary Information]

155E u P- [Proprietary Information] 126 Sb P- [Proprietary lnformation]

156Eu P- [Proprietary Information] 127 Sb P- [Proprietary Information]

151Eu 128Sb P- [Proprietary Information] P- [Proprietary Information]

1291 128msb P- [Proprietary Information] P- [Proprietary Information]

1301 129Sb P- [Proprietary Information] P- [Proprietary Information]

1311 P- 151sm

[Proprietary Information] P- [Proprietary Information]

132J P- [Proprietary Information] 153 Sm P- [Proprietary Information]

132ml 156Sm P- [Proprietary Information] P- [Proprietary Information]

133J P- [Proprietary lnformation] 89Sr P- [Proprietary Information]

133ml y [Proprietary Information] 9osr P- [Proprietary Information]

1341 P- [Proprietary Information] 91sr P- [Proprietary Information]

1351 P- [Proprietary Information] 92sr P- [Proprietary Information]

83mKr 99Tc y [Proprietary Information] P- [Proprietary lnformation]

85.Kr 99mrc P- [Proprietary Information] P- [Proprietary Information]

85mKr 125mre P- [Proprietary Information] y [Proprietary lnformation]

87.Kr P- [Proprietary Information] 127 Te P- [Proprietary Information]

88.Kr 121mre P- [Proprietary Information] P- [Proprietary Information]

140La P- [Proprietary Information] P- [Proprietary Information]

11-3

NWMl-2013-021 , Rev . 1 Chapter 11.0 - Radiation Protection and Waste Management Table 11-1. Gaseous Radioactive Source (2 pages)

Isotope 141La 142La 99Mo Principal radiations P-P-

P-Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

- 129mTe 1J1Te 13lmTe Principal radiations P-P-

P-Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

95Nb P- [Proprietary Information] inre P- [Proprietary Information]

95mNb P- [Proprietary Information] 133Te P- [Proprietary Information]

96Nb P- [Proprietary Information] 133mTe P- [Proprietary Information]

97Nb P- [Proprietary Information] 134Te P- [Proprietary Information]

97mNb y [Proprietary Information] mu a [Proprietary Information]

141Nd P- [Proprietary Information] 234u a [Proprietary Information]

236mNp P- [Proprietary Information] m u a [Proprietary Information]

231Np a [Proprietary Information] 236u a [Proprietary Information]

23sNp P- [Proprietary Information] 231u P- [Proprietary Information]

239Np P- [Proprietary Information] m u A [Proprietary Information]

233pa P- [Proprietary Information] 1J1mxe y [Proprietary Information]

133 234pa P- [Proprietary Information] Xe P- [Proprietary Information]

234mpa y [Proprietary Information] 133mxe y [Proprietary Information]

135 Xe 11 2pd P- [Proprietary Information] P- [Proprietary lnformation]

147Pm P- [Proprietary Information] 1Js mxe P- [Proprietary Information]

89my 148Pm P- [Proprietary Informati on] y [Proprietary Information]

148mPm y [Proprietary Information] 90y P- [Proprietary Information]

149pm P- [Proprietary lnformation] 90my P- [Proprietary Information]

150Pm P- [Proprietary Information] 91y P- [Proprietary Information]

91my 151Pm P- [Proprietary Information] y [Proprietary Information]

142Pr P- [Proprietary Information] 92y P- [Proprietary Information]

143 Pr P- [Proprietary Information] 93 y P- [Proprietary Information]

144Pr P- [Proprietary Information] 93Zr P- [Proprietary Information]

144mpr P- [Proprietary Information] 9szr P- [Proprietary Information]

145Pr P- [Proprietary Information] 97Zr P- [Proprietary Information]

238pu a [Proprietary Information] [Proprietary Information]

239Pu a [Proprietary Information] Total Ci [Proprietary Information]

Sources: Barrington, C., 2015, "NWMI Release # 11 - Process Vessel Ventilation (PVV) System Estimate," (memorandum to G. Dunford, May 26), AEM Consulting, LLC, Richland, Washington, 2015, and [Proprietary Lnformation).

a alpha. y = gamma.

p = beta .

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• • • NWMI NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management

' ~* * ~ NOlffifWESTM£DfCAl.ISOTOPfS 11.1.1.1.2 Release of Airborne Radionuclides 10 CFR 20.1101, "Radiation Protection Programs," paragraph ( d) requires the licensee to "implement the ALARA requirements of 10 CFR 20.1 lOl(b), and notwithstanding the requirements in 10 CFR 20.1301, a constraint on air emissions of radioactive material to the environment, excluding radon-222 and its daughter products, shall be established by licensees other than those subject to 10 CFR 50.34(a), such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent (TEDE) in excess of0.1 millisievert (mSv) (10 millirem [mrem]) per year from these emissions."

Regulatory Guide 4.20, Constraint on Releases ofAirborne Radioactive Materials to the Environment for Licensees Other than Power Reactors, provides guidance on methods that the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for meeting the constraint on airborne emissions of radioactive material to the environment, as described in 10 CFR 20.1101 (d). In 1996, the NRC added this constraint to 10 CFR 20 to remove dual regulation by the NRC and the U.S. Environmental Protection Agency (EPA) and to provide an "ample margin of safety" to the public from airborne emissions of radioactive material to the environment. As previously noted, the defined TEDE to a member of the public is not expected to exceed 0.1 mSv (10 mrem) per year from gaseous emissions released from the RPF into the unrestricted area. If a licensee subject to this requirement exceeds this dose constraint, the licensee shall report the exceedance as provided in 10 CFR 20.2203, "Reports of Exposures, Radiation Levels, and Concentrations of Radioactive Material Exceeding the Constraints or Limits," and promptly take appropriate corrective action to ensure against recurrence.

Regulatory Guide 4.20 notes that one method that the NRC staff considers acceptable for demonstrating compliance with 10 CFR 20.1101 (d) is the use of computer codes. To evaluate the dose associated with air emissions from the RPF and to ensure rates will meet the requirements of 10 CFR 20.1101 (d), the COMPLY computer model, Version 1.6, that assesses dose from airborne releases using varying amounts of site-specific information in four screening levels, was used. In Level 1, the simplest level, only the quantity of radioactive material processed during the monitoring period is entered. The calculations are based on generic parameters. Level 4 produces a more representative dose estimate and provides for a more complete treatment of air dispersion by requiring the greatest amount of site-specific information.

Licensees that do not pass at the lowest level in COMPLY must move to the next higher level until they can demonstrate compliance. The bases for the methods used in COMPLY are:

EPA 520/1-89-002, A Guide for Determining Compliance with the Clean Air Act Standards for Radionuclide Emissions from NRC-Licensed and Non-DOE Federal Facilities EPA 520/ 1-89-003, Users Guide for the COMPLY Code Level 4 of the COMPLY code was used to demonstrate compliance with 10 CFR 20.1101 ( d) for the RPF .

Table 11-2 provides the gaseous radionuclide stack release source term input to COMPLY. The source term was developed by combining the effiuent from each of the systems that is vented to the process vessel vent system and applying decontamination factors for each of the treatment subsystems (iodine absorbers, high-efficiency particulate air [HEP A] filters , etc.) (Barrington, 2015).

The source term calculations of airborne releases are based on the processing of eight targets at University of Missouri Research Reactor (MURR) and consistent with nominal operating conditions (i.e., irradiated targets beginning to be processed in the RPF at [Proprietary Information] for systems downstream of the impure uranium lag storage tanks). The primary dose contributor is the [Proprietary Information] noble gas, and the offgas system is designed to retain [Proprietary Information] to below release limits and bound the range of target processing. These calculations will be updated and described in the final safety analysis report (FSAR) as part of the Operating License Application.

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' ! * *! : NOIUMWUT MDMCAL ISOTDPU NWMl-2013-021 , Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management Table 11-2. Radionuclide Stack Release Source Term Input to COMPLY (2 pages)

• em*HM 241Am Release rate (Ci/yr) 8.67£-17 ll@i*MM 23 1Np Release rate (Ci/yr) 1.63£-12 11@(.J.lj 89Sr Release rate (Ci/yr) 3.69£-02 I36mBa 8.03£-08 23sNp 7.07£-10 9osr 3.02£-04 137mBa 2.89£-06 23 9Np 3.49£-04 91sr 2.36£-02 233pa 4.14£-13 92 Sr 5.57£-03 139Ba 1.03£-05 140Ba 7.99£-04 234pa 1.59£-12 99Tc 4.04E-08 141ce 6.02£-04 234mpa 1.23£-09 99mTc 3.29£-02 143Ce 3.52£-04 11 2pd 7.43£-07 12smTe 8.55£-07 144Ce 9.87£-05 147Pm 9.39£-06 121Te 9.76£-04 134 14Bpm 127mTe 1.1 2£-04 Cs 4.92£-10 5.81£-09 134mcs 2.08£-10 148mpm 4.62£-09 1.89£-03 136 149Pm Cs 7.16£-07 6.44E-05 9.73E-04 137 150Pm 6.75£-09 5.38£-04 Cs 3.06£-06 1ssEu 1.03£-07 151Pm 2.47£-05 2.39E-03 1s6Eu 2.06£-06 142Pr 2.22£-10 132Te 2.58£-02 1s1Eu 3.28£-07 143Pr 8.1 5E-04 133 Te 1.38£-05 1291 1.90£-13 144Pr 9.87£-05 133mTe 6.13£-05 130J 7.17£-08 144mpr 1.38£-06 134Te 1.78£-05 1311 5.97£-04 145Pr 1.1 4£-04 mu 2.63£-12 1132 1.56£-03 23Bpu l.3 lE-12 234U 2.48£-06 132ml 8.65£-08 239Pu 3.56£-09 23su 1.14£-07 1331 2.62£-03 24opu 2.58E-12 236u 3.82£-08 133ml 4.36£-07 241 Pu 9.31£-13 mu 2.40£-03 1341 2.69£-05 103mRh 2.74£-04 23su 7.15£-08 1351 1.36£-03 105Rh 6.40£-05 131mxe 1.66£+02 83 mKr 3.81£-10 106Rh 5.70£-06 133 Xe 4.98E+02 85Kr 5.84£+01 106mRh 1.05£-08 133mxe 9.77£-04 85mKr 1.92£-03 103Ru 2.75£-04 nsxe 9.51£-20 87Kr 1.80£-23 1osRu 2.12£-05 13smxe 1.46£-27 88Kr 1.16£-07 106Ru 5.70£-06 89my 3.43£-06 14oLa 122 90y 8.64£-04 sb 3.89£-11 2.94£-04 141La l.l lE-04 124Sb 6.56£-10 90my 5.70£-09 142La 1.24£-05 12ssb 1.67£-07 91y 4.18£-02 99Mo 3.40£-04 126Sb l.lOE-07 91my 1.50£-02 127 95Nb 1.68£-04 Sb 1.02£-05 2.03£-02 95mNb 4.64£-06 12ssb 9.42£-07 2.68£-02 96Nb 3.06£-08 12smsb 1.05£-07 6.12£-09 11-6

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...*.. NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management

~ * *! NORTHWEST MlDICAl. ISOTOPES Table 11-2. Radionuclide Stack Release Source Term Input to COMPLY (2 pages)

M!tf!U*J.!W Release rate (Ci/yr) M!@t.].!W Release rate (Ci/yr) l!@i*J.!M Release rate (Ci/yr) 97Nb 3.37E-04 129 Sb l .14E-05 9szr 4.29E-02 97mNb 2.98E-04 1s1sm 6.84E-08 97Zr 3.14E-02 147 153 Nd 2.81E-04 Sm 9.44E-06 Total Ci 7.24E+02 236 m Np 9.02E-15 156 S m 6.20E-07 Sources: Barrington, C., 2015 , "NWMI Release # 11 - Process Vessel Ventilation (PVV) System Estimate," (memorandum to G. Dunford, May 26), AEM Consulting, LLC, Richland, Washington, 2015 , and [Proprietary Information].

The weekly radionuclides (Ci/week) generated for the maximum dose case were multiplied by 52 weeks to obtain the release rates in Ci/year (yr) The radionuclide releases were adjusted to conservatively account for one HEPA filter in the Zone I heating, ventilation, and air conditioning (HV AC) offgas treatment system (Chapter 9, "Auxiliary Systems," Section 9.1.2.2) in accordance with EPA 520/1-89-003, which recommends that radionuclide particulate rel eases be reduced by an adjustment factor of 0.01. The noble gases and iodine releases were not reduced in the analysis. The following radionuclides were not available in the COMPL y database: I36mBa, 137mBa, I33mI, 97mNb, 236mNp, 234mPa, 112Pd, I44mPr, I06Rh, 128Sb, 12Bmsb, and 98my _

The following assumptions were used in the development of the analysis :

Meteorological data - COMPLY meteorological wind rose file for Columbia Stack data - Stack height 22.9 meter (m) (75 feet [ft]), diameter of 0.86 m (34 inches [in.])

Building data - Height 19.8 m (65 ft), width 24.4 m (80 ft), length 76 .2 m (250 ft)

Receptor location - Nearest receptor locations is the RPF fence line at 9.1 m (3 0 ft) from the stack

• Agricultural data - Food sources (e.g., milk, meat, vegetables) assumed to be home grown at receptor location The maximum dose to the public from the normal operational stack releases was calculated to be 0.036 mSv/yr (3 .6 mrem/yr) at 9.1 m (30 ft) from the RPF. The results of the COMPLY analysis determine that the requirement of 10 CFR 20.1101 ( d) will be met for the RPF , such that air emissions of radioactive material to the environment will not result in a member of the public receiving a TEDE in excess of 0.1 mSv/yr (10 mrem/yr) from these emissions.

The operating conditions in the above calculations are slightly more conservative than those described in Section 4.1.2.1 and will be aligned in the FSAR as part of the Operating License Application.

11.1.1.2 Liquid Radioactive Sources Liquid radioactive sources within the RPF will consist of process fluids created during Mo recovery and purification, liquids created during the recycling of LEU, and liquids resulting from the treatment of offgases. A detailed description of the RPF processes and associated locations are provided in Chapter 4.0, Section 4.3. Details of the process vessel vent system are provided in Chapter 9.0, Section 9.1.

During normal operating conditions, liquid radioactive materials will be contained within closed systems consisting of piping components and tanks. There will be no radioactive liquid discharges from the RPF into the sanitary sewer or the environment. Any liquid radioactive waste will be treated and/or solidified prior to being packaged and shipped to a disposal facility. Table 11-3 provides source term information for the liquid radioactive source. This source is based on the liquid in the target dissolver based

[Proprietary Information]. This source term is considered bounding ([Proprietary Information]).

11-7

NWMl-2013-021 , Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 241Arn 136mBa Principal radiations a

y Table 11-3 . Liquid Radioactive Source (2 pages)

[Proprietary Information] 241Pu Principal radiations A

P-Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

137mBa y [Proprietary Information] 103mRh y [Proprietary Information]

139Ba 105Rh P- [Proprietary Information] P- [Proprietary Information]

14oBa 106Rh P- [Proprietary Information] P- [Proprietary Information]

141ce 106mRh P- [Proprietary Information] P- [Proprietary Information]

143Ce 103Ru P- [Proprietary Information]

P- [Proprietary Information]

144Ce 1osRu P- [Proprietary Information]

P- [Proprietary Information]

134Cs 106Ru P- [Proprietary Information]

P- [Proprietary Information]

u4mcs y [Proprietary Information] 122sb P- [Proprietary Information]

136Cs 124Sb P-P- [Proprietary Information] [Proprietary Information]

137Cs P- [Proprietary Information] 12ssb P- [Proprietary Information]

1ssEu 126Sb P- [Proprietary Information]

P- [Proprietary Information]

1s6Eu P- [Proprietary Information] 121sb P- [Proprietary Information]

1s1Eu 128 Sb P- [Proprietary Information] P- [Proprietary Information]

1291 12smsb P-P- [Proprietary Information] [Proprietary Information]

1301 129Sb P- [Proprietary Information]

P- [Proprietary Information]

1311 151 Sm P- [Proprietary Information] P- [Proprietary Information]

153 Sm Il32 P- [Proprietary Information] P- [Proprietary Information]

132ml 1s6sm P- [Proprietary Information] P- [Proprietary Information]

1331 s9sr P- [Proprietary Information]

P- [Proprietary Information]

133ml 9osr y [Proprietary Information] P- [Proprietary Information]

1341 91sr P- [Proprietary Information] P- [Proprietary Information]

1351 92 P- [Proprietary Information] Sr P- [Proprietary Information]

83mKr y [Proprietary Information] 99Tc P- [Proprietary Information]

85Kr P- [Proprietary Information] 99mrc P- [Proprietary Information]

85mKr 12smre P- [Proprietary Information] y [Proprietary Information]

87Kr 127Te P- [Proprietary Information] P- [Proprietary Information]

88Kr 121mre P- [Proprietary Information] P- [Proprietary Information]

140La P- [Proprietary Information] 129Te P- [Proprietary Information]

141La 129mre P- [Proprietary Information] P- [Proprietary Information]

142La P- [Proprietary Information] 131Te P- [Proprietary Information]

99Mo 131mre P- [Proprietary Information] P- [Proprietary Information]

95Nb 132Te P- [Proprietary Information] P- [Proprietary Information]

95mNb 133 Te P- [Proprietary Information] P- [Proprietary Information]

96Nb 133mre P- [Proprietary Information] P- [Proprietary Information]

11-8

• ......NWMI

~.-.; *

. NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management

' ~* * ~ . NORTHWEST MEDICM. ISOTOPES

.. 97Nb 97mNb Principal radiations P-y Table 11-3. Liquid Radioactive Source (2 pages)

Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

- 134Te 232u Principal radiations P-a Radioactivity (Ci/wk)

[Proprietary Information]

[Proprietary Information]

141Nd P- [Proprietary Information] 234U a [Proprietary Information]

236mNp P- [Proprietary Information] mu a [Proprietary Information]

231Np a [Proprietary Information] 236u a [Proprietary Information]

23sNp P- [Proprietary Information] 237u P- [Proprietary Information]

239Np 23su P- [Proprietary Information] a [Proprietary Information]

233Pa P- [Proprietary Information] 1J1mxe y [Proprietary Information]

234pa 133 Xe P- [Proprietary Information] P- [Proprietary Information]

234mpa y [Proprietary Information] I33mxe y [Proprietary Information]

112pd 13sxe B- [Proprietary Information] P- [Proprietary Information]

147Pm I3smxe P- [Proprietary Information] P- [Proprietary Information]

14spm 89my P- [Proprietary Information] y [Proprietary Information]

148mpm 90y y [Proprietary Information] P- [Proprietary Information]

149Pm 90my P- [Proprietary Information] P- [Proprietary Information]

ISOPm 9Iy P- [Proprietary Information] P- [Proprietary Information]

IS I pm 9Jmy P- [Proprietary Information] y [Proprietary Information]

142Pr 92y P- [Proprietary Information] P- [Proprietary Information]

J43pr 93y P- [Proprietary Information] P- [Proprietary information]

144Pr 93zr P- [Proprietary Information] P- [Proprietary Information]

144mpr 9szr P- [Proprietary Information] P- [Proprietary Information]

145Pr 97Zr P-P- [Proprietary Information] [Proprietary Information]

238Pu a [Proprietary Information]

a [Proprietary Information] Total Ci [Proprietary Information]

Source: 0004, filtered dissolver product, in [Proprietary Information].

a alpha. y gamma.

p = beta.

11.1.1.3 Solid Radioactive Sources Solid radioactive sources within the RPF will consist of fresh LEU, irradiated LEU targets, LEU target material, and solidified waste. Details on these processes are provided in Chapter 4.0. During normal operating conditions, solid radioactive sources will be contained within tanks and shielded hot cells, all within restricted areas. Table 11-4 summarizes the solid radioactive source term in the RPF. This source term is based on the cumulative total of high- and low-dose waste, encapsulated waste, and one batch of

[Proprietary Information] .

11-9

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~

-- NWM I

~ * .* ~

• NOflTNWEST MtOICAl tSOTDHS NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Table 11-4. Solid Radioactive Source (2 pages)

Isotope 241Am Principal radiations a

Radioactivity (Ci/wk)

[Proprietary Information]

Isotope 240Pu rm.111 a

Radioactivity (Ci/wk)

[Proprietary Information]

136mBa y [Proprietary Information] 241 pu P- [Proprietary Information]

137mBa y [Proprietary Information] 103mRh y [Proprietary Information]

139Ba P- [Proprietary Information] 105Rh P- [Proprietary Information]

140Ba P- [Proprietary Information] 106Rh P- [Proprietary Information]

141ce P- [Proprietary Information] 106mRh P- [Proprietary Information]

143Ce P- [Proprietary Information] 103Ru P- [Proprietary Information]

144Ce P- [Proprietary Information] 1osRu P- [Proprietary Information]

134Cs P- [Proprietary Information] 106Ru P- [Proprietary Information]

134mcs y [Proprietary Information] 122sb P- [Proprietary Information]

136Cs P- [Proprietary Information] 124Sb P- [Proprietary Information]

137Cs P- [Proprietary Information] 125 Sb P- [Proprietary Information]

1ssEu P- [Proprietary Information] 126Sb P- [Proprietary Information]

1s6Eu P- [Proprietary Information] 121sb P- [Proprietary Information]

1s1Eu P- [Proprietary Information] 128 Sb P- [Proprietary Information]

1291 P- [Proprietary Information] 12smsb P- [Proprietary Information]

1301 P- [Proprietary Information] 129 Sb P- [Proprietary Information]

1311 P- [Proprietary Information] 1s1sm P- [Proprietary Information]

1321 P- [Proprietary Information] 153 Sm P- [Proprietary Information]

132ml P- [Proprietary Lnformation] 1s6sm P- [Proprietary Information]

1331 ~- [Proprietary Information] s9sr P- [Proprietary Information]

133mJ y [Proprietary Information] 9osr P- [Proprietary Information]

1341 P- [Proprietary Information] 91Sr P- [Proprietary Information]

135J P- [Proprietary Information] 92 Sr P- [Proprietary Information]

83mKr y [Proprietary Information] 99Tc P- [Proprietary Information]

85Kr P- [Proprietary Information] 99mTc P- [Proprietary Information]

85mKr 125mTe P- [Proprietary Information] y [Proprietary Information]

87Kr P- [Proprietary Information] 121Te P- [Proprietary Information]

88Kr ~- [Proprietary Information] 121mTe P- [Proprietary Information]

140La P- [Proprietary Information] 129Te P- [Proprietary Information]

141La P- [Proprietary Information] 129mTe P- [Proprietary Information]

142La P- [Proprietary Information] 131Te P- [Proprietary Information]

99Mo P- [Proprietary Information] 131mTe P- [Proprietary Information]

95Nb P- [Proprietary Information] 132Te P- [Proprietary Information]

95mNb

~- [Proprietary Information] 133Te P- [Proprietary Information]

11-10

NWMl-2013-021 , Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Table 11-4. Solid Radioactive Source (2 pages)

Isotope 96Nb Principal radiations Radioactivity (Ci/wk) Isotope t33mre 1111 Radioactivity (Ci/wk)

P- [Proprietary Information] P- [Proprietary Information]

97Nb P- t34Te

[Proprietary Information] P- [Proprietary Information]

97mNb y [Proprietary Information] 232u a [Proprietary Information]

t47Nd 234U P- [Proprietary Information] a [Proprietary Information]

236mNp 23su P- [Proprietary Information] a [Proprietary Information]

231Np a [Proprietary Information] 236U a [Proprietary Information]

23sNp 231u P- [Proprietary Information] P- [Proprietary Information]

239Np P- [Proprietary Information] mu a [Proprietary Information]

233Pa 1J1mxe P- [Proprietary Information] y [Proprietary Information]

234pa 133 P- [Proprietary Information] Xe P- [Proprietary Information]

234mpa y [Proprietary Information] t33mxe y [Proprietary Information]

112pd 135 P- [Proprietary Information] Xe P- [Proprietary Information]

147Pm 1Jsmxe P- [Proprietary Information] P- [Proprietary Information]

148Pm 89my P- [Proprietary Information] y [Proprietary Information]

148mpm 90y y [Proprietary Information] P- [Proprietary Information]

149Pm 90my P- [Proprietary Information] P- [Proprietary Information]

150Pm 9ty P- [Proprietary Information] P- [Proprietary Information]

151Pm 9Imy P- [Proprietary Information] y [Proprietary Information]

142Pr 92y P- [Proprietary Information] P- [Proprietary Information]

143Pr 93y P- [Proprietary Information] P- [Proprietary Information]

144Pr 93zr P- [Proprietary Information] P- [Proprietary Information]

144mPr 9szr P- [Proprietary Information] P- [Proprietary Information]

145Pr 97zr P- [Proprietary Information] P- [Proprietary Information]

23Bpu a [Proprietary Information] Total Ci [Proprietary Information]

239pu a [Proprietary Information]

Source: [Proprietary Information] and [Proprietary Information].

a a lpha. y gamma.

p = beta.

1 At discharge from the Oregon State University TRIGA Reactor (OSTR) (or third) reactor, the 30 targets will have essentially the same amount of radioactivity as eight targets being discharged from MURR.

Since the OSTR targets are not going to be received for 48 hr, the total radioactivity is significantly less than the eight MURR targets received in 8 hr. Therefore, other than grams of uranium, the radiation source for the 30 OSTR targets is lower.

1 TRJGA (Training, Research , Isotopes, General Atomics) is a registered trademark of General Atomics, San Diego, California.

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~* * ~ NOmfWUT MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management 11.1.2 Radiation Protection Program NWMI management is committed to protecting RPF workers, the public, and environment from unacceptable exposure to radiation sources. NWMl's policy is to conduct radiological operations in a manner that ensures the health and safety of employees, contractors, and the public. In achieving this objective, NWMI will ensure that radiation exposure to workers and the public, and releases of radioactivity to the environment, are maintained below regulatory limits. Deliberate actions will be taken to further reduce exposures and releases in accordance with a process focused on keeping exposures and releases ALARA.

NWMI is fully committed to implementing the NWMI radiation protection program to consistently reflect this policy. The NWMI radiation protection program will also meet the requirements of 10 CFR 20 and 10 CFR 19, "Notices, Instructions, and Reports to Workers: Inspection and Investigations."

The radiation protection program will be developed, documented, and implemented commensurate with the risks posed by RPF operations. NWMI will use, to the extent practicable, procedures and engineering controls based on sound radiation protection principles to achieve occupational doses and doses to the public that are ALARA.

The radiation protection program content and implementation will be reviewed at least annually, as required by 10 CFR 20. l lOl(c). In addition, in accordance with 10 CFR 20.llOl(d), constraints on atmospheric releases will be established for the RPF such that no member of the public would be expected to receive a TEDE in excess of 0.1 mSv/yr (I 0 mrem/yr) from these releases.

The NWMI RPF administrative Table 11-5. Estimated Radioisotope Production Facility exposure limits have been set below Controlled and Restricted Area Dose Rates the limits specified in 10 CFR 20 to ensure that regulatory radiation Location Dose rate exposure limits are not exceeded and to emphasize ALARA Target fabrication area 0.0003 mSv/hr 0.03 rnrem/hr principles. The administrative Irradiated target receipt area 0.005 mSv/hr 0. 5 rnrem/hr exposure limit (TEDE) for the RPF Hot cell operating and maintenance 0.005 mSv/hr 0.5 rnrem/hr is defined as 0.02 Sv/yr (2 roentgen galleries equivalent in man [rem]/yr). The Waste management loading area 0.005 mSv/hr 0.5 rnrem/hr ALARA goal and dose investigation level is set at Utility area 0 0 2

5 mSv/yr (500 mrem/yr). Laboratory area 0.005 mSv/hr 0.5 rnrem/hr Table 11-5 provides a conservative Administration and support area 0 0 estimate of the dose rates during normal operations for occupied areas. The high and very high radiation areas presented in Chapter 4, Section 4.2 will not be occupied by personnel during normal operations. Dose rates outside the controlled area are anticipated to be below 0.02 mSv (2 mrem) in any 1 hr and 0.001 Sv/yr (0.1 rem/yr) in accordance with 10 CFR 20.1302, "Compliance with Dose Limits for Individual Members of the Public."

2 A dose investigation level of 5 mSv/yr (500 rnrem/yr) is the TEDE above which would trigger an investigation by the Radiation Protection staff to determine why an individual received such a dose equivalent. The routine TEDE to workers is not anticipated to approach this level. An investigation might entail interviews with the individual and the immediate supervisor, review of radiation work permits (or equivalent), review of procedures, review of ALARA approaches, and providing feedback to management with recommendations on how to proceed.

11-12

NWMl-2013-021 , Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management The dose rates calculated for Table 11-5 are based on actual shielding calculations or were the goals/

endpoints of the shielding analysis. Table 11-5 will be updated in the FSAR as part of the Operating License Application when the final shielding design and calculations are completed. Areas identified as controlled access areas, restricted areas, radiation areas, and high radiation areas will also be designated, based on the definitions provided in 10 CFR 20, and the predicted doses rates presented by the shielding analysis. Although the Radiation Protection Plan has not yet been developed (i.e., this plan will be supplied with the Operating License Application), dosimetry is anticipated to be required in any restricted area.

In addition, although a dose rate of zero may not be achievable in the controlled areas, this is the goal. As stated in Section 11.1.5.5.2, an area monitoring program will be established in the controlled area to demonstrate compliance with public exposure limits in the FSAR as part of the Operating License Application.

To ensure that RPF personnel are protected, the radiation protection program will require surveillance of and control over the radiation exposure of personnel.

The radiation exposure policy and control measures for personnel will be established based on the requirements of 10 CFR 20 and the guidance of applicable regulatory guides. Recommendations from the International Commission on Radiological Protection (ICRP) and the National Council on Radiation Protection and Measurements (NCRP) will also be used in the formulation and evolution of the RPF radiation protection program.

Details on the area monitoring program will be provided in the FSAR as part of the Operating License Application. Area monitoring is anticipated to comprise a combination of passive (e.g., TLD or optically stimulated luminescence [OSL] monitors changed out monthly or quarterly) and active (e.g., energy-compensated Geiger-Mueller (G-M) detector systems with local and remote monitoring capability) monitoring systems located at points in the controlled area that would provide reasonable assurance that radiation areas are not present in the controlled area. The selection of specific instrumentation, range of detection, and alert/alarm setpoints will be consistent with the intent to detect radiation in areas where it should not be and alert personnel to this changing condition.

11.1.2.1 Responsibilities of Key Program Personnel This section provides the organizational structure of the NWMI radiation protection program, including the responsibilities of key personnel. Chapter 12.0, "Conduct of Operations," provides additional information on the facility management and technical organization. The NWMI Chief Operating Officer (COO) will have overall responsibility for the operation of the RPF , including radiation protection.

Detailed program procedures will be provided in the Operating License Application.

11.1.2.1.1 Plant Manager The Plant Manager will report to the COO and have direct responsibility for safe operation of the RPF ,

including the protection of workers and the public against radiation exposure resulting from facility operations and materials. Other responsibilities will include:

• Ensuring compliance with applicable NRC, State, and local regulations and license/permits

• Implementing the RPF conduct of operations program

• Establishing and managing the required training programs to support the operations organization 11-13

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~* * ~ NORTHWUT MEDtcAl. ISOTOrfS NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.1.2.1.2 Safety, Health and Licensing Manager The Safety, Health, and Licensing (SH&L) Manager will report to the COO, with overall responsibility for the development and implementation of programs addressing worker safety and health, NRC licensing, and State and local permitting (including monitoring compliance with those licenses and permits). Other responsibilities will include nuclear criticality safety, radiation protection/chemistry, environmental protection, integrated safety analysis, industrial hygiene and safety, chemical safety, fire protection, security, and emergency preparedness. With respect to operations, the SH&L Manager will be responsible to confirm the safety of those operations, and will have the authority to order facility shutdown for RPF operations that are judged to be unsafe for continued operations or noncompliant with applicable regulatory requirements and to approve restart of operations.

11.1.2.1.3 Radiation Protection Manager The Radiation Protection Manager will report to the SH&L Manager and have responsibility for the development and implementation of programs to limit personnel racliological exposures and environmental impacts associated with facility operations, including the ALARA program. Other responsibilities include implementation of the chemistry analysis programs and procedures for the RPF.

In matters involving radiological protection, the Radiation Protection Manager will have direct access to the COO. The Radiation Protection Manager, supported by his staff, will be responsible for:

• Establishing and implementing RPF radiation protection program Serving as Racliation Safety Officer Generating and maintaining procedures associated with radiation protection program Reviewing and auditing the efficacy of the program in complying with NRC and other government regulations and applicable regulatory guides Adequately staffing the Radiation Protection group to implement the radiation protection program Establishing and maintaining the ALARA program and ensuring personnel follow ALARA principles Establishing and maintaining a respirator usage program Monitoring worker doses, both internal and external Complying with raclioactive materials possession limits Performing calibration and quality assurance of all radiological instrumentation (e.g., verification of required lower limits of detection or alarm levels)

Establishing and maintaining a radiation safety training program Performing annual audits of the radiation protection program Establishing and maintaining the radiological environmental monitoring program Posting restricted areas and developing associated occupancy guidelines (e.g., racliation, airborne radioactivity, high radiation, and contaminated areas) 11.1.2.1.4 Operations Manager The Operations Manager will report to the Plant Manager and have responsibility for day-to-day RPF operations activities. Inherent in this responsibility is the assurance that operations are conducted safely and in compliance with license conditions.

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NWMl-2013-021 , Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.1.2.1.5 Employees NWMI employees will be responsible for conducting work safely and to follow the rules, regulations, and procedures that have been established for their protection and the protection of the public. Personnel whose duties require working with radioactive material, entering radiation areas, controlling facility operations that could affect effluent releases, or directing the activities of others, are trained such that they understand and effectively carry out their responsibilities.

11.1.2.2 Staffing of the Radiation Protection Program Only suitably trained radiation protection personnel will be employed at the RPF. For example, the Radiation Protection Manager will have a bachelor's degree (or equivalent), as a minimum, in an engineering or scientific field and 4 years of applicable nuclear experience. Other members of the radiation protection staff will be trained and qualified consistent with the guidance provided in American National Standards Institute (ANSl)/American Nuclear Society (ANS) 15 . 11 , Radiation Protection at Research Reactor Facilities. NWMI management is committed to providing sufficient resources in terms of staffing and equipment to implement an effective radiation protection program.

11.1.2.3 Independence of the Radiation Protection Program The radiation protection program will be independent of the RPF's routine operations. This independence ensures that radiation protection personnel maintain their objectivity and are focused on implementing sound radiation protection principles necessary to achieve occupational doses and doses to the public that are ALARA. As noted in Section 11 .1.2. 1.3, the Radiation Protection Manager has direct access to the COO for matters involving radiological protection.

11.1.2.4 Radiation Safety Committee A Radiation Safety Committee will meet periodically to review the status of projects, measure performance, look for trends, and review radiation safety aspects of facility operations, in accordance with 10 CFR 20.1101 (c). The Radiation Protection Manager will chair the Radiation Safety Committee.

The other Radiation Safety Committee members come from quality assurance, operations, maintenance, and technical support, as deemed appropriate by the Plant Manager.

The objectives of the Radiation Safety Committee wi ll be to maintain a high standard of radiation protection in all facility operations. The Radiation Safety Committee will review the content and implementation of the radiation protection program at a working level and strive to improve the program by reviewing exposure trends, the results of audits, regulatory inspections, worker suggestions, survey results, and exposure incidents. The maximum interval between meetings may not exceed 180 days. A written report of each Radiation Safety Committee meeting will be forwarded to all managers.

11.1.2.5 Training Programs The design and implementation of the radiation protection training program will comply with the requirements of 10 CFR 19.12, "Instruction to Workers." Records will be maintained in accordance with 10 CFR 20.2110, "Form of Record." The development and implementation of the radiation protection training program will be consistent with the guidance provided in the following regulatory guidance documents:

ASTM El 168-95, Radiological Protection Trainingfor Nuclear Facility Workers ANSI/ ANS 15 .11 , Radiation Protection at Research Reactor Facilities Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable 11-15

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*. NWMI NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management

~* * ~

• NORTHW£ST MEDICAL ISOTOPES Regulatory Guide 8.13, Instructions Concerning Prenatal Radiation Exposure

• Regulatory Guide 8.29, Instructions Concerning Risks from Occupational Radiation Exposure All personnel and visitors entering restricted areas, as described in Section 11 .1.5.5, will receive training that is commensurate with the radiological hazard to which they may be exposed. Visitors will be provided with trained escorts who have received radiation protection training.

The level of radiation protection training will be based on the potential radiological health risks associated with the employee's work responsibilities and incorporate the provisions of 10 CFR 19. 12. In accordance with 10 CFR 19.12, any individual working at the facility who is likely to receive a dose in excess of 1 mSv (I 00 mrem) in one year will be:

Kept informed of the storage, transfer, or use of radioactive material Instructed in health protection problems associated with exposure to radiation and radioactive material, in procedures to minimize exposure, and in the purposes and functions of protective devices employed Required to observe, to the extent within the worker ' s control, the applicable provisions ofNRC regulations and licenses for protection of personnel from exposure to radiation and radioactive material Instructed of the responsibility to promptly report to facility management any condition that may cause a violation of NRC regulations and licenses, or unnecessary exposure to radiation and radioactive material Instructed in the appropriate response to warnings made in the event of any unusual occurrence or malfunction that may involve exposure to radiation and radioactive material Advised of the various notifications and reports to individuals that a worker may request in accordance with 10 CFR 19.13, "Notifications and Reports to Individuals ."

The radiation protection training program will take into consideration a worker's normally assigned work activities. Abnormal situations involving exposure to radiation and radioactive material, which can reasonably be expected to occur during the life of the facility, will also be evaluated and factored into the training. The extent of these instructions will be commensurate with the radiological health protection considerations appropriate for the workplace.

Personnel who have previously been trained for radiological, chemical, industrial, and criticality safety will receive (retraining) refresher training at least annually. The retraining program will review procedure changes and any updates and changes in required skills. Changes to the training resulting from incidents potentially compromising safety or changes to the facility or processes will be incorporated as required.

Records of training are maintained in accordance with the NWMI records management system, which is described in the RPF Quality Assurance Program Plan (QAPP) (Chapter 12.0, Appendix C). Training programs will be established in accordance with Chapter 12.0, Section 12.10. The radiation protection sections of the training program are evaluated at least annually. The program content will be reviewed to ensure that it remains current and is adequate to ensure worker safety.

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...... NWMl-2013-021 , Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management

~ * *! NORTHWEST MUHCAl ISOTOf"U Radiation Protection Training The radiation protection training program will emphasize the importance placed on radiological safety of workers and the public. In-depth radiation protection training will be provided for the various types of job functions (e.g. , production operator, radiation protection technician, contractor personnel) commensurate with the radiation safety responsibilities associated with each position. Visitors to any of the RPF production areas will be trained commensurate with the planned activities and will be escorted by trained personnel while in those areas.

Personnel access procedures will be defined to ensure the completion of formal radiological safety training prior to being allowed unescorted access into the restricted areas. Training sessions covering safety, radiation protection, and emergency procedures will be conducted on a regular basis to accommodate new employees or those requiring retraining. Retraining will be conducted when necessary to address changes in policies, procedures, and requirements.

Specific topics covered in the training program are listed in Chapter 12.0, Section 12. l 0. The training provided will include the requirements of 10 CFR 19. Individuals attending these training classes must pass an initial examination covering the training contents to ensure an understanding of the requirements and to determine the effectiveness of the training. The effectiveness and adequacy of the training program curriculum and instructors will also be evaluated by audits performed by personnel responsible for safety and radiation protection.

If contractor employees are required to perform tasks in the restricted areas or controlled areas of the facility, formal training for these individuals will be designed to address the type of work being performed. In addition to applicable radiation safety topics, training contents may include radiation work permits (RWP), special bioassay sampling, and special precautions for activities performed.

Instructors approved by the Radiation Protection Manager will be responsible for conducting the radiation protection training program. The Radiation Protection Manager will be responsible for establishing and maintaining radiation protection training for all personnel, including contractor personnel who may be working at the facility. Records will be maintained by the Training Manager for each employee to document the training date, scope of the training, identity of the trainer(s), any test results , and other associated information.

Individuals requiring unescorted access to a restricted area receive annual retraining. Contents of the formal radiation protection training program will be reviewed and updated at least annually by the SH&L Manager and Radiation Protection Manager to ensure that the programs are current and adequate.

11.1.2.6 Document Control The RPF document control program is described in the RPF QAPP (Chapter 12.0, Appendix C).

11.1.2.7 Audits The radiation protection program will be audited annually, at a minimum, to review all functional elements of the program. This function is performed to meet the requirements of 10 CFR 20. 1 101 (c).

The audit activity is led by the Radiation Protection Manager, with the results being sent to the Radiation Safety Committee, COO, SH&L Manager, and Plant Manager for review. The corrective action program is used to address any deficiencies identified during the audit. Details on the corrective action plan are provided in the NWMI QAPP (Chapter 12.0, Appendix C).

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• i*:h NWMI

~* * ~ . NomrwuT MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.1.2.8 Radiation Work Control Procedures All work performed in restricted areas will be performed under an RWP. The procedures controlling RWPs will be consistent with the guidance provided in Regulatory Guide 8.10. An RWP may also be required whenever the Radiation Protection Manager deems that one is necessary or when activities involve licensed materials not covered by operating procedures where radioactivity levels are likely to exceed airborne radioactivity limits. Routine and nonroutine activities will be performed under an RWP that provides a description of the work to be performed (i.e. , defines the authorized activities). The RWP will summarize the results of recent dose rate surveys, contamination surveys, airborne radioactivity results, and other relevant information.

Precautions to be taken by those performing the task, including personal protective equipment to be worn while working (e.g., gloves, respirators, personnel monitoring devices), stay-times or dose limits for work in the area, recordkeeping requirements (e.g., time or dose spent on job), and the attendance of a radiation protection technician during the work, will be defined in the RWP. The Radiation Protection Manager or designee will approve the RWP. The designee must meet the requirements of Section 11 .1.2.2.

The RWPs will have a predetermined period of validity, with a specified expiration or termination time.

Standing RWPs will be issued for routinely performed activities, such as tours of the RPF by shift personnel.

Determining the need for issuing and closing out an RWP will be the responsibility of the Radiation Protection Manager, or designee. The RWP procedures will require the following:

Reviewing planned activities, changes to activities inside restricted areas, or work with licensed materials for the potential to cause radiation exposures that exceed action levels or produce raruoactive contamination. This review is also the responsibility of the Raruation Protection Manager, or designee.

Specifying requirements for any necessary safety controls, personnel monitoring devices, protective clothing, respiratory protective equipment, and air sampling equipment, and the attendance of raruation protection technicians at the work location Posting R WPs at access points to restricted areas, with copies of current R WPs posted at the work area location Clearly defining and limiting the work activities to which the RWPs apply Closing out the RWP when the applicable work activity for which it was written is completed and terminated Retaining the RWP as a record at least for the life of the facility 11.1.2.9 Recordkeeping For additional program commitments applicable to records and reports, the RPF will meet the requirements of the following regulations:

10 CFR 20 Subpart L, "Records," and Subpart M, "Reports" 10 CFR 50.71, "Maintenance of Records, Making of Reports" 10 CFR 70.51 , "Records Requirements" ANSI 15. 8, Quality Assurance Program Requirements for Research Reactors ANSI/ ANS 15. 11 , Radiation Protection at Research Reactor Facilities 11-18

.... . ...;. NWMI

.;........ NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management

' ~* * ~ NORTHWEST MEDICAL lSOTOPU The NWMI records management program is described in the NWMI QAPP (Chapter 12.0, Appendix C).

NWMI will maintain records of the radiation protection program (including program provisions, audits, and reviews of the program content and implementation), radiation survey results (air sampling, bioassays, external-exposure data from monitoring of individuals, internal intakes of radioactive material), and results of corrective action program referrals, R WPs, and planned special exposures.

By procedure, NWMI will report to the NRC any event that results in an occupational exposure to radiation exceeding the dose limits in l 0 CFR 20 within the time specified in 10 CFR 20.2202, "Notification oflncidents.". NWMI will prepare and submit an annual report of the results of individual monitoring to the NRC, as required by 10 CFR 20.2206, "Reports of Individual Monitoring."

11.1.3 ALARA Program This section provides a discussion of the NWMI ALARA program, as required by 10 CFR 20.110 I.

11.1.3.1 ALARA Policy NWMI management is committed to the ALARA philosophy for radiological operations. NWMI's policy is to conduct radiological operations in a manner that will ensure the health and safety of its employees, contractors, and the public. In achieving this objective, NWMI will ensure that radiation exposure to workers and the public, and releases of radioactivity to the environment, are maintained below regulatory limits. Deliberate actions will be taken to further reduce exposures and releases in accordance with a process focused on keeping exposures or releases ALARA. NWMI is fully committed to implementing an ALARA program that consistently reflects this policy.

11.1.3.2 Approach to ALARA Program As stated in the ALARA policy, NWMI management is committed to the implementation of an ALARA program. The objective of the program will be to make every reasonable effort to maintain facility exposures to radiation as far below the dose limits of 10 CFR 20.1201 , " Occupational Dose Limits," as practical, and to maintain radiation exposures to the public below the dose constraints of 10 CFR 20.1301 ,

"Dose Limits for Individual Members of the Public." Annual doses to personnel will be maintained ALARA. In addition, the annual collective dose to personnel (i.e. , the sum of all annual individual doses, expressed in person-Sv or person-rem) will be maintained ALARA. The dose equivalent to the embryo/fetus will be maintained below the limits of 10 CFR 20.1208, "Dose Equivalent to an Embryo/Fetus."

The goals of the ALARA program are to ensure occupational exposures and environmental releases are as far below regulatory limits as reasonably achievable. The RPF design incorporates ALARA principles .

As systems, components, and process areas are designed, radiation protection staff will evaluate the potential dose to workers and the public and provide suggested approaches to reducing dose.

Areas where facility personnel are expected to spend significant time are designed so that dose rates are maintained ALARA. The areas with higher doses rates will be minimized. Radiation areas will be established to minimize the spread of contamination and reduce unnecessary exposure of personnel to radiation. The potential radiation area designations within the RPF are shown in Figure 11-2, Figure 11-3, and Figure 11-4, and discussed further in Section 11.1.5 .5.

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[Proprietary Information]

Figure 11-3. Radioisotope Production Facility Radiation Zones (Second Floor)

[Proprietary Information]

Figure 11-4. Radioisotope Production Facility Radiation Zones (Basement) 11-21

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• NORTHWEST MEDICAL ISOTl>'ES The design and implementation of the ALARA program will be consistent with the guidance provided in:

Regulatory Guide 8.2, Administrative Practices in Radiation Surveys and Monitoring Regulatory Guide 8.13, Instructions Concerning Prenatal Radiation Exposure Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation Exposure Regulatory Guide 8.37, ALARA Levels for Effluents from Materials Facilities The operation of the RPF will be consistent with the guidance provided in Regulatory Guide 8.10. The guidance of Regulatory Guide 4.21 , Minimization of Contamination and Radioactive Waste Generation:

Life-Cycle Planning, will be followed to minimize, to the extent practicable, contamination of the facility and the environment, facilitate eventual decommissioning, and minimize, to the extent practicable, the generation of radioactive waste.

The radiation protection program will ensure that a comprehensive and effective program is implemented and document that policies are established to ensure that ALARA goals are met. Facility procedures will be written to incorporate ALARA principles into routine operations of the RPF and ensure that exposures are consistent with 10 CFR 20.1201 limits.

The Radiation Protection Manager will be responsible for implementing the ALARA program and ensuring that adequate resources are committed to support an effective program. An annual ALARA program evaluation report will be prepared that summarizes (I) radiological exposure and effluent release data for trends, (2) audits and inspections, (3) use, maintenance, and surveillance of equipment used for exposure and effluent control, and (4) other issues, as appropriate, that may influence the effectiveness of the radiation protection and ALARA programs. Copies of the report will be submitted to the COO, Radiation Safety Committee, and Plant Manager.

The Radiation Safety Committee will review the effectiveness of the ALARA program at least every quarter and determine if exposures, releases, and contamination levels are in accordance with ALARA principles. The committee will also evaluate the results of assessments made by the Radiation Protection organization and reports of facility radiation levels, contamination levels, and employee exposures for identified categories of workers and types of operations . The committee will be responsible for ensuring that the occupational radiation exposure dose limits of 10 CFR 20.1201 are not exceeded under normal operations and for evaluating alternatives to improve the effectiveness of equipment used for exposure control. The committee will determine if there are any upward trends in personnel exposures, environmental releases, and/or facility contamination levels.

11.1.4 Radiation Monitoring and Surveying Radiation monitoring and surveys will be conducted to (I) determine radiation levels, concentrations of radioactive materials, and potential radiological hazards that could be present in the facility , and (2) detect releases of radioactive material from facility equipment and operations. Radiation surveys will focus on those areas of the facility where the occupational radiation dose limits could potentially be exceeded.

Measurements of airborne radioactive material and/or bioassays will be used to determine that internal occupational exposures to radiation do not exceed the dose limits specified in 10 CFR 20, Subpart C, "Occupational Dose Limits ." NWMI has established written procedures to ensure compliance with the requirements of 10 CFR 20, Subpart F, " Surveys and Monitoring."

The radiation survey and monitoring programs will be consistent with the guidance provided in the following references:

Regulatory Guide 8.2, Administrative Practices in Radiation Surveys and Monitoring Regulatory Guide 8.4, Personnel Monitoring Device- Direct-Reading Pocket Dosimeters 11-22

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~* * ~ NOflTHWEST MEDICAL ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure Data Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program Regulatory Guide 8.25, Air Sampling in the Workplace Regulatory Guide 8.34, Monitoring Criteria and Methods to Calculate Occupational Radiation Doses

• ANSI N13.1, Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities ANSI Nl3.6, Practice for Occupational Radiation Exposure Records Systems

• ANSI Nl3.l l , Dosimetry-Personnel Dosimetry Performance-Criteria for Testing ANSI Nl3.27, Performance Requirements for Pocket-Sized Alarm Dosimeters and Alarm Ratemeters ANSI N323, Radiation Protection Instrumentation Test and Calibration ANSI/ ANS 15 .11, Radiation Protection at Research Reactors ANSI/HPS N13.22, Bioassay Program for Uranium ANSI/HPS Nl3.30, Performance Criteria for Radiobioassay NUREG-1400, A ir Sampling in the Workplace The procedures will include program objectives, sampling procedures, and data analysis methods.

Equipment selection will be based on the type of radiation being monitored. The procedures will be developed for each instrument used, including the frequency and method of calibration, and the maintenance and calibration requirements. Specific types of instruments used in the facility are discussed in Section 11.1.4.1. The survey program procedures will also specify the frequency of measurements and the recordkeeping and reporting requirements. Additional information on radiation monitoring and surveys will be provided in the Operating License Application.

All personnel who enter restricted areas will be required to wear National Voluntary Laboratory Accreditation (NA VLAP)-compliant personnel dosimeters.3 Personnel will also be required to survey themselves prior to exiting restricted areas that may have the potential for contamination.

Continuous airborne radioactivity monitors will provide indication of the airborne activity levels in the restricted areas of the facility. When deemed necessary, portable air samplers may be used to collect a sample on filter paper for subsequent analysis in the laboratory. Monitor data will be collected for regular analysis and documentation. Monitors will be equipped with alarms. The alarms activate when airborne radioactivity levels exceed predetermined limits. The limits will be set with consideration given to both toxicity and radioactivity. The volume of air sampled may have to be adjusted to ensure adequate sensitivity with minimum sampling time. The operating history of the RPF , changes in technology, changes in room functions and design, and changes in regulations may necessitate adjustment of the monitors.

3 Personnel dosimetry will provide a means to measure, assess, and record personnel exposures to ionizing radiation from external sources. Exposure to external sources ofradiation will be monitored by individual monitoring devices such as TLDs, OSL, CR-39, activation foils , or direct reading pocket dosimeters. Use of personnel dosimetry will be required for all personnel entering th e "restricted areas. Use of direct-reading personnel dosimetry and criticality monitoring will be required for all personnel entering "high radiation areas" and "very high radiation areas."

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• NCMtTHWUT MEDICAL ISOlWf.S NWMl-2013-021, Rev . 1 Chapter 11.0 - Radiation Protection and Waste Management Calibrations will be performed in accordance with written established procedures and documented prior to the initial use of each airflow measurement instrument (used to measure flow rates for air or effluent sampling) and each radioactivity measurement instrument. Periodic operability checks will also be performed in accordance with established written procedures.

Calibrations will be performed and documented on each airflow measurement and radioactivity measurement instrument, as follows:

At least annually (or according to manufacturers ' recommendations, whichever is more frequent)

After failing an operability check After modifications or repairs to the instrument that could affect its proper response When the instrument is believed to have been damaged Unreliable instruments will be removed from service until repairs are completed. Portal monitors, hand and foot monitors, and friskers will have the required sensitivity to detect alpha contamination on personnel to ensure that radioactive materials do not spread to the areas outside of the restricted areas.

Instruments will be calibrated with sources that are within +/-5 percent of the reference value and are traceable to the National Institute of Standards and Technology (NIST) or equivalent.

The background and efficiency of laboratory counting instruments, when used for radiation protection purposes, will be determined daily. This determination may be less :frequent only if necessary due to long counting intervals.

11.1.4.1 Monitoring Eq uipment The following subsections provide the procedures and equipment used at the RPF to routinely monitor and sample workplaces and other accessible locations to identify and control potential sources of radiation exposure and release.

11.1.4.1.1 Personnel Monitoring Three basic types of personnel monitoring equipment will be used at the facility : count rate meters (friskers), hand and foot monitors, and portal monitors.

Friskers - These devices typically consist of a handheld Thermo Scientific HP 210 (or equivalent) probe connected to an RM-25 (or equivalent) count rate meter. Instructions for the use of these instruments will be posted in a prominent location near the instrument. Handheld friskers will typically be placed in locations where conditions restrict the use of other monitors or for short-term use, as necessary, to ensure effective control of the spread of contamination.

Hand and foot monitors - These devices typically consist of multiple detectors arranged to monitor only hands and feet. Instructions for the use of these monitors will be prominently posted on or near the instrument. Hand and foot monitors will be used in applications where pass-throughs are frequent and where hand and foot monitoring is the major requirement.

Portal monitors - Portal monitors can quickly scan large surface areas of the body. Portal monitors will typically use large area beta and/or gamma sensitive detectors to monitor personnel passing through.

Additional detectors will be provided to monitor the hands, head, and feet. These monitors may be used where the number of personnel exiting an area, available space, etc. , makes their use advantageous.

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~-*~ ~MEDICAL ISOTOPES 11.1.4.1.2 Air Monitoring Continuous air monitors (CAM) will be provided within the RPF to provide indication of airborne activity. The CAMs will be operated to collect continuous samples. Portable CAMs may also be deployed when deemed necessary (e.g., non-standard maintenance activities). CAMs will be equipped with alarms. The alarm will be activated when airborne radioactivity levels exceed predetermined limits.

The limits will be set with consideration given to both toxicity and radioactivity. The volume of air sampled may have to be adjusted to ensure adequate sensitivity with minimum sampling time.

The operating history of the facility, along with changes in technology, room functions and design, and regulations, may necessitate adjustment of the monitors.

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 on air monitoring will be provided in the Operating License Application.

11.1.4.1.3 Radioactive Liquid Monitoring The RPF will not discharge radioactive liquids ; therefore, the RPF will not have liquid effluent monitors.

The monitoring of liquids within the RPF process systems is discussed in Chapter 4.0.

11.1.4.2 Technical Specifications Technical specifications associated with the contamination control are provided in Chapter 14.0, "Technical Specifications" and will be developed as part of the Operating License Application.

11.1.5 Radiation Exposure Control and Dosimeter The RPF is designed to prevent uncontrolled radiation releases to work areas or the environment during normal operations. This is accomplished through the process design and shielding discussed in Chapter 4.0 and the facility HVAC system design discussed in Chapter 9.0.

The goal of maintaining occupational internal and external radiation exposures ALARA encompasses an individual's dose and the collective dose of the entire working population. Because the TEDE is the sum of the internal and external exposures, the radiation protection program addresses both contamination control and external radiation protection. The fo llowing sections provide examples of how the RPF is designed to incorporate ALARA principles.

11.1.5.1 Process Design for ALARA Examples of process design and operating considerations that will be implemented to reduce personnel radiation exposures include the following :

Processing irradiated targets and purification of 99Mo under subatmospheric pressure (additional details are provided in Chapter 9.0, Section 9.1)

Limiting constant direct contact of personnel with radiological materials (additional detail provided in Chapter 4.0, Section 4.2)

Ensuring equipment and components are designed to include reliability, availability, maintainability, inspectability, constructability, and other design features to reduce or eliminate the need for repair or preventive maintenance Providing design redundancy of equipment or components to reduce the need for immediate repair when radiation levels may be high or when there is no feasible method avai lable to reduce radiation levels 11-25

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~e *~. NOtfTHWtST MIDK:Al ISOTOPH NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Designing equipment and piping to minimize the accumulation of radioactive materials Providing for draining, flushing, or, if necessary, remote cleaning or decontamination of equipment containing radioactive materials Designing airflow rates at exhausted enclosures and close-capture points, when in use, to preclude escape of airborne radioactive gases and particles and to minimize the potential for intake by workers. Airflow rates will be checked monthly when in use and after modification of any hood, exhausted enclosure, close-capture point equipment, or ventilation system serving these barriers. Additional detail is provided in Chapter 9.0, Section 9.1.

Handling accidental radioactive contamination by using the personnel decontamination room. A hand-washing sink and a shower will be provided for contamination removal.

11.1.5.2 Facility Design for ALARA Examples of facility design and operating considerations that are being implemented to reduce personnel radiation exposures include the following :

Incorporating ease of maintenance or repair, including ease of disassembly and modularization of components for replacement or removal to a lower radiation area for repair or disposal Providing the ability to remotely or mechanically operate, repair, service, monitor, or inspect equipment Laying out the facility so that access to a given area does not require passing through a higher radiation zone area Providing the ability to operate equipment from accessible areas both during normal and off-normal operating conditions Providing areas outside of high radiation areas that equipment can be moved to for service Incorporating ease of decontamination of potentially contaminated areas Providing control systems so that process controls (e.g. , essential instrumentation and controls) will be from the lowest radiation zone practicable Controlling HYAC system contamination by maintaining ventilation air flow patterns from areas of lower radioactivity to areas of higher radioactivity Requiring self-monitoring when exiting restricted areas; if contamination is detected, facility personnel will be required to notify radiation protection staff Training facility personnel in emergency evacuation procedures per the Emergency Preparedness Plan (Chapter 12.0, Appendix B)

All personnel whose duties require entry into restricted areas will wear individual external dosimetry devices (e.g. , passive dosimeters such as thermoluminescent dosimeters [TLD] that are sensitive to beta, gamma, and neutron radiation). External dosimetry devices will be evaluated at least quarterly to ascertain external exposures.

11.1.5.3 Control of Entry The RPF will include areas locked to limit access, and alarms and signals that alert workers to or prevent unauthorized entry into radiation areas, high radiation areas, and very high radiation areas. Additional information on control of entry will be provided in the Operating License Application.

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' ~* *~ HOflTMWEST MEDtCAl. ISOTOPES NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management As shown in Section 11.1.5.5.2, Figure 11-5, the entire RFP is considered a "controlled area."

Figure 11-2 shows five doors from the outside of the RFP to entrances into the "restricted area." Each door will have two credential access controls (e.g., fob/PIN, fob/biometric, or biometric/PIN). The RPF radiation protection program will require personnel to access assigned dosimetry and portable survey instrumentation (as needed, based on the work authorized) from an as-yet unspecified location within the RPF administrative area before entering the restricted area. Portal survey monitoring will be in place at the exit from the restricted area into the administrative area. The specifics on the type and instrument used will be described in the FSAR as part of the Operating License Application and will either be a control that allows standing passive detection or hand and foot monitors.

11.1.5.4 Protective Equipment and Materials Personnel working within the restricted area will be required to wear appropriate personal protective clothing.

Protective clothing, as prescribed by the RWP, will be selected based on the contamination level in the work area, anticipated work activity, worker health considerations, and consideration for nonradiological hazards present. Protective clothing of the following types will be made readily available as necessary:

Cloth and disposable coveralls Nonpermeable coveralls (plastic/rubberized)

Rubber and disposable shoe covers

• Rubber and disposable gloves Cotton liners Cloth and disposable hoods Full-face particulate respirators Eye goggles Face shields

• Supplied-air respirators Self-contained breathing apparatus Areas requiring protective clothing will be posted at each of the associated entry points. Radiation protection management and technical staff will be responsible for determining the need for protective clothing in each work area and documenting the requirements in the RWP.

Based on air sampling results and work evolutions, the Radiological Protection Manager will select the appropriate respiratory protection required. Airborne radioactivity concentrations will be minimized to the extent practical by the use of engineered controls (e.g. , containment, ventilation). When establishing radiological controls for work involving potential airborne radioactivity, the first consideration will be to use techniques that help prevent or reduce the potential for airborne radioactivity and maintain loose surface contamination in controlled areas within ALARA levels. Respiratory protection equipment requirements will be specified on the area RWP.

10 CFR 20, Subpart H, "Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas," defines the required elements of the facility respiratory protection program. Additional information on implementing and maintaining the respirator program are provided in Section 11.3.

11.1.5.5 Radiological Areas Radiological areas will be established to control (1) the spread of contamination, (2) personnel access to avoid unnecessary exposure of personnel to radiation, and (3) access to radioactive sources present in the facility. Table 11-5 lists the general dose rate estimates for the RPF. These dose estimates are based on shielding calculations (NWMl-2015-SHIELD-001, Radioisotope Production Facility Shielding Analysis).

Areas where facility personnel spend substantial amounts of time are designed, in accordance with ALARA principles, to minimize the exposure received when routine tasks are performed.

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~* * ~ . NomfWEST MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management The radiological areas within the RPF are shown in Figure 11-2 through Figure 11-4 (Section 11.1.3.2).

The following subsections provide the definitions of each of these areas and a description of how the RPF radiation protection program will be implemented to protect workers and the public.

11.1.5.5.1 Restricted Areas The NRC defines a restricted area as an area in which access is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Within the RPF, access to and egress from a restricted area will be through a radiation protection control point.

Monitoring equipment will be located at these points. All personnel will be required to self-monitor prior to exiting restricted areas that have the potential for contamination. Personnel who have not been trained in radiation protection procedures will not be allowed to access a restricted area without escort by trained personnel.

Additional temporary or permanent areas may be defined within the restricted area , including radiation areas, high radiation areas, airborne radioactivity areas, and contaminated areas. The radiation areas shown in Figure 11-2, Figure 11-3, and Figure 11-4 comprise the restricted areas within the RPF . These areas are defined in 10 CFR 20. l 003 , "Definitions." The entire basement and second story areas will be restricted. The areas will be posted in accordance with the requirements of 10 CFR 20 to inform workers of the potential hazards in the area and to help prevent the spread of contamination.

Radiation area - A radiation area (shown in green in Section 11 .1.3.2, Figure 11-2 and Figure 11-3) is where radiation levels could result in an individual receiving a dose equivalent in excess of 0.05 mSv (5 mrem) in 1 hr at 30 centimeters (cm) (11.8 in.) from the radiation source or from any surface that the radiation penetrates.

High radiation area - A high radiation area (shown in tan in Figure 11-2 and Figure 11-3) is an area accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 1 mSv (100 mrem) in 1 hr at 30 cm (11.8 in.) from the radiation source or from any surface that the radiation penetrates. Within the RPF, these areas will not normally be accessible to individuals during routine operations.

Very high radiation area - The NRC defines very high radiation areas (shown in pink in Figure 11-2, Figure 11-3, and Figure 11-4) as areas accessible to individuals, in which radiation levels exceed 5 sievert (Sv) (500 rem) in 1 hr at 1 m from the source or from any surface that the radiation penetrates (10 CFR 20.1003). The hot cells within the RPF (not normally accessible) are an example of a very high radiation area. The hot cells will be radiologically shielded and isolated from access to individuals by the use of engineered physical barriers, including structural shield blocks and locked shield doors.

Airborne radioactivity area - An airborne radioactive area is an area, room, or enclosure where airborne radioactive materials either exist in concentrations that exceed the derived air concentrations (DAC) specified in 10 CFR 20, Appendix B, or where an individual present in the area without respiratory protection equipment could exceed, during the hours the individual is present in a week, an intake of 0.6 percent of the annual limit on intake or 12 DAC-hr. There are no identified permanent airborne radioactive areas with the RPF .

11.1.5.5.2 Controlled Area The NRC defines a controlled area as an area outside of a restricted area but inside the site boundary, in which the licensee can limit access for any reason. For the RPF, the controlled area is the area within the perimeter fence but outside the restricted area and the Administrative Building, as shown in Figure 11-5.

The area fence will limit public access to the controlled area of the site. Training for access to a controlled area will be provided commensurate with the radiological hazard.

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' ~* * ~ . NOKTHWf.ST MEOtCAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Site visitors will include delivery people, site visitors, and service personnel who are temporary, transient occupants of the I controlled area. Area monitoring will demonstrate compliance with public exposure I

I limits for such visitors. All NWMI personnel I or contractor employees who work only in the I controlled area will be subject to the exposure limits for the public, as stated in /

10 CFR 20.1301.

Details on the area monitoring program will be provided in the FSAR as part of the Operating License Application. Area monitoring is anticipated to comprise a N

combination of passive (e.g., TLD or OSL monitors changed out monthly or quarterly) and active (e.g. , energy-compensated G-M detector systems with local and remote monitoring capability) monitoring systems t

Unrestricted area located at points in the controlled area that will provide reasonable assurance that Controlled area

\

radiation areas are not present in the controlled area. The selection of specific Figure 11-5. Controlled and Unrestricted Areas instrumentation, range of detection, and alert/alarm setpoints will be consistent with the intent to detect radiation in areas where it should not be and alert personnel to this changing condition.

11.1.5.5.3 Unrestricted Areas The NRC defines an unrestricted area as an area that is not controlled or limited by the licensee. For the RPF, the areas not specifically included within the definition of restricted and controlled areas will be considered unrestricted areas, as shown in Figure 11-5. These areas can be accessed by facility personnel and by the public. The unrestricted area is governed by the limits in 10 CFR 20.1301 , with the TEDE to individuals from the licensed operation not to exceed 1 mSv (100 mrem) in a year (exclusive of background radiation) nor exceed 0.02 mSv (2 mrem) in any one hour.

11.1.5.6 Personnel Monitoring and Assessment oflnternal and External Dose The following subsections describe personnel monitoring and provide an assessment of internal and external dose.

11.1.5.6.1 Internal Dose Internal exposures for selected personnel are evaluated via direct bioassay (e.g. in vivo body counting),

indirect bioassay (e.g., urinalysis), or an equivalent technique.

For soluble (Class D) uranium, 10 CFR 20.1201(e) limits worker intake to no more than 10 milligrams (mg) of soluble uranium in a week. This limit is to protect workers from the toxic chemical effects of inhaling Class D uranium. The RPF annual administrative limit for the TEDE will be 0.02 Sv (2 rem).

Internal doses will be evaluated at least annually.

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~* * ~ NORTHWEST MEOtCAl. ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management Continuous air monitoring in airborne radioactivity areas may be performed to complement the bioassay program. Alarm setpoints on the CAMs in the airborne radioactivity areas may be used to provide an indication that internal exposures may be approaching the action limit.

If the facility annual administrative limit is exceeded, as determined from bioassay results, an investigation will be performed to determine what types of activities may have contributed to the worker's internal exposure. The action limit is based on ALARA principles. This investigation may include procedural reviews, efficiency studies of the air-handling system, and work practices, and the results will be documented.

11.1.5.6.2 External Dose External dose will primarily be received from the fission products produced from irradiated targets and associated processing. The radionuclides of significance are identified in Section 11.1.1. All personnel whose duties require entry into restricted areas will wear individual external dosimetry devices (e.g. ,

passive dosimeters such as TLDs that are sensitive to beta, gamma, and neutron radiation). External dosimetry devices will be evaluated at least quarterly to ascertain external exposures. The ALARA goal on radiation exposure is set at 5 mSv/yr (500 mrem/yr) based on an administrative limit of 10 percent of the NRC limit of0.05 Sv/yr (5 rem/yr) given in IO CFR 20.1201.

If 25 percent of the ALARA goal (1.25 mSv [125 mrem]) is exceeded in any quarter, an investigation will be performed to determine what types of activities may have contributed to the worker's external exposure. This investigation may include procedural reviews, efficiency studies of the air-handling system, cylinder storage protocol, and work practices, and the results will be documented.

The Radiation Protection Manager will be informed whenever an administrative limit is exceeded. The Radiation Protection Manager will be responsible for determining the need for and recommending investigations or corrective actions to the responsible manager(s). Copies of the Radiation Protection Manager's recommendations will be provided to the Radiation Safety Committee.

11.1.6 Contamination Control Contamination will consist of two types:

Loose (removable) contamination, which can be removed from surfaces by smears and may contribute to airborne radioactivity and/or personnel contamination from routine activities . Loose contamination poses both an internal and external radiation hazard.

Fixed contamination, which is not smearable and may only be reduced by using approved decontamination techniques, procedures, and equipment. Fixed contamination does not readily contribute to airborne radioactivity and/or personnel contamination from routine activities. Fixed contamination poses an external radiation hazard.

When establishing radiological controls for work involving potential loose or airborne contamination, the first consideration is to use techniques that will help prevent or reduce the potential for airborne radioactivity and to maintain loose surface contamination in controlled areas within ALARA levels.

11.1.6.1 Routine Monitoring to Detect Contamination Contamination survey monitoring will be performed for all process areas and areas in which radioactive materials are handled or stored. Surveys will include routine checks of non-process areas, including areas normally not contaminated. Monitoring will include direct radiation and removable contamination measurements. Survey procedures will be based on the potential for contamination of an area and operational experience. All restricted areas will be surveyed at least weekly.

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NWMl-2013-021, Rev . 1 Chapter 11 .0 - Radiation Protection and Waste Management The change rooms will be surveyed at least daily. Various instruments (e.g., proportional counters and thin window G-M tubes) will be used at the RPF to evaluate contamination levels. Additional information on routine contamination survey monitoring will be provided in the Operating License Application.

11.1.6.2 Access Control to Contaminated Areas The access control program will be established to ensure that:

Signs, labels, and other access controls are properly posted and operative Restricted areas prevent spread of contamination and have appropriate signage Step-off pads, change facilities, protective clothing facilities, and personnel monitoring instruments are provided in sufficient quantities and locations For other areas, access control will be managed by administrative methods. Access to certain areas may be physically prevented for security reasons. Personnel who have not been trained in radiation protection procedures will not be allowed access to a restricted area without escort by other trained personnel.

Access to and egress from a restricted area will be through one of the monitor stations at the particular restricted area boundary. Access to and egress from each radiation area, contaminated area, or airborne radioactivity area within the restricted area may also be individually controlled. A contamination monitor (e.g. , frisker, hand and foot monitor, or portal monjtor), step-off pad, and container for any discarded protective clothing may be provided at the egress point from certain areas to prevent the spread of contamination.

Action levels for skin and personal clothing contamination at the point of egress from restricted areas and any additional designated areas within the restricted area (e.g., a contaminated area that is provided with a step-off pad and contamination monitor) will not exceed 2.5 becquerel (Bq)/100 square centimeter (cm 2 )

(150 disintegrations per minute [dpm]/ 100 cm2) alpha or beta/gamma contamination (corrected for background).

Clothing contaminated above egress limits will not be released unless laundered to within these limits. If skin or other parts of the body are contaminated above egress limits, reasonable steps that exclude abrasion or other damage will be undertaken to effect decontamination.

Areas that are designated as hjgh raruation and very high radiation areas will not be accessible to individuals during routine operation of the RPF. These areas wi ll be radiologically shielded and isolated from access to inruviduals by the use of engineered physical barriers that include structural shield blocks and/or locked shield doors.

11.1.6.3 Anti-Contamination Techniques The RPF is designed to limit contamination, with processes and equipment that contain radioactive material designed to require as little maintenance as possible to ensure personnel radiation exposures are ALARA. Additional exposure reductions will be achieved by:

Removing as much radioactive material as possible from equipment and area prior to maintenance, thereby reducing the intensity of the radiation field

• Providing adequate space for ease of maintenance to reduce the length of time required to complete the task, thereby reducing time of exposure Preparing and using procedures that include specifications for tools and equipment needed to complete assigned work activities Conducting proper job planning, including practice on mockups 11-31

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~ *.* ~ NOflTHWEST MEDtCAL ISOlW'ES Reviewing previous similar jobs Identifying highest contamination areas and communicating that information to workers prior to start of work.

11.1.6.4 Monitoring and Handling Contaminated Equipment and Components Outside Contaminated Areas The RPF processes and equipment that contain radioactive material are designed to require as little maintenance as possible to ensure that personnel radiation exposures are ALARA. Additional contamination controls are described in Section 11.1.6.3.

11.1.6.5 Criteria for Classification of Contaminated Material, Equipment, and Working Areas Contaminated material and equipment that are removed from a restricted area will be appropriately packaged in preapproved containers, inventoried, and monitored prior to release. The specific criteria for classifying contaminated materials and equipment will be provided in the Operating License Application.

Classification of the working areas is provided in Section 11.1.5.5.

11.1.6.6 Training Programs Details on the training program associated with the radiation protection program are discussed in Section 11.1.2.5.

11.1.6. 7 Recordkeeping Recordkeeping requirements associated with the radiation protection program are discussed in Section 11.1.2.9.

11.1.6.8 Technical Specifications As determined in Chapter 13.0, "Accident Analysis," the RPF is designed to control contamination consistent with occupational safety and protection of the public and environment. Contamination control will likely not require a technical specification.

11.1.7 Environmental Monitoring The RPF radiological environmental monitoring program will meet 10 CFR 20.1302. The radiological environmental monitoring program will be used to verify:

Effectiveness of plant measures are used to control release of radioactive material Measurable concentrations of radioactive materials and levels of radiation are not higher than expected based on effluent measurements and modeling of environmental exposure pathways.

Methods for establishing and conducting environmental monitoring are provided in Regulatory Guide 4.1, Radiological Environmental Monitoring/or Nuclear Power Plants. Regulatory Guide 4.1 refers to NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors, for detailed guidance on conducting effluent and environmental monitoring.

Although Regulatory Guide 4.1 and NUREG-1301 are written for nuclear power plants, due to the similarities between airborne releases of radioactivity from nuclear power plants and those released from the RPF, the guidance provided in Regulatory Guide 4.1 and NUREG-1301 was considered when developing radiological environmental monitoring for the RPF. Specifically, guidance provided in Figure 1 of Regulatory Guide 4.1 and Table 3.12-1 ofNUREG-1301 was considered when determining which exposure pathways to sample, sample locations, types of samples, and sample frequencies for the RPF.

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NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.1.7.1 Verification of Compliance Environmental monitoring data will be compared against permits and environmental reports, as appropriate, to ensure compliance. As described above, methods for establishing and conducting environmental monitoring are provided in Regulatory Guide 4.1 . Regulatory Guide 4.1 refers to NUREG-1301 for detailed guidance on conducting effluent and environmental monitoring.

11.1. 7.2 Identification of Potential Impacts Potential impacts of the RPF on the environment are addressed in Chapter 19 .0, "Environmental Review."

11.1. 7.3 Establishment of Baseline Environmental Quality Background radiation values will be obtained during the baseline environmental survey by monitoring TLDs at multiple locations (Section 11 .1. 7.4.2). This survey will be conducted prior to construction and prior to RPF operation.

11.1.7.4 Environmental Surveillance Program The following radiation exposure pathways will be considered for monitoring under the NWMI radiological environmental monitoring program:

Waterborne exposure pathway Direct radiation exposure pathway monitoring using TLDs Airborne exposure pathway monitored using continuous air samples Ingestion exposure pathway 11.1.7.4.1 Waterborne Exposure Pathway Monitoring The proposed RPF is designed to have zero liquid discharge from the radiologically controlled area, with no release of water from the facility to the adjacent environment that would affect surface water (e.g., Gans Creek). As such, surface water sampling will not be included in the radiological environmental monitoring plan. Similarly, aquatic life in the rivers is not expected to accumulate detectable levels of radioactivity, and sampling of fish or other aquatic creatures for the ingestion pathway will not be included in the radiological environmental monitoring plan.

The groundwater aquifer beneath the proposed RPF site is the Mississippian aquifer (also referred to as the Kimmswick-Potosi aquifer), which is discussed in detail in Chapter 19.0, Section 19.3.4.2. There are no defined liquid effluent release pathways, and the groundwater is not expected to be contaminated due to operation of the RPF. Therefore, groundwater sampling will not be included in the radiological environmental monitoring plan.

11.1. 7.4.2 Direct Exposure Pathway Monitoring TLDs will provide measurements of direct radiation from radioactive materials located at the RPF ,

radioactivity in airborne effluent, and deposition of airborne radioactivity onto the ground. NUREG-1301 recommends 40 TLD locations, consisting of an inner ring and outer ring of TLDs, with one TLD located in each ring at each of the 16 meteorological sectors (i.e., a total of 32 TLDs) and the remainder located at special interest areas. NUREG-1301 also recommends that at least one TLD be located a significant distance from the facility as a control station to measure background radiation dose.

At the RPF, seven TLDs will be located outside at entry points to the building where personnel may congregate or spend time outside of the RPF building. An additional TLD will be located on the outside wall near the target fabrication area to evaluate direct radiation from the hot cells and waste management area. The location of the on-site TLDs is shown in Figure 11-6.

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• NOllTHWEST ME.DICAl ISOTOf'(S Chapter 11.0 - Radiation Protection and Waste Management NWMl-2013-021 , Rev. 1 DISCOVERY RIDGE LOT 15 PROPERTY UNE PlllE WA T!lB. PUMP SK.ID 7.4ACllES WASTI! MANAGEMl!NT BUILDING

~ WASTE MANAOEMENTCANOPY AREA FOR MECHANICAl.CHJLLER SIDE SETBACK

• 15 Fl!t."T SPACB RESl!RVED POR PlllE WATER STORAGE TANll:. AND llECE!Vl!ll TANll:.

Legend

• On-Site TLDs

• Fence Line TLDs CAMs BBRM SIDE SET BACK

• U Fl!l!T PARK.INO LOT32 TOTAL PARKJNO SPAC6S .¢

• STEP V l>J;

~

~

STEP V AN GUARDHOUSI!

ADMIN BUUD!NG fOOTPIUNT N P.LCURVE L- 117.75' R-74.30' FRONT SETBACK

• 3HBIIT GUARD HOUSE AND \IElllCL.8 TRAP AREA SITE PLAN f

PARKING LOT 24 TOTAL 100" 0 200' PARKING SPACES P.L CURVE L-JS9.84' R* 1S42.83' Figure 11-6. Location of On-Site Environmental Thermoluminescent Dosimeters and Continuous Air Monitors TLDs will also be located at the site boundary (the perimeter of the lot) to evaluate the direct radiation dose. Sixteen TLDs will be placed on the Jot line, with a TLD placed at all four corners of Lot 15, and the remaining TLDs placed at approximately equal distances from each other. The sixteen TLDs will provide adequate coverage to ensure that direct doses to neighboring facilities on adjoining lots can be monitored and evaluated. The location of the perimeter TLDs is shown in Figure 11-6.

An additional TLD will serve as a control and will be located offsite at a significant distance from the RPF such that it represents a background dose. One TLD location will be provided with two TLDs so that data quality can be determined.

11.1.7.4.3 Airborne Exposure Pathway Monitoring Airborne effluent releases from the RPF will contribute to off-site doses. The airborne effluent exhaust from the vent stacks is expected to contain measurable quantities of noble gas radioactivity (e.g., Xe and Kr). Radioactive iodine, radioactive particulates, and tritium could also be present in the airborne effluent exhaust. However, most of the off-site exposure due to airborne effluent releases will be associated with noble gas and radioactive iodine releases. The tritium release rate would be a small fraction of the noble gas rates provided in Table 11-2 (several orders of magnitude less). The dose contribution from tritium would be a small fraction of the dose contributions, and the total public dose from all routine gaseous releases including tritium would remain well below 10 CFR 20 limits.

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• NOfUHWlST MEDICAL ISOTOPES Chapter 11.0 - Radiation Protection and Waste Management The final facility design strategy is to route the air stream from the evaporation tanks into the Zone I exhaust system. The Zone I exhaust stack will have continuous monitoring.

Environmental airborne sampling will be performed to identify and quantify particulates and radioactive iodine in airborne effluents. Regulatory Position C.3.b of Regulatory Guide 4.1 indicates that airborne sampling should always be included in the environmental monitoring programs for nuclear power plants.

Since the RPF includes airborne effluent releases and radioactivity in the airborne effluent can result in measurable off-site doses, the radiological environmental monitoring program will include airborne sampling.

The guidance provided in Table 3.12-1 ofNUREG-13 01 was used to establish locations for airborne sample acquisition, sampling frequency, and type of sample analysis. Continuous air sample locations will be specified in accordance with the guidance provided in Table 3.12-1 ofNUREG-1301. The CAMs that are used to obtain continuous air samples will include a radioiodine canister for weekly iodine-131 (1 31 1) analysis, and a particulate sampler that is analyzed for gross beta activity and for quarterly isotopic analysis.

Four CAMs (air samplers) will be located near the facility fence line, with one CAM being located in the direction of the prevailing wind (e.g. , north-northwest) and the other three CAMs being located in the remaining cardinal directions (e.g., 90 degrees) from the first CAM location (i.e., west-southwest, south-southeast, and east-northeast). The CAM locations are shown in Figure 11-6. An additional CAM will be located a sufficient distance from the RPF, in the least prevalent wind direction, to provide background information for airborne activity.

11.1.7.4.4 Ingestion Exposure Pathway Monitoring NUREG-130 I suggests sampling of various biological media (biota monitoring) to indirectly assess doses due to particulate and iodine ingestion. This type of monitoring may include sampling of soils and broad-leafed plants, fish, meat, or milk. Considering that particulates and iodine radionuclides are not expected to be present in measurable quantities within the RPF airborne effluent releases, biota monitoring will not be performed. In the event that environmental airborne sample results indicate the presence of iodine or particulates in measurable quantities, or if the effluent monitor sample results indicate the presence of iodine or particulates in quantities large enough to result in a calculated dose at the property line that exceeds 10 percent of the dose constraint (i.e., I mrem/yr), a sampling campaign will be undertaken.

Milk is an important food product that contributes to the radiation dose to people, most notably from radioactive iodine. If biota sampling is determined to be required as a result of radioactive iodine and particulate activity measured during effluent monitoring or air sampling, milk sampling will be performed following the guidance provided in Table 3.12-1 ofNUREG-1301 (e.g., sampling frequency and type of sample analysis). Cow and/or goat milk samples will be obtained from dairy production sites on a semi-monthly basis (when animals are on pasture) and on a monthly basis (at other times). A gamma isotopic analysis and 1311 analysis will be performed on the samples. Since milk samples are considered a better indicator ofradioactive iodine in the environment than vegetation, as long as milk samples are obtained, vegetation sampling (e.g. , broad leaf vegetation) is not expected to be included in the exposure pathway sampling, in accordance with guidance provided in Table 3.12-1 ofNUREG-1301.

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~* * ~ NOlmfWEST MEDICAL ISOTOPH NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management 11.2 RADIOACTIVE WASTE MANAGEMENT Radioactive waste management activities associated with the RPF liquid, gaseous, and solid waste systems will be performed in accordance with approved written procedures. A detailed description of the sources, types, and approximate quantities of waste within the proposed RPF is provided in Chapter 19.0, Section 19.2.7.

The Plant Manager will have responsibility for preparation and implementation of the radioactive waste management procedures.

11.2.1 Radioactive Waste Management Program The following subsections provide a description of the radioactive waste management program, including the philosophy and objectives. Organizational information is provided relative to the Waste Management organization. A detailed NWMI organization chart is provided in Chapter 12.0, Section 12.1.

The waste management program will be coordinated with the radiation protection program, and program management will report to the Plant Manager. Section 11.1 describes the program and procedures for controlling and assessing radioactive exposures associated with radioactive sources, including radioactive waste streams.

The goal of the waste management program is to minimize waste generation, minimize exposure of personnel, and to protect the public and environment. An official charter describing the authority, duties, and responsibilities of personnel in the Waste Management organization will be described in the FSAR as part of the Operating License Application.

11.2.1.1 Waste Management Policy NWMI management is committed to the ALARA philosophy for radioactive waste management.

NWMI's policy is to conduct waste management operations in a manner that ensures the health and safety of employees, contractors, and the public, and to comply with all Federal, State, and local Jaws and regulations for generation, storage, packaging, transportation, and disposal of wastes generated at the RPF.

11.2.1.2 Waste Management Procedures Radioactive waste management procedures will be developed and reviewed in accordance with the RPF procedure program, as discussed in Chapter 12.0, Section 12.3. These procedures will provide for the efficient and safe conduct of operations of the waste management program. Additional information on waste management procedures will be provided in the Operating License Application.

11.2.1.3 Organizational Responsibilities 11.2.1.3.1 Plant Manager The Plant Manager will report to the COO and have direct responsibility for the safe operation of the RPF, including the protection of workers and the public against radiation exposure resulting from facility operations and materials. Other responsibilities will include:

• Ensuring compliance with applicable NRC, State, and local regulations and license/permits

• Implementing the RPF conduct of operations program Establishing and managing the required training programs to support the operations organization 11-36

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' ~* * ~ NORTHWEST MlDICAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management 11.2.1.3.2 Waste Management Lead The Waste Management Lead will report to the Plant Manager and have responsibility for the following:

Implementing waste management policy Developing waste management procedures for the processing, packaging, and shipment of radioactive waste from the facility Processing, packaging, and shipping radioactive waste from the facility Providing technical input to the design of equipment and processes Providing technical input to the waste management training program

• Establishing and maintaining contractual relationships with waste disposal sites and radioactive waste carriers

• Maintaining working knowledge of the waste acceptance criteria, standards, guides, and codes with respect to waste disposal Conducting self-assessments of waste management practices and ensuring compliance with procedures in accordance with the waste management self-assessment program 11.2.1.4 Training The radioactive waste management training program will be closely coordinated with the radiation protection training program to emphasize the importance placed on radiological safety of RPF personnel and the public. In-depth waste management training will be provided for the various types of job functions (e.g. , operator, waste technician, and waste shipper) commensurate with responsibilities associated with each position and conducted in accordance with the procedures defined in Chapter 12.0.

Additional information on training will be provided in the Operating License Application.

11.2.1.5 Document Control and Recordkeeping Document control and recordkeeping will be in accordance with the procedures defined in the QAPP (Chapter 12.0, Appendix C) and will include waste manifests, audits, and program reviews. Additional information on document control and recordkeeping will be provided in the Operating License Application.

11.2.1.6 Reviews and Audits Audits of the RPF radioactive waste management program will be conducted, at a minimum, on an annual basis to review the functional and safety elements of the waste management program. An additional component of these audits will be to evaluate efforts to minimize the production of radioactive wastes.

Additional detail on the NWMI review and audit function is provided in Chapter 12.0.

11.2.1.7 Technical Specifications As discussed in Chapter 13 .0, the RPF waste management processes are designed with a focus on occupational safety and protection of the public and the environment. The technical specifications associated with waste management area, if applicable, will be discussed in Chapter 14.0 as part of the Operating License Application.

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• HOflllfWHT MEDICAL ISOTonS Chapter 11 .0 - Radiation Protection and Waste Management 11.2.2 Radioactive Waste Management Controls The NRC divides low-level radioactive waste into three different classes: A, B, and C. These classes are based on waste concentration and the half-lives and types of radionuclides in the waste.

Class A - Waste consisting of radionuclides with the shortest half-lives and lowest concentrations, with radioactivity levels that return to background levels within 100 years

• Classes Band C - Waste containing greater concentrations of radionuclides with longer half-lives, fading to background levels in less than 500 years (must meet stricter disposal requirements than Class A waste)

Greater than Class C - Waste exceeds the requirements for Class C waste and is the responsibility of the U.S. Department of Energy under Federal law Descriptions of the RPF processes that will produce radioactive waste are provided in Chapter 4.0.

NWMI will implement pollution prevention and waste minimization activities that review associated processes and procedures to ensure that the kinds and amounts of waste generated are minimized.

Waste management control will include methods to:

Avoid inadvertent exposure of personnel or uncontrolled escape of the radioactive materials Maintain continuous control of radioactive materials that require treatment and management as waste 11.2.2.1 Waste Designation The RPF will generate class A, B, and C low-level radioactive waste. Greater than Class C waste will not be produced in the RPF. The Class A waste is low-dose waste, while Class Band C wastes are high-dose waste. The solidified high-dose liquid waste from the RPF will be either Class B or Class C waste. As a result of reducing the waste volume and minimizing disposal costs, the liquid waste concentration endpoint may result in a change in the final waste classification from Class B to Class C.

The waste handling system for the generated wastes is discussed in Chapter 9.0, Section 9.7.4. High-dose liquids will be designated as waste once the liquids are collected in the waste concentrate collection tank.

The low-dose condensate from the high-dose concentrator held in the condensate collection tank will be used as much as practical by the uranium recovery process. Low-dose liquids will be designated as waste once the liquids are collected in the low-dose evaporation tank. Solids will be designated as waste once the solid materials are loaded into waste drums.

11.2.2.2 Waste Management Procedures Radioactive waste management operating procedures are discussed in Section 11.2.1.2. These procedures ensure proper identification, characterization, and separate treatment of radioactive wastes. Additional information on waste management procedures will be provided in the Operating License Application.

11.2.2.3 Airborne Radioactive Waste Management The RPF will not directly produce airborne radioactive waste. Chapter 9.0, Section 9.1, provides a detailed description of the process vessel vent system and the Zone I and Zone II HV AC treatment systems. Liquid waste resulting from these processes will be directed to the high-dose waste collection tank and processed through the high-dose waste treatment system, where the waste will be solidified.

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~ * *! . NORTHWEST M£01CA1. lSOTOPfS NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.2.3 Release of Radioactive Waste This section discusses the methods used to identify, characterize, package, and transport waste offsite from the RPF. The majority of the radioactive waste being shipped from the RPF will require special containers to provide for the protection of the public and environment. Each of these containers is designed to meet applicable NRC and U.S. Department of Transportation (DOT) standards. In general, waste released from the RPF will be processed and packaged to meet the waste acceptance criteria of an established disposal facility. The processing and packaging of routine waste is described in Chapter 9.0, Section 9.7.4. Table 11-6 summarizes the types waste generated and annual generation rate for the RPF.

Table 11-6. Waste Produced in the Radioisotope Production Facility Description 3 UMhdll!!!ml Annual generation High-doseb,c,d Solid B or C 200,000 L (52,834 gal)

Low-doseb,c Solid A 150,000 L (39,625 gal)

Target cladding materials from disassembly encapsulated in cement Solid c 1,100 L (290 gal)

Exchange resins and other solid waste Solid c 1,370 L (365 gal)

Solid wastes encapsulated in cement Solid A 8,000 L (2,113 gal)

HEPA filters Solid AorC 28 m 3 (977 ft3)

Carbon Solid AorC 0.14 m 3 (5 .1 ft 3 )° Iodine absorption Solid c 0.06 m 3 (2.1 ft3l Facility support waste (non-rad) Solid NIA 26,000 L (6,868 gal)

Facility support waste (rad) Solid A 40,000 L (10,566 gal)

Silicone oil Liquid A 100 L (26 gal)

Lab pack Liquid A 10 L (2.6 gal)

Solvent Liquid A 200 L (53 gal)

• Special nuclear material is not considered a waste. SNM will be returned to the U recovery and recycle system, purified, and reused. In addition, waste volume projections are based on the composite values from the MURR and OSTR mass balance calculations that assume an eight-target/week MURR processing rate plus a 30-target/week OSTR processing rate and will bound the planned operations.

b Caustic soda (NaOH) is included in the waste volume estimates.

c Waste solidification agents are included in the waste volume estimates.

d Nongaseous long-lived radioisotopes are contained in the high-dose liquid waste stream that is solidified and eventually sent offsite for disposal.

e Volume represents changeout of carbon beds every two years (1 12).

r Volume represent changeout of iodine absorption beds every five years (115).

HEPA high-efficiency particulate air. OSTR Oregon State University TRlGA Reactor MURR University of Missouri Research Reactor. SNM special nuclear material.

NIA not applicable. U uranium.

NaOH sodium hydroxide 11.2.3.1 Solid Radioactive Waste The majority of solid waste produced in the RPF will be the high- and low-dose waste discussed in Chapter 9.0. Samples of this waste will be analyzed in the RPF laboratory to ensure that the waste meets the disposal facility waste acceptance criteria. This waste will be stored for radioactive decay to meet shipping and disposal requirements, and then packaged in approved transportation casks for transport to the disposal facility.

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' ~* *!: NOflTHWESTMEDICALISOTOPES The RPF will also produce intermittent waste that includes HEP A filters, carbon absorption beds, and zeolite absorption beds. Samples of these wastes will be analyzed within the RPF laboratory to ensure that the waste meets the disposal facility waste acceptance criteria. These wastes will be encapsulated and packaged in appropriate waste containers for disposal. Depending on the location and use, HEP A filters are anticipated to have a changeout frequency from monthly to every 2 years. The iodine adsorber beds are designed to last 5 years before requiring changeout. Carbon beds have a 2-year design life.

Table 11-7, Table 11-8, and Table 11-9 list the low-dose radioactive waste sources, high-dose radioactive waste sources, and encapsulated solid radioactive waste sources, respectively.

11.2.3.2 Liquid Radioactive Waste The RPF does not release any radioactive liquid waste. As discussed in Chapter 9.0, Section 9.7, high-and low-dose liquid waste will be solidified prior to release.

11.2.3.3 Gaseous Radioactive Waste Gases from the RPF process and HVAC system will be processed as described in Chapters 4.0 and 9.0, respectively. The offgas system is designed to filter and/or retain these isotopes in the facility until the resulting release is at levels less than those defined in Table 2 of 10 CFR 20, Appendix B. The gaseous radioactive emissions will be released through the RPF ' s three exhaust stacks. Monitoring of the effluent is described in Section 11.1.4.1.2.

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• NORTHWEST MEDICAL ISOTOPES Table 11-7. Low-Dose Radioactive Waste Sources Radioactivity Radioactivity Isotope (Ci/wk) Isotope (Ci/wk) 241Am 2.26E-15 239Np 8.91E-03 136mBa l .77E-06 233p3 9osr 7.44E-05 137mBa 7.1 lE-05 234pa 4.lOE-11 91sr 2.00E-17 140Ba l. 75E-02 234mp 3 3. l SE-08 99Tc l.06E-08 141ce l.44E-02 11 2pd 2.45E- ll 99mTc 7.06E-04 143Ce 4.29E-06 147Pm 2.42E-04 125mTe 2.29E-07 144Ce 2.42E-03 148pm 8.l 7E-08 121Te 9.62E-05 134Cs l .21E-08 148mpm l.1 lE-07 127mTe 2.93E-05 136 Cs l.58E-05 149Pm 3.96E-05 129Te 1.53E-04 137 Cs 7.52E-05 151Pm 5.56E-08 129mTe 2.39E-04 1ssEu 2.54E-06 142pr l .41E-15 131Te 2.3 lE-08 1s6Eu 4.67E-05 143Pr l.89E-02 131mTe l.03E-07 1s1Eu l.95E-14 144Pr 2.42E-03 132Te l .09E-03 1291 7.96E-1 2 144mpr 3.39E-05 232u 7.85E-12 1301 2.36E- 18 238pu 3.37E-l 1 234u 7.40E-06 13 11 4.l 7E-03 239pu 9.06E-08 23su 3.38E-07 1132 7.61E-04 24opu 6.41E-l l 236u 1.14E-07 1331 6.88E-09 241Pu 2.31E- 11 231u 5.34E-03 85Kr 2.94E-10 103mRh 6.77E-03 238u 2. 13E-07 140La 2.0 lE-02 105Rh 1.63E-06 13 1mxe 3.87E-09 99Mo 3.12E-05 106Rh l.44E-04 133 Xe 2.58E-07 95Nb 4.47E-03 103Ru 6.78E-03 133mxe 3.20E-10 95mNb 1.21E-04 106Ru l.44E-04 89my 8.29E-07 96Nb 5.04E-12 122 sb 7.32E-l l 90y 7.43E-05 97Nb 2.02E-10 124Sb l.63E-08 90y l .02E-02 97~ 1.78E-10 12ssb 4.28E-06 91my l.27E-l 7 141Nd 5.95E-03 126Sb 2.46E-06 93 y l .80E-16 236mNp l.89E-1 8 121sb 7.36E-05 93Zr l.SlE-09 231Np 7.04E- l l 1s1sm l.71E-06 9szr 1.04E-02 23sNp 7.23E-1 0 1s3sm 2.26E-06 97Zr l.87E-1 0 Total Ci l.44E-01 Source: WO 17, solidified low-dose concentrate, in [Proprietary Information].

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~* * ~ NomrwtST MEDICAi. tsOTOP£S Table 11-8. High-Dose Radioactive Waste Sources Radioactivity Isotope (Ci/wk) 241 Am 9.93E+02 136mBa l.97E-01 234pa 4.57E-06 9osr 8.30E+OO 137mBa 7.92E+OO 234mpa 3.Sl E-03 9' Sr 2.22E-1 2 139Ba l .95E+03 11 2pd 2.73E-06 99Tc l .18E-03 140Ba l.60E+03 147Prn 2.70E+Ol 99mTc 7.87E+Ol 141ce 4.78E-01 148Pm 9.lOE-03 125mTe 2.SSE-02 143Ce 2.70E+02 148mpm l.24E-02 121Te l.07E+Ol 134Cs 1.35E-03 149Pm 4.41E+OO 127mTe 3.27E+OO 136 Cs 1.76E+OO 151Prn 6.19E-03 129Te l. 71E+Ol 137 Cs 8.39E+OO 142Pr l.57E-10 129mTe 2.66E+Ol 1ssEu 2.83E-01 143Pr 2.1 1E+03 131Te 2.58E-03 1s6Eu 5.20E+OO 144Pr 2.70E+02 131mTe l.14E-02 1s1Eu 2. 18E-09 144mpr 3.78E+OO 132Te l .22E+02 1291 8.87E-07 238Pu 3.71E-06 232u 2.88E-l 1 1301 2.63E- l3 m pu 9.97E-03 234u 2.71E-05 1311 4.65E+02 240pu 7.06E-06 235u 1.24E-06 11 32 8.48E+Ol 241Pu 2.54E-06 236u 4.l 7E-07 1331 7.67E-04 103mRh 7.54E+02 231u l.96E-02 85Kr 3.27E-05 105Rh l.82E-Ol 23su 7.80E-07 140La 2.24E+03 106Rh l.61E+Ol 1J 1mxe 4.31E-04 99Mo 3.47E+OO 103Ru 7.56E+02 133 Xe 2.87E-02 95Nb 4.98E+02 106Ru l.61E+O l 1JJmxe 3.57E-05 95mNb l .35E+Ol 122 sb 8.16E-06 13sxe 5.24E-17 96Nb 5.61E-07 124Sb l.82E-03 89my 9.23E-02 97Nb 2.25E-05 125 Sb 4.77E-Ol 90y 8.28E+OO 97mNb l .98E-05 126Sb 2.74E-Ol 90y 1.13E+03 141Nd 6.63E+02 127 Sb 8.20E+OO 91my l.41 E- 12 236~p 3.5 lE-17 128 Sb l.lSE-15 93y 2.0 lE-11 231Np l .30E-09 1s1sm l.90E-01 93zr 1.68E-04 23sNp l.34E-08 153 Sm 2.52E-01 95Zr 1.16E+03 239Np l .28E-02 1s6sm 3.68E-15 97 Zr 2.09E-05 Total Ci 1.42E+04 Source: W015, solidified high-dose concentrate, in [Proprietary Information].

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NWMl-2013-021 , Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Table 11-9. Encapsulated Solid Radioactive Waste So urces

~M Radioactivity Isotope Isotope (Ci/wk) 241 Am m p, l.60E+Ol 136mBa 3.18E-03 234pa 5.67E-05 9o sr l.34E-01 137mBa 1.28E-Ol 234mpa 4.40E-08 9 1sr 3.59E-l4 139Ba 3.14E+Ol 112pd 4.35E-01 99Tc l.90E-05 140Ba 2.58E+Ol 147Pm l .47E-04 99mTc l .27E+OO 141ce 7.72E-03 148Pm 2.00E-04 125 mTe 4.12E-04 143C e 4.36E+OO 148mpm 7.12E-02 121Te l.73E-Ol 134Cs 2.18E-05 149Pm l.OOE-04 127mTe 5.28E-02 136Cs 2.84E-02 151 Pm 2.53E- 12 129Te 2.76E-Ol 137 142Pr 129mTe Cs l.35E-Ol 3.40E+Ol 4.30E-Ol 1ssEu 4.57E-03 143Pr 4.36E+OO 131Te 4.16E-05 1s6Eu 8.40E-02 144Pr 6. lOE-02 13 1mTe l.85E-04 1s1Eu 3.5 IE-11 144mpr 5.99E-08 132Te l.97E+OO 1291 l.47E-08 238Pu l.61E-04 2nu 8.92E-10 1301 4.35E-15 239pu l.14E-07 234U 8.40E-04 1311 7.69E+OO 240Pu 4.l lE-08 mu 3.84E-05 1132 1.40E+OO 24 1Pu l. 22E+O l 236u l.29E-05 1331 l.27E-05 103mRh 2.94E-03 231u 6.07E-Ol 85Kr l .13E-02 105 Rh 2.59E-Ol 23su 2.42E-05 140La 3.62E+Ol 106Rh l.22E+Ol 13 1mxe l.49E-Ol 99Mo l.31E+OO 103 Ru 2.59E-Ol 133 Xe 9.90E+OO 95Nb 8.04E+OO 106Ru 1.32E-07 133 mxe l.23E-02 95 mNb 2. l 7E-Ol 122 sb 2.94E-05 13sxe l.8 IE-14 96Nb 9.06E-09 124Sb 7.70E-03 89my 1.49E-03 125 90y 97Nb 3.63E-07 Sb 4.43E-03 l .34E-Ol 97mNb 3.20E-07 126Sb l.32E-01 90y l.83E+Ol 141Nd l .07E+Ol 127 Sb l .85E-l 7 9 1my 2.28E-1 4 236mNp l.53E-15 12s sb 3.07E-03 93 y 3.24E-13 231N p 5.68E-08 1s1sm 4.07E-03 93z r 2.72E-06 23sNp 5.83E-07 1s3sm 5.94E-17 95Zr l.87E+Ol 23 9Np 5.58E-Ol 1s6 sm 97 3.37E-07 7.37E-08 Zr Total Ci 2.42E+02 Source : W022 encapsulated waste in [Proprietary Information].

Note: This table does not include carbon beds, iodine absorption beds, or HEPA filters .

HEPA = high-efficiency particulate air.

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NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management 11.3 RESPIRATORY PROTECTION PROGRAM 10 CFR 20, Subpart H, defines the required elements of the RPF respiratory protection and ventilation programs. The use of engineering controls is preferred over the use of respirators to minimize radioactive materials in the air. However, there may be a need for the following to control the concentrations of radioactive material in the air to maintain the TEDE ALARA.

Control of access Limitation of exposure times Use of respiratory protection equipment Other controls The RPF facility design and analysis of the RPF ventilation system ensures that no uncontrolled releas e of airborne radioactive material to the unrestricted environment could occur during normal operational states and to mitigate the consequences of design basis accidents (e.g., maintaining a series of cascading pressure zones to draw air from the cleanest area to the most contaminated area of the RFP). In addition, the preliminary design indicates that the distribution and concentrations of any airborne radionuclides are limited by operation of the ventilation system so that during the full range of facility operations, no potential occupation exposures would exceed the design bases (e.g., 10 CFR 20), as described in this chapter. The pressure relationship between the four ventilation zones and ambient atmospheric pressure is presented below.

Zone IV will be the cleanest zone and is slightly positively pressurized with respect to atmosphere.

Zone IV is independent of the other three ventilation zones. Zones I, II, and III will potentially be contaminated areas, with Zone III being the cleanest of the potentially contaminated areas, and each subsequent zone being more contaminated and having lower pressures, as shown below:

P Zone I < P Zone II < P Zone Ill The irradiated target receipt area and the irradiated target truck bay are two different areas in the RPF.

The truck bay is where trailers will be rinsed before entering the receipt area, and where the cask will be removed from the trailer. The irradiated target truck bay is Zone IV, while the irradiated target receipt area is normally Zone III. Details of how the irradiated target receipt area will transition between Zone II and III during operating/maintenance activities will be provided in the FSAR as part of the Operating License Application.

Section 3 .1 provides the codes and standards to which the ventilation system will be designed. The detailed ventilation system criteria, including minimum flow velocity at openings in each zone, maximum differential pressure across filters, and types of filters to be used (e.g. HEPA, high-efficiency gas adsorption [HEGA]), will be provided in the FSAR as part of the Operating License Application.

NWMI's radiological respiratory protection program is designed to comply with the requirements of ANSI Z-88.2, American National Standard for Respiratory Protection; 10 CFR 20, Subpart H; and 29 CFR 1910.134, "Respiratory Protection." Respirators will only be issued if the Radiation Protection Manager determines that engineering controls may be ineffective, the total effective dose will be reduced by wearing respirators, and/or the physical stress of wearing a respirator will not interfere with workers' health and safety.

If the decision is made to permit the use of respiratory protection equipment to limit the intake of radioactive and hazardous materials, only National Institute of Occupational Safety and Health (NIOSH)-

certified equipment will be used.

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• NORTHWEST MEDtCAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management The respiratory protection program will include the following elements:

Air sampling to identify the potential hazard, select proper equipment, and estimate doses

• Surveys and when necessary, bioassays, to evaluate actual intakes Performance testing of respirators for operability (user seal-check for face-sealing devices and functional check for others) immediately prior to each use Limitations on periods ofrespirator use and relief from respirator use Determination by a physician that the individual user is medically fit to use respiratory protection equipment. This evaluation will be done prior to initial fitting of a face sealing respirator, before the first field use of non-face sealing respirators, and either every 12 months thereafter, or periodically at a frequency determined by a physician.

A respirator fit test will require a minimum fit factor of at least I 0 times the assigned protection factor for negative pressure devices, and an overall fit factor of at least 500 for any positive pressure, continuous flow, and pressure-demand devices. The fit testing will be performed before the first field use of tight-fitting, face-sealing respirators. Subsequent testing will be performed at least annually thereafter. Fit testing must be performed with the face-piece operating in the negative pressure mode.

Personnel using respirators will be informed that they may leave the area at any time for relief from respirator use in the event of equipment malfunction, physical or psychological distress, procedural or communication failure, significant deterioration of operating conditions, or any other conditions that might require such relief. Respirator used within the RPF will provide for vision correction and adequate communication and allow for concurrent use of other safety or radiological protection equipment.

Radiological protection equipment will be used in such a way as to not interfere with the proper operation of the respirator.

Standby rescue personnel will be used whenever one-piece, atmosphere-supplying suits are in use.

Standby rescue personnel will also be available when any combination of supplied air respiratory protection device and personnel protective equipment is in use that presents difficulty for the wearer to remove the equipment. The standby personnel will be equipped with respiratory protection devices or other apparatus appropriate for the potential hazards. The standby rescue personnel will observe and maintain continuous communication with the workers (e.g., visual, voice, signal line, telephone, radio, or other suitable means). The rescue personnel will be immediately available to assist the workers in case of a failure of the air supply or for any other emergency. The Radiation Protection Manager, in consultation with the SH&L Manager, will specify the number of standby rescue personnel that must be immediately available to assist all users of this type of equipment and to provide effective emergency rescue if needed.

Atmosphere-supplying respirators will be supplied with respirable air of a quality that meets or exceeds the specifications of Compressed Gas Association (CGA) G-7, "Compressed Air for Human Respiration,"

and G-7.1, "Commodity Specification for Air," and the requirements included in the regulations of the Occupational Safety and Health Administration, 29 CFR 191O.l34(i)(l )(ii)( A) through (E).

No objects, materials, or substances (e.g., facial hair), or any conditions that interfere with the face-to-face-piece seal or valve function, and that are under the control of the respirator wearer, will be allowed between the skin of the wearer 's face and the sealing surface of a tight-fitting respirator face-piece. The dose to individuals from the intake of airborne radioactive material will be estimated by dividing the ambient air concentration outside the respirator by the assigned protection factor. If the actual dose is later found to be greater than that estimated initially, the corrected value will be used. If the dose is later found to be less than the estimated dose, the lower corrected value may be used.

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NOlmtW£ST MEDICAL ISOTOP£S NWMl-2013-021, Rev . 1 Chapter 11 .0 - Radiation Protection and Waste Management Records of the respiratory protection program (including training for respirator use and maintenance) will be maintained in accordance with the NWMI records management program, as described in Section 11.1.6.7.

The radiological respiratory protection program will include written procedures for each of the following :

Monitoring, including air sampling and bioassays Supervision and training of respirator users Fit testing Respirator selection Breathing air quality Inventory and control Storage, issuance, maintenance, repair, testing, and quality assurance ofrespiratory protection equipment Recordkeeping Respiratory protection procedures will be revised as necessary whenever changes are made to the facility, processing, or equipment.

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11.4 REFERENCES

10 CFR 19, "Notices, Instructions, and Reports to Workers: Inspection and Investigations," Office of the Federal Register, as amended.

10 CFR 19.12, "Instruction to Workers," Office of the Federal Register, as amended.

10 CFR 19.13, "Notifications and Reports to Individuals," Office of the Federal Register, as amended.

10 CFR 20, "Standards for Protection Against Radiation," Code ofFederal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1003, "Definitions," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1101, "Radiation Protection Programs," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1201, "Occupational Dose Limits," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1208, "Dose Equivalent to an Embryo/Fetus," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1301, "Dose Limits for Individual Members of the Public," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.1302, "Compliance with Dose Limits for Individual Members of the Public," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.2110, "Form of Record," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.2202, "Notification oflncidents," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 20.2203, "Reports of Exposures, Radiation Levels, and Concentrations of Radioactive Material Exceeding the Constraints or Limits," Code ofFederal Regulations, Office of the Federal Register, as amended.

10 CFR 20.2206, "Reports oflndividual Monitoring," Code ofFederal Regulations, Office of the Federal Register, as amended.

10 CFR 50.34, "Contents of Applications; Technical Information," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 50.71, "Maintenance of Records, Making of Reports," Code of Federal Regulations, Office of the Federal Register, as amended.

10 CFR 70.51, "Records Requirements," Code ofFederal Regulations, Office of the Federal Register, as amended.

29 CFR 1910, "Occupational Safety and Health Standards," Code of Federal Regulations, Office of the Federal Register, as amended.

ANSI 15.8, Quality Assurance Program Requirements for Research Reactors, American National Standards Institute, New York, New York, 1995, R2005/2013.

ANSI N 13 .1, Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities, American National Standards Institute, New York, New York, 1999.

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' ~e *~ NOITHWEST llEDtCAL ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 11.0 - Radiation Protection and Waste Management ANSI Nl3.6, Practice for Occupational Radiation Exposure Records Systems, American National Standards Institute, New York, New York, 1966 (R 1989).

ANSI N13 .1 l , Dosimetry-Personnel Dosimetry Performance-Criteria for Testing, American National Standards Institute, New York, New York, 2001.

ANSI N13.27, Performance Requirements for Pocket-Sized Alarm Dosimeters and Alarm Ratemeters, American National Standards Institute, New York, New York, 1981.

ANSI N323, Radiation Protection Instrumentation Test and Calibration, American National Standards Institute, New York, New York, 1978.

ANSI Z-88.2, American National Standard for Respiratory Protection, Rev. 15, American National Standards Institute, New York, New York, 2015.

ANSI/ ANS-15.11 , Radiation Protection at Research Reactors, American Nuclear Society, La Grange Park, Illinois, 2009.

ANSl/HPS NJ 3.22, Bioassay Program for Uranium, Rev. 13, American National Standards Institute/Health Physics Society, New York, New York, 1995 (R2013).

ANSl/HPS NJ 3.30, Performance Criteria for Radiobioassay, Rev. 1 American National Standards Institute/Health Physics Society, New York, New York, 2011.

ASTM El 168-95, Standard Guide for Radiological Protection Training for Nuclear Facility Workers ,

ASTM International, West Conshohocken, Pennsylvania, 2013 .

Barrington, C., 2015 , "NWMI Release # 11 - Process Vessel Ventilation (PVV) System Estimate,"

(memorandum to G. Dunford, May 26), AEM Consulting, LLC, Richland, Washington, 2015 .

CGA G-7, "Compressed Air for Human Respiration," Compressed Gas Association, Chantilly Virginia, April 2014.

CGA G-7.1 , "Commodity Specification for Air," Compressed Gas Association, Chantilly Virginia, October 2011 .

EPA 520/1-89-002, A Guide for Determining Compliance with the Clean Air Act Standards for Radionuclide Emissions from NRC-Licensed and Non-DOE Federal Facilities, Rev. 2, U.S. Environmental Protection Agency, Washington, DC, October 1989.

EPA 520/1-89-003 , Users Guide for the COMPLY Code, Rev. 2, U.S. Environmental Protection Agency, Washington, DC, October 1989.

NUREG-1301 , Offsite Dose Calculation Manual Guidance: Standard Radiological Effeuent Controls for Pressurized Water Reactors, Rev. 1, U.S. Nuclear Regulatory Commission, Office ofNuclear Reactor Regulation, Washington, D.C. , April 1991.

NUREG-1400, Air Sampling in the Workplace, U.S . Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington, D.C., September 1993 .

[Proprietary Information]

NWMI-2015-SHIELD-001, Radioisotope Production Facility Shielding Analysis, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 201 5.

Regulatory Guide 4.1, Radiological Environmental Monitoring /or Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.,

June 2009 (R2014).

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~ * *! NORTHWEST MEDtCAL ISOTOl'ES NWMl-2013-021, Rev. 1 Chapter 11 .0 - Radiation Protection and Waste Management Regulatory Guide 4.20, Constraint on Releases ofAirborne Radioactive Materials to the Environment for Licensees Other than Power Reactors, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C., April 2012.

Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., June 2008 (R2013).

Regulatory Guide 8.2, Administrative Practices in Radiation Surveys and Monitoring, Rev. 1, U.S . Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.,

May 2011.

Regulatory Guide 8.4, Personnel Monitoring Device- Direct-Reading Pocket Dosimeters, Rev. 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.,

June 2011.

Regulatory Guide 8. 7, Instructions for Recording and Reporting Occupational Radiation Exposure Data, Rev. 2, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., November 2005.

Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program, Rev. 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., July 1993.

Regulatory Guide 8. I 0, Operating Philosophy for Maintaining Occupational Radiation as Low as Is Reasonably Achievable, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., 1977 (R2014).

Regulatory Guide 8.13, Instructions Concerning Prenatal Radiation Exposure, U.S. Nuclear Regulatory Commission, Office ofNuclear Reactor Regulation, Washington, D.C., 1999 (R201 l).

Regulatory Guide 8.25, Air Sampling in the Workplace, Rev. 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., June 1992.

Regulatory Guide 8.29, Instructions Concerning Risks from Occupational Radiation Exposure, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.,

July 1981.

Regulatory Guide 8.34, Monitoring Criteria and Methods to Calculate Occupational Radiation Doses, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.,

July 1992.

Regulatory Guide 8.37, ALARA Levels for Effluents from Materials Facilities, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C ., July 1993.

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