ML17193A432

NWMI-2013-021, Revision 1, Chapter 8.0 - Electrical Power Systems, Construction Permit Application for Radioisotope Production Facility

ML17193A432
Person / Time
Site:	Northwest Medical Isotopes
Issue date:	06/30/2017
From:	Northwest Medical Isotopes
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17193A418	List: ML17193A425 ML17193A428 ML17193A432
References
NWMI-2017-007 NWMI-2013-021, Rev. 1
Download: ML17193A432 (235)

Text

• * * * * * * * * ****** * * ** * * * ** * ** * * * ** * ** * * ** * * . *. *. * . NORTHWEST MEDICAL ISOTOPES *

• Chapter 8.0 -Electrical Power Systems Prepared by: Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 1 June 2017 Northwest Medical Isotopes , LLC 815 NW g th Ave , Suite 256 Corvallis , OR 97330 This page intentionally left blank.

...... .. NWMI ....... *.:* ........... . * "NOmlW(STllEDICALISOTOPES NWMl-2013-021 , Rev. 1 Chapter 8.0 -Electrical Power Systems Chapter 8.0 -Electrical Power Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 8.0 -Electrical Power Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

C wJ r

..... .. NWMI ...*.. ..* .... ........ *.* 0

• 0

• lllOmfWEST MEOUL ISOTOPH NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems This page intentionally left blank.

Rev Date 0 6/29/2015 1 6/26/2017 REVISION HISTORY Reason for Revision Initial Application NWMl-2013-021 , Rev. 1 Chapter 8.0 -Electrical Power Systems Revised By Not required Incorporate changes based on responses to C Haass NRC Requests for Additional Information

.**.*.*. .; ... .. NWMI ........ *.* 0 !*. * . NOITHWEn MlDICAl lSOTOKS NWMl-2013-021 , Rev. 1 Chapter 8.0 -Electrical Power Systems This page intentionally left blank.

.; ... ; .. NWMI ..**.. ..* **.* ........ *.* .. *****. NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems *. * * . NOllTHWEU MEDICAL ISOTOPES CONTENTS 8.0 ELECTRICAL POWER SYSTEMS ................................................................

............................. 8-1 8.1 Norma l 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 Electrical Power Required .....................................

...............

............... 8-5 8.1.4 Use of Substation s 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 Radiation Monitor s .................................

8-8 8.2.5 Power for Physical Security Contro l , Information , and Communica tion Systems ......................

.....................

..................................................

...................

8-8 8.2.6 Power to Maintain Experimental Equipment in Safe Condition

.................

......... 8-8 8.2.7 Power for Active Confine men t/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 8.2.10 Power for Engineered Safety Feature Equipment

................................

.............

... 8-9 8.2.11 Power for Emergency Lighting ....................

....................

..................

..................

8-9 8.2.12 Power for Instrumentation and Contro l Systems to Monitor Shutdown .............. 8-9 8.2.13 Technical Specifications

......................................................................................

8-9 8.3 References

......................................................................................................

............

...... 8-10 8-i

..... ;. NWMI ...*.. ..* .... ........... ' *.* NOITlfWUT llllEDK:AL ISOTOPU NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems FIGURES F i gure 8-1. Radioisotope Production Facility Electrical One Line Diagram ..................................... 8-3 TABLES Table 8-1. Summary of Radioisotope Production Fa c ility and Ancillary Faciliti es Electrical Loads (2 p ages) ............................................................................................................

.... 8-1 8-ii

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

• 0 NOITifWEST MEDICAL tsOTOKS A cron y m s and Abbreviations TERMS AEC active e n gineering control A TS automatic transfer sw itch CAAS crit icality accide nt alarm system NWMl-2013-021 , Rev. 1 Chapter 8.0 -Electrical Power Systems HV AC heating , ventilation , and air conditioning IEEE Institute of Electrical and E l ectronics Engineers IROFS item relied on for safety MCC motor control center NEP normal electrica l power NFPA National Fire Protection Association NO x NWMI RPF SEP UPS U nits gal hp hr Hz km kV kW L rru rrun s ec v nitrogen oxides Northwest Medical Isotopes, LLC Radioisotope Production Faci li ty sta ndb y electrical power uninterruptable power supply ga llon horsepower hour hertz kilometer kilovolt kilowatt liter mile minute second volt 8-iii

.. .. NWMI ..**.. ..* .... ..... .. .. .. 0 * ! 0 HORTHW£ST MUHCAl ISOTOP'ES NWMl-2013-021 , Rev. 1 Chapter 8.0 -Electrical Power Systems T his page inte n ti o na ll y l eft bl ank. 8-iv

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

• NORTlfWEST MEDICAL ISOTOtaES NWMl-2013-021, Rev. 1 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 if NEP is interrupted or lost. Uninterruptable power s upplies (UPS) automatically provide power to systems that support the safety functions protecting workers and the public. The NEP system is de signe d to provide reasonable assurance that use or malfunction of electrical power sys tems will not damage the RPF or prevent safe RPF s hutdown. In addition , the RPF has a non-safety standby electrical power (SEP) system to reduce or eliminate proces s downtime due to electrical outages. A combination of UPSs 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 load s, which systems have UPSs , and the load s for those systems supported by the SEP sys tem. Table 8-1. Summary of Radioisotope Production Facility and Ancillary Facilities Electrical Loads (2 pages) Demand Target fabric a tion system Target receipt and disassembly system Target dissolution system Molybdenum recovery and purification system Ura nium reco very and re cycle sys tem Waste handling system Radiation monitoring and CAAS systems Standby electrical power system General facility electrical power Process vessel ventilation system Facility ventilation system Ventilation Zone I Ventilation Zone II/III Ventilation Zone IV Laboratory ve ntilation Supply air Fire protection syste m Plant and instrument air system Gas supply syste m Process chilled water system Normal electrical peak power load --125 16 8 30 40 40 54 30 40 IO 13 25 34 5 7 N I A 173 232 40 54 67 90 215 288 295 396 38 51 49 66 0.8 1 60 83 0.8 1 280 375 8-1 Uninterruptable power No No No No No No Yes* No Yes* No No No No No No Yes* No No No Standby electrical peak power load --0 0 0 0 40 54 25 34 10 13 5 7 5 7 N I A N I A 101 1 35 40 54 67 90 215 288 295 396 10 13 49 66 ob o b 60 83 0.8 1 140 188

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

• NomtWESTMlDfCALtSOTDKS NWMl-2013-02 1, Rev. 1 Chapte r 8.0 -E l ect r ical Powe r Systems Ta bl e 8-1. S umm a r y of R a dioi s o to p e Producti o n F acili ty and An cillar y Fac iliti es E le ct rical L oads (2 p ages) Demand F a cility chilled water system Facility h eate d wate r sys t e m Proce s s stream sys tem D emine r a l ized wate r system Supp l y a ir s y s t e m C h emica l s up ply system Fac il ity proce ss contro l and comm u nications s ystem s Energy recovery Safeguard s and s ecurity Administrative building Waste m a nagem e nt bui l ding Normal electrical peak power load --1 , 300 4 7 0.8 0.8 49 5 5 40 90 1 I 1 , 743 63 66 7 7 54 1 21 15 ' Uninterruptable power No No N o No No Yes No Y es No No a O n ly part s of the sy s tem are provided with uninterrupt a ble power s uppli es. Standby electrical peak power load ----0 0 0 0 0.8 0 0 49 66 5 7 0 0 40 54 18 24 3 4 b Th e fire detection and fire a l arm sub sys tem s w ill be pro v ided by a n uninterr u ptable power supp ly w ith a 2 4-hr capacity. C hapter 9.0 pro v ides a ddition a l detail. C AAS = c ritica li ty accident a larm s yst e m N I A = not a pplicable.

8.1 N O RMA L ELECT RI CAL PO WE R SYSTEMS The NEP system will connect to e l ectric utility power from t h e off-s ite utilit y tran s mis s ion and distri b ution system at a point of common coup l ing. This poin t of common co u pling wi l l b e located near the property line on t h e NWMI site. The NEP distribution system will operate in a redun d ant electrical syste m topo l ogy from t h e utility t ransmission and d istributio n system to the 480 volt (V) service entrance s witc h gear t h at s ervices t h e RPF e l ectrica l di s tri bu tion s ys t em an d the de v i ces and equipment within t he faci l ity. T h e RPF e l ectrica l distrib u tion s ystem is d esigned to s u pport the safe t y functions p rotecting workers , the pu b lic , special nuclear ma t erial activities , an d radioisotope production operation processes , as describe d in C h apter 4.0, "Rad i o i sotope Production Faci l ity D escript i o n ," and to minimize the num b er of poi nt s where a fai l ure in the RPF is a s ingle point of power conveyance.

Figure 8-1 provi d es e l ectrical one-line diagram s for the electrica l di s tri b ution topolog y. Power wi ll b e provided to the NWM I si t e from an undergroun d ut il ity fee d 0 t o the pad-m ounted swi t c h gear located outside of t h e RPF Power wi ll the n be routed un d erground from t h e switchgear to the Administra ti ve Bui l ding fl and th e RPF 8. The und ergroun d fee d ers 8 to t h e RPF wi ll comprise two re dun dant fu ll-capac i ty service l a t erals to the RPF. Eac h service l a t eral will s u p p ort redu nd an t fu ll-capac i ty service transformer s 0 t h at will norma ll y carry ha l f the RPF l oa d. E i ther of t h e RPF fee d ers can be o p ene d and the tie breaker c l ose d , as nee d ed , allowing t h e other fee d er to carry th e entire RPF l oad. Any RP F l oad s re qu iring SEP wi ll be provi d ed p ower from t h e d iesel g enerator when req ui red 0. 8-2

[Proprietary Infonnation]

Figure 8-1. Radioi so tope Production Faci lit y Electrical One Line Diagram NWMl-2013-021 , Rev. 1 Chapter 8.0 -Electrical Power Systems 8-3 NWMI ...... ..* ... **** .. .. .. * !*. * !

• NOITHWUT MEDICAi. ISOTOPES NWMl-2013-021 , Rev. 1 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 (HVAC) 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.

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. 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. 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 , Recomm e nd e d Pra ct i ce for the Desi gn of R e liabl e Industrial and Commercial Pow e r Systems, and IEEE 379 , Standard Application of the Single-Failure Criterion to Nuclear Pow er Generating Station Safety Systems. The electrical distribution system topology wi ll employ a redundant power conveyance system. The distribution system will include overcurrent protective devices , surge arresters , fusing , relays , and similar safety-rel ated protective devices. These safe ty 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 wi ll 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. 8-4 NWMl-2013-021, Rev. 1 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 of NEP , 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). 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 los s of normal power: * * * * *

• Inlet bubble-tight isolation dampers within the Zone I ventilation system will close , and the HV AC 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 off gas will be confined in the off gas piping , vessels , and pressure-relief tank Process vessel emer g ency 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 potentia l for transfer of process fluids to waste tanks Equipment providin g a motive force for process activities will cea s e , including:

Pumps performing liquid transfers of process solution s Pumps supporting operation of the steam and cooling utility heat transfer fluids Equipment supporting physical transfer of item s (primarily cranes) 8.1.3 Ranges of E l ectrical Power Required The RPF power service will be 480 V, 3-phase , 1 20 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. 8-5

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

• NORTHWEST MEDJCAL ISOTDPfS NWMl-2013-021, Rev. 1 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 s ystem will not require a technical specification per the guidelines in Chapter 14.0, " Technical Specifications

." 8-6

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

• 0 NOITHWEST MEDM:.Al lSOTOPO 8.2 EMERGENCY ELECTRICAL POWER SYSTEMS NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems E mergency electrical power i s defined by NUREG-1537 , Guidelines for Pr eparing and R eviewing Ap pli ca tions for the Licensing of Non-Power R eactors -Format and Content , as any temporary substitute for normal electrical service. A combination ofUPSs and the SEP system will provide emergency e le c trical power to the RPF , although only selected UPS sys tem s wi ll have a safety function.

A 1 , 000 kW (1 , 341 hp) diesel generator will provide SEP. Figure 8-1 also provide s the electrical distribution topology for the SEP system. Power from this ge nerator will serv ice the RPF through an automatic transfer switc h (A TS). The normal power side of th e ATS will be connected to the RPF service entrance switc hgear , with the load side of the ATS to be connected to the standby sw itchboard.

The SEP system is designed to support the safety function s durin g RPF operations to protect workers, the public , and environment.

The SEP system design include s recommended practic es from IEEE 446 , R ecommende d Pra ctice for E m erge n cy and Standby Power Systems for Indu strial and Commercial Applications, NFPA 110 , Standard for Eme r gency and Standby Power Systems , IEEE 379, and IEEE 49 3. The SEP system will include overcurrent protective de vices, surge arresters, fusin g, relays , and si milar safety-related protective devices. The se safety device s 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 di s tribution 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 ve ntilation syste m , and other loads requiring standby power. Feeders , busing , overcurrent protection , de vices, and equipment will provide the conveyance and conductor protection throughout the building.

During normal operations , loads connected to the standby switchboard will be service d through the ATS with normal and faci lit y electric power. In this way , any malfunctions of the SEP sys tem durin g RPF operation with NEP will not interfere with normal RPF operations or prevent safe facility shutdown.

When the A TS senses a l oss 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 limit s, 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 wi thin acceptable limit s and after a pre scri bed delay , the A TS will swi tch from the diesel generator power source to the normal power source. UPSs will be pro v ided , as required. The function of the UPSs is to pro v ide power to se lect 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 gallons [gal]) diesel tank. This capacity will enable the generator to operate for 11-14 hr , depending on actual loads , without requiring additional fuel. 8-7

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

• NOl11fWEST MEDICAL ISOTOPCS 8.2.1 Design Basis of the Emergency Electric Power System NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems 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 Joss of NEP. The system design basis also provides SEP to operate select process-related equipment to limit the impacts of loss of NEP 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 I , 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 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.

8-8

.; ... NWMI ...... ..* ... .*.* .. *.*.* ' *.* !' . NOllTHWUT MEDICAi. lSOTOP£S NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems 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 of HVAC 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. 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 system s is not considered an IROFS. 8.2.9 Power for Emergency Cooling Based on the analysis pro v ided 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 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.

8.2.11 Power for Emergency Lighting Power required for emergency lighting will be provided by 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.

8.2.12 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 s ystems , as required , are anticipated to be part of the technical specification for the system being supported.

The SEP system will not require a technical specification per the guidelines in Chapter 14.0. 8-9

8.3 REFERENCES

NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems 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 Commercia l 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 Jerse y, 2007. NFPA 70 , Nationa l Electri ca l Code (NEC), National Fire Protection Association, Quincy, Massachusetts, 2014. NFPA 110 , Standard for Emergency and Standby Pow e r Systems, Institute of E le ctrica l and Electronics Engineers , Piscataway, New Jer sey, 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. 8-10

.....

..*... ..* .... ..... .. .. .. . *****. * *** *

• NOflTHWEST M£lJICAl WITOPEI NWMl-2013-021, Rev. 1 Chapter 8.0 -Electrical Power Systems This page intentionally left blank. 8-11 -------1

• * * * * * * * * ****** * * ** * * * ** * ** * * * ** * ** * * ** * * . *. *. * . NORTHWEST MEDICAL ISOTOPES *

• Chapter 9.0 -Auxiliary Systems Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021, Rev. 1 June 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, OR 97330 I NWMI ..*...... * . .............. .

• NORTHWESTMEDICAUSOTOPES This page intentionally left blank.

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems I .. NWMI ...... .. .. . .......... * * !' * . NOllTHWEST MEDICAl ISOTOPES NWMl-20 1 3-021 , Rev. 1 C h ap t e r 9.0 -Aux il iary Sys t em s Chapter 9.0 -Auxiliary Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 1 Date Published:

June 26 , 2017 Document Number. NWMl-2013-021 I Revision Number. 1 Title: Chapter 9.0 -Auxiliary Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

c w.J"rfv<-

I .. NWMI *::.**.*.*. *.*.* .. *.*.* . NOmtWESTMEDtCAl.lSOTI>P£S This page intentionally left blank. NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems r I .; .* ;*.NWMI ...... .. .. . ........ *. * * * *

• NOPITitWEST MEDICAL ISOTOPES Rev Date 0 6/29/2015 1 6/26/2017 REVISION HISTORY Reason for Revision Initial Application NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Revised By Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional Information

.. .. NWMI ...... .. .... ......... *.* * * * . NOllTHWEST M£DICAl I SOTOPES This page intentionally left blank. NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems I . .. NWMI ...... .. .. .*.* .. *.*. e * . NOlmfW'EST MEDICAl I SOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary System s CONTENTS 9.0 RADIOISOTOPE PRODUCTION FACILITY AUXILIARY SYSTEMS ................................. 9-1 9.1 Heating Venti lation and Air Co nd itioning System s ..................................

.......................

9-1 9.1.1 Design Basis ......................................................................

......................

........ 9-2 9 .1.2 System D escription

...................

..................

..................................................... 9-2 9 .1.2.1 Confinement

.................................

...............

.................................... 9-6 9.1.2.2 Supply Air System .........................

..........................................

........ 9-7 9.1.2.3 Exhaust Air System ..............................................

....................

..... 9-10 9.1.2.4 CleanroomSubsystem

....................

........................

..............

.......... 9-13 9.1.2.5 Ph ys ical Layo ut and Location .................................................

....... 9-14 9.1.2.6 Principles of Operation

....................................

........................

...... 9-14 9.1.3 Operational Analysis and Safety Function ......................................................

9-15 9 .1.4 In s trumentation a nd Co nt rol Requirement s .....................................................

9-16 9.1.5 Required Technica l Spec i fications

..................................................................

9-16 9.2 Material Handling .............

....................

......................

....................

.............................

9-17 9.3 Fire Prote ction Systems and Program s .................................

.............

...........................

9-18 9.3. l De s ign Basis .........................................................

....................................

..... 9-18 9.3.2 System De sc ription ..............

...............

....................................................

....... 9-18 9.3.2.1 Fire Suppression Subsystem

...................................

........................

9-19 9.3.2.2 Fire Detection and Alarm Subsystem

......................................

....... 9-20 9.3.2.3 Fire Ext ingui s h ers ...............................................................

........... 9-21 9.3.3 Operat ion a l Ana l ys is and Safety Function ......................................................

9-21 9.3.3.1 Radioisotope P roduction Faci lit y Fire Areas ...............

.................... 9-22 9.3.3.2 Other Radioisotope Production Faci lit y Systems ............................

9-36 9.3.3.3 Architectura l Feature s ..........................

..........................................

9-36 9.3.4 In str umentation and Control Requirement s ......................

...............................

9-37 9.3.5 R eq uired Technical Specifications

.....................................

.............................

9-38 9.4 Co mmunic atio n Systems ...........

................

.................................

................................

.. 9-39 9.4.1 D es ign Ba s i s .................

...............................................

.................................. 9-39 9.4.2 System Description

.........................................................................

...............

9-39 9.4.3 Operational Anal ys i s and Safety Fu nction ...............................

....................... 9-39 9.4.4 In s trumentation and Co nt rol R e quirements

...............................

...................... 9-40 9.4.5 Required Technical Specifications

.......................

........................................... 9-40 9.5 Posses sio n and Use of Byproduct , Source, a nd Special Nuclear Material..

.................... 9-41 9.5.1 Desi g nBa s is ....................

.........................................

.....................................

9-41 9.5.2 System De sc ription ..................................................................

......................

9-41 9.5.2. l Special N u c l ea r Materials

................

.....................................

......... 9-41 9.5.2.2 Byproduct Material s ...................................

....................................

9-4 2 9.5.3 Operational Anal ys is and Safety Function ..............................

........................ 9-4 2 9.5.4 In str umentation an d Co ntr o l R e quirement s ..................

................................... 9-4 2 9.5.5 R equire d Technical Specifications

......................

..........................................

.. 9-42 9.6 Cover Gas Contro l in C lo se d Prim ary Coo l ant Systems ................

................................

9-43 9.6.1 D esign Basis ................................

...............

................................................... 9-43 9.6.2 System Description

........................................

...........

.....................................

9-43 9.6.3 Oper a tional Anal ys is an d Safety Function ...................................

...................

9-4 3 9.6.4 In s trumentation and Contro l R e quirement s ...........

..........................................

9-4 3 9.6.5 R eq uired Technical S pecifi cations .........................

...........

.............................. 9-4 3 9-i

.*; .. ;* .. NWMI ...... .. .. . ........... NORTHWESTME.DtCAl l SOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems 9.7 Other Auxiliary Systems .................................................................................

............. 9-44 9.7.1 Utility Systems ...........................................

.................

................

................... 9-44 9. 7. I. I Design Basis ...........................

............................

...........................

9-44 9.7.1.2 System Description

..............

..............................................

............ 9-44 9. 7. 1.3 Operational Analysis and Safety Function ................

......................

9-6 I 9. 7 .1.4 Instrumentation and Control Requirements

..................................

... 9-6 I 9.7.1.5 Required Technical Specifications

...........

......................................

9-61 9.7.2 Control and Storage of Radioactive Waste ..................

....................

................ 9-62 9. 7 .2. I Design Basis ................

....................

..............................................

9-62 9.7.2.2 System Description

...................

..........

..........

..................

............... 9-62 9. 7 .2.3 Operational Analysis and Safety Function ................

......................

9-76 9.7.2.4 Instrumentation and Control Requirement s ...............................

...... 9-76 9.7.2.5 Required Technical Specifications

............................

.....................

9-76 9.7.3 Analytical Laboratory

................................................

...........................

.........

9-77 9. 7 .3. I Design Basis ......................................................

.........................

... 9-77 9.7.3.2 System Description

...................................

.................................

.... 9-77 9.7.3.3 Operational Analysis and Safety Function ...................................

... 9-78 9. 7 .3 .4 Instrumentation and Contro l Requirements

...............

......................

9-79 9.7.3.5 Required Technical Specifications

..................

............................... 9-79 9.7.4 Chemical Supply ....................................

.........................................

............... 9-79 9. 7 .4.1 Design Basis ..............................................................

....................

9-79 9.7.4.2 System Description

.................................

.......................................

9-79 9. 7.4.3 Operational Analysis and Safety Function .......................

............... 9-86 9. 7.4.4 Instrumentation and Contro l Requirements

..................................... 9-86 9.7.4.5 Required Technical Specifications

................

............

.....................

9-86 9.8 References

........................................................................

..............

............................. 9-87 9-ii I

..... .. .. .. *. ." . NORTHWEST MEDICAL IS O T O PES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-1. Figure 9-2. Figure 9-3. Figure 9-4. Figure 9-5. Figure 9-6. Figure 9-7. Figure 9-8. Figure 9-9. Figure 9-10. Figure 9-11. Figure 9-12. Figure 9-13. Figure 9-14. Figure 9-15. Figure 9-16. Figure 9-17. Figure 9-18. Figure 9-19. Figure 9-20. Figure 9-21. Figure 9-22. Figure 9-23. Figure 9-24. Figure 9-25. Figure 9-26. Figure 9-27. Figure 9-28. Figure 9-29. Figure 9-30. Figure 9-31. Figure 9-32. Figure 9-33. Figure 9-34. Figure 9-35. Figure 9-36. Figure 9-37. Figure 9-38. Figure 9-39. FIGURES Ground Level Confinement

.....................................................................................

.... 9-4 Upper Level Confinement

..........................................

................................................. 9-5 Lower Level Confinement

..................................................

.........................................

9-6 Facility Ventilation System Diagram 1 ..................................

...................................

... 9-8 Facility Ventilation System Diagr am 2 .......................

.............................................

.... 9-9 Process Flow Diagram for Proces s Vessel Ventilation Treatment..

..........

................... 9-12 Life Safety Plan (First F loor) .....................

.................

...............................................

9-23 Life Safety Plan (Second Floor) ............................................................................

..... 9-24 Second Floor Mechanical Utility Area ........................

...............................

................ 9-45 Medium-Pressure Steam System ..............................

.................................................. 9-46 Low-Pressure Steam System ...................

..................

................................................. 9-47 Chilled Water System Large Geometry Hot Cell Loop ..........................................

..... 9-50 Chilled Water System Cr iticall y Safe Hot Cell Loop .................................................. 9-51 Chilled Water System Target Fabrication Loop .................................

..................

....... 9-52 Proce ss Chilled Water System ................

................................................................... 9-5 3 Demineralized Water System .........................................

..................

..........................

9-56 Plant Air System .................

..........................

.....................................

....................... 9-57 Nitrogen and Helium Supply System .................

....................

..........................

..........

9-58 H y drog e n and Oxygen Supply System ....................................................................... 9-59 Waste Management Process Flow Diagram and Process Flow Streams ......................

9-64 High-Dose Liquid Waste Solidification Subsystem and Low-Dose Collection Tank Location ......................

.....................

............................................

....................

9-65 S implifi ed High-Dose Waste Handlin g Process Flow Diagram ..............................

.... 9-66 High-Dose Waste Treatment and Handling Equipment Arrangement

.........................

9-67 Low-Dose Liquid Waste Evaporation System Location ...........................................

.. 9-68 Low-Dose Liquid Waste Dispo s ition Process .......................

................................

...... 9-69 Low-Dose Li quid Waste Solidification E quipment Arrangement

...............................

9-70 Spent Re sin Dewaterin g Operational Flow Diagram .................................................. 9-71 Spent Re sin Collection Tanks Location ................................

...................................... 9-71 Solid Waste Encapsulation Operation a l F low Di agram ............................................

.. 9-72 High-Dose Waste Deca y Operational Flow Diagram ..........................................

....... 9-72 High Dose Waste Deca y Cell Equipment Arrangement..

.....................

..............

......... 9-7 3 High Do se Waste Handlin g Operational Flow Dia gram ............................................. 9-7 3 Waste Handling Flow Diagram .......................

......................................

.....................

9-74 Waste Handling Equipment Arrangement

.................................................................. 9-75 Analytical Laboratory L ayo ut .....................................

...............................................

9-78 Chemical Supply Room E quipment Layout .........................................

...................... 9-80 Nitric Acid Flow Diagram ..........................

...........................................

....................

9-81 Sodium Hydroxide Flow Diagram .............................................

................................ 9-83 Hydrogen Peroxide Flo w Diagram ........................

.....................

............

...............

..... 9-84 9-iii I .. NWMI *********** .............. . . ." . NORTHWEST MEDICAl I SOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Table 9-1. Table 9-2. Table 9-3. Table 9-4. Table 9-5. Table 9-6. Table 9-7. Table 9-8. Table 9-9. Table 9-10. Table 9-11. TABLES Facility Areas and Respective Confinement Zones ........................

...............................

9-3 Indications for Facility Ventilation System Parameters

..................

................

............

9-16 Purge Gas Flows ............

...................................

...................

....................

.................

9-60 Tanks Requiring Purge Gas ................................................

.......................................

9-60 High-Dose Waste Tank Capacities

............

...................................................

..............

9-66 Low-Dose Waste Tank Capacities

....................................

......................................... 9-69 Subsystem 100, Nitric Acid Tank Sizes .......................................

.............................. 9-82 Subsystem 200, Sodium Hydroxide Tank Sizes ..........

...............................................

9-83 Subsystem 300 , Reductant and Nitrogen Oxide Absorber Solutions Tank Sizes ......... 9-84 Subsystem 400, Hydrogen Peroxide Tank Sizes ...................................

...................... 9-84 Subsystem 600 , Fresh Uranium Ion Exchange Resin Tank Sizes ................................

9-85 9-iv I NWMI *::.**.*.*. *.*.* .. *.*.* . *. "NORTHWESTMEDICAl.ISOTOPES TERMS Acronyms and Abbreviations 89 Sr strontium-89 9 0 Sr strontium-90 9 9 Mo molybdenum-99 23 0 Th thorium-230 231 Pa protactinium-231 233 Pa protactinium-233 233 U uranium-233 235 U uranium-235 237 Np neptunium-237 238 Pu plutonium-238 238 U uranium-238 239 Pu plutonium-239 24 0 Pu plutonium-240 2 41 Arn americium-241 ALARA as low as reasonably achievable CFR Code of Federal Regulations DBF design basis fire DOT U.S. Department of Transportation H 2 hydrogen gas HEGA high-efficiency gas adsorption HEPA high-efficiency particulate air HIC high-integrity container HN0 3 nitric acid HVAC heating , ventilation, and air conditioning IBC International Building Code NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems ICP-MS inductively coupled plasma mass spectrometry IROFS item relied on for safety IRU iodine removal unit IX ion exchange Kr krypton LAN local area network LEU low-enriched uranium Mo molybdenum MURR University of Missouri Research Reactor NaOH sodium hydroxide NESHAP National Emission Standards for Hazardous Air Pollutants NFPA National Fire Protection Association NO x nitrogen oxide NRC U.S. Nuclear Regulatory Commission NWMI Northwest Medical Isotopes , LLC OSTR Oregon State University TRIGA Reactor OSU Oregon State University PFHA preliminary fire hazards analysis RCA radiologically controlled area RPF Radioisotope Production Facility SNM special nuclear material Ti0 2 titanium dioxide U uranium 9-v

.. NWMI ...... .. .... ............ * * . NOllTHWEST M£DtCAl. tSOTOPES U.S. u.s.c. VoIP Xe Units o c O F µ cm cm 2 ft ft 2 ft 3 gal gmol hr in.

• 2 m. kg L lb m M m 2 m 3 mm mm w wt% United States United States Code Voice over Internet Protocol xenon degrees Celsius degrees Fahrenheit micron centimeter square centimeter feet square feet cubic feet gallon gram-mo I hour inch square inch kilogram liter pound meter molar square meter cubic meter minute millimeter watt weight percent 9-vi NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems

.;*. NWMI ...... .. .. . **** .. .. .. NORTHWESTMEDICALISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems 9.0 RADIOISOTOPE PRODUCTION FACILITY AUXILIARY SYSTEMS This chapter provides the descriptions of the auxiliary systems for the Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) that have not been addressed in previous chapters.

These auxiliary systems are important to the safe operation of the RPF and to protect the health and safet y of workers , the public , and environment.

The chapter is organized in accordance with NUREG-1537 , Guidelin e s for Pr e parin g and R ev i ew ing Appli c ation s for th e Li ce nsing of N on-Po wer R e actors -Format and Cont e nt , as augm e nted by the Final Interim Sta ff Guidan ce Augmentin g NU REG-153 7 , " Guid e lin es for Pr e paring and R e vi ew ing Appli c ations for th e Li ce n s ing of No n-Po we r R e a c tor s ," Part s 1 and 2 , for Li ce n s in g Radioi s otop e Produ c tion Fa c iliti es and A qu e ou s Homo ge neous R e a c tor s (NRC , 2012). The RPF auxiliary s ys tems include the following:

• * * * *

• Heating and ventilation , and air conditioning (HV AC) systems Fire protection systems Communication systems Possession and u s e of byproduct , source , and s p e cial nuclear material Co v er ga s control in the closed primary coolant sys tem Other auxiliar y sys tems , including utility sy stems , control and storage of radioactive waste, analytical laborator y, and chemical suppl y For each auxiliary system, a description is provided of the system's capability to function as designed without compromising RPF operations and to shut down the RPF during normal operations or under RPF accident conditions. Each auxiliary s y stem description includes:

• * * *

• 9.1 Design basis System description Operational anal y si s and safet y function Instrumentation and control requirements Required technical s pecification s and their ba s e s, including testin g and s urveillance HEATING VENTILATION AND AIR CONDITIONING SYSTEMS The RPF HVAC system , al s o referred to as the facilit y ventilation system , is de s igned to ensure that temperature , relative humidity , and air exchange rates are within the design-basi s limits for personnel and equipment and to ensure that all normal sources of airborne radioactive material are controlled s o that occupational doses do not e x ceed the requirements of Title I 0 , Cod e of Federal R eg ulations , Part 20 , " Standards for Protection A g ainst Radiation" (10 CFR 20). The s y stem design i s consistent with NWMI's as low a s rea s onabl y achievable (ALARA) program. The RPF design features en s ure that airflow and relati v e pressure will prevent inadvertent diffusion or other uncontrolled release of airborne radioactiv e material from the RPF. The facility is also designed and operated to ensure that no uncontrolled release of airborne radioactive material to the unrestricted e n v ironment can occur. The analyses of system operations show that planned releases of airborne radioactive material to the unrestricted environment will not expose the public to doses that exceed the limits of 10 CFR 20 and the NWMI ALARA program. NWMl's ALARA program is discussed in Chapter 11.0 , " Radiation Protection Program and Wa s te Management

," and a detailed airborne exposure analysis is provided in Chapter 11 ,Section I 1. 1 .1.1 .2. 9-1

.. ; ... ;.NWMI ...... .. .. . ..... .. .. .. .... . * * . NORTHWEST MEOtCAl ISOTOPES 9.1.1 Design Basis NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The facility ventilation system is designed to provide confinement of hazardous chemical fumes and airborne radiological materials and conditioning of the RPF environment for facility personnel and equipment.

The design basis of the facility ventilation system and the process vessel ventilation system is provided in Chapter 3.0, " Design of Structures , Systems , and Components

," Section 3.5.7.2; and the s afet y functions are provided in Chapter 6.0 , " Engineered Safety Features ," Section 6.2.1.1. 9.1.2 System Description The facility ventilation system will maintain a series of cascading pressure zones to draw air from the cleanest areas of the facility to the most contaminated areas. Zone IV will be a clean zone that is independent of the other ventilation zones. Zone IV will be slightly positively pressurized with respect to the atmosphere.

Zone III will be the cleanest of the potentially contaminated areas , with each subsequent zone being more contaminated and having lower pressures. A common supply air system will provide 100 percent outdoor air to all Zone III areas and some Zone II areas that require makeup air in addition to that cascaded from Zone III. Three separate exhaust s y stems will maintain zone pressure differentials and containment: (I) the Zone I exhaust system will service the hot cell, waste loading areas , target fabrication enclosures , and process vessel ventilation subs y stems in Zone I; (2) the Zone II/III e x haust sy stem will service exhaust flow needs from Zone II and Zone III in excess of flow cascaded to interior zones; and (3) a laboratory exhaust system will service fume hoods in the laboratory area. Supply air will be conditioned using filters , heater coils , and cooling coils to meet the requirements of each space. Abatement technologies , primarily high-efficiency particulate air (HEPA) filtration and activated carbon , will be used to ensure that air exhausted to the atmosphere meets 40 CFR 61 , " National Emission Standards for Hazardous Air Pollutants" (NESHAP), and applicable State law. A stack monitoring and sampling system will be emplo y ed to demonstrate compliance with the s tated regulatory requirements for exhaust. The RPF ventilation system will include the air supply , process ventilation, and exhaust air s y stems and associated filters , fans , dampers , ducts , and contro l instrumentation.

The supply air system will draw in and condition fresh air and distribute it throughout the facility. A portion of the supply air will enter the process ventilation system through fume hoods , open-front enc lo sures, gloveboxes , and hot cells , and will be removed with other exhaust air systems through the stacks to the environment after being treated. The safety functions of the ventilation s y stems will serve to protect workers , the public , and environment b y maintaining confinement barriers in a multiple confinement barrier system. The RPF will typically be ventilated such that airflows travel from areas oflower potential for contamination to areas of higher potential.

The ventilation system functions will include temperature and air quality control to meet production and worker needs. 9-2

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

• NORntWEST MEDICAl ISOTOPES The RPF building ventilation system will have four confinement zone designations, with airflow directed from lowest to highest potential for contamination:

Zone I, Zone II , Zone III, and Zone IV. Figure 9-1 through Figure 9-3 show the facility confinement boundaries on the ground level (first level), upper level (second level), and lower level (basement), respectively.

Table 9-1 defines the confinement zone applicable to major spaces. The zones are defined as follows: * * *

• Zone I, shown in pink, is the initial confinement barrier and includes gloveboxes, vessels, tanks , piping, hot cells, and the Zone I exhaust subsystem.

Zone II, shown in orange, is the secondary confinement subsystem and includes the walls, floors, ceilings, and doors of the laboratories with the gloveboxes, HEP A filter rooms, and the Zone II ventilation exhaust subsystem.

Laboratory gloveboxes and fume hoods are also Zone II. Zone III, shown in green, is the tertiary confinement barrier and includes the walls floor , ceilings, and doors of the corridor that surround the operating galleries, and the mechanical mezzanine.

Zone IV, shown in blue, is the confinement ventilation zone -the positively pressurized areas served by unitary , non-safety , and grade equipment.

These areas will include the administration support area , truck bays, and maintenance utility areas. NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Table 9-1. Facility Areas and Respective Confinement Zones Area Hot cells (production)

Tank hot cell Solid waste treatment hot cell High dose waste solidification hot cell Uranium decay and accountability hot cell HIC vault Analytical laboratory gloveboxes R&D hot cell laboratory hot cells Target fabrication room and enclosures Utility room Analytical laboratory room and hoods R&D hot cell laboratory room and hoods Waste loadin g hot cell Maintenance gallery Manipulator maintenance room Exhaust filter room Airlocks* Irradiated target basket receipt bay Waste loading truck bay Operating gallery and corridor E lectrical/mechanical supply room Chemical supply room Corridors Decontamination room Loading dock s *+],[* I I I I I II II II II II II II II II, III III III III III III III III IV Waste management loading bay IV Irradiated target receipt truck bay JV Maintenance room IV Support staff areas IV

• Confinement zone of airlocks will be dependent on the two adjacent zones being connected.

HIC high integrit y container.

R&D = research and dev elop ment. 9-3 I .. NWMI .*:.**.-.* . ......... . *. .°. NORTHWEST MEDICAL ISOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-1. Ground Level Confinement 9-4 I .*;.-.;*. NWMI *::**:::* ..*... e * ." . NOllllfWEST MEDICAL ISOTOPfS [Proprietar y Information]

Figure 9-2. Upper Level Confinement 9-5 NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems I ... .. NWMI ............ ......... *.*

• 0 * ! .' , NOltntwEST MEDICAl. ISOTOPES 9.1.2.1 Confinement

[Proprietary Information]

Figure 9-3. Lower Level Confinement NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Confinement is an engineered safety feature of the HV AC system. Confinement is the term used to describe the boundary that surrounds radioactive materials and the associated ventilation system. Confinement systems are designed to localize any release of radioactive material to controlled areas in normal operational states and to mitigate the consequences of design basis accidents.

Radiation protection control features (e.g., adequate shielding and confinement ventilation systems) minimize hazards associated with radioactive materials.

The principal design and safety objective of the confinement system is to protect on-site personnel and the off-site public. The second design objective is to minimize the reliance on administrative or complex active engineering controls to provide a confinement system as simple and fail-safe as reasonably possible.

The process vessel ventilation system will serve as the primary confinement pressure boundary and is safety-r elated. The Zone I exhaust subsystem is an engineered safety feature that (along with shielding) will create a secondary confinement boundary; enclosing the vessels and process off gas within the hot cells. Confinement of the hot cells will be achieved through both the confinement ventilation system and the shielding provided by the steel and concrete structures comprising the walls , roofs , penetrations , and covers of the cells. Secondary confinement will be accomplished by the zone boundaries , associated ventilation systems, and HEP A filter plenums to filter exhaust air prior to discharge at the facility ventilation stacks. Secondary confinement will also be accomplished through the use of bubble-tight isolation dampers. These dampers will isolate the ducts at the zone boundary under certain scenarios to ensure that all potential releases have been HEPA-filtered prior to exiting the facility (i.e., release to atmosphere).

The safety aspects of the confinement system are discussed in Chapter 6.0 , "Engineered Safety Features," Section 6.1, including the design response to off-normal conditions (e.g., Joss of power). 9-6

.*; .. NWMI ...... .. .. . ........... .' . NOPITlfWEST MEDICAl I SOTOPES 9.1.2.2 Supply Air System NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The RPF supply air system will provide conditioned air for facility workers and equipment and supply makeup air for RPF exhaust air systems. The supply air system will provide filtered and conditioned air to all Zone III spaces and some Zone II spaces at a ventilation rate of I 00 percent outside air. The three air supply handling units will be sized at 50 percent capacity each, for redundancy.

Two of the three units will be operating, while the third is on standby. If a single unit fails, the standby unit will start automatically.

Each unit will consist of an outdoor air louver, filters, cooling coil, heating coil, heat recovery coil , isolation dampers, and a fan. Variable-speed fans will be modulated to control the pressure in the common air plenum. The heating and cooling coils in each air-handling unit will be controlled based on a common supply air temperature sensor. Reheat coils will be provided in the supply ducts to each space, as required, to further condition the supply air , based on space temperature thermostats.

Outside air will be drawn into the RPF air supply system through air-handling units (Figure 9-4). The units will normally supply a constant volume of conditioned air to the Zone II and Zone III areas of the RPF. Zone III air will be cascaded into Zone II areas through engineered leakage pathways by a negative pressure differential, maintaining the desired pressure drop between the zones (Figure 9-4). Terminal unit components in the supply duct system will include airflow control valves and reheat coils. The terminal reheat coils will provide final tempering of the supply air to maintain the Zone II space temperature setpoint.

Zone II supply airflow control valves will operate in conjunction with exhaust valves to control the pressure differential in each zone by maintaining a fixed difference between the total supply and exhaust air flows for each Zone II space. Exhaust from Zone II will be expelled through the 23 meter (m) (75-foot [ft]) high Zone II exhaust stack. Additional detailed information on the Zone II stack design will be developed for the Operating License Application.

The isolation dampers and backdraft dampers in the supply duct system at the zone boundary (Figure 9-5) will close when required to provide confinement at the zone boundary. The supply air system HVAC controls will operate through the building management system. 9-7 I ... .. NWMI ..*...... *. .............. . *. ." . NORTifWEST MEDICAl. ISOTOPES [Propri e tar y Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-4. Facility Ventilation System Diagram 1 9-8 I ... NWMI .*:.**.*.* . .............. , *. * *. ! ." . NORTHWtST MEDICAL ISOTOPES [Proprietar y Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-5. Facility Ventilation System Diagram 2 9-9

.:; .. ; ... NWMI .*;.**.-.* . .............. * .", NORTffWESTMEDICAL ISOTOPES 9.1.2.3 Exhaust Air System NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The RPF will have four exhaust air subsystems:

Zone I exhaust, Zone II/III exhaust , laboratory exhaust , and process vessel ventilation exhaust. Each exhaust system will be provided with two 100 percent capacity exhaust fans and filter trains for complete redundancy on all exhaust subsystems. This redundancy is important to ensure confinement ventilation pressure differentials are maintained at all times. [Proprietary Information).

Exhaust ducts upstream of the filter trains will be round to minimize areas where contamination can accumulate, and are sized to minimize particulate settling in the duct. Each exhaust system will have a separate stack, with the exception of the process vessel ventilation subsystem , which will merge with the Zone I exhaust stream. A stack monitoring and sampling system will be provided on each stack to demonstrate compliance with applicable State law. 9.1.2.3.1 Zone I Exhaust System The Zone I exhaust system will serve the hot cell, high-integrity container (HIC) loading area, and solid waste loading area. This exhaust system will maintain Zone I spaces at negative pressure with respect to atmosphere. The disassembly hot cell station will be maintained at a slightly lower pressure due to the increased likelihood of contamination in that area. All makeup air to Zone I spaces will be cascaded from Zone II spaces. Space temperature control will not be provided for Zone I spaces unless thermal loads are expected to cause temperatures to exceed equipment operating ranges without additional cooling. HEPA filters will be included on both the inlet and outlet ducts to Zone I. The outlet HEPA filters will minimize the spread of contamination from the hot cell into the ductwork leading to the exhaust filter train. The inlet HEP A filters will prevent contamination spread in case of an upset condition that results in positive pressurization of Zone I spaces with respect to Zone II spaces. The process vessel ventilation subsystem will enter the Zone I exhaust subsystem just upstream of the filter train. The Zone I exhaust system will expel air from the hot cells and glovebox enclosures located within the RPF. The system will also capture exhaust from the process vessel ventilation system. The Zone I hot cell and glovebox enclosure will draw ventilation air from the surrounding Zone II spaces through HEPA filters. The exhaust air from each cell will pass through local HEP A filters. Negative space pressure in Zone I will be controlled through local exhaust airflow control valves for each cell. The exhaust from the cells will collect in a Zone I system duct header and then be drawn through final , testable, HEP A filters and carbon adsorbers prior to discharge into the exhaust stack. The speed of the Zone I exhaust fans will be controlled to maintain a negative pressure setpoint in the Zone I exhaust duct header. The exhaust fans will be fully redundant.

If the operating fan fails, the standby fan will start automatically.

Exhaust from Zone I will be expelled through the 23 m (75-ft) high Zone I exhaust stack. Detailed information on the Zone I stack design will be developed for the Operating License Application.

9.1.2.3.2 Zone 11/111 Exhaust System The Zone II/III exhaust system will serve the Zone II spaces and those Zone III spaces that do not provide cascaded air flow into Zone II. This exhaust system will maintain Zone II spaces at negative pressure and Zone III spaces at a less negative pressure with respect to atmosphere.

Makeup air to Zone II spaces will either be cascaded from Zone III spaces or supplied from the supp l y air subsystem to meet additional space conditioning needs. All makeup air to Zone III spaces will be provided from the supply air subsystem.

9-10

.*; .. ;.NWMI ...... .. .. ........ *. * .' NOftTMWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The RPF Zone II exhaust system will expel air from the operating areas, workrooms, and fume hoods to maintain confinement.

This confinement is important to safety to protect facility workers from radiological and hazardous chemical releases.

The exhaust air from these spaces will collect in a Zone II exhaust header and will then be drawn through final, testable, HEPA filters and carbon adsorbers prior to discharge into the exhaust stack (Figure 9-4). The exhaust fan speed will be controlled to maintain the desired negative pressure in the RPF Zone II exhaust header. The exhaust fans will be fully redundant.

If the operating fan fails , the standby fan will start automatically.

Air flow control valves in the Zone II room exhaust duct system will operate in conjunction with the zone supply valves to produce an offset between the exhaust and supply flow rates. The flow offset will enable a negative space pressure. Flow control valves in the fume hood exhaust ducts will maintain a constant volume through each fume hood. The control valves will automatically modulate to compensate for a drop in air pressure due to loading of local filters. 9.1.2.3.3 Laboratory Exhaust System The laboratory exhaust system will provide fume hood and glovebox exhaust capability.

This essentially is a Zone I system, but is separate from the main Zone I exhaust system to accommodate the large flow fluctuations from changing fume hood positions.

These highly variable flow conditions will be controlled better through a separate exhaust system. This exhaust system will minimize the potential pressure perturbations and control difficulties that cou ld result from including the fume hoods on the main Zone I exhaust system. Makeup air for increased fume hood exhaust flow will be supplied from the common supply air system. 9.1.2.3.4 Process Vessel Ve ntilation Treatment System Due to the relatively short timeframe from neutron fission operations at a reactor to target dissolution and processing in the RPF , there will be an amount of short-lived tellurium isotopes in some process streams. The decay of these tellurium isotopes will create iodine isotopes.

While most of these process streams will not likely evolve any iodine species into the offgas, this event cannot be precluded.

To ensure the safety of the facility, the off gas from these special process streams will be collected and routed to an iodine removal system. Figure 9-6 provides a flow diagram for the process vessel vent subsystems that flow to the process vessel vent iodine removal unit (IRU). The loc ations that are routed to the iodine removal subsystem include the following:

• * * * * [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

9-11 l I ... NWMI .*;.**.*.* . .............. * ." . NORTMWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-6. Process Flow Diagram for Process Vessel Ventilation Treatment Iodine removal unit for target dissolution offgas system -This system in the tank hot cell will include offgas from the target disassembly and the target dissolution offgas systems. *

• Target disassembly

-[Proprietar y Information].

Target dissolution

-[Proprietary Information].

After the offgas filter will be the dissolver offgas system's vacuum pumps and tanks , then the stream will flow through the secondary fission gas adsorbers and into the proces s vesse l vent header. 9-12

.; .. ; ... NWMI ...... .. .. . ........ *. .. NORTMWESTMEDICAllSOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Iodine removal unit for uranium, molybdenum, and waste accumulation tanks -Some of the liquids in the hot cell will contain tellurium isotopes that generate iodine isotopes during decay. A portion of the iodine will remain in the dissolver solution.

Although it is not likely that much of the iodine will evolve into the offgas, these streams will be passed through an IRU before the process vessel vent header. The expected off gas streams that feed this IRU will be from tank hot cell vessels, including the Mo feed tanks, impure U collection and lag storage tanks , U recovery waste tanks, and the liquid waste handling tanks. [Proprietary Information].

This offgas stream will flow into the process vessel vent header. General vessel vent system -This header system will service the remaining vessels in the tank hot cell , including the pure U lag storage tanks (14), recycled U collection tank, and tanks attributed to the U concentrators.

This offgas stream will flow into the process vessel vent header without additional treatment.

High volume evaporative vent from waste handling -This system will service the three waste solidification unit operations (low-dose liquid waste , high-dose liquid waste , and solid waste) and the low-dose evaporation tanks. The low-dose evaporation tanks will have high flowrate and elevated temperatures to allow water to evaporate.

The header will collect these humid air sweeps and dilute with additional air bleed to ensure that the evaporated water does not condense in the ducting or pipes. This offgas stream will flow into the process vessel vent header. Target fabrication vent -The target fabrication area ventilation is required for confining:

(1) off gas from the dissolver and other process vessel, and (2) offgas from the calcination or reduction furnace systems, where hydrogen will be diluted with air to less than the lower flammability limit. This offgas stream will flow into the process vessel vent header. Process vessel vent iodine removal unit -The process vessel vent IRU (VV-SB-520) system will consist of a sorbent bed of charcoal or activated carbon to remove iodine from the vessel vent gases. The process vessel vent IRU is part of an item relied on for safety (IROFS) RS-03, " Hot Cell Secondary Confinement Boundar y." Chapter 6.0 , Section 6.2.1 , and Chapter 13.0 , " Accident Analysis , Section 13 .2.2.8, provide additional detail on the safety function. Process vessel vent filter-This treatment operation will consist of HEPA filtration and the exhaust fan and will flow to the Zone I exhaust system. 9.1.2.4 Cleanroom Subsystem The Mo purification hot cell cleanroom subsystem is designed to provide filtered and conditioned air at an exchange rate to meet the standards of an ISO 14644-1 , " Cleanrooms and Associated Controlled Environments

-Part I: Classification of Air Cleanliness

, Class 8 cleanroom.

The cleanroom will be maintained at a slightly positive pressure relative to its surroundings to ensure that unfiltered air does not infiltrate the cleanroom.

Air inside the cleanroom will be continually recirculated through a dedicated filtration system to remove internally generated contaminants.

Air will be 100 percent recirculated, with the only air exchange with the surroundings of the c l eanroom occurring through ex filtration and makeup air entering on the suction side of the fan. The cleanroom air handling unit and filters will be located inside the hot cell and, therefore , must be remotely maintainable.

Periodic cleanroom certification testing will also need to be performed remotely with permanently installed instrumentation.

9-13

.:;.-.;* .. NWMI .*:.**.*.*. .............. . ** *. ." . NORTHWEST MEDtCAL ISOTOPES 9.1.2.5 Physical Layout and Location NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems All supply air handling units, supply fans, exhaust fans, and associated heat recovery coils will be located in the mechanical/electrical area (supply air handler room) located on the second floor. This area will house the Zone II and Zone III subsystem air-handling units and fans. The exhaust HEP A filter plenums and exhaust fans will be located in the mechanical area on the second floor. 9.1.2.6 Principles of Operation The RPF ventilation system will maintain the facility at the desired temperatures and negative pressurization during normal operations.

Supply air temperature from the air-handling units will be held constant through the use of heating and cooling coils. Reheat coils will be provided to further temper supply air to occupied areas based on local thermostat demand. The systems also have design features to maintain constant overall building pressures , the Zone I header pressure, and Zone II exhaust header pressures during normal operations.

Local room pressurization will be obtained by the airflow offset between supply and exhaust. Supply airflows will be held constant through the use of supply fan variable-frequency drives and flow measuring stations. Exhaust airflow will be controlled based on building pressure and exhaust header pressure demands and to ensure that the HEPA filter plenum rated airflows are not exceeded. frequency drives on the exhaust fans will be provided to maintain required exhaust flows when flow resistance resulting from exhaust filter dirt loading increases.

Makeup air to maintain a constant air pressure differential between the Zone II and Zone III areas will be provided by the Zone III supply air. Zone III will provide overall building pressure control during normal operations by modulation of the exhaust/return airflow path , while the supply air remains fixed. Pressure and flow conditions for the process enclosures and laboratory ventilation will be manually controlled using volume dampers and valves. Airflow control valves will be installed in each room's main supply and exhaust ducts to maintain laboratory design space pressure.

These valves will be l ocated outside of the laboratory modules. The Zone I exhaust system for each module will be adjusted manually using a valve located in the room duct header near the air inlet end to maintain minimum vacuum pressure.

A static pressure tap will be located near the air inlet end of the header and will be attached to a magnehelic gauge to monitor the header pressure relative to the laboratory module space pressure (on the radiologically controlled area [RCA]-designed portion of the system). The system is designed to maintain the Zone I process enclosures at their design pressure during normal operations and have the capacity to draw the required inflow of air in the event of a design breach of an enclosure. The Zone II exhaust system is designed to maintain the Zone II enclosures at their required pressure.

A balancing valve located in the exhaust duct of each enclosure will initially be partially closed. As the local filter of the enclosure loads up and a drop in pressure increases across the filter, the valve will be adjusted to reestablish flow in the design range. Differential pressure gauges will be provided at each enclosure to monitor the filter pressure drop and measure the pressure drop across only the enclosure.

The enclosure's pressure drop reading will be calibrated to its acceptable face velocity range to monitor enclosure performance.

9-14

.; .. ;*. NWMI *::**:*:* ..*...

• e * . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The Zone II supply air system is designed to provide the supply air volume rate required for each space. The system will supply makeup air as required for the Zone I and II process enclosures, genera l exhaust , and to maintain the design temperatures in the laboratories.

[Proprietar y Information]

to prevent the entrainment of potentially contaminated air back out of the process enclosures.

9.1.3 Operational Analysis and Safety Function Chapter I 1.0 and Chapter 13.0 pro v ide an analysis of normal and off-normal operation of the RPF HVAC system. Chapter 11.0, Section 11.1. I .1 presents that normal release analysis.

Chapter 13 .0 , Section 13.2 evaluates various accident sequences that involve failure of the ventilation components, radiological spills, and the release of high-dose solutions, vapors, or gases from within the hot cell liquid confinement , secondary confinement , or shielding boundary.

Defense-in-depth

-Failure of the air balance system is not in itself an accident, but represent s the failure of a system designed to mitigate other accidents that lead to an airborne release of radionucl ide s in the form of particulates or gases. Systems that will mitigate these releases include the primar y confinement and primar y offgas treatment system, which will capture particulates , absorb iodine, and absorb Xe and Kr and other gaseous radionuclides, to slow the release following deca y to more stable isotopes.

In the target fabrication proce sses, uranium will be handled in physical forms that do not contribute to a dose rate factor in airborne releases. Uranium solutions will also be processed in closed systems with filtered proces s ventilation systems to remove the small amounts of activity normall y released.

Item relied on for safety -Based on the Chapter 13.0 analysis, the hot cell secondary confinement (Zone I exhaust ventilation subsystem) has been de s ignated as an IROFS (RS-03, " Hot Cell Secondary Confinement Boundary"). The operations, equipment , and components of this sys tem will ensure the confinement of hazardous materials during normal and abnormal conditions , including natural phenomena , fires , and explosions.

Components of the dissolver off gas s ubsystem and the process vesse l ve ntilation syste m have also been designated as IROFS. The safety functions of the confinement system are discussed in more detail in Chapter 6.0 , Section 6.1. Chapter 13.0 evaluates a fire that could cause the carbon retention beds to ignite , leading to the release of radionuclides into the RPF exhaust stack. Based on analysis of thi s accident , the exhaust stack height wa s identified as an IROFS (FS-05 , " Exhaust Stack Height").

This analysis is discu sse d in more detail in Chapter 13.0. This passive engineered control is de signe d and fabricated with a fixed height for safe relea se of gaseous effluents.

9-15

.:;.-.;* .. NWMI ...... ..* **.* ............ *. NORTHWESTMEOICAllSOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems 9.1.4 Instrumentation and Control Requirements Section 9.1.2.6 provides a genera l description of the operation of the RPF venti lation system. Ventilation system control and monitoring is discussed in Chapter 7.0, "Instrumentation and Control Systems." Table 9-2 summarizes the system parameters (in general) and whether they are monitored or alarmed. The system seque nce of operation will be developed and provided in the Operating License Application.

9.1.5 Required Technical Specifications The technical specifications associated with the venti lation system, if applicable, will be discussed in Chapter 14.0, "Technical Specifications, as part of the Operating License App lication. Table 9-2. Indications for Facility Ventilation System Parameters Parameter E quipment operating status Damper position status Exha u st header pressure Fan speed Filter differential pressures Equipment bearing vibration Eq uipm e nt bearing temperatures HEPA filter unit air inlet temperature HEPA filter unit airflow rate First-stage HEPA inlet temperature Fan motor amperage Fan thermal overload Zone I header pressure Zone II header pressure Confinement zone pressure differentials -M 1'1 Mfui.IM ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ HEPA = high-efficiency particulate air. 9-16

.*;.-.; .. NWMI .*;.**.-.*. .......... * * ." . NOllTHWEST MEDJCAL tsOTOP£S 9.2 MATERIAL HANDLING NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The RPF does not handle or store reactor fuel. Material handling activities are discussed in Chapter 4.0 , " Radioisotope Production Fac ilit y D escript ion ," Sections 4.3 and 4.4, and are analyze d in Chapter 13.0. 9-17

.. ; .. ;*.NWMI ...... .. .. ........ *.* * * ! . , NOflTNWEST MEDICAL ISOTOPES 9.3 FIRE PROTECTION SYSTEMS AND PROGRAMS NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The frre protection system is designed to provide varying levels of notification of a fire event, suppress small fires, and prevent small fires from becoming large fires. Notification of personnel will be achieved through detection of a fire by automatic detection devices, manual pull stations, automatic sprinklers , and the use of alarm devices that broadcast within the building and transmit signals to the central alarm station and RPF control room. Suppression of fires will be accomplished through the use of automatic sprinklers where appropriate.

The suppression system will include all piping , valves , and fittings from the water supply (i.e., water storage tanks or municipal hydrants) to the automatic sprinklers and standpipes in the building.

9.3.1 Design Basis The fire protection system design provides detection and s uppression of fires in the RPF. The fire protection system design basis includes:

• *

• Providing varying levels of notification of a fire event and transmitting the notification to the site central alarm station and RPF control room Suppressing small fires Preventing small fires from becoming large fire s Additional information on the design basis is provided in Chapter 3.0 , Section 3.5.2.7. 9.3.2 System Description The fire protection system will provide detection and suppression of fires within the RPF , generation of alarm signals indicating the presence and location of fires , and execution of commands appropriate for the particular location of the fire. A complete addressable fire alarm system, with both automatic and manual initiation , will be provided throughout the RPF. Detection devices will report to a local alarm panel. AJI alarms (fire , supervisory and trouble) will transmitted to the site central alarm station and RPF control room. Fire protection sy stem components will have fail-safe features and audibl e/visual alarms for operability and trouble indication. The fire detection and alarm subsystem will include smoke detectors , heat detectors , water flow and tamper switches , manual pull stations , horns and strobes , and a notification system. The building fire s uppression subsystem will include automatic sprinkler , HEPA filter plenum deluge water sprays , and portable fire extinguishers.

Water will be supplied from the exterior fire hydrant s upply via connections to the domestic water system. Firewater booster pumps will increase the system pressure in the fire suppression subsystem piping. Space has been reserved so that if required, the fire protection system can have a dedicated water storage facility onsite. The need for dedicated storage will be dependent on the reliability and flow rate of the city water supply. The storage tank capacity is anticipated to be [Proprietary Information], and will be determined for the Operating License App lication. If an on-site water storage system is found to be necessary, an electric motor-driven fire pump will serve as the primary pressure source , and a redundant diesel engine-driven fire pump will provide backup. 9-18

.-.;*. NWMI *

:**:*:* ...... * . NORTHWEST MEDtCAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Fire protection water will be distributed throughout the building via a gridded water system. Vertical risers will supply various systems , with redundant risers also provided.

From the vertical risers, the automatic sprinkler part of the system will feed a series of sprinkler heads that have temperature-sensitive links. When a set temperature is reached at the sprinkler head, the links will melt or break (depending on type) and release water in an umbrella-shaped spray pattern. The fire protection system is designed to provide a constant flow of water to an area experiencing a fire for a minimum of 120 min. The size of that area will be determined using guidelines from the International Fire Code (IFC , 2012). For sprinkler systems, the International Fire Code uses a design based on the National Fire Protection Association (NFPA) 13, Standard for the Installation of Sprinkler S y stems. Fire hose stations will also provide flow for use in fighting fires. Because water from the sprinklers may become contaminated with materials it contacts, areas where hazardous materials are present are designed to hold firewater runoff for sampling prior to release to the env ironm ent. Additional detailed information on the firewater runoff storage will be developed for the Operating License Application.

The fire protection system is divided into two major subsystems.

The subsystems and components are categorized as follows: *

• Fire suppression subsystem consisting of automatic sprinklers, a HEPA filter plenum deluge, glovebox fire suppression, and fire hydrants Fire detection and alarm subsystem consisting of: Controls (e.g., fire alarm control panel , subpanels, or devices used for control of devices) Genera l area detection (e.g., room smoke and heat detectors, manual pull stations)

Duct smoke detection for non-nuclear ventilation systems , glovebox heat detection HEPA filter plenum heat detection Fire suppression subsystem monitoring devices (e.g., waterflow switches , tamper switches , fire pump, and water storage monitoring devices) Occupant notification Alarm transmission to the central alarm station and RPF control room 9.3.2.1 Fire Suppression Subsystem The fire suppression subsystem will include automatic sprinklers, HEPA filter plenum deluge, and fire hydrants.

The need for fire suppression in gloveboxes will be evaluated and additional information will be provided in the Operating License Application. In addition to the automatic features of the fire suppression subsystem, manual response capabilities will be provided by fire extinguishers with an appropriate classification (discussed further in Section 9.3.2.3). A 20.3 centimeter (cm) (8-in ch [in.]) network of main piping (commonly ca ll ed a grid) will be provided. Vertical piping, referred to as risers and sized at 15.2 cm (6 in.), will be provided to support the fire suppression subsystem components (sprinklers , HEPA filter plenum deluge, and hydrants).

The RPF will also be provided with redundant sprinkler risers. The connection between the risers and sprinkler piping will be provided with control va l ves, check va l ves, waterflow switches, and a test/drain assembly for detection of waterflow and system maintenance.

Piping from the risers will support automatic sprinklers located throughout the facility.

The automatic sprinkler system is designed in accordance with NFPA 13. 9-19

.......

.*;.**.*.* . ............. . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The HEPA filter plenum deluge will be also supplied by the 20.3 cm (8-in.) piping network and will be part of a larger filter plenum fire safety design that includes fire screens , demisters , plenum drains , and plenum dampers. The automatic feature will include a deluge valve that is activated via heat detectors in the ducts serving the plenum. When high temperatures are sensed in the air stream, the heat detector will send a signal to the fire alarm control panel, which in turn will send a s ignal to the deluge valve to operate. Water will flow through the deluge valve into the leading portion of the plenum to cool the air before it reaches the HEPA filters. The heat detectors and deluge valve for a particular plenum will be paired such that only plenums that are experiencing high temperatures will react. A manual bypass feature will be also provided to allow waterflow if the deluge valve fails to open. A separate, manually activated feature is designed to spray directly on the HEP A filters and is intended to only be used if the HEP A filter ignites. The manual feature wil I include a control valve connected via piping to a spray nozzle directed at the HEPA filters. The fire hydrants, located on the exterior of the building , will be supported by the 30.5 cm (12-in.) municipal water supply line provided for the RPF. Two 8-in. connections will support the 20.3 cm (8-in.) loop that surrounds the building.

Four fire hydrants , one at each corner of the building , will be provided. The fire hydrants are not designed for natural phenomenon hazards and cannot be relied on for seismic accidents.

The fire hydrant subsystem is designed in accordance with NFP A 24, Standard for th e Installation of Privat e Fir e S e rvic e Mains and Th e ir Appurt e nan c es , and the International Fire Code (IFC, 2012). The subsystem is designed to support fire flows of 5 , 680 Umin (1 , 500 gal/min) overall and at least 1,893 Umin (500 gal/min) at each fire hydrant. 9.3.2.2 Fire Detection and Alarm Subsystem The fire detection and alarm subsystem will provide a range of fire detection capabilities and notification methods. The primary means of detection will be by monitoring the fire suppression system devices, including flow switches that indicate release of water from automatic sprinklers or deluge valves, and tamper switches that supervise valve position.

Smoke and heat detection will be provided in specific locations to provide detection of fires in spaces where water damage concerns warrant improved manual intervention (e.g., computer server rooms), areas deserving additional life safety (e.g., egress locations), or other safety-driven functions.

As required by NFPA 101, Life Safety Code, and NFP A 72 , National Fir e Alarm Code , smoke detection will be provided above the main fire alarm control panel and any subpanels necessary to perform control functions for the system. For ventilation units, smoke and heat detection will be provided in support of several safety aspects. Smoke detectors will be provided in: * *

• Non-nuc l ear ventilation systems , in accordance with NFPA 90A, Standard for the Installation of Air-Conditioning and Ventilating S y stem s , and the International Fire Code (IFC , 2012) Air intakes , to address smoke infiltration from wild land fires and fires in other facilities that might spread smoke to the surrounding area Nuclear ventilation systems, to support shutdown and minimize the spread of contaminated smoke to other areas of the RPF Heat detectors will be provided in the Zone I and II ventilation system exhausts for both notification of high temperatures and release of the automatic portion of the HEP A filter plenum deluge capability. Control modules and relays will be integrated into the fire detection and a larm subsystem.

Control modules will provide signals for releasing the deluge valves for the HEP A filter plenum deluge capability, and control methods will be integrated for shutdown of non-safety HVAC systems. 9-20

. ; .. ... NWMI ...... .. .. . ..... .... .. .... . NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems * *

• NORTifWEST MEDfCAl ISOTOPES Alarms received by the fire alarm control panel will be transmitted via a copper cable or fiber optic cable network to monitoring stat ions in the RPF. The fire alarm control panel will also provide notification through the facility-wide infrastructure to the central alarm station. The central alarm station will provide data to the Columbia Fire Department for response. The fire detection and alarm subsystem will receive it s primary power supply from a dedicated circuit off of the normal building power. Internal batteries will provide a secondary power source, with connection to the standby generator.

The batteries will be sized to provide 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (hr) of backup power , plus 10 min of alarm power , as required b y NFPA standards.

9.3.2.3 Fire Extinguishers In addition to the automatic features of the fire suppression subsystem, manual response capabilities will be provided via fire extinguishers with an appropriate classification.

Fire extinguishers will be located throughout the building , as required by NFPA 10, Standard for Portable Fire Extingu.ishers.

Specific extinguisher types, such as those for metal fires or particular chemicals, will be spec ified depending on the hazard. 9.3.3 Operational Analysis and Safety Function Chapter 13.0 identifies fire hazards and evaluates adverse events and accident sequences.

The criticality safety evaluations discussed in Chapter 6.0 include the impact of fire suppression water in its analysis.

Chapter 13.0 provides an evaluation of the accident sequences that involve either combustible solids or liquids , or explosive gases , in close proximity to the hi g h uranium process streams or the high-dose process streams. As part of this analysis, an emergency purge gas system was identified to prevent flammable concentration in process vessel headspaces.

IROFS FS-03, " Process Vessel Emergency Purge System,

is discussed in Chapter 13.0 , Section 13.2. 7 , and in Chapter 6.0. The following summarizes NWMI-2013-039, Pr e liminary Fire Ha za rds Analysis (PFHA), which was prepared to demonstrate that the RPF will maintain the ability to perform safe-shutdown functions and minimize radioactive material releases to the environment in the event of a fire. The PFHA objectives were to: * *

• Consider potential in s itu and transient fire ha zar ds Determine the effects of a fire in any location in the RPF and the ability to safely shut down the facility and/or minimize and control the relea se of radioactivity to the environment Specify measures for fire prevention , detection , suppression , and containment for each fire area housing structures, systems, and components that are important to safety, in accordance with U.S. Nuclear Regulatory Commission (NRC) guidelines and regulations The PFHA assessed the fire hazards at the RPF , support facilities, and surrounding project site. The analysis also assessed the fire safety criteria identified in NRC Regulatory Guide 1.189 , Fir e Protection for Nuclea r Power Plant s. The PFHA provided a consequence evaluation of a de s ign basis fire (DBF) scenario within each fire area, assuming the loss of automatic and manual fire suppression.

The PFHA also identified fac ilit y design features and fire hazard mitigating features for personnel safety and property protection commensurate with the NRC criteria.

9-21

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

• NORTHWEST MEOIW ISOTOPES 9.3.3.1 Radioisotope Production Facility Fire Areas NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The fire hazards , life safet y considerations , fire protection features , and DBF for designated fire areas within the RPF are discussed below. The RPF will be subdivided into separate fire areas for the purposes of limiting the spread of fire , protecting personnel , and limiting the consequential damage to the facility.

Figure 9-7 and Figure 9-8 provide the delineation of fire areas on the first floor and second floor of the RPF, respectively. The determination of fire area boundaries was based on consideration of the following: * * * *

• Types , quantities, density , and location of combustible materials Location and configuration of equipment Consequences of inoperable equipment Location of fire detection and suppression systems Personnel safety and exit requirement s Fire areas will typically be bounded by 2-hr fire-rated barriers to separate:

• * * * * *

• Proce s sing areas and radioactive material storage areas from each other and adjacent areas Rooms with major concentrations of electrical and mechanical equipment from adjacent areas Computer and control rooms from adjacent area s Maintenance shops from adjacent areas Combustible storage areas from adjacent area s Fan rooms and plenum chambers from adjacent areas Office areas from moderate and high fire hazard areas In one case, two fire areas will be separated by 3-hr fire-rated barrier walls. The fire-rated barrier design and construction are in accordance with the International Building Code (IBC) (ICC, 2012) and NFPA 221 , Standard for High Challeng e Fir e Wall s , Fir e Walls , and Fire Barri e r Walls. Where fire-rated assemblies are partiall y or full y penetrated by pipes , ducts , conduits , raceways , or other devices , fire-rated barrier material will be placed in and around the penetrations to maintain the resistance rating of the assembl y. All openings in the fire barriers will be protected , consistent with the designated fire-resistance rating of the barrier. Fire doors will be rated commensurate with the fire-rated barrier in which they are installed , and comply with the requirements ofNFPA 80 , Standard for Fire Doors and Oth e r Openin g Prot e ctiv e s. 9-22 I NWMI *:::**:::.: ..*... . . ***** . . * .*. * . NOltTitWEST MEDICAL lSOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-7. Life Safety Plan (First Floor) 9-23

... ; ... ; ... NWMI -********** ..... * .. *.*:* . "NORTHWESTMEDICALISOTOPES

[Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-8. Life Safety Plan (Second Floor) 9.3.3.1.1 Hot Cell, Waste Handling, and Shipping Areas As the most consequential fire area within the RPF, the hot cell area will be a single-story, combustible, high bay structure.

The footprint of this area will be [Proprietary Information].

The hot cell area will include parts of the irradiated target receipt bay and waste management areas, Mo recovery and purification process , U recovery and recycle process , high bay above the hot cell area, and operating and maintenance galleries.

An overhead crane system will be used to transfer radioactive materials between the different operations.

Life Safety Considerations The hot cell area is anticipated to handle hazardous materials that exceed the maximum allowable quantity limits established in the IBC (ICC, 2012). Therefore , the hot cell area will be designed as High Hazard H-3 and/or H-4 occupancy in accordance with the IBC and will be provided with emergency lighting , illuminated exit signs, automatic sprinklers, and an automatic and manuall y actuated fire alarm system with audible and visual indicating devices as necessary. The common path of egress travel for an H-3 occupancy equipped throughout with an automatic sprinkler system will be 7.6 m (25 ft), in accordance with the IBC Section 1014.3. The exit access travel distance for a fully sprinklered H-3 occupancy will be limited to 45.7 m (150 ft), in accordance with the IBC. Dead-ends in corridors should not exceed 6.1 m (20 ft), in accordance with IBC Section 1018.4. 9-24

---

. ;:.;* .. NWMI ...... .. .. . .......... * * . NORTKWEST MEDtCAl I SOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Access to the crane platforms will be limited to maintenance and service personnel only. IBC Section 505.3 defines equipment platforms as not being habitable and are considered to not be occupiable space. Because the crane platforms will normally be unoccupied with limited access, these crane platforms will not be required to meet IBC means of egress requirements.

Exposure Fire Potential/Potential for Fire Spread between Fire Areas The hot cell, waste, handling and shipping areas will be separated from other fire areas of the building by 2-hr fire-rated barrier walls , with the exception of the wall between the production area and the administrative area, which will have a 3-hr fire-rated barrier wall. Penetrations in the fire-rated barrier walls will be protected with penetration seals, providing a fire rating equivalent to the barriers.

The hot cell area could be exposed to a fire in an adjacent fire area when the lar ge access doors are opened during radiological material transfer activities.

The primary areas of concern include the interface (open doors) between the unloading and waste truck bays with the production area. To prevent a fire from spreading between these areas, administrative controls will be implemented that dictate personnel procedures and limit combustibles around interface access doors. Fire spread between areas will be therefore mitigated b y personnel actions , limited combustibles, and fire-rated boundaries.

Fire Protection Features The hot cell area requires the following fire protection features to provide a defense-in-depth approach to fue protection.

This approach will result in a fire being quickly detected and suppressed, which will mitigate fire-induced damage. * *

• Automatic

-Automatic sprinkler systems will be installed throughout the production area, with the exception of the hot cell enclosure.

Self-contained fire suppression systems may be located on equipment such as cranes and forklifts.

An automatic fire detection and alarm system will be installed throughout the production area. Analysis of the need for sprinklers in the hot cell area and additional detailed information on these syste ms will be developed for the Operating License Application.

Manual -Manual fire suppression will consist of portable fire extinguishers and Class I standpipe system hose valves that will be provided within the production area. Manual fire alarm pull stations will be provided at exits from the production area. Passive -Passive fire protection will be provided in the form of fire-rated construction to protect the means of egress from the facility and separation between fire areas. Fuel traps will be provided where the diesel-powered vehicles interface with the production area. Underhung collection pans will be provided under the crane gearboxes. Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios The following fire hazards and ignition sources were considered for evaluation of a DBF scenario within the production area. *

• Scenario 1 -A fire starts within the irradiated target shipping cask that is caused by agitation and spontaneous ignition of the pyrophoric uranium dust or particulate.

Scenario 2 -A fire or explosion starts within a tank or exhaust system that is caused by the uncontrolled accumulation of hydrogen gas. Hydrogen generation represents a fire hazard, where the accident sequence is initiated by failure of the sweep gas subsystem.

9-25

.:;.-.;* .. NWMI .*;.**.*.* . .............. ' ." . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems * * *

• Scenario 3 -A fire starts within the exhaust stack system that is caused b y the ignition of the carbon retention bed and/or HEPA filters. Scenario 4 -A fire starts adjacent to a semi-tractor trailer that is caused by the rupture of the fuel tank and ignition of the unconfined (static) diesel spill. Scenario 5 -A fire s tarts on a diesel-driven forklift that is caused by the rupture of the fuel tank and ignition of the unconfined (static) diesel spill. Scenario 6 -A fire starts in a crane collection pan that is caused by the rupture of the gearbox and ignition of the confined (static) silicone oil pool. The DBF scenario for the production area consists of a diesel fuel spill and ignition from an unknown so urce caused by the operation of a semi-tractor trailer or forklift.

The semi-tractor trailer is assumed to have two 284 L (75-gal) diesel fuel tanks (568 L [150 gal total]), along with rubber tires, a batter y, and small amounts of other combustible material.

A small amount of permanent combustibles, including electrical cables , pol ye thylene tarps, isopropyl alcohol, v in y l , and trash bins, ma y also be present. These combustibles will be limited by administrative controls.

The DBF scenario postulates that the entire contents of the fuel tank s will spill, forming an approximately 15.5 m (50-ft) diameter pool with a 3 millimeter (mm) (0.12-in.) depth , and will then ignite. The DBF postulates that any combustibles located within the fuel spill diamet e r will also ignite and be completely consumed (NWMI-2013-039).

Consequences of an Automatic Fire Suppression Failure Failure of the automatic fire suppression sys tem will cause a delay in responding to a fire , resulting in the combustibles being completely consumed during the DBF. The adoption of administrative controls will limit combustibles and minimize the spread of fire. However , smoke and hot gases could damage equipment located within the production area. The Columbia Fire Department will be notified of a fire by either actuation of a manual fire alarm pull bo x station or the automatic smoke or temperature detection systems. The DBF would be contained within the irradiated target receipt bay and operating gallery by the 2-hr rated fire walls. If the automatic fire suppression system fails to operate, the fire department is expected to arrive well before the 2-hr fire walls have failed and extinguish the fire using portable extinguishers or the hose stream supported by the Class I standpipe system. The required response time of the fire department will be determined for the Operating License Application. Conclusion While the DBF for this area is unlikel y to result in a radiological release with the radioactive material being contained in a U.S. Department of Transportation (DOT) Type B cask, the potential exists for a release in some of the other scenarios described. Additional information , and a determination if the fire protection systems in this fire area will be considered IROFS, will be provided in the Operating License Application.

9.3.3.1.2 Target Fabrication Area The target fabrication area will be located adjacent to the production area on the east side of the RPF and will be a noncombustible structure with an industrial F-1 occupancy.

Two-hour fire-rated barrier walls will separate the target fabrication area from other fire areas of the building.

Penetrations in the fire-rated barrier walls will be protected with penetration seals, providing a fire rating equivalent to the barriers.

9-26

.; .. ;*. NWMI ...... .. .. . ........... * * ." . NOflTHWEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The footprint of the target fabrication area will be [Proprietary Information]

over most of the area. This area will be dedicated to the production of low-enriched uranium (LEU) targets. Life Safety Considerations The target fabrication area is required to meet IBC life safety criteria (ICC, 2012) and will be provided with emergency lighting, illuminated exit signs , automatic sprinklers, and an automatic and manually actuated fire alarm system with audible and visual indicating devices as necessary. An accessible means of egress will be provided in accordance with the IBC. Exit access will be provided to the target fabrication area , with direct exit discharge from the RPF. The maximum distances to the exit access in the target fabrication area will be within the following parameters for a High Hazard H-3 occupancy.

The common path of egress travel for an H-3 occupancy equipped throughout with an automatic sprinkler system will be 30.5 m (100 ft), in accordance with IBC Section 1014.3. The exit access travel distance for a fully sprinklered F-1 occupancy will be limited to 76.2 m (250 ft), in accordance with IBC Table 1016.2. Dead-ends in corridors will not exceed 15.2 m (50 ft), in accordance with IBC Section 1018.4 , Exception

2. No deviations from the IBC life safety criteria have been identified.

Exposure Fire Potential/Potential for Fire Spread between Fire Areas The target fabrication area could be exposed to a fire in an adjacent area when the large access doors are opened during target transfer or waste shipping activities.

The primary area of concern is an open doorway to the production area. To prevent a fire from spreading between these areas, administrative controls will be implemented that dictate personnel procedures and limit combustibles around interface access doors. Additional information on these controls will be provided in the Operating License Application.

Fire spread between areas will therefore be mitigated by personnel actions , limited combustibles , and 2-hr fire-rated boundaries.

Fire Protection Features The target fabrication area requires the following fire protection features to provide a defense-in-depth approach to fire protection. This approach will result in a fire being quickly detected and suppressed , reducing fire-induced damage. * *

• Automatic

-An automatic fire suppression system will be installed throughout the target fabrication area. An automatic fire detection and alarm system will be also installed throughout the target fabrication area. The system specifics will be determined during detailed design and included in the Operating License Application.

Manual -Manual fire suppression will be provided within the target fabrication area and consist of portable fire extinguishers and Class I standpipe system hose valves. Manual fire alarm pull stations will be provided at exits from the target fabrication area. Passive -Passive fire protection will be provided in the form of fire-rated construction to protect the means of egress from the facility and separation between fire areas. 9-27

.. ; .. ;* .. NWMI ...... .. .... *.-.* .. *:.* *. * ." . NORTl f WtsT M£DICAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios The following fire hazards and ignition sources were considered for evaluation of a DBF scenario within the target fabrication area. * *

• Scenario 1 -A fire or explosion starts within the reduction subsystem, caused by ignition of a nitrogen or hydrogen gas mixture by the high temperature created by the oven (determined to be highly unlikely based on credible physical conditions

[Chapter 13.0)). Scenario 2 -A pyrophoric fire of uranium metal (determined to be highly unlikely based on credible physical conditions

[Chapter 13 .OJ). Scenario 3 -A fire starts with combustible materials or equipment in the target fabrication area . The DBF event was determined to be a fire of combustible materials such as paper products (Scenario 3). The DBF for the target fabrication area involves ignition of in situ combustibles located within the area caused by an electrical short circuit or a maintenance welding operation.

The combustible loading of the area was considered low. The fire also consumes other transient combustibles located within the area. Consequences of an Automatic Fire Suppression Failure Failure of the automatic fire suppression system will cause a delay in responding to a fire, resulting in the combustibles being completely consumed during the DBF. The adoption of administrative controls will limit combustibles and minimize the spread of fire. However , smoke and hot gases could damage equipment located within the target fabrication area. In the event of a fire, the Columbia Fire Department will be notified by either actuation of a manual fire alarm pull box station or the automatic smoke or temperature detection systems. The DBF would be contained within the target fabrication area by the 2-hr rated fire walls. If the automatic fire suppression system fails to operate, the fire department is expected to arrive well before the 2-hr fire walls have failed and extinguish the fire using portable extinguishers or the hose stream supported by the Class I standpipe system. The required response time of the fire department will be determined for the Operating License Application. Conclusion The above analysis and description show that the fire protection and life safety systems within the target fabrication area are designed such that they will function in a manner, whether operational or not , consistent with occupational safety and protection of the public and environment.

Two of the three scenarios described are considered highly unlikely.

The DBF for this fire area would result in minimal or no release to the public because of the low radiological source term and the fact that the standard combustibles described are unlikely to be mixed with the LEU materials.

Therefore, this system will likely not be considered an IROFS. 9.3.3.1.3 Administration and Support Area The administration and support area will be l ocated adjacent to the production area on the south side of the RPF and will be a single-story, noncombustible structure with business (Group B) and assembly (Group A-2) occupancies. The administration and support area will be Type IIB construction and separated from the remainder of the RPF by 3-hr fire-rated barrier walls. 9-28 NWM I ..*... * * ."

• NOPITHWtST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The administration and support area will include the main entry and security access points, break room, control room, conference room, men's and women's lavatories, and several small offices. The control room will be separated from the remainder of the administration and support area by 2-hr fire-rated barrier walls. The operations performed within the administration and support areas will be consistent with office space uses. The occupant load of the administration and support area will include non-production work staff. Life Safety Considerations The administration and support area is required to meet IBC life safety criteria (ICC , 2012) and will be provided with emergency lighting, illuminated exit signs , automatic sprinklers, and an automatic and manually actuated fire alarm system with audible and visual indicating devices as necessary.

An accessible means of egress will be provided to the area in accordance with the IBC. Exit access will be provided to the administration and support area by one main exit at the front of the building and a secondary exit located to the south side of the RPF. A break room will also be provided with an additional exit. The maximum distances to the exit access within the administration and support area will be within the following parameters.

The travel distance for the common path of egress travel for a mixed use business (B) and assembly (A-2) occupancy equipped throughout with an automatic sprinkler system will be 23 m (75 ft), in accordance with IBC Table 1014.3. The exit access travel distance for a fully sprinklered mixed-use business (B) and assembly (A-2) occupancy will be limited to 76 m (250 ft), in accordance with IBC Table I 016.2. Dead-ends in corridors will not exceed 6.1 m (20 ft), in accordance with IBC Section 1018.4. Exposure Fire Potential/Potential for Fire Spread between Fire Areas The administration and support area will be separated from other fire areas of the RPF by 3-hr fire-rated barriers.

Penetrations in the fire-rated barrier walls will be protected with penetration seals, providing a fire rating equivalent to the barriers. Load-bearing structural elements are not required to be protected by fire-resistive construction. To prevent a fire from spreading between areas, administrative controls will be implemented that dictate personnel procedures and limit combustibles around access doors. Fire spread between areas will be therefore mitigated by personnel actions , limited combustibles , and 3-hr fire-rated boundaries. Fire Protection Features The administration and support area requires the following fire protection features to provide a in-depth approach to fire protection.

This approach will result in a fire being quickly detected and suppressed , reducing fire-induced damage. * *

• Automatic

-An automatic wet-pipe sprinkler system will be installed throughout the administration and support area. An automatic fire detection and alarm system will also be installed throughout this area. Additional detailed information will be developed for the Operating License Application.

Manual -Manual fire suppression will be provided within the administration and support area and consist of portable fire extinguishers and Class I standpipe system hose valves. Manual fire alarm pull stations will be provided at the exits from the administration and support area. Passive -Passive fire protection will be provided in the form of fire-rated construction to protect the administration and support area from other occupied areas of the facility.

9-29

.. ; .. ; ... NWMI ...... .. .. . .......... ! * * .' . NORTHWEST MEDtCAl ISOTOPES Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The DBF was determined to consist of ordinary combu s tible s (e.g. paper product s and office furniture) ignited within a closed office caused by an electrical short circuit. The combustible loading of the office was considered low. The fire also consumes other tran s ient combustibles located within the office and s preads to nearby cubicles.

The DBF would result in the comp l ete combustion of the combustible materials in the area of origin. No credit was taken for fire suppression activities.

The administration and support area was considered a single fire area, and the result of the DBF was the complete loss of function of the area. Consequences of an Automatic Fire Suppression Failure Failure of the automatic fire suppression system will cause a delay in responding to a fire, resulting in the combustibles being completely consumed during the DBF. The adoption of administrative controls will limit combustibles and minimize the spread of fire. However , smoke and hot gases could damage equipment located within the administrative and support area. In the event of a fire , the Columbia Fire Department will be notified of a fire by either actuation of a manual fire alarm pull box sta tion or the automatic s moke or temperature detection systems. The DBF would be contained within the administrative and support area b y the 3-hr rated fire walls. If the automatic fire suppression system fails to operate , the fire department is expected to arrive well before the 3-hr fire walls have failed and extinguish the fire using portable extinguishers or the hose stream supported by the Class I standpipe system. The required response time of the fire department will be determined for the Operating License Application.

Conclusion The above analysis and de sc ription show that the fire protection and life safety systems within the administration and support area are designed such that they will function in a manner , whether operational or not , consistent with occupational safety and protection of the public and environment.

Because this fire area is not expected to contain anything other than check sources for in s trumentation , no releases to the public are expected to occur. Therefore , this system will lik ely not be considered an IROFS. Additional detailed information will be developed for the Operating License Application. 9.3.3.1.4 Irradiated Target Receipt and Waste Management Truck Bay Areas The irradiated target receipt and waste management truck bay areas will be located adjacent to the production area on the north s ide of the RPF and will be a noncombustible enclosure that is considered a sto rage S-2 occupancy area. The truck bay will be capable of accepting three semi-tractor trailers at the same time. Each truck ba y will be separated from the production area (cask unloading) by a 2-hr rated rollup door. The door s to the production area w ill be closed when the doors to the outside are open. This area will be used for the receipt of irradiated LEU targets and shipments involved with the disposal of radiological waste material.

Radiological material will be transported in appro ve d containers.

The casks will reside on the heavy-duty tractor-trailer for delivery and removal from the RPF. The heavy duty tractor-trailer will be present when the retractable doors are open to the production area. 9-30

.. ;*. NWMI ...... .. **.* ..... *

NORTIMEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Life Safety Considerations The irradiated target receipt and waste management truck bay areas will be required to meet IBC life safety criteria (ICC, 2012) and will be provided with emergency lighting, illuminated exit signs, automatic sprinklers, and an automatic and manually actuated fire alarm system with audible and visual indicating devices as nece ssa ry. An accessible means of egress will be provided in accordance with the IBC. Exit access will be provided to the truck bays. The maximum distances to the exit access within the truck bay will be established and conform to IBC code based on industrial occupancies.

The common path of egress travel for an S-1 occupancy equipped throughout with an automatic sprinkler system will be 30.5 m (100 ft), in accordance with IBC Table 1014.3. The exit access travel distance for a fully sprinklered S-1 occupancy will be limited to 76.2 m (250 ft), in accordance with IBC Table 1016.2. Dead-ends in corridors will not exceed 15.2 m (50 ft), in accordance with IBC Section 1018.4 , Exception

2. No deviations from the IBC life safety criteria have been identified. Exposure Fire Potential/Potential for Fire Spread between Fire Areas The irradiated target receipt and waste management truck bay areas will be separated from other fire areas in the building by 2-hr fire-rated barriers.

Penetrations in the fire-rated barrier walls will be protected with penetration seals, providing a fire rating equivalent to the barriers.

Load-bearing structural elements are not required to be protected by fire-r esistive construction, as indicated in the IBC (ICC, 2012). The truck bay could be exposed to a fire in an adjacent fire area when the large access doors are opened to attach or disconnect a trailer to or from a tractor. To prevent a fire from spreading between these areas , administrative controls will be implemented that dictate personnel procedures and limit combustibles around the interface access doors. Personnel actions, limited combustibles, and 2-hr fire-rated boundarie s will therefore mitigate fire spread between areas. Fire Protection Features The irradiated target receipt and waste management truck bay areas will require the following fire protection features to provide a defense-in-depth approach to fire protection.

This approach will result in a fire being quickly detected and suppressed, reducing fire-induced damage. * *

• Automatic

-An automatic sprinkler system will be installed throughout the truck bay area . However, due to the large quantity of diesel fuel and number of tires on the heavy-duty trailer , alternative suppression systems may be considered.

An automatic fire detection and alarm system will be installed throughout the truck ba y area. Additional detailed information will be developed for the Operating License Application. Manual -Manual fire suppression wi ll be provided within the truck bay area and consist of portable fire extinguishers and Class I standpipe system hose valves. Manual fire alarm pull stations will be provided within the truck drive-through.

Passive -Passive fire protection will be provided in the form of fire-rated construction to protect the means of egress from the facility and separation between fire areas. Built-in fuel traps and sloped floors will be provided to control potential fuel spills within the area. The fuel traps and sloped floors will also be used for containment of potentially contaminated firefighting water. The fuel and/or water will drain to outdoor underground collection tanks for testing and removal. 9-31

.:;.-.;* .. NWMI ..*...*.. *. .............. . *. NORTHWEST MEDICAL ISOTOPES Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The following fire hazards and ignition sources were considered for evaluation of a DBF scenario within the truck bay area. *

• Scenario 1 -A fire starts due to maintenance activities (e.g., spark ignition or open flame) . Scenario 2 -A fire is caused b y hot work (e.g., welding, flame , or plasma cutting) . Scenario 3 -A fire starts adjacent to a semi-tractor trailer that i s caused by the rupture of the fuel tank and ignition of the unconfined (static) diesel spill.

• Scenario 4 -A fire sta rts on a diesel-driven forklift that is caused b y the rupture of the fuel tank and ignition of the unconfined (static) diesel spill. The DBF for the truck bay consists of a diesel fuel spill a nd ignition from an unknown source caused by the operation of a diesel-powered semi-tractor trailer (Scenarios 3 and 4). The truck is assumed to have two 284 L (75-gal) diesel fuel tanks, along with 32 hard rubber tires , a battery , and small amounts of other combustible material.

The truck may also carry some combustibles on noncombu sti ble pallets when supporting radiological material-handling operations.

Administrative controls will be used to limit temporary combustible item s within the production area. The DBF scenario postulates that the entire contents of the fuel tanks will spill and drain to the built-in fuel trap. The area of the fire will be limited to the area of the built-in fuel trap trench , which was estimated to be approximately 2.6 m 2 (28 ft 2). The results of the DBF were postulated as the complete combustion of the combustible materials in the irradiated target receipt truck bay area. No credit was taken for fire suppression activities.

The DBF fire could result in the complete loss of function for the systems and/or equipment in the area. Consequences of an Automatic Fire Suppression Failure Failure of the automatic fire s uppression system will cause a delay in responding to a fire, resulting in the combustibles being completely consumed during the DBF. The adoption of administrative controls will limit combustibles and minimize the spread of fire. However, smoke and hot gases could damage equipment located within the truck bay area. In the event of a fire, the Columbia Fire Department will be notified of a fire by either actuation of a manual fire alarm pull box station or the automatic smoke or temperature detection systems. The DBF would be contained within the truck bay area by the 2-hr rated fire walls. If the automatic fire suppression system fails to operate , the fire department is expected to arrive well before the 2-hr fire walls have failed and extinguish the fire using portable extinguishers or the hose stream s upported b y the Class I standpipe system. The required response time of the fire department will be determined for the Operating License Application. Conclusion The above analysis and description show that the fire protection and life safety systems within the truck bay are designed such that they will function in a manner , whether operational or not , consistent with occupationa l safety and protection of the public and environment.

Because the radioactive material will be contained in DOT Type B casks, a fire in this area should not result in a radiological release to the public. Therefore, this system will likely not be considered an IROFS. Additional detailed information will be developed for the Operating License Application. 9-32

.. NWMI *********** ......... ::-. NOR1ltWESTMEDICAllS O T O PES 9.3.3.1.5 Laboratory Area NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The laboratory area will be located adjacent to the production area on the west side of the RPF and will be a single-story, noncombustible structure with a High Hazard H-3 and H-4 occupancy.

The footprint of the laboratory area will be approximately

[Proprietary Information]

over most of the area. This area will process and analyze quality and process control samples during production of the molybdenum-99 (9 9 Mo) product , fabrication of targets for irradiation , and processing of waste for disposal.

Typical RPF analysis will include: * * * *

• An inductively coupled plasma mass spectrometry (ICP-MS) to analyze mass quantities of isotopic [Proprietary Information]

A kinetic phosphorescence analyzer for [Proprietary Information]

Alpha spectroscopy for [Proprietary Information]

Beta activity by liquid scintillation spectrometry for strontium-89

/strontium-90 (89 Sr!9°Sr) Gamma energy analysis A variety of gloveboxes and fume hoods will be located within the analytical laboratory area. Life Safety Considerations The laboratory area is required to meet IBC life safety criteria (ICC, 2012) and will be provided with emergency lighting , illuminated exit signs, automatic sprinklers, and an automatic and manually actuated fire alarm system with audible and visual indicating devices as necessary.

An accessible means of egress will be provided in accordance with the IBC. Exit access will be provided to the laboratory area, with direct exit discharge from the RPF. The maximum distances to the exit access within the laboratory area will be within the following parameters. The common path of egress travel for a mixed High Hazard H-3 occupancy equipped throughout with an automatic sprinkler system will be 7 .6 m (25 ft), in accordance with IBC Table 1014.3. The exit access travel distance for a fully sprinklered mixed H-3 occupancy will be limited to 45.7 m (150 ft), in accordance with IBC Table 1016.2. ends in corridors will not exceed 6.1 m (20 ft), in accordance with IBC Section 1018.4. No deviations from the IBC life safety criteria have been identified.

Exposure Fire Potential/Potential for Fire Spread between Fire Areas The laborator y area will be se parated from other fire areas of the building by 2-hr fire-rated barriers.

Penetrations in the fire-rated barrier walls will be protected with penetration seals, providing a fire rating equivalent to the barriers. The laboratory area could be exposed to a fire in an adjacent fire area when the large access doors are opened during material transfer activities.

The primary area of concern in this case is an open doorway to the production area. To prevent a fire from spreading between these areas , administrative controls will be implemented that dictate personnel procedures and limit combustibles around the interface access doors. Fire spread between areas will therefore be mitigated by personnel actions, limited combustibles, and 2-hr fire-rated boundaries.

9-33

... ; .. ;* .. NWMI *::.**.*.*.* ............ . * .' . NORTHWEST MEDICAL ISOTOPES Fire Protection Features NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The laboratory area requires the following fire protection features to provide a defense-in-depth approach to fire protection.

This approach will result in a fire being quickly detected and suppressed, reducing induced damage. * *

• Automatic

-An automatic fire suppression system will be designed and installed throughout the laboratory area. An automatic fire detection and alarm system will also be installed throughout the laboratory area. The system specifics will be determined during detailed design and provided in the Operating License Application.

Manual -Manual fire suppression will be provided within the laboratory area and consist of portable fire extinguishers and Class I standpipe system hose valves. Manual fire alarm pull stations will be provided at the exits from the laboratory area. Passive -Passive fire protection will be provided in the form of fire-rated construction to protect the means of egress from the facility and separation between fire areas. Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios The DBF scenario for the laboratory area will be developed for the Operating License Application. Consequences of an Automatic Fire Suppression Failure The consequences of the failure of the automatic fire suppression system in the laboratory area will be determined for the Operating License Application.

Conclusion More analysis is needed to determine if the fire protection system in this area would be considered an IROFS. Additional detailed information will be developed for the Operating License Application.

9.3.3.1.6 Utility Areas Utility areas (e.g., electrical rooms, mechanical rooms , fire riser rooms, etc.) will be noncombustible spaces separated from other fire areas by fire-rated barrier walls. The footprint of each utility room will vary, but will be classified as utility (Group U) occupancies in accordance with the IBC (ICC, 2012). These utility areas will include rooms that house electrical equipment (e.g., power and lighting panels , transformers, and associated operations equipment distribution systems) and other common industrial equipment (e.g., air handling units , boilers , fans, pumps, and associated piping distribution systems).

Personnel will not normally occupy the utility areas. Life Safety Considerations The utility areas are required to meet IBC life safety and means of egress criteria (ICC, 2012) and will be provided with emergency lighting, illuminated exit signs, automatic sprinklers, and an automatic and manually actuated fire alarm system with audible and visual indicating devices as necessary.

An accessible means of egress will be provided in accordance with the IBC. The maximum distances to the exit access within the utility areas will be within the following parameters for utility occupancies.

The common path of egress travel for a utility occupancy equipped throughout with an automatic sprinkler system will be 22.9 m (75 ft), in accordance with IBC Table 1014.3. The exit access travel distance for a fully sprinklered utility occupancy will be limited to 121.9 m (400 ft), in accordance with IBC Table 1016.1. Dead-ends in corridors will not exceed 15.2 m (50 ft), in accordance with IBC Section 1018.4 , Exception

2. 9-34

.*; .. ;* NWMI *:.**.*.*. ..... .. .. .. .... . NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems * * . NOlmfW(ST MElMCAI..

ISOTOPf.S Exposure Fire Potential/Potential for Fire Spread between Fire Areas For the purpose of this analysis, the utility areas are each considered separate areas , each with 2-hr rated barrier walls used to limit the spread of fire. HEPA filters and exhaust carbon beds will be encased by stainless steel housings that can be isolated from the inlet and outlet ductwork by isolation dampers. Fire detectors will also be provided in each HEPA filter housing and inlet ductwork.

Therefore, isolation dampers will prevent the fire from propagating from the filter housing to other fire areas. Fire Protection Features The utility areas require the following fire protection features to provide a defense-in-depth approach to fire protection.

This approach results in a fire being quickly detected and suppressed, reducing induced damage. * *

• Automatic

-An automatic wet-pipe sprinkler or other approved fire suppression system will be installed throughout each utility area. An automatic fire detection and alarm system will also be installed throughout each utility area. Additional detailed information will be developed for the Operating License Application.

Manual -Manual fire suppression will be provided within each utility area that consists of portable fire extinguishers.

Passive -Passive fire protection will be provided in the form of fire-rated construction to protect separation between fire areas. Isolation dampers will be provided in the inlet and outlet of each HEP A filter housing to prevent fire from spreading to other fire areas. Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios The following were considered DBF scenarios for the utility areas. * * *

• Scenar io 1 -A fire starts due to maintenance activities, ignited from a spark or open flame . Scenario 2 -A fire starts from overheated electrical systems and equipment.

Scenario 3 -A fire starts in or near a transformer . Scenario 4 -A natural gas leak occurs . The DBF for the utility area consists of a natural gas leak resulting in an explosive mixture of natural gas and a detonation or deflagration.

Additional information for this accident sequence will be provided in the Operating License Application.

Consequences of an Automatic Fire Suppression Failure The consequences of a failure of the automatic fire suppression system in the utility area will be determined for the Operating License Application.

Conclusion More analysis is needed to determine if the fire protection system in this area should be considered an IROFS. Additional detailed information will be developed for the Operating License Application.

9-35

... ;.-.;* .. NWMI .*;.**.*.*. ............. . *. . NORTHWEST MEDICAl. I SOTOPES 9.3.3.2 Other Radioisotope Production Facility Systems 9.3.3.2.1 Facility Ventilation and Smoke Management NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The RPF ventilation system requirements must satisfy the process, building, safety , and regulatory requirements unique to the 99 Mo production process. To assist in the confinement of airborne radioactive contamination, the RPF ventilation system is designed to create pressure gradients and cause air to flow from areas of lesser contamination potential to areas of greater contamination potential.

Confinement zone exhaust ductwork will have fire dampers consistent with NFPA 45 , Standard on Fire Prot e ction for Laboratories Using Chemicals , and will be constructed to maintain fire ratings where ducting penetrates fire-rated barriers, as appropriate.

The confinement ventilation systems will also include HEP A and efficiency gas adsorption (HEGA) filtration systems located in a dedicated mechanical area. The Zone I ventilation system will comprise the secondary confinement boundary and be classified as an IROFS (RS-03). Chapter 13.0 provides additional information on the accident analysis that identified this IROFS. A combination of passive and active smoke management strategies will be used to minimize the spread of smoke , maintain tenable conditions for the evacuation of building occupants, and limit the damage caused by smoke. These strategies will be designed in accordance with NFP A 92 , Standard for Smoke Control S ys tems. The smoke control methods for each fire area will be developed for the Operating License Application.

9.3.3.2.2 HEPA Filtration Systems The HEPA filters and housings are a component of the hot cell secondary confinement boundary that will be classified as an IROFS (RS-0 3). The HEPA filters are expected to contain low levels of radiological material and will be located in designated fire areas. The filter housings are expected to be large, with a maximum size being approximately

[Proprietary Information]

in face area. The large filter face area will require automatic and manual sprinklers in the plenum housings and contaminated water collection or retention.

In addition, the HEP A filter housings will be located within 2-hr fire-rated barrier walls that are protected b y automatic sprinkler systems. 9.3.3.2.3 Crane Superstructure The structural steel supporting the facility overhead crane has been classified as an IROFS (FS-02, " Overhead Cranes"). Therefore , the crane superstructure must remain standing during and after a fire event to prevent damage to irradiated material.

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

9.3.3.2.4 Security and Safeguard Components Security systems are discussed in Chapter 12.0 , " Conduct of Operations

." 9.3.3.3 Architectural Features The codes and standards applicable to the RPF are defined in Chapter 3.0. The objectives of the NRC fire protection program will primarily be achieved through comp lianc e with prescriptive criteria, as defined by the PFHA (NWMI-2013-039).

9-36

.; .. ;*.NWMI ...... .. .. . ........ *. NORTifWESTMEOfCALISOTOP£S Types of Construction NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems All structures within the RPF complex confines will be constructed of Type JIB , noncombustible material, as defined by IBC Chapter 6 (ICC, 2012). Additional detailed information will be developed for the Operating License Application.

9.3.4 Instrumentation and Control Requirements The fire protection system will report the status of the fire protection equipment to the central alarm s tation and the RPF control room, with sufficient information to identify the general location and progres s of a fire within the protected area boundarie s. Initiatin g device s for the fire detection and a larm subsystem, which will include monitoring devices for the fire suppression s ubs ys tem , will indicate the presence of a fire within the facility.

Once an initiatin g device activates, signals will be sent to the fire alarm control panel. The fire alarm control panel will transmit signals to the central alarm s tation and perform any ancillary functions, such as shutting down the ventilation equipment or actuating the deluge valves. As required b y NFP A 101 a nd NFP A 72, smoke detection will be pro v ided above the main fire alarm control panel and any s ubp a nels neces sary to perform co ntrol functions for the syste m. For ventilation units, smoke and heat detection will be pro v ided in support of severa l safety aspects. Smoke detectors will be provided in non-nu c lear ventilation sys tem s in accordance with NFPA 90A and the IFC. Smoke detector s will also be provided in air intakes to address smoke infiltration from wi ldland fires and fire s in other facilities that mi ght spread smoke to the surrounding area. Smoke detection will be provided in ventilation systems servicing potentially contaminated zones to s upport shutdown and minimize the spread of contaminated smoke to other areas of the RPF. Heat detectors will be provided in these ventilation system exhausts for both notification of high temperatures and rel ease of the automatic portion of the HEPA filter plenum d el uge s ub system. Co ntrol modules and relays will be integrated into the fire detection and alarm subsys tem to initi a te reactions required for safety. Control modules will provide s ignals for relea s in g of deluge valves on the HEPA filter plenum delu ge s ubs ystem. Co ntrol m e thod s will also be integrated for s hutdown of the HY AC sys tems. Shutdown of electrical equipment or co mputer s will also occur as deemed ne cessa r y b y the design effort. Alarms recei ve d by the fire alarm control panel will b e transmitted via a copper cable or fiber optic cable network to monitoring stations in the RPF. The fire a larm control panel will also provide notification through the sitewi de infrastructure to the central alarm sta tion. The alarm stations will provide data to th e Columbia Fire Departm e nt for re s pons e. System Monitoring The fire protection system will be monitored by the fire alarm control panel, which will transmit signals to the central alarm station v ia a digital alarm communicator transmitter and to the RPF control room. Command and control functions will be exclusively available at the fire alarm control panel. Loca li zed monitoring of the various fire pumps will occur at the respective pump contro ll ers. 9-37

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

• NOlffifWEST MEDICAL ISOTOPES Control Capability and Locations NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The fire detection and alarm subsystem will be controlled exclusively from the fire alarm control panel. Numerous devices in the fire suppression subsystem can be operated manually.

The fire pumps can be started manually via their respective controllers.

Valves and hydrants will be turned manually, and no air or electrically operated valves will be provided.

Deluge valves for the HEPA filter plenum water spray can be activated manually, in addition to the bypass valves that are integrated into the design. Automatic and Manual Actions The fire detection and alarm subsystem is intended to operate automatically.

Manual intervention will be required for some operations, such as shutdown of outside air intake fans or dampers , due to the need to avoid false activation or to maintain operational status in emergency conditions.

The fire suppression subsystem will be split between automatic and manual operations.

The sprinkler systems (including the pumps) and the demister section of the HEPA filter plenum deluge subsystem are designed to operate automatically.

The filter section of the plenum deluge subsystem and fire hydrants are designed for manual operation.

Certain portions , however, can be operated manually as necessary.

The demister section of the HEP A filter plenum deluge subsystem will have a manual bypass and a manual actuator as part of the deluge valve. Portable fire extinguishers will be manually operated.

Maintenance and testing activities on both systems will require manual interaction.

The maintenance and testing requirements included in NFPA 25 , Standard for the Inspection , Testing , and Maint e nance of Water-Based Fir e Protection Systems, and NFPA 72 require manual operation of valves, starting of pumps, testing of circuits with meters, and other functions that necessitate manual actions. Interlocks, Bypasses, and Permissives The fire protection system, as designed, will not be subject to external interlocks , bypasses , or permissives (i.e., those outside the system itself). There will be inherent interlocks , bypasses , and permissives within the various fire protection system equipment , which will be designed to the criteria and requirements discussed in Chapter 3.0. For example , the fire detection and alarm subsystem can be controlled via passwords and allow for bypassing certain functions; however, the passwords will be limited to testing technicians and are not availab l e to general building personnel.

Thus, there will be no ability for the system to be locally manipulated without proper authorization.

Additional information will be provided in Chapter 7.0 for the Operating License Application. 9.3.5 Required Technical Specifications The technical specifications associated with the fire protection systems , if applicable , will be discussed in Chapter 14.0 as part of the Operating License Application.

9-38

.. ; .. ; .. NWMI ...... .. .. . .... .. .. .. * * ." . NORTHWEST MEDICAL lSOTOPlS 9.4 COMMUNICATION SYSTEMS NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The RPF communication systems will relay information during normal and emergency conditions for general operations and emergencies within the RPF. These 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. 9.4.1 Design Basis The communications system design basis is to provide communications during normal and emergency conditions between vital areas of the RPF and the Administration Building.

This communications capability will include the ability of operators or other designated staff members to announce an emergency in all areas of the RPF and provide two-way communications between all operational areas and the control room. Design of the telecommunication system also compl i es with Electronic Industries Alliance and Telecommunications Industry Association requirements.

9.4.2 System Description The communication system is designed to provide two-way communication between the RPF control room and other site locations necessary for safe RPF operations.

This system will provide (1) communications capability between RPF operators , their supervisor , health physics personnel, and other personnel as required by the technical specification , and (2) the ability to make faci lit y-wide emergency announcements and summon emergency assistance.

The telephone and data/local area network (LAN) telecommunications system will include a service entrance communications room. The service provider's outside plant optical fiber will terminate on a wall-mounted service provider entrance patch panel. An optional outside plant copper telephone cable from the service provider will terminate at the wall-mounted overvoltage entrance protection terminal modules for use in l egacy non-Voice over Internet Protocol (VoIP)-based equipment.

The main entrance room will be connected with a telecommunications room with fiber and copper backbone cable. The telecommunications room will support the offices, laboratory area , target fabrication area, shipping and receiving areas , and other required telephone and data/LAN outlets. Grounding of the telecommunication system will comp l y with Telecommunications Industry Association and NFP A requirements.

The process control system will be physically separated from and not connected to the communication system. Additional information will be provided in the Operating License Application.

9.4.3 Operational Analysis and Safety Function Chapter 13.0 identifies and evaluates adverse events and accident sequences.

The accident analysis has not identified the need to credit the communication system. The communication system is designed such that it will function in a manner, whether operational or not , consistent with occupational safety and protection of the public and environment.

9-39

.:;.-.;* .. NWMI .*:.**.*.* . .............. . *. NORTHWEST MEDICAL ISOTOPES 9.4.4 Instrumentation and Control Requirements NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Chapter 7.0 discusses the instrumentation and control requirements associated with the communication systems. 9.4.5 Required Technical Specifications The technical specifications associated with the communication systems, if applicable, will be discussed in Chapter 14.0 as part of the Operating License Application.

9-40

.:;.-.;* .. NWMI ..*...... * . ......... :: . * * * ." . NORTHWEST MEDICAL ISOT O PES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems 9.5 POSSESSION AND USE OF BYPRODUCT, SOURCE, AND SPECIAL NUCLEAR MATERIAL The RPF is designed to ensure that:

• No uncontrolled release of radioactive materials (solid , liquid, or airborne) from the facilities can occur

• Per so nnel exposures to radiation , including ingestion or inhalation, do not exceed limiting values in 10 CFR 20 , as defined in Chapter 11.0 , and are consistent with the NWMI ALARA program. The operating procedure s developed for the Operating License Application will ensure that only radioactive byproducts handled by the RPF are permitted , unless specifically authorized by the 1 0 CFR 50 , " Domestic Licensing of Production and Utilization Facilities," license or an additional license. 9.5.1 Design Basis The design basis for the possession and use of special nuclear material (SNM) and byproduct material is to ensure that no uncontrolled release of radioactive materials (solid, liquid , or airborne) from the facilities can occur and that personnel exposures to radiation, including ingestion or inhalation , do not exceed limiting values in 10 CFR 20 and are consistent with the NWMI ALARA program. Additional information on the design basis is pro v ided in Chapter 3.0. 9.5.2 System Description SNM is defined by Title I of the Atomic Energy Act of 1954 (42 U.S.C. 2011 et seq.) as plutonium , uranium-233 (2 33 U), or uranium enriched in the isotopes 233 U or 235 U. The RPF will receive , store, and process fresh unirradiated uranium metal and irradiated uranium with an enrichment of 19.75 weight percent (wt%) +/-0.20 wt% 235 U (LEU). Byproduct material, as defined by the Atomic Energy Act, is radioactive material (except SNM) yielded in or made radioactive by exposure to radiation incident to the process of producing or using SNM. As defined by NRC regulations , byproduct material includes any radioactive material (except enriched uranium or plutonium) produced by a nuclear reactor. The RPF will handle byproduct material during the separation of 99 Mo and the recycling of the irradiated LEU. Source material is defined as the element thorium or the element uranium , provided that the uranium has not been enriched in the isotope 235 U. Source materials will not be present in the RPF. 9.5.2.1 Special Nuclear Materials SNM will be handled in two areas of the RPF: the target fabrication and irradiated material areas (i.e. hot cells). The target fabrication area SNM inventory is li s ted in Chapter 4.0, Table 4-1, and the irradiated material area SNM inventor y is provided in Chapter 4.0 , Table 4-2. Chapter 4.0 also provides a description of the design of spaces and equipment to ensure that there i s no uncontrolled release of radioactive materials (solid, liquid , or airborne) from the RPF and that personnel exposures to radiation, including ingestion or inhalation, do not exceed limiting values in 10 CFR 20 consistent with the RPF ALARA program, as described in Chapter 11.0. Associated procedures are defined in Chapter 12.0. The NWMI emergency preparedness and physical security plans are provided in Chapter 12.0 , Appendix A and B , respectively.

Fire protection provisions are described in Section 9.3.2.1. 9-41

.:;:.;* .. NWMI ..**..... * . ........... :. *. NORTlfWESTMEDICAllSOTOPES 9.5.2.2 Byproduct Materials NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Byproduct materials handled in the RPF include 99 Mo and radioactive waste materials. A description of the Mo recovery process design is provided in Chapter 4.0 , Section 4.3.5. A description of the waste processing design is provided in Chapter 11.0, Section 11.2. A detailed inventory of byproduct materials within each of the main systems within the RPF is provided in the following chapters:

• * * * *

• Target fabrication

-Chapter 4.0 , Section 4.4.2 Target receipt and disassembly

-Chapter 4.0 , Sections 4.3.2 and 4.3.3 Target dissolution

-Chapter 4.0, Section 4.3.4 Molybdenum recovery and purification

-Chapter 4.0, Section 4.3.5 Uranium recovery and recycle -Chapter 4 0 , Section 4.4.1 Waste handling-Chapter 11.0 , Section 11.2 Chapter 4.0 and Section 9.7.2 provide descriptions of the design of spaces and equipment to ensure that there is no uncontrolled release ofradioactive materials (solid, liquid , or airborne) from the RPF and that personnel exposures to radiation, including ingestion or inhalation, do not exceed limiting values in 10 CFR 20 consistent with the NWMI ALARA program (Chapter 11.0). Associated procedures will be defined in Chapter 12.0 , as part of the Operating License Application.

9.5.3 Operational Analysis and Safety Function The criticality safety of SNM is discussed in Chapters 4.0 and 6.0 , and the material control and accounting of SNM is discussed in Chapter 12.0 , Section 12.13. The byproduct materials associated with the RPF process are addressed in Chapter 4.0, and byproduct materials within the waste processing and storage areas are described in Section 9.7.2 and Chapter 11.0 , Section 11.2. 9.5.4 Instrumentation and Control Requirements Instrumentation and control requirements for the processes associated with the possession and use of byproduct materials and SNM are discussed in Chapter 7.0 and Chapter 12.0 , Section 12.13. 9.5.5 Required Technical Specifications The technical specifications associated with the possession and use of byproduct materials and SNM, if applicable, will be di s cussed in Chapter 14.0 as part of the Operating License Application.

9-42

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

. e * .' . NORTHWEST MEDtCAL ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems 9.6 COVER GAS CONTROL IN CLOSED PRIMARY COOLANT SYSTEMS As discussed in Section 9.7.1.2.2, the RPF provides coo lin g water to numerous process tanks. The radiolytic decomposition of water within this system co uld result in the production of h ydrogen (H 2) and oxygen mi xtures. This sectio n provides a discussion of the cover gas co ntrol system within the process coo lant system. 9.6.1 Design Basis Information on the d esign basis of cover gas control in the closed primary coolant system (process ch ill ed water system) is provided in C h a pt er 3.0, Section 3.5.2.7. 9.6.2 System Description The process chilled water system i s described in Section 9.7.1.2.2.

The accumulation of combustible gases within this system wi ll be controlled b y the "sweep" gas system that is described in Sect ion 9. 7.1.2.6. Gases entrai ned in the chilled water syste m will be released in the cooling water collection tanks. Hydrogen, which is the primary component of evolved combustible gases , diffuses very rapidly and will be diluted by the airflow provided by the sweep gas flow. The plant air supp ly system (descr ibed in Section 9. 7.1.2.4) will provide low-flo w [Proprietary Information]

purge gases to Tanks TK-420 and TK-320. The process vessel vent system will co llect the purge gas from each of the tanks and merge the collected vent subsystems into the main facility ve ntilation syste m for treatment an d filtration.

These systems will work together to prevent exp lo sive gas mi x tures from developin g. 9.6.3 Operational Analysis and Safety Function Chapter 13.0 eva luate s the acci dent sequences that involve eit h er combustible so lid s or liquids , or explosive gases, in close proximity to the high uranium process streams or the high-dose process streams. This analysis d etermined that if the purge air system was not operational , a h ydrogen-air concentration in se l ecte d tanks co uld rise above 25 percent of the low er exp losiv e limit, a nd an ignition source could cause a detlagration or detonation , resulting in the r e le ase of radionuclides into the air. The tanks associate d wit h th e cooling syste m are not anticipated to require IROFS controls.

9.6.4 Instrumentation and Control Requirements Instrumentation and control requirements for the cover gas control in the c lo sed primary coolant system are discu sse d in C hapt er 7.0. 9.6.5 Required Technical Specifications The technical specifications associated wit h the cover gas control in the closed primary coolant system , if applica ble , will be discus sed in Chapter 14.0 as part of the Operating License Application.

9-43

.. ; ... ; .. NWMI ...... .. ... *.*.* .. *:.* e * * . NORTHWEST MEDtcAl I SOTOPES 9.7 OTHER AUXILIARY SYSTEMS NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Other RPF auxiliary systems that are important to the safety of workers , the public , and environment will include the following:

• * *

• Process utilities Control and storage of radioactive waste (waste management)

Analytical laboratory Chemical supply The followings subsections describe these auxiliary systems , including their design basis , system description, operational analysis and safety function , instrumentation and control requirements , and technical specifications.

9.7.1 Utility Systems The utility systems will provide heating , cooling , process water, compressed gases , instrument, motive force, and other functions to support uranium processing , waste handling , and ventilation.

The utility sy stems will include the following subs y stems: * * *

• Process steam Proce s s chilled water Demineralized water Plant and instrument air *

• Gas supply , which supplies nitrogen , helium , hydrogen , and oxygen Purge/sweep gas The utility systems are designed to ensure that any potential malfunctions do not cause accidents in the RPF or an uncontrolled release of radioactivity.

The systems are designed to ensure that in the event radioactive material is released by the operation of one of these systems , potential radiation exposures would not exceed the limits of 10 CPR 20 and are consistent with the NWMI ALARA program. No function or malfunction of the auxiliary systems will interfere with or prevent safe shutdown of the RPF. 9.7.1.1 Design Basis The utility systems design ba s is is provided in Chapter 3.0 , Section 3.5.2.7. 9.7.1.2 System Description Figure 9-9 shows the second floor mechanical utility area where the process steam , chilled process water, demineralized water , and plant or in s trument supply air units will be housed. Helium , hydrogen , and oxygen will be provided b y bottled gases located near the point of use either in the laboratory area or the target fabrication area. Nitro g en will be provided by a tube trailer for nitrogen located out s ide of the laboratory area. 9-44 I . ... NWMI ...... .. .. . ........... * ." , NOllTifWHT MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-9. Second Floor Mechanical Utility Area 9.7.1.2.1 Process Steam The process steam system will be divided into a medium-pressure central heating loop (Figure 9-10) and a low-pressure secondary loop within the hot cell (Figure 9-11 ). Medium pressure steam will be generated b y a natural gas-fired boiler (ST-H-100).

Low-pressure steam in the secondary loop will be generated by medium-pressure steam in a shell-and-tube heat exchanger (ST-E-200).

Medium-pressure steam will be at least 4.2 kilograms (kg)/square centimeter (cm 2) (60 pounds [lb]/square inch [in.2]) gauge , to provide an adequate temperature differential to generate 1. 7 k g/cm 2 (25 lb/in. 2) gauge steam for the low-pressure steam loop. 9-45 I ... .. NWMI .*:.**.*.*. ......... !:* NORTHWESTMfDfCAllSOTOPES

[Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-10. Medium-Pressure Steam System 9-46 I .. NWMI .*;.**.*.* . .............. . * .'

• NORTHWEST MEDICAL ISOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-11. Low-Pressure Steam System 9-47

.;.-.;*.NWMI *;.**.*.*. ........... .', NORTHWEST ME.DICAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Low-pressure steam will be generated in a vertical shell-and-tube heat exchanger.

Automatic blowdown and makeup water streams will limit the content of sludge or dissolved solids in the boiler and steam generation heat exchanger.

9.7.1.2.2 Chilled Water Process Chilled Water The process chilled water system is a central process chilled water loop that will cool the three secondary loops:

• One large geometry secondary loop in the hot cell (Figure 9-12)

• One criticality-safe geometry secondary loop in the hot cell (Figure 9-13)

• One criticality-safe geometry secondary loop in the target fabrication area (Figure 9-14) . The central process chilled water loop will rely on three air-cooled chillers, each sized to accommodate 50 percent of the process cooling demands (Figure 9-15). The secondary loops will be cooled by the central chilled water system through plate-and-frame heat exchangers.

Several proce ss demands will require cooling at less than the freezing point of water. These demands will be met with water-cooled refrigerant chiller units , cooled by the secondary chilled water loops. The chilled water system will operate with cascading pressure differentials. The central system will operate at the highest pressure , and the secondary loops will operate at a pressure between the central syste m and the process fluid. The large-geometry secondary loop in the hot cell will meet the cooling demands where fissile material leaking through a heat exchanger is not a credible event. The other cooling loops will be inherently criticality-safe by geometry, so active controls will not be required to keep fissile material out of the chilled water return. At each process cooling demand where fissile material may be present , conductivity sensors will monitor the chilled water return to detect heat exc hanger leaks. Facility Chilled and Heating Water The HVAC system will maintain the occupied space at 24°C (75°F) (summer) and 22°C (72°F) (winter), with active ventilation to support workers and equipment.

The facility chilled water and heating water sys tems will provide heating and cooling media to the HVAC system. The facility chilled water system (FCW) will supply the HVAC system with cooling water that is circulated through the chilled water coils in the air-handling units. The air will be drawn across the coils and cooled to be delivered to the RPF production area to maintain temperature.

The FCW will provide cooling water at a temperature of 9°C ( 48°F) to the HV AC air-handling unit cooling coils. There will be three equal-sized facility chillers located adjacent to the RPF: two in operation and one spare. The heating water system (H W) will supply the HV AC system with heating water that is circulated through the heating water coils in the air handling units. The air will be drawn across the coils and cooled to be delivered to the RPF production area to maintain temperature.

The HW will provide heating water at a temperature of 82°C (180°F) to the HV AC air-handling unit heating coils and reheat coils. The heating water will be generated as a byproduct stream of the steam boilers. 9-48

... ; ... ; ... NWMI ...... .. .... ............ e * . NOR'TlfWEST MEDK:Al ISOTOPES This page intentionally left blank. 9-49 NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems I .. .. NWMI .*;.**.*.** .............. . *. N O llTlfWESTMEOJCAl.

I SOTOPES [Proprietar y In fo rmation] NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-12. Chilled Water System Large Geometry Hot Cell Loop 9-50 I .. ... NWMI ...... ... .... *.*.* .. *.*. NOmrwESTMEDICAllSOTOPES

[Proprietar y Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-13. Chilled Water System Critically Safe Hot Cell Loop 9-51 I NWMI ............ ........... * * * . NOlmtWEST MEDfCAl ISOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-14. Chilled Water System Target Fabrication Loop 9-52 I

.*.* .. *.*.* NOmrwEnMEDtCAll$0TOPlS

[Proprietary Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-15. Process Chilled Water System 9-53

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

• NORTHWEST MfDICAl ISOTOPES 9.7.1.2.3 Demineralized Water NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Demineralized water will be produced by removing mineral ions from municipal water through an ion exchange (IX) process (Figure 9-16). Water passes through anion and cation exchange media located in separate IX tanks (DX-IX-100 and 110), and the demineralized water will accumulate in a storage tank (DW-TK-120).

A feed pump will provide the water at 4.2 kg/cm 2 (60 lb/in.2) gauge (DW-P-1 25) for RPF process activities.

The IX media will be regenerable using a strong acid and a strong base (DW-P-105 and 115). Acid and base will be fed from local chemical drums by toe pumps. 9.7.1.2.4 Plant and Instrument Air Plant air will be provided for seve ral activities (e.g., tool operation, pump power , purge gas in tanks , valve actuation, and bubbler tank level measurement) (Figure 9-17). Small, advective flows of plant air will be used throughout the RPF to prevent accumulation of combustible gases to hazardous concentrations.

Combustible gases will be evolved from process liquid s due to exposure of these liquids to ionizing radiation.

The plant air system will provide air to the instrument air subsystem.

The instrument air subsystem will use plant air that is filtered and dried (IA-V-11 OA, 11 OB , and IA-F-110).

Plant air w ill be generated b y a compressor (PA-K-100) and cooled to near-ambient temperatures by an aftercooler (PA-E-100).

The lead/lag configuration can s uppl y reduced flow after a si ngle compressor failure. The plant air receiver wi ll provide buffer capacity to make up the difference between peak demand and compressor capacity (PA-V-100).

Instrument air will be dried in regenerable desiccant beds to a dew point of no greater than -40°C (-40°F) and filtered to a maximum 40 micron (µ) particle size. The instrument air receiver will provide buffer capacity (IA-V-120) to make up the difference between peak demand and compressor capacity.

9. 7.1.2.5 Gas Supply Gas supply of helium (Figure 9-18), hydrogen , and oxygen (Figure 9-19) will be supplied b y standard gas bottles. Nitrogen will be provided from a tube truck (Figure 9-18). The nominal capacity of the gas bottles will be 8,495 L (300 ft3). The nitrogen will be fed from the tube truck (GS-Z-100) to the chemical su pply room where manifold piping will be used to distribute the gas. The primar y use of nitrogen will be in the reducing furnaces during target fabrication.

Helium, hydrogen, and oxygen gas bottles will be loc ated near the points of use. Gas supply pressures will be regulated to 1. 7 k g/c m 2 (25 lb/in.2) gauge at the bottle (Figure 9-19. Where lower pressures are required, point-of-use gas regulators will be installed.

Automatic gas cylinder changeover valves will provide a continuous gas supply when one bottle (or rack of bottles) is empty, and alert the operator when bottles need to be replaced.

H y drogen and oxygen gas bottles will be stored in ventilated gas cabinets with 13 air changes/min to mitigate the risk of leaks. The ventilation demand will be 8.8 Umin (250 ft3/min) air for each gas cabinet. 9-54

.:; .. ;* .. NWMI ...... .. .... ......... *.* *****. * * . NOllTlfWEST MEDtCAl. ISOTOPES T h is p age in tentiona ll y l eft bl ank. 9-55 NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems I ... NWMI .*:.**.*.*. .............. *. * ! 0

• NOfmfWtST MEOICAl lSOTOPES [Proprietary Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-16. Demineralized Water System 9-56 I ... NWMI .*:.**.*.* . ..... * .. *::* . *. ... .* . NORTHWEST MEDICAL ISOTOPES [Proprietar y Information]

Figure 9-17. Plant Air System NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems 9-57 I . NWMI .*;.**.*.*. ............ e * ." . NOllTifWEST MEDfCAl. ISOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-18. Nitrogen and Helium Supply System 9-58 I . .. NWMI .*;.**.*.*. ........ !. * * ." . NORTlfWEST MEOtCAl ISOTOPES [Proprietary Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-19. Hydrogen and Oxygen Supply System 9-59

............ ... ; ... ; ... NWMI ............ . *. ." . NORTHWEST MEDICAL ISOTOPES 9. 7.1.2.6 Purge Gas NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The plant air and nitrogen supply systems, described in Section 9.7.1.2.4 and Section 9.7.1.2.5 , provide purge gases to the required tanks. Depending on the tank, the purge gas will be provided through the bubbler tank level measurement device or other means. The purge gas flow rates are specified as either high flow for conditions of a large tank or high radioactivity , or low flow where the tank is small and radioactivity is low. Table 9-3 provides the purge gas flows for both the high and low flow rates. Table 9-3. Purge Gas Flows -Mi@j,MWM@!.i Flow rate Units (basis) High purge [Propri e tar y [Propri e t a r y [Proprietary Information]

In fo rmation] In fo rm a tion] Low purge [Proprietary

[Proprietary

[Proprietary Information]

information]

Information]

a NWMI-2013-CALC-005 , Tank Air Bl ee d Estimat e , Rev. B, Northwe s t Medical I s otopes , LLC, Corvallis , Oregon , 2014. b NWMI-2013-CALC-009 , U ranium Purifi c ation Sy s t e m Equipm e nt Si z in g, Rev. B , Northwest Medical I s otopes , LL C, C or v allis , Ore g on , 2014. = h y dro ge n ga s. u = u ra nium. The process vessel vent system will collect the purge gas from each of the vessels and treat it before discharge to the Zone I exhaust. The process vessel vent system merges the collected vent subsystems into the main facility ventilation system for treatment and filtration. These systems will work together to prevent explosive gas mixtures from developing in the headspace of the process v es s els. The tanks anticipated to require purge gas are listed in Table 9-4. Additional information on the purge gas system will be developed for the Operating License Application.

Table 9-4. Tanks Requiring Purge Gas Tank number Tank name Tank number Tank name DS-D-100 Dissolver l UR-TK-120A Impure uranium collection tank 2A DS-D-200 Dissolver 2 UR-TK-120B Impure uranium collection tank 2B DS-TK-800 Waste collection and sampling tank 1 UR-TK-140A Impure uranium collection tank 3A DS-TK-820 Waste collection and sampling tank 2 UR-TK-140B Impure uranium collection tank 3B MR-TK-100 Feed tank lA UR-TK-160A Impure uranium collection tank 4A MR-TK-140 Feed tank lB UR-TK-160B Impure uranium collection tank 4B MR-TK-180 U solution collection tank UR-TK-200 IX feed tank l MR-TK-200 Feed tank 2 UR-TK-900 IX waste collection tank l MR-TK-340 Waste collection tank UR-TK-920 IX waste collection tank 2 UR-TK-lOOA Impure uranium collection tank lA WH-TK-100 High-dose waste collection tank UR-TK-1008 Impure uranium collection tank lB WH-TK-240 High-dose concentrate collection tank IX = ion e xchan g e. u uranium. 9-60

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

. NOflllfWEST MEDICAL ISOTOP£S 9.7.1.3 Operational Analysis and Safety Function NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Chapter 13.0 eva luate s the accident sequences that involve fissile solution or solid materials being introduced into systems not normall y designed to process these solutio ns or so lid materials.

The accident analysis associated with utilities a ddr esses fissile solutio n leaks across a mechanical boundar y between process vesse l s or backflows into a utility system. Defense-in-depth

-The tank and vessel wa ll s will be made of corrosion-resistant materials and ha ve wall thicknesses that are rated for lon g service with harsh acidic or basic chemica l s. Items relied on for safety-Based on the analysis conducted in Chapter 13.0 , Section 13.2, the fo llowin g IROFS are implemented.

• * * *

• IROFS CS-I 0 , "C losed Safe Geometry Heatin g/C ooling Loop w ith Monitoring and Alarm" IROFS CS-20, "Eva porator/Concentrator Condensate Monitoring" IROFS CS-27 , "C losed Heatin g/C ooling Loop with Monitoring and A l arm" IROFS FS-03 , " Process Vessel Emergency Purge System" IROFS CS-18 , " Backflow Prevention Device" The analyses that identified these IROFS and the associated system descriptions are addressed in Chapter 13.0 and Chapter 6.0 , respectively. 9.7.1.4 Instrumentation and Control Requirements Instrumentation and control requirements for the proce sses associated wit h the utilit y system are discussed in Chapter 7.0. 9.7.1.5 Required Technical Specifications The technical spec ification s associated with the utilit y system, if applicable , will be discu ssed in Chapter 14.0 as part of the Operating License Application.

9-61

.. ;*. NWMI ...... .. .. . .......... * * .' , NOIOltWEST MEDtCAl. ISOT0'1S 9.7.2 Control and Storage of Radioactive Waste NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The radioactive waste control and storage systems are designed to ensure that (I) any potential malfunctions do not cause accidents in the RPF or uncontrolled release of radioactivity , and (2) in the event radioactive material is released by the operation of one of these systems, potential radiation exposures would not exceed the limits of 10 CFR 20 and remain consistent with the NWMI ALARA program. No function or m a lfunction of the auxiliary syste m s will interfere with or prevent safe s hutdown of the RPF. 9.7.2.1 Design Basis The waste handling system design ba s is is provided in Chapter 3.5.2.7. 9.7.2.2 System Description To fulfill the de s ign basis , the control and s torage of radioactive waste will include the following functions

• * * * * * *

• High-dose liquid waste handling (collection, concentration, and solidification)

Low-dose liquid waste handling (co ll ectio n , evaporation, r ecycle and so lidification)

Spent resin dewaterin g Solid waste encapsulation High-dose waste d ecay High-dose waste handlin g Waste handling Waste Staging and Shipping Buildin g (Class A storage) These function s are described in detail in the following s ubs ec tion s. Figure 9-21 summa ri zes the wee kl y de sign basis volumes and the average annual week l y volumes of all waste handlin g proce ss streams. The de sign basi s vo lum e is based on eight University of Missouri R esea rch Reactor (MURR) targets and 30 Oregon State University (OSU) TRI GA 1 R eac tor (OSTR) targets per week to provide appropriately s i ze d tanks. The annual weekly average is based on processin g eight MURR targets per week for 44 weeks per yea r and 30 OSTR targets per week for eight weeks per yea r and is used in the sizing of the high-dose decay s tora ge. 1 TRI GA (Trai nin g, R esearch, I soto p es , Genera l Atomics) is a registered trademark of General Ato mi cs , San Diego, Ca li forn ia. 9-62

.:;.-.; ... NWMI .*;.**.*.*. .*.* .. *.*:-. * * * ." . NORTHWEST MEDtCAl I SOTOPES This page intentionally left blank. 9-63 NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems I NWMI ...... .. .... ......... *.* * * * .' . NORTHWEST MEDtCAl lSOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-20. Waste Management Process Flow Diagram and Process Flow Streams 9-64

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

• NORTHWEST MEOtcAl ISOTOPES 9.7.2.2.1 High-Dose Liquid Waste Handling NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-21 shows the location in the hot cell area where the high-dose liquid waste will be processed.

High-dose liquid waste will be collected in th e high-dose waste collection tank (s hown in Figure 9-22), which will provide the ne e ded handling capacity to match the volume of liquid waste generated by the upstream processes.

Chapter 4.0 provides description s of the high-dose liquid st r ea ms that will be directed to the collection tank. [Propriet ary Information]

Figure 9-21. High-Dose Liquid Waste Solidification Subsystem and Low-Dose Collection Tank Location The proce ss strea m volumes are summarized in Figure 9-20, and Table 9-5 pro vi des the high-dose waste tank capacities.

The process s treams include: * * *

• Caustic sc rubb e r waste Oxidizing column waste NO x absorber waste Reg enera tion waste from Ti0 2 #1 IX * *

• 9-65 Raffinat e/rinsate from #2 IX Raffinat e/rinsate from #3 IX U IX waste I ..** .. NWMI ...... .. .. .... .. .. .. 0 * * .'

• NORTHWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-22. Simplified High-Dose Waste Handling Process Flow Diagram Tank ID WH-TK-100 WH-TK-240 Table 9-5. High-Dose Waste Tank Capacities Description/purpose High-dose waste accumulation tank High-dose concentrate accumulation tank 9-66 Tank capacity 5,050 1,270 19,000 4,800

... ; ... ;-.. NWMI .*:.**.*.* . .............. e * ." . NOllTlfWEST MEDICAL I SOTO P ES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Additions to the collection tank are in discrete, analyzed batches. Sodium hydroxide solution will be added as needed to neutralize any excess acidity. The neutralized liquid will be forwarded to the dose waste concentrator, where water is evaporated from the high-dose liquid , condensed, and directed to the condensate collection tank. The evaporator bottoms will be directed to a high-dose concentrate collection tank. Figure 9-23 shows the arrangement of the high-dose waste handling equipment.

A HIC will be transferred into the high-dose waste treatment hot cell through the HIC transfer drawer, and docked with the high-dose solidification mixer. Solidification agent will be transferred to the designated bin from th e distribution hopper, which will be loaded by operators in the low-dose waste solidification area. dose liquid waste concentrate from the waste concentrate collection tank and solidification agent will be metered into the HIC by the high-dose solidification mixer that may consist of an in-line mixer or a sacrificial paddle within the HIC. After filling and mixing are complete, the high-dose solidification mixer will be disengaged, and the HIC lidded and prepared for transfer to the high-dose waste decay subsystem for storage. [Proprietary Information]

Figure 9-23. High-Dose Waste Treatment and Handling Equipment Arrangement 9.7.2.2.2 Low-Dose Liquid Waste Handling Figure 9-24 shows the location of the low-dose liquid waste collection tank. Low-dose condensate from the high-dose concentrator will be held in the condensate collection tank (Figure 9-25). Chapter 4.0 provides descriptions of the low-dose liquid streams that will be directed to the collection tank. The process stream volumes are summarized in Figure 9-20 , and Table 9-6 provides the low-dose waste tank capacities. Low-dose liquid received from other upstream processes , combined with the low-dose condensate not recycled, will be transferred to the low-dose waste collection tank where the contents of the tank will be ana l yzed and adjusted with sodium hydroxide (NaOH) to neutralize any residual acids. Once neutralized, the low-dose waste will then be forwarded to the first of two evaporation tanks located on the second floor (Figure 9-24). In these heated tanks , the liquid will be held at elevated temperatures (60°C [140°F]), and high rates of ventilation air will be passed through the tank. The heated tank contents, plus the high rate of ventilation, will evaporate excess water , reducing the volume of solid waste generated. Samples will be collected and analyzed to ensure compliance with waste acceptance criteria.

9-67 I .. NWMI ..**.*... *. .............. . *. .". NORTHWE S T MEDICAllSOTOPES

[Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-24. Low-Dose Liquid Waste Evaporation System Location 9-68 I *.*... NWMI ...... .. .. . ..... .. .. .. * * ." . NOllTHWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-25. Low-Dose Liquid Waste Disposition Process Table 9-6. Low-Dose Waste Tank Capacities Tank capacity Tank ID Description/purpose WT-TK-400 Condensate tank for high-dose evaporator 4 , 300 16,250 WH-TK-420 Low-dose waste accumulation tank 5,900 22,300 WH-TK-500 Low-dose waste evaporation tank (LO-I) 5 , 900 22 , 300 WH-TK-530 Low-dose evaporation tank (LD-2) 2 , 600 9,800 9-69

... ;.-.;* .. NWMI .*;.**.*.*. : ........... . , 0 .'

• NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The partially concentrated low-dose liquid waste will be transferred to the low-dose waste solidification area (Figure 9-26), where the waste will be metered into a drum that has been placed in the low-dose solidification hood (WH-EN-600).

Solidification product vendor information indicates that a ratio of 56. 7 to 79.4 kg (125 to 175 lb) of solidification agent is sufficient to solidify 59 to 178 L ( 42 to 4 7 gal) of liquid waste within a 55-gal drum. The drum will be lidded at the drum lidding station. With time, the mixture will solidify within the waste drum. The filled waste drum will be loaded onto a shipping pallet and transferred by pallet jack to the shipping and receiving airlock door. [Proprietary Information]

Figure 9-26. Low-Dose Liquid Waste Solidification Equipment Arrangement 9.7.2.2.3 Spent Resin Dewatering Spent resin dewatering will be conducted in the high-dose waste treatment hot cell. Figure 9-27 provides the flow diagram for the spent resin dewatering subsystem.

This subsystem will transfer uranium recovery and recycle system spent IX resin slurry from the spent resin collection tanks located in the tank hot cell (Figure 9-28) to the dewatering filling head in the high-dose waste treatment hot cell (Figure 9-23). The dewatering filler head will remove liquid from the resin. Dry resin will be collected in a waste drum , and the liquid returned to the low-dose waste collection tank. The solid waste drum transfer drawer (WH-TP-810) (Figure 9-23) will be opened , and the high-dose waste handling crane will be used to lift the drum and place it in the solid waste drum scan for characterization.

After characterization is complete , the drum will be transferred by the high-dose waste handling crane from the solid waste drum scan feed conveyor and placed into a five-drum rack. As determined by characterization, the drum will either be held for decay storage or transferred to the dose waste handling system for transfer to a shipping cask. 9-70

.;.-.;*. NWMI ...... .. .. . ........ *.* NORTHWESTMEOICAl.ISOTOPES

[Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-27. Spent Resin Dewatering Operational Flow Diagram [Proprietary Information]

Figure 9-28. Spent Resin Collection Tanks Location 9-71

...... .. NWMI .*;.**.*.* . ......... !:* . NORTHWESTMEDICAllSOTOPES 9.7.2.2.4 Solid Waste Encapsulation NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-29 provides the flow diagram for the solid waste encapsulation subsystem.

Operators will enter the maintenance gallery and retrieve the solid waste drum cart from the waste collection port and transfer the drum cart into the high-dose waste treatment hot cell (Figure 9-23). The solid waste drum access port will be opened, and the solid waste encapsulation grout mixer (WH-Z-800) filling nozzle will be docked for waste encapsulation.

After the grout filling is complete, the solid waste encapsulation grout mixer filling nozzle will be removed , and the solid waste drum access port closed. The solid waste drum transfer drawer will be opened, and the high-dose waste handling crane will be used to lift the drum and place it in the solid waste drum scan for characterization.

After characterization is complete, the drum will be transferred by the high-dose waste handling crane from the solid waste drum scan feed conveyor and placed into a five-drum rack. As determined by the drum's characterization, the drum will either be held for decay storage or transferred to the high-dose waste handling subsystem for transfer to a shipping cask. [Proprietary Information]

Figure 9-29. Solid Waste Encapsulation Operational Flow Diagram 9.7.2.2.5 High-Dose Waste Decay Figure 9-30 provides the flow diagram for the high-dose waste decay subsystem.

This subsystem will provide Jag storage capability for solidified liquid waste and the five-drum racks with high-dose source terms. After HICs or five-drum racks have been filled and lidded in the high-dose waste treatment hot cell , they will be transferred to the high-dose waste decay subsystem. [Proprietary Information]

Figure 9-30. High-Dose Waste Decay Operational Flow Diagram 9-72

. .. NWMI .... ** ..... .... .. .. .. e * ." . NOmtWEST MEOtcAl I SOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The high-dose waste decay cell lift (WH-L-900) (Figure 9-31) will lower the HIC or five-drum rack into the high-dose waste decay cell , where the high-dose waste decay cell conveyor (WH-CN-900) will transfer the HIC or five-drum rack to its decay storage position.

The HIC or five-drum rack will remain in storage for a set amount time to allow for short-lived radioisotopes in the waste to decay to lower levels. When the HIC or five-drum rack has decayed to an acceptable activity level , the high-dose waste decay cell conveyor (WH-CN-900) will transfer the HIC or five-drum rack to the high-dose waste deca y cell lift , where the HIC or rack will be raised into the high-dose waste treatment hot cell and then transferred to the high-dose waste handling area. [Proprietary Information]

Figure 9-31. High Dose Waste Decay Cell Equipment Arrangement 9.7.2.2.6 High-Dose Waste Handling Figure 9-32 provides the flow diagram for the high-dose waste handling subsystem.

This subsystem will provide the capability to remotely transfer high-dose waste containers into a shipping cask. When a HIC or two five-drum racks are ready for shipment , the high-dose waste handling crane will be used to open the high-dose waste shipping transfer port (WH-TP-1000) and then transfer the HIC or two five-drum racks , from [Proprietary Information]

Figure 9-32. High Dose Waste Handling Operational Flow Diagram within the high-dose wa s te handling area , through the high-dose waste shipping transfer port, and into a s hipping cask. 9-73

.:;:.;* .. NWMI .*:.**.*.* . .............. *. .". NORTHWESTMED I CAl I SOTOPES 9.7.2.2.7 Waste Handling NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The simplified operational flow diagram for the waste handling subsystem is shown in Figure 9-33. [Proprietary Information]

Figure 9-33. Waste Handling Flow Diagram The waste handling subsystem will have multiple material handling capabilities.

The liquid high-dose radiological waste and solid radiological waste handling will begin with the arrival of a truck and lowboy trailer transporting an empty DOT-appro ve d cask (Figure 9-34). The truck, trailer , and shipping cask will enter the RPF to the waste management loading bay via an exterior facility high-bay door. The shipping cask will then be documented for material tracking and accountability per the safeguards and security system requirements.

Operators will use the utility system's truck bay spray wand for any necessary wash-down of the truck, trailer, or shipping cask while located in the waste management loading bay. The operators will remove the shipping cask's upper impact limiter using the waste shipping overhead crane (WH-L-1100) (Figure 9-34). The upper impact limiter will be placed in the designated impact limiter landing zone and secured. Operators will unbolt the lid and prepare the DOT-approved s hipping cask for loading per the cask loading and unloading procedure.

At this point , the truck, trailer , and shipping cask will enter the waste loading area via a high-bay door. The trailer containing the approved shipping cask will be positioned below the high-dose waste s hipping transfer port (WH-TP-1000) of the contaminated waste system. The truck will be disconnected from the trailer and exit the RPF via the high-ba y doors in which the vehicle entered. All high-bay doors will be verified as closed and the shipping cask will then be in position and ready for loading per the contaminated waste system procedures.

9-74

.*; .. ; ... NWMI ...... .. .. . .......... * * ! * . NORTHWEST MEDICA.l ISOTOPE.S

[Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-34. Waste Handling Equipment Arrangement After the DOT-approved cask has been loaded, the shipping cask will be separated from the high-dose waste shipping transfer port (WH-TP-1000).

The truck will enter the RPF into the waste management loading bay via an exterior facility high-bay door , and operators will use the utility system's truck bay overhead spray wand for any necessary wash-down of the truck while located in the waste management loading bay. The truck will then enter the waste loading area via a high-bay door. The truck will be connected to the trailer and exit to the waste loading area in the waste management loading bay. At this point, the facility process control and commun i cations system will allow operators to replace the shipping cask's upper impact limiter using the waste shipping overhead crane (WH-L-1100). The shipping cask will be documented for material tracking and accountability per the safeguards and security system requirements (Chapter 12.0). The truck , trailer , and shipping cask will exit the RPF through the high-bay doors in which the vehicle entered. The liquid low-dose radiological waste handling process will begin with the arrival of a truck transporting the empty waste drum pallets to the fresh and unirradiated shipping and receiving area. The receiving area door will be opened, and the truck will be docked to the receiving bay, allowing for transfer of the waste drum pallets into the RPF. Pallet-loaded empty waste drums will be unloaded from the truck using the waste handling pallet jack (WH-PH-1100). All unloaded empty waste drum pallets will then be documented for material tracking and accountability per the safeguards and security system requirements. The pallet jack carrying an empty waste drum pallet will be transferred to the shipping and receiving airlock door , where the empt y waste drums will enter the contaminated waste system for loading. After the waste drums have been loaded with liquid low-dose radiological waste and re-palletized, a pallet containing full waste drum s will be transferred via the waste handling pallet jack (WH-PH-1100) from the s hipping and receiving airlock door to the waste loading area. The waste handling forklift (WH-PH-1110) will then enter the waste management loading bay via an exterior facility high-bay door. 9-75

.; .. ;.NWMI ...... .. .. . ........... .... . NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems * * . N O mfWEST MEDICAL ISOTOPES A waste shipping truck will also enter the waste management loading bay via an exterior facility high-bay door. Operators will open the high-bay door to the waste loading area and use the forklift to load the waste drum pallet into the truck. The shipping truck will then be documented for material tracking and accountability per the safeguards and security system requirements. The truck containing the waste pallets will exit the RPF through the high-bay doors in which the vehicle entered. 9. 7.2.2.8 Waste Staging and Shipping Building (Class A Storage) The Waste Staging and Shipping Building will be approximately

[Proprietary Information]

and will provide additional waste storage and shipping preparation for Class A radioactive waste prior to disposal.

9.7.2.3 Operational Analysis and Safety Function Chapter 13.0, Section 13.2 evaluates the accident sequences that involve fissile solution or solid materials being introduced into systems not normally designed to process these solutions or solid materials. The waste handling system is not geometrically safe; therefore , a number oflROFS have been identified. * * * * * * * * * * * * * * * * *

• IROFS RS-01 , " Hot Cell Liquid Confinement Boundar y" IROFS RS-03 , " Hot Cell Secondary Confinement Boundary" IROFS RS-04 , "Hot Cell Shielding Boundary" IROFS RS-08 , "Sample and Analysis of Low Dose Waste Tank Dose Rate Prior to Transfer Outside the Hot Cell Shielding Boundary" IROFS RS-10 , "Active Radiation Monitoring and Isolation of Low Dose Waste Transfer" IROFS CS-14 , " Active Discharge Monitoring and Isolation" IROFS CS-15, "Independent Active Discharge Monitoring and Isolation" IROFS CS-16, " Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal" IROFS CS-17 , " Independent Sampling and Analysis of U Concentration Prior to Discharge or Disposal" IROFS CS-18, " Backflow Prevention Device" IROFS CS-21, " Visual Inspection of Accessible Surfaces for Foreign Debris" IROFS CS-22 , " Gram Estimator Survey of Accessible Surfaces for Gamma Activit y" IROFS CS-23 , " Non-Destructive Assay (NDA) of Items with Inaccessible Surfaces" IROFS CS-24 , "Independent NDA of Items with Inaccessible Surfaces" IROFS CS-25 , " Target Housing Weighing Prior to Disposal" IROFS CS-26 , " Active Discharge Monitoring and Isolation" IROFS FS-01 , "Enhanced Lift Procedure" IROFS FS-02 , " Overhead Cranes" Additional information on the analyses that identified these IROFS is provided in Chapter 13.0. 9. 7.2.4 Instrumentation and Control Requirements Instrumentation and control requirements for the processes associated with the control and storage of radioactive waste are discus s ed in Chapter 7.0. 9.7.2.5 Required Technical Specifications The technical specifications associated with the control and storage of radioactive waste , if applicable, will be discussed in Chapter 14.0 as part of the Operating License Application.

9-76

.; .. ;* .. NWMI ...... .. .. . .......... * * . NOlmfWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems 9. 7.3 Analytical Laboratory The analytical laboratory will support production of the 99 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. 9.7.3.1 Design Basis The RPF analytical laboratory design basis is to provide on-site analysis to support the production of 99 Mo product and fabrication of targets for irradiation. This analysis will be used to determine (I) mas s, concentration and purity of SNM , (2) concentration of 99 Mo product and product impurities , (3) process stream chemical and radionuclide concentrations, and ( 4) chemical and radionuclide analysis for waste handling and disposition.

Analysis will be required to: * * * *

• 9.7.3.2 Verify acceptable 99 Mo product to ship Confirm uranium content Determine adjustments for feed tanks and other associated adjustments Verify that recycled uranium product complies with product specification Ensure compliance with waste acceptance criteria System Description The RPF analytical laboratory space will include the following:

• * * * * *

• Hoods to complete sample preparation , waste handling , and standards preparation Hoods for specialty instruments , including an ICP-MS and kinetic phosphorescence analyzer Glovebox for ICP-MS Gloveboxes for sample delivery and preparation prior to sample transfer to hoods Countertops for the gamma spectroscopy sys tem , low-energy photon s pectroscopy , alpha spectroscopy system , liquid scintillation system , and beta-counting system Storage for chemical and laboratory supplies Benchtop systems , s uch as balances, pH meter s, and ion-chromatography The analytical laboratory layout is pre se nted in Figure 9-35 and provides space for eight hoods , four gloveboxes, and two countertops.

9-77 I .;-..;:.NWMI ...... .. .... ........... * * * . NORTHWlST ME.DH:Al.

ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-35. Analytical Laboratory Layout Analytical instrumentation will include the ICP-MS , kinetic phosphorescence anal yz er , gamma energ y analysis, alpha spectroscopy , liquid scintillation spectrometry , and gamma energy anal y sis. 9.7.3.3 Operational Analysis and Safety Function Chapter 13.0 evaluates the accident sequences that involve miscellaneous chemical safety process upsets in areas without significant fissile or high-dose licensed material present (chemical storage areas and the laboratory).

The accidents analyzed that are associated with the analytical laboratory include Accident Sequence S.R.31, "Chemical Burns from Contaminated Solutions During Sample Analysis." No laboratory IROFS have been identified.

Defense-in-depth

-Operators and laboratory technicians will follow set protocols on sampling and analysis to identify the sampling locations , sampling techniques , containers to be used , transport routes to take , analysis procedures , reagents to use , equipment requirements, and disposal protocol for the sample residue material.

Each of these procedures will be evaluated for standard safety protocols, including requirements in the safety datasheets for the chemicals used and safety requirements for the equipment used. 9-78

.:;.-.;* .. NWMI .*;.**.*.* . .......... *:. . * .. NORTHWESTMEDICAJ..

I SOTOP E S 9.7.3.4 Instrumentation and Control Requirements NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Instrumentation and control requirements for the processes associated with the analytical laboratory will be discussed in Chapter 7.0 as part of the Operating License Application.

9. 7.3.5 Required Technical Specifications The technical specifications associated with the analytical laboratory, if applicable, will be discussed in Chapter 14.0 as part of the Operating License Application.

9. 7.4 Chemical Supply The chemical supply system will include tanks supplying aqueous chemicals to the process systems, flammable material storage cabinets used to segregate incompatible materials , and storage of chemical solids used in the process systems. 9.7.4.1 Design Basis The chemical supply system design basis is to provide chemical solutions mixed to the required concentrations that are used within the target fabrication, target dis so lution, Mo recovery and purification , and waste management systems. The system will provide nitric acid, NaOH , reductant and NO x absorber solutions, hydrogen pero xi de , and fresh uranium IX resin. Additional information is provided in Chapter 3.0 , Section 3.5.2.7. 9. 7.4.2 System Description Figure 9-36 shows the la yo ut of the chemical supply room within the RPF. Tanks are sized to provide s upport to the process requirements.

9. 7.4.2.1 Subsystem 100, Nitric Acid Subsystem 100 will consist of five tanks that provide the following functions:

• [Proprietary Information]

• [Proprietary Information]

• [Proprietary Inform at ion] * [Proprietary Information]

• [Proprietary Information]

• [Proprietary Inform atio n] 9-79 I .. NWMI ...... .. *... *.*.* .. *:.* NORTHWESTMEDfCAllSOTDPE.S

[Proprietary Information]

NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-36. Chemical Supply Room Equipment Layout 9-80 NWM I .*;.**.*.* . ......... !:* .

• NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-37 provides the flow diagram for Subsystem I 00, and Table 9-7 provides a s ummar y de scription of the tanks in this subsystem.

[Proprietary Information]

Figure 9-37. Nitric Acid Flow Diagram 9-81 I NWMI *:::**:*:**: ..*... .

• NORTHWESTMEDICAllSOTOPES NWMl-2013-021 , Re v. 1 Chap t e r 9.0 -Aux i l i ary Systems Tab le 9-7. Su b sys t e m 100 , N it r ic A cid Ta nk S i zes Tank number Chemical CS-TK-100 [Proprietary Information]

CS-TK-13 0 [Proprietary Informat i o n] CS-TK-150A [Propri e tary Information]

CS-TK-l 50 B [Proprietary In fo r matio n] CS-TK-180A [Proprietary Information]

CS-TK-180 B [Proprietary Information]

CS-TK-300 [Propri e tary Information]

CS-T K-320 [Pro p rietary Informat i o n] CS-TK-600A/B

/C/D [Proprietary Information]

HN0 3 = nitric ac id. 9.7.4.2.2 S ub sys t e m 2 00 , S odium H y dr ox id e Working volume (L) [Proprietary Information]

[Prop r ietary I nformat i on] [P rop ri etary I n for m a tion] [Prop r ietary Informa t io n] [Propri e tar y In for m at i o n] [Proprietary Infor m atio n] [P rop ri e t a r y In format i o n] [Propr i etary Informatio n] [P ro pr i etary In fo rm at ion] I X Total volume (L) [Proprietary In fo rmati o n] [Proprietary Information]

[Proprietar y Informatio n] [Proprietary Infor m ation] [Propriet a r y Information]

[Proprietary Informa ti o n] [Proprietary I nfor m at i o n] [Proprietary Informatio n] [Proprietary In for mati o n] ion exchange. S u bsystem 200 will co n sist of three tanks that provide the fo ll owing functions:

• * * * [P roprietary I n formatio n] [Proprietary Information

] [P roprieta r y Information]

[Proprietary Information

] 9-82 Diameter (in.) 120 1 2 84 84 84 84 18 1 8

• 135 20 83 83 110 11 0 2 1 2 1

.*; .. ; ... NWMI ...... .. .. . ........ :.* e * . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-38 provides the flow diagram for the NaOH subsystem, and Table 9-8 provides a summary description of the tanks in this subsystem. [Proprietary Information]

Figure 9-38. Sodium Hydroxide Flow Diagram Table 9-8. Subsystem 200, Sodium Hydroxide Tank Sizes Tank number Chemical ----CS-TK-200 [Proprietary Information]

[Proprietary

[Proprietary 84 96 In for m a ti on] In format i o n] CS-TK-230

[Proprietary Information]

[Proprietary

[Proprietary 18 18 Information] (nformation]

CS-TK-260 [Propri etary Information]

[Prop ri etary [Proprietary 24 27 Information]

lnformation]

CS-TK-350

[Proprietary Information]

[Proprietary

[Proprietar y 6 8 Information]

Information]

Na OH sod ium h y dro x ide. NO , n itroge n oxide. 9-83

..... NWMI .*:.**.*.*. ........ !. * * ! . . NOJITHWEST MEDICAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems 9. 7.4.2.3 Subsystem 300, Reductant and NO. Absorber Solutions Subsystem 300 will consist of three tanks that provide the following functions: * * [Proprietary Information]

[Proprietary Information]

Table 9-9 provides a summar y description of the tanks in Subsystem 300. Table 9-9. Subsystem 300, Reductant and Nitrogen Oxide Absorber Solutions Tank Sizes Tank number Chemical C S-TK-300 [Propri e tar y Information]

CS-TK-320

[Proprietary Information]

CS-TK-340 [Propri e tar y Informati o n] NO x ni troge n ox id e. 9. 7.4.2.4 Subsystem 400, Hydrogen Peroxide Subs y stem 400 will provide the followin g functions:

• * [Proprietary Information]

[Proprietar y Information]

Working volume (L) [P ro pri e t a r y In fo rm a ti o n] [Proprietary Information]

[Proprietary In fo rm a t io n] Total volume (L) [Propri eta r y Inform a ti o n] [Proprietary Information]

[P ro p r i etary In fo rm a t io n] Diameter (in.) 1 8 18 6 [Proprietar y Information]

I Height (in.) 2 1 21 8 Figure 9-39 provides the flow diagram for the h y drogen pero x ide subsy s tem. The subsystem w ill consist of one tank (CS-TK-400

), which i s s ummarized in Table 9-10. Figure 9-39. Hydrogen Peroxide Flow Diagram Tank number C S-TK-400 Table 9-10. Subsystem 400, Hydrogen Peroxide Tank Sizes Chemical Hydro g en p e roxide Working volume (L) [Propri e tary In fo rm a ti o n] 9-84 Total volume (L) [Proprietary Inform a ti o n] I Diameter (in.) 9 I Height (in.) 12

.;.-.;* .. NWMI .... ** . .. .. . .*.* .. *.*.* * * ." . NORTHWEST MEDfCAl lSOTOPfS NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems 9.7.4.2.5 Subsystem 600, Fresh Uranium Ion Exchange Resin Subsystem 600 will consist of four tanks (one tank to support each uranium IX col umn) that provide the following functions:

• * * [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

Tab le 9-11 provides a summary description of the tanks in Subsystem 600. Table 9-11. Subsystem 600, Fresh Uranium Ion Exchange Resin Tank Sizes Tank number Chemical CS-TK-600 A [Proprietary Information]

CS-TK-600B [Proprietary Information]

CS-TK-600C

[Proprietary Information]

CS-TK-600D

[Proprietary Information]

IX = ion exc han ge. Working volume (l) [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

Total volume (l) [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

Diameter (in.) 24 24 24 24 Height (in.) 24 24 24 24 These tanks will support preparation of fresh resin for addition to an IX column after spent resin has been removed. A description of the fresh resin makeup activity is summarized as follows: * [Proprietary Information]

• [Proprietary Information]

• [Proprietary Information]

• [Proprietary Information]

• [Proprietary Information]

Once resin has been prepared by fines removal and washing , the makeup tank will be adjusted to contain a total volume of [Proprietary Information]. The makeup tank low-speed agitator will be started to suspen d the resin inventor y, and the valve opened to route the suspension to an IX column. 9-85

.. ; ... ; ... NWMI ...... .. .. . .... -..*... e * . NORTHWEST MEDICAL ISOTDPf.S

9. 7.4.3 Operational Analysis and Safety Function NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Chapter 13.0 evaluates accident sequences that involve miscellaneous chemical safety process upsets in areas without significant fissile or high-dose licensed material present (e.g., chemical storage areas and the laboratory).

The backflow of fissile or radioactive solutions into auxiliary systems (e.g., chemical supply) was also analyzed and two preventive IROFS identified. Defense-in-depth

-NWMI will comply with U.S. Environmental Protection Agency and Occupational Safety and Health Administration regulations for the design , construction , and operation of chemical preparation and storage areas in the RPF. Chemical handling procedures will be provided to operators to ensure safe handling of chemicals according to applicable regulatory requirements and consistent with the material safety datasheets. Items relied on for safety -Based on the analysis conducted in Chapter 13 .0, Section 13 .2, the following IROFS will be implemented:

• 9.7.4.4 CS-18 , " Backflow Prevention Device" CS-19 , " Safe Geometr y Day Tanks" Instrumentation and Control Requirements Instrumentation and control requirements for the processes associated with the chemical supply system will be discussed in Chapter 7.0 as part of the Operating License Application.

9.7.4.5 Required Technical Specifications The technical specifications associated with the chemical supply system , if applicable , will be discussed in Chapter 14.0 as part of the Operating License Application.

9-86

.:;:.;* .. NWMI .*;.**.*.* . ......... !:* .. ***** . . * * * . NORTHWEST MEDICAl ISOTOPES

9.8 REFERENCES

NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems 10 CFR 20, "Standards for Protection Against Radiation ," Code of Federal Regulations, Office of the Federal Register , as amended. 10 CFR 50, " Domestic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register , as amended. 40 CFR 61, "Na tional Emission Standards for Hazardous Air Pollutants

," Code of Federal R egu lation s, Office of the Federal Register , as amended. 42 U.S.C. 2011 et seq., "Ato mic Energy Act of 1954 ," U nit ed States Code, as amended. ICC, 2012, " International Building Code (IBC) and Commentary 2012," International Code Council, Falls Church, Virginia, 2012. IFC, 2012, International Fire Code, International Code Council, Falls Church, Virginia, 2012. ISO 14644-1 , "Cleanrooms and Associated Controlled Environments

-Part 1: Classification of Air Cleanliness," International Organization for Standardization, Geneva, Switzerland, 1999. NFPA 10 , Standard for Portable Fire Extinguis h ers, National Fire Protection Association, Quincy, Massachusetts, 2013. NFPA 13 , Standard for the Installation of Sprinkler Systems, National Fire Protection Association, Quincy, Massachusetts , 2013. NFP A 24, Standard for the Installation of Private Fire Service Mains and Their Ap purtenanc es, National Fire Protection Association, Quincy, Massachusetts , 2013. NFPA 25, Standard for the Inspection, T esti ng , and Maintenance of Water-Based Fire Protection Systems, National Fire Protection Association, Quincy, Massachusetts, 2014. NFP A 45, Standard on Fire Protection for Laboratories Using Chemicals, National Fire Protection Association, Quincy, Massachu setts, 2015. NFP A 72, Nationa l Fire A larm and Signaling Code, National Fire Protection Association, Quincy, Massachusetts , 2013. NFP A 80, Standard for Fire Doors and Other Op ening Prot ectives, National Fire Protection Association, Quincy, Massachusetts , 2013. NFPA 90A, Standard for the Installation of Air-Conditioning and Ventilating Systems, National Fire Protection Association, Quincy, Massachusetts, 2015. NFPA 92, Standard for Smoke Control Systems, National Fire Protection Association, Quinc y, Massachusetts, 2015. NFPA 101, Life Safety Code, National Fire Protection Association, Quincy, Massachusetts , 2015. NFPA 221, Standard for High Challenge Fire Walls, Fire Walls, and Fire Barri er Walls, National Fire Protection Association, Quincy , Massachusetts , 2015. NRC, 2012, Final Interim Staff Guidance Augmenting NUREG-1537 , "G uidelin es for Preparin g and R eviewing Applications for the Licensing of No n-Power R eacto rs ," Part s 1 and 2, for Licensing Radioisotope Produ c tion Facilities and Aqueous Homog eneous Reactors , Docket ID: NRC-2011-0135 , U.S. Nuclear Regulatory Commission, Washington, D.C., October 30, 2012. 9-87

... ;.-.;* .. NWMI .*:.**.*.* . . * ....... !:* .. ***** . . * *. *

• NORTHWEST MEDJCAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems NUREG-1537 , Guidelines for Preparing and Reviewing Applications for the Lic e nsing of Non-Po wer Reactors -Format and Content, Part 1 , U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation , Washington, D.C., February 1996. NWMI-2013-039, Preliminary Fire Ha z ards Analysis, Rev. C, Northwest Medical Isotopes, LLC, Corvallis, Oregon , 2015. NWMI-2013-CALC-005 , Tank Air Bl eed Estimate, Re v. B , Northwest Medical Isotopes , LLC , Corvallis , Oregon , 20 14. NWMI-2013-CALC-009 , Uranium Purification S ys tem Equipment Si z ing, Rev. B , Northwest Medical Isotopes, LLC , Corvallis , Oregon, 2014. Regulatory Guide 1.189, Fire Protection for N ucl e ar Po wer Plants , U.S. Nuclear Regulatory Commission , 2009. 9-88

• * * * * * * * * ****** * * ** ** * ** * ** * * * ** * ** * * ** * * . *. *. * . NORTHWEST MEDICAL ISOTOPES Prepared by: *

• Chapter 13.0 -Accident Analysis Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 June 2017 Northwest Medical Isotopes , LLC 815 NW gth Ave , Suite 256 Corvallis , Oregon 97330 This page intentionally left blank.

...... .. NWMI ...... ..* .... ........ *.* . * ! : . NOITlfWUT MEDICAL tsOTDPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analys i s Chapter 13.0 -Accident Analysis Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

June 26 , 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 13.0 -Accident Analysis Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

.**.*.*.* ..... .. NWMI .*.* .. *.*.* * *

• NORTHWEST M£DICAI. ISOTDf'U This page intentionally left blank. NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis

.; ... .. NWMI ..*... ..* .... ........... 0 "NORT'NWEST MEDtCAL ISOTOffl Rev Date 0 6/29/2015 1 6/26/2017 REVISION HISTORY Reason for Revision Initial Application NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analys is Revised By Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional Information

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

• IMMITllMST IEMtAL tsOnftl This page intentionally left bl ank. NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis CONTENTS NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis 13.0 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSIS .................................

1 3-1 13.1 Accident Analysis Methodology and Preliminary Hazards Analysis ....................

...........

13-3 13.1.1 Methodologies App lied to the Radioisotope Production Faci lity Integrated Safety Analysis Process ............

................................................

.........................

13-3 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and fusk Matrix ..............................................................

13-5 13. I .1 .2 Accident Consequence Analysis ....................................................... 13-7 13. l. l.3 What-If and Structured What-If..

......................................................

13-7 13.1.1.4 Hazards and Operability Study Method ............................................

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

.....................................................

1 3-8 13.1.1.6 Fault Tree Analysis ................................

...........................................

13-8 13.1.1. 7 Failure Modes and Effects Analysis ..................................................

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

............

......... 13-8 13 .1.3 Preliminary Hazards Analysis Results ...................................

..........................

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

........................

...................................

..... 13-12 13.1.3.2 Radioisotope Production Facility Accident Sequence Evaluation

................................

.............

.....................................

..... 13-13 13.2 Analysi s of Accidents with Radiological and Criticality Safety Consequences

............

1 3-39 13.2.1 Reserved ........................................................

................................................... 13-40 13.2.2 Liquid Spills and Sprays with Radiological and Critica lit y Safety Consequences

.............

......................

................................................................ 13-40 13.2.2.1 Initial Conditions

.............................................................................

13-40 13.2.2.2 Identification of Event Initiating Conditions

..................................

13-45 13.2.2.3 Description of Accident Sequences

...............................

.................. 13-45 13.2.2.4 Function of Components or Barriers ...............................................

13-45 13.2.2.5 Unmitigated Likelihood

..............................................

....................

13-46 13.2.2.6 Radiation Source Term .....................

..............................

..............

.. 13-46 13.2.2.7 Evaluation of Potential Radiological Consequences

....................... 13-48 13.2.2.8 Identification ofltems Relied on for Safety and Associated Functions

.........................

..........................

..........................

............ 13-51 13.2.2.9 Mitigated Estimates

.........................................................................

13-55 13.2.3 Target Dissolver Off gas Accidents with Radiological Consequences

.............

13-55 13.2.3.1 Initial Conditions

.............................................................................

13-56 13.2.3.2 Identification of Event Initiating Conditions

..................................

13-57 13.2.3.3 Description of Accident Sequences

..........

..........

.............................

13-57 13.2.3.4 Function of Components or Barriers ....................

............

...............

13-57 13.2.3.5 Unmitigated Likelihood

..................

.............................

................

... 13-57 13.2.3.6 Radiation Source Term ...................................................................

13-58 13.2.3.7 Evaluation of Potential Radiological Consequences

.......................

13-58 13.2.3.8 Identification ofltems Relied on for Safety and Associated Functions

...................................................

.................

.....................

13-59 13.2.3.9 Mitigated Estimates

..........................

............

...................

................

13-60 13-i

...... .. NWMI ...... ..* .... ............ . ". NOftTKWESTMEDtCAI.

ISOTOPH 13.2.4 13.2.5 13.2.6 13.2.7 NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Leaks into Auxiliary Services or Systems with Radiological and Criticality Safety Consequences

.....................................

................................. 13-60 13.2.4.1 Initial Conditions

............................................................................. 13-60 13.2.4.2 Identification of Event Initiating Conditions

.................................. 13-64 13.2.4.3 Description of Accident Sequence s ..................

....................

........... 13-65 13.2.4.4 Function of Components or Barriers ............

........................

........... 13-65 13.2.4.5 Unmitigated Likelihood

....................................

.......................

....... 13-65 13.2.4.6 Radiation Source Term ..........

..........

....................

................

........... 13-66 13.2.4.7 Evaluation of Potential Radiological Consequences

..................

..... 13-66 13.2.4.8 Identification ofltems Relied on for Safety and Associated Functions

................................................................

.................

........ 13-66 13.2.4.9 Mitigated Estimates

..............

....................

....................................... 13-70 Loss of Power. ....................................................

.............................................. 13-70 13.2.5.l Initial Condit i ons ..............................

............................................... 13-70 13.2.5.2 Identification of Event Initiating Conditions

................

.................. 13-70 13.2.5.3 Description of Accident Sequences

...............................................

.. 13-70 13.2.5.4 Function of Components or Barriers ..............

.............

.................... 13-71 13.2.5.5 Unmitigated Likelihood

................

............................................

...... 13-71 13.2.5.6 Radiation Source Term ..................

................................................. 13-71 13.2.5.7 Evaluation of Potential Radiological Consequences

....................... 13-71 13.2.5.8 Identification of Items Relied on for Safety and Associated Functions

.............................................

............................................ 13-71 Natural Phenomena Events ..............................................................

................ 13-72 13.2.6.1 Tornado Impact on Facility and Structures , Systems , and Components

..................................................................................... 13-72 13.2.6.2 High Straight-Line Winds Impact the Facility and Structures , Systems , and Components

............

..................................

................

13-73 13.2.6.3 Heavy Rain Impact on Facility and Structures , Systems, and Components

..................................................................................... 13-73 13.2.6.4 Flooding Impact to the Facility and Structures , Systems , and Components

..................................................................................... 13-74 13.2.6.5 Seismic Impact to the Facility and Structures , Systems , and Components

.....................

................................................................ 13-74 13.2.6.6 Heavy Snow Fall or Ice Buildup on Facility and Structures , Systems , and Components

...........................................

................... 13-75 Other Accidents Analyzed ....................

........................................................... 13-76 13.2.7.1 Items Relied on for Safety for Radiological Accident Sequences (S.R.) ..............................

...............................................

13-86 13.2.7.2 Items Relied on for Safety for Criticality Accident Sequences (S.C.) ............................................................................................... 13-88 13.2.7.3 Items Relied on for Safety for Fire or Explosion Accident Sequences (S.F.) .............................................................................. 13-94 13.2.7.4 Items Relied on for Safety for Natural Phenomena Accident Sequences (S.N.) ...................................

.......................................... 13-94 13.2.7.5 Items Relied on for Safety for Man-Made Accident Sequences (S.M.) ..........................................

.................................................... 13-95 13.2.7.6 Items Relied on for Safety for Chemical Accident Sequences (S.CS.) .................................................................

............................ 13-95 13-ii

......... * .. :.;.-.;* .. NWMI ........ *.* * . NORTHWEST MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.3 Analysis of Accidents with Hazardous Chemicals

...................................................

...... 13-96 13.3.1 Chemical Burns from Contaminated Solutions During Sample Analysis ....... 13-96 13.3.1.1 Chemical Accident Description

...................................................... 1 3-96 13.3.1.2 Chemical Accident Consequences

.................................................. 13-9 6 13.3.1.3 Chemical Proce ss Co ntrols ............................................................. 13-96 13.3.1.4 Chemical Proce ss Surveillance Requirement s ................................ 13-9 6 13.3.2 Nitric Acid Fume Rele ase ............

..........................................................

.......... 13-97 13 .3 .2.1 Chemical Accident Description

...................................................... 13-97 13.3.2.2 Chemical Accident Consequences

.............................

..................... 13-97 13.3.2.3 Chemical Proce ss Con trols ................

............................................. 13-97 13.3.2.4 Chemical Proce ss Surveillance Requirements

................................ 13-97 13.4 Reference s ......................................

............

....................................................................

13-9 8 13-iii

.. NWMI ..*... ..* .... ........... . * " "NORTifWHTllEOICALISOTOPfl NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis FIGURES Figure 13-1. Integrated Safety Analysis Process Flow Diagram ........................................................ 13-4 Figure 13-2. Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident...

................... 13-50 TABLES Table 13-1. Likelihood Categories

........................................................

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

................................................................. 13-5 Table 13-3. Radioisotope Production Facility Consequence Severity Categories Derived from lOCFR 70.61 ................................................................................................................. 13-6 Table 13-4. Radioisotope Production Facility Risk Matrix .............................................................. 13-6 Table 13-5. Radioisotope Production Facility Preliminary Hazard Analysis Accident Sequence Category Designator Definitions

................................................................... 13-9 Table 13-6. Crosswalk ofNUREG-1537 Part 1 Interim Staff Guidance Accident Initiating Events versus Radioisotope Production Facility Preliminary Hazards Analysis Top-Level Accident Sequence Categories

............................................................

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

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

.................................... 13-12 Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) ....................................................... 13-14 Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation ( 4 pages) ....................................................... 13-18 Table 13-11. Adverse Event Summary for Molybdenum Recovery and Identification of Accident Sequences Needing Further Evaluation (3 pages) ........................................ 13-22 Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) ....................................................... 13-25 Table 13-13. Adverse Event Summary for Waste Handling and Identification of Accident Sequences Needing Further Evaluation (2 pages) ....................................................... 13-29 Table 13-14. Adverse Event Summary for Target Receipt and Identification of Accident Sequences Needing Further Evaluation (2 pages) ....................................................... 13-31 Table 13-15. Adverse Event Summary for Ventilation System and Identification of Accident Sequences Needing Further Evaluation

....................................................................... 13-33 Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) .......................................

..................................

13-34 Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) ............................... 13-41 Table 13-18. Source Term Parameters

................................................................................

.............. 13-47 Table 13-19. Release Consequence Evaluation RASCAL Code Inputs ........................................... 13-49 13-iv

... .. NWMI ...... ... .... ..... .... .. * ' NORTHWEST MEDICAL ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-20. Spray Release Consequence Summary ...................

.........................

............................

13-50 Table 13-21. Maximum Bounding Inventory of Radioiodine

[Proprietary Information]

................. 13-56 Table 13-22. Target Dis s olver Offga s Accident Total Effective Dose Equivalent..

................

......... 13-59 Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 page s) ...........................

.... 13-61 Table 13-24. Analyzed Accidents Sequences (9 pages) ....................

......................

..........................

13-76 Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages) ...........

...........................................

....................

...............

..............................

13-84 Table 13-26. Accident Sequence Category Definitions

..........

.................

..............

........................... 13-86 13-v

.. .. NWMI ...*.. ..* **.* ........... * * * .° NOflTHWEST MEDICAl lSOTO'll TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 99 mTc technetium-99m 235 U uranium-235 2 41Am AAC AC ACI AEC AEGL AISC ALARA ALOHA ARP ASCE COE CEDE CFR DAC DOE DOT DR EDE EOI ETA FEMA FMEA FTA HAZOP HEGA HEPA HIC HN0 3 HVAC IBC IROFS IRU ISA ISG IX americium-241 augmented administrative contro l administrative contro l American Concrete Institute active engineered control Acute Exposure Guideline Level American Institute of Steel Construction as low as reasonably achievab l e areal lo cations of hazardous atmospheres airborne release fraction American Society of Civil Engineers committed dose eq ui va l ent committed effective dose equiva l ent Code of Federal Regulations derived air concen tration U.S. Department of Energy U.S. Department of Transportation damage ratio effective dose equivalent end of irradiation event tree analysis Federa l Emergency Management Agency failure modes and effects ana l ysis fault tree analysis hazards and opera bility high-efficiency gas adsorption high-efficiency particulate air high-integrity canister nitric acid heating , ventilation, and air cond itionin g International Building Code items relied on for safety iodine removal unit integrated safety analysis Interim Staff Guidance ion exchange low enriched uranium leak path factor material at risk maximum hypothetical accident molybdenum University of Missouri Research Reactor sodium hydroxide nondestructive assay NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis LEU LPF MAR MHA Mo MURR Na OH NOA NIOSH National Institute for Occupational Safety and Health 13-vi

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

• NOITHWUT M£DICAL ISOTOP£S NO , NOAA NRC NWMI NWS OSTR osu P&ID PEC PFD PHA PMP QRA RASCAL RF RPF RSAC SNM SSC ST TCE TEDE u U.S. UN nitrogen oxide NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis National Oceanic and Atmospheric Administration U.S. Nuclear Regulatory Commission Northwest Medical Isotopes, LLC National Weather Service Oregon State University TRIGA Reactor Oregon State University piping and instrumentation drawing passive engineered control process flow diagram preliminary hazards analysis probable maximum precipitation quantitative risk assessment Radiological Assessment System for Consequence Analysis respirable fraction Radioisotope Production Facility Radiological Safety Analysis Code special nuclear material structures , systems , and components source term trichloroethylene total effective dose equivalent uranmm United States uranyl nitrate 13-vii

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

• NORTHWEST MEDICAi. ISOTDf'£S Units o c o p Ci Cm ft ft 3 g hr in.2 kg km km 2 L lb m M m 3 mg ITI1 mi 2 mil ITiln mrem oz ppm rem sec Sv wk wt% yr degrees Celsius degrees Fahrenheit cune centimeter feet cubic feet gram hour square inch kilogram kilometer squa re kilometer liter pound meter molar cubic meter milligram mile square mile thousandth of an inch minute millirem ounce parts per million roentgen equivalent man second sievert week weight percent year 13-v ii i NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis

.-.; ... NWMI

: ....... .. ........ *.* .. ***** NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis . * *

• NOAllfW(n llEDtcAL ISOTOPES 13.0 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSIS The proposed action is the issuance of a U.S. Nuclear Regulatory Commission (NRC) Construction Permit and Operating License under Title 10 , Cod e of F e deral R e gulations, Part 50 (10 CFR 50) "Domestic Licensing of Production and Utilization Facilities," and provisions of 10 CFR 70 , "Domestic Licensing of Special Nuclear Material,'

' and 10 CFR 30 , "Rules of General Applicability to Domestic Licensing of Byproduct Material,'

' that would authorize Northwest Medical Isotopes , LLC (NWMI) to construct and operate a molybdenum-99 (99 Mo) Radioisotope Production Facility (RPF) at a site located in Columbia , Missouri. The RPF is being designed to have a nominal operational processing capability of one batch per week of up to [Proprietary Information].

The primary mission of the RPF will be to recover and purify radioactive 99 Mo generated via irradiation of low-enriched uranium (LEU) targets in off-site non-power reactors.

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

Irradiation services and transportation activities are not analyzed in this chapter. This chapter evaluates the various processing and operational activities at the RPF , including: Receiving LEU from U.S. Department of Energy (DOE) Producing LEU target materials and fabrication of targets Packaging and shipping LEU targets to the university reactor network for irradiation Returning irradiated LEU targets for dissolution , recovery , and purification of 9 9 Mo Recovering and recycling LEU to minimize radioactive , mixed , and hazardous waste generation Treating/packaging wastes generated by RPF process steps to enable transport to a disposal site Chapter Organization Section 13.1 describes hazard and accident analysis methodologies applied to the RPF integrated safety analysis (ISA) (Section 13.1.1 ). Section 13.1.2 identifies the accident initiating events, and Section 13.1.3 summarizes the results of the RPF preliminary hazards analysis (PHA) (NWMI-2015-SAFETY-001 , NWMI Radioisotop e Production Facili ty Prelimina ry Ha z ard s Analy s i s). The PHA discussion in Section 13.1.3 identifies the accident scenarios that required further evaluation.

Section 13.2 presents analyses of radiological and criticality accidents , including: Section 13.2.1 (Reserved)

Section 13.2.2 discusses spills and spray accidents Section 13.2.3 discusses dissolver offgas accidents Section 13.2.4 discusses leaks into auxi li ary systems accidents Section 13.2.5 discusses loss of electrical power

• Section 13.2.6 discusses natural phenomena accidents Section 13.2. 7 identifies the additional accident sequences evaluated and associated items relied on for safety (IROFS) 13-1

...... NWMI ...... ...* ... .... .. .. .. ' ! . NOATMWUT MlDICAl lSOTOPEI NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Section 13.3 presents bounding accidents involving hazardous chemicals.

The data presented in the following subsections are based on a comprehensive PHA , conservative assumptions , the MHA result s , draft quantitative risk assessments (QRA), and scoping calculations. These items provide an adequate basis for the construction application. 13-2 NWMI ..*... ... .... ........ *.* .. *****. * * *

• MEDlCAI. ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.1 ACCIDENT ANALYSIS METHODOLOGY AND PRELIMINARY HAZARDS ANALYSIS 13.1.1 Methodologies Applied to the Radioisotope Production Facility Integrated Safety Analysis Process This section de sc ribes methodologies app lied to the RPF ISA. The ISA process comprises the PHA and the follow-on development and completion of QRAs to address events and hazards identified in the PHA as requiring further evaluation. The ISA process flow diagram is provided Figure 13-1. The ISA process (being adapted for this application) consists of conducting a PHA of a system using a combination of written process description s , process flow diagrams (PFD), process and instrument drawings (P&ID), and supporting calculations to identify events that could lead to adverse consequences.

Those adverse consequences are evaluated qualitatively by the ISA team members to identify the likelihood and severity of consequences using guidance on event frequencies and consequence categories consiste nt with the regulatory guidelines. Each event with an adver se consequence that involve s licensed material or it s byproducts is evaluated for risk using a risk matrix that enables the user to identify unacceptable intermediate-and high-consequenc e risks. For the unacceptable intermediate-and high-consequence risks events , the IROFS developed to prevent or mitigate the consequences of the events and an event tree analysis are used to demon s trate that the risk can be reduced to acceptable frequencies through preventati ve or mitigati ve IROFS. Fault trees and failure mode and effects analysis can be used to (I) provide quantitative failure analysis data (failure frequencies) for use in the event tree analysis of the IROFS, as nece ssary, or (2) quantitatively anal yze an event from its basic initiators to demon strate that the quantitati ve failure frequency is already highl y unlikely under normal s tandard industrial conditions , thus not needing the application of IROFS. Once the IROFS are developed , management measure s are identified to ensure that the IROFS failure frequency u se d in the analysi s i s preserved an d th e IROFS a re able to perform th eir intended function when needed. The following s ub sec tions s ummari ze the RPF ISA methodologies.

13-3

.. ... ; .. NWMI ...... *.** *.. ........ *.* ' *. ! . NOIT1fW£ST MEOtcAL ISOTOKS Desl1n and Safety Functions Develop process descriptions, PFDs, P&IDs Identify preliminary hazards and consequences (radiological, criticality, chemical, fire, external) using regulatory guides where applicable Develop CSAs, FHA, and other support documents Initiate ISA process by collecting prelimina data Perform PHA on facility operations Categorize events for likelihood, consequence, and risk lndeter-Yes Perform QRA to quantitatively evaluate risk and identify IROFS High or intermediate Yes Identify "accident sequence" and develop IROFS and basis for each i n complete QRA Document identified low-risk events (no IROFS) No Start Phase 1 development of IROFS boundary definition packages for each IROFS NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Deslpand EnstnHrtn1 Functions Design function development of IROFS specifications/

conceptual drawin s Develop PSAR, ISA '------+ ... summary, technical specifications Complete Phase 1 development of IROFS boundary definition packages ISA team review and recommendation for a roval Management approval of ISA basis NRC review of document i----------+---------1.

license submit to NRC application .m:_r Figure 13-1. Integrated Safety Analysis Process Flow Diagram 13-4

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

• NOmfWEST llEDICAL tsOTOPlS NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.1.1.1 Accident Likelihood Categories, Consequence Severity Categories, and Risk Matrix Table 13-1 shows the accident likelihood categories applied to the RPF ISA proces s. Table 13-2 shows qualit at ive guidelines for applying the likelihood categories from Table 13-1. Table I 3-3 s how s accident co n se quence seve rity categories from Table 13-1. Likelihood Categories I 0 CFR 70.61, " Perform a nc e Requirements

." Table 13-4 s ho ws the RPF risk matri x, w hich is a produ ct of the likelih oo d a nd consequence sever ity categories from Table 13-1 and Table 13-3 , respectively.

Not unlikely Unlikely Highly unlik e l y 2 I Event frequency limit Between I 0-3 and I 0-5 event s per year Less than I 0-5 per eve nt s per year Table 13-2. Qualitative Likelihood Category Guidelines

., .* Initiator 3 An eve n t initiated by a human error 3 An event initiat e d by failure of a proces s system processing corrosive materials 3 An eve nt initi ate d by a fire or exp lo s ion in areas w h ere comb u st ibl es or flammable materials are present 3 An event initiated by failure of an active control system 3 A dama gi n g se i s mic eve nt 3 A damaging high wind event 3 A sp ill of material 3 A failure of a process variable monitored or unmonitored by a control system 3 A va l ve o ut of position or a va l ve that fails to sea t and i solate 3 Most standard industrial component failures (valves, sensors , safety devices , gauges, etc.) 3 An adverse chem i ca l reaction caused by improp e r quantities of r eactants , out-of-d a t e reactants, o utof-s p ec ifi ca ti on r eac tion environment , or the wrong reactants are u se d 3 Most external man-made events (until confirmed using an approved method) 2 A n eve nt initi ate d by th e failure of a robust passive d es i gn feat ur e wit h no s ignificant int erna l or externa l c h a ll enges a pplied (e.g., s pont a n eous rupture of an all-we ld ed dr y nitrogen system pipe operat in g at o r below d es i g n pressure in a clean , vi bration-free e n viro nm ent) 1-2 An adverse chemical reaction when proper quantities of in-date chemicals are reacted in the proper environment Natura l phenomenon s uch as tsunami , volcanos , a nd asteroids for th e Missouri faci lity site 13-5

..... ; .. NWMI ...*.. ..* *.. ........ *. 0 ! * * . NOIO'NWEST lffotCAL tsOTOfl'£S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-3. Radioisotope Production Facility Consequence Severity Categor i es Derived from 10 CFR 70.61 lllill Consequence category Workers Off-site public Environment High conse qu e nc e Intermediate consequence Low conseq u e nc e 3 2

• Radiolo g i ca l do se* > I Sv (I 00 rem) . A irborn e, rad iol ogica ll y co nt aminated nitric aci d > 1 70 ppm nitric ac id (AEGL-3 , 10-min exposure limit) . Uns hi e ld edb nuclear cri ti cality

• Radiological dose* between 0.25 Sv (25 rem) and I Sv (100 rem)

• Airborne, radiologically contaminated nitric acid > 43 ppm nitric acid (AEGL-2, 10-min exposure limit) Acc id e nt s with lower radio lo g ic a l , c h e mical , and/or toxicological expos ur es than tho se above from licensed material and byproducts of li cense d material

• R a diol og ic a l do se* > 0.2 5 Sv (25 rem) . Toxic intake> 30 m g so lubl e U

• Airborne , co nt aminate d nitri c aci d > 24 ppm nitri c ac id (AEGL-2, 60-min expos ur e limi t)

• Radiological dose* 24-hr radioactive between 0.05 Sv (5 rem) re l ease> 5,000 x and 0.25 Sv (25 rem) Table 2 of

• Airborne , contaminated IO CFR 20 , 0 nitric acid > 0.16 ppm Appendix B nitric acid (AEGL-1, 60-min exposure limit) Accidents wit h lo wer Radiological radiological , c h em i ca l , releases producin g a nd/or toxicological low er effects than e xposures than tho se above those li s ted a b ove from li censed m ateria l a nd from li ce n se d byproducts of li ce n sed m ate ri al material Sou r ce: I 0 CFR 70.6 1 , " Per formance R e quirement s," Code of Fed er a l R eg ulati ons , Office of the Fede r al R eg i s ter , as amended.

• As tota l effec ti ve do se equivale nt. b A s hi e ld ed criticality accide nt is a l so considered a hi g h-co n se qu e n ce event. c I 0 CFR 20, "S tandard s for Prote c ti on Against R a di ation ," Code of Federal R egu lation s, Office of the Fed eral R egis ter , a s ame nd e d. AEG L Acute Ex po s ur e G u ide lin e Level. u = uranium. Table 13-4. Radioisotope Productio n Facility Risk Matrix Severity of consequences High consequence (Consequence category 3) Intermediate consequence (Consequence category 2) Low consequence (Consequence category 1) Highly unlikely (Likelihood category 1) Risk index = 3 Acceptable risk Ri s k ind ex = 2 Acceptable ri sk Risk index = l Acceptable risk Likelihood of occurrence Unlikely (Likelihood category 2) Risk index= 6 Unacceptable risk Ri s k ind ex = 4 Acceptabl e risk Risk index = 2 Acceptable risk 13-6 Not unlikely (Likelihood Category 3) Risk index = 9 Unacceptable risk Risk index = 6 Unacceptable risk Risk index = 3 Acceptable risk

..... ; .. NWMI ..**.. ..* ... ..... .. .. .. 0 *.* ! 0 NOmfW£ST tllDtCAl ISOT0.-£1 NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.1.1.2 Accident Consequence Analysis The ISA process requires an understanding of the source terms and consequences of an adverse event to determine if the event is low , intermediate, or high consequence, as compared with the hazard criteria identified in Table 13-4. NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook , offers methodologies to calculate the quantitative consequences of events. For simplicity and prudent expenditure of resources , the RPF ISA assumes a worst-case approach using a few bounding evaluations of events that are identified through either: Calculations (e.g., the source term and radiation doses caused by contained material in the system) Studies of representative accidents (e.g., comparison of accidental criticalities in industry with processes similar to those at the RPF) Bounding release calculations using approved methods (e.g., using RASCAL [Radiological Assessment System for Consequence Analysi s] to model bounding facility releases that affect the public) Reference to nationally recognized safety organizations (e.g., use of Acute Exposure Guideline Le ve ls [AEGL] from the U.S. Environmental Protection Agency to identify chemical exposure limits for each consequence category)

• Approved method s for evaluation of natural and man-made phenomenon and comparison to the design basis (e.g., calculation of explosive dama ge potential from the n eares t railroad line on the facility)

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

material spills; wrong materials , place, or time for activities; etc.). The key words for each structured what-if evaluation are documented in the PHA. 13-7

....... .. NWMI ...... ... .... ..... .. .. .. ' !*. ! 0 NOmlWUT MEDtcAl tsOTOPES 13.1.1.4 Hazards and Operability Study Method NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis For processes that are part of a processing system and have well-defined PFDs and/or P&IDs , the more structured hazards and operability (HAZOP) approach was used. This method systematically evaluates each node of a process using a set of key words that enables the team to systematically identify adverse changes in the process and evaluate those changes for adverse consequences.

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

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

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

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

This is an exhaustive analysis technique that can be used to evaluate the reliability of a comp le x , active engineered control (AEC) type of IROFS. 13.1.2 Accident-Initiating Events Each of the following accident initiating events was included in the PHA. Loss of power as an accident event is discussed further in Section 13.2.5. Criticality accident Loss of electrical power External events (meteoro log ical, seismic , fire , flood)

• Critical equipment malfunction Operator error Facility fire (explosion is included in this category)

Any other event potentially related to unique facility operations 13-8

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

• HOITifWEST MEOfCAl tSOTOP'lS The PHA (NWMI-2015-SAFETY-001) i dentifi es and categorizes accident seq u e nce s th a t require further evaluation.

T a ble 1 3-5 define s the level accident se quen ce notation u se d in the RPF PHA. Table 13-6 provides a crosswalk between the PHA top-level accident sequence categories and the NUREG-15 3 7 , Guidelin es fo r Pr eparing and R ev iewing App li c ation s for t h e Lic e n s ing of NonPower R eacto r s -Format and Co nt ent, Part 1 Interim Staff Guidance (ISO) accident initiating events listed above. As noted at the bottom of Table 13-6 , PHA accident se quence s involve one or more of the NUREG-1537 Part 1 ISO accident initiating event categories, as not ed b y ../ in the corres ponding table ce ll , but the PHA acc ident seq uence s themselves a r e not ne cessari l y initiated b y the IS O acc id e nt initi ating eve nt. Table 13-6 NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-5. Radioisotope Production Facility Preliminary Hazard Analysis Accident Sequence Category Designator Definitions PHA top-level accident sequence categorya S.C. S.F. S.R. S.M. S.N. S.C S. Definition C riti cality Fire or explosion Radiological Man-made Natural phenomena Chemical safety

• The a lph a category de s i g n ato r i s followed in th e PHA b y a two-di git num ber "X X" th a t refers to the specific acci d ent seq u e n ce (e.g., S.C.01, S.F.07). Specific acc id ent se quence s are di sc u ssed in Sect i o n s 1 3.1.3 and 1 3.3. PHA = prelimi n ary haz a rd ana l ysis. s how s how PHA acci d ent se quence s correspond with ISO accident initiating events, a nd demonstrat es that the PHA co n s ider s the full range of acc id e nt events id e ntifi e d in the ISO. Table 13-6. Crosswalk of NUREG-1537 Part 1 Interim Staff Guidance Accident Initiating Events versus Radioisotope Production Facility Preliminary Hazards Analysis Top-Level Accident Sequence Categories NUREG-1537 a Part 1 ISG accident initiating event category Critica lit y accident Loss of electrical power Exte rn a l events (meteoro l ogica l , se ismic , fire, flood) Critical equipment malfunction Operator error Facility fire (explosion is included in this category)

Any other event potent i a ll y rel a ted to unique faci li ty operat i o n s PHA Top-Level Accident Sequence Categoryb

--mJ:m--,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/

• NUREG-153 7, Guidelines for Preparing and R eview in g Appli c ations for the Licensing of Non-Power R eactors -F or mat and Co nt e nt , Part I , U.S. uclear Regulatory Co mmission , Office of N u c l ear Reactor Regulation , Was hin g ton , D.C., February 1 996. b PHA accident se quence s in vo l ve one or more of the NUREG-15 3 7 Part 1 l SG acci d e nt initi ati n g eve nt ca t egor i es , as not e d by a n ./ in the co rr esponding table ce ll , but the PHA seq u ences them se l ves are not n ecessa ril y init iat ed by the lSG acc id e nt initiating event. I SG = Int e rim Staff Guidance.

PHA = preliminary h azard ana l ys i s. 13-9

.. ... ; .. NWMI ..*... ..* .... .*.* .. *.*.* * * . NORTHWUT MEDICAL ISOTDPH NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Th e RP F P HA s ubdi v id es th e RPF pro cess i nto e i g ht p ri m ary n o d es b ase d o n fac il ity de s i gn d oc um e nt a t ion. Ta bl e 1 3-7 l ists t h e RPF p rim ary n o d es a nd co rr es p on d i ng s ubp rocesses, as id enti fi e d in t h e PH A. Table 13-7. Radioisotope Production Facilit y Preliminar y Hazards Analysis Primar y Proce s s Nodes and S ubproce ss e s (2 pages) Node no. Node name Subprocesses encompassed in node 1.0.0 2.0.0 3.0.0 4.0.0 T arge t fa b rica t ion process Targ e t di s solution proces s Mo l y bd e num recovery an d pur i fi ca ti on process Uran i um reco v ery and r ec y cle proce ss

• Fres h ur a nium rece i p t a nd s t orage

• Fresh u ra nium d isso luti o n

• Ura n y l ni trate bl e n di ng a n d feed pre p ara ti o n

• Nitrate ex t ractio n

• R ecyc l ed u ra n y l nitrate c on centration

• [Prop ri e t ary I nformat i o n] * [Pro p r i e t ary I nformat i o n] * [Pro pri e t a r y In forma t io n] * [Pro pri e t ary In format i o n] * [Pro pri etary I n format i o n] * [Prop ri etary In format i o n]

• Uran ium scrap recovery

• Target asse mbl y , l oa din g, in spect i o n , qu ality check in g, verification , packagi ng a nd storage * [Proprietary Information]

• [Proprietary Informat i on]

• Primary proces s offgas treatment

• F iss ion gas retention

• Fee d pr e p aratio n

• Firs t stage r ecovery

• First stage p u rificat i o n pr e p aratio n

• First stage purificatio n

• Seco nd s tage p u rifica ti o n pr eparat i o n

• Seco nd s t age purification

• Fina l purifi ca ti o n a dju s tm e n t

• 99 Mo pr e p ara ti o n fo r s hippin g

• Impure uranium lag s tora ge

• First-cycle uranium recovery

• Second-cycle uranium purification

• Product uranium lag s torag e

• Other s upport (s torage ve s sels , transfer lines, s olid waste hand l ing for resin bed replacement) 13-10

.....

..**.. ... **: ............ .. *****. NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis *. *

• NOATllWEST MEOICAl. ISOTill'(S Table 13-7. Radioisotope Production Facility Preliminary Hazards Analysis Primary Process Nodes and Subprocesses (2 pages) Node no. Node name Subprocesses encompassed in node 5.0.0 Waste handlin g syste m

• Liquid waste storage 6.0.0 7.0.0 8.0.0 9 9 Mo HEPA process

• High do se liquid waste vo lum e re duction Target receipt and disassembly process Ventilation system Natural phenomena, man-made external events, and other facility operations

• Co nd e n sate storage and recycling

• Co nc e ntr ated high dos e liqu id waste storage/preparation

• Low do se liquid waste vo lum e reduction and storage

• Liquid waste solid ific at ion

• So lid waste handling

• Waste e ncap su l a ti on

• TCE solvent reclamation

• Mixed waste accumulation

• Cask receipt and target unloading

• Target Inspection

• Target disassembly

• [Proprietary Information]

• Target disassembly stations

• Gaseous fission product control * [Proprietary Information]

• Empty target hardware handling * (No s ubproc esses identified in PHA. Ventilation sys tem provides casca din g pr essure zones, a common a ir s uppl y sys t e m wi th m akeu p air as necessary , h eat r ecove r y for preconditioning incoming air , a nd HEPA filtration

.)

• Natural phenomena

• Man-made external events

• Chemical storage and preparation areas

• On-site vehicle operation

• General storage, utilities, and maintenance activities

• Laboratory operations

• Hot cell support activities

• Waste storage operations including packaging and shipment molybdenum-99 hi g h-efficiency particulate a ir. PHA TCE preliminary h azards a n a l ys i s. = tric hl oroet h y l ene. Tabl e 13-8 s ho ws a crosswalk that identifie s the applicability ofRPF PHA top-le ve l accident se quence ca tegorie s to the primary process node s. The information in this table i s referenceabl e to Table 13-6 and ultimately s how s the relationship between the PHA proces s node s an d the NUREG-1537 Part I ISG acc ident initi a ting event categories via the PHA top-le ve l accident sce nario categories.

13-11

.: . NWMI ............. ............ '!*.*!:. NOflTMWESTMEOICAllSOT0.-£1 NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Tab l e 13-8. Crosswalk of Radioisotope Production Facility Preliminary Hazards Analysis Process Nodes and Top-Level Accident Sequence Categories PHA Top-Level Accident Sequence Category *-llil

• Target fabrication (Node 1.0.0) ./ ./ ./ Target dissolution (Node 2.0.0) ./ ./ ./ -Molybdenum recovery and ./ ./ ./ purification (Node 3.0.0) Uranium recovery and recycle ./ ./ ./ .. _ (Node 4.0.0) Waste h a ndling sys tem ./ ./ ./ (Node 5.0.0) Target receipt and disassembly

./ ./ -(Node 6.0.0) Ventilation system (Node 7.0.0) ./ ./ ./ Natural phenomena , man-made ./ ./ ./ ./ ./ external events, and other facility operations (Node 8.0.0) Note: The ,/ in a table ce ll indicates that the acc id en t se qu ence ca te go r y applies to the proce ss node. If it does not , the cell i s blank. PHA = preliminary ha zards a n a l ysis. 13.1.3 Preliminary Hazards Analysis Results This section presents the radiological, criticality, and chemical hazards that could result in high or intermediate consequences.

13.1.3.1 Hazard Criteria Methodologies and hazard criteria are identified in Section 13 .1.1. Numerous hazards are present during the handling and processing the materials in the RPF. The target material is fissile LEU consisting of uranium enriched up to 19.95 weight percent (wt%) uranium-235 (235 U). This material presents a criticality accident hazard in the processes that involve high concentrations of uranium. Both 10 CFR 50 and 10 CFR 70 require that accidental nuclear criticalities be prevented using the double-contingency principle, as defined in ANSI/ ANS-8.1 , Nuclea r Criticality Safety in Operations wit h Fissionable Material Out side R eactors. The RPF separates 99 Mo from among the fission products in the irradiated LEU target material.

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

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

The facility can also be impacted by various internal and external man-made and natural phenomena events that ha ve the potential to damage structures, systems, and components (SSC) that control the licensed material , thereby leading to intermediate

-and high-consequence events. 13-12

...... .. NWMI ::.**.-.*.* *.*.* .. *.*.* . ' !*. * . NOITHWEST MEDK:Al tsOTOl'f.S Known and credited safety features for normal operations include: NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The hot cell shielding boundary , credited for shielding workers and the public from direct exposure to radiation (an expected operational hazard)

• The hot cell confinement boundaries, credited with confining fissile and high-dose solids , liquids , and gases, and controlling gaseous release s to the environment Administrative and passive engineered design features that control uranium batch size, volume, geometry and interaction are credited for maintaining critically safe (i.e., subcritical) configurations during normal operations with fissile material.

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

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

All process system nodes were analyzed using a HAZOP approach with special emphasis on criticality , radiological , and chemical safety hazards. Fire safety issues are addressed in every node and addressed generally in Node 8.0.0. Fire safety issues include the explosive hazard associated with hydrogen gas generation via radiolytic decomposition of water in process solutions and due to certain chemical reactions encountered during dissolution proce sses. Most hot cell processing areas contain very few combustible materials , either transient or fixed. The RPF PHA has identified adverse events listed in Table 13-9 through Table 13-16. Adverse events are identified as: Standard industrial events that do not involve licensed material

• Acceptable accident sequences that satisfy performance criteria by being low consequence and/or low frequency Unacceptable accident sequences that require further evaluation via the QRA process An accident sequence number was assigned to each accident initiator that results in the same , or similar, bounding accident sequence result and consequence.

The same accident sequence designator can appear in multiple nodes. (Table 13-5 provide s definitions of accident sequence category designators

.) 13-13

...... ; ... NWMI ..*... ..* .... ..... .. .. .. . * *, * ! . NORTHWEST MEOtcAl ISOTOrf.S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.1.1.1 , 1.1.1.2 , 1.6.1. l , Op e rat or d o ubl e b atc h es A cc id e nt a l c ritic a lity S.C.0 2, F a ilur e of 1.8.1.1 , 1.8.2.1 , a nd 1.8.3.1 a ll o tt e d a m o unt of m a t e r ial i ss ue -T oo mu c h fi ss ile a dmini s trat ive co ntrol o n (fres h U , scra p U, [Prop r i e t ary m ass in o n e l oca ti o n m ass (b a tch lim i t) durin g In fo rm a ti o n], target b a t c h) m ay b eco me c riti c al h a ndling o f fres h U , int o on e l ocat i o n o r cont a in e r scra p U , [Propri e t ary durin g h a ndl i n g In fo rm at ion], a nd t a r ge t s 1.1.1.3 Supplier ships greater than Accidental criticality S.C.01 , Failure of site 20 wt% m u to site issue -Too much m u enrichment limit put into a container or solution vessel , exceeding assumed amounts 1.1.1.6 , 1.1.1.7 , 1.6.1.2, Op e rat or h a ndlin g va riou s Acc id e nt a l c ri t i ca lit y S.C.0 3, F ailur e of 1.6.1.4 , 1.8.1.2 , 1.8.1.3 , c o nt a in e r s of ur a nium or i ss u e -T oo much a dmini s t ra ti ve co ntrol on 1.8.1.6 , 1.8.2.2 , 1.8.2.3, b a t c h es o f uranium uramum m ass m one i nt e racti o n li mit durin g 1.8.3.2, 1.8.3.3, 1.8.3.4 , and c ompon e nt s b ri n gs two l oca ti o n h a ndlin g of fres h U, 1.8.3.5 co nt a in e r s or b a t c h es c l oser scra p U , [P ro pri etary to ge th e r th a n t he a ppro ve d In fo rm a tion], a nd tar ge t s i nt e ra c tion c o n t rol di s tanc e 1.2.1.1 , 1.2.1.11 , 1.2.1.14 , Failure of safe geometry Accidental criticality S.C.04 , Spill of fissile 1.2.1.25 , 1.3.1.1 , 1.3.1.6 , confinement from fissile solution not material from safe 1.3.1.11, 1.3.1.17, 1.4.1.19 , confined in safe geometry system 1.4.1.20 , 1.4.1.21 , 1.4.1.23 , geometry confinement 1.4.2.6 , 1.4.2.10 , 1.4.2.15 , 1.4.3.14 , 1.4.3.26 , 1.4.3.31 , 1.4.4.1 , 1.4.4.6 , 1.4.4.10, 1.4.4.15 , 1.5.1.21 , 1.5.1.23 , 1.5.1.26 , 1.5.2.16 , 1.7.1.1 , 1.7.1.11, 1.7.1.14 , 1.7.1.25 , 1.9.1.1 , 1.9.1.6 , 1.9.1.10, and 1.9.1.15 1.2 .1.2 a nd 1.7 .1.2 Ura nium-co nt a ining s oluti o n Accid e nt a l cr i t ic a lit y S.C.05 , L e ak of fi s sil e l ea ks o ut of safe g e om e tr y from fi ss il e so lution n o t so lution int o h eat in g/ c o nfin e m e n t int o th e c onfin e d in sa fe c o o lin g j a ck et on v e sse l h ea tin g/coolin g ja c keted s p ace ge om e tr y 1.2.1.3, 1.4.3.33 , 1.4.3 .34 , Uranium solution is Accidental criticality S.C.07 , Leak of fissile and 1.7.1.3 transferred via a leak between from fissile solution not solution across auxiliary the process system and the confined in safe system boundary (chilled heater/cooling jackets or coils geometry water or steam) on a tank or in an exchanger 13-14

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

• NOITHWEST MEDtcAl tsOTOPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.2.1.8 , 1.3.1.4 , 1.4.1.1 5 , Fa ilur e of safe geo m e tr y A cc id e n ta l c riti ca lit y S.C.1 9 , F a ilure o f 1.4.2.4 , 1.4.3.1 8 , 1.4.4.4 , dim e n s ion ca u sed b y fro m fi ssi le s olu t i on n ot p ass i ve d esig n fea tur e -1.5.1.2 0 , 1.5.2.11 , 1.7.1.8 , co nfi g urati on m a n age m e nt c onfin e d i n safe C ompon e nt safe a nd 1.9.1.4 (in sta ll at i o n , m a int e n a n ce), geo m etry geo m e tr y d i m e n s i o n int e rn a l or exte rn a l eve n t 1.2.1.12 , 1.3.1.9 , 1.4.2.8 , Tank overflow into proces s Accidental criticality S.C.06 , Overfill of a tank 1.4.4.8 , 1.4.5.4 , 1.7.1.12, and ventilation system issue -Fissile solution or component causing 1.9.1.8 entering a system not fissile solution entering necessarily designed for the process vessel fissile solutions ventilation system 1.3.1.2, 1.4.2.2 , 1.4.4.2 , a nd Uranium pr ec ipit a te o r o th er A cc id e nt a l c riti ca lit y S.C.2 0 , F a ilure o f 1.9.1.20 hi gh u ra nium s o lid s fro m fis s i l e so lution n ot co n ce n tra ti o n limit s -acc umul a t e in safe geometry co n fi n ed to s afe Pr ec ipit a ti o n of ura n i u m vesse l geo m e try a nd in safe ge om et ry t a nk i n terac t ion co nt ro l s w i t hin a ll owa bl e c on ce ntrati o n s 1.2.1.26 , 1.3.1.7 , 1.5.1.3 , and Uranium solution backflows Accidental criticality S.C.08 , Fissile solution 1.5.2.5 into an auxiliary support issue -Fissile solution backflow into an system (water line , purge line, entering a system not auxiliary system at a fill chemical addition line) due to necessarily designed for point boundary various cau s es fissile solutions 1.4.1.6 , 1.4.1.1 2 , a nd 1.4.1.16 Fa ilur e of safe geo m e t ry Acc id en t al cri ti ca l i t y S.C.1 1, F i ssi l e ma t e ri a l co nfin e m e n t du e t o fro m fi ss i le so luti on n o t co nt a min ation of i n a d ve rt e nt tra n sfer to c on fi n e d i n safe co nt a ct or r ege n era ti o n U-b ea rin g so lu t i o n across a geo m etry a qu eo u s was te s tr ea m -b o und a r y in to n o n-favora bl e bo und ary t o un sa f e geo m e try geo m e tr y syste m 1.4.3.1 , 1.4.3.9, 1.4.3.19 , F a ilure of safe geometry Accidental criticality S.C.09 , Fissile material 1.4.3.21 , 1.4.5.9 , and 1.4.5.11 confinement due to from fi s sile solution not contamination of inadvertent transfer to confined in safe evaporator condensate

-U-bearing solution across a geometry boundary to unsafe boundary into non-favorable geometry system geometry 1.6.1.3 Fa ilur e o f safe ge om e tr y A cc iden ta l c riti ca lit y S.C.1 2 , W as h of co nfin e m e n t d ue t o fro m fi ssile so luti on n o t [P ro pri e t ary In fo rm a ti o n] i n a d ve rt e n t tr a n sfer t o co nfin e d i n safe w i t h wro n g r eage nt U-b ea rin g so lu t ion ac ro ss a ge om e tr y co nt a min a tin g was h bound a r y int o n o n-favora bl e s o lution with fi ss il e U; geo m e tr y b o undar y t o un safe geo m e try s ys t e m 1.1. l.l l Dusty surface generated Potential exposure to S.F.01 , Pyrophoric fire during shipping on uranium workers due to airborne in uranium metal pieces s pontaneously ignite s uranium generation due to pyrophoric nature of uramum 13-15

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

• N0<<\1MUT MEOICAl tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summar y for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.2.1.6 , 1.2.1.11 , 1.7.1.6 ,a nd 1.7.1.11 1.4.1.17, l.4.l.21, and 1.4.1.23 Hydrogen buildup in tanks or sys tem , l ea din g to exp lo sive concentrat io ns Fire in process system containing high concentration uranium spreads the uranium Explosion l eadi n g to ra diolo g i cal a nd criticality co n cerns Radiological and criticality issue -Radiological airborne release of uranium and uncontrolled spread of uranium outside safe geometry confinement 1.6.1.6, 1.6.1.9 , a nd 1.6.1.1 2 Air inle akage into th e Accidental cri ti ca lit y 1.6.1.8 1.2.1.l I , 1.2.1.14 , 1.4.1.17 , 1.4.1.19 , 1.4.1.20 , 1.4.1.21 , 1.4.1.23 , 1.4.2.6 , 1.4.3.14 , 1.4.3.26 , 1.4.3.31 , 1.4.3.32 , 1.7 .1.11 , 1.7 .1.14 , a nd 1.9 .1.6 1.2.1.11, 1.2.1.12, 1.2.1.14, 1.2.1.25, 1.3.1.1, 1.3. l.6, 1.3.1.11, 1.3.1.17, 1.4.1.17, 1.4.1.18, 1.4.1.19, 1.4.1.21, 1.4.2.1 , 1.4.2.6 , 1.4.2.8, 1.4.2.l 0 , l.4.2.15 , 1.4.3.14 , 1.4.3.26, 1.4.3.31, 1.4.4.6, 1.4.4.10, 1.4.4.15 , 1.5.1.21, 1.7.1.11 , 1.7.1.14, 1.7.1.2 5 , 1.9.1.1 , 1.9.1.6 , 1.9.l.8, l.9.1.10, and 1.9.1.15 reduction furnace during H 2 issue -Uncontro ll ed purge cycle or H 2 inleakage s pread of uranium into re du ction furnace befor e outside safe geo metr y inerting with n i trogen can l ea d confinement to an exp lo s ive mixture in the pre se n ce of an ignition source Loss of cooling of exhaust or fire in the reduction furnace leads to high temperatures in downstream ventilation component and accelerated release of adsorb radionuclides High concentration uranium so lution is spraye d from the sys tem , causing hi g h a irborn e ra dio act i vity High concentration uranium solution is spilled from the system 13-16 Radiological issue -Potential accelerated release of high-dose radionuclides to the stack (worker and public exposure)

R a diological release of uranium so lution spray that remains s u spe nd ed in the a ir , expos in g wo rk ers or t h e publi c Potential radiological exposure to workers from contaminated solution S.F.02, Accumulation of flammable gas in tanks or sys t e ms S.F.07, Fire in nitrate extraction system -flammable solvent with uranium S.F.03 , H ydrogen d eto nation in reduction furnace S.F.04, High temperature damage to process ventilation system due to loss of cooling in reduction furnace exhaust or fire in reduction furnace S.R.03 , Solution spray re l ease pot e ntially c reating airborne uranium above DAC limits S.R.O 1, contaminated solution spill

.; ... .. NWMI ...... ..* **: ........ *.* * ! . NOlmfWUT MUHCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-9. Adverse Event Summary for Target Fabrication and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 1.2.1.2 1 , 1.2.1.22, 1.4.5.1 3 , 1.7 .1.2 1 , an d 1.7 .1.22 1.3.1.16 and 1.4.1.24 1.8.3.7 uran i u m-235. B o ilin g or carryover of stea m or hi g h co n ce nt ra ti o n wa t e r va p o r into t h e prim a r y ve n t il a tion sys t e m , affecti n g r ete n t i on b e d s fro m p a rt ia l or co mpl ete l oss of coo l i n g system c a p a biliti es High-dose solution (failure of the uranium recovery proce ss) resu l ts in high-dose radionuclide s entering the first s tage of proc e ssing uranium [Proprietary Inform a tion] (eventually handled by the worker) Loa din g li mjts a r e n ot a dh ere d t o b y th e opera t o rs o r t h e c l os ur e requ i re m e n ts a r e not sa ti sfie d, a nd th e cask d oes n o t pro v id e th e co nt a inm e nt o r s hi e ldin g fu n c ti o n th a t i t is d es i gned t o perfo rm 23s u DAC H z IR U de r ived a i r concentratio

n. hy d roge n gas. io din e r e m ova l un it. 13-17 R a di o l og i cal r e l ease fro m rete n tion b e d s Potentially high radiological exposure to workers Hi g h-dose to workers or t h e p ublic fro m imp roper ly s hi e ld e d cas k S.R.04 , L iqu i d e nt e r s p rocess vesse l ve nt i l a ti on sys t em d a m agi n g IR U or r ete nti on b eds r e l eas in g r eta in e d ra d io nucl i d es S.R.05, High-dose solution enters the UN blending and storage tank S.R.28 , Ta r get o r was t e s hip pi n g cask n ot lo a d ed or sec ur e d accor din g t o pro ce dur e, l ea din g t o p e r s onn e l ex p os ur e PHA u UN proce ss hazards a n a l ysis. uran i um. u ra n y l n i trate.

..... NWMI ...... ..* ... .....

• 0 NORTHWEST llEDK:Al tsOTOPO NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Tab l e 13-10. Adverse Event Summary for Target Dissolution and Identification of Acci d ent Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 2.1.1.1 , 2.1.1.11 , Fai lu re of safe geo m etry Acc id e nt al c r iti ca l ity fro m S.C.04 , Fa ilur e of 2.1.1.1 3, 2.1.1.1 7, co n fi n e m e nt fi ss il e so luti on n o t co nfin e d in co n fi n e m e nt in safe 2.2.1.5, 2.2.1.1 2 , safe geo m e tr y geo m e t ry; s pill of fiss il e 2.2.1.1 5 , 2.3.6.5 , mater i a l so lu t i o n 2.3.6.1 2, a nd 2.3.6.1 3 2.1.1.2 Uranium-containing Accidental criticality from S.C.05 , Leak of fissile solution leaks out of safe fissile solution not confined in s olution in to geometry confinement into safe geometry heating/cooling jacket the heating/coolingjacketed on vessel space 2.1.1.3 Ura n i um so lution i s Acci d e nt a l c riti ca lit y fr om S.C.07 , L ea k of fi ss il e tra n sfe rr e d v ia a l eak fissi l e so lu t i o n not confi n e d i n so lu t i o n ac ro ss a u x ili ary b etwee n th e pro cess sys t e m safe geo m e t ry sys t e m bound ary a nd t h e h ea t e r/c o o l ing (c hill ed w ate r o r s t ea m) jackets o r coi ls o n a tank o r in a n exc h a n ge r 2.1.1.8 , 2.2.1.11 , and Failure of s afe geometry Accidental criticality from S.C.19 , Failure of 2.3.6.11 dimension fissile solution not confined in passive design feature; safe geometry component safe-geometry dimension 2.1.1.1 2 , 2.1.1.1 5 , a nd Fa il ure of safe-geome t ry Acc id e nt a l c rit ica l ity fro m S.C.1 3 , F i ss il e so luti o n 2.3.1.4 co n fi n e m e nt fiss il e so luti on n o t co n fi n e d in e n te rs t he N O x sc rubb e r safe geo m et r y w h ere hi gh u ran iu m so luti on i s n o t inte nd e d 2.1.1.14 and 2.3.4.14 Tank overflow into proce s s Accidental criticality i s sue -S.C.06 , System ventilation system Fis s ile solution entering a o v erflow to process system not necessarily designed ventilation involving for fissile solutions fissile material 2.3.4.11 U r a nium e nt e r s c a rb o n Acci d e nt a l c riti c alit y fr o m S.C.2 4 , Build-up of hi g h r e t e n t ion b ed d rye r w h ere it fissi l e m a t e ri a l or s oluti o n no t ura nium parti c ul a t e in ca n mix wi th co nd e n sa t e to co n fi n e d in sa f e geomet r y t h e ca rbon re t ention b e d form a fiss ile s oluti o n dryer sys t e m 2.1.1.33 and 2.1.1.34 Uranium solution backflow s Accidental criticality and high S.C.08, System into an auxiliary support radiological dose -High-dose backflow into auxiliary system (water line , purge and fissile sol u tion entering a support system line , chemical addition line) system not necessarily designed due to variou s causes for fissile solutions that exist outside of hot cell walls 13-18

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

• 0 NO<<TMWEST MEDICAl ISOTOHS NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 2.1.1.1 8 , 2.3.1.2 1 , H y dro ge n build-up in t a nk s Ex plo s ion l ea ding to S.F.0 2, A cc umulation of 2.3.2.2 1, 2.3.3.2 4 , or sys t em l ea din g to ra di o lo g i ca l a nd cr it i ca li ty fl a mm a bl e gas in t a nk s 2.3.4.3, a nd 2.3.5.5 ex plo s i ve con ce ntration s co n ce rn s or sys t e m s 2.3.4.20, 2.3.5.2 , A fire develops through Radiological is s ue -Potential S.F.05 , Fire in a carbon 2.3.5.6 , 2.3.5.10 , and exothermic reaction to accelerated rele a se of high-dose retention bed 2.3.5.13 contaminants in the carbon radionuclide s to the stack retention bed and rapidly (worker and public expo s ure) releases accumulated gaseous high-do s e radionuclides 2.1.1.1, 2.1.1.2 , H ig h-do se a n d/or hi g h-P ote nti a l ra di o l ogica l ex p os ur e S.R.O I , R a d io l og i ca l 2.1.1.1I , 2.1.1.13 , co n ce ntrati on u ra nium to wo rk e r s fr o m hi g h-do se r e l ease in th e fo rm of a 2.1.1.17 , 2.2.1.5 , so lut ion is s pill e d fr o m t h e and/o r hi gh u ranj um-li q uid s pill of hi g h-d ose 2.2.1.1 2 , 2.2.1.15 , syste m co n ta min a t e d so lution a n d/or hi g h ur a nium 2.3.6.5 , 2.3.6.1 2 , a nd co n ce nt ra ti on so luti o n 2.3.6.1 3 2.1.1.3 High-dose solution is Radiological exposure to S.R.13 , High-dose transferred via a leak workers and the public from solution leaks to chilled between the process system high-radiological dose not water or steam and the heater/cooling contained in the hot cell condensate system jackets or coils on a tank or containment or confinement in an exchanger boundary 2.1.1.1I ,2.1.1.1 7 , S pill l ea din g to s p ray-typ e R a di o lo g i c al d ose fro m S.R.0 3 , S p ray o f produ c t 2.2.1.1 5 , a nd 2.3.6.1 3 re l ease, ca u s ing a irb o rn e airbo rne s p ray of p ro du ct so luti on i n h ot ce ll area radioac ti v it y ab ove DA C so lution from sys t e m s l im it s fo r ex p os ur e 2.1.1.23, 2.1.1.26 , Carryover of rugh vapor High airborne radionuclide S.R.04 , Carryover of 2.1.1.27 , 2.3.4.1 , content gases or entrance of release , affecting workers and heavy vapor or solution 2.3.4.12 , and 2.3.4.1 7 solutions into the process the public into the process venti l ation header can cause ventilation header poor performance of the causes downstream retention bed materials and failure of retention bed , release radionuclides releasing radionuclides 13-19

...... ; ... NWMI ..*... ..* **: ..... .... .. . ' * ! ' NOITNWUT MEOtCAl. t$01VH NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 2.3.1.1 7, 2.3.1.22 , A s p i ll o flo w-d ose P oten ti a l ra di o l ogi c al d ose to S.R.02, S pill of l ow-2.3.1.2 4 , 2.3.2.17 , co nd ensa t e occ u rs for a worke rs a nd the p ubli c fro m dose co nd e n sate 2.3.2.22, 2.3.2.24 , varie t y of reaso n s fro m t he s p i ll e d liquid 2.3.3.8, 2.3.3.20 , co n fi n e m e nt ta n ks or vesse l s 2.3.3.27 , 2.3.4.3 , 2.3.4.5 , 2.3.4.6 , a nd 2.3.4.8 2.3.3.l , 2.3.3.2 , 2.3.3.3 , High flows through the IRU Potential radiological dose to S.R.06 , High flow 2.3.3.6 , 2.3.3.12 , incre a ses the relea s e of the workers and the public from through IRU causes 2.3.3.13 , 2.3.3.16 , retained iodine and iodine above regulatory limit s premature release of 2.3.3.17' 2.3.3.23 , increases the high-dose high-dose iodine ga s 2.3.4.13 , 2 , 3.5.l , concentration of this ga s in 2.3.5.6 , 2.3.5.8 , and the st a ck 2.3.5.10 2.3.3.1 5 and 2.3.5.8 Low tem p erat ur e s in t h e Potentia l ra di o l ogical d ose to S.R.07 , Loss of I R U in l e t gas s t ream d rives wor k ers a nd th e publi c fro m tem p era tu re contro l o n re l ease of i o d i n e from th e i o d ine a b ove r eg ul atory limit s the IR U l ea d s to uni t pre m a tur e re l ease of h i g h-d ose io d i n e 2.3.3.22 and 2.3.5.8 Liquid and water vapor in Potential radiological dose to S.R.04 , Liquid/high the IRU inlet gas s tream work e rs and the public from v apor in the IRU leads drives release of iodine from iodine above regulatory limits to premature release of the unit high-dose iodin e 2.3.4.4 , 2.3.4.5 , an d Loss of vac uum pu m p s i n P otentia l ra di o l ogical d ose t o S.R.08 , L oss of vac uum 2.3.4.6 t h e disso l ve r offgas workers a nd the p u b li c fr o m p ump s treatme nt s ys t em leads t o s pi ll ed liqu i d press ur e buildup i n s ide t h e process a nd p o t entia l re l ease ofradion uclid es from t h e syste m up s tr ea m 2.3.4.l l Uncontrolled lo ss o f media Pot e ntial radiological do s e to S.R.09 , Lo ss ofIRU and contact with a liquid workers and the public from media to downstream with potential for premature iodine abo v e regulatory limits dryer relea s e of the adsorbed iodine 13-20

..

...... ..* ... ....... !.* * *,*

• NOmfWEST MEOtCAl ISOTOPlS NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-10. Adverse Event Summary for Target Dissolution and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 2.3.3.28, 2.3.4.19 , 2.3.5.9 , 2.3.4.15, a nd 2.3.5.11 2.3.4.16, 2.3.5.5 , and 2.3.5.12 2.1.1.33 a nd 2.1.1.34 Us in g th e wron g r e tention m e dia (IR U or c a rbon bed s) or u s in g sa tu ra t ed m e dia w i t h pot e nti a l fo r in e ff e ctiv e a d so rption of high-do se gaseo u s radionuclid es An event causes damage to the structure holding the retention media , and retention media is released to an uncontrolled environment H ig h-do se proc ess s olution b ac kfl o w s into a n a u x ili ary s upp o rt sys t e m (wa t e r lin e , pur ge lin e, ch e mi ca l a ddition lin e) due t o v ariou s ca u ses D AC IR U d er i ve d a ir co n centra t io n. iodine r e m ova l un i t. Pot e nti a l radiol o gical do se t o w ork e r s and th e publi c from ra di o nu c lide s ab ove r eg ul a to ry l i mi ts Potential radiological dose to workers and the public from radionuclides above regulatory limits H ig h radiolo g ic al d ose -Hi g h d ose proc ess so lu t i o n e nt e rs a sys t e m th at ex its o ut s ide o f th e h ot ce ll wall s n itroge n ox id e. S.R. l 0 , Wron g r e t e ntion m e dia a dd ed t o b e d or sa tu ra t ed r e t e ntion m e di a S.R.09 , Breach of an IRU or retention bed resulting in release of the media S.R.11 , Sys t e m b ac kflo w of hi g h-d ose so luti o n into a n a u x ili ary s upp o rt sys t e m a nd out s id e th e hot c e ll boundar y NO x PHA p roce s s h azar d s a n a l y s i s. 13-21

.; ... ;. NWMI ...... ..* .... .... .. .. .. * !*. * ! . NC*THWUT MlDtCAl. ISOTOPES NWMl-2 01 3-021, Rev. 1 Chapter 13.0 -A ccident Analysis Table 13-11. Adverse Event Summary for Molybdenum Recovery and Identification of Accident Sequences Needing Further Eva l uation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 3.3.1.2 4 Hi g h er ra di a tion do se du e to Hi g her l oca li ze d do se in N I A hold-up acc umulation or hot c e ll b o und a r y tran s i e nt b a t c h di ffe r e n ces (uno cc upi ed b y wo rk e r s) 3.2.3.7 , 3.2.4.7 , 3.4.3.7 , 3.4.4.7 , Chemical spills of Standard industrial N I A 3.6.3.7 , and 3.6.4.7 nonradiologically accident -Chemical contaminated bulk exposure (not involving chemicals licensed material) to workers 3.7.4.5 a nd 3.7.4.6 Dropp e d cas k or cas k St a ndard indu s tri a l N I A c ompon e nt durin g l oa din g ac cid e nt -W o rk e r inju ry or h a ndlin g 3.7.4.2 , 3.7.5.2 , and 3.7.5.3 Mo product is exposed with Potential dose to the N I A -Addressed by no shielding as the result of public and/or environment DOT packaging and an accident , shipment due to release or transportation mishap , or shipment mishandling of Mo regulations mishandling after lea v ing product during transit (10 CFR 713) the site 3.1.1.9 , 3.1.1.14 , 3.1.1.23, 3.1.2.4 , Fa ilur e of safe-ge om e t ry A cc id e nt a l c riti ca lit y fro m S.C.04 , F a ilur e o f 3.1.2.7, 3.1.2.1 3 , 3.1.2.1 6 , c onfin e m e nt fi ss il e so lution not co nfin e m e nt in s a fe 3.1.2.1 7, 3.2.1.6, 3.2.1.10 , confin e d in safe geo m e tr y ge om e tr y; s pill of 3.2.1.2 0 , 3.2.1.22, 3.2.1.23 , fi ss il e m a t e ri a l 3.2.2.9, 3.2.2.13 , 3.2.3.6 , 3.2.3.8, so luti o n 3.2.5.9, 3.2.5.14 , 3.2.5.23, 3.8.1.9 , 3.8.1.1 3 , a nd 3.8.1.22 3.1.1.4 , 3.1.1.16 , 3.2.5.4 , 3.2.5.16 , Tank overflow into process Accidental criticality issue S.C.06, System and 3.8.1.4 ventilation system -Fissile solution entering overflow to process a system not necessarily venti l ation involving designed for fissile fissile material solutions 3.1.1.23, 3.2.1.23, 3.2.5.23, an d Ura n i um so lution is A c cid e nt a l c ritic a lit y fro m S.C.07 , L ea k of 3.8.1.22 tra n sferre d v ia a l eak fiss il e so luti o n not fi ss il e so lution b e t ween t h e pro cess syste m co n fi n e d in sa f e geo m etry ac ro ss a u x ili a r y an d th e h ea t e r/coo l ing system b o und a r y jac k e t s or c oil s o n a t a nk or (c hill e d wa t e r or in a n exc h a n ger s t eam) 3.2.1.4 , 3.2.1.5 , 3.2.2.3, 3.2.2.4 , Fissile product solution Criticality safety issue -S.C.10, Inadvertent 3.2.2.5 , 3.2.3.6, and 3.2.4.6 transferred to a system not Fissile solution directed to transfer of solution designed for safe-geometry a system not intended for to a system not confinement fissi l e solution designed for fissi l e so l utions 13-22

..... .. NWMI ...... ... ... .......... *

• 0 NORTHWEST MEDfCAl lSOTOJU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-11. Adverse Event Summar y for Mol y bdenum Recover y and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 3.1.1.1 3, 3.1.2.9 , 3.2.1.1 5 , 3.2.5.1 3, a nd 3.8.1.1 2 3.1.1.25 , 3.2.5.25 , 3.3.1.25 , 3.5.1.25 , and 3.8.1.24 3.7.1.1 , 3.7.l.2, 3.7.2.1, 3.7.3.1 , 3.7.3.2, a nd 3.7.4.1 3.1.1.9 , 3.1.1.14 , 3.1.1.23 , 3.1.2.7 , 3.1.2.13 , 3.1.2.16 , 3.1.2.17 , 3.2.1.6 , 3.2.1.20 , 3.2.1.22 , 3.2.1.23 , 3.2.2.7 , 3.2.2.9 , 3.2.2.13 , 3.2.3.6 , 3.2.3.8 , 3.2.3.10 , 3.2.4.10 , 3.2.5.9 , 3.2.5.14 , 3.2.5.23, 3.3.1.9 , 3.3.1.14 , 3.3.1.18 , 3.3.1.22 , 3.3.1.23 , 3.3.2.4 , 3.3.2.7 , 3.3.2.13 , 3.3.2.16 , 3.3.2.17, 3.4.1.5 , 3.4.1.9 , 3.4.1.19, 3.4.1.21, 3.4.1.22 , 3.4.2.6, 3.4.2.7' 3.4.2.12 , 3.4.3.6, 3.4.3.8 , 3.4.3.10 , 3.4.3.14 , 3.4.4.6 , 3.4.4.10 , 3.4.4.14 , 3.5.1.9 , 3.5.1.14, 3.5.1.16 , 3.5.1.23 , 3.5.2.4 , 3.5.2.7 , 3.5.2.13 , 3.5.2.16 , 3.5.2.17 , 3.6.1.5 , 3.6.1.6 , 3.6.1.10 , 3.6.1.20, 3.6.1.20 , 3.6.1.23 , 3.6.2.7 , 3.6.2.9 , 3.6.2.13 , 3.6.3.8 , 3.6.3.10, 3.6.3.14, 3.6.4.10 , 3.6.4.14 , 3.8.1.9 , 3.8.1.13, and 3.8.1.22 F a ilur e of safe-geo m e t ry d i m e n sion Hydrogen buildup in tanks or system , leading to explosive concentration s Ope ra t o r s pill s Mo p ro du ct s oluti o n durin g r e m o t e h a ndl ing o p era ti o n s Spill of product s olution in the hot cell area 13-23 Acc id e nt a l c riti ca lit y fro m fiss il e so luti o n not co n fi n e d in safe geo m e tr y S.C.1 9 , Fa ilur e of p ass i ve d es i g n fea tur e; co mp o n e n t safe-geo m e t ry d i m e n sio n Explosion leading to S.F.02 , radiological and criticality Accumulation of concerns R a diol og ic a l s pi II o f hi g hd ose M o so luti o n Radiological dose from spill of product solution from systems flammable gas in tanks or systems S.R.O I , R a diol og i ca l s pill of M o pr o du c t d urin g re m o t e h a ndlin g S.R.01 , Spill of product solution in hot cell area

.. ... .. NWMI ...... ..* **.* .*...*. *.* * *.

• NORTHWEST MEDtCAL ISOTDPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-11. Adverse Event Summary for Molybdenum Reco very and Identification of Accident Sequences Needing Further Evaluation (3 pages) Bounding accident PHA item numbers description Consequence Accident sequence 3.1.1.9 , 3.2.1.10 , 3.2.1.22 , 3.2.2.7, 3.2.2.9, 3.2.3.8 , 3.2.3.10 , 3.2.4.10, 3.2.5.9, 3.3.1.9, 3.3.1.1 8 , 3.3.1.22, 3.3.2.7, 3.4.1.10 , 3.4.1.22, 3.4.2.7, 3.4.3.8, 3.5.1.9 , 3.5.1.23 , 3.6.1.10, 3.6.2.7, 3.6.3.8, a nd 3.8.1.9 3.1.1.7, 3.1.1.22 , 3.2.5.7, 3.2.5.22, 3.3.1.4 , 3.3.1.7, 3.3.1.16, 3.5.1.4, 3.5.1.7, 3.5.1.16, 3.5.1.22, 3.8.1.7, and 3.8.1.13 3.7.4.3 3.3.1.23, 3.3.2.16, 3.4.1.22, 3.5.1.23, and 3.6.1.23 Spill 1eading to spray-type rele ase , causing airborne radioactivity above DA C limits for ex posure Boiling or carryover of steam or high-concentration water vapor into the primary process offgas ventilation system affecting retention beds with partial or complete loss of cooling system capabilities A Mo product cask i s remov ed from the hot cell boundary with improp er s hield plug in s tallation High-dose radionuclide solution leaks through an interface between the process system and a heating/cooling jacket coil into a secondary system (e.g., chilled water or steam condensate) releasing radionuclides to workers, the public , and environment Radiological dose from airborne spray of product so luti on from systems Radiolog ical release from retention beds Potenti a l dos e to workers, the public , and/or environment due to relea se or mishandling of Mo product during tran sit High-dose radionuclide solution that leaks to the environment through another system to expose workers or the public S.R.03 , Spray of product solution in hot ce ll area S.R.04, Loss of cooling, leading to liquid or steam carryover into the primary offgas treatment train S.R.12 , Mo product is released during s hipment S.R.13, High dose radionuclide containing solution leak s to chilled water or steam condensate system

• I 0 CF R 71, " Packagin g and Transportation of Radioacti ve Material," Cod e of Federal R egulations , Office of th e Federal Register , as amended. DAC derived air concentration.

N I A not applicable. DOT U.S. Departm ent of Tran s p ortation.

PHA process ha zards analys is. Mo molybdenum.

13-24

.**.*.*. ..... .. NWMI ........ *.* * *.

• 0 NOITNWEST MEDltAl tsOTOfl(S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.1.1.4 , 4.1.1.18 , 4.2.1.4 , 4.2.1.6 , 4.2.1.17' 4.2.1.18, 4.2.3.6 , 4.2.8.4 , 4.2.8.1 8 , 4.2.10.4 , 4.3.1.4 , 4.3.1.6 , 4.3.1.1 8, 4.3.1.19 , 4.3.3.6 , 4.3.8.4 , 4.3.8.18 , 4.3.10.4 , 4.4.1.4 , 4.4.1.17' 4.5.1.4, 4.5.1.17' 4.5.2.4 , 4.5.2.17 , 4.5.3.4 , a nd 4.5.3.14 4.1.1.6, 4.2.1.7, 4.2.2.4 , 4.2.3.4 , 4.2.3.7, 4.2.3.8, 4.2.8.7 , 4.3.1.7 , 4.3.2.4, 4.3.3.4 , 4.3.3.7, 4.3.3.8 , 4.3.8.7, 4.4.1.6, 4.5.2.6, and 4.5.3.6 4.1.1.14 , 4.2.1.14 , 4.2.3.1 6 , 4.2.8.15 , 4.3.1.15 , 4.3.3.1 6 , 4.3.8.15 , 4.3.9.20, 4.4.1.14 , 4.5.1.14 , 4.5.2.14 , and 4.5.3.11 4.1.1.8, 4.1.1.9 , 4.1.1.12 , 4.1.1.13 , 4.1.1.16, 4.2.1.9, 4.2.1.13 , 4.2.5.11 , 4.2.8.10 , 4.2.8.13 , 4.2.8.14 , 4.2.8.17 , 4.2.9.18 , 4.3.1.10, 4.3.1.1l , 4.3.1.l4 , 4.3.1.17, 4.3.1.18 , 4.3.5.l l , 4.2.8.10, 4.3.8.13, 4.3.8.14 , 4.3.8.17 , 4.3.9.18 , 4.4.1.8 , 4.4.1.9 , 4.4.1.12, 4.4.1.13 , 4.4.1.16 , 4.5.1.16 , 4.5.2.8, 4.5.2.9 , 4.5.2.12 , 4.5.2.13, and 4.5.2.16 4.1.1.10, 4.1.1.15 , 4.1.1.23, 4.2.1.11 , 4.2.1.15 , 4.2.1.24, 4.2.2.1 , 4.2.3.11 , 4.2.3.13 , 4.2.3.1 8 , 4.2.3.22 , 4.2.3.23 , 4.2.3.24 , 4.2.4.10 , 4.2.5.10 , 4.2.7.8 , 4.2.8.11 , 4.2.8.16 , 4.2.8.23 , 4.2.9.1 6 , 4.2.9.29 , 4.2.9.34 , 4.3.1.12 , 4.3.1.16 , 4.3.1.25 , 4.3.2.1 , 4.3.3.11 , 4.3.3.13 , 4.3.3.18 , 4.3.3.22, 4.3.3.23 , 4.3.3.24 , 4.3.4.10 , 4.3.5.10 , 4.3.7.8, 4.3.8.11 , 4.3.8.16 , 4.3.8.23, 4.3.9.16, 4.3.9.28 , 4.3.9.3 4 , 4.4.1.10, 4.4.1.15 , 4.4.1.23, 4.5.1.23, 4.5.2. l 0, 4.5.2.15 , 4.5.2.23, 4.5.3.8 , 4.5.3.12, an d 4.5.3.19 T a nk overflow into Accidental cr iticalit y S.C.06 , System overflow pro cess ve ntilation syste m issue -F i ssi l e so luti on to proce ss vent il a tion Uranium solution back:flows into an auxiliary support system (water line, purge line , chemical addition line) due to various causes Failure of safe geometry dim e n sio n caused by config u ra tion m a n age m e nt (in stallat ion , mainten a nc e) o r externa l event Uranium precipitate or other high uranium solids accumulate in geometry vessel Failure of safe-geo m etry confinement du e t o sp ill of uranium so lution from the system 13-25 enters a system not in vo l v in g fissile m ater i a l n ecessarily d es i g n ed for fissile so luti ons Accidental criticality S.C.08, System backflow issue -Fissile solution into auxiliary support enters a system not system necessarily designed for fissile solutions Acci d enta l criticality from fissile so luti o n not confined in safe geometry Accidental criticality from fissile solution not confined to safe geometry and interaction controls within allowable concentrations Accidental criticality from fi ssi l e solution not co n fi n ed in safe geo m etry S.C.1 9 , Failure of passive d es i g n feature; co mponent geo m e tr y dimension S.C.20, Failure of concentration limit s S.C.04 , Failure of confinement in safe geo m e tr y; s pill of fissile material so lu tion

.; .. ;: NWMI ...... ..* ... ........ *. ' * ! ' lfOITMW£n MEDICAl tsOTOPH NWM l-201 3-0 21, Rev. 1 Chapter 13.0 -A ccident Analysis Table 13-12. Ad v erse E v ent Summar y for Uranium Recover y and Identification of Accident Sequences N eeding F u rther Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.2.3.21 , 4.2.4.11 , 4.2.6.12 , Failure of safe-geometry Accidental criticality S.C.14 , Failure of 4.3.3.2 1 , 4.3.4.11 , and 4.3.6.1 2 confinement du e to from fissile s olution confinement in s afe inadvertent tran s fer to not confined i n safe ge ometry; tran s fer of U-bearin g re si n to the U g eometry U-bearing resin to U IX IX waste collection tank s w a s te collection tank s through a broken ret e ntion e lemen t 4.2.5.5 , 4.3.1.9, 4.3.5.5 , a nd Fa ilur e of safe-geo m etry Acci d ental cr i tica lit y S.C.14 , Fa ilur e of 4.5.1.5 co nfin eme n t du e to fro m fissi l e so lu t i o n co n fi n e m e n t i n safe in a d vertent tra n sfer t o n ot co n fi n e d in safe geo m et r y; transfer o f U-b earing so lu t i o n to the geo m e t ry U-bea rin g so l utio n to U I X waste co ll ec t ion U IX was t e co ll ect i o n ta nk s tanks 4.2.7.7 , 4.3.7.7, a nd 4.5.3.10 Inadv e rten t transfer of high Accidental criticality S.C.15 , Too high of uranium-c oncentration too high of uranium uranium ma ss in s pent solution or r e sin s to s pent m ass in waste s tre am r es in wa s te s tream r es in tank s 4.2.9.10 , 4.2.9.19 , 4.2.9.2 1 , Ura niu m is i n a d verte n t l y Acc id e nt a l critica l ity S.C.09 , Carryove r of 4.2.9.23 , 4.2.10.l 0, 4.2.1 0.1 2 , ca rri e d ove r fro m th e fro m fissi l e so luti o n uran ium t o t he co nd e n se r 4.3.9.10 , 4.3.9.19, 4.3.9.2 1 , co n centrato r (1 o r 2) to t h e n o t co n fined in safe or co nd e n sate tan k s 4.3.9.23 , 4.3.1 0.10 , a n d 4.3.10.1 2 co nd ense r a nd geo m e t ry s ub sequen tl y , th e co nd e n ser co nd e n sate co ll ectio n ta nk s 4.2.9.36 and 4.3.9.36 Uranium s olution i s Accidental criticality S.C.07 , Uranium-tran s ferred v ia a lea k from fissile s olution c ont a ining solution leak s b e t w een the proce ss not confined in safe t o chilled water or s te a m s y s tem a nd h e ater/cooling g eometry conden s ate system jacket s or coil s on a tank or in an exchanger 4.1.1.8 , 4.1.1.22 , 4.2.1.9 , 4.2.1.17 , Carryove r of hi g h-v a por Hi g h a ir bo rn e S.R.04 , Carryove r of 4.2.1.23, 4.2.9.11 , 4.2.9.1 4 , co n tent gases o r e n trance ra di o nu clide r e l ease , h eavy va p o r or so lu t i o n 4.2.9.1 7 , 4.2.9.23 , 4.2.9.30 , of so l u t io n s int o th e affec tin g wo rk e rs a n d i n to t h e p rocess 4.2.9.32 , 4.2.10.14 , 4.3.1.l 0 , p rocess ve n t il at i on h eader th e publi c ve n t il ation h eader ca u ses 4.3.1.1 8 , 4.3.1.2 4 , 4.3.9.11 , ca n cause p oo r downstream fai l ure of 4.3.9.1 4 , 4.3.9.17, 4.3.9.23 , p e r formance of t h e retent i on b e d , re l eas i ng 4.3.9.30 , 4.3.9.3 2, 4.3.10.1 4 , r etention be d material s ra d io nu c l i d es 4.4.1.8, 4.4.1.22 , 4.5. l .9, 4.5.1.22, and r e l ease ra di o nu c l ides a nd 4.5.2.8 13-26

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

• NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation ( 4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4. l.1.10, 4.1.1.15, 4. l.1.23 , 4.2.1.11 , 4.2.1. I 5 , 4.2. I .24 , 4.2.2. I, 4.2.2.4, 4.2.3.1I,4.2.3.13 , 4.2.3.18 , 4.2.3.22, 4.2.3.23 , 4.2.3.24 , 4.2.4.10 , 4.2.5.10, 4.2.6. I I , 4.2.7.8 , 4.2.8.11, 4.2.8. I 6 , 4.2.8.23 , 4.2.9.16 , 4.2.9.28 , 4.2.9.34 , 4.3.1.12, 4.3.1.16 , 4.3.1.25 , 4.3.2. l, 4.3.2.4 , 4.3.3.11, 4.3.3.13 , 4.3.3. J 8, 4.3.3.22 , 4.3.3.23 , 4.3.3.24 , 4.3.4.10 , 4.3.5.10, 4.3.6.J I , 4.3.7.8 , 4.3.8.11 , 4.3.8.16, 4.3.8.23 , 4.3.9.16, 4.3.9.28, 4.3.9.34 , 4.4.1.10, 4.4. I .15 , 4.4.1.23 , 4.5.1.11, 4.5.1.15 , 4.5.1.23, 4.5.2.10 , 4.5.2.15, 4.5.2.23 , 4.5.3.8, 4.5.3.12, and 4.5.3.19 4.2.1.1 2 , 4.2.1.2 4 , 4.2.2.1 , 4.2.3.11 , 4.2.3.13 , 4.2.3.18 , 4.2.3.22, 4.2.3.23 , 4.2.4.10 , 4.2.5. I 0 , 4.2.6.11 , 4.2.8.11 , 4.2.8.1 6 , 4.2.8.23 , 4.2.9.16 , 4.2.9.28 , 4.2.9.34 , 4.2.9.35, 4.3.1.1 2 , 4.3.1.16 , 4.3.1.12 , 4.3.1.25 , 4.3.2.1 , 4.3.3.11 , 4.3.3. 13 , 4.3.3. I 8 , 4.3.3.22 , 4.3.3.23 , 4.3.4. I 0 , 4.3.5.10 , 4.3.6. I I , 4.3.8. I I , 4.3.8.1 6 , 4.3.8.23, 4.3.9.1 6 , 4.3.9.28 , 4.3.9.34, 4.3.9.35 , 4.4.1. I 0 , 4.4. I .1 5 , 4.4.1.23 , 4.5.1. I I , 4.5. I .23, 4.5.2. I 0 , 4.5.2.15 , 4.5.2.23, a nd 4.5.3.19 4.2.9.37 , 4.2.9.36, 4.3.9.36 , and 4.3.9.37 High-dose radionuclide solution is spilled from the system Hi gh-d ose ra di onuclide so lution is s pray e d from the syste m , ca u s in g high a irborn e radioactivity High-dose radionuclide solution leaks through an interface between the process system and a heating/cooling jacket coil into a secondary system (e.g., chilled water or steam condensate), releasing radionuclides to workers , the public , and environment 13-27 Radiological release of high-dose solution with potential to impact workers , the public , or environment R a di o l ogica l relea se of hi g h-d ose s pra y that remain s s u spe nd ed in the a ir , giv in g hi g h d ose to workers o r the public High-dose radionuclide solution that leaks to the environment through another system to expose workers or the public S.R.01, Spill of product solution in hot cell area S.R.03 , Spray of produ ct s olution in hot cell area S.R.13 , High-dose , radionuclide-containing solution leaks to chilled water or steam condensate system

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

• NCMtTHWUT MEOM:Al tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-12. Adverse Event Summary for Uranium Recovery and Identification of Accident Sequences Needing Further Evaluation (4 pages) Bounding accident PHA item numbers description Consequence Accident sequence 4.1.1.2 5 , 4.2.1.26 , 4.2.8.25 , 4.3.1.27 , 4.3.8.25 , 4.4.1.2 5 , 4.5.1.25 , 4.5.2.25 , a nd 4.5.3.2 1 4.1.1.24, 4.2.1.25 , 4.2.8.24, 4.2.10.18 , 4.3.1.26 , 4.3.8.24 , 4.3.10.18 , 4.4.1.24, 4.5.1.24 , 4.5.2.24 , and 4.5.3.20 4.2.4.8 a nd 4.3.4.8 4.2.10.6 and 4.3.10.6 4.2.10.8 , 4.2.1 0.1 1 , 4.2.10.1 7 , 4.3.10.8 , 4.3.l 0.11 , a nd 4.3.1 0.1 7 IROFS IX N I A it e m s r e li e d o n for sa fe t y. i on e x c h a n ge. not a ppli ca bl e. Uranium Recover y Open Item H y dro ge n buildup in t a nk s Ex plo s i on l ea din g to o r sys t em , l ea din g t o ra diolo g i ca l a nd ex plo sive co n ce nt ra ti o n s c ritic a lit y co nc e rn s Higher dose than normal due to double-batching an activity or due to buildup of radionuclides in the system over time Hi gh t empera tur e p re-e luti o n o r rege n erat i o n reage n t ca u ses unkn own i mp act on I X r es in Radiation dose is elevated over normal operational levels , but does not exceed low consequence values for exposure to workers due to s hielding Co n se qu e n ce i s n o t full y und er s too d S.F.0 2, Ac c umul a tion of fl a mm a bl e gas i n t a nk s o r sys t e m s Hot cell shielding is credited as the normal condition , mitigating safety feature for this hazard (adverse condition does not represent failure of the safety function of the IROFS) Te nt a ti ve l y S.R.14 Same a s S.C.08 except Low consequence N I A with low-dose solution resulting in from condenser condensate contaminated s y s tem S pill o r s p ray of l ow-d o s e co nd e n s a t e Low c o n se qu e n ce r es ultin g in co nt a mi na t e d s ur faces a nd d ose to wo rk er b e lo w int e rm e di a t e co n se qu e n ce d ose l eve l s N I A P HA u proc ess h aza rds a n a l ys i s. = u ra n i um. T h e following a d ve r se e ve nt n ee d s t o b e furth er r esea r c h e d. PHA it e m s 4.2.4.8 and 4.3.4.8 po s tulat e h ig h-t e mp e rature 2 molar (M) nitri c acid (HN0 3) s oluti o n bein g used on the ur a nium purifi ca tion ion-e xc han g e (IX) media a s a pre-elut io n rin s e. Th e con se quen c e of the b o undin g acc i dent wa s n o t full y und ers t oo d and n e ed s t o b e furth e r re se arch e d. The likeliho o d w a s i d e ntified as l ow , as there are no g ood ca u ses of the hi g h temperature from th e suppl y t a nk other than a n i mprop e r mi xi n g se quen ce. Thi s ups e t wo uld not ca u se ex tr e mely e l eva t e d tempe ra ture s nor go und e t e ct e d. 13-28

.: .... ; ... NWMI ...*.. ..* **.* ......... *.* . "NORTMWUTMEDICALISOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-13. Adverse Event Summary for Waste Handling and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item Bounding accident numbers description Consequence Accident sequence 5.1.1.13 Hi g h ur a nium co nt ent Solution fro m thi s tank is so lidifi e d S.C.10 , F i ssile so lution in produ ct so luti on i s in a n on-favorab l e geometry proc ess high-d ose waste c oll ect ion dir ec t ed to the hi g h-d ose with potential to result in acci d e nt tanks (a non-fissile so lution waste co ll ect i o n tanks by nucl ear critica l ity at the high boundary) accide nt u ra nium co n ce nt ra tion 5.2.1.13 and High uranium content Solution from this tank is solidified S.C. l 0, Fissile solution is 5.2.2.13 product solution enters the in a non-favorable geometry process directed to the low-dose low-dose waste collection with potential to result in accidental waste collection tank tanks by accident nuclear criticality at the high uranium concentration 5.4.1.1 Hi gh u ran ium co nt e n t The m ass of uranium may excee d a S.C.22, High concentrati on accu mul ates in th e TCE safe m ass an d result in an accide nt a l of uranium in the TCE reclamation eva porat or nucle ar cr i tica lit y without evapora tor residue monito r in g and controls 5.4.2.1 Dissolved uranium The mass of uranium may exceed a S.C.23, High concentration products may accumulate safe mass and result in an accidental in the spent silicone oil in the silicone oil wa s te nuclear criticality without waste stream monitoring and controls 5 .1.1.24 a nd Hydrogen buildup in Ex plo s ion leads to radiological a nd S.F.0 2, Accumulation of 5.1.4.23 tanks or sys tem l ea d s to c riti ca lit y conce rn fl am mabl e gas in t a nks o r ex pl osive co n ce ntrati ons syste m s 5.1.1.4, 5.1.1.16 , Several tank or Radiological release may cause a S.R.04 , High-dose solution 5 .1.4.4 , 5.1.4.15, components vented to the high-dose exposure to workers and from a tank or component and 5.1.4.17 process vessel ventilation the public overflows into the proces s system overflow and send ventilation system , high-dose solution into compromising the retention proc ess ventilation system beds components that exit the hot cell boundary 5.1.1.6 a nd 5.1.4.6 The purge a ir syste m (an R ad iol ogical r e l ease m ay cause a S.R.1 6 , Hi g h-do se so lution a u x ili ary system t h at hi g h-d ose exposure to workers a nd backflows int o the purge a ir or i g i nates o ut side th e hot the public syste m cell b o und a r y) allows hi g h-d ose ra di o nu c lid es to ex it the boundary in an un co nt ro ll ed m a nn er 5.1.1.10 , 5.1.1.14 , Spills from multiple Radiological release may cause a S.R.01, High-dose solution 5.1.1.22 , 5.1.2.26, sources; materials high-dose exposure to workers and spill in the hot cell waste 5.1.2.31, 5.1.4.10 , originating from high-the public handling area 5.1.4.13 , 5.1.4.2 1 , dose process solutions are 5.1.5.16 , 5.1.5.19 , spilled from the system or 5.1.5.20 , 5.3.1.14 , process that normally 5.3.1.17 , and confines them 5.3.1.18 13-29

...... .. NWMI ...... ..* **.* ..... .. .. .. . ' !

• NOlmM£ST MEDICAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-13. Adverse Event Summary for Waste Handling and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item Bounding accident numbers description Consequence Accident sequence 5.1.1.21, 5.1.2.28, Several t anks or and 5.1.4.20 components ve nted to the proce ss vessel vent il ation sys tem evolve hi g h liquid va por concentra tion s, resulting in acce lerat e d hi g h-do se rad ionuclid e release to t h e stack from we tt ed r etentio n bed s 5.1.1.22 , 5.1.2.26 , 5.1.2.31, 5.1.2.32, 5.1.4.10, and 5.1.4.21 5.1.2.9, 5.1.2.18 , 5.1.2.19 , a nd 5.1.2.2 1 5.1.2.33 5.1.5.8 5.5.1.1 Catastrophic failure of a component (high pressure or detonation) leads to rapid release of solution and higher airborne levels A d verse even t s in the concentrator or evaporator sys t e m s l ea d to carryover o f hi g h-d ose so lution into the condenser, resulting in hi g h-do se ra di o nuclid es in the lo w-d ose waste co llecti o n tanks Normally low-dose vapor in the condenser leaks through the boundary into the chilled water system Hi g h-d ose so lution i s in a dv ertent l y mi sfe d into the so lidi ficat ion h o pp er Due to several potential initiators, the payload container or the shipping cask of high-dose encapsulated waste is dropped during transfer from the storage location to the conveyance PHA process h azar d s a nalysis. R a diolo g ic al r e l ease m ay ca u se a hi g h-do se expos ur e to wo rk e rs a nd t h e public Radiological release may cause a high-dose exposure to workers and the public Radiological ex po s ur e l eve l s on th e l o w-do se encapsulated waste m ay excee d int er m e diat e or hi g h co n se qu e n ce l eve l s Radiological release may cause a high-dose exposure to workers and the public Radiolo gical r e le ase m ay cause a hi g h-d ose expos ur e to workers a nd t h e public Radiological issue -Depending on damage from the drop, workers could receive high-dose radiation exposure.

Unshielded package may impact dose rates at the controlled area boundary. S.R.04, High-do se ra di o nuclid e r e le ase due to hi gh vapo r content in exhaust S.R.03, High-dose solution spray events from equipment upsets may cause high airborne radioactivity S.R.17 , Carryover of hi gd ose so lution int o conde n sa t e (a low-d ose waste s tr ea m) S.R.13 , Process vapor from the evaporator leaks across the condenser cooling coils into the chilled water system S.R.18 , Hi g h-do se so lution flows into th e soli dification h opper S.R.32 , Container or cask dropped during transfer TCE tric hloro et h y l ene. 13-30

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

• NOmfWEST MEDtCAl tsOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-14. Adverse Event Summary for Target Receipt and Identification of Accident Sequences Needing Further Evaluation (2 pages) PHA item numbers Bounding accident description Consequence Accident sequence 6.1.2.4, 6.1.2.8, 6.1.2.9 , H a ndlin g d a m age to th e t a r ge t A cc id e nt a l nucl e ar c ritic a li ty S.C.2 1 , T a r get b as ket 6.1.2.11, 6.1.2.14, a nd b aske t fi xe d-int erac tion p ass i ve l ea d s to hi gh d ose t o work ers p ass i ve d es i g n control 6.1.2.1 5 d esign fea tur e l ea ds t o acc id e nt a l a nd pot e nti al d ose t o th e fa ilur e on fi xe d nu clea r c ritic a lit y publi c inte rac ti on s p ac in g 6.1.2.7 , 6.1.2.10 , Too much uranium mass is Accidental nuclear criticality S.C.02 , Operator 6.2.1.1 , 6.2.1.5 , 6.2.2.1 , handled at once either through leads to high dose to workers exceeds batch handlin g 6.2.2.2 , 6.2.2.4 , 6.2.2.5 , operator error or inattention to and potential dose to the limits during target 6.2.3.3 , 6.2.4.1 , 6.2.4.2 , housekeeping public disassembly operation s 6.2.4.4 , 6.2.6.1 , 6.2.6.3 , in the hot cell and 6.2.6.4 6.2.1.6 , 6.2.2.9 , 6.2.3.4 , Op era t or a c cumul a t es mor e A cc id e nt a l nu c l ea r c riti ca l ity S.C.0 3, F ai lur e o f a nd 6.2.6.6 t a r ge ts o r [Propri e t ary l ea d s to hi g h do se to work e rs a dmini s t ra ti ve c ontrol In fo rm at i o n] co n ta in e r s into a nd p o t e n t i a l do se t o the o n int e ra ct i o n limit spec i fic roo m th an a ll owe d a nd publi c durin g h a ndlin g of vio l ates inte ra cti o n co ntrol ta r ge ts a nd i rr a di a t e d [Propri e t ary In fo rm a ti o n] 6.2.1.3, 6.2.1.4 , 6.2.1.5 , Too much uranium in the solid Accidental nuclear criticality S.C.17 , [Proprietary 6.2.2.2 , 6.2.2.4 , 6.2.2.6 , waste container (that is not safe-leads to high dose to workers Information]

residual 6.2.3.1 , 6.2.3.2 , 6.2.3.3 , geometry) entering the solid and potential dose to the determination fails , and 6.2.5.1 , 6.2.5.3 , 6.2.5.4 , waste encapsulation process public used target housings 6.2.5.8, 6.2.6.1, 6.2.6.2 , (where moderator will be added have too much uranium 6.2.6.3 , and 6.2.6.5 in the form of water) in solid waste encapsulation waste stream 6.1.1.5 , a nd 6.1.1.9 Cask in v ol ve d in a n in-t ra n s i t Hi g h do se t o wo rk e r s dur i n g S.R.2 8, Hi g h do se t o acci d e nt o r imp ro p e rly c lo sed r ece ipt in s p ect i o n a nd wo rk e r s durin g p rior to s hipm e n t , l ea din g to o p e nin g act i vities s hipm e nt r ece ipt s t rea min g ra di a ti o n in s p e ction a nd cas k pr e p ara ti o n a cti v iti es du e to dam age d irr a di a t ed t a r ge t c a s k 6.1.1.10 Cask involved in in-transit High dose to workers during S.R.29 , High dose to accident or targets failed during receipt inspection and workers from release of irradiation , leading to excessive opening acti v ities gaseous radionuclides offgasing from damaged targets during cask receipt inspection and preparation for target basket removal 6.1.1.11 , 6.1.1.1 2, Seal b e tw ee n ca s k a nd hot cell Hi g h do se to w ork e r s from S.R.30 , C as k dockin g 6.1.2.1 , 6.1.2.1 3 , a nd d ock in g port fa il s fr om a numb e r s tr ea min g ra d iat ion a n d/or p o rt failur es l ead t o 6.1.2.16 of c a u ses hi g h a irborn e ra dio ac ti v it y hi g h d ose t o wo rk e r s du e to s tr ea min g ra di a ti o n a n d/o r hi g h a irb o rn e ra d ioac ti v it y 13-31

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

• 0 NOITHW£ST MEmCAL tSOTOPU NWMl-2 01 3-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-14. Adverse Event Summary for Target Receipt and Identification of Acci d ent Sequences Needing Further Evaluation (2 pages) PHA item numbers Bounding accident description Consequence Accident sequence 6.1.1.1 Cask involved in a crane High dose to workers during S.R.32 , High dose to movement incident, leading to receipt inspection and workers during streaming radiation opening activitie s shipment receipt inspection and cask preparation acti v ities due to damaged cask i n crane movement incident 6.1.2.3 a nd 6.1.2.5 Imp ro p er h a ndlin g ac ti v iti es Hi g h ex t e rn al d ose t o S.R. I 9, Hi g h ta r ge t res ul t in hi g h ex t e rn a l do se ra t es wo rk e r s b as k e t retrie va l do se th ro u g h th e hot c e ll wa ll wh e n ra t e r e m ovi n g th e t a r ge t b as k e t a nd se tting i t i n th e t arget b as k e t carouse l s hi e ld e d we ll 6.1.2. JO , 6.1.2.15 , [Proprietary Information]

spilled High dose to worker s or the S.R.20 , Radiological 6.2.1.5 , 6.2.2.2 , 6.2.2.4 , or ejected in an uncontrolled public may result from spill of irradiated 6.2.3.3 , 6.2.4.2 , 6.2.5.4 , manner during various target and uncontrolled accumulation of targets in the hot cell 6.2.6. I , and 6.2.6.3 container-handling activities or irradiated

[Proprietary area during target-cutting activities Information]

6.1.2. I 5 Op era tion s r e m o vin g th e t a r ge t Hi g h do se to wo rk e r s du e t o S.R.2 I , Dam age to the b as k et (pot e nti a ll y in a h eavy d eg rad e d s hi e ldin g h o t ce ll wall pro v idin g s hi e l d in g hou s in g) w ith a ho is t s hi e ldin g l ea d s to s trikin g th e wa ll a nd d a m agi n g th e h o t ce ll wa ll s h ie ldin g fun c tion 6.2.4.5 Delays in processing a batch of High dose to workers from S.R.22 , Decay heat remo v ed [Proprietary high airborne radioactivity buildup in unprocessed Information]

results in long-term

[Proprietary heating outside of target housing Information]

removed from targets leads to higher high dose radionuc l ide offgasing 6.2.4.6 a nd 6.2.4.7 Imp ro p e r ve ntin g of the c h a mb e r Hi gh d ose to w ork e r s fr o m S.R.23, O ffgas in g from o r pr e m a tu re o p e n i n g of th e hi g h a irborn e ra dio ac ti v it y irra diat ed t a r ge t va l ve durin g pro cess ing o f a d iss oluti o n ta nk occ ur s pr ev iou s ly a dd e d batch r es ult s in w h e n th e upp e r v alv e i s r e l ease of hi g h-d ose op e n e d ra di o nuclid es to t h e h o t c e ll s p ace 6.2.5.5, 6.2.5.6 , and The seal on the bagless transport High dose to workers from S.R.24 , Bag l ess 6.2.5.7 door fails and l eads to high dose high airborne radioactivi t y tra n sport door fail u re radionuclides escaping the hot cell containment or confinement boundary PHA pro cess h azards ana l ysis. 13-32

... NWMI ::.**.*.*.* ..... .. .. .. , ' *. * !° ." NORTHWEST MEOICAl. ISOTOPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-15. Adverse Event Summary for Ventilation System and Identification of Accident Sequences Needing Further Evaluation PHA item numbers 7.1.1.7 a nd 7.1.1.8 7.1.1.2, 7.1.1.3, and 7.1.1.6 7.1.1.10 and 7.2.1.19 7 .1.1.11 and 7.2.1.20 7.1.1.12 , 7.1.1.14 , and 7.2.1.21 7.2.1.4 , 7.2.1.7, 7.2.1.8 , 7.2.1.9, 7.2.1.1 3, 7.2.1.14, 7.2.1.17 , and 7.2.1.22 7.2.1.12 and 7.2.1.17 Bounding accident description Too much uranium accumulated on the HEPA filter allows an accidental criticality when l eft in the wrong configuration Hydrogen buildup in the ventilation system, due to insufficient flow to sweep it away, leads to fire in the HEPA filters or carbon beds Ignition source causes fire in the carbon bed Overloading of HEPA filter leads to failure and release of accumulated radionuclide particulate The accum ul ated high-dose (and low-d ose) ra dionuclid es retained in th e carbon bed are r e leased through a flow, h eat, or chemica l reaction from the medi a (or the media is released)

Loss of the negative air balance between zones (a confinement feature that prevents migration of radionuclides from areas of high dose and high concentration to areas of low concentration)

Durin g an extended power outage, so me so lution sys tems free ze and ca u se fai lur e of the piping sys t e m , l eading to radiological spi ll s H E PA hi g h-efficiency p a rti c ulat e air. Consequence Accidental nuclear crit icalit y l eads to high dose to workers and potential dose to the public A detonation or deflagration event in the ventilation system rapidly releases retained high-dose radionuclides , causing high airborne radioactivity Fire eve nt in th e ve ntilation system rapid l y releases retained hi g h-d ose radionuclides , causing high airborne radioactivity High dose to workers from high airborne radioactivity High dose to workers from high airborne radioactivity High dose to workers from high airborne radioactivity Hi g h dose to workers from high airborne radioactivity Accident sequence S.C.24, High uranium content on HEP A filters S.F.06, Accumulation of flammable gas in ventilation system components S.F.05 , Fire in th e carbon b e d S.R.25, HEPA filter failure S.R.04, Carbon bed radionuclide retention fa ilur e S.R.26, Failed negative a ir balance from zone to zone or failure to exhaust a radionuclide buildup in an area S.R.27, Extended outage of heat , l ead in g to freezing , pipe fa ilur e, and release of radion uclid es from liquid process systems PH A pro cess h aza rd s analysis.

13-33 PHAitem NW Ml-2 0 13-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) numbers Bounding accident description Consequence Accident sequence 8.2.1.5 L a r g e leak l e ad s to l oca li ze d lo w S t a nd a rd indu s tri a l ha za rd -L oc a li ze d N it roge n s to rage or o xygen l eve l s th a t a d ve r se l y as ph yx i a nt di s tributi o n sys t em l ea k imp a ct w ork e r per for m a nce a nd m ay l ea d to d ea th 8.5.1.l and Operator double-batches allotted Accidental criticality i s sue -Too much S.C.02 , Failure of AC 8.5.1.5 amount of material (fresh U , scrap fissile mass in one location may become on mass (batch limit) U , [Proprietary Information], critical during handling of target batch) into one location or fresh U, scrap U , container during handling [Proprietary Information], and targets 8.5.1.3 a nd Op era tor h a ndlin g va riou s A c c id e nt a l c riti ca lit y i ss u e -T oo mu c h S.C.0 3, Fa ilure o f A C 8.5.1.5 con t a in e r s of u ra nium or b a t c h es uranium m ass in one l oca t io n o n int eract i o n limit of u r a n i um co mp one nt s br i n gs durin g h a ndling o f t wo co nt a in e r s or b a tch es c lo ser fres h U , scra p U , t oge th e r than the a pp r o ve d [Propri e t ary in tera cti on c o ntr o l d i s tan ce In fo rmati o n], a nd ta r ge t s 8.6.1.7 A liquid spill of recycle uranium Criticality issue -Fissile solution may S.C.04 , A liquid spill or target dissolution solution collect in unsafe geometry of fissile solution occurs within the hot cell occurs boundary 8.6.1.9 P rocess so lution s b ac kfl ow C riti ca lit y i ss u e -F i ss il e so lu tion m ay S.C.0 8 , F i ss il e proc ess thr o u g h c h e mic a l a dditi o n lin es t o co ll ec t in un safe geo m e t ry s olution s b ac kfl o w l oca tion s out s ide t h e hot ce ll thr o u g h c h e m ica l b o und ary a ddition lin es 8.6.1.13 Improper installation of HEP A Accidental nuclear criticality leads to S.C.24 , High uranium filters (and prefilters) leads to high dose to worker and potentia l dose content on HEP A transfer of fissi l e uranium to public filters particulate into downstream sections of th e ventilation system with uncontrolled geometries 8.5.1.2 a nd O p e ra t or h a ndling e nrich e d C riti ca li ty haz ard -T oo mu c h ur a nium S.C.27 , F a ilu re of AC 8.5.1.5 so lu t i o n s pour s so lution into a n m ass in o n e pl a c e can l ead to a c c id e nt a l on vo lum e limit durin g un a pp rove d con t a in e r nucl ea r crit ica lit y s amplin g 8.4.1.8 and Drop of a hot cell cover block or Criticality issue -Structural damage S.C.28 , Crane drop 8.6.1.12 other heavy object damages SSCs cou l d adverse l y damage SSCs relied on accident over hot cell relied on for safety for safety , leading to accidents with or other area with SSCs intermediate or high consequence re l ied on for safety 13-34

..... ;* .. NWMI ...... ... *.. .......... . *.* ! . NOmlWUT MfD1CAl tsOTOl'U NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident An alysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Acc i dent Sequences Needing Further Evaluation (5 pages) PHA item numbers Bounding accident description Consequence Accident sequence 8. 1.2.7 and A general facility fir e (caused by Uncontrolled fire can lead to dama ge to S.F.08, General facility 8.1.2.12 vehicle acci d en t inside or outside SSCs r e lied on for safety , re s ultin g in fire of the facility , wildfire , chemical , radiological , or criticality com bu s tibl e fi r e in non-indu stria l ha zards that repre se nt intermediate to a r eas , or fire in n o n-lic e n sed hi g h co n seq u e n ce to worke r s , the material processing a re as) sp r ea d s publi c , and enviro nm e nt to a re as in the building that contain lic ense d mat e ri a l 8.2.1.7 Leak of hydrogen in the facility May lead to an explosion (detonation or S.F.09, Hydrogen attains an explosive mixture and detlagration), depending on the location exp l osion in the facility finds an ignition source, leading in the facility where the hydrogen leaks due to a leak from the to detonation or detlagration of from. Explosion may compromise hydrogen storage or the mixture SSCs to various degrees and may lead distribution system to intermediate or high consequence events. 8.6.1.11 E l ectrica l fire spa rk s l a r ger R a diolo gica l a nd criticality issue -S.F. I 0 , Com bu st ible com bu st ibl e fire in one of the h ot D e p e nding o n the lo catio n a nd qu a nti ty fire occurs in hot ce ll ce ll s of co mbu s tibl es or fl am m a bl es l eft in a r ea the area, a fire in th e hot cell area co uld ruptur e syste ms with hi g h-do se fi ss ion products a n d/or high uranium content , l ea din g to s pills a nd airborne r e l eases 8.1.2.9 and A natural gas leak develops in the Potential explosion that could S.F.11, Detonation or 8.4.1.9 steam generator room and finds catastrophically damage nearby SSCs. detlagration of natural an ignition source , resulting in a Depending on the extent of the damage gas leak in steam detonation or detlagration that to SSCs, an accidental nuclear criticality generator room damages SSCs or an intermediate or high consequence exposure to workers could occur. 8.1.2.7 , Ve hicle in s ide building s trik es Accidental nucl ea r criticality l ea d s to S.M.01 , Vehicle s trik es 8.3. 1.2 , and fresh ur a nium dissolution syste m h i g h dose to workers and potential do se SSC relied on for 8.6.1.5 co mpon e nt , l eading t o a s pill or to public safety and causes acc id e ntal criticality due to damage or l eads to a n disruption of geome tr y and/or acc id e nt se quen ce of interaction interm e di a t e or hi g h consequence 8.4.1.6 TBD (impact must be evaluated TBD (impact must be evaluated after S.M.02, Facility after determining all IROFS that determining all IROFS that re l y on evacuation impacts on rely on personnel action) personnel action) operation 13-35

.. .. NWMI ...*.. ..* *... ........ *.* NORTHWESTllEDICAl.ISOTOPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-16. A d ve r s e Ev ent S ummar y for N ode 8.0 and Identification of Accident S equences N e eding Further Ev aluation (5 pages) PHAitem numbers Bounding accident description Consequence Accident sequence 8.1.2.1 3 8.1.1.1 8.1.1.2 8.1.1.3 8.1.1.4 8.1.1.6 F l oo din g fro m externa l eve n ts an d i n terna l eve nt s compro mi ses the safe geo m etry s l ab a r ea un der certai n ta nk s. D epe n d ing o n the li quid l eve l , int erspe r sed mo d erat i o n of compo n e n ts m ay be im p acte d. Floor storage arrays are s u b j ect to stored co n tai n ers floati n g (l oss of i n teract i o n co n trol). L a r g e tornado s trik es th e facilit y Straig ht-lin e wind s str i ke t h e faci li ty A 48-hr prob a b le m ax imum precipitat i on e ve nt strike s the fac ility Floo din g occ u r s in t he a r ea in excess of 500-ye a r re tu rn freq u e n cy S a fe shutdown e arthquake strikes -Sei s mic s hakin g can lead to damage of the facility and parti a l to complete collapse. Thi s damage impact s SSCs inside and outside the hot cell boundary. Leaks of fiss i le solution , c ompromise o f safe-geometry , and safe interaction storage in s olid material storage arrays and pencil tanks or v essels containing enriched uranium solution s. C riti ca li ty iss u e -Wa t e r acc umul a ti o n und er safe geo m e t ry storage vesse ls o r in safe i nt eract i o n sto r age a rr ays , ca u s in g i nte r s p erse d moderat i o n. Fl oo di ng coul d co mp ro m ise geo m e t ry storage ca p acity for s u bse qu ent s pill s of fi ss il e so lu tio n. Ei th e r event co uld com p romise cr iti ca li ty s afety. Radiologi c al , chemical , and criticality i s sue -Structural damage could a d v ersely damage SSCs relied on for s afet y. F acility could lo se all electrical di s tribution. Facility could lose chilled water s ys t e m function (coolin g tow e r outside of building). R a di o l og i ca l, c h e mi ca l , a n d cr iti ca lit y iss u e -Structural d amage co uld a d verse l y da m age SSCs r e li ed o n fo r safety. Facility co uld l ose a ll e l ectrica l di s tribu tion. Faci l ity co ul d l ose c h i ll ed wa t e r system fu n c ti o n (coo li ng t ower o u ts id e of b uildi ng). Radiological , chemical, a nd criticalit y issue -Structural damage from roof collapse could adversely damage SSC s r e lied on for s afety R a di o l ogical i ss u e -M i nor struc tu ra l d a m age is n ot a nti ci p ated to imp ac t SSCs r e li e d o n for safety exce pt t h a t the fac ilit y could l ose a ll e l ectrica l di s t r ibu t i o n and/o r c h i ll ed wate r s y s tem fu n c t ion (coo l i n g t ower outs id e of build i n g) Radiological , chemical , and criticalit y i s sue -Structural damage could adversely damage SSCs relied on for safety. Facil i ty could lose all electrical distribution.

Facility could lose chilled water s ystem function (cooling tower outside of building). 13-36 S.M.0 3. Floo din g occ ur s in bu i ld i n g du e to i n te rn a l system l ea k or fire s up pressio n s yste m ac t ivatio n (like l y) S.N.01 , Tornado impact on faci l ity and SSC s S.N.0 2 , Hi g h s traig htline w in d i mpact o n faci l ity a nd SSCs S.N.03 , H e avy r a in impact on facility and SSC s S.N.0 4 , Floodi n g impac t o n facility a nd SS Cs S.N.05 , Sei s mic impact on facility and SSC s r .; ... ;. NWMI ...... ..* ... ........ *.* * *.

• NORTHWEST llEOtCAL lSOTOftS NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident A na l ysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Evaluation (5 pages) PHAitem numbers Bounding accident description Consequence Accident sequence 8.1.1.9, H eavy s n owfa ll or i ce buildup R a di o l og i ca l, c h e mi cal, a nd cr it ica li ty S.N.0 6 , H eavy 8.1.1.10 excee d s d es i gn l oa din g of th e i ss u e -St ru c tur a l d a m age fro m ro of snowfa ll or ice buildup roof , r es ult ing in co ll a p se of t h e co ll a p se co uld a d versely d a m age SSCs o n fac ili ty a nd SSCs roo f a nd d a m age t o SSCs (e.g., r e li e d o n fo r safe t y. Loss of s it e those out s id e of t h e h o t ce ll s) e l ectrica l power is hi g h ly lik e l y i n h eavy ice sto rm eve n t. 8.6.1.8 Any stored high-dose product Radiological i s s ue -High-dose solution S.R.01 , A liquid spill s olution spills within the hot cell is unconfined or uncontrolled and can of high-dose fission boundary cause expo s ures to workers , th e public , product solution occur s and environment 8.5.1.5 Op e r a t o r s pill s d i lut e d sa mpl e R a di o l ogical i ss u e -P ote n t i a l s p ray or S.R.O 1 , S pill of p ro du c t o u ts id e of t h e h ot ce ll area va p o ri zatio n of ra di on u cli d e co nt a ini ng so luti on i n l aboratory va p or-ca u sing a d verse wo r ker ex p osu r e (b ase d o n typ i cal l ow qu a n titie s h a ndl ed i n th e l a b ora t o r y , th is i s p os tul a t ed t o be a n in termed i ate co n se qu e n ce eve nt) 8.6.1.10 Recycle uranium transferred out Radiological issue -High radiation may S.R.05 , High-dos e before lag storage decay complete occur in non-hot cell areas, impacting solution exits hot cell or with significant high-dose worker s with higher than normal shie l ding boundary radionuclide contaminants external doses (d e stined for UN b l ending and storage tank) 8.6.1.9 Process so lu t i o n s b ack fl ow R a di o l ogical i ss u e -H igh ra d ia ti on m ay S.R.1 6 , Hi g h-dose thro u g h c h e mi ca l a dditi o n lin es to occ u r in n o n-h o t ce ll areas , im pac tin g p rocess so lu t i o n s locat i o n s ou ts ide t h e h o t ce ll wor k e r s wit h hi g h er t h an normal b ac kfl ow t hr o u g h boun d a r y ex t e rn a l doses c h e mi ca l a ddi tio n lin es 8.6.1.2 and An improperly s ealed co v er block Radiologi cal i ss ue -Depending on S.R.21 , Damage to the 8.6. l.3 or transport door (e.g., for cask location of damage , some streaming of hot cell wall transfers) offer large opening high radiation may occur , impacting pen e tration , potentials for radiation streaming workers with higher than normal compromising external do se s s hielding 8.6.1.1 Th e sea l o n th e bag l ess tra n s p ort R a di o l ogical i ss u e -D egrade d or l oss of S.R.2 4 , B ag l ess d oor fai ls a nd l eads to hi g h-dose casca din g n egat i ve a ir pr ess ur e b e t wee n tra n s p ort d oo r fa ilu re ra d io nu cl id es escap ing t h e h o t zo n es m ay a ll ow hi g h ra d iologica l ce ll confi n e m ent bo und a r y a irb o rn e conta min a ti on to r elease w ith o ut proper fi lt ra ti on an d a d so rpti o n , l ea din g to h ig h er th a n a ll owe d exposure rates t o worke rs a nd th e publi c 8.6.1.13 Following process upsets and Radiological and criticality issue -S.R.25 , HEPA filter over long periods of operation , Following process upsets and over long fai l ure contamination levels in periods of operatio n , contamination downstream components l eads to leve l s in downstream components can high dose during maintenance and lead to high dose during maintenance to uncontrolled accumulation of and to uncontrolled accumulation of fissile material fissile mat e ri a l 13-37

... NWMI ...... ... *.. ........... ' *. * ! . NOITifWUT MEDtCAl tsOTOfllS NWM l-201 3-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-16. Adverse Event Summary for Node 8.0 and Identification of Accident Sequences Needing Further Eva l uation (5 pages) PHAitem numbers Bounding accident description Consequence Accident sequence 8.6.1.2, 8.6.1.3, a nd 8.6.1.6 8.5.1.7 and 8.5.1.8 8.4.1.8, 8.6.1.4 , a nd 8.6.1.1 2 8.2.1.1 AC HEPA IROFS PHA An improp e rl y sea l e d cover bl ock or t ransport d oor (e.g., fo r cask transfers) compromises n egative air press u re b a l ance Laboratory technician is burned by solutions containing radiological i s otop e s during samp l e analysi s activities D ro p of a h o t cell cover bl ock or o t her h eavy object d amages SSCs re l ied on fo r safety All nitric acid from a nitric a cid s torage tank is released in 1 hr from the chemical preparation and storage room admi ni strat i ve contro l. high effic i ency part i cu l ate a i r. ite m s r e l ie d o n for safety. process hazards analysis.

R a di o l og i ca l iss u e -D egra d e d or l oss of S.R.26, Fai l e d nega ti ve casca d i n g n ega ti ve a ir p r ess u re b etwee n air b a l a n ce from zo n e zo n es may a ll ow h ig h radio l og i ca l to zo n e o r fail u re t o ai rb orne co n tami n atio n to re l ease ex h a u s t a radion u c lid e wit h out proper fi l trat i on a n d adsorp t ion, b uil d up in an area l ea di ng to hig h er th a n a ll owe d ex p osure rates to worke r s a nd t he p ubl ic R a diologic al i ss ue -Bums may lead to S.R.31 , Chemical burn s intermediate con s equence event s if eyes from contamin a ted are involved s olutions during sample analy s is R a d io l ogica l a nd cri ti ca l ity iss u e -S.R.32, Crane d rop Struct u ra l da m age cou l d adverse l y acc id e nt over h ot ce ll d amage SSCs re li e d o n for safety, or o th er a r ea w i t h SSCs l ea di ng to acc i de n ts wit h intermed i ate re l ie d o n for safety o r hi g h co n se qu e n ce St a ndard indu s trial accident with S.CS.01 , Nitric acid potential to impact SSC s or cause fume relea s e additional a ccidents of concern SSC TBD u UN structu r es , systems, a n d co mp onents. to be determ i ned. uranium. uranyl n itrate. T he i d e nt i fi e d acc id e n t seq u e n ces a r e furt h er eva lu ate d i n QRA s to co ntinu e th e acc id e nt a n a l ysis a nd t o i d e nti fy IRO FS fo r th ose acc id e nt se qu e n ces th a t excee d t h e p e rf o rm a n ce c r i t er i a as s p ec i fie d i n NW MI-2 014-051 , I ntegrated Safety Ana l ysis Pl an for t h e R adio i soto p e Pr od u ctio n Fac il ity. 13-38

.**.*.*.* NWMI ........ *.* ' !*. * ."

• NORTHWEn MEDtCAl ISOTOPH NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis 13.2 ANALYSIS OF ACCIDENTS WITH RADIOLOGICAL AND CRITICALITY SAFETY CONSEQUENCES This section presents an analysis of accident sequences with radiological and criticality safety consequences. In Section 13 .1.3 , a number of the hazards and accident sequences identified in the PHA that require further evaluation are grouped and identified.

These accident sequences were evaluated using both qualitative and quantitative techniques. Accidents for operations with SNM (including irradiated target processing, target material recycle , waste handling , and target fabrication), radiochemical , and hazardous chemicals were analyzed.

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

Shielded and unshielded criticality accidents are assumed to have high consequences to the worker if not prevented.

Most of the quantitative consequence estimates presented in these accident analyses were for releases to an uncontrolled area (public). The worker safety consequence estimates are primarily qualitative.

As the design matures , quantitati v e worker safety consequence analyses will be performed.

Updated frequency (likelihood) and the worker and public quantitative safety consequences will be provided in the Operating License Application.

Sections 13.2.2 through 13.2.5 present key representative sequence s for radiological and criticality accidents.

Section 13.2.2 discusses spills and spray accidents with both radiological and criticality safety consequences Section 13.2.3 discusses dissolver offgas accidents with radiological consequences Section 13.2.4 discusses leaks into auxiliary system accidents with both radiological and criticality safety consequences Section 13.2.5 discusses loss of electrical power These accidents cover failure of primary vessels and piping in the processing areas , loss of fission product gas removal efficiency , leaks into auxiliary systems , and loss of power to the RPF. Section 13.2.6 briefly presents evaluations of natural phenomena events. The stringent design criteria and requirements for the RPF structure , as discussed in Chapter 3.0 , " Design of Structures , Systems , and Components

," will require the RPF design to survive certain low-return frequency events. Therefore , the return frequency of most of the external events that the RPF will be designed to withstand are highly unlikely per Table 13-1. The remainder of the accident sequences , identified in the PHA as requiring further evaluation , are summarized in Section 13.2.7. Each sequence is identified and the associated IROFS (if any) listed. The IROFS not discussed in Sections 13.2.2 through 13.2.6 are also discu s sed in this section. Numerous accident sequences with both radiological and criticality safety consequences have been evaluated. Some accident sequences are bounded or covered in the preceding accident analysis; others , on further evaluation , have an unmitigated likelihood or consequence that does not require IROFS-level controls.

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

The additional required information will be provided in the Operating License Application. 13-39

...... ; .. NWMI ...... ..* .... ........ *.* ' ! * ! ' . N°"11M'En MUHCAl ISOTOPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.1 Reserved 13.2.2 Liquid Spills and Sprays with Radiological and Criticality Safety Consequences Liquid solution spi ll and spray events causing a radiological exposure hazard were identified by the PHA that represent a hazard to workers from direct exposure or inhalation and an inhalation exposure hazard to the public in the unmitigated scenario.

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

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

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

A process tank containing low-dose uranium so lutions , with no or trace quantities of fi ss ion product radionuclides located in a contact maintenance-type of enclosure typical of the target fabrication syste m s A process tank containing high-dose uranium so lutions located in a hot cell-type of enclosure typical of the irradiated target di sso lution system A process tank containing 99 Mo product solution located in a hot cell-type of enclosure typical of the molybdenum (Mo) purification system (this condition does not lead to a criticality safety concern) In each case, a vesse l is assumed to be filled with process so lution appropriate to the process location with the process off gas ventilation system operating.

A level monitoring system is available to monitor tank transfers and stagna nt storage volumes on all tanks processing LEU or fission product solutions.

Bounding radionuclide concentrations in liquid streams were developed for five regions of the process in NWMI-20I3-CALC-O11, Source T erm Calculations:

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

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

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

This adds a 1.32 margin to the radionuclides stream compositions presented in Chapter 4.0 , " Radioisotope Production Facility Description." Two high-dose uranium solutions located in hot cell enclosures have been evaluated for the Construction Permit Application:

Dissolver product in the target dissolution system -Based on a minimum radionuclide decay time of [Proprietary Information], representing the minimum time for receipt of targets at the RPF Uranium separation feed in the uranium recovery and recycle system -Based on a radionuclide decay time of [Proprietary Information], representing the minimum lag storage time required for impure uranium solution prior to starting separation of uranium from fis si on products 13-40

....... .. NWMI ............ ......... *.* * *. ." , NOmfWEST MEDtcAL ISOT OH:I NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The source term used in this analysis is from NWMI-2013-CALC-01

l. The breakdown of the radionuclide inventory u se d in NWMI-2013-CALC-Ol l is extracted from NWMI-2013-CALC-006, Overall Summary Mat er ial Bal ance -MURR Tar get Bat c h , using the reduced set of 123 radioisotopes.

NWMI-2014-CALC-014 , Selection of Dominant Tar get I sotopes for NWM! Mat e rial Balan ces, identifies the 123 dominant radioisotopes included in the MURR material b ala nce (NWMI-2013-CALC-006). NWMI-2014-CALC-014 provides the ba s is for using the 123 radioisotopes from the total list of 660 radioisotopes potentially present in irradiated targets. The majority of omitted radioisotopes exist in trace quantities and/or decay swiftly to stable nuclides.

The reduced set of 123 radioisotopes consists of those that dominate the radioactivity and decay heat of irradiated targets. Bounding solution concentrations from NWMI-2013-CALC-Ol l are s ummarized in Table 13-17. Additional conservatism has been incorporated in the di sso lver product radionuclide concentrations.

Th e nominal diluted dissolver product volume is [Proprietary Information]

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

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

batch size. The uranium separation feed composition reflects planned processing a djustments that reduce the so lution uranium concentration to [Proprietary Information].

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

[2 4 1 Am]) ha ve increased due to parent decay. Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay , hours a fter EOI Stream description Isotope 24 1Am 13 6 m B a I 37m Ba 1 39 B a 14 0 Ba 14 1ce 14 3 Ce 1 44 Ce 242Cm 24 3 C m 244Cm 1 3 4Cs 1 34m Cs 1 36 Cs 137 Cs 1 s s Eu 1s6Eu 1 s 1 Eu Target dissolution

[Propriet a ry Information]

Di sso lver product Bounding concentration

{Ci/L) [Proprietary Information]

[Propriet ary Inform a ti o n] [Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propriet ary Inform a tion] [Proprietary Information]

[Propriet ary Inform a tion] [Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Propriet ary Inform a tion] [Proprietary Information]

[Propriet ary Inform a tion] 13-41 ' Uranium recovery and recycle [Propri e tary Inform ation] Uranium se paration feed Bounding concentration

{Ci/L) [Proprietary Information]

[Propri e tary Information]

[Proprietary Information]

[Propri e tary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Prop r ietary Information]

.: . .. NWMI ..**.. ..* **.* ......... *.* . *. " "." NOtlTHWEST MEDICAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-17.Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay , hour s after EOI Stream description Isotope 1291 13 0 1 1311 1 32 J 13 2m I 133 1 133m I 1 3 4 1 1351 83 m Kr 85 Kr 85 m Kr 8 7 K_r 88 K_r 140La 141La 142La 99 Mo 95 Nb 9 5mNb 96 Nb 97 Nb 97m Nb 14 1 Nd 236mNp 2 31 Np 23sNp 2 39 Np 233 Pa 23 4 pa 234 m Pa 112pd 141Pm 1 4 8 pm 148mpm Target dissolution

[Proprietary Information]

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

[Proprietary Inform atio n] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Inform atio n] [Proprietary Information]

[Propriet ary Inform atio n J [Proprietary Information)

[Proprietary Inform atio n] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Inform atio n] [Proprietary Information]

[Proprietary Inform atio n J [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

13-42 Uranium recovery and recycle [Propriet ary Information]

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

[Proprietary Information]

[Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri e tary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri e tary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

..... .. NWMI ::.**.*.*. ..... .. .. . . ' NOmfWEST MEDtCAl ISOTDPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident A nalysis Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay , hour s after EO I Stream description Isotope J4 9 pm 1sopm 1 s 1 pm 14 2 pr J4 3 pr 144pr 14 4 m pr 145pr 2Js pu 239 pu 2 4 0 pu 241pu 10 3 m Rh 10 5 Rh 10 6 Rh 1 06m Rh 1 03 Ru 1osRu 1 06 Ru 122sb 1 24 Sb 125 Sb 126Sb 127 Sb 12 s sb 12smsb 1 29 Sb 1s1sm 153 Sm 1s 6 sm s9 sr 9osr 91sr 92 Sr 99 Tc 99mTc Target dissolution

[Proprietary Information]

Dis so lver product Bounding concentration (Ci/L) [Proprietary Information]

[Proprietary Information]

[Proprietary Inform a ti o n] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Inform atio n] [Proprietary Information]

[Propri eta ry Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

[Propriet ary Inform at ion] [Proprietary Information]

[Propriet a ry Inform at ion] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propriet a ry Inform at ion] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

1 3-43 Uranium recovery and recycle [Proprietary Information]

Uranium se paration feed Bounding concentration (Ci/L) [Proprietary Inform atio n] [Proprietary Information]

[Proprietary Inform ation] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Inform ati on] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri e tary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

.. ;. NWMI ...*.. ..* *.. .......... . MlDICAl tSOlOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-17. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation D ecay, hours after EOI Stream description Isotope 125mTe 121Te 12 7m Te 129Te 12 9m Te 1 31 Te 1 3 1mTe 1 32 Te 1 33 Te 133mTe J 3 4Te 231 Th z 3 4Th 232u 234U 23su 236 u 231u 23s u 1J1mxe 133 Xe uJmxe u s xe mmxe 89 my 90 y 90 my 9J y 9 Jmy n y 93 y 93 zr 9s zr 91 zr Totals Target dissolution

[Proprietary Information]

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

[Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propriet ary Inform at ion] [Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Inform atio n] [Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Propriet ary Inform atio n] [Proprietary Information]

[Proprietary Inform atio n] [Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Inform ation] [Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Propriet ary Information]

Uranium recovery and recycle [Proprietary Information]

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

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Inform ation] [Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Informati on] [Proprietary Information]

[Propriet ary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

So ur ce: Table 2-1 ofNWMl-2013-CALC-O 11 , Sou r ce T e rm Cal c ulations , R ev. A , Northwest Medical I soto pe s , LLC , Corva lli s, Oregon , February 2015. EO I = e nd of irradiation.

13-44

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

• NOITifWUT ME.DICAl ISOTOKS 13.2.2.2 Identification of Event Initiating Conditions NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The accident initiating event is generally described as a process equipment failure, but also could be operator error or initiated by a fire/explosion.

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

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

5. A final stable condition is achieved when solution accumulated in the sump has been transferred to a vesse l available for the particular sump material and removed from the enclosure floor. Maintenance activities to repair the cause of a tank or spray l eak are initiated after achieving the final stable condition.

13.2.2.4 Function of Components or Barriers The process vessel enclosure floor , walls , and ceiling will provide a barrier that prevents transfer of radioactive material to an uncontrolled area during a liquid spill or spray accident.

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

The process enclosure secondary confinement (or ventilation) system will provide a barrier to prevent transfer of radioactive material to an uncontrolled area during a liquid spill or spray accident from radioactive material in the airborne particulate and aerosols generated by the event. The secondary confinement system is to function throughout the accident until a stable condition has been achieved. 13-45

... NWMI ..*... .. **.* ..... .. .. .. *

• NCMnlCWHT MEDICAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The process enclosure sump system represents a component credited (part of the double-contingency analysis) for preventing the occurrence of a solution-type accidental nuclear criticality due to spills or sprays of fissile material.

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

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

Failure rates for tanks , vessel s, pipes , and pumps are estimated from WSRC-TR-93-262 , Sa v annah Riv e r Sit e Gen e ric Data Base D eve lopm e nt. Table 13-2 (Section 13.1.1.1) shows qualitative guidelines for applying the likelihood categories.

Operator error and tank failure as initiating events are estimated to have an unmitigated likelihood of"not unlikely." Additional detailed information describing a quantitative evaluation , including assumptions, methodology , uncertainties , and other data , will be developed for the Operating License Application. 13.2.2.6 Radiation Source Term The following source term descriptions are based on information developed for the Construction Permit Application.

Additional detailed information describing source terms will be developed for the Operating License Application.

13.2.2.6.1 Direct Exposure Source Terms Liquid spill source terms are dependent on the vessel location in the process sy s tem. The following source terms describe the three configurations used to span the range of initial conditions

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

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

Dissolution of the targets produces an aqueous solution containing uranyl nitrate, nitric acid , and fission products. The primary attribute of high-dose uranium solutions used for consideration of direct exposure consequences is that equipment operation and maintenance must be conducted in a shielded hot cell environment due to the presence of fission products. 99 Mo product solution was bounded by a small solution volume (less than 1 L) containing the weekly inventory of product from processing MURR targets. The product is an aqueous solution containing M sodium hydroxide (Na OH) with a total inventory of 1.3 x 10 4 curies (Ci) 99 Mo. 13.2.2.6.2 Confinement Release Source Terms Confinement release source terms are based on the five-factor algebraic formula for calculating source terms for airborne release accidents from NUREG/CR-6410, as shown by Equation 13-1. where , ST MAR ST= MAR x DR x ARF x RF x LPF Source term (activity)

Material at risk (activity) 13-46 Equation 13-1 NWMI ..**.. ... **: ......... *.* . !*. * ! * . NORTHWEST MEOtcAl ISOTOPU DR ARF RF LPF Damage ratio (dimensionless)

Airborne release fraction (dimensionless)

Respirable fraction (dimensionless)

Leak path factor (dimensionless)

NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Confinement release source terms for spray used the source term parameters listed in Table 13-18. Four source term cases were developed for evaluation based on the two bounding liquid concentrations shown in Table 13-17 and the source term parameter alternatives.

Parametera Material at risk (MAR) Damage ratio (DR) Airborne release fraction (ARF) Respirable fraction (RF) Leak path factor (LPF) Table 13-18. Source Term Parameters Unmitigated spray release Table 13-17 1.0 0.0001 (1.0 for Kr , X e, and iodine)b 1.0 1.0 Mitigated spray release Table 13-17 1.0 0.0001 (1.0 for Kr , Xe , and iodine)b 1.0 0.0005 (1.0 for Kr , Xe; 0.1 for iodine) Source: Tabl e 2-1 ofNWMI-20 l 5-RPT-009 , F i ss i o n Pr o du c t R e l e ase Evaluati o n , R ev. B , North wes t Medical I so top es, LL C, C orvalli s, Or eg on , 2015. a P a ram e t e r d e finition s deri v ed from NURE G/C R-6410 , N u cle ar Fu e l Cycle F ac ili ty Acc id e nt A nal ys i s Han d b oo k , U.S. Nuclear R e gulatory Commi ss ion , Office o f Nuclear Mat e rial S a fet y and Sa fe gu a rds , Washington , D.C., Mar c h 1998. b A c cid e nt do s e c ons e quen ce s were found to b e s e nsitiv e to iodin e source t e rm parameters. Further work m ay allow for a lower iodine ARF. Kr = krypton. Xe = x en o n. The DR was set to 1.0 for all cases. The assumed volume was 100 L of solution contained b y a vessel being affected by the spill or spray release. The ARF and RF values are functions of the release mechanism and do not enter into consideration for a mitigated versus unmitigated release. Thus, for both the unmitigated and mitigated cases , the ARF and RF were set to representative values based on the guidance in NUREG/CR-6410 and DOE-HDBK-3010 , DOE Handbook -Airborn e R e l e a se Fra c tions/Rat es and R es pirabl e Fra c tion s for Nonr e actor N ucl e ar Facilities. A spray release due to rupture of a pressurized pipe (transfer line) is modeled as depressurization of a liquid through a leak below the liquid surface level. Both NUREG/CR-6410 and DOE-HDBK-3010 report an ARF of 1 x 10-4 and a RF of 1.0 for a spray leak involving a low temperature aqueous liquid. These values take into consideration upstream pressures as high as 200 pounds (lb)/square inch (in.2) gauge. The spray mechanism is also bounded by a droplet size distribution produced from commercial spray nozzles. This approach is conservative , as the effective nozzle created by a pipe failure is unlikely to be optimized to the extent of a manufactured spray nozzle. Therefore , an ARF of 1 x 10-4 and a RF of 1.0 were used for all isotopes , except iodine and the noble gas fission products Kr and Xe. Radioisotopes of Kr , Xe , and iodine were assigned an ARF of 1.0 for all cases. For the unmitigated evaluations , the LPF was set to 1.0 , since the unmitigated release scenario credits no confinement measures (i.e., no credit was taken for any aspect of the facility design or equipment performance). The gravitational settling associated with flow throughout the facility and the removal action of high-efficiency particulate air (HEPA) filtration may be lumped into an effective value for LPF. The performance of different filtration systems is presented in Appendix F of DOE-HDBK-3010.

For scoping purposes, a HEPA filtration efficiency of 99 .95 percent was selected for all mitigated cases , which corresponds to an LPF of 0.0005. 13-47

.**.*.*.* NWMI ..... .. .. .. . NOlmrwmMEOICAllSOTOf'U NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The HEPA filter LPF was applied to all isotopes except Kr , Xe , and iodine. An LPF of 1.0 was selected for Kr and Xe in the mitigated spray release evaluation , assuming these isotopes behave as a gas when airborne and are not removed by HEPA filtration or sufficiently retained on the high-efficiency gas adsorption (HEGA) modules. The mitigated analysis credits an iodine removal capability in the facility ventilation exhaust gas equipment, with an iodine removal efficiency of 90 percent. The credited removal efficiency corresponds to an LPF of 0.1 for iodine due to the HEGA modules co-located with the HEPA filters. 13.2.2. 7 Evaluation of Potential Radiological Consequences Confinement release consequence estimates for the Construction Permit Application are based on NUREG-1940 , RASCAL 4: D e scription of Models and M e thod s, and Radiological Safety Analysis Code (RSAC), Version 6.2 (RSAC 6.2). Additional detailed information describing validation of models , codes , assumptions , and approximations will be developed for the Operating License Application.

13.2.2.7.1 Direct Exposure Consequences The potential radiological exposure hazard of liquid spills depends on the vessel location in the process system. Low-dose uranium solutions are generally contact-handled , and direct radiation exposure to the worker is expected to be slightly elevated but well within ALARA guidelines.

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

Vessels located within hot cells require s hielding to control worker radiation exposure independent of whether process solution is contained in the vessel or spilled to the enclosure floor. High-dose uranium solutions are assumed to require hot cell shielding. Spills of 99 Mo solution from the Mo recovery and purification processes, and during handling prior to shipment of the product , involve product solution that c ontains high-dose 99 Mo. The direct whole-bod y exposure from radiation does not change from the normal case and must always be shielded to reduce the dose rate for workers to ALARA. As a preliminary estimate using a point-source dose rate conversion factor for 99 Mo of 0.112924 roentgen equivalent man (rem)/hr at 1 meter (m) per Ci 99 Mo , the unshielded dose rate for the product is: MAR= 1.3 x l 0 4 Ci 99 Mo 99 Mo dose rate at 1 m = 1.30 x 10 4 Ci 99 Mo x 0.1129 rem/hr/Ci 99 Mo = 1.5 x 10 3 rem/hr In a very short period of time , a worker can receive a significant intermediate or high consequence dose. Therefore , both high-dose uranium and 99 Mo product solution vessels must be located in hot cells for normal operations to control the direct exposure to workers. Based on the analysis of several accidental nuclear critical i ties in industry , LA-13638 , A R e vi e w of Criticality A c cidents , identifies that a uranium solution criticality can yield between 10 1 6 to 10 17 fissions.

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

Consequences for a s hielded hot cell criticality will be developed for the Operating License Application. 13.2.2.7.2 Confinement Release Consequence Receptor dose consequences were originally evaluated in NWMI-2015-RPT-009 , Fis s ion Produ c t R e lease Evaluation , using the RASCAL code. Since the submission of the application , NWMI has selected RSAC 6.2 for off-site accident consequence modeling. For the liquid spills and spray accident , NWMI has rerun the dissolver product off-site dose calculations using RSAC 6.2. Four release consequence estimates were prepared to support the Construction Permit Application based on unmitigated and mitigated spray release events using the two liquid radionuclide concentrations shown in Table 13-18. The RSAC inputs for the dissolver product accident are listed below , and the RASCAL inputs for the high dose uranium solution are listed in Table 13-19. 13-48

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

• NOlmfW£ST MEDICAL ISOTOPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-19. Release Consequence Evaluation RASCAL Code Inputs Input Prim a r y t oo l Event type Fac ilit y lo ca ti o n* County Tim e zo n e Latitude/longitude E l eva ti o n Plume rise Me t eoro l ogy Receptor distance D ose co n versio n fac t o r s Description S TD ose -So u r c e t e rm t o do se opti o n se l ecte d as th e pr i m a r y to o l in RA SC AL for all c ases. Other release -RASCAL includes separate models for nuclear power plant accidents involving spent fuel , accidents involving fuel cycle acti v ities , and other radioactive material releases at non-reactor facilities.

The other radioactive material releases option was selected for all cases. Co lumbi a, M i sso uri Boone Ce nt ra l 38.9520° N/92.3290° W 231 m None -For scoping purpose s, the enthalp y and momentum of the RPF stack exhaust was a ssumed negligible. S umm er-ni g ht-c a lm -Se l ec t e d fo r scop in g purp oses a nd fea tur es wi nd s p ee d of 6.4 km/h r (4 mi/hr), P as quill C l ass F s t abi l i t y, no pr ec ip ita ti o n , r e l a ti ve humidit y o f 8 0%, a nd a mbi e nt t e mp era tur e o f 1 2.8°C (55°F). Low w ind s p eed a nd s tabl e co ndition s se l ec t ed t o p rov id e m ax imu m d ose t o n e ar-fie ld r ece ptor s. 100 m -Selected to approximate site boundary. Input represents minimum value for RASCAL input. I C RP-72b -Se l ec t e d as th e m os t c u rre n t an d a u t h or it a t ive se t of d ose co n ve r sio n fac to rs availa bl e. So ur ce: Ta ble 2-1 ofNWM I-20 1 5-RP T-0 09, Fiss i o n Produc t Re l e ase Eva lu at i on, Rev. A, No rth west M e di cal I sotopes, LLC , Corva lli s, Oregon, Fe bru ary 20 1 5. a L oca ti o n i n fo rm a ti o n obtaine d fro m Wikipe d ia. b I C RP-72, Age-Dependent Doses to t h e Members of th e Pub l ic from I ntak e of R ad i on u clid es -Part 5 Compilation of I ngest i o n and I n h a l at i o n Co effic ients , Int e rn a ti o n a l Co mmi ss i o n on R a di o l og i cal Protect i o n , Ottawa , Ca n a d a, 1 995. RASCAL = Ra di o l og i ca l Assess m e nt System for RPF = Ra di o i so t ope Prod u c ti o n Fac ili ty. Co n se qu e n ce Ana l ys i s. R S A C 6.2 was u se d to mod e l th e di s p e r s i on r e sultin g fr om a s p ray l eak. Th e fo ll ow in g p ara m e ter s we r e u s ed for m o d e l run s: Mi x in g d e pth: 40 0 m (1 , 312 f ee t [ft]) (d efa ult) A ir d e n s it y: 1 ,2 40 g/c ubi c m e t e r [m 3] (1.2 4 o un ce [o z]/c ubi c feet [ft 3]) (sea le v el) P as qu i ll-Gi f ford a (N R C R eg ul a to ry Guid e 1.1 45, At m osp h eric Di spers i o n Mo d e l s for P ote nt ia l Accident Co n sequence Assess m ents a t Nuclear P owe r Pl ants) N o plum e ri s e (i.e., buo ya n cy o r s t ac k mom e ntum e ff ec t s) N o plum e d e pl e ti o n (we t or dry d e po s it i on) 1-hr r e l eas e (con s t a nt r e l e a s e of a ll a c ti v i ty) I -hr ex p os ure I C RP-3 0 , Limit s for I nt ak es of R adio nu cl i des by Wo rk e r s, inh a lation m o d e l Finite cloud immer s i o n model Bre a th i n g rate: 3.4 2E-4 m 3/s e co nd (se c) (1.2E-2 ft 3/se c) (I C RP-30 h eavy ac ti v it y) 13-49

.. NWMI ............ .*.* .. *.*.* ' ! *. * ." . NOmlW£ST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Consequence evaluation results are shown in Figure 13-2 and Table 13-20 for a I 00 L (26.4 gal) spray release event. The unmitigated spray release of dissolver product solution is an immediate consequence event. The nearest permanent resident, at 432 m (0.27 miles [mi]), dissolver product spray unmitigated dose estimate is 300 mrem , while the maximum receptor location (1,100 m [0.68 mi]) has a total effective dose equivalent (TEDE) of 1.8 rem. The mitigated consequences are an order of magnitude lower due to the credited IROFS in the Zone I exhaust system. Therefore the nearest permanent resident ( 432 m [0.2 7 mi]) dis so lver product spray mitigated dose estimate is 30 mrem, while the maximum rec eptor location (1,100 m [0.68 mi]) has a TEDE of0.18 rem. 2.0 1.8 1.6 1.4 E 1.2 Q) ,_ a) 1.0 <J) 0 Cl 0.8 0.6 0.4 0.2 -e-Inhalation CEDE -e-External EDE -e-TEDE -e-Thyroid Inhalation WCDE 100 200 300 400 500 600 700 800 900 1000110012001300140015001600 Distance, meters Figure 13-2. Unmitigated Off-Site Dose of Dissolver Product Spray Leak Accident Table 13-20 shows that the uranium separation feed solution spray release unmitigated dose is below the immediate consequences thre s holds of 10 CFR 70.61. Table 13-20. Spray Release Consequence Summary Process stream Case Mitigation Receptor dose, total EDE Stack height Release mechanism Release duration Uranium separations feed Unmitigated 0.078 rem 10 m (33 ft)" Spray leak, I 00 L 1 hr Mitigated 0.006 rem 23 m (75 ft) Source: Table 2-1 an d Table 2-7 ofNWMI-2015-RPT-009 , Fission Product Rel e as e Evaluation , Rev. A , Northwest Medical Isotopes , LLC, C orvallis , Oregon, February 2015. EDE = effec tiv e dose equiva l ent. 13-50

..

...... ..* .... ........ *.* ". NOflTKWHTMEOtcAl.ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.2.8 Identification of Items Relied on for Safety and Associated Functions Unmitigated spill and spray releases have the potential to produce direct exposure and confinement releases with high consequence to workers and the public. Hot cell shielding is designed to provide protection from uncontrolled liquid spills and sprays that result in redistribution of high-dose uranium and 99 Mo product solution in the hot cell. From a direct exposure perspective , a liquid spill does not represent a failure or adverse challenge to the hot cell shielding boundary function.

However , the hot cell shielding boundary must also function to prevent migration of liquid spills to uncontrolled areas outside the shielding boundary. Liquid spill and spray-type releases occur as a result of the partial failure of process vessels to contain either the fissile solution (for areas outside of the hot cell) or to contain fissile or high-dose radiological so lutions (for areas inside the hot cell). In either case, the process vesse l spray release result s in an event that carries with it a higher airborne radionuclide release magnitude than a simple liquid spill. The type release also carries the extra hazard of potential chemical burn s to eyes and skin, with the complication of radiological contamination.

Consequently, spray protection i s a secondary safety function needed to satisfy performance criteria.

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

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

• IROFS RS-0 l , "Hot Ce ll Liquid Confinement Boundary" IROFS RS-03 , " Hot Ce ll Secondary Confinement Boundary" IROFS RS-04 , "Hot Cell Shielding Boundary" Liquid spill and spray events involving solutions containing fissile material have the potential for producing liquid nuclear criticalities that must be prevented.

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

13.2.2.8.1 IROFS RS-01, Hot Cell Liquid Confinement Boundary IROFS RS-01 functions to mitigate the impact of liquid spills from process vessels in the hot cells. As a passive engineered control (PEC) and safety feature, the hot cell liquid confinement boundary will provide an integrated system of features that protects workers and the public from the high-dose radiation generated during primary confinement releases of primarily liquid solutions during the 99 Mo recovery process. The hot cell liquid confinement boundary will also protect the environment from releases of product solution from the primary confinement of the processing vessels. In addition, the barrier wi ll provide a function of confining spills of irradiated LEU target solid material in some of the irradiated target handling hot cells. 13-51

... .. NWMI ...*.. ..* **.* ..... .. .. .. . ' ! ; NOmfWfn MlDtCAl tsOTOf'H NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The primary safety function of the hot cell liquid confinement boundary is to capture and contain liquid releases and to prevent those releases from exiting the boundary , causing high dose to workers or the public , or contaminating the environment.

A secondary function of the liquid confinement boundary is to prevent contact chemical exposure to workers from acidic or caustic solutions contaminated with licensed material that exceeds the performance criteria established by NWMI for the RPF. As a PEC to contain spills and sprays of high-dose product solution , the hot cell liquid confinement boundary will consist of sealed flooring with multiple layers of protection from release to the environment.

Various areas will be diked to contain specific releases , and sumps of appropriate design will be provided with remote-operated pumps to mitigate liquid spills by capturing the liquid in appropriate safe-geometry tanks. Additional IROFS apply to the flooring and sumps for criticality safety double-contingency controls in some areas. In the 99 Mo purification product and sample hot cell , smaller confinement catch basins will be provided under points of credible spill potential in addition to use of a sealed floor. Entryway doors into a designated liquid confinement area will be sealed against credible liquid leaks to outside the boundary.

This continuous barrier is also credited to prevent spills or sprays of high-dose product solutions that are acidic or caustic from causing adverse exposure to personnel through direct contact with skin, eyes, and mucus membranes , where the combination of the chemical exposure and the radiological contamination would lead to serious injury and long-lasting effects or even death. Specific design features of the liquid confinement barrier , a liquid barrier to uncontrolled areas and worker radiation exposure from leaked solution , include: * * * *

• Continuous , impervious floor with an acid-or caustic-resistant surface finish Hot cell walls and ceiling designed to control worker dose from liquids accumulated in sumps Monitors with alarms to indicate a liquid release has occurred Sealed penetrations designed to prevent liquid leaks through the barrier to uncontrolled areas Sump solution collection vessels for accumulating leaked process solution 13.2.2.8.2 IROFS RS-03, Hot Cell Secondary Confinement Boundary IROFS RS-03 functions to mitigate the impact of liquid spills and sprays from process vessels in the hot cells. As a system of PECs and AECs , the hot cell secondary confinement boundary safety feature is engineered to provide backup to credible upsets in the primary confinement system using the following safety functions: Provide negative air pressure in the hot cell (Zone I) relative to lower zones outside the hot cell using exhaust fans equipped with HEP A filters and HEGA modules to remove the release of radionuclides (both particulate and gaseous) to outside the primary confinement boundary to below 10 CFR 20 release limits during normal and abnormal operations.

• Components credited include: Zone I Inlet HEPA filters to provide an efficiency of 99.97 percent for removal of radiological particulates from the air that may reverse flow from Zone I to Zone II Zone I ducting to ensure that negative air pressure can be maintained by conveying exhaust air to the stack Zone I exhaust train HEPA filters to provide 99.97 percent removal of radiological particulates from the air that flows to the stack Zone I exhaust train HEGA modules to provide 90 percent removal of iodine gas from the air that flows to the stack Zone I exhaust stack to provide dispersion of radionuclides in normal and abnormal releases at a discharge point of 22.9 m (75 ft) above the building ground level 13-52

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

• NOATHWEST M£DICAl llOTOPlS NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Stack monitoring and interlocks to monitor discharge and signal changing on service filter trains during normal and abnormal operations As a system of PECs and AECs, the purpose of this IROFS is to mitigate high-dose radionuclide releases to maintain exposure to acceptable levels to both the worker and the public in a highly reliable and available manner. The hot cell secondary confinement boundary will perform this function using the following engineered features to ensure a high level of reliability and availability. As a PEC , the hot cell floor, walls , ceilings , and penetrations are designed to provide an air intrusion barrier sufficient to allow the exhaust system to maintain negative air pressure under normal and credible abnormal conditions.

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

Design features associated with this function include airlocks for normal egress , cask and bagless transfer ports that can only open when the cask or container is properly sealed to the port , and appropriately sized ventilation ports between zones. Along with the AECs of the filtered ventilation system , this boundary will provide secondary confinement and prevent uncontrolled release of general radiological airborne gases and particulates that escape the primary confinement to reduce releases to the monitored stack to acceptable release levels during normal and abnormal operations.

The Zone I exhaust system will serve the hot cell , high-integrity canister (HIC) loading area , and solid waste loading area. This exhaust system will maintain Zone I spaces at negative pressure with respect to atmosphere.

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

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

13-53

.; ... ; .. NWMI ...*.. ..* **.* .********** . ' " "* . NOmlWUT MElHCAl ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The hot cell shielding boundary will provide shielding for workers and the public during normal operations to reduce worker exposure to an a v erage of 0.5 mrem/hr , or less , in normally accessible workstations and occupied areas outside of the hot cell. The hot cell shielding boundary will provide s hielding for workers and the public during process upsets to reduce the worker exposure to a TEDE of 5 rem , or less, at workstations and occupied areas outside of the hot cell. As a PEC , shielding will be provided by a thick concrete , steel-reinforced wall with steel cladding that reduces the normally expected operational exposures from within the boundary to an average of 0.5 mrem/hr , or less , outside of the boundary.

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

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

These incidents include spills , s prays, fires , and other releases of radionuclides contained within the boundary. The shield may be divided into protection areas for the purposes of applying limiting conditions of operation. Each shielded protected area will be operable when the equipment in that area is in the operating or standby modes. 13.2.2.8.4 IROFS CS-07, Pencil Tank and Vessel Spacing Control using Fixed Interaction Spacing oflndividual Tanks or Vessels IROFS CS-07 functions to ensure that potential interactions between full ves s els and a sump filled by a liquid spill or spray have been considered to prevent a nuclear criticality event. As a PEC , pencil tanks and other standalone vessels (controlled with safe geometry or volume constraints) are designed and fabricated with a fixed interaction spacing for safe storage and processing of the fissile solutions.

The safety function of fixed interaction spacing of individual barrels in pencil tanks and between other single processing vessels or components is designed to minimize interaction of neutrons between vessels such that under normal and credible abnormal process upsets , the system s will remain subcritical.

The fixed interaction control of tank s, vessels, or components containing fissile solutions will prevent accidental nuclear criticality , a high consequence event. The fixed interaction contro l distance from the safe slab depth spill containment berm is specified where applicable. 13.2.2.8.5 IROFS CS-08, Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Spill Containment Berms IROFS CS-08 functions to ensure that sump designs have been considered to pre v ent a nuclear criticality event by geometry if filled with liquid from a spill or spray release. As a PEC , the floor under designated tanks , vessels , and workstations will be constructed with a spill containment berm that maintains a geometry s lab depth to be determined with final design , and one or more collection sumps with diameters or depths to be determined in final design. The safety function of this spill containment berm is to safely contain spilled fissile solution from systems overhead and prevent an accidental nuclear criticality if one of the tanks or related piping leaks , ruptures , or overflows (if so equipped with overflows to the floor). Each spill containment berm will be sized for the largest single credible leak associated with the overhead systems. The interaction distance for the spill containment area is provided in IROFS CS-07. The sump will have a monitoring system to alert the operator that the IROFS has been used and may not be available for a follow-on event. A spill containment berm will be operable if it contains reserve volume for the largest single credible spill. Spill containment berm sizes and locations will be determined by the fina l design. 13-54

.; ... ; .. NWMI ...... ..* ... ........ *.* . ! *.* ! . NO<<nfW£n *DtCAL lSOTOl'U NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.2.8.6 IROFS CS-09, Double-Wall P i ping IROFS CS-09 functions to control liquid spills or sprays in a similar manner to IROFS CS-08. As a PEC , the piping system conveying fissile solution between credited locations will be provided with a wall barrier to contain any spills that may occur from the primary confinement piping. IROFS CS-09 is used at locations that pass through the facility where creating a spill containment berrn (IROFS CS-08) under the piping is neither practical nor desirable for personnel chemical protection purposes.

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

The secondary safety function of the double-wall piping is to prevent personnel injury from exposure to acidic or caustic licensed material solutions that are conveyed in the piping. Defens i ve-in-Depth The following defense-in-depth features were identified by the liquid spill and spray accident eva l uation s. Alarming radiation area monitors will provide continuous monitoring of the dose rate in occupied areas , and alarrn at an appropriate setpoint above background.

Continuous air monitoring will be provided to alert operators of high airborne radiation levels that exceed derived air concentration (DAC) limit s. HEPA filters on hot cell outlets are not credited and will reduce the impact of spills or sprays to the public. Most product solution and uranium solution processing systems will operate at or slightly below atmospheric pressure , or solutions will be pumped between tanks that are at atmospheric pressure to reduce the likelihood of system breach at high pressure.

• Tanks , vessels , components , and piping are designed for high reliability with materials that will minimize corrosion rates associated with the processed solutions. 13.2.2.9 Mitigate d Estimates The controls selected will mitigate both the frequency and consequences of this accident.

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

IR Us Pressure-relief vessel Primary adsorbers (carbon media beds for 6 days noble gas holdup) 13-55

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

lSOTOPO NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Iodine guard beds (remove any iodine not trapped in the IRUs)

• Filters Vacuum receiver tanks Vacuum pumps (draw a downstream vacuum on the target dissolver offgas treatment train) Secondary adsorbers (additional carbon media beds to hold up noble gases for an additional 60 days) The IR Us nominally removes about 99.9 percent of the iodine in the off gas stream after the NO x scrubbers.

NWMI expects the availability and operation ofIRUs will become part of the technical specification to meet annual release limits. The iodine released from dissolution of the irradiated targets will have three primary pathways:

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

[see Section 13.2.2]), (2) a significant portion of the iodine gas exiting the dissolver will be captured in the caustic scrubber (and other NOx treatment absorbers) and end up in the high dose liquid waste tanks, and (3) the remainder of the iodine will be captured in the IRUs. These IR Us will remove the bulk of the radioactive iodine that passes through the dissolver scrubbers during the dissolution process. As demonstrated by the MHA analysis discussed in Chapter 19.0 , iodine will be the greatest contributor to the effective dose equivalent (EDE) for gaseous accident-related releases from the RPF. The primary and secondary adsorbers will be important for delaying the release of radioactive noble gases (radioisotopes of Kr and Xe) until these isotopes have had time to decay. However , as shown in the MHA analysis in Chapter 19.0, the dose impact of noble gases will be orders of magnitude below that of radioiodine.

Therefore, this evaluation focuses on accidents or upsets negatively impacting the IRU performance as the bounding off gas accident.

13.2.3.1 Initial Conditions The target dissolver and associated off gas treatment train are assumed to be operational and in service prior to the occurrence of any accident sequence that affects the IRUs. The IRUs are assumed to be loaded with the conservative bounding holdup inventory of iodine , as determined in Ol 1. There is no credible event where the inventory on the IRUs would be released.

Therefore, this evaluation focuses on accident sequences where the inventory at risk is from a single dissolution of [Proprietary Information].

The maximum amount of iodine [Proprietary Information]

is shown in Table 13-21. The mass balance projects about 20 percent of the iodine will stay in the dissolver solution and Table 13-21. Maximum Bounding Inventory of Radioiodine

[Proprietary Information]

Isotope 1 3 3 1 Total I Ci = iodine. [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

nearly 50 percent of the elemental iodine (h) that does volatize will be captured in the NO x scrubbers (primary the caustic scrubber) and transferred to the high dose liquid waste system. However , for this analysis , all of the iodine is assumed to evolve and remain in the off gas stream going to the IRUs. 13-56

..... ; .. NWMI ::.**.*.*.* ..... .. .. .. . NOmtWtnUEDICALISOTOPU 13.2.3.2 Identification of Event Initiating Conditions NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis There are a number of events identified in the PHA that have the potential to impact the normal efficient operation of the target dissolver off gas treatment train. The three most likely sequences with the potential to impact efficient operation include: (1) excessive moisture carryover in the gas stream due to a process upset in the NO x units , (2) high gas flow rates due to process conditions in the dissolver (e.g., excessive sweep air) or poor NO x recovery , and (3) loss of temperature control (loss of power or failure of temperature controller) to the IRU. All three of these accidents have the potential to reduce the IRU efficiency. 13.2.3.3 Description of Accident Sequences The accident sequences for loss of IRU efficiency include the following.

[Proprietary Information]

is being dissolved.

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

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

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

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

Failure rates for tanks , vessels , pipes , and pumps are estimated from WSRC-TR-93-262. Table 13-2 shows qualitative guidelines for applying the likelihood categories.

Operator error and equipment failure as initiating events are estimated to have an unmitigated likelihood of "not unlikely." Additional detailed information describing a quantitative evaluation , including assumptions , methodology , uncertainties , and other data , wi ll be developed for the Operating License Application. 13-57

..... ; ... NWMI ...... ..* **: ........ *.* . * *, * ! ." . NOITifWEST llB>tCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.3.6 Radiation Source Term The radioiodine inventory is given in Section 13.2.3.1.

As discussed with regard to the MHA in Chapter 19.0, the dose consequences of noble gas radioisotopes are orders of magnitude less than that of iodine radioisotopes.

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

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

Conversely , in this accident, the previous capture iodine is not part of the source term. Therefore , the source term is 27 , 100 Ci. Additional detailed information describing the validation of models, codes , assumptions , and approximations will be developed for the Operating License Application. The source term for this accident is based on a set of initial conditions that were designed to bound the credible offgas scenarios. These assumptions include: [Proprietary Information]

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

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

l. The breakdown of the radionuclide inventory used in NWMI-2013-CALC-Ol l is extracted from NWMI-2013-CALC-006 using the reduced set of 123 radioisotopes. NWMI-2014-CALC-014 identifies the 123 dominant radioisotopes included in the MURR material balance (NWMI-2013-CALC-006).

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

The reduced set of 123 radioisotopes consists of those that dominate the radioactivity and decay heat of irradiated targets. 13.2.3. 7 Evaluation of Potential Radiological Consequences Radiological consequences are bounded by those of the MHA (Section 19.4 .. The unmitigated dose consequences should be about 3.4 times Jess than the MHA results for the public , based on the source term ratio. Realistic radiological consequences are negligible due to the presence of defense-in-depth iodine capabilities in the dissolver offgas system and in the process vessel vent system that backs up the performance of the target dissolver off gas treatment train IR Us. Additional detailed information describing validation of the models , codes , assumptions , and approximations will be developed for the Operating License Application. Assuming this accident has similar release characteristic as the MHA , the radiological dose consequences can be estimated using the ratio of source terms. This is reasonable since a dissolution takes 1 to 2 hr. The entire inventory would also be released over a 2-hr period directly to the 22.9 m (75-ft) stack and into the environment.

RSAC 6.2 was used to model the dispersion resulting from the MHA. The following parameters were used for model runs: Mixing depth: 400 m ( 1,312 ft) (default)

Air density: 1,240 g/m 3 ( l.24 o z/ft 3) (sea level) Pasquill-Gifford cr (NRC Regulatory Guide 1.145) No plume rise (i.e., buoyancy or stack momentum effects) No plume depletion (wet or dry deposition) 2-hr release (constant release of all activity) 2-hr exposure 13-58

...... ; ... NWMI ...... ..* .... *.*.* .. *.*.* . ' . NOllTHWtST MEDICAL ISOTOPE S ICRP-30 inhalation model Finite cloud immersion model Breathing rate: 3.42E-4 m 3/sec (l.2E-2 ft 3/sec) (ICRP-30 heavy activity)

Respiratory fraction: 1.0 Table 13-22 shows the distance-dependent total receptor accident doses versus distance from the RPF stack for 2-hr exposure. This table was developed using MHA dose consequences and dividing by a ratio of the MHA and the accident source term. The maximum public dose is 6.65 rem at 1 , 100 m. RSAC 6.2 calculates inhalation doses using the ICRP-30 model with Federal Guidance Report No. 11 dose conversion factors (EPA 520/1-88-020, Limiting Valu e s of Radionuclide Intak e and Air Conc e ntration and Dos e Conver s ion Factors for Inhalation, Submersion, and Ingestion).

The committed dose equivalent (CDE) is calculated for individual organs and tissues over a 50-year period after inhalation.

The CDE for each organ or tissue is multiplied by the appropriate ICRP-26, Recommendations of the International Commission on Radiological Protection, weighting factor and then summed to calculate the committed effective dose equivalent (CEDE). N WM l-2013-0 2 1, Rev. 1 C ha pter 13.0 -A ccident A nalysis Table 13-22. Target Dissolver Offgas Accident Total Effective Dose Equivalent TEDE (rem) Distance (m) Total 100 2.05E-Ol 200 I .98E-Ol 300 2.21E-01 400 6.41E-OI 500 1.76E+o0 600 3.18E+OO 700 4.50E+OO 800 5.47E+OO 1,000 6.50E+OO 1 , 100 6.65E+OO 1,200 6.62E+OO 1 , 300 6.50E+OO 1,400 6.29E+OO 1,500 6.06E+OO 1,600 5.82E+o0 1 , 700 2.05E-OI P e ak total dose i s bolded and it a lici z ed. TED E = t o tal effective do se e quivalent.

The RSAC 6.2 gamma dose from the cloud is the EDE (the person may or may not be immersed in the cloud depending on the plume position in relation to the ground surface), which is the sum of the products of the dose equivalent to the organ or tissue and the weighting factors applicable to each of the body organs or tissues that is irradiated.

The summation of the two RSAC 6.2 doses is the TEDE , which is the sum of the EDE (for external exposures) and the CEDE (for inhalation exposures).

The RSAC 6.2 dose calculations and dose terminology are consistent with 10 CFR 20 terminology based on ICRP-26/30. The doses and dose commitments rem) are within intermediate consequences severity categories

(<25 rem). 13.2.3.8 Identification of Items Relied on for Safety and Associated Fu n ctions IROFS RS-03, Hot Cell Secondary Confinement Boundary The applicable part ofIROFS RS-03 that specifically mitigates target dissolver offgas treatment train IRU failures is the process vessel vent iodine removal beds. These beds are located downstream of where the target dissolver off gas treatment train discharges into the process vessel vent system; hence, the beds provide a backup to the target dissolver offgas treatment train IRUs. IROFS RS-03 is categorized as an AEC. 13-59

...... ; ... NWMI ..*... ..* *.. ......... *.* . ', *. * !

• NOl11fWEST MEOK:Al tSOTOPlS IROFS RS-09, Primary Off gas Relief System NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis As an AEC, a relief device will be provided that relieves pressure from the system to an on-service receiver tank maintained at vacuum, with the capacity to hold the gases generated by the dissolution of one batch of targets in the target dissolution tank. The safety function of this system is to prevent failure of the primary confinement system by capturing gaseous effluents in a vacuum receiver.

To perform this function , a relief device will relieve into a vacuum receiver that is sized and maintained at a vacuum consistent with containing the capacity of one batch of targets in dissolution.

Defensive-in-Depth The following defense-in-depth features preventing target dissolver off gas accidents were identified by the accident evaluations. Releases at the stack will be monitored for radionuclide emissions to ensure that the overall removal efficiency of the system is reducing emissions to design levels and well below regulator limits. A spare dissolver off gas IRU will be available if the online IRU unit loses efficiency.

The primary carbon retention bed will include an iodine adsorption stage that reduces iodine as a normal backup to the IRU. 13.2.3.9 Mitigated Estimates The controls selected do not affect the frequency ofthis accident but mitigate the consequences. The process vessel vent iodine removal bed and the HEGA module in the Zone I exhaust system will mitigate the dose consequences by a factor of I 00. The selected IROFS have reduced the off-site consequences to acceptable levels (less than 66 mrem to the public). Additional detailed information , including worker dose estimates and detailed frequency , will be developed for the Operating License Application. 13.2.4 Leaks into Auxiliary Services or Systems with Radiological and Criticality Safety Consequences In the unmitigated scenario , liquid solution leaks into secondary containment (e.g., cooling water jackets) were identified by the PHA to represent a hazard to workers from direct radiological exposure or inhalation and an inhalation exposure hazard to the public. The PHA also identified fissile solution leaks into secondary containment as an event that could lead to an accidental nuclear criticality. The accidents covered by this analysis bound the family of accidents where highly radioactive or fissile solution leaves the hot cell or other shielded areas via auxiliary systems and creates a worker safety or criticality concern. 13.2.4.1 Initial Conditions Initial conditions are described as a tank or vessel (with a heating or cooling jacket) filled with process solution. Multiple vessels are projected to be at this initial condition throughout the process. The second primary configuration of concern is the hot cell and target fabrication condensers associated with the four concentrator or evaporator systems. The evaporator(s) initial conditions are normal operations , in which boiling solutions generate an overhead stream that needs to be condensed.

The bounding source term is expected to be the dissolvers or the feed tanks in the Mo recovery and purification system. Table 13-23 lists the radionuclide liquid concentration for [Proprietary Information]. The [Proprietary Information]

stream is used to represent and bound the uranium recovery and recycle and target fabrication evaporators feed streams. 13-60

.**.*.*.* ...... NWMI .*.* .. *.*.* NO<<THWESTMlDtCAl.ISOTOf'fl NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Target dissolution Uranium recovery and recycle Decay , hours after EOI [Proprietary Information]

[Proprietary Information]

Stream description Dissolver product Uranium se aration feed 2 41Am [Proprietary Information]

[Proprietary Information]

136mBa [Proprietary Information]

[Proprietary Information]

1 37m Ba [Proprietary Information]

[Proprietary Information]

139 Ba [Proprietary Information]

[Proprietary Information]

1 40 Ba [Proprietary Information]

[Proprietary Information]

141ce [Proprietary Information]

[Proprietary Information]

1 43 Ce [Proprietary Information]

[Proprietary Information]

144Ce [Proprietary Information]

[Proprietary Information]

242 cm [Proprietary Information]

[Proprietary Information]

2 4 3 Cm [Proprietary Information]

[Proprietary Information]

244 Cm [Proprietary Information]

[Proprietary Information]

1 3 4Cs (Proprietary Information]

[Proprietary Information]

1 34m Cs [Proprietary Information]

[Proprietary Information]

1 3 6Cs [Proprietary Information]

[Proprietary Information]

137 Cs [Proprietary Information]

[Proprietary Information]

1s s Eu [Proprietary Information]

[Proprietary Information]

1 s6 Eu [Proprietary Information] (Proprietary Information]

1s 1 Eu [Proprietary Information]

[Proprietary Information]

129 1 [Proprietary Information]

[Proprietary Information]

130I [Proprietary Information]

[Proprietary Information]

1 3 1 I [Proprietary Information]

[Proprietary Information]

1321 [Proprietary Information]

[Proprietary Information]

1 32 m I [Proprietary Information]

[Proprietary Information]

133J [Proprietary Information]

[Proprietary Information]

1 33 m I [Proprietary Information] (Proprietary Information]

1 3 41 [Proprietary Information]

[Proprietary Information]

135£ [Proprietary Information]

[Proprietary Information]

8 3 m Kr [Proprietary Information]

[Proprietary Information]

85 Kr [Proprietary Information]

[Proprietary Information]

85m Kr [Proprietary Information]

[Proprietary Information]

87 Kr [Proprietary Information]

[Proprietary Information]

8 8 Kr [Proprietary Information]

[Proprietary Information]

1 40 La [Proprietary Information]

[Proprietary Information]

141La [Proprietary Information]

[Proprietary Information]

1 42 La [Proprietary Information]

[Proprietary Information]

99 Mo [Proprietary Information]

[Proprietary Information]

95 Nb [Proprietary Information]

[Proprietary Information]

95mNb [Proprietary Information]

[Proprietary Information]

96 Nb [Proprietary Information]

[Proprietary Information]

1 3-61

..... .. NWMI ...... ... *.. **** .. .. .. " "NOmfWESTllEDICAllSOTOPE S NWMl-2 0 13-021, Rev. 1 Chapter 13.0 -A ccident Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay , hour s after EOI Stream description 97 mNb 14 1N d 23 6mN p 231 Np 23s Np 239 Np 2 33 pa 234 pa 23 4 m Pa 112pd 14 7 pm I4 8 pm 1 4 8m pm I4 9 pm 1 so pm 1s1pm I 42 pr I4 3 pr 1 44 pr I44m pr I45 pr 2 38 pu 239 pu 240 pu 2 41p u !03mRh 1 05 Rh 106Rh !06 m Rh 103Ru i os Ru 10 6 Ru 122 sb 1 24 Sb 125 Sb 126Sb 127 Sb 128 Sb Target dissolution

[Proprietary Inform at ion] Di sso l ver roduct [Proprietary Inform at ion] [Proprietary Information]

[Propri etary Inform at ion] [Proprietary Information]

[Proprietary Inform at ion] [Proprietary Information]

[Propri eta ry Inform atio n] [Proprietary Information]

[Propri etary Inform atio n] [Proprietary Information]

[Propr ietary In formation] [Proprietary Information]

[P roprietary Inform atio n] [Proprietary Information]

[Propriet ary Inform a tion] [Proprietary Information]

[Propriet ary Inform ation] [Proprietary Information]

[Proprietary In formatio n] [Proprietary Information]

[Proprietary In formatio n] [Proprietary Information]

[Propri e tary Inform atio n] [Proprietary Information]

[Proprietary Inform a t io n] [Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Proprietary In format i o n] [Proprietary Information]

[Propriet ary Informati o n] [Proprietary Information]

[Propriet a ry Inform a tion] [Proprietary Information]

[Propri etary Inform a tion] [Proprietary Information]

[Proprietary In format ion] [Proprietary Information]

13-62 Uranium recovery and recycle [Proprietary Information]

Uranium se ara tion feed [Propri etary Information]

[Proprietary Information]

[Propri etary Inform at ion] [Proprietary Information]

[Proprietary Inform a tion] [Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Inform ation] [Proprietary Information]

[Propri etary Inform at i on] [Proprietary Information]

[Propr ietary Inform ation] [Proprietary Information]

[Propri etary Inform at ion] [Proprietary Information]

[Propri etary Inform ation] [Proprietary Information]

[Proprietary Inform ation] [Proprietary Information]

[Proprietary In fo rm ation] [Proprietary Information]

[Propri etary Inform ation] [Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Proprietary Inform ation] [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Information]

[Proprietary Information]

[Propri etary Inform atio n] [Proprietary Information]

[Proprietary Inform atio n] [Proprietary Information]

..... .. NWMI ::.**.*.*. .*.******** ' . NOmfWln MEDICAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Deca y, hour s after E OI Stream description Target dissolution

[Propri etary Information]

Di sso lver product Uranium recovery and recycle [Propri etary Inform ati on] Uranium se paration feed Isotope Bounding concentration (Ci/L) Bounding concentration (Ci/L) 1 2s msb [Propri etary Inform atio n] [Proprietary Inform ation] 129Sb [Proprietary Information]

[Proprietary Information]

1 s 1sm [Proprietary Information]

[Proprietary Inform ation] 153 Sm [Proprietary Information]

[Proprietary Information]

1 s6 sm [Proprietary Inform a tion] [Proprietary Information]

s 9 sr [Proprietary Information]

[Proprietary Information]

9o sr [Proprietary In forma tion] [Propri etary Inform at ion] 9'Sr [Proprietary Information]

[Proprietary Information]

92 Sr [Proprietary In format ion] [Propri etary Inform at ion] 9 9 Tc [Proprietary Information]

[Proprietary Informat i on] 99m Tc [Propr i etary In formation] [Propri etary Inform atio n] 125m Te [Proprietary Information]

[Proprietary Information]

1 21 Te [Proprietary Inform atio n] [Proprietary Inform at ion] I27m Te [Proprietary Information]

[Proprietary Information]

1 29 Te [Propri etary Inform at ion] [Propri eta ry Inform atio n] 129m re [Proprietary Information]

[Proprietary Information]

1 3 1Te [Propri e t ary Informati o n] [Proprietary Inform atio n] 13Im Te [Proprietary Information]

[Proprietary Information]

132 Te [Propri e t a ry In fo rm atio n] [Proprietary Inform ation] 133 Te [Proprietary Information]

[Proprietary Information]

1 33 mTe [Propri e t a r y Inform atio n] [Proprietary Inform ation] t 3 4Te [Proprietary Information]

[Proprietary Information]

23 1T h [Proprietary Inform atio n] [Proprietary Inform ation] 234 Th [Proprietary Information]

[Proprietary Information]

232 u [Propriet a r y In formatio n] [Proprietary Information]

234u [Proprietary Information]

[Proprietary Information]

23 su [Proprietary In formation]

[Proprietary Inform at i on] 236u [Proprietary Information]

[Proprietary Information]

231 u [Proprietary Information]

[P roprietary Inform ation] 23su [Proprietary Information]

[Proprietary Information]

1 3 1mxe [Propri etary In fo rm ation] [Proprietary Inform at i o n] 133 Xe [Proprietary Information]

[Proprietary Information]

1 33 mxe [Propri etary Inform a tion] [Proprietary Information]

135 Xe [Proprietary Information]

[Proprietary Information]

t 35 mxe [Proprietary Inform atio n] [Propri etary Inform at ion] 89m y [Proprietary Information]

[Proprietary Information]

90 y [Proprietary inform ation] [Propri etary Inform atio n] 90m y [Proprietary Information]

[Proprietary Information]

9 1y [P ropr i etary In format i o n] [Propri etary Information]

13-63

...... NWMI ..**.. ..* **.* ........ *.* " "NomtWESTUEDK:AltsOTOP£S NWM l-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-23. Bounding Radionuclide Liquid Stream Concentrations (4 pages) Unit operation Decay , hours after EOI n y 9 3y 93 zr 95zr 9 7 zr Totals Target dissolution

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

Uranium recovery and recycle [Proprietary Information]

Uranium se aration feed [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

S o urc e: Table 2-1 ofNWMI-201 3-C AL C-011 , So ur ce T e rm C al c ulati o n s , R e v. A , Northwe st M e di cal I s otope s, LL C , Co rvallis , Or eg on , F e bru a ry 2015. E OI = e nd of irradi a ti o n. In each case , a jacketed vessel is assumed to be filled with process solution appropriate to the process location, with the process offgas ventilation system operating. A level monitoring system will be available to monitor tank transfers and stagnant store volumes on all tanks processing LEU or fission product solutions. The source term used in this analysis is from NWMI-2013-CALC-011.

The breakdown of the radionuclide inventory used in NWMI-2013-CALC-011 is extracted from NWMI-2013-CALC-006 using the reduced set of 123 radioisotopes. NWMI-2014-CALC-014 identifies the 123 dominant radioisotopes included in the MURR material balance (NWMl-2013-CALC-006). NWMI-2014-CALC-014 provides the basis for u s ing the 123 radioisotopes from the total list of 660 radioisotopes potentially present in irradiated targets. The majority of omitted radioisotopes exist in trace quantitie s and/or decay swiftly to stable nuclide s. The reduced s et of 123 radioisotopes consists of those that dominate the radioactivity and decay heat of irradiated targets. 13.2.4.2 Identification of Event Initiati n g Conditions The accident initiating event is generally described as a process equipment failure. The PHA identified similar accident sequences in four nodes associated with leaks of enriched uranium solution into heating and/or cooling coils surrounding safe-geometry tank s or vessels. The PHA identified predominately corrosive degradation of the tank or overpressure of the tank as potential cause s that might damage this interface and allow enriched uranium solution to leak into the cooling system media or into the steam condensate for the heating s ystem. The primary containment fails , which allows radioactive or fissile solutions to enter an auxiliary s ystem. Radioactive or fissile solution leaks across the mechanical boundary between a process vessel and as s ociated heating/cooling jacket into the heatin g/cooling media. Where heatin g/cooling jackets or heat exchangers are used to heat or cool a fissile and/or high-dose proces s solution , the potential exists for the barrier between the two to fail and allow fissile and/or high-dose process solution to enter the auxiliary system. If the auxiliary system is not designed with a safe-geometry configuration , or if this system exits the hot cell containment , confinement , or shielding boundary in an uncontrolled manner , either an accidental criticality is po s sible or a high-dose to workers or the public can occur. 13-64

..... ;. NWMI ...... ..* ... **** .. .. .. * *. * ! 0 NORTHWEST llEDtCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Where auxiliary services enter process solution tanks , the potential exists for backflow of high-dose radiological and/or fissile process solution into the auxiliary service systems (purge air , chemical addition line , water addition line , etc.). Since these systems are not designed for process solutions , this event can lead to either accidental nuclear criticality or to high-dose radioactive exposures to workers occupying areas outside the hot cell confinement boundary. 13.2.4.3 Description of Accident Sequences The PHA made no assumption about the geometry or the extent of the heatin g/cooling subsystem. Consequently , an assumption is made that without additional control , a credible accidental nuclear criticality could occur , as the fissile solution enters into the heatin g/cooling system not designed for fissile s olution , or as the solution exits the shielded area and creates a high worker dose consequence.

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

The accident sequence for a tank leak into the cooling water (or heating) system includes the following.

The process vessel wall fails and the tank content s leak into the cooling jacket and medium , or the process medium leaks into the ves s el. Tank liquid level monitoring and liquid level in s trumentation are functional

however , dependin g on the size of the leak , the tank level instrumentation may or may not detect that a tank has leaked. The cooling water system monitor (conductivity or pH) detects a change in the cooling water , and an alarm notifies the operator.

The operator place s the system in a safe configuration and trouble s hoots the source of the leak. Maintenance activities to identify , repair , or replace the cause of the leak are initiated after achieving the final s table condition. Additional PHA accident sequences include the back.flow (siphon) or backup of process solution s into the chemical or water addition s y s tems. The controls for these accidents are de s cribed in Section 13.2.4.8. 13.2.4.4 Function of Components or Barriers This accident sequence requires the failure of the primary confinement in a safe-geometry vessel or tank , the normal condition criticality safety control for the process. This same barrier will provide primary containment of the high-dose process solution to maintain the solution within the hot cell containment , confinement , and shielding boundary. The heating and cooling systems will have secondary loops (closed loops), so a s econd failure i s required for the fissile solution to enter into a non-geometric-safe auxiliary system or into a non-shielded auxiliary system out of the hot cell s. 13.2.4.5 Unmitigated Likelihood Leaks into auxiliary services can be initiated by mechanical failure of equipment boundaries between the process solutions and auxiliary system fluids, or backflow of high-dose radiological or fissile solution to a chemical supply system. Failure rates for tanks , vessel s, pipes , and pumps are estimated from WSRC-TR-93-262.

Table 13-2 shows qualitative guidelines for applying the likelihood categories.

Failures resulting in leaks or back.flows as initiating events are estimated to have an unmitigated likelihood of" not unlikely." 13-65

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

• NOmfWEJT MEDICAi. tSOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Additional detailed information describing a quantitative evaluation , including asswnptions, methodology , uncertainties , and other data , will be developed for the Operating License Application.

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

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

Solution leaks into the cooling or heating system were bounded by the irradiated target dissolver after dissolution is complete.

The target dissolution process produces an aqueous solution containing uranyl nitrate, nitric acid , and fission products.

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

13.2.4. 7 Evaluation of Potential Radiological Consequences The following evaluations are based on information developed for the Construction Permit Application. Additional detailed information describing radiological consequences will be developed for the Operating License Application.

13.2.4.7.1 Direct Exposure Consequences The potential radiological exposure hazard ofliquid spills discussed in Section 13 .2.2 bound the consequences from radiation exposure for these accident sequences.

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

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

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

13.2.4.7.2 Confinement Release Consequences Not applicable to this accident sequence. 13.2.4.8 Identification of Items Relied on for Safety and Associated Functions Hot ce ll shielding is designed to provide protection from leaks into the heating and cooling closed loop auxiliary systems that result in redistribution of high-dose uranium solutions in the hot cell. From a direct exposure perspective , this type of accident does not represent a failure or adverse challenge to the hot cell shielding boundary function. 13-66

• .**.*.*. .; ... NWMI ........ *.* ' !*.

• 0 NORTHWEST MEOICAl lSOTOP£S 13.2.4.8.1 IROFS RS-04, Hot Cell Shielding Boundary NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis IROFS RS-04 functions to prevent worker dose rates from exceeding exposure criteria due to the presence of radioactive materials in the hot cell ve s sels before or after a leak to the cooling and heating auxiliary systems. As a PEC and safety feature , the hot cell shielding boundary will provide an integrated system of features that protect workers from the high-dose radiation generated during radioisotope processing.

A primary safety function of the hot cell shielding boundary will be to reduce the radiation dose at the worker/hot cell interface to ALARA. While protecting workers , the shield will also protect the public at the controlled area boundary.

The hot cell shielding boundary will provide shielding for workers and the public during normal operations to reduce worker exposure to an average of 0.5 mrem/hr , or less , in normally accessible workstations and occupied areas outside of the hot cell.1 The hot cell shielding boundary will also provide shielding for workers and the public during process upsets to reduce worker exposure to a TEDE of 5 rem , or less , at workstations and occupied area s outside of the hot cell.2 As a PEC , shielding will be provided by a thick concrete , steel-reinforced wall with steel cladding that reduces the normally expected operational exposures from within the boundary to an average of 0.5 mrem/hr , or less , outside of the boundary. Where direct visual access is required , leaded-glass windows with appropriate thicknesses will be used to reduce normally expected operational exposures from within the boundary to an average of 0.5 mrem/hr , or less , outside of the boundary. Some shielding will be movable , such as around the high-dose waste ca s k loading area. Where penetrations are required , the engineered design pro v ides for access-controlled , non-occupied corridors or airlocks where potential radiation streaming is safely mitigated by multiple layers of shielding or through a torturous path. The shielding is also designed to reduce the exposure from postulated upsets within the hot cell shielding boundary to less than a low consequence exposure to workers and the public of 5 rem , or less , per incident.

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

Each shielded protected area will be operable when the equipment in that area i s in the operating or standby modes. 13.2.4.8.2 IROFS CS-06, Pencil Tank and Vessel Spacing Control using the Diameter of the Tanks, Vessels, or Piping All tanks, vessels , or piping systems involved in a proce s s upset will be controlled with a safe-geometry confinement IROFS that consists ofIROFS CS-06 to provide a diameter of the v e ssels confinement or IROFS CS-26 to provide safe volume confinement.

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

1 Some op e ration s m ay ha v e hi g h e r do ses during s hort p e ri o ds o f th e op e ration. Th e ave ra ge wor k e r ex po s ure r a t e i s d es i g n e d to be 0.5 mr e m/hr , or l ess. Ar e a s n o t normall y ac cess ibl e b y th e worker m ay h ave hi g h e r do se ra t es (e.g., s tr e amin g radi a tion a round n o rmall y ina c c ess ibl e r e mot e man i pul a tor p e n e tration s well abo v e th e w o rker's h ea d). 2 The s hie l din g i s not cr e dit e d fo r mitig a tin g do se rat es during an a cc idental nucl e ar criti ca lity; in s t ea d , additional IROFS a r e identified to pro v id e doubl e-c ontingenc y protecti o n to pre v ent (reduc e the lik e l ih oo d o f) a n accid e ntal nuclear criticality.

13-67

...... ; ... NWMI ..*... ..* .... ........ *. * !*:! ." NOmfWUT MEDICAL ISOTOP'U NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The heat exchanger materials will be compatible with the harsh chemical environment of the tank or vessel process (this may vary from application to application). Sampling of the heating or cooling media (e.g., steam condensate conductivity , cooling water radiological activity , uranium concentration , etc.) 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 according to IROFS CS-16 and CS-17 prior to discharge to a non-safe geometry. 13.2.4.8.4 IROFS CS-27, Closed Heating or Cooling Loop with Monitoring and Alarm As a PEC, on the evaporator or concentrator condensers , a closed cooling loop with monitoring for breakthrough of process solution will be provided to contain process solution that leaks across this boundary , ifthe boundary fails. IROFS CS-27 is applied to those high-heat capacity cooling jackets (requiring very large loop heat exchangers) servicing condensers where the leakage is always from the cooling loop to the condenser , reducing back-leakage, and the risk of product solutions entering the condenser is very low by e v aporator or concentrator design. The purpose of this safety function is to monitor the condition of the condenser cooling jacket to ensure that in the unlikely event that a condenser overflow occurs , fissile and/or high-dose process solution will not flow into this non-safe geometry cooling loop and cau s e nuclear criticality. The closed loop will also isolate any high-dose fissile product solids (from the same event) from penetrating the hot cell shielding boundary , and any high-dose fission gases from penetrating the hot cell shielding boundary during normal operations. The heat exchanger materials will be compatible with the harsh chemical environment of the tank or vessel process (this may vary from application to application).

Sampling of the cooling media (e.g., cooling water radiological activity , uranium concentration , etc.) will be conducted to alert the operator that a breach has occurred and that additional corrective actions are required to identify and isolate the failed component and to restore the c lo sed loop integrity.

Closed loop pressure will also be monitored to identify a leak from the closed loop to the process system. Discharged solutions from this system will be handled as potentially fissile and sampled according to IROFS CS-16 and CS-17 prior to discharge to a non-safe geometry. 13.2.4.8.5 IROFS CS-20, Evaporator or Concentrator Condensate Monitoring As an AEC, the condensate tanks will use a continuously active uranium detection system to detect high carryover of uranium that shuts down the evaporator feeding the tank. The purpose of this system is to (I) detect an anomaly in the evaporator or concentrator indicating high uranium content in the condenser (due to flooding or excessive foaming), and (2) prevent high concentration uranium solution from being available in the condensate tank for discharge to a non-favorable geometry system or in the condenser for leaking to the non-safe geometry cooling loop. The safety function of this IROFS is to prevent an accidental nuclear criticality.

The detection system will work by continuously monitoring condensate uranium content and detecting high uranium concentration , and then shutting down the evaporator to isolate the condensate from the condenser and condensate tank. At a limiting setpoint , the uranium monitor-detecting device will close an isolation valve in the inlet to the evaporator (or otherwise secure the evaporator) to stop the discharge of high-uranium content solution into the condenser and condensate collection tank. The uranium monitor is designed to produce a valve-open permissive signal that fails to an open state , closing the valve on loss of electrical power. The isolation valve is designed to fail-closed on loss of instrument air , and the solenoid is designed to fail-closed on loss of signal. The locations where this IROFS is used will be determined during final design. 13-68

...... ;* .. NWMI ..*... ..* .... ........ *.* NOtmMmM£DICALtsoTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.4.8.6 IROFS CS-18, Backflow Prevention Device As a PEC or AEC, chemical and gas addition ports to fissile process solution systems will enter through a backflow prevention device. This device may be an anti-siphon break, an overloop seal, or other active engineering feature that addresses the conditions of backflow and prevents fissile solution from entering non-safe geometry systems or high-dose solutions from exiting the hot cell shielding boundary in an uncontrolled manner. The safety function of this IROFS is to prevent fissile solutions and/or high-dose solutions from backflowing from the tank into sys tems that are not designed for fissile solutions that could lead to accidental nuclear criticality or to locations outside of the hot cell shielding boundaries that might lead to high exposures to the worker. Each hazardous location will be provided an engineered backflow prevention device that provides high reliability and availability for that location. The backflow prevention de v ice features for high-do se product solutions will be located inside the hot cell shielding and confinement boundaries ofIROFS RS-04 and RS-01 , respectively. The feature is designed such that spills from overflow are directed to a safe geometry confinement berm controlled by IROFS CS-08 (described in NWMI-20 l 5-SAFETY-004, Quantitati ve Risk Analysis of Pro cess Upsets Assoc iat ed with Pas sive Engineering Controls Leadin g to Criticality Accident S e quences , Section 3.1.6.3) or to safe-geometry tanks controlled by IROFS CS-11. 13.2.4.8.7 IROFS CS-19, Safe Geometry Day Tanks As a PEC , safe-geometry day tanks will be provided where the first barrier cannot be a backflow prevention device. The safety function of this PEC is to prevent accidental nuclear criticality by providing a safe-geometry tank if a fissile s olution backs-up into an auxiliary chemical addition system. IROFS CS-19 will be used where conventiona l backflow prevention in pressurized systems i s not reliable. The safe-geometry day tank will be provided for those chemical addition activities where the reagent cannot be added via an anti-siphon break since the tank or vessel is not vented and operates under some backpressure conditions. The feature works by providing a safe-geometry vessel that is filled with chemical reagent using the conventional backflow prevention devices , and then provides a pump to add the reagents to the respective process system under pressure. Safe-geometry day tanks servicing dose product solutions systems will be located in the hot cell shielding or confinement boundaries of IROFS RS-04 and RS-01 , respectively.

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

All tanks wi ll be vented and unpressurized under normal u se. The heating and cooling systems will operate at pressures that are higher than the processing systems that they heat or cool. The majority of system leakage would typically be in the direction of the heat transfer media to the processing system. All vented tanks are designed with level indicators that are available to the operator to detect the level of solution in a tank remotely.

Operating procedures will identify an operational high-level fill operating limit for each tank. As part of the level detector , a high-level audib le a larm and li ght will be provided to indicate a high lev el above this operating limit so that the operator can take action to correct conditions leading to failure of the operating limit. With batch-type operation with typically low volume transfers , the sizing of the tanks will include sufficient overcapacity to handle reasonable perturbations in operations caused by variations in chemical concentrations and operator errors (adding too much). 13-69

.. ; ... ; .. NWMI ...*.. ..* .... ............ ' ! ' NOWTHWEn MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Tank and vessel walls will be made of corrosion-resistant materials and have wall thicknesses that are rated for long service with harsh acid or basic chemicals. Purge and gas reagent addition lines (air, nitrogen , and oxygen) will be equipped with check valves to prevent flow of process solutions back into uncontrolled geometry portions (tanks, receivers , dryers , etc.) of the delivery system. 13.2.4.9 Mitigated Estimates The controls selected will mitigate both the frequency and consequences of this accident.

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

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

I. Electrical power to the RPF is lost due to an initiating event. 2. The uninterruptible power supply automatically provides power to systems that support safety functions , protecting RPF personnel and the public. The following systems are supported with an uninterruptible power supply: Process and facility monitoring and control sys tems Facility communication and security systems Emergency lighting Fire a l arms

• Criticality accident alarm systems Radiation protection sys tems 3. Upon loss of power , the following actions occur: Inlet bubble-tight isolation dampers within the Zone I ventilation system c l ose, and the heating, ventilation , and air conditioning (HV AC) system is automatically placed into the passive ventilation mode of operation. Process vessel ve nt system is automatically placed into the passive ventilation mode of operation , and all electrical heaters cease operation as part of the passive operation mode. Pressure-r elief confinement system for the target dissolver off gas system is activated on reaching the system re l ief setpoint , and dissolver off gas is confined in the off gas piping , vessels, and pressure-relief tank (IROFS RS-09). 13-70

---

---

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

• NOITHWEST MEDtCAl tSOTOP£S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Process vessel emergency purge system is activated for hydrogen concentration contro l in tank vapor spaces (IROFS FS-03). Uranium concentrator condensate transfer line va l ves are a utom atically con figur ed to return condensate to the feed tank due to residual heating or coo l ing potential for t ransfer of process fluids to waste tanks (IROFS CS-l 4/CS-15). All equipment providing a motive force for process activities cease, including:

Pumps performing liquid transfers of process solutions Pumps supporting operation of the steam and cooling utility h eat transfer fluids Eq uipm ent supporting physica l transfer of items (primari l y cranes) 4. Operators follow alarm response procedures. 5. The facility is now in a stable condition. 13.2.5.4 Function of Components or Barriers All faci li ty structural components of the hot ce ll secondary confinement boundary (in a passive venti lation mode) and hot cell s hield ing boundary (walls , floors, and ceilings) will remain intact and functional.

The engineered safety features requiring power wi ll activate or go to their fail-safe config u ration. 13.2.5.5 Unmitigated Likelihood Loss of power can be initiated by off-site events or mechanical failures of equipment.

Fai lur es resulting in l oss of power as initiating events are estimated to have an unmitigated lik elihood of "not unlikely." Additional detailed information describing a quantitative eva lu ation , including assumptions , methodology , uncertainties, and other data, wi ll be developed for the Operating License Application.

13.2.5.6 Radiation Source Term The lo ss of power eva lu ation is based on information developed for the Construction Permit Application.

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

13.2.5. 7 Evaluation of Potential Radiological Consequences The lo ss of power eva lua tion is based on information developed for the Construction Permit Application.

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

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

A standby diesel generator will be available at the RPF. 13-71

.. ; ...

........... * ........... . * * .. N09'TMWHT MEOtCAL lSOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.6 Natural Phenomena Events Chapter 2.0 , " Site Characteristics," and Chapter 3.0 discuss the design of SSCs to withstand external events. The RPF is designed to withstand the effects of natural phenomena events. Consequences of natural phenomena accident sequences have been evaluated.

Sections 13.2.6.1 through 13.2.6.6 provide event descriptions and identify any additional controls required to protect the health and safety of workers , the public , and en v ironment.

13.2.6.1 Tornado Impact on Facility and Structures, Systems, and Components The adverse impact of a tornado on facility operations has a number of facets that must be evaluated. This evaluation addresses the facility design as impacted by the maximum-sized tornado with a return frequency of I0-5/year (yr). High winds can lead to significant damage to the facility structure.

Damage to the structure is a function of the strength of the tornado winds , duration , debris carried by the winds , direction of impact , and facility design. This evaluation determines the impact of tornado winds on the facility from a design basis perspective to ensure that the design prevents impact to SSCs in the building.

The local area impact may result in loss of utilities (e.g., electrical power) and reduced access by local emergency responders.

Loss of power i s evaluated (Section 13.2.5) as a potential cause for all adverse events. The individual PHA nodes evaluate the loss of site power and loss of power distribution to each s ubsystem.

High winds may directly impact SSCs important to safety (e.g., components of the fire protection system are located in areas adjacent to the building) and reduce the reliability of those SSCs to respond to additional events (like loss of electrical power) that can be initiated concurrently with the tornado (either as an indirect result or as an additional random failure). This evaluation analyzes the impact of tornado winds on these SSCs. Tornado impact on the facility structure

-High wind pressures could cause a partial or complete collapse of the facility structure , which may cause damage to SSCs important to safety or impact the availability and reliability of those SSCs. A partial or complete structural collap s e may also lead directly to a radiological or chemical release or a potential nuclear criticality , if damage caused by the collapse c reates a violation of criticality spacing requirements.

Tornado wind-driven missiles could penetrate the facility building envelope (walls and roof), impacting the availability and reliability of SSCs important to s afety , or may lead directly to a radiological or chemical release. Tornado impact on SSCs important to safety located outside the main facility -High wind pressures and tornado wind-driven missiles could damage SSCs important to safety located outside the RPF building envelope.

The damage sustained may impact the availability and reliability of the SSCs important to safety. Loss of site power may affect the ability of SSCs important to safety located within the facility building envelope to respond to additional events. A partial or complete collapse of the facility structure could also lead directly to an accident with adverse intermediate or high consequences. The only IROFS located outside the RPF building envelope is the exhaust stack. Buckling or toppling of the exhaust stack would affect its ability and availability to mitigate other events with intermediate consequences.

The return frequency of the design basis tornado is 10-5/yr , making the initiating event highly unlikely.

No additional IROFS are required.

13-72

.

......... *.* ......... *.* .. *****. *. . *

• NORTKWtn MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.6.2 High Straight-Line Winds Impact the Facility and Structures, Systems, and Components Similar to the tornado, high straight-line winds can also damage the facility structure, which in turn can lead to damage to SSCs relied on for safety. This evaluation demonstrates how the facility design addressed straight-line winds with a return interval of 100 years or more , as required by building codes. Buckling or toppling of the exhaust stack would affect the ability and availability to mitigate other events with intermediate consequences.

A partial or complete collapse of the facility struc ture may a lso lead directly to an accident with adverse intermediate or hi g h consequences.

The facility is designed as a Risk Category IV structure, a standard industrial facility with equivalent chemical hazards, in accordance with American Society of Civil Engineers (ASCE) 7 , Minimum Design Loads for Buildings and Other Structures.

The return frequency of the basic (design) wind speed for Ri sk Category IV structures i s 5.88 x l0-4/yr (mean return interval, MRI= 1 , 700 yr). At this return frequency , the straight-line wind event is not likely but credible during the de sign life of the facility and is considered in the structural design as a severe weather event. The provision s of the ASCE 7 s tandard , when used with companion standards suc h as American Concrete Institute (ACI) 318, Building Code R equirements for Structural Concrete, and American Institute of Steel Construction (AISC) 360, Specification for Structural Steel Buildin gs, are written to achieve the target maximum annual probabilities of established in ASCE 7. The highest maximum probability of failure, which is for a failure that is not sudden and does not lead to a wide-spread progression of collapse , targeted for Risk Category IV s tructures is 5.0 x I o-6. Therefore , the likelihood of failure of the structure when subjected to the design basi s straight-line wind in conjunction with other loads , as required by ASCE 7, is highly unlikely.

No additional IROFS are required.

13.2.6.3 Heavy Rain Impact on Facility and Structures, Systems, and Components Localized heavy rain can overwhelm the structural integrity of the RPF roofing system. This evaluation determines the impact of probable maximum precipitation (PMP) in the form of rain on the roofing s tructure.

The PMP is defined as " theoretical greatest depth of precipitation for a given duration that is physically possible over a particu l ar drainage area at a certain time of year." In other words, the PMP represents the theoretical worst-case of the most precipitation the atmosphere is capable of discharging to a particular area over a selected period of time. The PMP is based on an empirical methodology with no defined annual exceedance probability. For impact on the facility, the PMP for 25 .9 square kilometers (km 2) (10 square mi Jes [ mi 2]) is evaluated.

Large amounts of rain water accumulating on the roof could lead to collapse of the roof. A partial or complete collapse of the facility roof may impact the availability or reliability of SSCs relied on for safety located within the RPF building envelope to respond to other events of high consequence.

From the National Weather Service (NWS)/National Oceanic and Atmospheric Administration (NOAA) Hydrometeorological Report No. 51, Probable Maximum Precipitation Estimates , United States East of the 105th Meridian, the PMP is defined as " theoretical greatest depth of precipitation for a given duration that is physically possible over a particular drainage area at a certain time of year." In other words, the PMP represents the theoretical worst-case of the most precipitation the atmosphere is capable of discharging to a particular area over a selected period of time. The PMP is based on an empirical methodology with no defined annual exceedance probability. Although the NWS/NOAA has historically sta ted that it is not possible to assign an exceedance probability to the PMP (NOAA Technical Report NWS 25, Comparison of Generalized Estimates of Probabl e Maximum Pr eci pitation with Greatest Observed Rainfalls), several academic studies and papers have undertaken the exercise to determine the annual exceedance probability for PMP using modern probabilistic techniques and storm modeling and have found that the exceedance probability varies by location but is quite low (NAP, 1994). As such, the PMP event has been determined to be highly unlikely. 13-73

.. .. NWMI ...... ..* .... ........ *.* , ' ", NOmfWEST MlOtcAL ISOTOP£S No additional IROFS are required.

NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The roof of the RPF is designed to prevent rainwater from accumulating on the roof. In accordance with 2012 International Building Code (IBC) and ASCE 7, the roof structure is designed to safely support the weight of rainwater accumulation with the primary drainage system blocked and the secondary drainage system at its design flow rate when subjected to a rainfall intensity based on the 100-yr hourly rainfall rate. Deflections of roof members are limited to prevent rainwater ponding on the roof. The roof structure is also designed to support the extreme winter precipitation load discussed in Section 13.2.6.6.

13.2.6.4 Flooding Impact to the Facility and Structures, Systems, and Components Regional flooding from large precipitation events raising the water levels of local streams and rivers to above the 500-yr flood level can have an adverse impact on the structure and SSCs within. These impacts include the structural damage from water and the damage to power s upplies and instrument control systems for SSCs relied on for safety. The infiltration of flood water into the facility could cause the failure of moderation control requirements and lead to an accidental nuclear criticality. Direct damage or impairment of SSCs could also be caused by flooding in the facility. The site will be graded to direct the stormwater from localized downpours with a rainfall intensity for the 100-yr storm for a 1-hr duration around and away from the RPF. Thus , no flooding from local downpours is expected based on standard industrial design. Rainwater that falls on the waste management truck ramp and accumu l ates in the trench drain has low to no consequence for radiological, chemical , and criticality hazards. Situated on a ridge , the RPF will be located above the 500-yr flood plain according to the flood insurance rate map for Boone County , Missouri , Panel 295 (FEMA, 2011). The site is above the elevation of the nearest bodies of water (two small ponds and a lake), and no dams are located upstream on the local s treams. This data conservatively provides a 2 x 10-3 year return frequency flood , which can be considered an unlikely event according to performance criteria. However , the site is located at an elevation of 248.4 m (815 ft), and the 500-year flood plain starts at an elevation of 231.6 m (760 ft), or 16.8 m (55 ft) below the site. Since the site is located only 6.1 m (20 ft) below the nearest high point on a ridge (re l ative to the l ocal topography), is well above the beginning of the 500-yr flood plain , and is considered a dry site , the probable maximum flood from regional flooding is considered highly unlikely , without further evaluation

.3 No additional IROFS are required. 13.2.6.5 Seismic Impact to the Facility and Structures, Systems, and Components Beyond the impact on the facility structure and the potential for falling facility components impacting SSCs or direct damage to SSCs causing adverse events , other activities were identified as sensitive to seismic events. During the irradiated target shipping cask unloading preparations , the shield plug fasteners will be removed from an upright cask before mating the cask to the cask docking port. During the short period between that activity and installing the cask, a seismic event could dislodge the lift/cask combination and result in dislodging the shield plug in the presence of personnel.

This event would result in potentially lethal doses to workers in a short period of time. Seismic ground shaking can directly damage SSCs relied on for safety or lead to damage of the facility, including partial or complete collapse, which could impact SSCs relied on for safety inside and outside the RPF. Damage to the facility could also impact SSCs , causing radiological and chemical releases of intermediate consequence. 3 Th e recomm e nded st a ndard for determinin g th e probably m a ximum flood , AN S 2.8 , D e t e rmin i n g D e si g n Ba s i s Floodin g at P owe r R e a c t o r S it es , h a s b ee n withdrawn. 13-74

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

• 0 NOmlWEST MEDICAL tsoTOPES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Leaks of fissile solution, co mpromising the safe-geometry and safe interaction storage in solid material storage arrays and pencil tanks or vessels containing enriched uranium solutions , could lead to a criticality accident , a high consequence accident.

NWMI-20 l 5-SAFETY-004 , Section 3.1 , identifies IROFS to prevent and mitigate this accident scenario. Dislodging the irradiated target shipping cask during unloading preparations could expose workers to a potentially lethal radiation dose. This event is considered a high consequence accident.

The safe-shutdown earthquake , or design basis earthquake , for the RPF is specified as the risk-targeted maximum considered earthquake (MCER), as determined in accordance with ASCE 7 and Federal E mergency Management Agency (FEMA) P-753 , N EHRP R ec ommended Sei s mi c Provisions for New Buildings and Oth e r Stru ct ur es. The MCER for this site is governed by the probabilistic considered earthquake ground-shaking , which has a n annual frequency of exceedance of 4 x 10 4 (2 , 500-yr return period). This event is considered unlikely.

Using the provisions of ASCE 7 for standard industrial facilities with equivalent chemical hazards , Risk Category IV results in a design basis earthquake equal to the safe-shutdown earthquake s pecified. When designed in accordance with ASCE 7 and companion standards , the maximum probability of a com plete or partial structural failure is 3 percent conditioned on the occurrence of the considered earthquake ground-shaking , or a probabilit y of failure of 1.2 x 10-5. Therefore , failure of the facility subject to the maximum-considered earthquake ground-shaking is considered highly unlikely.

No credit can be taken for phy s ical features of the irradiated target cask lifting fixture for the unmitigated case; therefore , the unmitigated likelihood is equal to the annual probability of exceedance for the safe s hutdown earthquake, fearrhquak e = 4 x 10 4. 13.2.6.5.1 IROFS FS-04, Irradiated Target Cask Lifting Fixture As a PEC , the irradiated tar ge t cask lifting fixture will be designed to pre ve nt the cask from tipping within the fixture and prevent the fixture itself from toppling during a seismic event. 13.2.6.6 Heavy Snow Fall or Ice Buildup on Facility and Structures, Systems, and Components Thi s evaluation addresses s now loading on the facility s tructure.

The facility prot ects the SSCs , and an extreme snow-loading event may cause failure of the roof , impacting the SSCs' ability to perform associated safety function s. NRC DC/COL ISG-07 , Int e rim Staff Guidan ce on A ssess ment of Nor mal and Extre m e Winter Pr ec ipitation Loads on th e Roofs of S e ismic Category I Stru c tur es, provides guidance on the design of Category I structures for s now load that conservatively bound s the RPF. The normal snow load as defined in the NRC ISG is the 100-yr s nowpack , which is equivalent to th e de s ign s no w load for Risk Category IV s tructure s determined in accordance wit h ASCE 7. Co llapse of the roof may d a ma ge SSCs that are relied on for safety , leading to accident sequences s uch as accidental nuclear criticalit y (e.g., a pencil tank was crushed and interaction controls violated) or a radiological release (e.g., if a hot cell confinement boundary was breached and a primary confinement boundary dama ge d), or ma y prevent an SSC from bein g avai l able to perform it s function.

The extreme winter precipitation load , as defined in the NRC ISG , is a combination of the 100-yr snowpack and the liquid weight of the probable maximum winter precipitation.

The probable maximum winter precipitation i s based on the seasonal variation of the PMP , given in NWS/NOAA Hydrometeorological Report 53 , S e asonal Variation of JO-Squar e Mile Pr obab l e Maximum Precipitation Estimates , United Stat es East of th e JOS h M e ri dian, for winter month s. The PMP is defined in Section 13.2.6.3. Considering the extreme winter precipitation load is a co mbination of the 100-yr s nowpack and the theoretical worst-case precipitation event , the extreme w inter precipitation load is highly unlikel y. 13-75

...... .. NWMI ...... ... **: ........ *.* ' *. . NOITHWEST MEDICAL NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis The normal snow load is the 100-yr snowpack , which is equivalent to the design s now load for a Risk Category IV structure determined in accordance with ASCE 7. The return frequency of the normal snow load is relatively high and expected to be likely to occur during the design life of the facility. The pro v isions of the ASCE 7 standard , when used with companion standards such as ACI 318 and AISC 360, are written to achieve the target maximum annual probabilities of failure established in ASCE 7. The highe s t probability of failure , which is for a failure that is not sudden and does not lead to a wide-spread progression of collapse , targeted for Risk Category IV s tructures is 5.0 x 1 o-6. Therefore , the likelihood of failure of the structure when s ubjected to the normal design snow load in conjunction with other loads as required by ASCE 7 i s highl y unlikely. No additional IROFS are required.

13.2. 7 Other Accidents Analyzed A total of 7 5 accident sequences identified for further evaluation by the PHA were analyzed for the Construction Permit Application.

A summary of all accidents analyzed is provided in Table 13-24. This table includes the accidents evaluated in Section 13.2.2 to 13.2.6 for completeness. Table 13-24 lists each accident sequence number , a descriptive title of the accident , and IROFS identified (if needed) to prevent or mitigate the consequences of the accident sequence.

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

• IROFS RS-01, Hot Ce ll Liquid Confinement Bound ary

• IROFS RS-03 , Hot Ce ll Secondary Confi nement Boundary

• IROFS RS-04 , Hot Cell Shielding Boundary

• JRO FS CS-07 , P e n c il Tank a nd Vessel Spacing Control using Fixed Int eract ion Spacing oflndividual Tanks or Vessels

• IROFS CS-08 , Floor and Sum Geometry Control on Slab Depth , Sump Di a m e ter or Depth for Floor Spill Co nt a inment Berm s

• JROFS CS-09 , Doubl e-W a ll Pipin g

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

• Bounded by S.R.O 1 from primary offgas treatment solutions, resulting in radiological consequences S.R.03 Spray release of high-dose or

• Bounded by S.R.01 enriched uranium-containin g product s olution , resulting in radiological consequences 13-76

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

• NOmfWUT MEDICAL ISOTOP£1 NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.04 S.R.05 S.R.06 S.R.0 7 S.R.08 S.R.0 9 S.R.10 S.R.1 2 S.R.13 S.R.14 S.R.16 Liquid enters process vessel ventilation system damaging IRU or retention beds, releasing retained radionuclides High-dose so lu tion e nt ers th e UN bl e ndin g and storage tank High flow through IRU causing premature release of high-dose iodine gas Loss of t emperat u re co ntrol on the IRU l eading to r elease of high-dose iodine Loss of vacuum pumps Loss ofIRU o r ca rbon b e d media to d ownstream p art of the syste m Wrong retention media added to bed or saturated retention media Mo produ ct cask removed from the hot ce ll boundary wit h improper s hi e ld plug in sta llat ion High-dose containing solution leaks to chilled water or steam condensate system IX resin fai lur e due to wrong r eage nt o r hi gh temperature Backflow of high-dose radiological and/or fissile solution into auxiliary system (purge air , chemical addition line , water addition line , etc.)

• IROFS RS-09, Primary Offgas Relief System

• IROFS RS-03 , Hot Cell Secondary Confinement Boundary

• See Section 13.2.3.8

• Not c r edib le or l ow conseq u ence

• Bounded by S.R.04

• Bounded b y S.R.04

• Bounded by S.R.04

• Bound ed by S.R.04

• Event unlikely with intermediate consequence

• Eve n t unlikely with int e rm e di a t e co n se qu e nc e

• IROFS RS-04, Hot Cell Shielding Boundary

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

• IROFS CS-I 0 , Closed Safe-Geometry Heating or Cooling Loop with Monitoring and Alarm

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

• IROFS CS-20, Evaporator or Concentrator Condensate Monitoring

• IROFS CS-18, Backflow Prevention Device

• IROFS CS-19, Safe-Geometry Day Tanks

• See Section 13.2.4.8

• Bound ed by S.R.01

• Bounded by S.R.13 13-77 "NWMI ...... ..* .... .......... * ! *. * !

• NOATMWUT MEDK:Al tscnwn NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Accident sequence designator Table 13-24. Analyzed Accidents Sequences (9 pages) from PHA Descriptor Preliminary IROFS Identified S.R.17 S.R.18 S.R.19 S.R.20 S.R.21 S.R.22 S.R.23 S.R.24 S.R.25 S.R.26 S.R.27 S.R.28 Carryover of hi g h-do se so lution into condensate (a l ow-do se waste s tr eam) High-dose solution flows into the solidification media hopper High target b asket retrieva l dose rate Radiological spill of irradiated LEU target material in the hot cell area D a m age to th e hot cell wall providing s hi e ldin g Decay heat buildup in unprocessed LEU target material removed from targets leads to higher-dose radionuclide offgasing

• IROFS R S-08, Sample an d Analysis of Low Do se Waste Tanlc Do se R ate Prior to Transfer Outside the Hot Ce ll Shielded Boundary

• IROF S R S-I 0 , Active R a di a tion Monitoring and I so l ation of Low-Dose Waste Tran sfer

• See Section 1 3.2.7.1

• Low consequence event that does not challenge IROFS RS-04

• D es i g n evo l ve d after PHA , acc id e nt seq uence e limin ated

• Bounded by S.R.01

• Low co n seq u e n ce eve nt that does not damage shie ldin g function of IRO FS RS-04

• Low consequence event Off gas in g from irradiated target

• IROFS RS-03, Hot Cell Secon d ary Confineme nt Bound a ry dissolution tank occ ur s when the

• See Section 1 3.2.2.8 upper valve is opened Bagless transport door failure ----H E PA filter failure Failed negative air balance from zone-to-zone or failure to exhaust a radionuclide buildup in an area Exte nd e d outage of heat l ea din g to freezing , pipe fai lur e , an d release ofradionuclides from liquid proc ess syste ms

• IROFS RS-03, Hot Cell Secondary Confinement Boundary

• IROFS RS-04, Hot Cell Shielding Boundary

• See Section 13 .2.2.8

• IROFS RS-03 , Hot Cell Secondary Confi n e m e nt Boundary

• See Section 13.2.2.8

• IROFS RS-03, Hot Cell Secondary Confinement Boundary

• See Section 13.2.2.8

• Highly unlikely event for process solutions con t aining fissio n products

• Bound ed by S.C.04 for target fa bri cation sys t e m s Target or waste shipping cask or

• Information will be provided in the Operating License container not loaded or secured Application according to procedure, leading to personnel exposure 13-78

..... .. NWMI ...*.. ... *.. ....... ' *: NOmfWUT MEDJCAl tsOTOPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident A nalysis Table 13-24. Ana l yzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.R.2 9 S.R.30 S.R.3 1 S.R.32 S.C.01 S.C.02 S.C.0 3 S.C.04 Hi gh d ose t o wo rk e r from re l ease of gaseo u s r a di o nuclid es durin g cas k rece ipt in s p ec ti o n a nd pr e p arat i on for ta r get b asket re m ova l Cask docking port failures lead to high-dose to worker due to streaming radiation and/or high airborne radioactivity C h e mi ca l bum s fro m co nt a mi nate d so lu t i o n s durin g sa mple a n a l ys i s Crane load drop accid e nts Fa ilur e of fac ili ty e nri c hm e nt limit Failure of administrative control on mass (batch limit) during handling of fresh U , scrap U , LEU target material , targets , and samples Fa ilu re of i n t erac ti o n limit durin g h a ndlin g of fr es h U , scra p U , L EU t a r get m a t e ri a l , t a r ge t s, co nt a in e r s , a nd s a mpl es Spill of process solution from a tank or process vessel leading to accidental criticality

• IRO FS R S-1 2, Cas k C ont a inm e nt Sa mplin g Prior t o C l os ure Lid R e m ova l

• IRO FS R S-1 3 , Cas k Loca l Ve ntil at i on Durin g C l os ur e Lid R e m ova l a nd D oc kin g Pr e p ara tion s

• See Sect ion 1 3.2.7.1

• IROFS RS-04 , Hot Cell Shielding Boundary

• IROFS RS-15 , Cask Docking Port Enabling Sensor

• See Sections 13.2.2.8 and 13.2.7.1

• Jud ge d u nlik e l y eve nt w ith int e rm e di a te c on se qu e n ce

• IROFS FS-01 , Enhanced Lift Procedure

• IROFS FS-02 , Overhead Cranes

• See Section 13.2.7.l

• Jud ged h ig hl y unlik e l y b ase d o n s uppli e r's c he c ks a nd b a l a n ces

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

• IROFS CS-03 , Interaction Control Spacing Provided by Admini s trative Control

• IROFS CS-04 , Interaction Control Spacing Provided by Passively Designed Fixtures and Work s tation Placement

• See Section 13.2.7.2

• IRO FS CS-0 2 , M ass a nd B at ch H a ndling L imit s fo r Ura nium M eta l , [Prop r i e t ary In fo rm a tion], T a r ge t s, a nd L a b ora t ory S a mp l e Out s id e Pro cess Syste m s

• IRO FS CS-0 3 , Int erac ti o n Co ntrol Sp ac in g Pro v id ed b y Admini s trat i ve C ontr o l

• IRO FS CS-04 , Int eract i o n Co ntrol S p aci n g Pro v id ed b y P ass i ve l y D es i g n e d Fi xtu r es a nd W o r k s ta tion Pl ace m e nt

• See Sect i on 1 3.2.7.2

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

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

• IROFS CS-08 , F l oor and Sump Geometry Co n tro l of S l ab Depth , Sump Diameter or Depth for Floor Spi ll Containment Berms

• IROFS CS-09 , Double-Wall Piping

• IROFS CS-26, Processing Component Safe Volume Confinement

• See Section 13.2.7.2 1 3-79

... ;. NWMI ..*... ..* ... ..... "0 NORTHWESTMBMCAl..tSOlOPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.05 S.C.06 S.C.07 S.C.08 S.C.09 S.C.10 S.C.11 Leak of fissile sol ution into the heating or cooling jacket on th e tank or vesse l System overflow to process ventilation involving fissile material Fissile so lution leaks across mechanical boundary b etween process vesse l s a nd heating/cooling jackets into heating/cooling media Backtlow of high-dose radiological and/or fissile solution into auxiliary system (purge air , chemical addition line , water addition line , etc.) High concentrations of uranium enter the concentrator or evaporator condensates High concentrations of uranium enter the low-dose or high-dose waste collection tanks High co ncentration s of uranium in co nta c tor so l ve nt regeneration aqueo u s waste

• Bound e d by S.R.13

• IROFS CS-11, Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm

• IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line

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

• See Section 13.2.7.2

• Bounded by S.R.13

• Bounded by S.R.13

• IROF S CS-06 , P enc il Tank, Vesse l , or Pipin g Safe Geometry Co nfin ement u si n g the Di amete r of Tanks , Vessels , or Piping

• IROFS CS-07 , P encil Tank an d Vessel Spaci n g Control Us in g Fixed Interaction Spacing oflndividual Tanks or Vesse l s

• IROFS CS-26, Proc essing Com p onent Safe Volume Confineme n t

• See Section 13.2.7.2

• IROFS CS-14 , Active Discharge Monitoring and Isolation

• IROFS CS-15 , Independent Active Discharge Monitoring and Isolation

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

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

• See Section 13.2.7.2

• Bound e d by S.C.04 and S.C. l 0 13-80

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

• NORTHWEST llEDtc.Al ISOTl>'U NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.12 S.C.13 S.C.14 S.C.15 S.C.17 High concentrations of uranium in the LEU target material wash solution High concentrations of uranium in the nitrou s oxide sc rubber High concentrations of uranium in the IX waste collection tanks effluent High co n ce ntration s of uranium in the IX resin waste High concentrations of uranium in the solid waste encapsulation process

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

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

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

• See Section 13.2.7.2

• IROF S CS-06 , P e n ci l Tank , Vessel, or Piping Safe Geometry Co nfin eme nt u s ing the Di a m e t e r of Tanks , Vessels , or Pipin g

• IROF S CS-1 6 , Sampling and Analysis of Uranium Mass or Co n centrat ion Prior to Dis c har ge or Disposal

• IROF S CS-1 7 , Ind epen d ent Samp lin g a nd Analysis of Ura nium Co n centrat ion Prior to Di sc h arge or Disposal

• See Section 1 3.2.7.2

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

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

• See Section 13.2.7.2

• IRO FS CS-06 , P e n ci l Tank , Vessel , o r Pipin g Safe Geometry Co nfin eme nt u s in g the Diameter of Tanks , Vessels , or Pipin g

• IROF S CS-07 , P e n cil Tank a nd Vessel S p acing Co nt rol Usi n g Fixed Int eractio n Spacing ofl ndi v idu al Tanks or Vessels

• IRO FS CS-16 , Sampling an d Analysis of Ura nium Mass or Co n centrat ion Prior to Discharge or Di s posal

• IRO FS CS-1 7, Indep endent Sa mplin g and Analysis of Uran ium Concentratio n Prior to Discharge or Disposal

• See Section 13.2.7.2

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

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

• IROFS CS-2 I, Visual Inspection of Accessible Surfaces for Foreign Debris

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

• IROFS CS-23 , Nondestructive Assay of Items with Inaccessible Surfaces

• IROFS CS-24, Independent Nondestructive Assay of Items with Inaccessible Surfaces

• IROFS CS-25, Target Housing Weighing Prior to Disposal

• See Section 13.2.7.2 13-81

.; ...

...... ..* *.. .... .. .. .. * ! * * !

• NOflTifWlST MEDICAL ISOTOf'lS NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.C.19 S.C.20 S.C.2 1 S.C.22 S.C.23 S.C.24 Failure of PEC -Component safe geometry dimension or safe volume Failure of concentration limits Target basket passive design contro l fa ilur e on fixed in teract ion spacing High concentration of uranium in the TCE evaporator residue High co n centration in the spent silicone oil waste High uranium content on HEP A filters and subsequent failure

• IROFS CS-06 , Pencil Tank , Vessel , or Piping Safe Geometry Co n finement using the Diameter of Tanks , Vesse l s, or Piping

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

• IROFS CS-26 , Processing Component Safe Vo lum e Confinement

• See Section 13.2.7.2

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

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

• IR OFS CS-03 , Interaction Contro l Spacing Provided by Admini s trative Control

• See Section 13.2.7.2

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

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

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

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

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

• See Section 13 .2. 7 .2

• IRO FS CS-04 , Int eraction Contro l Spacing Provided by Passively Designed Fixtures and Workstation Placement

• IROFS CS-05 , Container Batch Volume L imit

• IROFS CS-06 , P enci l Tank , Ve s sel , or Piping Safe Geometry Confinement Using the Diamet e r of Tanks , Vessels , or Piping

• IROFS CS-07 , Pencil Tank and Ves s el Spacing Control Using Fixed Interaction Spacing oflndividual Tanks or Vessels

• IROFS CS-16 , Sampling a nd Analy s is of Ura nium Mass or Concentration Prior to Discharge or Disposal

• IRO FS CS-17 , Independent Sampling a nd Analysi s of Uranium Concentrat i o n Prior to Discharge or Disposal

• See Section 1 3.2.7.2

• Bounded by S.C.17 13-82

.; ... ; .. NWMI ..*... ..* **.* ........ *.* ' *. * ! 0 NOITNWlST lllDecAl tSOTOPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Accident sequence designator Table 13-24. Analyzed Accidents Sequences (9 pages) from PHA Descriptor Preliminary IROFS Identified S.C.27 S.C.28 S.F.01 S.F.02 S.F.03 S.F.04 S.F.05 S.F.06 S.F.07 S.F.08 S.F.09 S.F.10 S.F.11 S.N.01 S.N.02 S.N.03 Failure of administrat i vely controlled container vo lum e limits Crane load drop accidents Pyrophoric fire in uranium metal Accumulation and ignition of flammable gas in tanks or systems Hydrogen detonation in reduction furnace Fire in reduction furnace Fire in a carbon retention bed Accumulation of flammable gas in ventilation sys tem components

• IROFS CS-0 3, Interaction Contro l Spac in g Provided by Administrative Control

• IROFS CS-04 , Int eraction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement

• IROFS CS-05 , Container Batch Volume Limit

• See Section 13.2.7.2

• IROFS FS-01, Enhanced Lift Procedure

• IROFS FS-02, Overhead Cranes

• See Section 13.2.7.2

• Event highly unlikely based on credible physical conditions

• IROFS FS-03, Process Vessel Emergency Purge System

• See Section 13.2.7.3

• Judged highly unlikely based on credible physical conditio n s

• Judged unlikely ba se d on event frequency

• IRO FS FS-05, Ex h a u st Stack Height

• See Section 13.2.7.3

• Bounded by S.F.02 Fire in nitrate extraction system -* Event unlikely with immediate or low consequences combustible so l vent with uranium General facility fire Hydrogen exp l osion in the facility due to a leak from the hydrogen storage or distribution system Combustible fire occurs in hot cell area Detonation or detlagration of natural gas l eak in steam generator room Tornado impact on facility and SSCs important to safe ty High straig ht-lin e winds impact the facility and SSCs import ant to safety Heavy rain impact on facility and SSCs important to safety

• Information will be provided in the Operating License Application

• In formation wi ll be provided in the Operating License App li cation

• Information will be provided in the Operating License Application

• Information wi ll be provided in the Operating License Application

• Judged highly unlikely event based on return frequency

• Judged hi gh l y unlikely to result in structure fai lur e

• Bounded by S.N.06 13-83

....... ; .. NWMI ::.**.*.*.* ..... .. .. .. . * " " "NOllTKWHTMEDICAltsOTOHS NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-24. Analyzed Accidents Sequences (9 pages) Accident sequence designator from PHA Descriptor Preliminary IROFS Identified S.N.04 Fl o odin g imp ac t to the fa cili ty a nd SSCs imp o rt a nt to safe t y S.N.05 Seismic impact to the facility and SSCs important to safety

• Jud ge d hi g hl y unlik e l y eve nt b ase d on fac ilit y loc a ti o n a b ov e th e 500-ye ar flood pl a in

• Judged highly unlikely to result in structure failure

• IROFS FS-04 , Irradiated Target Cask Lifting Fixture

• See Section 13.2.6.5 S.N.06 H ea vy s nowfall or i ce bu i ldup o n

• Jud g ed hi g hl y unlik e l y t o r es ult in s tru c tur e failur e fac ilit y and SS Cs import a nt to S.M.01 S.M.02 S.M.03 S.C S.01 H EPA IROF S I RU I X L EU Mo safe t y Vehicle strikes SSC important to safety and causes damage or leads to an accident sequence of intermediate or high consequence Fa cilit y evac u a t i on imp a ct s o n op e ration s Loca l ized flooding due to internal system leakage or fire suppression sprinkler activation N itric a cid fum e r e l ease hi g h-effic i e n cy p art i c ul a te a ir. ite m s r e li e d o n fo r safety. io din e r e m ova l uni t. io n exc h a n ge. l ow-e nri c h ed u ran ium. m o l y bd e num.

• Judged likely event with low consequence

• Jud ge d lik e l y eve nt wi th l ow con se qu e n ce

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

• See Section 13.2.7.2

• No IRO FS c urr e ntl y id e ntifi e d P EC p ass i v e e n g in eere d co nt ro l. PH A p re limin ary h azar d s a n a l ysis. SSC s tru c tur es , syste m s , a nd compo n e nt s. TCE tri c hl oroe th y l e n e U u ra nium. UN uran y l nitr ate. Table 13-25 provide s a s ummary of all IROFS identified by the acc ident anal yses performed for the Construction Permit Appli c ation. T a bl e 13-25 al s o identifie s wh e ther the IROFS w ere con s ider e d engineered s afety fe a ture s or a dmini s t ra ti v e control s. En g in e er e d sa fety featur es are described in Chapter 6.0 , and the admini s trative contr o l s are di s cu ss ed in Chapt e r 14.0 , "Te c hnic a l Sp e cific a tion s." A dditional IROFS are anti c ipated to be i dentified (or the current IROFS modifi e d) b y addit i onal de s ign d e tail de ve l o ped for the Operating Licen s e Applic a tion. Tab l e 13-25. Summary of Items Relie d o n for Safety I d entified b y Accident Analyses (3 pages) IROFS Engineered Administrative designator Descriptor safety feature control R S-01 Hot c e ll liquid c onfin e m e nt bound a ry ,/ RS-02 Reserved RS-0 3 H o t ce ll s econd a r y confin e m e nt bound a r y ,/ RS-04 Hot cell shielding boundary ./ RS-05 R ese r ve d 13-84

..... .. NWMI ..**.. ... *.. ..... .. .. .. ' ! *. * ! ." NOflTIIWUT MEDtCAl tSOTOPH NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages) IROFS Engineered Administrative designator Descriptor safety feature control RS-06 Reserved RS-0 7 R eserve d RS-08 Sample and analysis of low-dose waste tank dose rate prior to tran s fer outside the hot cell shielded boundary R S-0 9 P r im ary off gas re li ef sys t e m ./ RS-10 Active radiation monjtoring and i s olation oflow-dose waste tran s fer ./ R S-I I R eserve d RS-12 Cask containm e nt sampling prior to closure lid r e mo v al R S-1 3 Cask l oca l ve n ti l a ti o n du r ing c l os u re lid r e m ova l a nd d oc ki ng pr e p a r a t io n s RS-14 Re s erved R S-1 5 Cas k d oc kin g port e n a blin g se n so r CS-01 Re s erved CS-0 2 Mass a nd b atch h a ndl i n g limi ts for ur a nium meta l , [Pro pri etary I nfo rm at i o n], ta r gets, a nd l a b oratory sa mple o u ts id e pro cess syste m s CS-03 Interaction control spacing provided by administrative control ./ CS-04 In terac ti o n co nt ro l s p ac in g p rov id ed b y p ass i ve l y d es i gne d fi xtu r es ./ a nd w ork s t a ti o n pl ace m e nt C S-05 Container batch v olume limit CS-06 P e n c il t a nk , vesse l , or p i pin g safe geometry confine m e n t using t h e ./ d iame t e r of tanks , vesse l s, o r pi ping CS-07 P e ncil tank and vesse l s pacing control using fixed inter a ction ./ spacin g of indi v idual tanks or ves s el s CS-0 8 Floo r a nd s ump geo m etry co nt ro l of s l ab d e p t h , su m p di a m e t er o r ./ d e pth fo r fl oo r s p i ll co nt a inm e n t b e rm s CS-09 Double-wall pipin g ./ CS-IO Close d safe geometry h ea tin g or coo l i n g l oop with m o ni tori ng a nd ./ a l arm CS-I I S i mple overflo w to normally empty s afe geometry tank with level ./ alarm CS-1 2 Con d e n s in g p o t or seal p o t in ve ntil at i o n ve n t l i n e ./ CS-13 Simple overflow to normally empty safe geometry floor with level ./ alarm in the hot cell containment boundary CS-14 Ac ti ve di sc h a r ge m o ni to ring a nd i so l a tion ./ CS-15 Independent acti v e discharge monitoring and i s olation ./ CS-1 6 Sa mpl i n g a nd a n a l ys i s of u ra n i um m ass o r conce n trat i o n pri or to ./ di sc h a r ge o r di s p osa l CS-17 Independent sampling and analysis of uranium concentration prior ./ to di s charge or disposal 13-85

..... .. NWMI ...*.. ..* **.* ..... .. .. .. ' *. * !

• NORTHWEn llDHCAl tsOTDPES NWMl-2 01 3-021, Rev. 1 Chapter 13.0 -Accident Analysis Table 13-25. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages) IROFS Engineered Administrative designator Descriptor safety feature control CS-18 Backflow prevention device ./' CS-19 Safe-geometry day tanks ./' CS-20 Evaporator or concentrator condensate monitorin g ./' CS-21 Visual inspection of accessible surfaces for foreign debris ./' CS-22 Gram estimator survey of accessi ble surfaces fo r gamma activity ./' CS-23 Nondestructive assay of items with inaccessible surfaces ./' CS-2 4 Independent nondestructive assay of items with inaccessible surfaces ./' CS-25 Target housing weighing prior to disposal ./' CS-26 Processing component safe volume confinement

./' CS-27 Closed heating or cooling loop with monitoring and alarm ./' FS-01 Enhance d lift procedure

./' FS-02 Overhead cranes ./' FS-03 Proc ess vesse l emergency purge system ./' FS-04 Irradiated target cask lifting fixture ./' FS-05 Exhaust stack height ./' IROFS items r e li e d on for safety. The following subsections describe the IROFS that are not previously di sc ussed elsewhere in this chapter. The IROFS are grouped according to their respective accident sequence categories , as s hown in Table 13-26. Tab l e 13-26. Accident Sequence Category Definitions 13.2. 7 .1 Items Relie d on for Safety for Radio l ogical Acci d ent Sequences (S.R.) The following IROFS fall under the radiological accident sequence category and are not discussed elsewhere in this chapter. * -. S.R. S.C. S.F. S.N. S.M. s.cs. IROFS Definition Radiological Criticality Fire or explosion Natural phenomena Man-made Chemical safety Section containing related IROFS description 1 3.2.7.1 13.2.7.2 1 3.2.7.3 13.2.7.4 13.2.7.5 13.2.7.6 items r e li ed on for s afety. 13.2.7.1.1 IR O FS RS-08, Sample and A n alysis of Low Dose Waste Tank Dose Rate Prior to Transfer Outside the Hot Ce ll S h ie ld ed Boundary As an augmented administrative control (AAC), prior to transferring the solution from the low-dose waste tank to the low-dose waste encapsulation system outside of the hot cell shielded boundary, the low-dose waste tank will be administratively locked out, sampled, and the sample analyzed for high radiation.

Batches that satisfy the sample criteria can be transferred to the low-dose waste encapsulation system. The safety function of this AAC is to prevent transfer of low-dose solution to outside the shielded boundary at radiation dose rates that would lead to intermediate-or high-dose consequences to workers. 13-86

... .. NWMI ...... ..* ... ........ *. * *. * * . NOmlWEST ll(DCAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.1.2 IROFS RS-10, Active Radiation Monitoring and Isolation of Low Dose Waste Transfer As an AEC, the recirculating stream and discharge stream of the low-dose waste tank will be simultaneously monitored in a background shielded trunk outside of the hot cell shielded cavity. The continuous gamma-ray instrument monitoring the recirculation line and the transfer line will provide an open perrrussive signal to a dedicated isolation valve in the transfer line. The safety function of the system is to prevent transfer of low-dose waste solutions with exposure rates in excess of approved limits (safety limits and limiting safety system settings to be deterrruned later) to outside the shielded boundary at radiation dose rates that would lead to intermediate-or high-dose consequences to workers or the public. The system functions by monitoring both the recirculation line for the low-dose waste collection tank and the transfer line to the low-dose waste encapsulation system outside of the hot cell shielded boundary. Monitoring will be performed in a shielded trunk, which reduces the background from the normally shielded hot cell areas to acceptable levels for monitoring.

In this closed-loop system , the gamma monitor will provide an open perrrussive signal to a fail-closed isolation valve in the transfer line , allowing the isolation valve to open. If the radiation levels exceed a safety limit setpoint during recirculation for sampling or during transfers , the isolation valve will be closed. The isolation valve will also fail closed on loss of power and loss of instrument air. 13.2.7.1.3 IROFS RS-12, Cask Containment Sampling Prior to Closure Lid Removal As an AEC , a sampling system will be connected to the cask vent to sample the atmosphere within the cask prior to closure lid removal. The system will sample the contents of the cask and have the ability to remediate the atmosphere using a vacuum system if dose rates are too high (safety limits to be deterrruned). The safety function of IROFS RS-12 i s to prevent personnel exposure to high-dose gaseous radionucl ides. The system will identify a hazardous concentration of high-dose gases in the cask , and if a high dose is identified , will remediate the situation through evacuation to a safe processing system. The system works by evacuating a sample of the gas and analyzing the sample as it passes by a detector.

If high activity is detected, the system will alarm. The operator will use the system to evacuate and backfill the cask with fresh air (from a protected pressurized source such as a compressed bottle) until the atmospheres are within approved safety limits. 13.2.7.1.4 IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations As an AEC , a local capture ve ntilation system will be used over the closure lid to remove any escaped gases from the breathing zone of the worker during removal of the closure lid , removal of the shielding block bolts , and installation of the lifting lugs. The safety function of IROFS RS-I 3 is to prevent exposure to the worker by evacuating any high-dose gaseous radionuclides from the worker's breathing zone and preventing immersion of the worker in a high-dose environment.

The system will use a dedicated evacuation hood over the top of the cask during containment closure lid removal. The gases will be removed to the Zone 1 secondary containment system for processing. 13-87

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

• NORnfWEST MEOM:Al ISOTOPES 13.2.7.1.5 IROFS RS-15, Cask Docking Port Enabling Sensor NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis As an AEC, the cask docking port will be equipped with sensors that detect when a cask is mated with the cask docking port door. The sensors feed an enabling circuit that will prevent the door from being opened when no cask is present. The safety function of IROFS RS-15 is to prevent the cask docking port door from being opened, allowing a streaming radiation path to an accessible area and to prevent Zone II to Zone I air pressure imbalances that would allow air to migrate into the Zone II airlock. The system will also prevent a high streaming dose to workers from targets inside the hot cell , if the cask lift fails following mating. The system is designed to provide an enabling contact signal and positive closure signal when the sensor does not sense a cask mated to the door , causing the door to close. 13.2.7.1.6 IROFS FS-01, Enhanced Lift Procedure As an Administrative Control (AC), lifts of high-dose rate containers or casks or of heavy objects (weight limit to be determined in final design) that move over hot cells in the standby or operating modes will use an enhanced lift procedure to reduce the likelihood of an upset. Enhancements will use the guidelines in DOE-STD-I 090-2011, Hoi s ting and Rigging, for critical lifts (for nonroutine cover block lifts) and engineered production lifts (for routine container and cask lifts using pre-engineered fixtures).

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

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

are allowed as a safe batch , the operator will be provided with a carrying fixture that allows only [Proprietary Information]).

For targets , the housing is credited for maintaining the contents dry. Final limits for each activity will be set in final design. 13.2.7.2.2 IROFS CS-03, Interaction Control Spacing Provided by Administrative Control As a simple AC , while handling approved quantities of uranium metal , approved quantities of [Proprietary Information] (LEU target material), batches of targets, or batches of samples , an interaction control will be maintained between quantities being handled; fissile solution tanks , vessels, or IX columns; and safe-geometry ventilation housings.

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

13-88

..... ; .. NWMI ...*.. ..* *.. .*.* .. *.*.* ' ! *. * ! ' NORTKWUT MEDtcAl lSOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.2.3 IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement As a PEC, fixed interaction control fixtures or workstations will be provided for holding or processing approved containers with designated quantities of uranium metal, quantities of [Proprietary Information] (LEU target material), batches of targets, and batches of samples. The fixtures are designed to hold only the approved container or batch and are fixed with 61 centimeter (cm) (2-ft) edge-to-edge spacing from all other fissile material containers , workstations, or fissile solution tanks, vessels, or IX columns. Where LEU target material is handled in open containers , the design should prevent spills from readily spreading to an adjacent workstation or storage location.

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

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

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

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

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

13.2.7.2.5 IROFS CS-11, Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm As a PEC, for each vented tank containing fissile or potentially fissile process solution for which IROFS CS-11 is assigned , a simple overflow line will be installed below the level of the process vessel ventilation port and any chemical addition ports (where an anti-siphon safety feature will be installed).

The overflow drain will prevent the process solution from entering the respective non-geometrically favorable portions of the process ventilation system and any chemical addition ports (where the solutions will enter through anti-siphon devices).

The safety function of this feature is to prevent accidental nuclear criticality in geometrically favorable portions of the process ventilation system. The overflow will be directed to a safe-geometry storage tank , which will normally be empty. The overflow storage tank will be equipped with a level alarm to inform the operator when use of the IROFS has been initiated so that actions may be taken to restore operability of the safety feature by emptying the tank. The locations where this IROFS is used will be determined during final design. 13-89

.; ... ;. NWMI ...... ..* *.. .*.* .. *.*.* * " ". NOmlWUT MEOM:Al tsOTOf'fS NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.2.6 IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line As a PEC , downstream of each tank for which IROFS CS-12 is assigned , a safe geometry condensing pot or seal pot will be installed to capture and redirect liquids to a safe-geometry tank or flooring area with safe-geometry sumps. One such condensing or seal pot may service several related tanks within the geometry boundary of the ventilation system. The condensing or seal pot will prevent fissile solution from flowing into the respective non-geometrically favorable process ventilation system by directing the solution to a safe-geometry tank or flooring area with safe-geometry sumps. The safety function of IROFS CS-12 is to prevent accidental nuclear criticality in non-geometrically favorable portions of the proces s ventilation system. The safe-geometry tank or s umps will be equipped with a level alarm to inform the operator when use of the IROFS has been initiated. Each indi v idual tank or vessel operation must be evaluated for required capacity for overflow to ensure that a suitable overflow volume is available.

A monitoring and alarm circuit will be provided so that common overflow tanks or safe slab flooring or sumps may be used for multiple tanks or vessels , and limiting conditions of operation will be defined to ensure that the IROFS is made available in a timely manner or operations are suspended following an overflow event of a single tank. Where independent seal or condensing pots are credited , the drains of the s eal or condensing pot s must be directed to independent locations to prevent a common clog or overcapacity condition from defeating both. 13.2.7.2.7 IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary As a PEC for each vented tank containing fissile or potentially fissile process solution for which IROFS CS-13 is assigned , a simple overflow line will be installed above the high alarm setpoint.

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

13.2.7.2.8 IROFS CS-14, Active Discharge Monitoring and Isolation As an AEC for discharges from safe-geometry systems to non-favorable geometry systems , an active uranium detection system will be used to close an isolation valve in the dischar g e line at a uranium concentration limit and/or cumulative mass limit (the limit[s] to be set sufficiently low to preclude on process upsets and sufficiently high to maintain an operating limit setpoint below the safety setpoint).

This system will prevent a high-concentration uranium solution from being discharged to a non-favorable geometry system. The safety function of IROFS CS-14 is to prevent an accidental nuclear criticality.

The closed-loop system is designed to isolate the discharge points listed below by actively monitoring the solution stream for uranium concentration using a suitable uranium monitor. At a limiting setpoint , the uranium monitor will close an isolation valve in the discharge line to stop the discharge. The uranium monitor is designed to produce a valve-open permissive signal that fails to an open state, closing the valve on loss of electrical power. The isolation val v e is designed to fail-closed on loss of instrument air , and the solenoid is designed to fail-closed on loss of signal. The locations where this IROFS is used will be determined during final design. 13-90

...... ; .. NWMI ...... ..* **; ........ *.* * "NOllTHWESTMlDtCAllSOTOrf.S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.2.9 IROFS CS-15, Independent Active Discharge Monitoring and Isolation As an AEC for discharges from safe-geometry systems to non-favorable geometry systems , an independent active uranium detection system will be used to close an independent isolation valve in the discharge line at a uranium concentration limit and/or cumulative mass limit (the limit[s] to be set sufficiently low to preclude follow-on process upsets and sufficiently high to maintain an operating limit setpoint below the safety setpoint). This system will prevent a high concentration uranium solution from being discharged to a non-favorable geometry system. The safety function of IROFS CS-15 is to prevent an accidental nuclear criticality.

The closed-l oop system is designed to isolate the discharge points listed below by actively monitoring the solution stream for uranium concentration using a suitable monitor to detect uranium. At a limiting setpoint , the monitor will close an isolation val v e in the discharge line to stop the discharge. The monitor is designed using a different monitoring method and isolation valve than u s ed in IROFS CS-14 to produce a valve-open permissive signal that fails to an open state , closing the v alve on loss of electrical power. The isolation valve is designed to fail-closed on Joss of instrument air , and the solenoid is designed to fail-closed on loss of signal. The locations where this IROFS is used will be determined during final design. 13.2.7.2.10 IROFS CS-16, Sampling and Analysis of U Mass/Concentration Prior to Discharge/Disposal As an AAC , prior to initiating discharge from the safe-geometry container, tanks , or vessels assigned IROFS CS-16 to non-favorable geometry systems , the container , tank , or vessel will be isolated and placed under administrati v e control , recirculated or otherwise uniformly mixed, sampled , and the sample analyzed for uranium content. The discharge or disposal will only be approved following independent review of the sample results to confirm that the uranium content is below a concentration or a ma s s limit (to be determined for each individual application based on expected volumes and follow-on processing needs) and under the independent oversight of a supervisor (who administrati v ely controls the locks on the discharge system). Uranium mass in the disposal container or vessel will be tracked to en s ure that the mass or concentration limit for the container is not exceeded. The safety function ofIROFS CS-16 is to prevent accidental nuclear criticality caused by discharging or disposing of high-concentration uranium to an uncontrolled system. The IROFS functions as described by ensuring, through physical sampling and analysis , that the uranium content of an isolated container , tank , or vessel (both inlets and outlets isolated , as appli c able) is below a safe , single parameter limit on solution concentration or under a safe mass for the disposal container.

Systems, tanks , or ves s el s for which IROFS CS-16 applies , include: TCE recycle tanks Spent s ilicone oil Condensate tanks (either as normal or backup controls) 13.2.7.2.11 IROFS CS-17, Independent Sampling and Analysis of U Concentration Prior to Discharge/Disposal As an AAC , prior to initiating discharge from the safe-geometry tanks or vessels assigned IROFS CS-17 to non-favorable geometry systems , the tank or vessel will be isolated and placed under administrative control, recirculated , sampled , and the sample analyzed for uranium content. The recirculation or uniformly mixing , sampling , and analysis activities will be independent (performed at a different time, using different operators or laboratory technicians , and different analysis equipment , checked with independent standards) of that performed in IROFS CS-16. 13-91

.. ;*.-.; .. NWMI ...*.. ..* **: ..... .. .. . . . * ! *. * !

• NOmfWEST llEDtcAl tSOTOflES NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis The discharge or disposal will only be approved following independent review of the sample results to confirm the uranium content is below the limiting setpoint for uranium concentration or batch mass for the contents and under the independent oversight of a supervisor (who administratively controls the locks on the discharge system). Uranium mass in the disposal container or vessel will be tracked and independently verified to ensure that the mass or concentration limit for the container is not exceeded. The safety function of IROFS CS-17 is to prevent accidental nuclear criticality caused by discharging concentration uranium to an uncontrolled system. The IROFS functions as described by ensuring, through physical sampling and analysis , that the uranium content of an isolated tank or vessel is below a safe, single parameter limit on solution concentration or mass for a disposal container.

Systems, tanks , or vessels for which IROFS CS-17 applies include: TCE recycle tanks Spent silicone oil Condensate tanks (either as normal or backup controls) 13.2.7.2.12 IROFS CS-21, Visual Inspection of Accessible Surfaces for Foreign Debris As a simple AC, a visual inspection will be performed to identify foreign matter on accessible surfaces of equipment and waste materials approved for this method prior to disposal.

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

Materials involved must be solids (no solutions or liquids present). All surfaces must be visually accessible either directly or through approved inspection devices. The inspection criterion is for no foreign material of discernible thickness to be visible (transparent films allowed).

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

The control will be exercised at designated waste consolidation stations , holding specifically approved waste containers , and on the items approved by the Criticality Safety Manager. The waste will not be consolidated until independent measurements conducted according to IROFS CS-22 or IROFS CS-24 have been completed. The item will be controlled during the waste measurement analysis period. Items initially approved include disassembled irradiated or scrap target housing parts or pieces. 13.2.7.2.13 IROFS CS-22, Gram Estimator Survey of Accessible Surfaces for Gamma Activity As an AAC , a gram estimator survey will be performed on all accessible surfaces of equipment and waste materials approved for this method prior to disposal.

The survey will be performed on low-risk waste streams that have surfaces that are 100 percent accessible with the measurement instrument.

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

The purpose of this IROFS is to provide a backup instrument AAC to visual inspection (IROFS CS-21) for bulking and disposal oflow-r isk waste to prevent accidenta l nuclear criticality. All surfaces will need to be accessible to the instrument used. The waste stream must not be contaminated with significant fission product radionuclides since all activity is attributed to uranium. This survey wi ll be performed as backup to the visual inspection described in IROFS CS-21. An independent person from the one performing the visual inspection ofIROFS CS-21 will perform the survey. The control will be exercised at designated waste consolidation stations , holding s pecifically approved waste containers, on the waste items using survey instrument(s) and setpoint(s) approved by the Criticality Safety Manager. Waste consolidation will be conducted after independent verification of the two methods of quantifying uranium mass has been performed.

IROFS CS-22 is applicable to radiological waste generated outside the hot cell boundary that has had a low risk for direct contact with uranium-bearing materials.

13-92 NWMI ...... ... .... ........ *.* ' ! *. *

• NOmtW£n MEDICAl tsOTOP'£S NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.2.14 IROFS CS-23, Non-Destructive Assay of Items with Inaccessible Surfaces As an AAC, a nondestructive assay (NDA) method will be used on approved waste streams to quantify the uranium mass prior to disposal.

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

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

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

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

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

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

The control will be used as a backup to IROFS CS-16 , IROFS CS-21 or IROFS CS-23 , as approved by the Criticality Safety Manager for each waste stream. At designated waste consolidation stations holding specifically approved waste containers , the control will be exercised on the waste items using NDA techniques and mass or concentration limits approved by the Criticality Safety Manager. Waste consolidation will be conducted after independent verification of the two methods of quantifying uranium mass has been performed. 13.2.7.2.16 IROFS CS-25, Target Housing Weighing Prior to Disposal A s an AAC , on disposal of empty target housings , target housing pieces will be weighed and the weight compared to the original housing tare weight. The removed LEU target material will be weighed , and the weight compared to the original loading of LEU target material prior to disposal.

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

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

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

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

13-93

..

..*... ..* .... ........ *.* ' * ! * . NORTIIWm MfDtcAL ISOTOf'U NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.2.17 IROFS CS-26, Processing Component Safe Volume Confinement As a PEC , some processing components (e.g., pumps , filter housings , and IX columns) will be controlled to a safe volume for safe storage and processing of fissile solutions. The safety function of the safe volume component is also one of confinement of the contained solution.

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

Where insulation is used on the outside wall of the component, the insulation will be closed foam or encapsulated type (so as not to soak up solution during a leak) and will be compatible with the chemical nature of the contained solution. 13.2.7.3 Items Relied on for Safety for Fire or Explosion Accident Sequences (S.F.) The following IROFS fall under the fire or explosion accident sequence category and are not discussed elsewhere in this chapter. 13.2.7.3.1 IROFS FS-05, Exhaust Stack Height As a PEC , the exhaust stack is designed and fabricated with a fixed height for safe release of the gaseous effluents.

13.2.7.3.2 IROFS FS-02, Overhead Cranes Overhead cranes will be designed, operated, and tested according to ASME B30.2 , Ov e rhead and Gantry Cran e s (Top Running Bridge , Singl e or Multiple Gird e r , Top Running Trolley Hoist). Lifting devices for shipping containers will be designed , operated, and tested according to ANSI N14.6 , Standard for Special Lifting Devic es for Shipping Contain e r s W e ighing 10 , 000 Pound s (4 , 500 kg) or More for Nuclear Mat e rial s. The safety function of IROFS FS-02 is to prevent (by reducing the likelihood) mechanical failure of cranes during heavy lift activities. This IROFS will be implemented through the facilities configuration management and management measures programs.

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

The system will monitor the pressure or flow going to the header and open an isolation valve on low pressure or flow (setpoint to be determined) to restore the sweep gas flow to the system using nitrogen.

The system will be configured to provide more than 24 hr of sweep gas for the required tanks. The safety function oflROFS FS-03 is to prevent a hydrogen-air mixture in the tanks from reaching lower flammability limit conditions to prevent the deflagration or detonation hazard. The purge gases will be exhausted through the dissolver off gas or the process vessel ventilation system. The system is designed to sense low pressure or flow on the normal sweep system and introduce a continuous purge of nitrogen from a reliable emergency backup station of bottled nitrogen into each affected vessel. 13.2.7.4 Items Relied on for Safety for Natural Phenomena Accident Sequences (S.N.) The IROFS under the natural phenomena accident sequence category are discussed in Section 13.2.6. 13-94

..... ; .. NWMI ...*.. ..* *.. ..... .. .. .. ' *. * ! 0 NOITNWUT llEIMCAl tSOTOPU NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis 13.2.7.5 Items Relied on for Safety for Man-Made Accident Sequences (S.M.) There are no IROFS specifically identified for the man-made accident sequence category.

13.2.7.6 Items Relied on for Safety for Chemical Accident Sequences (S.CS.) There are no IROFS specifically identified for the chemical accident sequence category.

13-95

.:.;.-.;* .. NWMI ...... ..* **.* ......... *.* * *. * ! : , NORTHWEST MEDICAL ISOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 13.3 ANALYSIS OF ACCIDENTS WITH HAZARDOUS CHEMICALS This section analyzes the hazardous chemical-based accident sequences identified in the PHA. 13.3.1 Chemical Burns from Contaminated Solutions During Sample Analysis 13.3.1.1 Chemical Accident Description This accident sequence occurs during sampling and analysis activities performed outside the hot cell confinement and shielding boundary where facility personnel (operators and/or technicians) may handle radioactively contaminated acidic or caustic solutions. There are two possible modes of occurrence for this accident.

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

This accident sequence may result in long-term or irreversible tissue damage , particularly to the eyes. 13.3.1.3 Chemical Process Controls Facility personnel will be required to follow strict protocols for sampling and analysis activities at the RPF. Sampling locations , techniques , containers to be used , routes to take through the RPF when transporting a sample, analysis procedures , reagents , analytical equipment requirements, and sample material disposal protocol s will all be specified per procedures and/or work plans prepared and discussed prior to sampling or analytical activities.

Operators and technicians will be required to wear personal protective equipment , specifically for eye and skin protection. Radiologically contaminated acidic and caustic solution samples will be handled in approved containers.

Containers will be properly sealed when removed from sa mple locations and vent hoods during transport and/or storage. Sample containers will also be opened only when securely located in an approved laboratory hood , with the hood lowered for spray protection. This process will provide an additional layer of protection for eyes and skin (e.g., protective eyewear/face shield , laboratory coat or apron , anti-contamination chemical resistant gloves, etc.). 13.3.1.4 Chemical Process Surveillance Requirements Specific surveillance requirements will be identified in the Operating Permit Application.

For this accident sequence, surveillance may consist of management auditing or oversight of sampling and analysis activities to ensure adherence to the specified protocol of procedures , personal protective equipment usage , approved container usage , and laboratory hood etiquette. 13-96

... :.**.*.*.* ..... ; .. NWMI ............ ' ! *. * ! * . NORTHWEST MEDICAi. ISOTOPH 13.3.2 Nitric Acid Fume Release 13.3.2.1 Chemical Accident Description NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analys is This accident consists of a release of nitric acid fumes inside or outside of the RPF originating from one of the nitric acid storage tanks in the chemical storage and preparation room. 13.3.2.2 Chemical Accident Consequences Chapter 19.0 identifies hazardous chemical release scenarios for the facility using several of the stored chemicals. A I-hr release of the bounding RPF inventory of 5 , 000 L of nitric acid was shown to cause a concentration of 1,200 parts per million (ppm) at the controlled area fence line and 19.1 ppm at 434 m (1,425 ft) (nearest resident location) under dispersion conditions of moderate wind. Unmitigated exposure to a nearby worker would be much higher. The AEGL-2 , 60-minute (min) exposure limit for nitric acid is 24 ppm , which is high consequence to the public. AEGL-3, the 10-min exposure limit , is 170 ppm for a high consequence exposure to the worker. These determinations were made using the ALOHA (Areal Locations of Hazardous Atmospheres) computer code for estimating the consequences of chemical releases.

The use of ALOHA is recognized by the NRC in NUREG/CR-6410. The impact and consequences of a chemical release on RPF operations would require personnel to either evacuate the facility or , under some circumstances, shelter in place depending on the location of the event. 13.3.2.3 Chemical Process Controls The RPF will follow U.S. Environmental Protection Agency and Occupational Safety and Health Administration regulations for design , construction, and operation of chemical preparation and storage areas. Chemical handling procedures will be provided to operators to ensure safe handling of chemicals according to applicable regulatory requirements and consistent with the applicable material safety data sheets. IROFS to prevent or mitigate events that could impact the chemical storage tanks in the RPF chemical storage and preparation room are addressed in Section 13.2.5. 13.3.2.4 Chemical Process Surveillance Requirements Specific surveillance requirements for chemical use and storage at the RPF will be identified in the Operating Permit Application.

13-97

.. ;.-.;* .. NWMI ...... ..* **.* ........ *.* . ' ! *. * ! ." NORTHWEST Mmtc:Al ISOTOPES

13.4 REFERENCES

NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis 10 CPR 20, "Stan dards for Protection Against Radiation ," Cod e of F e deral R egulations, Office of the Federal Register , as ame nded. 10 CPR 30, " Rules of General Applicability to Domestic Licensing of Byproduct Material," Code of Federal R egu lation s, Office of the Federal Register , as amended. 10 CPR 50, " Domestic Licensing of Production and Utilization Facilities

," Code of Federal R egu lation s, Office of the Federal Register , as amended. 10 CPR 70, " Domestic Licensing of Special Nuclear Material," Code of Federal Regulations , Office of the Federal Register , as amended. 10 CPR 70.61, " Performance Requirements," Code of Federal R e gulations, Office of the Federal Register , as amended. 10 CPR 71 , " Packaging and Transportation of Radioactive Material ," Code of Federal R egu l ations, Office of the Federal Register , as amended. ACI 318 , Building Code Requirements for Structural Concrete, American Concrete Institute , Farmington Hills , Michigan , 2014. AISC 360 , Specification for Structural Steel Buildin gs, American Institute of Steel Construction , Chicago , Illinois , 2010. ANS 2.8, Determining Design Ba sis Flooding at Power Rea ctor Sites, American Nuclear Society, La Grange Park , Illinois , 1992 ,W2 002. ANSI Nl4.6 , Standard for Special Lifting D evices for Shipping Containers Weighing 10 , 000 Pounds (4 , 500 k g) or More for Nuclear Materials, American Nuclear Society, La Grange Park , Illinois , 1993. ANSI/ ANS-8.1 , Nuclear Criticality Safety in Op erations with Fissionable Material Out side Reactors, American Nuclear Society, La Grange Park , Illinois , 1998 (Reaffirmed 2007). ASCE 7 , Minimum D esign Loads for Buildings and Oth er Structur e s , American Society of Civil Engineers, Reston , Virginia , 2010. ASME B30.2, Overh ead and Gantry Cranes (Top Running Bridg e, Single or Multiple Girder , Top Running Troll ey Hoist), American Society of Mechanical Engineers , New York, New York, 2005. CDC , 2010 , N IOSH Pock e t Guide to Chemical Ha z ards , 2010-168c, Centers for Disease Control and Prevention , http://www.cdc.gov

/niosh/np g/, downloaded February 27, 2015. DC/COL ISG-07 , Int erim Staff Guidance on Assessment of Normal and Extreme Winter Pr ecipitatio n Loads on the Roofs of Seismic Category I Structures, U.S. Nuclear Regulatory Commission , Washington, D.C., 2008. DOE-HDBK-3010 , DOE Handbook-Airborne R elease Fractions/Rat es and R espira bl e Fractions for Non reactor Nuclear Facilities , Change Notice No. 1 , U.S. Department of Energy , Washington, D.C., December 1994 (R2013). DOE-STD-1090-2011 , Hoisting and Rigging , U.S. Department of Energy, Wa s hington , D.C., September 30, 2011. 13-98

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

• NORTifWtsT MEDtCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 13.0 -Accident Analysis EPA 520/1-88-020, Federal Guidance Report No. 11 , Limiting Values of Radionuclid e Intak e and Air Concentration and Dose Conversion Factors for Inhalati on, Submersion, and Ing est ion , U.S. Environmental Protection Agency, Washington , D.C., September 1988. FEMA, 2011, "F lood Insurance Rate Map, Panel 295 of 470, Boone County, Missouri and Incorporated Areas, Map# 29019C0295D, Federal Emergency Management Agency, Washington, D.C., March 17 , 2011. FEMA P-753 , NEHRP R ecom mend ed Seismic Provi sions for New Buildings and Other Structures, Federal Emergency Management Agency, Washington , D.C., 2009. Hydrometeorological Report No. 51 , Pr o babl e Maximum Pr ecipitat ion Estimates , United States East of th e 105th Meridian , U.S. Department of Commerce, National Oceanic and Atmospheric Administration , Washjngton, D.C., 1978. Hydrometeorological Report No. 53 (NUREG/CR-1486), Seasonal Variation of JO-Square Mile Probable Maximum Pr ecipitatio n Estimates, United States East of the 105 1 h M e ridian , U.S. Department of Commerce, National Oceanic and Atmospheric Administration, U.S. Nuclear Regulatory Commission, Office of Hydrology National Weather Service, Washington , D.C., April 1980. IBC , 2012 , Int ernat ional Building Code, as amended, International Code Council, Inc., Washington , D.C., February 2012. ICRP-26 , R ecom m en dation s of th e Int er national Commission on Radiological Protection , International Commission on Radiological Protection , Ottawa , Canada, 1977. ICRP-30, Limits for Intak es of Radionuclid es b y Workers, International Commission on Radiological Protection , Ottawa , Canada, 1979. ICRP-72 , Age-Dependent Do ses to the Members of the Public from Intake of Radionuclides

-Part 5 Compilation of Ing estio n and Inhalation Coefficients, International Commission on Radiological Protection , Ottawa , Canada, 1995. LA-13638, A R eview of Criticality Accidents, Los Alamos National Laboratory, Los Alamos, New Mexico, 2000. NAP 1994 , Estimating Bound s on Extreme Precipitation Events, N a tional Academy Press , National Research Council , Washington, D.C., 1994. NOAA Technical Report NWS 25, Comparison of Generali z e d Estimates of Probable Maximum Precipitation with Greatest Ob served Rainfall s , National Oceanic and Atmospheric Administration, Washington , D.C., 1980. NUREG-153 7 , Guidelines for Pr epa rin g and R ev i ew in g Applications for the Li censi n g of Non-Pow er R eacto rs -Format and Content, Part l , U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation , Washington , D.C., February 1996. NUREG-1940 , RASCAL 4: Description of Models and Methods, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards , Washington , D.C., December 2012. NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Wa s hington , D.C., March 1998. NWMI-2013-CALC-006 , O vera ll Summary Material Balan ce -MURR Targ e t Batch , Rev. D , Northwest Medical Isotopes , LLC , Corvallis, Oregon , 2015. NWMI-2013-CALC-Ol l , Source T e rm Calculations, Rev. A, Northwest Medical Isotopes , LLC , Corvallis, Oregon , 2015. 13-99

............ .. .. NWMI ..... .. .. .. * *

• NORTNWEST llEDICAL ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 13.0 -Accident Analysis NWMI-2014-051, Integrat ed Safety Analysis Plan for the Radioi sotope Production Facility, Re v. A, Northwest Medical Isotopes, Corvallis, Oregon , 2014. NWMI-2014-CALC-014, Selection of Dominant Targ et Isotopes for NWMI Material Balan ces, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2014. NWMI-2015-RPT-009, Fission Product Release Evaluation, Rev. B, Northwest Medical Isotopes , LLC , Corvallis, Oregon, 2015. NWMI-2015-SAFETY

-001 , NWM I Radi oisotope Production Facility Preliminary Ha zards Analysis, Rev. A, Northwest Medical Isotopes , Corvallis, Oregon, 2015. NWMI-20 l 5-SAFETY-004, Quantitative Ri sk Analysis of Process Upsets Associated with Passive Engineering Controls Leading to Criticality Accident Sequences , Rev. A , Northwest Medical Isotopes , Corvallis, Oregon, 2015. Regulatory Guide 1.145 , Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Rev. 1 , U.S. Nuclear Regulatory Commission, Washington, D.C., February 1983. WSRC-TR-93-262, Savannah Riv er Site Generic Data Base Development, Rev. 1, Westinghouse Savannah River Company, Savannah River Site, Aiken, South Carolina, May 1988.13-100