Text

Official Exhibit - NRC-006E-MA-CM01 - Northwest Medical Isotopes, LLC, Construction Permit Application - PSAR, NWMI-2013-021, Rev. 3, Chapters 5 Through 10 (Sep. 2017)
	ML18025B162
Person / Time
Site:	Northwest Medical Isotopes
Issue date:	09/30/2017
From:	; NRC/OGC
To:	; NRC/OCM
	SECY RAS
References
	50-609-CP, Construction Permit Mndtry Hrg, RAS 54182
	Download: ML18025B162 (293)
	v • d • e

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

Chapter 5.0 -Coolant Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 3 September 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, Oregon 97330 This page intentionally left blank.

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NORTirWUT MEOICAl ISOTOPU NWMl-2013-021 , Rev. 3 Chapter 5.0 -Coolant Systems Chapter 5.0 -Coolant Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 3 Date Published:

September 5 , 2017 Document Number. NWMl-2013-021 I Revision Number. 3 Title: Chapter 5.0 -Coolant Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

.; .. ;. NWMI ...... .. .. . .... .... .. NORTHWEST MEOtCAl tSOTOflfS This page intentionally left blank. NWMl-2013-021 , Rev. 3 Chapter 5.0 -Coolant Systems Rev Date 0 6/29/2015 1 5/19/2017 2 N/A 3 9/5/20 1 7 REVI S IO N HISTOR Y Reason for Revision Initial Applicat i on NWMl-2013-021 , Rev. 3 Chapter 5.0 -Coolant Systems Revised By Not required Incorporate changes based on responses to NRC C. Haass Requests for Additional Information Incorporate final comments from NRC Staff and ACRS; C Haass ful l document rev i sion

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NOlllTlfWEST MEOtCAl T h is p age intenti o na ll y left bl ank. NWMl-2013-021 , Rev. 3 Chapter 5.0 -Coolant Systems

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NORTHWEST MfDICIJ. ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 5.0 -Coolant Systems CONTENTS 5.0 COOLANT SYSTEMS .................................................................................................................. 5-1 5.1 S umm ary Description

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............................................................. 5-1 5 .1.1 Irradiated Target Basis ..............................

........................................................... 5-1 5.1.2 Vessels Considere d for Thermal Characterization

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5-3 5.1.3 Heat Load and Thermal Flux ............................................

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5-4 5.1.4 Maximum Vesse l Temperature and Pressure Est im ates ..........................

............ 5-7 5.1.5 Potential Impact of Overcooling Proces s So lution s ............................................. 5-8 5.1.6 Potential Impact on Gas Management System .............

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5-9 5.1. 7 Co n c lu sion ............................

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5-9 5.2 Coolant Systems De sc ription ..............

............................................................................... 5-9 5.3 References

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........................ 5-10 Figure 5-1. Figure 5-2. Table 5-1. Table 5-2. Table 5-3. Table 5-4. Table 5-5. FIGURES [Proprietary Information]

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................................................................................... 5-2 TABLES Vesse l s Selected to Describe Radioisotope Production Facility Thermal Characteristics (2 pages) ...............

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......................................................... 5-3 Heat Load and Thermal Flux for Se l ected Water-Cooled Vesse l s .................................. 5-5 Heat Load and Th e rmal Flux for Se l ected Vessels without Water Coo l ing ............

........ 5-5 Estimate of Maximum Temperature and Pr ess ure in Water-Cooled Vessels (2 pages) .........................................

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..................................... 5-7 Estimate of Maximum Temperature and Pressure in Vessel s wit hout Water Cooling .................

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... ;. NWMI *::**:*:* ...... ' *. * ! . NORTHWEST MEDtcAl ISOTOPlS TERMS Ac r o n y m s and Abbrev i ations z34 u m u 236 u m u 23s u 239 pu CFR EOI I IROFS Kr Mo MURR NWMI OSTR osu RPF u [Proprietary Information]

Xe U n its o c O f BTU cm cm 2 cm 3 ft 2 g hr m. in.2 kW L lb rem w wk wt% uranium-234 uranium-235 uranium-236

uranium-237 uranium-238 plutonium-239 Code of Federa l Regulations end of irradiation iodine item relied on for s afety krypton molybdenum Univer s ity of Mi s souri Research Reactor Northwest Medical Isotopes , LLC Oregon State University TRIGA Reactor Oregon State Univer s ity radioisotope production faci l ity uranium [Proprietary Information]

xenon degree s Celsius degrees Fahrenheit British thermal unit centimeter square centimeter cubic centimeter square feet gram hour inch square inch kilowatt liter pound roentgen equivalent in man watt week weight percent 5-i i NW Ml-201 3-02 1, Re v. 3 Cha pter 5.0 -C ool a n t Sy stems

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SUMMARY

DESCRIPTION 5.0 COOLANT SYSTEMS NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems Cooling water systems are used to control the temperature of process solutions in the Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) from process activities a nd the heat load resulting from radioactive decay of the fission product inventory. The RPF is located at a separate site , independent from the reactors used to irradiate the targets. Therefore , the RPF cooling s ystem does not influence operation of a reactor primary core cooling system. Chilled water is used a s the primary cooling fluid to proces s vess els. A central process chilled-water loop is used to cool three secondary loops: one large geometry s econdary loop in the hot cell , one safe geometry secondary loop in the hot cell , and one criticality-safe geometry secondary loop in the target fabrication area. The central process chilled-water loop relies on air-cooled chillers , while the secondary loops ar e cooled by the central chilled-water system through plat e-and-frame heat exchanger s. Selected process demands require cooling at le s s than the freezing point of water. These demands are met with water-cooled refrigerant chiller packages , cooled by the secondary chilled water loops. 5.1.1 Irradiated Target Basis Thermal characteristics of irradiat e d target s entering the RPF depend on the s ource reactor and decay time prior to receipt. Heat load estimates are currently based on preliminary calculations for targets irradiated at the Oregon State University (OSU) TRIGA 1 Reactor (OSTR) [Proprietary Information][Proprietary Information]

[Proprietary Information]

[Proprietary Information].

The calculations are based on the OSTR operating at a power of [Proprietary Information]

irradiating a target for [Proprietary Information]. The charged target is assumed to contain [Proprietary Information]

comprising

* * * * [Proprietary Information]

[Proprietary Information]

Estimates are limited to prediction of actinide s and fission products during irradiation of a fresh uranium target containing a limited set of assumed impurities. Calculations for rec y cled uranium , a broader set of impurities , and potential activation products are not currently available. The OSTR calculations resulted in an average power p e r target of [Proprietary Information].

The preliminary OSTR calculations have been extrapolated to e s timate the heat load of a target irradiated at the University of Missouri Resear c h Reactor (MURR). The basis for thi s extrapolation is discussed in Chapter 4.0 , " Radioisotope Production Facility Description

," Section 4.2 (biological shielding). Assuming a similar cycle time produces an average target power of [Proprietary Information].

The MURR target (in prototypical reactor location s), radionuclide inventory , and thermal characteristics modeling is underway and will be completed to support the Operating License Application. 1 TRJGA (Trainin g, Research , I s otopes , General Atomics) is a registered trademark of General Atomics , San Die g o , C a lifornia.

5-1

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NORTHWtn MEDICAi. ISOTOf'U Figure 5-1 de scribes the variation of heat generation with deca y time for an individual average target irradiated at MURR and OSTR over I week. Du e to location of the RPF r e lative to the reactor sites, the minimum decay time for receipt of targets [Proprietary Information].

The combination of reactor source and minimum decay time produces an estimated individual target heat load of NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems [Proprietary Inform a tion] Source: [P roprietary Inform a tion] Figure 5-1. [Proprietary Information)

[Proprietary Information]

So urce: [Propri e tary Inform a ti on] Figure 5-2. [Proprietary Information]

[Proprietary Information]

for MURR and OSTR irradiated targets, respecti ve l y. Several material-handling steps mu s t occur after the EOI within the reactor before a cask c o ntainin g irradiated targets can be transport e d to the RPF. Examples include tran sfe r of targets into the cask, remo va l of the loaded cask from the reactor pool , asse mbly of the cask lid , removal of water from the cask, drying the cask , performing the cask leak-check procedure , and cask decontamination a nd verification.

At-reactor handling procedures are projected to require significantly

[Proprietary Information]

for an individual cask. Independent of the actual cask handling time required , the clock time for EOI of a target batch become s a datapoint recorded on tran sfe r papers, a nd a cask will not be unloaded until the minimum decay time after EOI u se d in safety evaluations has elapsed. The number of irradiated target s received b y the RPF in a si ngle week also varies with the so urce reactor. The MURR operation i s based on irra diating eight targets per week , while the OSTR opera t ion i s based on irradiating 30 targets per week. Th e numb er of irradiated targets will be optimized as p a rt of the Operating License Application.

Figure 5-2 indicate s that the total heat load from target s receiv e d by th e RPF is ap proximately the sa me from either reactor as a function of dec ay time. The weekly heat load from radionuclide dec ay is estimated at [Proprietary Information].

Therefore , heat load from receipt of MURR targets ha s been used as an upper bound for irradiated target receipt s at the RPF. 5-2 NWM I ...... * *

NOfllnfWHT MEOtCAl tSOTOl'U NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems The thermal load is characterized by radial heat transfer in a vessel and the uranium concentration of so lutions held within vesse l s throu g hout the RFP. Increa s ing the number of targets proce sse d during a given week increases the total liquid vo lum e contained in geometrically favorable vessels (or liquid le ve l height), but does not change the uranium concentration or radial thermal flux. 5.1.2 Vessels Considered for Thermal Characterization T herma l characteristics ofRPF process vessels are eva luated in NWMI-2015-CALC-022, Maximum V e ssel H eat Load, T e mperature , and Pr essure Estimat e s. The vesse ls li sted in Table 5-1 were se lected to describe the RPF thermal characteristics.

The therma l characteristics of every vessel containing radionuclides in the RPF have not been developed by the preliminary evaluation.

Howe ver, the selected vesse l s were considered s ufficient to span the range of potential heat generation rates anticipated to be contained in proce ss vesse l s. Table 5-1. Vessels Selected to Describe Radioisotope Production Facility Thermal Characteristics (2 pages) Process location Description V essels Equipped with Water-Cooling Jackets Di sso l ver I /2 (DS-D-100/200) -Dissolver vesse l after in sertion of di sso l ver basket. This configurat ion i s Start of dissolution cycle included for co mplet e nes s , but is not yet a nal yze d. Require s consideration of di sso l ver b asket both b efore a nd after process so lution ad ded to the dissolver containing a dis so lver ba s ket. Dissolver 1/2 (DS-D-100/200)E nd of dissolution cycle Mo syste m feed tank IN IB (MR-TK-100/140) Dissolver so lution after dissolution complete, prior to combination with transfer flush water. Assumes Kr/Xe and I isotopes transfer to dissolver offgas equipment during dissolution. Di sso l ve r so lution after transfer to Mo sys t em feed vesse l , but prior to combination with transfer flush water. I mpure uranium collection tanks Process solution after recovery of Mo isotopes from the uranium-bearing (e.g., UR-TK-lOOA/B)

-lnput from process solution.

Mo recovery I mpure uranium collection tanks (e.g., UR-TK-lOOA/B)

-Output to uranium recovery Ion exchange feed tank I (UR-TK-200)

High-dose waste concentrate collection tank (WH-TK-240)

Uranium-bearing proc ess so lution input to uranium reco very after [Proprietary Information]. Process solution feed to the first-cycle uranium ion exchange columns after composition adjustment for ion exchange feed. Accumulated hi g h-do se liquid waste after co nc e ntration b y the waste handlin g system conce ntr ator. Vessels without Water-Cooling Jackets Uranium deca y tank (e.g., UR-TK-700A)

-lnput from separatio n Uranium decay tank (e.g., UR-TK-700A)

-Output to target fabrication Uranium-bearing pro cess sol ution after separatio n of uranium from other isotope s. Uranium bearing process solution after separation of uranium from other isotopes and [Proprietary Information).

5-3 NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems Table 5-1. Vessels Selected to Describe Radioisotope Production Facility Thermal Characteristics (2 pages) Process location Description Solid Transfer Containers (No Cooling Jackets) High-dose waste disposal container High-dose waste concentrate after addition of solidification agent. Irradiated target in cask at receipt [Proprietary Information]

in annular target cladding on receipt in the transfer cask. Flux based on both internal and external surfaces.

Temperature not yet evaluated.

Dissolver basket in air [Proprietary Information]

in dissolver basket for transfer between target disassembly and the target dissolver.

Annular configuration between basket wall and lifting post. Flux based on external surface only. I Kr Mo iodine. krypton. molybdenum.

[Proprietary Information]

Xe = xenon. Three groups of vessels are shown in Table 5-1. The first group contains vessels that include coolingjackets to control process solution temperatures. The solution temperature control facilitates solution transfer from one vessel to another, minimizes solution evaporation during storage, or maintains conditions for operation of subsequent unit operations.

The second group contains vessels that are not projected to require cooling. The third group contains vessels used for transfer or storage of solid material in air and are not influenced by the cooling water system. Uncooled vessels are included in the evaluation to provide a more complete description of the RPF vessel thermal characteristics.

Heat flux is estimated based on a simple steady-state heat balance for an individual vessel containing a heat-generating material.

Only radial heat flow is considered, neglecting heat flow in the axial direction.

The simplified heat balance neglects heat losses associated with evaporation of the liquid p hase that might be present in the vessel. This type of heat balance is equivalent to modeling each vessel as an unvented vessel , even though most vessels in the RPF will be either open containers or vented by the vessel vent system. The high-dose waste disposal container and irradiated target in the cask at receipt represent the only two process conditions listed in Table 5-1 that are actually closed containers in the RPF. The irradiated target in the cask at receipt and the dissolver basket in the air process locatio ns are included in Table 5-1 , even though the temperatures are not influenced by the coolant system, to indicate vessels will exist with relatively high surface temperatures within the RPF during operation.

Estimates of the irradiated target temperature and pressure on receipt at the RPF will be developed as part of the cask licensing activity.

Detailed design of the dissolver basket has not been completed. However , a preliminary calculation indicates that a dissolver basket with a lifting post diameter of [Proprietary Information].

The dissolver basket is not currently anticipated to be a completely enclosed vessel with the potential to build pressure on heating. The estimated dissolver basket temperature indicates that the containers of irradiated target material have the potential to achieve relatively high equilibrium temperatures.

5.1.3 Heat Load and Thermal Flux The volumetric heat load contained by process vessels varies throughout the RPF system as radioisotopes decay, selected radioisotopes are separated, and solution compositions are adjusted by the unit operations.

Conservatism is included in the thermal flux estimate by assuming heat transfer is limited to a radial direction and neglecting heat loss from solution evaporation.

Table 5-2 provides estimates of the volumetric heat load and radial thermal flux at the containment apparatus wall for selected vessels where cooling water is used to control the process solution temperature shown in Table 5-1. 5-4 NWMl-2013-021 , Rev. 3 Chapter 5.0 -Coolant Systems Table 5-2. Heat Load and Thermal Flux for Selected Water-Cooled Vessels Process Location Dissolver 1/2 (DS-D-100/200)-Start of dissolution cycle" Di ss ol ve r I /2 (DS-D-l 00/2 00) -E nd of di sso lution cycle Mo system feed tank lA/IB (MR-TK-100/140)

Impur e uranium collectio n tanks (e.g., UR-TK-1 OOA/B) -Input from Mo recovery Impure uranium collection tanks (e.g., UR-TK-IOOA/B)-

Output to uranium recovery I on excha n ge feed Tank I (UR-TK-200) High dose waste concentrate collection tank (WH-TK-240)

Thermal characteristics

- ********* -. -... -. * * [Proprietary

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[Proprietary (Proprietary Inform atio n] infonn a tion] infonnati on] lnfonn a tion] [Propri e t ary [Proprietary [Proprietary

[Propri e t ary I nformat ion] In formation] ln fo nn ation] In formation] (Proprietary (Propriet a ry (Proprietary (Proprietary information]

information]

Information]

Radial thermal flux W/cm 2 (BTU/hr-ft

2) [Proprietary information]

[Proprietary Informati on] (Proprietary lnfonnation]

[Propri e t ary Informati on] [Proprietary information]

[Proprietary Information] (Proprietary information]

So ur ce: NWMl-2015-CALC-022 , Max imum V es s e l H e a t Load , T e mp e ratu r e, and Pr es sur e Est im at es, R ev. A , orthwest M e dic a l Isotope s, LLC , Corva lli s, Oregon , 2015.

ot eva lu ated by thi s calculation.

The s implified eva lu at ion m e th odo l ogy was not cons id e r e d a ppli cab l e. b High-dose waste vesse l s co ll ec t was te from multipl e weeks of pro cess opera ti on th at i s domin a t ed by a [Proprietary In formatio n] tim e peri o d. C urr e nt plans a r e ba se d on co ll ec tin g hi g h-d ose waste as concentrate fro m [Propri e t ary Information]. Optimiza ti o n ma y allow ex t e n s i o n of the waste coll ec tion time period. Eva lu a tion indi cates the accum ul a t e d hi g h-d ose waste h eat lo a d approaches a n asy mpt ote of [Proprietary Inform a tion]. c B ase d on hi g h-do se waste concentrate t a nk th at i s [Proprietary In format ion] of h eat-ge n e r a tin g i so t opes. E OI e nd of irradiat i on. N I A n ot applicab l e. Mo mol y bdenum. TBD t o b e determined.

Table 5-3 provides s imilar est im ates for se l ected vesse l s where coo lin g wat er i s not provided to control the process so luti on temperature.

The vessels listed in Table 5-2 and Tabl e 5-3 were selected to indicat e the range of condition s experienced as proces s so lution i s tran sfe rred throu g h the RPF process eq uipm ent. Table 5-3. Heat Load and Thermal Flux for Selected Vessels without Water Cooling Thermal characteristics Process location ** ********* -.. -. .. -. *

Uranium decay tank (e.g., UR-TK-700A)

-[Proprietary

[Proprietary (Proprietary (Propriet a ry Input from separation Inform a tion] in format i on] lnfonnation]

Information]

Uranium deca y tank (e.g., UR-TK-700A)-[Prop r ietary [Prop ri etary (Propriet a ry (Proprietary Output to tar ge t fabrication Information]

In formation]

Radial thermal flux W/cm 2 (BTU/hr-ft

2) [Proprietary Infonn ation] (Propri e t ary Inform a ti on] High-dose waste disposal container [Proprietary [Proprietary

[Proprietary (Proprietary

[Proprietary Information]

information]

lnfonn atio n] lnfonnation]

Information]

So ur ce: NWMI-20 15-CALC-022 , Max imum Vess el H eat L oa d , T e mp e ratur e, and Pr e ssure Estimat e s, R ev. A, Northwest Medica l I so top es, LLC, Corva lli s , Oregon , 2015.

High-do se waste vesse l s co ll ec t waste from multipl e weeks of proc ess operat i on th a t i s domin a ted by a [Prop ri etary Inform a tion] tim e p eriod. C urr e nt pl ans a re b ased on collecting hi g h d ose waste as con ce ntrat e from [Propri etary Informati on]. Optimi zat ion may a ll ow ex ten s i o n of the waste co ll ection tim e p e ri od. Eva lu a tion indic ates th e acc umul a t e d hi g h-do se waste h eat l oa d approaches an as ympt o t e of [Proprietary Inform a ti o n]. b Ba se d on hi g h-d ose waste di sposa l contai n er that is [Proprietary Inform a tion] of heat-generating isotopes. EO I e nd of irradiation.

N/ A n o t ap plic a ble. 5-5

NWM I ...... NOflTHWESTMlDtcALISOTOPES NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems The heat load of process solutions in unit operations prior to the start of separating uranium from other radionuclides can be characterized by the solution uranium concentration.

Planned operat i ng conditions are used to support characterization of the thermal heat load. The process solution uranium concentration at the end of the dissolution cycle [Proprietary Information]

is estimated based on the mass of uranium input by a dissolver basket, combined with the volume of acid charged to the d i ssolver. The resultant dissolver solution is transferred to the molybdenum (Mo) system feed tank , and thermal characteristics are evaluated neglecting mixing with dissolver vessel flush solutions.

The uranium concentration in the impure uranium collection tanks [Proprietary Information], ion exchange feed tank [Proprietary Information], and uranium decay tanks [Proprietary Information]

represent goal compositions for process solutions during operation. Three radionuclide decay times , summarized below , are used to describe the RPF thermal characteristics based on currently planned decay limits within the process operation:

*

A decay time of [Proprietary Information] (minimum decay time for targets in transfer casks received at the RPF outer door) is used to describe process solutions in the dissolver, Mo system feed tanks , and solution transferred into the impure uranium collection t anks, neglecting the time required for cask receipt, target disassembly, and target dissolution. A decay time of [Proprietary Information] (minimum decay time required to control in-growth of plutonium-239

[2 3 9 Pu] in recycled uranium after separations) is used to describe process solution at the end of the impure uranium collection tank storage period, solution in ion exchange feed tank 1 , solution transferred into the uranium decay tanks , and waste entering the high-dose waste concentrate collection tank, neglecting time required to complete separation activities. A decay time of [Proprietary Information] (minimum decay time to reduce uranium-23 7 [2 3 7 U] in recycled uranium, allowing contact operation and maintenance in the ta r get fabrication system , is used to describe the process solution at the end of the storage period in the uranium decay tanks. Target heat generation (shown by Figure 5-1) is placed on a unit uranium mass basis to support the estimate of heat load in the selected vessels. The unit uranium mass input is modified to approximate the impact of radionuclide separations that occur in unit operations. The unit mass heat genera t ion shown in Table 5-2 for a dissolver vessel at the start of dissolution cycle [Proprietary Information]

re p resents material containing all radionuclides in an irradiated target at [Proprietary Information].

All isotopes of krypton (Kr), xenon (Xe), and iodine (I) are assumed to be evolved to the dissolver offgas system during dissolution, reducing the unit mass heat generation to [Proprietary Information].

The molybdenum isotopes are assumed to be separated from the dissolver solution by the Mo recovery and purification system, reducing the unit mass heat generation to [Proprietary Information]

for solution entering the impure uranium collection tanks. The unit mass heat generation is reduced to [Proprietary Information]

after solution in the impure uranium collection tanks is decayed to [Proprietary Information].

The thermal characteristics of recycled uranium process so lut ion after separation in the uranium recovery and recycle system are shown in Table 5-3. Minima l separation of neptunium from uranium is projected to be obtained by the process, and the heat load is approximated by a unit mass heat generation dominated by the isotopes of neptunium and uranium [Proprietary Information]

entering the uranium decay tanks. The unit mass heat generation of recycled uranium solution transferred to target fabrication is reduced to [Proprietary Information].

The thermal characteristics of waste handling vessels are not characterized by the process solution uranium concentration and are expected to collect solution containing radionuclides from multiple weeks of operation.

The waste handling vessel thermal characteristics are described by the high-dose waste vessels that contain a majority of the waste radionuclides. Weekly input to the high-dose waste vessels is dominated by wastes from the uranium recovery and recycle separation system and described by radionuclides in a target decayed to [Proprietary Information]

with isotopes of Kr, Xe , I , and Mo removed. 5-6 NWMI ...**... * * *

NOfmfW'EST MfotCAl ISOTOfl(S NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems A c cumulation of waste from a week of operations increases the waste vessel heat load that decreases by decay while awaiting waste input from a subsequent week of operation.

Current plans are based on accumulating waste from [Proprietary Information].

However, system optimization may increase the goal high-dose waste accumu lation time period. Evaluation of the heat load sequence indicates that the waste heat load approaches an asymptote of [Proprietary Information]

after accumulating waste for 16 weeks. Therefore, the waste vessel heat loads were characterized by a total heat load of [Proprietary Information]

contained in the vessel capacity using current estimates of the vessel dimensions.

5.1.4 Maximum Vesse l Temperature and Pressure Estimates An estimate of vessel temperature has been obtained using an overall heat transfer coefficient obtained from handbook values for a tank on legs containing water and an assumed cell air temperature of 35°C (95°F). Temperatures are estimated assuming no water-cooling system is active , and pressures are estimated assuming each vessel is unvented to approximate maximum values. Note that the preliminary estimate assumes that radial temperature variations within the generating heat material are not significant , which may be appropriate for vessels containing liquids , but could be questionable for containers generating solids. The vapor pressure of water at the estimated vessel temperature is used to approximate the maximum pressure.

The vapor pressure of water was considered a conservative estimate of the pressure developed within a process apparatus, as the tota l vapor pressure of a solution is decreased by the addition of nitric acid or uranyl nitrate to the liquid phase. Table 5-4 provides estimates of the maximum temperature and pressure predicted for selected vesse l s where cooling water is used to control the process so lution temperature shown in Table 5-1. Table 5-5 provides similar estimates for uncooled vessels and the high-dose waste disposal container.

The maximum temperature and pressure that could be observed in representative vessels without o peration of the coolant system is shown in Table 5-4 as [Proprietary Information], absolute for the Mo system feed tanks. However , the evaluation approach was not considered applicable to the vessel configuration representing a dissolver at the start of the dissolver cycle. This configuration has the potential to produce higher temperatures and pressures than the vessels that could be evaluated using the current approach.

Table 5-4. Estimate of Maximum Temperature and Pressure in Water-Cooled Vessels (2 pages) Process location Dissolver l/2 (DS-D-100/200) -Start of dissolution cycle d Dissolver 1/2 (DS-D-100/200)-End of dissolution cycle Mo system feed tank I A/ 1 B (MR-TK-100/140)

Impure uranium collection tanks (e.g., UR-TK-lOOA/B)-

Input from Mo recovery Impure uranium collect i on tanks (e.g., UR-TK-IOOA/B)

-Output to U recovery Ion exchange feed tank 1 (UR-TK-200)

Rad i al thermal flux ,* BTU/h r-ft 2 [Proprietary Information]

[Proprietary Information]

5-7 Maximum heat transfer surface temperature b oc (o F) [Proprietary Information]

[Proprietary Information]

Estimated maximum unvented vessel pressure c lb/in 2 , absolute [Proprietary Information]

[Proprietary Information]

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NOflllfWUT M£DtCAl ISOTOr£5 NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems Table 5-4. Estimate of Maximum Temperature and Pressure in Water-Cooled Vessels (2 pages) Process location High-dose waste concentrate collection tank (WH-TK-24 0) Radial thermal flux,* BTU/hr-ft 2 [Proprietary Information]

Maximum heat ' transfer surface temperatureb oc (oF) [Proprietary Information]

Estimated maximum unvented vessel pressurec lb/in 2 , absolute [Propriet ary Information]

Sou r ce: NWMJ-2015-CALC-022, Maximum Vessel H ea t Load , Temperature, and Pr ess ur e Estimat es, Rev. A, Northwe s t Medical Isotope s, LLC , Corva lli s, Oregon , 2015

Radial th ermal flux from Table 5-2. b Maximum h eat transfer s urfac e tempera tur e assuming overall h ea t transfer coefficient at walls of 1.8 BTU/hr-ft 2-°F and ambient cell a ir temperature of 35°C (95°F). c U n vented vessel pressure ba sed on water vapor pressure at the m aximum heat transfer s urface t emperature.

Actual estimated water vapor pre ss ure s hown in parentheses for pressures le ss than 14. 7 lb/in.2 , abso lut e. d Not eva luat ed by this calculation.

The s implified methodology was not conside r ed applicable.

Mo TBD mol ybdenu m. U = uranium. = to be determined.

Table 5-5. Estimate of Maximum Temperature and Pressure in Vessels without Water Cooling Process location Uranium decay tank (e.g., UR-TK-700A)Input from se paration Uranium decay tank (e.g., UR-TK-700A)Output to target fabrication Hi g h-dose waste di s po sa l container Radial thermal flux* BTU/hr-ft 2 [Propriet ary Information]

[Proprietary Information]

Maximum heat transfer surface temperatureb oc (oF) [Proprietary Information]

[Proprietary Information]

Estimated maximum unvented vessel pressurec lb/in 2 , absolute [Propri e ta ry Information]

[Proprietary Information]

[Propri etary Information]

Source: NWMI-2015-CALC-022, Maximum Vesse l H ea t Load , T e mp er atur e , and Pr ess ur e Estimat es, Rev. A, Nort hw est Medical Isotopes , LLC , Corva lli s , Oregon , 2015.

Radial th e rmal flux from Table 5-3. b Maximum h eat transfer s ur face temperature assuming overa ll h eat tran s fer coefficient at wa ll s of 1.8 BTU/hr-ft 2-°F and ambient ce ll air t e mp eratu r e of 35°C (95°F). c U nvented vessel pre ss ur e based on wate r va por pressure at the m aximu m he at transfer s ur face temperature. Actual estimated water vapor pres s ure s hown in parentheses for pres s ure s l ess th an 14.7 lb/i n.2 , absolute. 5.1.5 Potential Impact of Overcooling Process Solutions Overcooling of uranium-b ea ring process solutions has the potential to [Proprietary Information].

Precipitation as a solid form effectively increase s the uranium concentration of material con t ained by a proces s vessel and potentially results in a nuclear criticality. The [Proprietary Information].

Criticality evaluations are de sc ribed in the following three documents for current equipmen t configurations of the irradiated target disa sse mbly/dissolution , target fabrication, and uranium recycle separation syste ms , re spec ti ve ly. * * * [Proprietary Information]

[Proprietary Information]

5-8 NWM I ...*.. * *

NotmfW(ST MEDICAl ISOTDPH NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems T he impact of uranium precipitation upset conditions on nuclear criticality calculations was evaluated by interspersing selected tanks containing

[Proprietary Information]

among vessels containing uranium at a conservative nominal process concentration.

The results indicate that precipitation upset conditions are predicted to remain below an upper subcritical limit of [Proprietary Information]

for the configurations evaluated.

Therefore , overcooling process solutions is not predicted to pose a nuclear criticality hazard for the current RPF equipment configuration. 5.1.6 Potential Impact on Gas Management System Coolant system operation has the potential to impact the performance of the gas management system cooled sections.

The primary gas management system cooled section controls the decay time provided for noble gases (isotopes of Kr and Xe) by holdup in the dissolver offgas system. The maximum hypothetical accident evaluated in Chapter 13.0, "Accident Analysis," Section 13.2.1 indicates the dose consequences from a bounding release of Kr and Xe isotopes alone is less than 0.15 roentgen equivalent i n man (rem). The bounding release of noble gases is less than the performance requirement of 5 rem for an intermediate consequence event defined in Title I 0 , Code of F e deral R e gulations, Part 70.61 , "Performance Requirements" ( 10 CFR 70.61 ). Therefore, the cooling water system is not considered to be an item relied on for safety (IROFS) based on the potential impact on the gas management systems. 5.1. 7 Conclusion The evaluation focused on vessels equipped with water cooling jackets. Typical process vessels of a pencil tank configuration are anticipated to be constructed from material similar to Schedule 40 stainless steel pipe. The pressure rating of seamless standard stainless steel pipe ranges from: *

4-in., Schedule 40 -1 , 500 to 900 lb/in 2 , gauge for 37.8 to 398.9°C (100 to 750°F), respectively 5-in., Schedule to 800 lb/i n 2 , gauge for 37.8 to 398.9°C (100 to 750°F), respectively 6-in., Schedule 40-to 725 lb/in 2 , gauge for 37.8 to 398.9°C (100 to 750°F), respectively The maximum temperature and pressure in vessels without cooling and ventilation is estimated at [Proprietary Information]

in Table 5-4 in the Mo system feed tanks, which are projected to be 5-in. diameter pencil tanks. The high-dose concentrate collection tank is a standard tank design such that the stainless steel pipe comparison is not applicable.

Maximum temperature and pressure for this vessel is estimated at [Proprietary Information].

Standard tank designs are capable of containing process solution at the dose concentrate collection tank conditions.

Based on the above comparisons , the maximum temperature and pressure within RPF vessels are anticipated to not result in failure of a process apparatus , and the cooling water system is not selected as an IROFS. The approach used to evaluate vessels was not considered applicable to a dissolver at the start of a dissolver cycle (non-uniform distribution of the heat-generating material).

Future evaluation of this vessel configuration has the potential to impact the im p ortance of the coolant system. 5.2 COOLANT SYSTEMS DESCRIPTION The above analysis and description show that the cooling water system is designed such that the system will function in a manner, whether operational or not , consistent with occupational safety and protection of the public and environment.

Therefore, the cooling function is not considered an IROFS. A description of the coolant systems for the RPF is provided in Chapter 9.0, "Auxiliary Systems," Section 9.7. 5-9

5.3 REFERENCES

NWMl-2013-021, Rev. 3 Chapter 5.0 -Coolant Systems 10 CFR 70.61, " Performance Requirements

," Code of F e deral R e gulations , Offic e of the Federal Register , as a mended. [Proprietary Information]

NWMI-2015-CALC-022 , Maximum Vessel H e at Load, Temperature, and Pressure Estimates, Rev. A , Northwe st Medical I sotopes, LLC , Corvallis, Oregon , 2015. [Proprietary Inform ation] [Proprietary Inform atio n] [Proprietary Information]

OSU-RC-1301 , Pr e limina ry Isotop e Inv e nto ry Estimat e of the 99 Mo Targets After Irradiat io n In the Or ego n Stat e TRIGA R e a c tor, Rev. 1 , Oregon State University , Corvallis , Ore go n , March 2013. 5-10

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Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021, Rev. 3 September 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, OR 97330 This page intentionally left blank.

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NORTHWEST MEOICAl LSOTOPES NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 3 Date Published: September 5 , 2017 Document Number. NWMl-2013-021 I Revision Number. 3 Title: Chapter 6.0 -Engineered Safety Features Construct i on Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

c u.J ir-e_

NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features This page intentionally left blank.

Rev Date 0 6/29/2015 1 6/26/2017 2 8/5/2017 3 9/5/2017 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Eng i neered Safety Featu r e s RE V I S IO N HISTORY Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to NRC C. Haass Requests for Additional Information Modification based on ACRS comments C. Haass Incorporate final comments from NRC Staff and ACRS; C. Haass full document revision "NWMI *::**::* ...... ' * * ' NORTHWEST MlDICAl tsOTOflU NWMl-2013-021 , Rev. 3 Chapter 6.0 -E n gineered Safety Features T hi s pa ge i ntenti on a ll y l eft blank.

CONTENTS NWMl-2013-021 , Rev. 3 Cha p te r 6.0 -Eng i neered Safety F eatures 6.0 ENGINEERED SAFETY FEATURES ................

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6-1 6.1 Summary De sc ription .............................................................................

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... 6-1 6.2 Det ailed De sc riptions .....................................

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6-5 6.2.1 Confinement

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6-5 6.2.1. l Confi nement System .....................

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6-7 6.2.1.2 Acci dent s Mitigated

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6-11 6.2.1.3 Fu nctional R e quir eme nt s .............................................

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6-11 6.2.1.4 Co nfinement Components

................................................................. 6-11 6.2.1.5 Test Requir eme nts ............

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.......... 6-12 6.2.1.6 De s ign Ba s i s ...........................................

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6-13 6.2.1. 7 Derived Confinement Item s R e li e d on for Safety .............................

6-13 6.2.1.8 Dis so lv e r Off gas Systems ..............

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........ 6-23 6.2.1.9 Ex h a ust System .......................

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6-26 6.2.1.10 Effl uent Monitoring System .........................

..................................... 6-26 6.2.1.11 Radioactive R e l ease Monitoring

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6-26 6.2.1.12 Co nfinement Syste m Miti gat ion Effects ........................

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6-26 6.2.2 Containment

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6-27 6.2.3 Emergency Cooling System ......................................

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.............. 6-27 6.3 Nuclear C riticality Safety in the Radioi so top e Product i on Facility ...................

.............. 6-28 6.3.1 Criticality Safety Controls .............

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............................ 6-36 6.3.1.1 Prelimin ary Critica l ity Safety Evaluations

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6-36 6.3.1.2 Derived Nuclear Criticality Safety Item s Relied on for Safety .........

6-59 6.3.2 Surveillance R e quirement s ...................................................................

.............. 6-71 6.3.3 Technical S pecific a tion s .............................................

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6-71 6.4 Ref erences ....................

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6-72 6-i

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NORTHWEST MEDtCAl ISOTO!tf.S NWM l-2013-0 21 , Rev. 3 Chapt e r 6.0 -E n gi ne er e d Saf e ty Featu re s Figure 6-1. Figure 6-2. Figure 6-3. Figure 6-4. Figure 6-5. Figure 6-6. Figure 6-7. Table 6-1. Table 6-2. Table 6-3. Table 6-4. Table 6-5. Table 6-6. Table 6-7. Table 6-8. Table 6-9. Table 6-10. Table 6-11. T a ble 6-12. T a ble 6-13. FIGURES Simplified Zone I Ventilation Schematic

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....... 6-6 Ground Level Confinement Boundary ...........

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6-8 Mechanical Level Confinement Boundary ...........................

........................................... 6-9 Lower Level Confinement Boundary ...............................

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.................. 6-10 Di sso lver Offgas System Engineered Safety Features .......................

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................ 6-14 Dissolver Offgas Hot Cell Equipment Location ...........

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6-15 Proposed Location of Double-Wall Piping (Example)

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............. 6-21 TABLES Summary of Confinement Engineered Safety Features (2 pages) ...................................

6-2 Summary of Criticality Engineered Safety Features (2 pag es) ........................................

6-3 Confinement System Safety Functions

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................................ 6-7 Area of Applicability Summary ................................

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6-37 Controlled Nuclear Criticality Safety Parameters

.......................................................... 6-38 [Proprietary Information]

Double-Contingency Controls .............................................. 6-39 [Proprietary Information]

Double-Contingency Controls (2 pages) .............

................. 6-40 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-41 [Proprietary Information]

Double-Contingency Controls (8 pages) ...........................

... 6-43 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-51 [Proprietary Information]

Double-Contingency Controls (3 pages) ..............................

6-53 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-56 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-57 6-ii TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 235 U uranium-235 ADUN acid-deficient uranium nitrate AEC active engineered control NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features ANECF average neutron energy causing fission ANS American Nuclear Society ANSI American National Standards Institute CAAS criticality accident alarm system CFR Code of Federal Regulations CSE criticality safety evaluation DBE design basis earthquake HEGA high-efficiency gas adsorber HEPA high-efficiency particulate air HVAC heating , ventilation , and air conditioning IEU intermediate-enriched uranium IX ion exchange IROFS item relied on for safety Kr krypton L EU low-enric hed uranium MCNP Monte-Carlo N-Particle Mo N0 2 NO x NRC NWMI PEC PHA RPF SSC SPL UN [Proprietary Information]

USL Xe molybdenum nitrogen dioxide nitrogen oxide U.S. Nuclear Regul atory Commission Northwest Medic al I soto p es, LLC pa ssive engineered control preliminary hazard s analysis radioisotope production facility structures, systems, and components single parameter limit uranium nitride [Proprietary Inform atio n] upper s ubcritical limit s xenon 6-iii NWM I ...... ' * * ! NOfl1lfWHT MEDfCAl ISOTOPES Units o c O f atm cm cm 3 ft ft 2 ft 3 g hr Ill. L m m 2 mm mL mol rad wt% yr degrees Ce l sius degrees Fahrenheit atmosphere centimeter cubic centimeter feet square feet cubic feet gram hour inch liter m eter square m eter minute milliliter mole radiation absorbed dose weight percent year 6-iv NWMl-2013-021 , Rev. 3 Chapter 6.0 -E n gineered Safety Features NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features 6.0 ENGINEERED SAFETY FEATURES 6.1

SUMMARY

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

Confinement is a general engineered safety feature that i s credited as being in place as part of the preliminary hazards analysis (PHA) described in Chapter I 3.0, "Accident Analysis." Additional items relied on for safety (IROFS) associated with the confinement system were derived from the accident analyses in Chapter 13.0. The derived IROFS are also listed in Table 6-1 , with reference to more detailed descriptions in Section 6.2.1. The current design approach does not anticipate requiring containment or an emergency cooling system a s engineered safety features , as discussed in Section s 6.2.2 and 6.2.3. Nuclear criticality safety is discussed in Section 6.3. Criticality safety controls are described in Section 6.3.1. The currently defined criticality safety controls are derived from a combination of pr e liminary criticality safety evaluations (CSE) and accident analyses , which are described in Chapter 13 .0. The criticality safety analyses produce a set of features needed to satisfy the contingency requirements for nuclear criticality control. These features are evaluated by major systems within the RPF and listed by major system in Section 6.3. I. I , Table 6-6 through Table 6-13. The accident analyses in Chapter 13.0 identify IROFS for the prevention of nuclear criticality , which are summarized in Table 6-2, with reference to more detailed descriptions in Section 6.3.1.2. 6-1 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-1. Summary of Confinement Engineered Safety Features (2 pages) Detailed Engineered safety SSCs providing engineered description feature IROFS Accident(s) mitigated safety features section Co nfin e m e nt . Equipm e nt . Co nfin e m e nt e n c lo s ur es 6.2.1.1 in c lud es: m a lfun c ti o n a n d/or in c ludin g p e n e tr a ti o n s e a l s throu g h . H o t ce ll liquid R S-0 1 m a int ena n ce Zo n e I ex h a u st ve ntil a tion 6.2.1.6 c o nfin e ment . H aza rd o us c h e mi ca l sys t e m , includ i n g du c tin g , bound ary s pill s filt e r s, and ex h a u s t stac k H ot ce ll R S-03 . Zo n e I inl et ve ntil at i o n system , . seco nd a r y i n c ludin g du ct in g , fi lt e r s , and c onfin e ment bubble-ti g ht is ol a tion damp e r s b o und a ry . V e ntilation co ntr o l sys t e m Hot ce ll s hi e ldin g R S-04 . Sec ond a r y i o din e r e m ov al b e d . boundar y . B e rm s Confinement IROFS Derived from Accident Analyses and Potential Technical Specifications Prim ary o ff g a s r e li ef R S-0 9 Di ss o l ve r offgas failur e . Pr ess ur e r e li ef d ev i ce 6.2.1.7.1 syste m durin g di sso lu t i o n . Pr ess ur e r elief tank o p era ti o n Active radiation RS-10 Transfer of high-dose Radiation monitoring and isolat i on 6.2.1.7.2 monitoring and process liquid outside the system for low-dose liquid isolation of low-hot cell shielding transfers dose waste transfer boundary Cas k l oca l R S-1 3 T a r g et claddin g le a k age L oca l c a ptur e ve nti l a tion sys t e m 6.2.1.7.3 ve nt i l a ti o n durin g during s hipm e nt ove r c lo s ur e lid durin g lid r e m ova l c l os ur e lid remo va l a nd d oc kin g pr e p ara tion s Cask docking port RS-15 Cask not engaged in cask Sensor system controlling ca s k 6.2.1.7.4 enabler docking port prior to docking port door operation opening docking port door Pro cess v e ss el FS-0 3 SS C d a m age du e t o B a ckup bottl e d nitr oge n gas 6.2.1.7.5 e m e r ge nc y pur ge h y dro gen d e tl agra tion o r s uppl y syste m d e ton a tion Irradiated target FS-04 Dislodging the target . Cask lifting fixture design th a t 6.2.1.7.6 cask lifting fixture cask shield plug while prevents cask tipping workers pre s ent during . Cask lifting fixture design th a t target unloading prevents lift from toppling activities during a seismic event 6-2

- - - .. NWMI ..... .*.* .. *.*. NOflllfWHT MCOtcAl ISOTOl'U NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-1. Summary of Confinement Engineered Safety Features (2 pages) Detailed Engineered safety SSCs providing engineered description feature IROFS Accident(s) mitigated safety features section Exhaust stac k height FS-0 5 . Equipment . Zone I exhaust stac k 6.2.1.7.7 malfunction resulting in liquid spill or s pray . Carbon bed fire Double-wall piping CS-09 Solution spill in facility Double-wall piping for selected 6.2.1.7.7 area where spill transfer lines containment berm is neither practical nor desirable for personnel chemical protection purposes Backflow CS-1 8 High worker expos ur e Backflow prev e ntion devices 6.2.1.7.9 p r evention devices from backflow of high-lo cate d on pro cess line s cro ss in g Safe geometry day CS-1 9 dose solution the hot ce ll shielding boundary t a nks Dissolver offgas . Potential limiting Di sso lver offgas iodine removal 6.2.1.8 iodine removal unit* control for operation units (DS-SB-600AIB

/C) . Primary iodine control system during normal operation Dis so lver offgas . Potential limiting Dissolv e r offgas primary adsorber 6.2.1.8.2 primary ad sor ber* contro l for operat ion unit s (DS-SB-620NB

/C) . Primary noble gas contro l system durin g normal operation Dis so l ver offgas . Potential limiting . Dissolver offgas vacuum 6.2.1.8.3 vacuum receiver or control for operation receiver tanks (DS-TK-700AIB) vacuum pump* . Motive force for . Dissolver offga s vacuum pump s dissolver offgas (DS-P-710NB) a Exa mple s of candidate technical specifica tion rather than e ngin ee r ed safe ty f ea ture. IROFS it em r e li ed on fo r safety. SSC = s tructure s , sys t e m s , and components. Table 6-2. Summary of Criticality Engineered Safety Features (2 pages) Engineered safety feature Interaction control spacing provided by pa ss ively designed fixtures and workstation placement Pencil tank, vesse l , or piping safe geometry confinement using the diameter of tanks , vessels, or pip i ng SSC features providing engineered safety features CS-04 Defines spacing between SSC components using geo metry to prevent nuc l ear criticality CS-06 Defines dimensions of SSCs using geometry to prevent nuclear criticality 6-3 -. . 6.3.1.2.1 6.3.1.2.2

..... NWMI *::**:*:* ..**.. " NOKTlfWHT MEDK:AL tSOTOPU NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-2. Summary of Criticality Engineered Safety Features (2 pages) Engineered safety feature P e n c il ta nk ge om etry c o ntr o l on fixe d int erac ti o n s pa c in g o f indi v idu a l tank s Floor and s ump geometry control on s lab depth , and s ump diameter or depth for floor dikes D o ubl e-w all pipin g Clo s ed s afe-geometry heating or cooling loop with monitoring and alarm S im p l e over fl ow to no rm a lly e mp ty safe-ge om e t ry t a nk w i th l eve l a l arm Condensing pot or seal pot in ventilation vent line S im p l e ove rflo w t o n o rm a lly e mpt y s a fe ge om e tr y fl oor w ith l e v e l al ar m in t he h o t ce ll co nt ai nm e nt b o und ary Acti v e discharge monitoring and isolation Ind e p e nd e nt acti ve di sc har ge m o ni to ring a nd i so l at i o n Backflow prevention device Safe g eom e tr y d ay tanks Evaporator or c oncentrator condensate monitoring P rocess ing c omp o n e nt s afe v olum e con fi n e m e nt Closed heating or cooling loop with monitoring and alarm IR OFS it e m re li e d on fo r safety. ' SSC features providing engineered safety features CS-0 7 D efi n es s pa c in g b e t wee n diff e r e nt SSCs u s in g geo m e tr y to pr eve nt nucl ea r c rit ica lit y CS-08 Defines sump geometry and dimensions for SSCs u s ing geometry to prevent nuclear criticality CS-0 9 D efi n es tran sfer lin e leak co n fi n e m e nt in l oca tion s wh e re s ump s und e r piping a r e n e ith e r feas ibl e no r d esira bl e -.

6.3.1.2.3 6.3.1.2.4 6.3.1.2.5 CS-10 Closed-loop heat transfer fluid systems to 6.3. I .2.6 prevent nuclear criticality or transfer ofhigh-dose material across shielding boundary in the event of a leak into the heat tran s fer fluid C S-11 Ove rflo w to p reve nt nucl ea r c ri t ic a lit y from 6.3.1.2. 7 fi ss ile s oluti o n e n te rin g n o n-geo m etr ic a ll y favora bl e ve n t il a ti o n e quipm e nt CS-12 Seal pots to prevent nuclear criticality from 6.3.1.2.8 fissile solution entering non-geometrically favorable ventilation equipment C S-1 3 Ove rfl ow t o pr eve nt nu c l ea r c ri t i ca l i ty from 6.3.1.2.9 fiss ile s oluti o n e nt e rin g n o n-geo m e tric a ll y fav orabl e ve ntil a ti o n e quipm e n t CS-14 Information to be provided in the Operating 6.3.1.2.10 License Application C S-1 5 Inform a tion w ill b e pro v id e d in t h e Op era tin g 6.3 .1.2.11 L i ce n se Appli ca ti o n CS-18 Backflow prevention to preclude fi s sile or high 6.3.1.2.12 dose solution from crossing shielding boundary to non-geometrically favorable chemical supply tanks and prevent nuclear criticality C S-1 9 A ltern a te b ac kfl ow pr eve nti on d ev i ce 6.3.1.2.1 3 CS-20 Prevent nuclear criticality from high-volume 6.3.1.2.14 transfer to non-geometrically favorable vessels in solutions with normally low fissile component concentration s C S-26 D e fin es v olum e of SS Cs t o pr eve n t nucl ea r c ritic a lity CS-27 Closed-loop , high-volume heat transfer fluid systems to prevent nuclear criticality or transfer of high-dose material across shielding boundary in the event of a leak into the heat transfer fluid with normally low fissile component concentrations SSC = s tru c tur es , s ys t e m s , a nd co mp o n ent s. 6-4 6.3.1.2.1 5 6.3.1.2.16 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2 DETAILED DESCRIPTIONS The PHA used to identify accidents in C hapt er 13.0 , Section 13.1.3 , assumed the fo ll owing known and credited safety features, or IROFS , are in place for normal operations:

*

Hot cell shielding bound ary, credited for shie ldin g workers a nd the public from direct exposure to radiat ion (a normal h azard of the operation)

Hot cell confinement boundaries , credited for confining the fissile and high-dose solids, liquid s , and gases , and contro lli ng gaseous releases to t he environment Administrative an d passive design features on uranium batch , volume, geo m etry, and interaction co ntrol s on the act i vities, credited for maintaining normal operations invo l vi n g the h and lin g of fissile material s ubcritic al (the PHA identified initi ators for abnormal operations that require further evaluat ion for IROFS satisfying the double-co ntingency principle)

This sec tion provides detailed descriptions of the engineered safety features identified by the acc id ent analyses s ho w n in Chapter 13.0. 6.2.1 Confinement The PHA was based on a definition for confinement , as follows: Confinement -An enclosure of the facility (e.g., the hot cell area in the RPF) that i s designed to limit the exc han ge of effluents between the e ncl osure and its externa l environment to controlled or defined pathways.

A confineme nt should include the capab ility to maintain sufficient internal negative pressure to ensure inleakage (i.e., prevent uncontrolled l eakage outside the co nfin ed area), but n eed not be capable of supporting positive intern a l pressure or significantly shieldi n g the externa l enviro nment from internal sources of direct radiation.

Air movement in a confinement area could be integrated into the h eating, venti l atio n , and air co nditionin g (HV AC) systems, including exhaust stac ks or vents to the external environment, filters, blo wers, and dampers (ANSI/ ANS-15.1 , Th e Development of Tec hni cal Specifications for R esea r c h Reactors).

Confinement describes the low-leakage boundary s urroundin g radioact i ve or hazardous chemica l materials released during an accident to fac ili ty regio n s surrounding the physical process equipment containing process materials.

The confinement systems localize releases of radioact i ve or hazardous materials to controlled areas and mitigate the consequences of accidents.

The principal design and safety objective of the confine m e nt system i s to protect on-site workers, the public , a nd environment.

Per sonnel protection contro l features (e.g., adequate s hi e ldin g and venti l ation control) wi ll minimize h azar d s normally associated w ith radioactive or chemical material s. The seco nd design objective is to minimize the reliance on admin i strative or complex active eng in eering controls and provide a co nfinem ent syste m that i s as sim pl e and fa il-safe as reasonably possible. This s ubsection describes th e confinement systems for the RPF. The RPF confine m ent areas will consist of hot cell and glovebox e nclosur es housing process operations , tanks , and piping. Co nfin eme nt will be provided by a combination of the enc lo s ure bound aries (e.g., wa ll s, floor , and cei lin g), enclos ure ventilation , and ventilatio n control syste m. The enc lo s ur e boundari es will restrict bulk quantities of process material s, potentially present in so lid or liquid forms , to the confinement a nd limi t in-leakage of gaseous components co ntrolled by the ve ntila tion system. The ventilation and ventilation co ntrol systems will restrict the gaseous components (i ncludin g gas phase components and so li d/liquid dispersions) to the co nfin ement. Figure 6-1 provide s a simp lifi ed sc hematic of the confinement venti l at ion system, which i s described in mor e detail as the Zone I ventilation sys tem in Chapter 9.0, "Auxiliary Systems." 6-5 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [P rop r ietary I n fo rm atio n] So u rce: Figure 2-5 ofNWMI-2015-SDD-O 13 , Syste m D e sign D e s c r i ption for Venti l ation , Rev. A , Nort h west Me d ical I sotopes , LLC , Corva ll is, O r egon , March 20 1 5. Figur e 6-1. Simplified Zone I V entilation Schematic 6-6 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safet y Features A typical glovebox enclosure is shown in Figure 6-1, and the inlet does not have an automatic closure on the isolation damper. During development of the final safety analysis and Operating License Application , each glovebox will be evaluated based on inventory of concern (e.g., fission product gases) and hazards to determine ifthe inlet isolation damper is required to be an IROFS confinement control. Until the analysis is complete, the design of gloveboxes will include a bubble-tight isolation damper, as required , for the hot cells. T he enclosure boundary of the hot cells will also function as biological shielding for operating personnel.

Shielding functions of the hot cells are discussed in Chapter 4.0, "Radioisotope Production Facility Description." Hazardous chemical confinement will be provided by berms located within the RPF to confine spilled material to the vicinity where a spill may originate.

6.2.1.1 Confinement System Confinement system enclosure structures , ventilation ducting, isolation dampers , and Zone I exhaust filter trains are designated as IROFS. Table 6-3 provides a description of the system component safety functions.

Figure 6-2, Figure 6-3 , and Figure 6-4 indicate the general location of confinement structure boundaries to the facility ground level, mechanical level, and lower level layouts , respectively.

The confinement system is an engineered safety feature that performs the functions identified by IROFS RS-01 , RS-03, and RS-04 in Chapter 13.0. Table 6-3. Confinement System Safety Functions System, structure, component Zone I enclosure inlet isolation dampers and ducting leading from isolation dampers to enclosures Zone I enclosure exhaust ducting leading from enclosures to the exhaust stack, filters, and exhaust stack Process vessel vent exhaust ducting leading from process vessels to Zone I exhaust plenum Ventilation control system Secondary iodine removal bed Hot cells, tank vaults, and g l ovebox enclosure structures IROFS = item relied on for safet y. Description lff 6W1itt§ii!!.1.1 Provide confinement isolation at Zone I/Zone II IROFS enclosure boundaries Provides confinement to the confinement exhaust IROFS boundary Provides confinement to the confinement exhaust IROFS boundary Provides stack monitoring and interlocks to IROFS monitor discharge and signal changing on service filter trains during normal and abnormal operation Mitigates a release of the iodine inventory in the IROFS dissolver offgas treatment system Provide solid, liquid , gas confinement IROFS 6-7 NWM I ..**.. *

NO<<fHWlst MfDK:M. ISOTOPU NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Propr ietary In formation] So u rce: Figure 2-1 ofNWMI-20 1 5-SDD-0 1 3 , Syst e m D es ign D e scripti o n for V e ntilation , Rev. A , Northwest Med i cal I sotopes , LLC , Corva lli s, Oregon , Marc h 20 1 5. Figure 6-2. Ground Level Confinement Boundar y 6-8 -----------1

NWM I ...... NOlmfWEST MfDtCAl tSOTDPH NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [Propri etary inform a tion] So ur c e: Figu r e 2-2 ofNWM I-20 1 5-SDD-013 , S y st em D es i g n D esc ripti o n for V e ntilation , Re v. A , Nort h west Medical I so t opes , LLC , Corva lli s , Oregon , Marc h 20 1 5. Figure 6-3. Mechanical Level Confinement Boundar y 6-9 NWM I ......

HOflTHWEST M£DtcAl ISOTOPU NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

Sour c e: Figure 2-3 ofNWM l-2015-S DD-O 1 3, S ys t em D es i gn D e s c ripti o n fo r Ve ntilati o n , R ev. A , Northwe s t Medical I s otop e s , LLC , C orva lli s , Ore g on , March 2015. Figure 6-4. Lower Level Confinement Boundary During normal operation , passive confinement is provided by the contiguous boundary between the hazardous materials and the surrounding environment and is credited with confining the hazards generated as a result of accident scenarios.

The boundary includes the enclosure structure s and extension of the structures through the Zone I venti l ation components. The intent of the passive boundary is to confine h azardous materials while a l so preventing disturbance of the h azardous material inventory by external energy sources. This passive confinement boundary extends from the isolation valve downstream of the intake high-efficiency particulate air (HEPA) filter to the ex h a u st stack. An event that result s in a release of process material to a confinement enc l osure wi ll be confined by the enclosure structural components. Eac h process line that connects with ves s els located outside of a confinement boundary with vessels located inside a confinement boundary will be provided with back.flow prevention devices to prevent releases of gaseous or liquid material.

The backflow prevention devices on piping penetrating the confinement boundary are designed as p a ssive devices and will be located as near as practical to the confinement boundary or take a position that provides greater safety on loss of actuating power. The consequences of an uncontrolled release within a confinement enc lo sure , and the off-site consequences ofreleasi n g fission products through the ventilation system , will be mitigated by use of an active component in the form of bubble-tight isolation dampers as IROFS on the inle t ve ntil ation ducting to each enclosure.

This engineered safety feature reduces the ducting to the confinement vo lum e that needs to remain intact to achieve e nclo sure confinement.

The dampers wi ll close automatica ll y (fail-closed) on lo s s of power , and the venti l ation system wi ll automatica ll y be placed into the passive ventilation operating mode. Overall performance assurance of the active confinement components will be achieved through factory testing and in-place testing. Duct and hou sing leak tests will be performed in accordance wit h minimum acceptance criteria , as specified in ASME AG-I , Cod e on Nucl e ar Air and Gas Treatm e nt. Specific owner requirements wit h respect to acceptable leak rates will be based on the safety analysis. 6-10 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Berms wi ll employ a passive confinement methodology. Passive confinement wi ll be achieved through a continuous boundary between the hazardous materials and the s urrounding area. In the event of an accidental relea se, the hazardous liquid wi ll be confined to limit the exposed surface area of the liquid. 6.2.1.2 Accidents Mitigated The hot ce ll confinement system and shie ldin g boundary are credited as being in place by the accident analysis in Chapter 13.0 , Section 13.1.3.1.

Accidents mitigated co n s i st of equipme nt malfunction events that re s ult in the release of radioactive material or hazardou s chemica l s to a confi nem e nt enclosure.

The confinement system is also credited with mitigating the imp act of a non-specific initiating event resulting in release of the iodine inventory in the dissolver offgas treatment system. 6.2.1.3 Functional Requirements Functional requirements of the confineme nt structura l components include: *

Capt urin g and containing liquid or solid releases to pre ve nt the material from exiting the boundary and causing hi gh dose to a worker or member of the public or producing significant environment contamination Preventing spills or sprays of rad i oactive solut ion that are acidic or caustic from causing adverse exposure to personnel through direct contact with skin , eyes, a nd mucus membranes where the com bin ation of c h emical expos ur e and radiologica l contamination would lead to serious inju ry and long-lasting effects Functional requirements of the confineme nt venti l ation components include: * *

Providing negative air pres s ure in the hot ce ll (Zone I) relative to lower zones outside of the hot ce ll using exhaust fans equipped with HEPA filters a nd high-efficiency gas adsorbers (HEGA) to reduce the release ofradionuc lid es (both particulate and gaseous) outside the primary confinement boundary to below Title 10 , Code of F ederal R egu l ations , Part 20, "Sta ndards for Protection Against Radiation" (10 CFR 20) release limi ts during normal and abnormal operations. Mitigating high-do se radionuc lid e relea ses to maintain exposure to acceptable l evels to workers and the public in a highly reliable and availab l e manner. The hot cell secondary confinement boundary will perform this function using a system of passive and active e n gi n eered features to ensure a high level of reliability and availability.

Removing iodine i sotopes present in the process vessel ve nt und er accident conditions to comply with I 0 CFR 70.61, "Performance Requirements

," for an intermediate consequence re l ease. Berm s confi nin g potential h azardous c hemic a l spills are de s igned to hold the entire contents of the container in the event the container fai l s. 6.2.1.4 Confinement Components The following components are associated with the confinement barriers of the hot ce ll s, tank vau lt s, and g lo veboxes. The specific materials, construction , installation , and operating requirements of these components are evaluated based on the safety analysis.

6-11

.... ;. NWMI ::.**.*.* NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features * * *

NOATHWUT MEDICAL I SOTOPH Confinement structural components include the following.

* * *

Sealed flooring will provide multiple layers of protection from release to the environment.

Diked areas will contain specific releases. Sumps of appropriate design will be provided with remote operated pump s to mitigate liquid spills by capturing the liquid in appropriate geometry tanks. In the molybdenum-99 (99 Mo) purification clean room, s maller confinement catch basins will be provided under points of credible spill potential in addition to the sealed floor. Entryway doors into a designated liquid confinement area will be sealed against c r edible liquid leaks to outside the boundary.

Piping penetrations and air ducts will be located to minimize the potential for liquid leak s across the confinement boundary.

Ventilation system components that are credited include the following.

* * *

Zone I inlet HEPA filters will provide an efficiency of greater than 99.9 percent for removal of radiological particulates from the air that may reverse flow from Zone I to Zone II. Zone I ducting will ensure that negati ve air pressure can be maintaine d by conveying exhaust air to the stack. Bubble-tight dampers will be provided to comply with the requiremen t s of ASME AG-1 , Section DA-5141. Ventilation ductwork and ductwork support material s will meet the requirements of ASME AG-1. Supports will be de signe d and fabricated in accordance with the requirements of ASME AG-1. Zone I exhaust train HEPA filters will provide an efficie ncy of greater than 99.95 percent for removal of radiological particulates from the air that flows to the stack. Zone I exhaust train HEGA filters will provide an efficiency of greater than 90% for removal of iodine. The Zone I exhaust stack will provide disper s ion ofradionuclides in n orm al and abnormal releases at a discharge point of 23 meters (m) (75 feet [ft]) above the building ground level. Stack monitoring and interlocks will monitor discharge and signal changing of service filter trains during normal and abnormal operations. Secondary process offgas treatment iodine removal beds (VV-SB-520) will m i tigate an iodine release. 6.2.1.5 Test Requirements Engineered safety features will be tested to ensure that components maintain operability and can provide adequate confidence that the safety system performs satisfactorily during postulated events. The confinement engineered safety features that initiate the system interlocks are des igned to permit testing during plant operation.

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

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

6-12 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2.1.6 Design Basis Codes and standards are discussed in Chapter 3.0 , " D esign of Structures, Systems, and Compo nent s." The design bases for Zone I and Zone II ventilation syste m s are described in C hapt er 9.0. The design b a sis of con finem ent enc lo s ur e structures i s de scribe d in C h apter 4.0. C h apter 7 .0, "Instrume ntation and Control Systems," identifies the engineered safety feature-r elated design basis of the ventilation control system. The following information was developed for the Construction Permit Application to describe the process offgas seco nd ary iodine removal bed: * * * * *

Sorbe nt bed of [Proprietary Information]

Iodine removal efficiency greater than [Proprietary Information]

Nomina l superficia l gas flow ve lo c i ty of [Proprietary Information]

Nominal sorbent bed operating temperature of l ess than [Proprietary Information]

Nominal sorbent bed depth of [Proprietary Information]

Nominal gas relative humidity l ess than [Proprietary Information]

Additional detailed information on the process offgas iodine retention b ed design basis will be developed for the Operating License Application. Pot ential va riabl es , cond ition s, or ot her item s that w ill be probable s ubje cts of a technical speci fication associated with the RPF confinement systems and co mpon ents are discu ssed in C h a pt er 14.0, "Tec hnical Specifications." 6.2.1.7 Derived Confinement Items Relied on for Safety The following subsections describe addit ion al engineered safety featu r es that are derived from the acc ident ana ly ses described in Chapter 13.0 and are projected technical specificatio n s d efining limited conditions for operation.

6.2.1.7.1 IROFS RS-09 , Primary Offgas Relief System IROFS RS-09, "Primary Off gas Relief System," is identified by the accident ana l ysis in Chapter 13.0. As a n active eng ine ered control (AEC), the primary offgas relief system will be a component included in th e offgas train for the two irradiated target dissolvers.

The dissolver offgas system is intended to operate at a pressure that is l ess than the confinement enclosures to maintain gaseo u s components generated during dissolution within the vesse l s and route the gaseou s components through the offgas treatment unit operations.

The primary offgas relief system , or pressure r e l ief tank , will be used to confine gases to the dis s olver and a portion of the dis s olver offgas equipme nt , ifthe offgas m otive force (vacuum pumps) ceases operation during dissolution of a dissolver batch. 6-13 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Figure 6-5 is a diagram of the dissolver off gas system process , which shows the pressure relief tank position in the off gas treatment equipment train. Figure 6-6 shows the location of the pressure relief tank within the RPF hot ce ll (identified as "pressure relief').

[Proprietary Information]

Figure 6-5. Dissolver Offgas System Engineered Safety Features 6-14 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Proprietary Inform ation] Figure 6-6. Dissolver Offgas Hot Cell Equipment Location The pressure relief tank will be evacuated to a specified , subatmospheric pressure prior to initiati n g dissolution of a target batch and selected valves (indicated as 2 , 3 , and 4 on Figure 6-5) closed. Valve 1 w i ll be open during normal dissolver operation.

An upset during the dissolver operation (e.g., loss of vacuum pump operation) will result in closing Valve 1 a nd opening Valve 2 to contain dissolver offgas w i thin the dissolver and off gas vesse l s. Due to the s hort duration of dissolver operation , dissolution is assumed to go to comp l etion ind ependent of an off gas system upset. The pressure relief tank wi ll contain the offgas as dissolution i s completed. Va l ves 3 , 4 , and 5 are provided for upset recovery.

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

6-15

.. ;.;: .. NWMI ..... .......... ' * * ' NO<<THWUT MfDICAl ISOTOl'ES NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Accident Mitigated

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

Functional Requirements

As an AEC , use relief device to relieve pressure from the system to an on-service r eceiver tank maintained at vacuum with the capacity to hold the gases generated by the dissolution of one batch of targets in the target dissolver Prevent a failure of the primary confinement system by capturing gaseous effluents in a vacuum receiver tank Design Basis The following information was developed for the Construction Permit Applica t ion describing the pressure relief tank. * * * * * *

Pressure-relief tank sizing is based on a maximum dissolver batch of [P roprietary Information]

that has just started dissolution when the pressure relief event is initiated. The non-condensable gas vo lume to the pressure relief tank is equivalent to all nitrogen oxide (NOx) generated by dissolution, plus the sweep gas flow for flammable hydrogen gas miti gation. Worst-case reaction stoichiometry of [Proprietary Information]

dissolved i s used . No credit is taken for reaction ofN0 2 with water to produce nitric acid . Dissolver gas ad dition s , other than the minimum sweep gas flow for hydrogen mi t igation, are terminated by the pressure relief event. Gas contained by the pressure relief tank and associated dissolver offgas piping is saturated with water vapor. The pressure change from [Proprietary Information], absolute activates the pressure relief tank . Additional detailed information on the pressure relief tank design basis will be developed for the Operating License App lication. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additiona l detailed information on test requirements will be developed for the Operating L icense Application.

6-16 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2.1.7.2 IROFS RS-10, Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer IROFS RS-10, "Active Radiation Monitoring and I so l at ion of Low-Do se Waste Transfer ," is identified by the accide nt analyses described in Chap t e r 1 3.0. As a n AEC , the r ec ircul ating strea m a nd the di sc harg e st r eam of the lo w-d ose waste tank wi ll be s imult aneo u s l y monitored in a b ackgro und s hi elded trunk outside of the h ot cell shiel d ed cavity. The continuous gamma instrument will monitor the transfer lin es to provide a n open permissive signa l to d e dic ated isolation valves. Accident Mitigated Transfer of high-dose process liquid solutions o ut s ide the hot cell s hielding boundary System Components Additional d etai l ed information of the radiation monitor a nd isolation oflow-dose waste transfe r s will be developed for the Operati n g License Ap pli cation. Functional Requirement Maintain worker a nd public exposure rates w ithin approve d limit s Design Basis Additional d etailed informat ion of the rad i ation monitor and isolation of low-do se waste transfers wi ll be developed for the Operati n g Licen se App lic ation. Test Requirement s The above a nal ysis is based on information d evelope d for the Co n s tructi on Permit App li cation. Additional d etai l ed information on test requirements wi ll be developed for the Operating License Application.

6.2.1.7.3 IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations IRO FS RS-13 , "Cask Loca l Ventilat ion During C l osu r e Lid R emova l and Docking Preparations

," is id en ti fied by th e accident a n a l yses described in Chapte r 1 3.0. As a n AEC , a local capture venti l at ion sys t e m will be u sed ove r the ir radiate d target cask c lo s ur e lid to remove any esca p e d gases from the worker breathing zone during removal of the closure lid , remo va l of the shielding block bolts , and in sta llation of the lifting lugs. Accident Mitigated

Irradiated target c l a ddin g fails during transportation , releasi n g gaseous rad ionu c lid es wit hin th e cask containment boundary System Components Use a d edicated evac u ation hood over the top of t h e cask during containme n t clo s ure lid removal

R emove gases to the Zone I seco nd ary confi n e m e nt syste m for processing 6-17 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Functional Requirement

Prevent exposure to workers by evacuating any high-dose gaseous radionuclides from the worker breathing zone a nd preventing immersion of the worker in a high-d ose enviro nm ent Design Basis The following information was developed for the Construction Permit Application describing the cask local vent ilation system: Use the lo cal capture ventilat ion system to evacuate and backfill the cask with fresh air (from a protected pressurized source suc h as a compressed bottle) until the atmosp h eres are within approved safety limit s Additiona l detailed information on the cask local venti l ation system design basis will be developed for the Operating License App lication. Test Requirements The above analysis is based on information developed for the Const ruction Permit Application.

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

n. 6.2.1.7.4 IROFS RS-15 , Cask Docking Port Enabling Sensor IROFS RS-15, "Cask Docking Port Enabling Sensor ," i s identifi ed by the acci d e nt analyses described in Chapter 13 .0. As an AEC , the cask docking port will be equipped with sensors that detect when a cask is mated wit h the cask docking port door. Accident Mitigated

Cask lift failure occurs after shie ld plug removal (but before target basket removal) with targets inside the cask System Components Ena blin g contact signal and positive closure signa l when the sensor does not sense a cask mated to the cask docking port , ca u sing the cask docking port door to close Functional Requirement

Prevent the cask docking port door from being opened and allowing a strea min g radiatio n path to areas accessible by workers Design Basis Detailed information on the system design basis wi ll be developed for the Operating License Application.

Test Requirements The a bove analysis is ba sed on information developed for the Co n st ruction Permit Application.

Additional det ai l ed information on test requirements will be developed for the Operating L icense Application.

6-18 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2.1.7.5 IROFS FS-03, Process Vessel Emergency Purge System IROFS FS-03, "Process Vessel Emergency Purge Syste m ," is identified by the accident analyses described in Chapter 13.0. Hydrogen gas will be evo l ved from process solutions through radio lytic decomposition of water in the high radiat ion fields. An air purge to the vapor space of selected tanks will be provided by the facility air compressors to control the hydrogen concentration from radiolysis in vessel vapor space to below the flammability limit for hydrogen. As an AEC , an emerge n cy backup set of bottled nitrogen gas will be provided for all tanks that ha ve the potential to evo l ve significant vo lum es of h ydrogen gas through the radiolytic decomposition of water (in both a s h ort-and lon g-term storage condition).

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

Prevent development of an exp lo s iv e hydrogen-air mixture in the tank vapor spaces to prevent the deflagration or detonation hazard Design Basis The following informatio n was developed for the Constr u ct i on Permit Application describing the process vessel eme rgenc y purge system: * * *

Monitor the purge pressure going into the individual tanks and open an isolation valve on low pressure (setpoint to b e determined) to restore the continuo u s sweep of the system using nitro gen Provide sweep gas sufficient for the facility to a llo w repair of a compressed gas system outage Activate by se nsing lo w pressure on the normal sweep air system, introducing a continuous purge of nitro gen from a reliable emergency backup s t ation of bottled nitrogen into each affecte d vessel near the bottom (e.g., through a liquid level detection leg) of the vessel Dilute hydrogen as it rise s to the top of the vesse l and is vented to the respective vent system A d ditional detailed informat ion on the process vessel emergency purge system design basis will be developed for the Operating License Application. Test Requirements The above ana ly sis is based on information developed for the Const ruction Permit Application. A dditi ona l detailed information on test requirements will be developed for the Operating License App li cation. 6.2.1.7.6 IROFS FS-04, Irradiated Target Cask Lifting Fixture IROFS FS-04, "Irradiated Target Cask Lifting Fixture," i s identified by the accide nt a nal yses described in Chapter 13.0. As a passive e n gineered control (PEC), the irradiated target cask liftin g fixture will be designed to prevent the cask from tipping within the fixture and the fixture itself from toppling during a seismic event. 6-19 Accident Mitigated NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Dislodged irradiated target shipping cask shie ld plug in the presence of workers during target unloading activities System Components Detailed information on the system components wi ll be d eveloped for the Operating License Application.

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

Design Basis Detailed information on the system design basis w ill be developed for the Operating License Application. Test Requirements The above a n a l ysis is based on informat i on developed for the Construction Permit Application. Additional det ailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.7 IROFS FS-05, Exhaust Stack Height IROFS FS-05, " Exhaust Stack Height," i s identified by the acciden t analyses described in C hapt er 13.0. Accidents Mitigated Process solutio n spills and sprays Carbon bed fire System Component Zone I exhaust stack Functional Requirement

Provide an offgas release height for venti l ation gases consistent with the stack height u sed as input to mitigated dose conseq u ence evaluations.

Design Basis The Zone I ex h aust stack height is 23 m (75 ft). Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additiona l detailed information on test requirements will be developed for the Operating Lice n se Applicat ion. 6-20 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2.1.7.8 IROFS CS-09, Double Wall Piping IROFS CS-09, "Double Wall Piping ," is identified by the accident analyses in Chapter 13.0. This IROFS has both a confinement and nuclear criticality prevention function.

As a PEC , the piping system conveying fissile so lution between credited confinement locations will be provided with a double-wall barrier to contain any spills that may occur from the primary confinement

[Proprietary Information]

Figure 6-7. Proposed Location of Double-Wall Piping (Example) pi pi ng. This IROFS will be used at those location s that pa ss through the facility , where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection p u rposes. Figure 6-7 provides an example location where IROFS CS-09 will be applied (e.g., the transfer li ne between the recycle uranium decay tanks and the [Proprietary Information]). Accident Mitigated Leak in piping that passes between confinement enclosures System Components The following double-wall pipin g segme nts are identified at this time: * *

Transfer piping containing fissile solutions traversing between hot cell walls Transfer piping connecting the uranium product transfer send tank (UR-TK-720) and uranyl nitrate storage tank (TF-TK-200)

Other locations to be identified in final design Functional Requirements

Double-wall piping pre vents personnel injury from exposure to acidic or caustic licensed material so lution s conveyed in the piping that runs outside a confinement enclos ure Double-wall piping route s pipe leaks to a critically-safe leak collection tank or berm as a nuclear criticality control feature Design Basis The double-wall piping arrangement is designed to gravity drain to a safe-geometry set of tanks or to a safe geometry berm. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6-21 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2.1.7.9 IROFS CS-18, Backflow Prevention Devices, and IROFS CS-19, Safe-Geometry Day Tanks IROFS CS-18, "Backflow Prevention Devices ," and IROFS CS-19 , "Safe-Geometry Day Tanks," are identified by the accide nt analyses in Chapter 13.0. As a PEC or AEC, chemical and gas addition ports to fissile process solution systems wi ll enter a confinement enclosure through a backflow prevention device. Backflow prevention devices and safe-geometry day tanks will provide alternatives for preventing process addition backflow across confinement boundaries. The device may be an anti-sip hon break , an overloop sea l , or other active engineering feature that addresses the conditions ofbackflow and prevents fissile solution from entering non-safe geometry systems or high-dose so lution s from exiting the hot cell shielding boundary in an uncontrolled manner. Therefore , these IROFSs have both a confinement and a nucle ar criticality prevention function.

Accident Mitigated Backflow of process material located inside a confinement bound ary to vessel loc ate d outside confinement via connected piping due to process upset. System Components System component information will be provided in the Operating License Application.

Functional Requirements

* * *

Prevent fissile so lution s and/or high dose so lution s from backflowing from the tank into systems outside the confinement boundaries that may l ead to accidental nuclear criticality or high exposures to workers Provide each hazardous location with an eng ine ered backflow prevention device that provides high reliability a nd availability for that location Locate the backflow prevention device features for high-dose product so lution s inside the confinement boundaries Support the backflow prevention devic es with safe-geometry day tanks located inside the confinement boundary Direct spills from the backflow prevention device to a safe-geometry co nfin emen t berm Design Basis Design basis information will be provided in the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

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

6-22 6.2.1.8 Dissol ver Offgas Systems 6.2.1.8.1 Dissolver Offgas Iodine Removal U nit NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features A significant fraction of iodine entering the RPF in targets is projected to be released to dissolver offgas during target dissolution.

The dissolver offgas iodine removal units will be included in the RPF as the primary SSCs for contro llin g the release of iodine isotopes to the environment or faci li ty areas occupied by workers. Components of the dissolver offgas system , beginning with the iodine removal unit , will also be used to treat vent gas from the target disassembly system. Target disassembly vent gas is treated by dissolver offgas components for the Construction App l i cat i on Permit configuration as a measure to mitigate the unverified potential for a release of fission gas radion u c lide s during target transportation. Figure 6-5 (Section 6.2.1.7.1) shows the iodine removal unit position in the offgas treatment equ ipm ent train. The dissolver offgas iodin e remova l unit location in the facility i s shown in Figure 6-6 (ide ntifi ed as " primary fission gas treatment"). Accidents Mitigated Projected limiting control for operation Required for norma l operation and not for accident mitigation System Components Iodine removal unit A (DS-SB-600A)

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

Iod ine removal unit C (DS-SB-600C)

Functional Requirement Remove iodine isotopes from the dissolver offgas during normal operations s u ch that the dose to workers complies with 10 CFR 20.1201 , " Occupational Dose Limits for Adu lt s," and the do se to the public complies with 10 CFR 20.1301 , "Dose Limits for Individual Members of the Public." Design Basis The followi ng in formation was developed for the Construction Permit Application describing eac h individual iodine removal unit: Sor bent bed of [Proprietary Information]

Iodine removal efficiency greater than [Proprietary Information]

Nomina l superficia l gas flow velocity of [Proprietary Information]

No minal sorbent b ed operating temperature of [Propri etary Information]

Nominal sorbent bed depth of [Propri etary Information], providing iodine removal capacity of greater than 1 year (yr). Additional det ai l ed information on the iodine removal unit design basis will be developed for the Operating License Application.

Te s t Requirements Th e above analysis is based on informa tion developed for the Const ruction Permit App lic ation. Additional detailed information on test requirements wi ll be developed for the Operating License Application. 6-23 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.2.1.8.2 Dissolver Offgas Primary Adsorber Noble gases (krypton [Kr] and xe non [Xe]) entering the RPF in targets are proj ecte d to b e released to dissolver offgas during target dissolution. The dissolver offgas primary adsorber units will be included in the RPF as the primary SSCs for controlling the release of noble gas isotope s to the environment or facility areas occupied by workers. Components of the dis so l ve r offgas syste m will also be used to treat vent gas from the target disassembly system. Target disassembly vent gas is treated by di ss olver offgas components for the Construction Application P e rmit configuration as a mea s ure to mitig ate the unverified potenti a l for a release of fi ssion gas radionuclides durin g target transportation. Figure 6-5 (Section 6.2.1. 7 .1) shows the primary adsorber position in the off gas treatment e quipment train. The dissolver offgas prim ary adsorber location in the facility is s hown i n Figure 6-6 (identified as "primary fission gas treatment").

Accidents Mitigated Projected limiting control for operation Requir e d for normal operation and not for accident mitigation System Components Primary adsor ber A (DS-SB-620A)

Primary adsorber B (DS-SB-620B)

Primary adsorber C (DS-SB-620C)

Functional Requirement Delay the release of nobl e gas isotopes v ia the di ssolver off gas during normal operations such that the do se to workers complies with I 0 CFR 20.1201 and the do se t o the public complies with 10CFR 20.130 1. Design Basis The following information was developed for the Construction Permit Application describing each individual primary adsorber unit: * * * * *

Sorbent bed of [Proprietary Information]

Nominal sor bent bed operating temperature of [Proprietary Information]

Nominal gas relative humidity le ss than [Proprietary Information]

Average gas flow rate of [Proprietary Information]

Nominal su perficial gas flow velocity of[Proprietary Information]

Delay time for relea se of Xe isotopes of I 0 days and Kr isotopes of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s (hr) (ad ditional delay time is provided by the seco ndary adsorber)

Additional detailed information on the primary adsorber unit design ba s is will be developed for the Operating Licen se Application. 6-24 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Test Requirements The above analysis is based on information developed for the Construction Permit App li cation. Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.8.3 Dissolver Offgas Vacuum ReceiverNacuum Pump The dissolver offgas vacuum pump wi ll provide the motive force for transferring offgas, generated in the dissolvers and disassembly equipment during operation, through the dissolver offgas equipment train while maintaining dissolver vessels at a pressure less than the equipment enclosure pressure.

Vacuum receiver tanks will be provided as part of the motive force system to allow the vacu um pumps to cycle on and off less frequently and accommodate the wide variations in gas flow rate associated with a target dissolution cycle. Figure 6-5 (Section 6.2.1. 7.1) shows the vacuum receiver tank and vacuum pump positions in the off gas treatment equipment train. The vacuum receiver tank and vacuum pump location in the facility is s hown in Figure 6-3 in the vicinity of equipment identified for the process offgas secondary iodine removal bed. Accidents Mitigated Projected limiting control for operation Required for normal operation and not for accident mitigation System Components Vac uum receiver tank A (DS-TK-700A) Vacuum receiver tank B (DS-TK-700B)

Vacuum pump A (DS-P-710A)

Vacuum pump B (DS-P-710B)

Functional Requirements

Maintain the dissolver vessel gas space at a pressure less than the dissolver vessel enclosure pressure throughout the target dissolution cyc l e Accommodate pressure drops associated with dissolver offgas unit operations over the ra n ge of gas flow rates generated in both dissolvers and the target disassembly equipment vent throughout a target dissolution cycle Design Basis The following information was developed for the Construction Permit Application describing the vacuum receiver tanks and vacuum pump: * * *

Minimum inlet setpoint pressure of [Proprietary Information]

Maximum inlet setpoint pressure of [Proprietary Information]

Outlet pressure of [Propri etary Information]

Maximum sustained gas flow into [Proprietary Information]

Receiver tank provides a [Proprietary Information]

with the vacu um pump off and inlet at the maximum sustained gas flow 6-25

..... NWMI *::**::* ...... ' * * ' NCHITlfWUT MEDtcAl ISOTOH.S NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Additional det ailed information on the vacu um receiver tank and vacuum pump design basis wi ll be developed for the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating Lice n se Application. 6.2.1.9 Exhaust System The venti l ation exhaust system is described in Chapter 9 .0 , Section 9 .1.2. Additional detailed information will be developed for the Operating License Application , including: *

Describing changes in operating conditions in response to potential accide nt s and the mitigation of accident radio l ogical conseq u ences Demonstrating how dispersion or distribution of contaminated air to the environment or occupied spaces is contro ll ed Identifying the design bases for location and operating characteristics of the exhaust stacks 6.2.1.10 Effluent Monitoring System Each RPF ex h aust stack will include an effluent monitoring system. The mon i toring system sample lines are designed to comply with ANSI Nl3.l , Sampling and Monitoring Rel e a se s of Airborn e Radioactiv e Substanc es from th e Sta c ks and Du c ts of N ucl e ar Fa c iliti es. Add itional detailed information on the effluent monitoring systems will be developed for the Operating License App li cat ion. 6.2.1.11 Radioactive Release Monitoring The effluen t monitoring system wi ll provide flow rate , temperature , and compos ition inputs for dispersion modeling of releases from the exhaust stacks. These input s will provide the capa bili ty for calculating potential exposures as a basis for actio n s to en s ure that the public is protected during both normal operation and accident conditions.

Ad diti onal detailed information on radioact i ve release monitoring will be developed for the Operating License Application.

6.2.1.1 2 Confinement System Mitigation Effects Detailed information describing the co nfin ement system mitigation effects will be developed for the Operating License App li cation. This information wi ll compare the radiolog i ca l expo s ure s to the facility staff and the public with and without the confinement system e n gineered safety feature. The comparison will be based on ana l yses showing airflow rates , reduction in quantities of airborne radioactive material by filter systems, s ystem isolation , and other parameters that demonstrate the effect i veness of the system. 6-26 6.2.2 Containment NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Containment for the RPF is defined based on NUREG-1537 , Guid e lin es for Pr e parin g and R ev iewing Appli c ations for the Li c ensing of Non-Pow er R e a c tors -Format and Cont e nt , Part 1 int erim staff guidance. Containm e nt s ar e r e quir ed a s an e ngin ee r e d s af ety f e ature on th e ba s i s of th e radi o isotop e production fa c ility de s ign, op e rating charact e ristics , a cc idents sc e narios , and lo c ation. A pot e ntial sce nario for s u ch a r e l e as e co uld b e a s ignifi c ant loss o f int eg rity o f th e radioisotope e xtra c tion sy st e m or th e irradiat e d fu e l pro ce ssing s ys t e m. The c ontainment is d es ign e d to co ntrol th e r e l e a se to the e n v ironm e nt of airborn e radi o a c tiv e mat e rial that is r e leas e d in th e fa c ility eve n if the accid e nt is a c compani e d by a pr ess ure s urg e or s t e am r e l e as e. The NUREG-15 37 Part 1 interim staff guidance has been applied to the RPF target processing systems. The current accident ana ly s i s described in C hap ter 13.0 h as not identified a need for a containme nt system as an engineered safety feature. 6.2.3 Emergency Cooling System An emergency cooling system for the RPF is defined by NUREG-153 7 Part 1 interim staff guidance. In the ev e nt of th e loss o f an y requir e d primary or normal cooling sys t e m , an e m e rg e ncy c ooling sys t e m ma y b e r e quir e d to r e m ove d ec a y h e at fr o m th e fu el t o pr e v e nt th e failur e or d e gradation of th e gas managem e nt s ys t e m, th e isotope e xtraction s ys t e m , or th e irradiated fue l pro cess ing s y st e m. An evaluation of RPF coo lin g requirements provided in Chapter 5.0 , " Coolant Systems ," indicates that an emergency cooling system will not be required to avo id rupture of the primary process vessels. In addition , the current accident analysis described in Chapter 13.0 has not identified a need for an emergency cooling system as an engineered safety feature. 6-27

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... NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features ' * *.* *

Notn'HWEST MEOtCAl ISOTO"S 6.3 NUCLEAR CRITICALITY SAFETY IN THE RADIOISOTOPE PRODUCTION FACILITY The RPF design will provide adequate protection against critica lity hazards related to the storage, handling, a nd processing of SNM outside a reactor. This is accomp li shed by: * *

Including equipment, facilities, and procedures to protect h ea lth and minimize danger to life or property Ens urin g that the design provides for critica li ty control, including adherence to the double-contingency principle Incorporating a criticality monitoring and a l arm system into the facility design For the Construction Permit Application , the design has assumed that a nuclea r crit i cality accident is a hi gh-consequence event independent of w h ether sh i e ldin g or other i so l ation is avai l ab l e between the source of radiation and facility personnel.

While not considered lik e l y at this time, justificatio n for considering criticality events as other than a hi gh-consequence event will be provided in t he Operating License App li cation , if this assumption i s changed for specific locations by future design act i vities. The nuclear criticality safety program defines the programmatic e l ements that work in concert to maintain criticality contro l s throug h out the operating l ife of the RPF. The nuclear critica lity safety program and facility design are developed based on the following American Nat i ona l Standards In stitu t e/Amer i can Nuclear Society (ANSI/A N S) s tandards, wit h exceptions described in U.S. Nuc l ear Regulatory Commiss i on (NRC) Regulatory Guide 3. 71 , Nucl e ar Criti c ali ty Saf e ty Standards for Fu e l s and Mat e rial Fa c iliti es. * * * * * * * *

ANSI/ ANS-8 .1 , Nucl e ar Criticality Saf e ty in Op e rations with Fis s ionable Materials Outsid e R e actor s ANSl/ANS-8

.3 , Criti ca li ty A c cid e nt Alarm Syst e m ANSI/ ANS-8. 7 , Nucl ea r Criticality Saf e ty in the Storag e of Fissil e Mat e ria l s ANSI/ ANS-8.10 , Crit e ria for Nuclear Criticality Saf ety Controls in Op e rations With Shie ld ing and Confin e m e nt ANSl/ANS-8

.19 , Admini s trative Pra c ti ce s for Nucl e ar Criti c ali ty Saf e ty ANSl/ANS-8

.20 , Nucl e ar Criticality Saf ety Training A N SI/ ANS-8.22 , Nucl e ar Criticality Safety Bas e d on Limiting and Contro llin g Moderators ANSl/ANS-8.23, Nucl e ar Criticality A c cid e nt Em e rg e ncy P l anning and R e spon se ANSl/ANS-8.24 , Validation of Ne utr o n Transport M et hods for N ucl e ar Criticali ty Saf e ty Ca l cu l ation s ANSl/ANS-8

.26 , Criti c ality Saf e ty Engineer Training and Qualifi c ation Program For the Co n s tru ction Permit Application , no deviations from standards or requirements have b een identified that would require development of equiva l ent requirements for the RPF. NWM I commits to the following standa rd s and guides during design and construction

ANSI/ ANS-8.1 -N u clear criticality safety practices , including administra ti ve practices , technical practices, and va lid ation of a calculational method 6-28

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NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features * * * * * *

ANSVANS-8.3

-Criticality accident alarm system (CAAS) placement ana l ysis and procedure development

the standard is used as modified by NRC Regulatory Guide 3.71 ANSVANS-8.19

-NWMI nuclear criticality safety program development as it applies to organization, administration, ro l es , a nd responsibilities ANSVANS-8.20

-Nuclear criticality safety staff and contractor qualification and training procedure development ANSVANS-8.24 -Validation of a calculationa l method NUREG-I 520 , Standard R e view Plan for the R e vi e w of a Licens e Appli c ation for a Fue l C y cle Facility -Guidance for meeting 10 CFR 70.61 NUREG/CR-4604, Stati s ti c al Method s for Nuclear Mat e rial Managem e nt-Guidance for normality testing of the data from critical experiment calculation s NUREG/CR-6698, Guid e for Validation of Nuclear Criticality Safety Calcu l ational M e thodologyGuidance for validation of a calculational method The nuclear criticality safety program includes the following eleme nts: Responsibilities Critica li ty safety eva lu ations Criticality safety control implementation Nuc lear criticality safety training Crit icality safety assessments Critica li ty prevention specifications Operating procedures and maintenance work Crit icality safety postings Fiss il e material container labeling , storage , and transport Crit icali ty safety nonconformance respon s e Critica lity safety configuration control Critica li ty detector and alarm system Critica li ty s afety guidelines for firefighting Emergency preparedness plan and procedures Components of the nuclear criticality safety program specifically implemented during the design and construction phases of the RPF will include: Nuc lear criticality safety program policy Nuc lear criticality safety program procedure Nuclear criticality safety evaluation procedure Nuc lear criticality safety technica l/peer review procedure Nuclear criticality safety engineer training and qualification procedure Nuclear criticality safety validation procedure Preliminary descriptions of the nuclear cr iti ca lit y safety program elements developed for the Const ruction Permit App lic ation are summarized below. Modifications to the nuclear criticality safety program e l ements are anticipated as the design matures and wi ll b e included in the Operating License Application. Re s ponsibilities Th i s element describes the responsibilities of management and staff in implementing the nuclear criticality safety program. 6-29

.. NWM I ...... .......... ' NOmfWHT MEDM:AllSOTOPU NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features * * * * *

General facility management will ensure that the nucl ear safety function is as independent as practical from the facility operating functions. A Nuclear Criticality Safety Manager will be assigned and responsible for overall coordination , maintenance , and management of the nuclear criticality safety program. A Criticality Safety Representative will be assigned who i s qualified to interpret critica lity safety requirements and serve as a liaison between custodians of fissionable material and ot h er operations , advis in g operating personnel and supervisors on questions concerning conformance to crit i ca li ty safety requirements. Qualified Critica lity Safety Engineers will responsible for performing c ritic ality ana l yses and eva lu ations of systems , maintaining current verified and va lid ated critica lity computer codes, adv i s in g staff on technical aspects of criticality contro l s , and supporting/participa t ing in inspections and management assessments.

Operations management will be r esponsib l e for estab li s hin g the responsibility for crit icality safety throughout the operations organization , commun i cat in g criticality safety responsibilities for each individual involved in operations, ensuring that co ntrol s identified by CSEs are implemented, ens urin g eac h worker h as necessary training and qualifications , and e n suri n g that procedures that include contro ls s igni fic a nt to criticality safety are prepared before operations commence. Supe r vi s ors and workers wi ll be responsible for completing training before performing fissile material operations , understanding and ens urin g compliance with all app lic a bl e critica lity safety controls , and reporting any proposed change in fissile material operations to the Critica lity Safety Representative for evaluation and approva l before the operation commences.

Criticality Safety Evaluations This eleme nt describes the proce s s for preparing CSEs that demonstrate fissile material operat ion will be subcrit i ca l und er both normal and cre dibl e abnormal conditions. * * * *

CSEs will determine, identify , a nd document the controlled parameters and associated limits on which criticality safety depends. CSEs will be required to eva lu ate normal operations, and co ntin gent and up set conditions . Preliminary CSEs prepared for the Construction Permit Application , including verification and va lid ation of s upportin g co mput er codes , are described in Section 6.3.1.1 and provide example s of the CS Es. Design changes impacting criticality will be reviewed by the Criticality Safety Representative . CSEs will be independently reviewed to confirm the technical adequacy of the evaluation prior to commencing new or modified fissile material operations. Nuclear criticality safety limit s estab li s h ed for co ntrolled parameters in the NWMI facility processes will ensure that all nuclear processes are su bcritic al , including an adequate margin of s ubcriti cality for safety in accorda nce with the Interim Staff Guidance a ugmentin g NUREG-15 37 , Guid e lines for Preparing and Reviewing Appli c ations for th e Licensing of Non-Power Reactors:

Standard Review Plan and Acceptan ce Criteria , Part 2 , Sectio n 6.b.3 (NRC, 2012). Monte-Carlo N Particle (MCNP) calculatio n results u se d to set limits on parameters are compared to the upper s ubcritic al limjt (USL) establis h ed in the NWMI MCNP code validation report ([Proprietary Information]), after applying a 2cr calculation uncertainty. 6-30 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features The USL includes the method bias and uncertainty estab li shed in [Proprietary Information]

and a 0.05 margin of subcriticality.

In addition, the area of applicability , also established in [Proprietary Information], is checked to ensure that the NWMI RPF process model physics and materials are within the bands of applicability.

If either the physics or materials are outside the bands of applicability, an additional margin of subcriticality will be applied. Criticality Safety Control Implementation Thjs element describes the process for implementing cr i ticality safety controls defined by the CSEs. * *

Implementation includes confirmjng that: All required engineered criticality safety controls are maintained by a configuration management system. Equipment dimensions, volumes, or other features relied on for controls are with limits documented in the CSEs. Adrrunistrative criticality safety controls from CSEs are implemented in written operating and maintenance procedures.

Fissile material inventories will be monitored and incorporated into implementation of criticality safety controls. Access to fissionable material will be controlled . Nuclear Criticality Safety Training This element describes the training program for nuclear criticality safety based on the worker's duties and responsibilities. * *

This training program is developed and implemented with input from the nuclear criticality safety staff, training staff , and management, with a focus on: Knowledge of the physics associated with nuclear critica li ty safety Analysis of jobs and tasks to determine the knowledge a worker must have to perform tasks efficiently Design and development of learning objectives based on the analysis of jobs and tasks that reflect the knowledge , skjlls, and abilities needed by the worker Implementation of revised or temporary operating procedures Testing methods to demonstrate competence in trainjng materials dependent on an individual's responsibility Trailing records maintenance General training on criticality hazards and alarm responses wi ll be provided to all RPF personnel and visitors.

Operators responsib l e for some aspect of nuclear criticality safety will: Satisfy defined rrunimum initial qualifications Complete an initial criticality safety training course designed for operators Perform periodic requalification training Management , operations supervisor, and technical staff responsible for some aspect of nuclear criticality safety will: Satisfy defined rrunimum initial qualifications Complete an initial criticality safety training course designed for managers and engineers 6-31

Perform periodic requalification training The Criticality Safety Representative will: Satisfy defined minimum initial qualifications NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Complete an initial critica li ty safety program de signe d for the Criticality Safety Representative Demon stra te competence in understanding facility nuclear criticality controls and procedure s Perform periodic requalification training Criticality Safety Engineers will be trained and qualified in accordance with ANSI/ANS-8.26 . Nuclear criticality safety staff members and contract s upport will meet the qualification and training requir e ments specified in the NWMI nuclear criticality safety qualification an d training program. The NWMI nuclear criticality safety qualification and training program is compliant with ANSI/ANS 8.26. Criticality Safety Assessments Thi s element describ es the periodic criticality safety in s p ec tion s and assessments conducted to ensure that the criticality safety program i s maintained at an adequate le ve l for the RPF. * * * *

Annual criticality safety inspections will be conducted to sa tisfy the requirement of ANSI/ ANS-8.1 and 8.19 for operational reviews to be conducted at lea st annually.

Procedur es wi ll be de ve lop e d for performin g p e riodic critica lity safety inspections. The facility Criticality Safety Repre se ntati ve and inspection team wi ll comprise individuals (typ ically from Engineering) who are know l edgeable of criticality safety, and who, to the extent practicab l e , are not immediately responsible for the operation being inspected. Facility inspections are conducted to verify that the facility configuration and act ivities compl y with the nucle a r criticality safety program. Facility inspections genera lly consist of observation of task preparation and verificat ion of field procedure s and trainin g. Management assessments will be conducted of the nuclear criticality safety pro gram. These assessments will be led by the Nuclear Critica li ty Safety Manager , with assistance from other member s of the criticality safety staff. The criticality safety staff is independent of the operating organization and not dir ect ly responsible for the operations.

Records generated durin g performance of criticality safety inspe ctio ns and assessments will b e included in a criticality safe ty inspection report or specia l ty assessment report. An audit to asse ss the overall effectiveness of the nuclear criticality safety program will be performed at lea st once every three years. The audit will be led b y a qualified senior criticality safety e ngineer from outside the NWMI organization.

The se nior nucl ear criticality safety engineer conducting the audit will b e independent of the NWMI progr a m and will not have participated in any nuclear criticality safety evaluation that will be a subject of the audit. In addition to the triennial audit from an outside org a nization , NWMI senior management will perform periodic audits of the NWMI nuclear criticality safety program. The se nior manager will be chosen from an NWMI organizat i on other than the nuclear criticality safety group. The NWMI Quality Assurance Manager will select and assign auditors who are independent of the NWMI nucl ea r critica li ty safety program. Criticality Prevention Specifications Thi s element de scri b es the requir e ments for the criticality prevention s pecifications u se d t o implement limits and controls esta blished in the CSEs for safe handling of fissionable ma ter ial and implement the ANSI/ANS-8 series requirement for clear communication of criticality safety l imit s and controls.

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NOflTlfWHT MEOK:Al tSOTOPU NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features

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

Procedures will meet the intent of ANSI/ ANS-8.19.

Procedures for operations and maintenance work will be prepared according to approved procedure control programs, developed and maintained to reflect changes in operations, and written so that no single inadvertent failure to follow a procedure can cause a criticality accident.

Operating procedures will include: Controls and limits significant to nuclear criticality safety of the operation Periodic revisions, as necessary Periodic review of active procedures by supervisors Operating procedures will be supplemented by criticality safety postings on equipment or incorporated in operating checklists.

Maintenance work procedures associated with SSCs affecting nuclear criticality safety will be reviewed by the Criticality Safety Representative or a Criticality Safety Engineer for compliance with nuclear criticality safety limits based on current RPF conditions present prior to initiating each maintenance evolution. Criticality Safety Postings

Criticality safety postings will be developed for the Operating License Application . Fissile Material Container Labeling, Storage, and Transport

Fissile material container labeling , storage , and transport will be developed for the Operating License Application.

Criticality Safety Nonconformance Response This element describes the response to deviations from defined nuclear criticality safety controls.

Deviations from procedures and unforeseen alterations in process conditions that affect criticality safety will be immediately reported to management and the Criticality Safety Representative or a Criticality Safety Engineer.

NWMI management will provide the required notifications of the deviation to the U.S. Nuclear Regulatory Commission Operations Center. 6-33 NWM I ...**... NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features * * *

NOllTHWHT MEDtCAl ISOTOPU * *

The Criticality Safety Representative or a Criticality Safety Engineer will support an investigative team comprising, at a minimum, the Operations Manager and operations personnel familiar with the operation in question during the development of a recovery plan fo r safely returning to compliance with the procedures. The deviation will be corrected per the recovery plan and the incident documented . Action is to be taken to ensure that a similar situation does not exist in another part of the facility and to prevent recurrence of the non conformance. Criticality Safety Configuration Control This element describes the criticality safety configuration controls.

* * * *

The primary criticality safety control , performed at the start of a proposed activity or equipment change , is for the Criticality Safety Representative to confirm if an existing active CSE is applicable.

All dimensions , nuclear properties , and other features on which reliance is placed will be documented and verified prior to beginning operations, and control will be exerci s ed to maintain them. The nuclear criticality safety staff will provide technical guidance for the design of equipment and proce ss es and for the development of operating procedures. All proposed criticality safety-related changes to design or process configuration will be reviewed by a Criticality Safety Representative or Criticality Safety Engineer to ensure that the change can be performed under an approved CSE. All operational changes that impact criticality safety will be documented and include proper approval designation.

The project manager will request a CSE applicability review at the earliest practical stage of a project to determine ifthere could be criticality safety impacts. If the potential exists for the physical configuration or operating parameters for new or revised equipment to affect criticality safety , the drawings and process control plans will be reviewed and approved by a Criticality Safety Representative or Criticality Safety Engineer , in compliance wi t h standard engineering practices and procedures. Facility and process change control will include the following . The change management process will be in accordance with ANSI/ ANS-8.19. All dimensions , nuclear properties , and other features on which reliance is placed will be documented and verified prior to beginning operations , and control will be exercised to maintain them. Changes that involve or could affect nuclear criticality controls will be evaluated under 10 CFR 50.59 , "Changes , Tests , and Experiments

." Changes include new designs, operation , or modification to exi s ting SSCs , computer programs , processes , operating procedures , or management measures.

Changes that involve or could affect nuclear criticality controls will be reviewed and approved by the Criticality Safety Representative.

Prior to implementing the change , the process will be determined to be subcritical (with an approved margin for safety) under both normal and credible accident scenarios. 6-34 NWMI *:****-:-..*..*.. '

NCMITHWHT MEDtCAl ISOTOl'lS NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features T es ting and Calibration of Active Engineered Controls

Testing and ca l i bration of AECs will be developed for the Operating License Applica tion . Criticalit y Safet y Guidelines for Firefighting

Cri ticali ty safety g uid e lin es for firefighting will b e d eve lop e d for t h e Operating License Application. Emergency Preparedness Plan and Procedures Thj s ele m ent d escribes t h e response to c ritic a l ity acc id e nt s. * * * * * * * * * * * *

The CAAS wi ll b e used as d esc r i b e d in Sect ion 6.3.1.1 and pro v id es for detection a nd a nnunci ation of criticality acc id e nt s. E m e r gency procedures will be pr epa r ed an d a ppro ved by management.

Fac ility and off-site or ganizat ion s expecte d to respond to e m e r ge n cies wi ll be in forme d of co nditions t h at mi g ht b e enco unt e r e d. Procedures wi ll: D esignate evac uati on rou t es that are clea rl y id e nt ifie d a nd fo ll ow the qui c k est , mos t dir ect rout es practica l Includ e assess ment of expos ure to individuals De signate p ers onn e l asse mbly stat i o n s outside th e areas to b e evac u a t e d. A method to acco unt for personnel wi ll be es tabli s h e d a nd arrangements mad e in a dvance for the ca r e and tre a tm e nt of injur ed a nd ex po sed personnel.

The po ssibility of per sonne l co nt a mina tion by radioactive m aterial will be co n s id ere d . Personnel wi ll b e train ed in eva lu at ion methods , inform e d of routes and assembly stat ions , an d drills p erformed at l east ann u a lly. Instrumentation a nd pro cedures will be provided for determining radiat i o n in a n evac u ated area fo llo wi ng a critica lity acciden t a nd inform atio n co ll ected in a central location.

E m e r gency procedures will be maintained for eac h area in which specia l nuclear material is handled , used , or s tored to ens ure that all personnel withdraw to an area of safety on so undin g th e a l arm. E mer gency procedures wil l include conducti n g drill s to fami liari ze personnel with the evacuation plan , designation of re sponsib l e individuals to d e t ermine t h e ca u se of the a l arm, and placement of ra diation survey in st rum ents in accessib l e locations for use in suc h an emerge nc y. The current emerge nc y procedures for eac h area will be retained as a record for as l ong as licensed specia l nucl ear material i s handl ed, u se d , or store d in th e area. Superseded sectio ns of emergency pro cedu r es will b e retained for three years after the sectio n is s up e r seded. Fixed and personnel accident dosimeter s will b e pro vide d in areas that require a CAAS . Dosimeters wi ll be read i l y ava ilabl e to personnel responding to a n emerge n cy and a method provided for prompt on-site do s im eter rea d o ut s. 6-35 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features 6.3.1 Criticality Safety Controls The following sections describe critica lity safety controls based on information developed for the Construction Permit Application.

Section 6.3 .1.1 summarizes the results of preliminary CS Es that defi ne PECs and AECs cred ited to satisfy the doubl e-cont in gency contro l principle.

Section 6.3.1.2 summarizes IROFS related to preventing a nuclear critica lity identified by the accide nt ana l yses described in Chapter 13.0. 6.3.1.l Preliminary Criticality Safety Evaluations A series of calculations were performed to s upp ort the Construction Permit Appl i cation investigating parameters associated wit h prevention of nuclear criticality in the current equipme nt configuration of major process systems. The calculations are described in the following documents:

NWMI-2015-C RIT CALC-OO 1 , Single Parameter Subcritica l Limits for 20 wt% 235 U -Uranium Metal, Uranium Oxide, and Homogenous Water Mixtures * *

NWMI-2015-CRITCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution NWMI-2015-CRITCALC-003, 55-Gallon Drum Arrays NWMI-201 5-CRITCALC-005, Target Fabrication Tanks, Wet Processes, a nd Storage NWMI-2015-CRITCALC-006 , Tank Hot Cell Calculations were performed u sing the MCNP 6.1 code (LA-CP-13-00634, MCNP6 Us e r Manual). Validation of the MCNP 6.1 code used in the calculations is described in [Proprietary Information].

The va lid atio n report documents the methodology a nd r esu lt s for the bi as and bias uncert ainty va lue s calcu l ated for homogeneous and heterogeneous uranium systems for the MCNP 6.1 code syste m. The bias is exp re ssed as USLs calculated using a facility-specific

[Proprietary Information].

The primary focus of the validation was to determine the bias an d bi as uncertainty for intermediate-enr i c hed uranium (IEU) syste m s. However , s uffi cient exper im ents fo r low-enriched uranium (LEU) and high-enriched uranium were included to demonstrate that there i s no variation in the USL wi th varying enric hment. Similarly, the primary focus of the va lidation was on thermal neutron energy syste m s. Su ffi cient experiments for intermediate and fast energy experime nt s were a l so included to demonstrate that there is no variatio n in the USL w ith increasing neutron energy. The purpose of the comp ut er code va lid ation is to determine va lu es of k eff that are d emonstrated to be subcritica l (at or below the USL) for areas of appl icabilit y simi l a r to systems o r operations being analyzed.

The USL is defined by Eq u at ion 6-1. USL = 1.0 -Bias -Bias Uncertainty

-Margin of Subcriticality Eq u ation 6-1 [Propr ietary Information]

rearranges Eq u ation 6-1 to produce a criterion for model cases t ha t are considered acceptable as s ubcriti cal, as s ho wn by Eq u ation 6-2, and incorporates the margin of su bcritic a lity in t h e USL as required by ANSVANS-8.1.

k eff + (2 X a caiJ USL Eq u at i on 6-2 where keff is the MCNP calc ulat ed k-effective and O" c alc is the MCNP calculation uncertainty.

[Proprietary Information]

6-36

NWMI ...... ' *.* ." NOWTtfWfST MEOK:Al ISOTOPl S NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

indicates the validation is appropriate for homogeneous and heterogeneous IEU sy s tems. A summary of the area of applicability is provided in Table 6-4. For systems outside the validation area of applicability, an increased margin of subcriticality value may be warranted , depending on the specific problem being analyzed.

The analyst must document any extrapolation beyond the validation area of applicability , and justification must be documented for no adjustments to the margin of subcriticality when extrapolating. Table 6-4. Area of Applicability Summary Parameter Fissile material Fissile material form H/2 3 5 U ratio Average neutron energy causing fission Enrichment Moderating materials Reflecting materials Absorber materials Geometry

Source: [Proprietary Information]. ANE C F = avera ge neutron energy causing fi s sion. Area of Applicability

[Proprietary Information]

The RPF was divided into 13 activity groups for development of preliminary CSEs of the activities and associated equipment.

Controlled nuclear criticality safety parameters vary with the activity group and are summarized in Table 6-5. A minimum of two nuclear criticality safety parameters are controlled to satisfy the double-contingency principle.

6-37

.. ;.;; .. NWMI ..... .*.******** NOflHWHTMEDtCAl.ISOTOPfS NWM l-2 0 1 3-021 , Re v. 3 C h apte r 6.0 -E ng i ne e red Saf e ty F e atu r e s Tab le 6-5. Co n t roll e d N ucl ea r C rit ic alit y Sa f ety Pa r a m e t e r s NWMI criticality safety evaluation (NWMl-2015-CSE*)

' Nuclear parameter

Mass y y y y y y y N y y y b y y Geometry y y y y y ye ye y N y y y y Moderation y N N N N N N N N N N N N Interact i o n y y y y y y y y N y y y y Volume y y y y y y y N N N y N y Concentration/

N yd yd yd yd N N N ye ye ye N N density Reflection N N N N N N N N N N N N N Abso r bers N N N N N N N N N N N N N Enrichmentr N N N N N N N N N N N N N

Derived from the indicated CSE reference document.

b Limited by n a ture of proces s in the air fi l tration. c Limited by tar g et de s ign. d Contro ll ed through i n put fissile mass. 0 Limited by total urani u m ma s s allowed i n the s y stem. r Fac il ity licen se limited to ::;20 wt% m u. m u uranium-235. NWMI Northw es t Medica l Isotop es, LLC. CSE = criticality safety evalu a tion. y yes. N = no. The preliminary CSEs define a series of PECs , AECs, a n d administrative controls that are credited to satisfy the d ouble-contingency control principle for prevention of n uclear critica l ity events such that at least two c h anges in process conditions must occur before criticality is possib l e. PECs , AECs, a n d administrative contro l s are described for the 13 activ i ty groups in the fo ll owing referenced tab l es: * * * * * * * * * * *

NWMI-2015-CSE-01 , Irradiat e d Targ e t Handling and Disassembly (Tab l e 6-6) NWMI-20 l 5-CSE-02 , Irradiated Low-Enriched Uranium Targ e t Mat e ria l Disso lut ion (Ta b le 6-7) NWMI-20 l 5-CSE-03, Molybdenum-99 Recovery (Tab l e 6-8) NWMI-2015-CSE-04 , Low-Enriched Uranium Targ e t Mat e rial Production (Table 6-9) NWMI-2015-CSE-05, Target Fabrication Uranium Solution Pro ce sses (Table 6-9) NWMI-20 l 5-CSE-06 , Target Finishing (Tab l e 6-9) NWMI-2015-CSE

-07 , Target and Can Storage and Carts (Table 6-9) NWMI-2015-CSE-08 , Hot C e ll Uranium Purification (Table 6-10) NWMI-2015-CSE-09 , Waste Liquid Processing (Table 6-11) NWMI-2015-CSE-10 , Solid Wast e Coll e ction , Encapsulation , and Staging (Table 6-1 1) N WMI-2015-CSE-1 l, Offgas and Ventilation (Table 6-12) NWMI-20 l 5-CSE-12 , Targ e t Transport Cask or Drum Handling -The shipping packages dictate design features t h at m ust be properly imp l emented for lega l over-the-road tra n sport. T h is CSE does not impose or credit add i tional passive controls other t h an those a l ready incorporated in the respective shipping packages. 6-38 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features

NWMI-2015-CSE-13, Analytical Laboratory (Table 6-13) The CSEs wi ll be updated for final design and the Operating License App li cation. Cri ti cality controls are se l ected based on the following order of preference

Passive engineered control s Active engineered controls Enhanced administrative controls Admini s trative control s Note that a number of features li s ted in the preliminary CSEs are duplicated in multiple activity groups (e.g., the floor of cells is verified to be flat , with no collection points deeper than 3.5 centimeters

[cm]). Duplications are in clude d in the curre nt li st in gs to c l early identify minor dimension variat i ons that may exist in the d efined featu r es for different activity groups. Table 6-6. [Proprietary Information)

Double-Contingenc y Controls Identifier" CSE-01-PD Fl [P roprietary Information

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CSE-01-AC 1 [Proprietary Inform ation] CSE-0 1-AC2 [P roprietary Information]

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CSE-01-AC4 [Proprietary Inform a tion] * [Proprietary In fo rm a ti o n]. H E P A = hi g h-e ffici e n cy p a rticul a te a ir. Feature description and basis SPL s in g l e p ara m e t e r limit. 6-39

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...... .......... NORTHWESTMfDICAUSOTOf'lS NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-7. [Proprietary Information)

Double-Contingency Controls (2 pages) ldentifier a Feature description and basis CSE-02-PDFl

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CSE-02-PDF2

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NWMl-2013-0 21, Rev. 3 Chapter 6.0 -Engineered S a fety Features T able 6-8. [Propri e t ary Information]

Doubl e-Co ntin ge nc y Co ntrol s (2 p ages) Identifier

CSE-03-PDF l [Proprietary Information]

CSE-03-PD F2 [Proprietary Information

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CSE-03-P D F6 [Proprietary Informatio n] CSE-03-PDF7

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CSE-03-P D F8 [Proprietary I n formation] CSE-03-PDF9

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CSE-03-AEF 1 [Proprietary I nformatio n] CSE-03-AC l [Proprietary Informatio n] IX Mo * [Proprietary Inform a tion]. ion exchange. = molybd e num. Feature description and basis [Propri etary Information]

[Pro pri e tary In fo rmati o n]. 6-4 1

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

NORTKWHT MEDICAL ISOTOPU NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [P rop r ietary I n formation] 6-42 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-9. [Proprietary Information]

Double-Contingency Controls (8 pages) Identifier Feature description and basis CSE-04-PDFP

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PDF IO* CSE-04-[Proprietary Information]

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6-43

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

NORTHWEST MEOICAl ISOTOPES NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-9. [Proprietary Information]

Double-Contingency Controls (8 pages) Identifier CSE-05-AEFl b [Proprietary Information]

CSE-05-AEF2b

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CSE-07-AC7d

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a [Proprietary Information]

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ADUN DBE u acid-deficient uranium nitrate. de s ign basis earthquake. uranium. Feature description and basis UN = uranium nitride. [Proprietary Information]

[Proprietary Information]

6-44 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Propri etary In formation]

6-45 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-46 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [Propri etary Inform atio n] 6-47 NWM I ...... * *

NOflllfW(Sf MEOtCAL ISOTOPH NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [P roprietary I nformation]

6-48 NWM I ...... ' *

NomtWEST MEDM:Al. &SOTOf'U NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Proprietary In formation]

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

S NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [Prop rietary In fo rm atio n] 6-50 NWM l-2 01 3-021 , Rev. 3 C hapte r 6.0 -Eng i neered Sa f ety Featu r e s Ta bl e 6-10. [Propr ie ta ry Inform at i o n] D o ubl e-Co ntin ge n cy Co ntr o ls (2 p ages) Identifier*

' Feature description and basis CSE-0 8-PDF l [Proprietary Information]

CSE-08-P D F2 [Proprietary Information

] CS E-0 8-PDF3 [Proprietary Inform a tion] CSE-08-P D F4 [Proprietary Informatio n] CSE-0 8-PDF5 [Proprietary Information]

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CS E-0 8-PDF9 [Proprietary Inform a tion] CSE-08-[Pro p rietary I n formation] PDF I O C S E-0 8-[P ro pri e tary Inform a ti o n] PD F l l CSE-08-[Proprietary I nformatio n] PDF12 CS E-08-AE F l [P ro pri e tary Inform a tion] CSE-08-AC l [Proprietary Informatio n] C S E-0 8-A C2 [P ro pri e tary Informati o n] * [P ro pri e t ary In forma tion] DB E = d es i gn b as i s ea rt h q u a k e. IX i o n ex ch a n ge. 6-51

.. ; .. ;. NWMI ...... ..* *.. .......... ' *.* NCMITHWHT MU>tCAt. ISOTOPll NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [Propri etary Inform atio n] 6-52 NWM I ..*... * * . NOtmfWEST M£DtCAl lSOTDPU NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-11. [Proprietary Information]

Double-Contingency Controls (3 pages) Identifier Feature description and basis CSE-09-[Proprietary Information]

AEF I' CSE-09-ACl" [Proprietary Information]

C S E-09-AC2 a [Propri e tary Inform a ti o n] CSE-09-AC3 a [Proprietary Information]

CSE-10-PDFl b [Propri e t ary Inform a tion] CSE-I 0-[Proprietary Information]

AEFib CSE-I 0-AC I b [Propriet a ry Inform a tion] CSE-I O-AC2b [Proprietary Information]

CS E-I O-AC3 b [Propriet a ry Inform a tion] CSE-10-AC4b

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C SE-I O-AC5 b [Propri e t a ry Inform atio n] CSE-10-AC6b

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CS E-I O-AC 7b [Proprieta ry Inform a tion] CSE-I O-AC8b [Proprietary Information]

CS E-I O-AC9 b [Proprietary Inform a tion] 2 3 5 LJ HI C RPF * [Proprietary In fo rmation] b [Pro pri e ta ry In fo rm a ti o n] u ra nium-235. hi g h-int egr i ty c ont a in e r. R a di o i so t o p e Producti o n Fac ility. S PL u 6-53 s ingl e p arameter limit. uranium.

NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [Propri e tary Information]

6-54 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features [Propri etary Inform ation] 6-55 NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-12. [Proprietary Information]

Double-Contingency Contro l s (2 pages) Identifier*

CSE-I I-PDF I [Proprietary Inform atio n] CSE-I l-PDF2 [Proprietary Information]

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CSE-I l-PDF6 [Proprietary Inform at ion] CSE-l l-PDF7 [Propri etary Inform at ion] CSE-l l-PDF8 [Proprietary Information]

CSE-11-AEF l [Proprietary Inform a tion] CSE-11-ACl

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[Proprietary In fo rm a tion] DB E H E P A de s i g n basis ea rthquak e. = hi g h-effic i e n cy p a rti c ul ate a ir. Feature description and basis M o NO , 6-5 6 mol y bd e num. nitrogen oxi d e.

NWM I ...... ' * *

NOITHWHT MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features Table 6-13. [Proprietary Information]

Double-Contingency Contro l s (2 pages) Identifier" CSE-1 3-PDF I [Proprietary Information]

CSE-13-PDF2

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CSE-13-AC6 [Proprietary Inform atio n] * [Proprietary In formatio n] R&D RPF r esearch and d eve l opment. = R ad i oi s otope Production Fac ili ty. Feature description and basis SPL u 6-57 s ingle p aramete r limit. uranium.

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

NOITHWUT M£DtCAl ISOTOH.S NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features [P ro p r i etary In forma t io n] 6-58 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features Each of the prelimin ary CSEs indicates that the proce ss areas evaluated will be within the detector and alarm coverage of the CAAS. Evaluation of the CAAS coverage will be performed after final design is complete and prior to facility startup. To ensure the CAAS coverage is adequate for the facility, NWMI will conduct a coverage analysis using the minimum accident of concern that produces a detector re s ponse when the dose rate at the detector is equivalent to 20 rad/min at 2 m (6.6 ft) from the reacting m a terial. Using the so urce from the minimum accident of concern, NWMI will conduct one-dimensional de te rministic computations , when practical , to evaluate CAAS coverage.

For areas of the facility where the u se of one-dimensional determini st ic computations i s not practical, NWMI will use 3D Monte Carlo analysis to determin e adequate CAAS coverage.

The CAAS will be designed to meet the following. * *

6.3.1.2 The facility CAAS: Will be capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 radiation dose absorbed (rad) of combined neutron and gamma radiation at an unshielded distance of 2 m from the reacting material within 1 minute; two detectors will cover each area needing CAAS coverage Will use ga mma and neutron sensitive radiation detectors that energize clearly audible alarm signals if a n accidental criticality occurs Will comply with ANSVANS-8.3 , as modified by NRC Regulatory Guide 3.71 Will be ap propriate for the type of radiation detected , the intervening shielding , and the magnitude of the minimum accident of concern Will be designed to remain operational during design basis accidents Will be clearly audible in areas that must be evacuated or there will be alternative notification method s that are documented to be effective in notifying per sonne l that evaluation is necessary Operations will be rendered safe, by shutdown and quarantine , if necessary , in any area where CAAS coverage has been lost and not re store d within a s pecified number of hour s. The number of hours will be determined on a process-by-process basis , because shutting down certain processes , even to make them safe, may carry a larger risk than being without a CAAS for a s hort time. Compensatory mea s ures (e.g., limiting access, halting SNM movement , or restoring CAAS coverage with an alternate instrument) when the CAAS is not functional will be determined for inclusion in the Operating License Application.

Emergency power will be provided to the CAAS by the uninterruptable power s upply system . Derived Nuclear Criticality Safety Items Relied on for Safety The following subsections describe engineered safety features that are derived from the accident scenarios that could result in a nuclear criticality, as described in Chapter 13.0. 6.3.1.2.1 IROFS CS-04, Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement IROFS CS-04, " Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement ," is identified by the accident analyses in Chapter 13.0. During handling of uranium solids and solutions outside of processing systems under normal conditions , the material will be handled in safe ma sses controlled by either physical measurement or batch limit s on well characterized de vices. 6-59 NWMl-2013-0 21, R ev. 3 Chapter 6.0 -Engineered Safety Features Solid uranium will be handled outside of processing systems during: * * * *

Receipt and processing of fresh uranium (and presumably shipment of spent uranium back to the supplier)

[Proprietary Information]

Fabrication of targets using [Proprietary Information]

LEU target material (including movement of LEU target material to and from the fabrication workstation and handling of the completed targets) Disassembly of targets following irradiation Laboratory sampling and analysis activities (a l beit in smaller quantities) . Each activity is assigned a mass or batch l imit for safe handling. Accident Mi t igated The accident occurs when a safe mass or batch limit is exceeded beyond some bounding extent based on the management measures on the contro l. Note that this accident involves normal condition criticality controlled limits for safe handling , and the upset represents failure of an associated administrative control. The most limiting activity would involve processing the LEU target material from [Proprietary Information].

If the IROFS fails , accidental nuclear criticality is possible without additional control. Syste m Co mp onents As a PEC, fixed interaction control fixtures or workstations will be provided for ho l ding or processing approved containers containing approved quantities of uranium metal, [Proprietary Information], batches of targets , and batches of samples. F u nct i on al R equireme n ts The fixtures are designed to hold only the approved container or batch and are fixed with 2-ft edge-edge spacing from all other fissile material containers, workstations , or fissile solution tanks , vessels, and ion exchange (IX) columns. Where LEU target materia l is handled in open containers , the design will prevent spills from readily spreading to an adjacent workstation or storage location. Des i g n Basis Final workstation and fixture spacing will be determined in final design when all process upsets are evaluated.

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

*

[Proprietary Information]

[Proprietary Information]

Target basket fixture that provides safe spacing of a batch of targets from one another in the target receipt cell Test Req uir e m ents The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6-60

NWM I ...... NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.3.1.2.2 IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confinement Using the Diameter of Tanks, Vessels, or Piping IROFS CS-06 , "Pencil Tank , Vessel , or Piping Safe Geometry Confinement using the Diameter of Tanks , Vessels , or Piping,

is identified by the accident analyses in Chapter 13.0. The PHA in Chapter 13.0 identified a number of individual potential initiating events that could lead to a spill of fissile solution from the geometrically safe confinement tanks, vessels, or piping that provide the primary safety functions of the processes.

Four processing systems will handle fissile solutions: Target fabrication (from the [Proprietary Information])

Target dissolution system First stage of molybdenum recovery and purification Entire uranium recovery and recycle system Three of these systems will be at least partially located within the hot cell wall boundary due to the dose of the fission products.

Initiating events include the general categories of tank, vessel , or piping failure due to operator error (valves out of position), valves leaking , equipment leaking (pumps , piping , vessels , etc.), high pressure events from various causes including high temperature solutions (locked in boundary valves), hydrogen detonation, and exothermic reactions with the wrong resins or reagents used in the respective systems. Some of the initiators result in small leaks that are identified and mitigated (e.g., pump seal and small valve leaks). Over the life of the facility , these types of leaks are to be expected , but do not challenge the overall safety of RPF operations.

Accident Mitigated The accident of concern involves fissile process solution in quantities necessary to sustain accidental nu c lear criticality. Larger catastrophic leaks or ruptures of equipment must occur for enough material to be released. Such leaks would represent a failure of the safe-geometry confinement IROFS for the re s pective equipment.

Thus , scenarios leading to this accident sequence involve the failure of these IROFS. Due to the nature of the process, the worst-case accident involves the tanks with the largest capacity and the highest normal case concentrations. System Components As a PEC, pencil tanks and other standalone vessels are designed and will be fabricated with a geometry diameter for safe storage and processing of fissile solutions. The safe diameters of various tanks , vessels , or components will be provided in the Operating License Application.

Functional Requirements The safety function of safe diameter vessels is also one of confinement of the contained solution.

The safe-geometry confinement of fissile solutions will prevent accidental nuclear criticality, a high consequence event. The safe-geometry confinement diameter will conservatively include the outside diameter of the tank wall or out to the outside diameter of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the vessels. Where insulation is used on the outside wall of a vessel, the insulation will be closed foam or encapsulated type (so as not to soak up solution during a leak) and will be compatible with the chemical nature of the contained solution.

Design Basis The safe-geometry diameter of tanks , vessels, and piping will be determined in final design after finalizing the reference CSEs. Note that preliminary vessel sizes for activity groups are listed in the double-contingency parameters described in Section 6.3 .1.1. 6-61 Test Requirements NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.3 IROFS CS-07, Pencil Tank Geometry Control on Fixed Interaction Spa ci ng of Individual Tanks IROFS CS-07, "Pencil Tank Geometry Control on Fixed Interaction Spacing oflndividual Tanks," is identified by the accident analyses in Chapter 13.0 (see description in Section 6.3.1.2.2). Accident Mitigated See description in Section 6.3.1.2.2. System Components As a PEC , pencil tanks and other standalone vessels (controlled with safe geometry or volume constraints) are designed and will be fabricated with a fixed interaction spacing for safe storage and processing of the fissile solutions.

Tanks, vessels , and components requiring fixed interaction control spacing of the barrels within each set of pencil tanks and between various tanks , vessels, or components will be provided in the Operating License Application. Functional Requirements The safety function of fixed interaction spacing of individual tanks in pencil tanks and between other single processing vessels or components is designed to minimize interaction of neutrons between vessels such that under normal and credible abnormal process upsets , the systems remain subcritical.

The fixed interaction control of tanks, vessels , or components containing fissile solution s will preve nt accidental nuclear criticality, a high consequence event. The fixed interaction spacing will be meas ure d from the outside of the respective tanks, vessels, or component or from the outside of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the vessels or component.

The fixed interaction control distance from the safe slab depth spill containment berm will also be specified where applicable.

Design Basis Actual interaction control parameters will be defined during final design. In addition, the following generic interaction control parameters apply during design. *

Connecting piping between fissile material components will not exceed a cross-sectional density to be determined during final evaluation of systems. Edge-to-edge spacing between fissile material-bearing vessels and components and the concrete reflector presented by the hot cell shielding walls will be fixed at a distance to be determined during final evaluation of all components. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6-62 NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 6.3.1.2.4 IROFS CS-08, Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Dikes IROFS CS-08, "Floor and Sump Geometry Control on Slab Depth , Sump Diameter or Depth for Floor Dikes," is identified by the accident analyses described in Chapter 13.0 (see description in Section 6.3.1.2.2). Accident Mitigated See description in Section 6.3.1.2.2. System Components As a PEC , the floor under designated tanks , vessels, and workstations will be constructed with a spill containment berm using a safe-geometry slab depth , and one or more collection sumps with diameters or depths , to be determined in final design. Functional Requirements The safety function of a spill containment berm is to contain spilled fissile sol ution from systems ov e rhead and prevent an accidental nuclear criticality if one of the tanks or related piping leaks , ruptures , or overflows (if so equipped with overflows to the floor). Each spill containment berm will be sized for the largest single credible leak associated with overhead systems. The sump will have a monitoring system to alert the operator that the IROFS has been used and may not be available for a follow-on event. A spill containment berm is operable if it contains reserve volume for the largest single credible spill. Spill containment berm s izes and locations will be determined during final de s ign. Design Basis The safe-geometry s lab depth under designated tanks , vessels, and workstations will be det e rmined during final design after finalizing the reference CSEs. Note that the preliminary slab depth for the activity groups are listed in the double-contingency parameters described in Section 6.3.1.1. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

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

6.3.1.2.5 IROFS CS-09, Double-Wall Piping IROFS CS-09, "Double Wall Piping ," is identified by the accident analyses described in Chapter 13.0. As a PEC, a piping system for conveying fissile solution between confinement structures will be provided with a double-wall barrier to contain any s pills that may occur from the primary piping. Accident Mitigated

Leak in piping that passe s between confinement enclosures 6-63 System Components NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features IROFS CS-09 is used at the locations listed below that pass through the facility where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes.

The following double-wall piping segments are identified for criticality safety:

Transfer piping containing fissile solutions traversing between hot cell walls

Transfer piping connecting the uranium product transfer send tank (UR-TK-720) and the uranyl nitrate storage tank (TF-TK-200)

Any other locations in final design where fissile solution piping exits a safe-slab spill containment berm and enters another Functional Requirements The safety function of this PEC is to safely contain spilled fissile solution from system piping and prevent an accidental nuclear criticality ifthe primary confinement piping leaks or ruptures.

The double-wall piping arrangement will maintain the safe-geometry diameter of the solution. The double-wall piping will also function as a barrier to prevent fissile solution from soaking into the concrete from lines passing through concrete walls where required by the criticality safety analysis (e.g., see PDF2 of Table 6-9). The secondary safety function of double-wall piping is to prevent personnel injury from exposure to acidic or caustic licensed material solutions that are conveyed in the piping. Design Basis The double-wall piping arrangement is designed to gravity-drain to a safe-geometry set of tanks or a geometry containment berm. Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.6 IROFS CS-10, Closed Safe Geometry Heating/Coo l ing Loop with Monitoring and Alarm IROFS CS-10, "Closed Safe Geometry Heating or Cooling Loop with Monito r ing and Alarm," is identified by the accident analyses in Chapter 13.0. As a PEC , a closed-loop , safe-geometry heating or cooling loop with monitoring for uranium process solution or high-dose proce s s solution will be provided to safely contain fissile process solution that leaks across the heat transfer fluid boundary if the primary boundary fails. Accidents Mitigated The dual-purpose safety function of the closed-loop system is to prevent (1) fissile process solution from causing accidental nuclear criticality , and (2) high-dose process solution from exiting the hot cell containment, confinement, or shielded boundary (or to prevent low-dose solution from exiting the facility , for systems l ocated outside of the hot cell containment, confinement, or shielded boundary), and causing excessive dose to workers and the public, and/or causing a release to the environment.

System Components The closed loop steam and cooling water loop design is described in Chapter 9.0. 6-64 Functional Requirements NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features The hea t exchanger materials wi ll be co mpatibl e wit h the har sh c h emica l environme nt of the tank or vesse l process (this may vary from app lication to app li cat ion). Sampling of the h eat in g or cooli n g media (e.g., s team co nd ensate co ndu ctivity , coo lin g water ra di o lo gical act i vity , or uraniu m co n centration) will b e cond uct e d to alert t h e operator that a br each has occ urr ed , and that a ddi tiona l corrective act ion s are required to id e nti fy and i so lat e the fa iled component a nd restore the closed l oop integrity. Discharged so lut ions from this s y s tem wi ll be handled as potentially fis s ile an d sa mpl ed prior to di scharge to a safe geometry.

Design Basis The c lo se d loop ste a m a nd cooling wate r loop design i s d escribed in C h apter 9.0. Test Requirements The above analysi s i s based on informatio n developed fo r the Construction Permit App li cation. Additional detailed information on test requirements wi ll be developed for the Operating License App l ication. 6.3.1.2.7 IROFS CS-11 , Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm IROFS CS-11 , "Simple Overflow to Normally Empty Safe Geometry Tank with Leve l Alarm," i s identified b y the accide nt a n a l yses described in Chapter 1 3.0. As a P EC , a simple overflow l i n e will b e in sta ll ed b e low the level of the process vesse l venti l at ion port and any c h emica l addition ports (where a n a n t i-sipho n safety feature will b e insta ll e d) for each ve nt e d tank conta inin g fissi l e or potentially fissile proces s so lution for whic h this IROFS is assigned. Accident Mitigated The overflow d rain wi ll prevent the process so lu tio n fro m entering the respective non-geometrically favorab l e sect ion s of the process venti l ation system a nd any chemica l addition ports (where chemica l a ddi tion ports enter thro u gh anti-s iphon device s). System Components Locations of th e overflow a nd overflow co ll ection tanks wi ll be prov id e d with the fina l design. F unction a l Requirements The s afety function of this feature is to prevent accidental nuclear criticality in non-geometrically favorable sections of the process v enti l ation system. T h e overflow wi ll be directed to a s afe-geometry storage tank. Th e overflow storage tank will normally be maintained empty. The ove rfl ow storage tank will be eq uipped with a l eve l alarm to inform th e operato r when u se of the IROFS h as b een initiated , so that actions can be taken to re store operab ili ty of the safety feature b y emptying the tank. Design Basis D es i gn basis inform ation will be provided in the Opera tin g License App li cation. 6-65 Test Requirements NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Construction Permit Application.

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

6.3.1.2.8 IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line IROFS CS-12, "Condensing Pot or Seal Pot in Ventilation Vent Line ," is identified by the accident analyses described in Chapter 13.0. As a PEC, a safe-geometry condensing pot or seal pot will be installed downstream of each tank for which this IROFS is assigned to capture and redirect liquids to a safe-geometry tank or flooring area with safe-geometry sumps. One such condensing or seal pot may service several related tanks within the safe-geometry boundary of the ventilation system. The condensing or seal pot will prevent fissile solution from flowing into the respective geometrica ll y favorable process ventilation system by directing the solution to a safe-geometry tank or flooring area with s afe-geometry sumps. Accident Mitigated Where independent seal or condensing pots are credited , the drains of the seal or condensing pots must be directed to independent locations to prevent a common clog or over-capacity condition from defeating both. System Components Locations of the condensing pots or seal pots and associated drain points will be provided with the final design. Functional Requirements The safety function of the condensing or seal pots is to prevent accidental nuclear criticality in geometrically favorable sections of the process ventilation system. The safe-geometry tank or sumps will be equipped with a level alarm to inform the operator when use of the IROFS has been in i tiated. Each individual tank or vessel operation must be evaluated for required overflow capacity to ensure that a suitable overflow volume is available. A monitoring and alarm c ircuit will be provided so that common overflow tanks or safe s l ab flooring or sumps can be used for multiple tanks or vesse l s , and limitin g conditions of operation will be defined to ensure that the IROFS is made available in a timely manner or operations are suspended following an overflow event of a single tank. Design Basis Design basis information will be provided in the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements wi ll be developed for the Operating License Application.

6-66 NWM I ...... ' NORTlfWCST MEIHCAL ISOTOflU NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features 6.3.1.2.9 IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary IROFS CS-13, "Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Contain m ent Boundary," is identified b y the acci d e nt analyses described in C hapt er 13.0. As a PEC , a s imple overflow line wi ll be installed above the high a l arm setpoint for each vented tank containing fissile or potentially fissile process so lution for whic h this IROFS is assigned.

The overflow will be directed to one or more safe-geometry flooring configurations with safe-geometry sumps. Accident Mitigated This IROFS prevents accidental critica lity by ensuring that overflowing fis s ile sol ution s are captured in a safe-geometry slab configuration with safe-geometry s ump s. System Components System component information will be provided in the Operating License Application.

Functional Requirements The floor areas (separated as needed to support operations in different hot cell areas) will normally be maintained e mpt y. The floor area(s) wi ll be equipped with a sump l evel alarm to inform the operator when use of the IROFS has been initiated. Design Basis Design basis information will be provided in the Operating Licen se Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additiona l detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.10 IROFS CS-14, Active Discharge Monitoring and Isolation IROFS CS-14, "Active Discharge Monitoring and Isolation ," is identified by the accident analyses described in Chapter 13.0. Additional detailed information describing active discharge monitoring and isolation will be developed for the Operating License App li cation. System Components System component information will be provided in the Operating License Application.

Functional Requirements Functional requirements information will be provided in the Operating License App lication. Design Basis Design basis information will be provided in the Operating License Application. 6-67 Test Requirement s NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Construction Permit Application. Additiona l detailed information on test requirements will be developed for the Operating License Application. 6.3.1.2.11 IROFS CS-15, Independent Active Discharge Monitoring and Isolation IROFS CS-1 5, "Independent Active Di sc harge Monitoring and Isolation ," is identified by the accident analyses described in C h apter 13.0. Additional detailed information describing independ e nt active discharge monitoring and isolation wi ll be dev e loped for the Operating License Application. System Components System component information will be provided in the Operating Licen se Appl i cat ion. Functional Requirements Functional requirements inform ation w ill be pro v ided in the Operating License Application.

Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analy s is i s based on information de ve loped fo r the Construction Permit Application.

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

6.3.1.2.12 IROFS CS-18, Backflow Prevention Device IROFS CS-18, " Backflow Pre ve ntions Device ," is identified by the accident ana l yses de scr ibed in Chapter 13.0. See d esc ription in Section 6.2.1.7.9. Accident Mitigated See description in Section 6.2.1.7.9. System Components See de sc ription in Section 6.2.1. 7 .9. Functional Requirements See de sc ription in Section 6.2.1.7.9. Design Basis See description in Section 6.2.1.7.9. 6-68 Test Requirements See description in Section 6.2.1.7.9. 6.3.1.2.13 IROFS CS-19, Safe-Geometry Day Tanks NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features IROFS CS-19 , "Safe Geometry Day Tanks," is identified by the accident analyses described in Chapter 13.0. See description in Section 6.2.1.7.9. Accident Mitigated See description in Section 6.2.1.7.9. System Components See description in Section 6.2.1.7.9.

Functional Requirements See description in Section 6.2.1.7.9. Design Basis See description in Section 6.2.1.7.9. Test Requirements See description in Section 6.2.1.7.9. 6.3.1.2.14 IROFS CS-20, Evaporator/Concentrator Condensate Monitoring IROFS CS-20 , "Evaporator

/Concentrator Condensate Monitoring,

is identified by the accident a nal yses described in Chapter 13.0. As an AEC , the condensate tanks will use a continuous active uranium detection system to detect hi gh carryover of uranium that shuts down the evaporator feeding the tank. The purpose of this system is to (I) detect an anoma ly in the evaporato r or concentrator indicating hi gh u ranium content in the condenser (due to flooding or excessive foaming), and (2) prevent high concentration uranium so lution from being available in the condensate tank for discharged to a favorable geo m etry syste m or in the con d e n ser for l eaki n g to the non-safe geometry cooling loop. Accident Mitigated The safety function of this IROF S is to prevent an accidental nuclear criticality because of excessive uranium in the condensate carryover to a non-geometrically favorab l e waste collection tank. System Components System components consist of: Condensate samp l e tank 1 A Conde n sate delay tank 1 Co nden sate samp le tank l B Co nden sate samp l e tank 2A Condensate delay tank 2 Con den sate samp l e tank 2B Condensate sampling systems Condensate monitors (UR-TK-340) (UR-TK-360) (UR-TK-370) (UR-TK-540) (UR-TK-560) (UR-TK-570) 6-69 Functional Requirements NWMl-2013-021, Rev. 3 Chapter 6.0 -Engineered Safety Features The detection system works by continuously monitoring condensate uranium content and detecting high uranium concentration, and then shutting down the evaporator to isolate the condensate from the condenser and condensate tank. At a limiting setpoint , the uranium monitor detecting device will close an isolation valve in the inlet to the evaporator (or otherwise secures the evaporator) to stop the discharge of high uranium content solution into the condenser and condensate collection tank. The uranium monitor is designed to produce a valve-open permissive signal that fails to an open state , closing the valve on loss of electrical power. The isolation valve is designed to fail-closed on loss of instrument air , and the solenoid is designed to fail-closed on loss of signal. Locations where these IROFS are used will be determined during final design. Design Basis Design basis information will be provided in the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

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

6.3.1.2.15 IROFS CS-26, Processing Component Safe Volume Confinement IROFS CS-26, "Processing Component Safe Volume Confinement," is identified by the accident analyses described in Chapter 13.0 (see description in Section 6.3.1.2.2). Accident Mitigated See description in Section 6.3.1.2.2. System Components As a PEC, some processing components (e.g., pumps, filter housings, and IX columns) w i ll be controlled to a safe volume for safe storage and processing of the fissile solutions.

Components that may be controlled to a safe volume will be described in the Operating License Application. Functional Requirements The safety function of a safe-volume component is also one of confinement of the contained solution.

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

Where insulation is used on the outside wall of the component, the insulation will be foam or encapsulated type (so as not to soak up solution during a leak) and will be compatible with the chemical nature of the contained solution. Design Basis The safe-volume confinement components will be determined in final design after finalizing the referenced CSEs. 6-70 NWM I ...*.. *

NORTHWEST MEDICAL ISOTOl"ES Test Requirements NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features The above ana ly sis is based on information developed for the Co n struct i on Permit Application.

Additional detailed information on test requirements w i ll be developed for the Operating License App li cation. 6.3.1.2.16 IROFS CS-27, Closed Heating or Cooling Loop with Monitoring and Alarm IROFS CS-27, "Closed Heating or Coo lin g Loop wit h Monitoring and A l arm ," is identified by the accident analyses in Chapter 13.0. As a PEC, closed coo lin g water loop s with monitoring for br ea kthrough of process so luti on will be provided on the evaporator or concentrator condensers to contain process so lution that leaks across this boundary, ifthe boundary fai l s. This IROFS will be app li ed to those high-heat capacity cooling jackets (requiring very l arge loop heat exchangers) servicing condensers where the leakage is always from the coo lin g l oop to the condenser.

The inherent characteristics of the l eak path will reduce back-leakage into the closed loop system , and the risk of product solutions e nterin g the condenser will be very low by evaporator and concen trat or design. System Components The purpose of this safety function is to monitor the he a lth of the con d enser cooling jacket to ensure that in the unlikely event that a condenser overflow occurs, fissi l e and/or hi gh-dose process solution will n ot flow into this non-safe-geometry coo lin g loop and cause nuclear crit icali ty. The closed loop wi ll also isolate any hi g h-d ose fissi l e product solids , from the same event , from penetrating the hot cell sh i e ldin g boundary, and any high-dose fission gases from penetrating the hot cell shielding boundary during normal operations. Functional Requirements The heat exc han ger materials will be compatib l e with the har sh che mic a l environment of the tank or vessel process (this may vary from app lic atio n to app lic a tion). Samp lin g of the coo ling media (e.g., cooling water radiological activity , or uranium concentration) will be conducted to a l ert the operator that a breach has occurred, and that additiona l corrective actio n s are required to identify and isolate the failed component a nd restore the closed-loop integrity.

Closed-loop pressure will also be monitored to identify a l eak from the closed loop to the process system. Discharged so luti o n s from this system will be h andled as potentia ll y fissile and samp l ed prior to discharge to a non-safe geometry. Design Basis Design basis inform ation wi ll be provided in the Operating License Application. Test Requirement s The above ana l ysis is based on information developed for the Construction Permit App li cation. Additional detailed informat ion on test requirements will b e developed for the Operating License Application.

6.3.2 Surveillance Requirements A review of s urv eillance requirements to e n s ur e the ava ilabilit y and reliability of safety controls when required to perform safety functions wi ll be included in the Operating License App li cation. 6.3.3 Technical Specifications Th e technical specificatio n s will be prov id e d in the Operating License Application.

6-71

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6.4 REFERENCES

NWMl-2 0 13-021, Rev. 3 Chapter 6.0 -Engineered Safety Features 10 CFR 20 , "Standards for Protection Against Radiation ," C o d e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 20.1201 , "Occupational Dose Limits for Adults," C o d e of F e d e ral R e gulations , Office of the Federal Register , as amended. 10 CFR 20.1301 , "Dose Limits for Individual Members of the Public," Cod e of F e deral R e gulation s, Office of the Federal Regi s ter , as amended. 10 CFR 50.59, "Changes, Te s ts , and Experiments

," Cod e of F e d e ral R eg ulation s, Office of the Federal Register , as amended. 10 CFR 70.61 , " Performance Requirements

," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. ANS V ANS-8.1 , N ucl e ar Criti c ali ty Saf e ty in Op e ration s with Fi ss ionabl e Mat e rial Out s id e of Rea c tor s, American National Standards In s titute/American Nuclear Society , LaGrange P a rk , Illinois , 2014. ANS V ANS-8.3 , Criti c ality A cc id e nt Alarm S ys t e m , American National Standards Institute/American Nuclear Society , La Grange Park , Illinoi s , 1997 (Reaffirmed in 2012). ANSVANS-8.7 , N ucl e ar Criticali ty Saf e ty in th e Storag e of Fi s sil e Mat e rials , American National Standards Institute/ American Nuclear Society , La Grange Park , Illinois , 1998 (Reaffirmed in 2007). ANSVANS-8.10 , Crit e ria for N ucl e ar Criti c ality Saf e ty Controls in Op e ration s with Shi e lding and Confin e m e nt , American National Standards Institute/Am e rican Nucle a r Society , La Grange Park , Illinois , 2015. ANS V ANS-8.19 , Administrati ve Practi ces for N ucl e ar Criti c ali ty Saf ety, American National Standards Institute/American Nuclear Society, La Grange Park , Illinois , 2014. ANS V ANS-8.20 , N ucl e ar Criti c ali ty Saf ety Training , American National Standards Institute/ American Nuclear Society , La Grange Park , Illinois , 1991 (Reaffirmed in 2005). ANSVANS-8.22 , N ucl e ar Criti c ali ty Saf e ty Ba se d on Limitin g and Contr o llin g Mod e rat o r s, American National Standards Institute/American Nuclear Society , La Grange Park , Illinoi s, 1997 (Reaffirmed in 2011 ). ANSVANS-8.23 , N ucl e ar Criti c ali ty A cc id e nt Em e r ge n cy Plannin g and R e spon se , American National Stand a rds In s titute/ American Nuclear Society , La Grange Park , Illino i s, 2007 (Reaffirmed in 2012). ANSVANS-8.24 , Validation o f Ne utron Tran s p o rt M e thod s f or Nucl e ar Criti c ality Saf ety Cal c ulations , American National Standards Institute/American Nuclear Society , La Grange P a rk , Illinois , 2007 (Reaffirmed in 2012). ANS V ANS-8.26 , Criticali ty Saf ety Engin ee r Trainin g and Qualifi c ation Pro gr am, American National Standard s Institute/American Nuclear Society , La Grange Park , Illino i s , 2007 (Reaffirmed in 2012). ANSVANS-15 .1 , Th e Dev e lopm e nt of T ec hni c al Sp ec ification s for Res e ar ch R e a c tor s, American National Standards Institute/ American Nuclear Society , LaGrange Pa r k , Illinois , 2013. ANSI Nl3.l, Sampling and Monitoring Relea s es of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities , American Nuclear Society , La Grange Park , Illinoi s, 201 I. 6-72 NWMl-201 3-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features ASME AG-1 , Cod e on Nuclear Air and Gas Tr e atm e nt , American Soc i ety of Mechanical Eng in eers, New York, New York, 2003. LA-CP-13-00634 , MCNP6 Us e r Manua l , R ev. 0 , Los A l amos Natio n a l Laboratory , Los Alamos, New Mexico, May 20 1 3. N R C , 2012, Final Int e rim Staff Guidanc e Augm e nting NUREG-I 53 7 , " Guid e lin es for Pr e paring and Reviewing Applications for the Licensing of Non-Power Reactors ," Parts I and 2 , for Licensing Radioisotope Produ c tion Fa c iliti e s and A qu e ou s Homog e n e ous R e a c tor s, Docket Number: NRC-2011-0135 , U.S. Nuclear Regulatory Commission , Washington , D.C., October 30 , 2012. NUREG-1 520 , Standard R e vi ew Plan for the R e vi e w of a Li ce ns e Application for a Fu e l C y cle Fa c ili ty, Rev. 1 , U.S. Nuc l ear Regulatory Commission , Office of Nuclear Materia l Safety and Safeguards , Washington , D.C., May 2010. NUREG-1537 , Guidelin e s for Pr e paring and R e vi e wing Appli c ations for th e Li ce nsing of N on-Power R e actors -Format and Content, Part 1 , U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation , Washington , D.C., February 1 996. NUREG/CR-4604 1 PNL-5849 , Statistical M e thod s for Nucl e ar Material Manag e m e nt, Pacific Northwest Laboratory , Richland, Washington, December , 1 988. NUREG/C R-6698 , Guid e for Validation of Nucl e ar Criti c ali ty Saf e ty Cal c ulational M e th o d o lo gy, U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety and Safeguards, Was hin gton , D.C., January 2001. [P roprietary Information]

[Proprietary In formation]

NWMI-2015-SD D-013 , S y st em D es ign D e s c ription for V e ntilation , R ev. A, Nort h west Medi ca l Isotopes , LLC, Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-001 , Single Paramet e r Subcriti c al Limits for 20 wt% 235 U-Uranium M e tal , Uranium O x id e, and Homogenous Wat e r Mi x tur e s , Rev. A , Northwest Medical Isotopes , LLC , Corvallis , Oregon , 2015. NWMI-20 l 5-CRITCALC-002 , Irradiat e d Targ e t Low-Enrich e d Uranium Mat e rial Dissolution , Rev. A Nort h west Medical I sotopes , LLC , Corvallis , O r egon , 2015. NWMI-2015-CRITCALC-003, 55-Gallo n Drum A rrays , Rev. A Nort h west Medical I sotopes, LLC , Corva lli s , Oregon , 2015. NWMI-2015-CRITCALC-005 , Targ e t Fabricati o n Tank s, W e t Pro ce ss e s , and Storag e, Rev. A , Northwest Medical Isotopes , LLC , Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-006 , Tank Hot C e ll , Rev. A , Northwe s t Medical I s otopes , LLC , Corva ll is, Oregon , 2015. NWMI-2015-CSE-OO 1 , NWMI Pr e liminary Criti c ali ty Saf e ty Evaluation

Irradiat e d Tar ge t Handling and Disas se mbl y, Rev. A , Nort h west Medical I sotopes, LLC, Corvallis , Oregon , 20 15. NWMI-20 l 5-CSE-002 , NWM I Pr e liminary Criti c ality Saf e ty Evaluation:

Irradiat e d Low-Enriched Uran i um Targ e t Material Di s solution , Rev. A , Northwe s t Medical Isotopes , LLC , Corva lli s, Oregon , 2015. NWMI-20 l 5-CSE-003 , N WMI Pr e liminary Criti c ali ty Saf ety E v aluation:

M o l y bd e num-99 R e cov e ry, Rev. A , Northwest Medical Isotopes, LLC , Corva lli s , Oregon, 2015. 6-73 NWM I ...... *

NCMl'TNWEST 111£DtCAl ISOTOPU .. NWMl-2013-021 , Rev. 3 Chapter 6.0 -Engineered Safety Features NWMI-2015-CSE-004 , NWMI Pr e liminary Criticality Safety Evaluation:

Low-Enrich ed Uranium Targ et Material Produ c tion , Rev. A , Northwest Medical Isotopes , LLC , Corvallis, Oregon , 2015. NWMI-2015-CSE-005 , NWMI Preliminary Criti c ality Saf ety Evaluation:

Targ et Fabri ca tion Uranium Solution Pro cesses, Rev. A , Northwest Medical I sotopes, LLC, Corva lli s, Oregon , 2015. NWMI-2015-CSE-006 , NWMI Preliminary Criticality Safety Evaluation:

Target Finishing, Rev. A, Northwest Medical Isotopes , LLC , Corvallis , Oregon , 2015. NWMI-2015-CSE-007 , NWMI Preliminary Criticality Safety Evaluation:

Target and Can Storage and Carts, Rev. A , Northwest Medical Isotopes, LLC , Corva lli s, Oregon , 2015. NWMI-2015-CSE-008 , NWMI Preliminary Criticality Safety Evaluation:

Hot Cell Uranium Purification , Rev. A, Northwest Medical Isotopes , LLC, Corvallis , Oregon , 20 I 5. NWMI-2015-CSE-009 , NWMI Pr e liminary Criticality Saf ety Eva luati on: Liquid Wa ste Proc ess ing, Rev. A , Northwest Medical I sotopes, LLC , Corvallis , Oregon, 2015. NWMI-2015-CSE-010 , NWMI Pr e liminary Criticality Safety Evaluation:

Solid Wast e Collection , Encapsulation , and Staging, Rev. A, Northwest Medical Isotopes , LLC , Corvallis, Oregon, 2015. NWMI-2015-CSE-Ol I , NWMIPreliminary Criticality Saf ety Evaluation:

Off gas and Ventilation, Rev. A , Northwest Medical Isotopes, LLC, Corvallis, Oregon , 2015. NWMI-20I5-CSE-O12 , NWMJ Pr e liminary Criticality Safety Eva luati on: Targ e t Transport Cask or Drum Handling , Rev. A , Northwest Medical Isotopes, LLC, Corvallis , Oregon , 2015. NWMI-2015-CSE-013 , NWMI Preliminary Criticality Saf ety Evaluation

Analytical Laboratory, Rev. A , Northwest Medical I sotopes, LLC , Corvallis, Oregon , 2015. Regulatory Guide 3. 71 , Nuclear Criticality Saf e ty Standards for Fuels and Material Fa c iliti es, Rev. 2, U.S. Nuclear Regulatory Commission , Washington , D.C., December 20 10. 6-74

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, * ** * ** * * ** * * . *. *. * . NORTHWEST MEDICAL ISOTOPES *

Chapter 7.0 -Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021 , Rev. 3 September 2017 Northwest Medical Isotopes , LLC 815 NW g th Ave , Suite 256 Corvallis , OR 973 3 0 This page intentionally left blank.

NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Chapter 7.0 -Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 3 Date Published:

September 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 3 T i tle: Chapter 7.0 -Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

c w.J11t+-c.

NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems This page intentionally left blank.

Rev Date 0 6/29/2015 1 6/26/2017 2 8/5/2017 3 9/5/2017 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems REVISION HISTORY Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to NRC C. Haass Requests for Additional Information Modification based on ACRS comments C. Haass Incorporate final comments from NRC Staff and ACRS; C. Haass full document revision NWMI .... ** ...*... .......... . * *! .' NOftTHWEST MEIHCAL lSOTOPU NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems This page intentionally left blank.

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NORTifW'EST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems CONTENTS 7.0 INSTRUMENTATION AND CONTROL SYSTEMS ................................................................. 7-1 7.1 Summary Description

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................. 7-1 7 .2 Design oflnstrumentation and Control Systems ....................

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7-4 7.2.1 Design Criteria .....................

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........................................................ 7-4 7.2.2 Design Basis and Safety Requirements

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..................................................... 7-4 7.2.3 System Description

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........... 7-13 7.2.3.1 Facility Process Control System ................................................

....... 7-14 7.2.3.2 Engi neered Safety Feature Actuatio n Systems ................................. 7-14 7.2.3.3 Control Room/Human-Machine Interface Description

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7-14 7.2.3.4 Building Management System ................

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.......................... 7-15 7.2.3.5 Fire Protection System ............................................

.......................... 7-15 7.2.3.6 Facility Communication Systems .........................................

............. 7-15 7.2.3.7 Ana lytical Laboratory System ..................................................

......... 7-15 7.2.4 System Performance Analysis ................

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7-16 7.2.4.1 Facility Trip and Alarm Design Basis ............

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7-16 7.2.4.2 Ana lysi s ..................................

........................................................... 7-17 7.2.4.3 Concl usion .................................................................

....................... 7-17 7.3 Process Control Systems ............................................

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7-22 7.3. l Uranium Recovery and Recycle System ............................................................

7-23 7.3.1. l Design Criteria ....................

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........ 7-23 7.3.1.2 Design Basis and Safety Requirements

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.. 7-23 7.3.1.3 System Description

........................................................................... 7-23 7.3.1.4 System Performanc e Analysis and Conclus i on .................................

7-28 7.3.2 Target Fabrication System ..................

............................................................... 7-28 7.3.2.1 Design Criteria ...........................

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.................. 7-29 7.3.2.2 Design Ba sis and Safety Requirements

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...... 7-29 7.3.2.3 System Description

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.... 7-29 7.3.2.4 System Performance Analysis and Conclusion

................................. 7-32 7.3.3 Target Receipt and Disa ss embly System ....................................

....................... 7-32 7.3.3.1 Design Criteria ..................................................................................

7-32 7.3.3.2 Design Basis and Safety Requirements

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............... 7-32 7.3.3.3 System Description

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................... 7-33 7.3.3.4 System Performance Analysis and Conclusion

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..... 7-33 7.3.4 Target Di ssol ution System ..................................................

............................... 7-33 7 .3 .4.1 D e sign Criteria ........................

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... 7-34 7.3.4.2 Design Basis and Safety Requirements

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7-34 7.3.4.3 System De scriptio n .............................

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.......... 7-34 7.3.4.4 System Perform ance Analysi s and Conclu s ion ............

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........... 7-37 7.3.5 Molybdenum Recovery and Purification System ............................................

... 7-37 7.3.5.1 DesignCriteria

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7-37 7.3.5.2 Design Basis and Safety Requirement s ..........................

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7-37 7.3.5.3 System De scriptio n ........................................................................... 7-38 7.3.5.4 System Performance Analysis and Conclusion

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............ 7-39 7-i NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.3.6 Waste Handling System ..................................................................................... 7-39 7.3.6.1 Design Criteria .....................................................................

............. 7-40 7.3.6.2 Design Basis and Safety Requirements

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................. 7-40 7.3.6.3 System Description

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7-40 7.3.6.4 System Performance Analysis and Conclusion

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.... 7-43 7.3.7 Criticality Accident Alarm System ................

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...... 7-43 7.3.7.1 DesignCriteria

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.... 7-43 7.3.7.2 Design Basi s and Safety Requirement s ..............

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............. 7-43 7.3.7.3 System De sc ription ....................................................

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......... 7-43 7.3.7.4 System Performance Analysis and Conclusion

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7-43 7.4 Engineered Safety Features Actuation Systems ...................................................

............ 7-44 7.4.1 System Description

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........... 7-44 7.4.2 Annunciation and Display .................................................................................. 7-45 7.4.3 System Performance Analysis ............................................................................ 7-45 7.5 Contro l Console and Display In st ruments ...........

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7-46 7.5. l Design Criteria ..........................................

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....... 7-46 7.5.2 Design Basis and Safety Requirements

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........................................... 7-46 7.5.3 System De sc ription .................................

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7-46 7.5.4 System Performanc e Analysis and Co nclusion .................................................. 7-46 7.6 Radiation Monitoring Systems ....................................

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.............. 7-47 7.6.1 Design Criteria ........................................

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...... 7-47 7.6.2 Design Basi s and Safet y R e quirement s .....................................................

......... 7-47 7.6.3 System Description

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7-47 7.6.3.1 Air Monitoring

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............. 7-48 7.6.3.2 Stack Release Monitoring

................................................................. 7-49 7.6.4 System Performance Analysis and Conclusions

................................................ 7-49 7.7 Reference s ...................................................................................

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7-50 7-ii Figure 7-1. Table 7-1. Table 7-2. Table 7-3. Table 7-4. Table 7-5. Table 7-6. Table 7-7. Table 7-8. Table 7-9. Table 7-10. Table 7-11. Table 7-12. Table 7-1 3. NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems FIGURES Radioisotope Production Facility Instrumentation and Control System Configuration

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................. 7-2 TABLES Instrumentation and Control System Design Criteria (9 pages) ...................................... 7-5 Instrumentation and Control Criteria Crosswalk with Design Basis App li cability and Funct ion Means ( 4 pages) ...........

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......................................................... 7-I 8 Uranium Recovery and R ecycle Contro l and Monitoring Parameters (2 pages) ........... 7-24 Uranium Recycle and Recovery System Interlocks a nd Permi s sive Signals (4 pages) ..................................................

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........................... 7-25 Target Fabrication Syste m Control and Monjtoring P arameters (2 pages) .................... 7-29 Target Fabrication System Interlocks and Permissive Signals (2 pages) ...................... 7-31 Target Dissolution System Contro l and Monitoring Parameters

................................... 7-35 Target Dissolution System Interlocks and Permissive Signals (2 pages) ...................... 7-36 Molybdenum Recovery an d Purification System Co ntrol and Monitoring Parameters

...................................................................................................................... 7-38 Molybdenum Recovery and Purification System Interlocks and Permissive Signal s ......................

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.............................. 7-39 Wa s te Handling System Control and Monitoring Parameters

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...................... 7-41 Waste Handling System Interlocks and Permissive Signals .......................................... 7-42 Engineered Safety Feature Actuation or Monitoring System s (2 pages) ....................... 7-44 7-iii NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems TERMS Acronyms and Abbreviations 9 9 Mo molybdenum-99 ADUN acid-deficient uranyl nitrate ALARA as low as reasonably ac hi evable BMS building management system CAAS criticality acci d ent alarm system CAM continuous air monitor CFR Code of Federal Regulations CGD commercial grade dedication COTS commerc i a l off-the-shelf DCS digital co ntrol syste m ESF engineered safety feature FPC facility process contro l HMI human-machine interface I iodine I&C instrumentation and co ntrol IEEE Institute of Electrical and E l ectronics Engineers IROFS items relied on for safety ISA integrated safety ana l ysis IX ion exc h a n ge Kr krypton LEU low-e nri c h ed uranium Mo molybdenum NAVLAP National Voluntary Laboratory Accreditation NO , nitrogen oxide NRC U.S. Nuclear R egulatory Commission NWMI Northwes t Medical Isotopes, LLC PLC programmable lo gic contro ll er RAM radiation area monitor RPF Radioisotope Production Facility SDOE secure development and operational environment SIF safety instrumented function.

S IL safety integrity l evel. SIS safety instrumented system SNM special nu c l ear material SSC structures, systems , and components TCE trichloroethy l ene U.S. United Sta t es [Proprietary Inform atio n] UPS V&V Xe Units m mm rad [P roprietary Inform ation] uninterruptible power s uppl y verificatio n and validation xenon meter minute radiation absorbed dose 7-iv NWMl-2013-021, Rev. 3 Chapter 7.0 -I nstrumentation and Control Systems 7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.1

SUMMARY

DESCRIPTION The Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) preliminary instrumentation and control (I&C) configuration includes the special nuclear material (SNM) preparation and handling processes (e.g., target fabrication, and uranium recovery and recycle), radioisotope ex t raction and purification proces ses (e.g., target receipt and disassembly, target dissolution , molybdenum

[Mo] recovery and purification , and waste handling), process utility systems, criticality accident alarm system (CAAS), and sys tems associated with radiation monitoring.

The SNM processes will be enclosed predominately by hot cells except for the target fabrication area. The facility process control (FPC) syste m will provide monitoring and control of the process systems wi t hin the RPF. In addition, the FPC system will provide monitoring of safety-related components within the RPF. The proce ss strategy for the RPF involves the use of batch or semi-batch proces ses with relatively simple control steps. The building management system (BMS) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (tum on and off) the mechanical utility systems. Engineered safety feature (ESF) systems will operate on actuation of an alarm setpoint reached for a specific monitoring instrument/device.

For redundancy , this will be in addition to the FPC system or BMS ability to actuate ESF as needed. Each ESF safety function will use hard-wired analog controls/interlocks to protect workers , the public, and environment.

The ESF parameters and alarm functions will be integrated into and monitored by the FPC system or BMS. The preliminary concept for the RPF I&C system configuration is shown in Figure 7-1. The green circles identify the FPC and the BMS distributed process control or programmable logic controller (PLC) systems. The solid lines and dashed lines show how the SNM processes , support systems, utilities , ra d iation and criticality systems, and building functions relate to the FPC and BMS and to local machine interface (HMI) stations. Solid lines indicate the control functions , and dashed lines indicate the monitoring function s. The FPC system will perform as the overall production process controller.

This system will monitor and control the process instrumented functions within the RPF, including monitoring of process fluid transfer s and controlled inter-equipment pump transfers of process fluids. Process control systems are described further in Section 7.3. Th e fire protection system will have its own central alarm panel (green circle). The fire protection system will report the status of the fire protection equipment to the central alarm station and the RPF control room. The fire protection system is discussed further in Section 7.2.3.5. 7-1 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Process Support Systems ---e Waste Handling System ,--1 L------------------

.... , 1 I ,--1 I I I 1--1 I I I I I I I I I I I I I ----__ I I I I I I I I I I I I I I I ---e Process Utility Systems Fire Protection System I I I I --: Facility Ventilation Sys t em I I I _J Process Systems In Hot Cell Area ---l> Monitor i ng Only Control and Monitoring Process Systems In Target Fabrication Area Figure 7-1. Radioisotope Production Facility Instrumentation and Control System Configuration Special nuclear material preparation and handling processes

-The FPC system wi ll co ntrol and/or monitor the SNM preparation and handling proce sses, the following.

Target fabrication

-Batch processe s located in the target fabrication area will be contro ll ed by operators at local HMi s , with surveillance monitoring in the control room. Uranium recovery and rec y cle -Batch processes located inside the hot cell area will be monitored and controlled by operators in the control room. Radioisotope extraction and purification processes

-The FPC system will contro l and/or monitor the radioi so tope production processe s, including the fo llowin g. *

Target receipt and disassembly

-Hardware/target movement located in irradi ated target b asket receipt bay a r ea, target cask preparation ai rlock , target receipt hot cell , and target disassembly hot cell will normally be controlled by operators at lo cal HMis , with survei ll ance monitoring in the contro l room. Target dissolution

-Batch process located inside the dissolution hot cell wi ll occur at loc a l HMis in the operating gallery, and off gas operations in the tank hot ce ll wi ll be contro ll ed by operators in the control room , with survei ll ance monitoring at both location s. 7-2

NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Mo recovery and purification

-Batch processes lo cated inside the Mo hot cells wi ll be contro lled by operators at a local HMI in the operating ga ll ery , with surve ill a n ce monitoring in t he control room. Waste handling -This system includ es liquid waste handling, liquid waste so lidifi cat ion , a nd so lid waste handling.

Operators in the contro l room will control liquid waste handlin g, while operators at local HMis in the low-do se liq uid so lidific ation room (W 107) will monitor and control liquid waste so lidifi cation, and solid waste nondestructive examination and so lidification. Process utility and support systems -The FPC system will control and monitor the process utility and process sup port systems. Operators in the control room wi ll control the following subsystems:

*

Process chilled water hot cell seco ndary loops Process steam hot cell secondary loop s Process vessel ve ntil a tion system Operators at local HMis wi ll control the following subsystems, with s ur veillance monitoring in the control room u s in g the FPC system or BMS. * * *

* *

Plant air system Gas s uppl y system Process chilled water chillers Process steam boilers Demineralized water system Chemical sup ply system Standby e l ectrica l power syste m Cr i ticality accident alarm system -The CAAS will be provided as an integrated vendor package. The detectors an d alarm response are integral to the individual units/locations.

The FPC system will monitor the CAAS status in the contro l room. The CAAS is described further in Section 7.3. Radiation monitoring system -The FPC system will monitor the vario u s radiation monitoring systems , in c ludin g continuous air monitors (CAM), air samp l ers , radiation area monitors (RAM), and exha u st stack monitors.

The CAMs and RAMs will be strateg ic a ll y placed throughout the RPF to alert personnel of a n y potential radiation hazard s. The CAMs and RAMs will alarm in the control room a nd locally at loc a tions throughout the RPF. The rad i at ion monitorin g systems are d escribed further in Section 7.6. Facility ventilation system and mechanical utility sys t ems -The control function for most of the RPF ve ntil atio n system and mechanical utility systems wi ll be loc al HMis and h ard-wired interlocks for the ESF functions.

The BMS will monitor the systems and provide venti l ation and mechanical utility system status as an input to the FPC process contro l s. Th e following s ub syste m s will be monitored by the BMS: * * *

Facility ventilation Zones I , II, III , and IV Supply air system Facility chilled water system E nergy reco very and heatin g water 7-3 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Safety-Related Components and Engineering Safety Features The ESF safety functions will operate independently from the FPC sys tem s as hard-wired analog controls or interlocks.

Th e FPC system will be a digital control system (DCS) that monitors safety-related components within the RPF. The ESFs will be integrated into the FPC syste m s and provide a common point ofHMI, monitoring, and alarming at the control room and, as nece ssary, local HMI workstations. Control Console and Display Instruments Th e control room will be the primary interface location for the RPF s upport systems and provide centralized proc ess controls, monitoring , alarms, a nd acknowledgement.

Mechanical util i ty systems with vendor package s and integrated controls will be controlled at associated local HMis. The BMS will provide primarily on/off control and system monitoring from the control room. The tank hot cell processes will be controlled primarily in the control room, with surveillance monitoring of the FPC subsystems.

The FPC sys tem will h ave annunciation, alarms, and HMI di splays. From the consoles, operator s will view and trend essential measurement values from the HMI di sp lay , and evaluate real-time data from the essential measurements u se d to control and monitor the RPF process. This system is further described in Section 7.5. Proces s utility and s upport systems with vendor p ac kage a nd integrated controls will be operated at associated local HMis. These syste ms are discu sse d further in Section 7.5. Local HMis are anticipated in the following loc ations: * * * * *

7.2 Irradiated target basket receipt b ay N B (R102NB) Cask preparation airlock (RO 12) Operating gallery (GIOI N B I C) Target fabrication (Tl 04 NB) Low-do se liquid waste so lidification (W 107) Chemical su pply room (L102) Local to equipment with integrated control sys tems DESIGN OF INSTRUMENTATION AND CONTROL SYSTEMS The design criteria a nd the codes a nd standards for I&C systems are outlined in Chapter 3.0, " Design of Structures, System s, and Components," and di scusse d below. 7.2.1 Design Criteria Th e ap plicable de s ign criteria and guidelines that apply to the RPF I&C sys tems are summarized in column one of Table 7-1. Additional, design criteria for I&C sys tems are provided in Chapter 3.0. Th e detailed and specific de s ign criteria for I&C systems will be confirmed in the Operating License Application.

7.2.2 Design Basis and Safety Requirements The design basi s for l&C systems used in the RPF are presented in the seco nd column of Table 7-1. Th e second column maps the criteria to l&C systems or components and how compliance will be ensured. Note that the FPC syste m callouts may also apply to the BMS. The de s ign ba s is requirements for facility and process systems are described in Chapter 4.0 , "Radioisotope Production Facility Description

," and Chapter 9.0 , "Auxiliary Systems." The I&C system will use hard-wired interlock s for actuated engineered safety function s. Section 7.4 s ummarizes the I&C ESFs. 7-4 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description 3 Design bases as applied to RPF IEEE 379-2014, IEEE Standard Application of th e Application:

Single-Failure Criterion to N uclear Power Generating

Design ofFPC system, ESFs , and other Station Safety System s instrumentation SSCs that a r e id e nt ified as IROFS

Description:

Application of the single-fail ur e criterion Compliance:

to e l ectrica l power, in st rumentation , and control portion s

Ensure FPC system is a DCS designed , rated, and of nuclear power generating safety systems. approved for use in safety instrumented systems , as K e yw ords: Actuator, cascaded failure , common-cause determined by ANSI/ISA 84.00.0 I fa i lur e, design b asis event, detectable fai lur e, effects

Use a safety PLC , as recognized by IEC 6 150 8, in the analysis , safety system , s ingle-failure criter ion , system FPC system with r edunda nt power s upplies , actuation, sys t em l ogic pr ocessors, a nd inpu t/output c h anne l s IEEE 577-2012, IEEE Standard Requirements for Reliability Analysis in the Design and Operation of Safety Systems for Nuclear Facilities

==

Description:==

Sets minimum acceptable requirements for the performance of reliability analyses for safety systems when used to a ddress the reliability considerations discussed in industry standards and guidelines.

The requirement that a reliability analysis be pe r formed does not originate with thi s standard. However, when reliability analysis i s u se d to demonstrate compliance with reliability requirements , this standard describes an acceptable response to the requirements.

Keywords: Nuclear facilities, reliability analysis, safety systems

Eva lu ate contro l s that are classified as IROFS in C h apters 6.0 a nd 13.0 , or NWMI-20 I 5-SAFETY-002, aga in st si ngle-failure criteria Exception:

NUREG-1 537 a ll ows for shar in g and com binin g of systems and components with justification

The RPF is not considered a nuclear power reactor but a production facility. The facility will not have all of the systems detailed in thi s standard and guidance wi ll be applied as appropriate.

Application:

Use for design ofFPC system, ESFs, and other instrumentation SSCs that a re identifi ed as IROFS Compliance:

Perform a reliability analysis of the proposed design solution for IROFS functions , as identified in Chapters 6.0 and 13.0 , or NWMI-20 l 5-SAFETY-002.

The analysis can be qualitative or quantitative in nature , as de scri b e d in the stan dard 7-5 NWMI ...... ' * *

HOKTifW(ST MEDICAL I SOTOPES NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description 3 IEEE 603-2009, IEEE Standard Criteria for Safety Systems for N uclear Power Generating Stations Description

Establishes minimum functional and design cr iteria for the power , instrumentation, and control portions of nuclear power generating station safety systems. Criteria are to be applied to those systems required to protect public health and safety by functioning to miti gate the consequences of design basis events. The int ent is to promote appropriate practice s for design and evaluation of safety sys tem performance and reliabi li ty. The s tandard is limited to safety syste ms; many of the principles ma y have applicability to equipme nt provided for safe s hutdown , post-accident monitoring displa y in s trumentation , preventive interlock features , or any other systems , structures , or equipment related to safety. Keywords:

Actuated equipment, assoc i ated circuits , Class 1 E , design , failure, maintenance bypass , operatin g bypass , safety function , sense and command features , se n so r IEEE 384-2008, IEEE Standard Criteria for Independence of Class JE Equipment and Circuits

Description:

Describes independence requirements of circuits and equipment comprising or associated with Class 1 E systems. Identifies criteria for independence that can be achieved by physical separation, and electrical isolation of circuits and equipment that are redundant.

The determination of what is to be considered redundant is not addressed.

Keywords: Associated circuit, barrier , Class IE , independence, isolation, isolation device, raceway, separation Design bases as applied to RPF Application:

Use for de s ign ofFPC system, ESFs, and other instrumentation SSCs that are identified as IRO FS

Apply minimum functional and design criteria to safety syste m s Compliance:

Ensure design conforms to the practices detailed in the s tandard for the IROFS function s identified in Chapters 6.0 and 13.0 , or NWMI-20 1 5-SAFETY-002 Exception:

The RPF is not considered a nuclear power reactor but a production facility. The facility wi ll not ha ve all of the systems detailed in this standar d and guidance will b e applied as appropriate.

Application:

Use for design ofFPC system, ESFs, and other instrumentation SSCs that are identified as IROFS

Apply minimum criteria for separation and independence of sys tems in a physical way Compliance:

Ensure design conforms to the practices detailed in the standard for the IROFS functions identified in Chapters 6.0 and 13.0, or NWMI-2015-SAFETY-002 Exception:

The RPF is not considered a nuclear power reactor but a production facility.

The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

7-6 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description*

IEEE 323-2003, IEEE Standard for Qualifying Class 1 E Equipment for Nuclear Power Generating Stations Description

Identifies requirements for qualifying Class 1 E equipment and interfaces that are to be used in nuclear power generating stations.

The principle s, methods , and procedures are intended for use in qualifying equipment , maintaining and extending qualification, and updating qualification , as required , if the equipment is modified.

The qualification requirements of the stan dard demonstrate and document the ability of equipment to perform safety function(s) under applicable service conditions , including design basis events, reducing the risk of common-cause equipment failure. Keywords:

Age conditioning , aging, condition monitoring , design basi s event, equipment qualification , qualification methods , har s h environment , margin , mild environment, qualified life , radiation , safe ty-related function , significant aging mechani sm, test plan , test sequence , type testing IEEE 344-2004, IEEE Recommended Practice for Seismic Qualification of Class JE Equipment for Nuclear Power Generating Stations Description

Identifies recommended practices for establishing procedures that will yield data to demonstrate that the Class IE equipment can meet performance requirements during and/or following one safe shutdown earthquake event, preceded by a number of operating basis earthquake events. This recommended practice may be used to establish tests , analyses, or experience-based evaluations that will yield da t a to demonstrate Class 1 E equipment performance claims or to evaluate and verify performance of devices and assemblies as part of an overall qualification effort. Common methods currently in use for seismic qualification by test are presented.

Two approaches to seismic analysis are described:

one based on dynamic analysis, and the other on static coefficient analysis. Two approaches to experience-based seismic evaluation are described, one based on earthquake experience and the other on test experience.

Keywords:

Class IE, earthquake, earthquake experience, equipment qualification, inclusion rules, nuclear, operating basis earthquake, prohibited features, qualification methods, required response spectrum, response spectra, safe shutdown earthquake, safety function, seismic, seismic analysis, test response spectrum, test experience Design bases as applied to RPF Application:

Use for equipment qualification when needed to qualify equipment for applications or environments to which the equipment may be exposed

Use for qualification of Class IE equipment located in har sh environments and for certain post-accident monitoring equipment; may also be used for the qualification of equipment in mild environments Compliance:

Ensure design conforms to the practices detailed in the standard for those systems determined to be Class IE and located in harsh environments for safety functions identified in Chapters 6.0 and 13 , or NWMI-2015-SAFETY-002

Apply to SSCs within the hot cell area; not all safety components reside in the hot cell area

Apply sta ndard using a graded approach Exception:

The RPF is not considered a nuclear power reactor but a production facility. The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

Application:

Apply seismic design requirements for equipment used in Class 1 E systems Compliance:

Use in design ofFPC system, ESFs , and other instrumentation SSCs that are identified as a Class IE system Exception:

The RPF is not considered a nuclear power reactor but a production facility.

The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

7-7 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Crit e ria (9 pages) Design criteria description*

IEEE 338-2012, IEEE Standard for Criteria for the Periodic Surveillance Testing of N uclear Power Generating Station Safety Syst e ms Description

Pro vides cr it eria for the perform ance of periodic s urveill a n ce testing of nu c l ea r power generating stat ion safe ty sys tem s. The sco p e of p e riodic surve ill a nce te s tin g co n s i s ts of functional te s t s and checks , calibration ve rificati o n , and time response m eas ur e ments , as r eq uired , to verify th a t the safety sys tem perform s it s defined safety function. Po sm ai nt e nance and po st-modifi catio n testing a re not covere d by thi s do cu m e nt. This sta ndard amplifies the periodic surveillance testing requirements of other nuclear safety-r e l ated IEEE standards.

Keywords:

Functional t ests , IEEE 338 , p erio di c testing, ri sk-i nform e d testing, surve ill ance testing IEEE 497-2010, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations Description

Establishes criteria for variable selection, performance , design , and qualification of accident monitoring instrumentation , and includes the requirements for display alternatives for accident monitoring instrumentation , documentation of design bases , and use of portable instrumentation.

Keywords: Accident monitoring , display criteria , selection criteria , type variables IEEE 7-4.3.2-2010 , IEEE Standard Criteria for Digital Computers in Safety Sy s tems of N uclear Power Generating Station s Abstract:

Specifies a dditional co mput er-sp ec ifi c requirements to s uppl e m e nt IEEE 603-20 09. The s t an d a rd define s the term co mput e r as a sys t e m that includes computer hardware , software, firmware , a nd interfaces , and esta bli s h es minimum functional an d d esign requirements for co mput e r s u se d as components of a safe ty sys tem. Keywords: Commercial-grade it e m , div e rsity , safety sys t e ms , s oftware , softwa re tool s , software verification and va lidation Design bases as applied to RPF Application:

Use for d es i g n of FPC sys tem , ESFs , a nd o th e r in st rum e ntation SSCs that a r e id e n tified as IROF S

Use methods and c rit e ria to es t a bli sh a periodic s u rvei llan ce pro gram Compliance:

E n sure d es i g n conforms to the pr actices d e t a il e d in th e sta nd a rd for the IROFS fun c tion s identifi e d in C h apte r s 6.0 a nd 13.0, or NWMI-2015-SAF E TY-00 2 Exception:

The RPF i s not consi der e d a nu c l ear power reactor but a production facility.

Th e facility w ill not h ave all of the sys t e m s detailed in this s t andar d a nd g uid a n ce will be a pplie d as appropriate. Application:

Use as selection , design , performance, qualification , and display criteria for accident monitoring instrumentation

Apply guidance on the use of portable instrumentation and for examples of accident monitoring display configurations Compliance:

Ensure design conforms to standard for the monitoring functions de te rmined to be required for health and safety of workers or the public during normal operation and design basis accidents Exception:

The RPF is not con s idered a nuclear power reactor but a production facility. The facility will not have all of the systems detailed in this standard and guidance will be applie d as appropriate. Application:

In co njun ct ion with IEEE 603-2009 , u se to es tabli sh minimum fun c tion a l a nd d es i g n requirements for co mput e r s th at are co mponent s of a safe t y sys t em

D esig n F P C sys t e m as a D CS, and app l y this sta nd ard to sys t em d eve l o pm e nt , s pecific a ll y softwa r e de ve l op ment

A pply s tandard to CGD and impl eme nt a n approach Compliance:

D eve lop FPC system so ftware u s in g this s tandard Exception:

The RPF i s not considered a nucl ear power r eac tor but a production facility.

Th e facility will not ha ve a ll of the sys tem s d eta iled in this s t a nd a rd and guida n ce will be a pplie d as appropriate. 7-8 NWM I ......

e * . NOflTlfWUT MEDICAL ISOTOrEI NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description*

IEEE 828-2012, IEEE Standard/or Configuration Management in Systems and Software Engineering Description

Establishes minimum requirement s for c o nfiguration management in system s and software e n gineering. This s tandard applie s to any form , class , or type of software or s y s tem , and explains configuration management , including identifying and acquiring c o nfiguration items , controlling chan g es , reporting the status of configuration items , and performing software build s and release engineering. This s tandard addresse s what configurat i on management activities are to be done , when they are to happen in the life-cycle , and what planning and resources are required. The content areas for a configuration management plan are also identified.

The standard supports IEEE STD 12207 and ISO/IEC/IEEE 15288 , and adheres to the terminolo gy in ISO/IEC/IEEE STD 24765 and the information item requirements ofIEEE STD 15939. Keywords:

Change control , configuration accounting , configuration audit, configuration item , IEEE 828 , release engineering , software builds , software configuration management , system configuration management Design bases as applied to RPF Application:

Use to establish configuration management processes , define how configuration management is to be accomplished , and identify who is re s ponsible for performing specific activities , when the activities are to happen , and what s pecific re s ource s are required

Design FPC system as a DCS , and appl y standard during the development of s oftware for sys tems with IROFS functions Compliance:

Develop FPC system software using this standard for safety function implementation IEEE 1028-2008 , IEEE Standard for Software R e vi e w s Application:

and Audits D e scription:

Id e nt i fi es five typ es of softw ar e r ev i ews a nd a ud i t s , to ge th er w i t h pro ce dur es r e quir e d for th e exec ut io n of each t ype. This s t a nd a rd is c oncern e d o nl y with rev i e ws a nd audit s; p ro ce dur es fo r det e rmining t h e n e c ess it y of a r ev i ew o r a udit a r e n ot d e fin e d , a nd th e di s po s ition of th e r es ult s of the re v ie w or audit i s n o t s p e cifi e d. Typ es in c lud e d a re m a n age m e nt revi e w s , t ec hni ca l re v iew s , in s p ec tion s, wa lk-throu g hs , and audit s. K ey words: Audit , in s p ec tion , r ev i ew , wa lk-throu g h ANS 10.4-2008, Verification and Validation of Safety-Re/ated Scientific and Engineering Computer Programs for the Nuclear Industry Description

Provides guidelines for V&V ofnonsafety-related scientific and engineering computer programs developed for use by the nuclear industry.

Scope is restricted to research and other related, noncritical applications.

Use t o id e ntify minimum acc eptabl e r e quirement s fo r sys t e m a tic s oft w ar e r ev i ews

Id e nti fy or g ani za tion al m ea n s for c ondu c t i ng a re v i ew a nd d oc um e ntin g th e fi ndin gs

D es i gn F P C sys t e m as a D C S , a nd a pply s t a ndard during t h e de ve lopm e n t of so ftw a r e for sys t e m s with IROF S fun c tion s Compliance:

D eve l o p F P C s ys t em u s in g this s t a nd a rd Application:

Perform s oftware V&V to build quality into the software during the s oftware life-cycle

Use to verify and validate software development for non-safety-related systems

Use for software development in the RPF that is not safety significant (e.g., not safety-related or IROFS) Compliance:

Keywords:

Software integrity level, software life-cycle ,

Develop non-safety-related software us i ng this val i dation, verification , V & V standard 7-9 NWM I ...... *. *

NOllTHWEST MEDICAL tsoTOPlS NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description a ANSI/ISA 67 .04.01-2006, Setpoint s for N uclear Safety-Related Instrumentation Description

D e fin es r e quirement s for a s s ess in g , es t a bli s hing , and m ai nt a ining nucl ea r sa fet y-relat e d a nd oth e r important in s trum e nt s etp o int s ass ociated w ith nu c l ea r power pl a nts o r nucl e ar r eac t o r fa c ilitie s. Ke y words: S e tp o in t , drift , a n a l og c h a nn e l , r e liabili ty a n a l ys i s ANSI/ISA 84.00.01-2004, Functional Safety: Safety Instrumented Systems for the Process Industry Sector Part 1: "Framework, Definitions, System, Hardware and Software Requirements" Part 2: "Guidelines for the Application of ANSI/ISA-84.00.01-2004 Part 1 (IEC 61511-1 Informative" Part 3: "Guidance for the Determination of the Required Safety Integrity Levels -Informative"

Description:

Provides requirements for the specification , design , installation , operation , and maintenance of a safety instrumented system , so the s ystem can be confidently entru s ted to place and/or maintain the process in a safe state. This s tandard has been developed as a process sector implementation of IEC 61508. Keywords: Safety in s trumented system (SIS), safety integrated level (SIL), safety instrumented function (SIF) Design bases as applied to RPF Application:

Us e m e thod s and c rit e ri a to es t a bli s h se tpoint s for s afety s y s t e m s and t o m a int a in th e d o cument a tion

A ppl y t o the de s i gn o f th e F P C sys t em a nd oth e r in s trum e nt a ti o n SSCs th a t a r e id e n t i fied a s IRO FS fo r th e RP F Compliance:

E n s ur e d es i gn c o n fo rm s to the p rac ti c es d e t a il e d in the s t a nd a rd for IRO F S fun c tion s with inher e nt s etpoint s id e ntified i n C h a pt e r s 6.0 an d 13.0 , o r NW MI-20 I 5-SAFETY-00 2 Application:

Apply to the de s ign of safety sy s tem s (standard specifically designed for industrial processes)

Standard is made up of three parts: -Use Part l to lay the groundwork fo r the safety sy s tem life-cycle , overall structure of s afety systems , definitions u s ed , and to implement safety system de s ign engineering Use Part 2 guidance for the specification , design , installation , operation , and maintenance of safety instrumented functions and related safety instrumented sy s tems, as defined in Part 1 -Use Part 3 to develop underlying concepts of risk in relation to safety integrity , iden t ify tolerable risk, and determine the safety integrity levels of the safety functions

Design physical hardware of the FPC system based on this standard and IEC 61508

Evaluate the IROFS functions requir e d to be implemented by the FPC system usi n g Parts 1 , 2 , and 3 of this standard

Use to demonstrate reliability and risk reduction of the FPC system , while having simila r or higher documented and tested ability to reduce risk as fulfillment through other channels Compliance:

Use for the design and implementation for IROFS function s that are required of the FPC system 7-10 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description*

Design bases as applied to RPF NUREG-0700, Human-Sy s tem Interfa ce Design Application:

R eview Guidelines

Use compre h ens i ve design review guidance to Description

Provides guida nc e to the NRC on the develop information displayed in human-interface evaluatio n of hum an factors e n gi n eering aspects of systems nuclear power plants in accordance with NUREG-0800.

Develop informative and effect i ve designs that will Detailed design review procedures are provided in assist operators in the performance of their duties NUREG-0711.

As part of the review process , the Compliance:

interfaces between plant personnel and t h e plant systems

and compo nent s are evaluated for conforma nc e with Design FPC system to provide infonnation to operators in a di sp l ay format hum an facto r s enginee ring g uidelin es. Keywords: Di sp l ay , HMI , hum an-interface sys t em , human-system int erface NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems

Description:

Provide s guidance to the NRC on auditing programs for safety systems written in the following six high-level languages:

Ada, C and C++, PLC Ladder Logic, Sequential Function Charts, Pascal , and PL/M. The guidance could also be used by those developing safety significant software as a basis for project-specific programming guidelines. Keywords: Pascal , C, Ladder Logic , PL/M , Ada , C++, PLC, programming, seq uential function charts NUREG/CR-6090, The Programmabl e Logic Controller and It s Ap pli catio n in Nuclea r Reactor Systems Abstract:

Outlines recomme nd ations for review of the app li catio n of PLCs to the co ntrol , monitoring, a nd protection of nucl ear reactors.

Keywords:

PLC , progr a mming , protection systems

Display development used in connection wit h th e FPC system will be provided in the Operati n g License Application Application:

Use guidance to revie w high-integrity softwa re in a nuclear facility

Develop FPC system as a DCS , with associated programming development need s for the RPF

Use guideline as a means to review FPC system programming code Compliance:

Develop FPC system software programs using this guidance Exception:

The RPF is not considered a nuclear power reactor but a production facility.

The facility wi ll not ha ve all of the sys tems detailed in thi s sta ndard and guidance will be applied as appropriate. Application:

Use guidance to implement PLCs for nuclear app l ication and as a forum for what const itut es good practices of previously insta ll ed systems

Use guida n ce durin g select ion process for hardware , fa ilu re analys i s , a nd product li fe-cycle wit hin th e fac ili ty Compliance:

Design FPC system to use a PLC-type DCS

Select design a nd implement PLCs based on t hi s guide , as applicab l e Exception:

The RPF is not considered a nuclear power reactor but a production facility.

The fac ilit y wi ll n ot h ave a ll of the syste m s detailed in t hi s standard and gu id ance wi ll be applied as appropriate.

7-11

.. NWM I ..... .......... *. * ' NOflTifWUT MCDtCAl. tSOTI,,H NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description a EPRI TR-106439, Guideline on Evaluation and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications

==

Description:==

Provides a consistent , comprehensive approach for the evaluation and acceptance of Design bases as applied to RPF Application:

Use to identify appropri at e critical characteristics with subsequent verification through testing, analysis , vendor assessments, and careful review of operating experience commercial digital equipment for nuclear safety systems.

Keywords:

Commercial off-the-shelf(COTS), programming, software, commercial grade dedication Use guidance for digital upgrades to s afety-related systems and for non-safety-related applications that require high reliability or are compat i ble with specific change processes, including graded approaches for quality assurance Regulatory Guide 1.152, Criteria/or Use of Computers in Safety Systems of Nuclear Power Plants

Description:

De scribes a method that the NRC staff d eems acce ptable for co mplying with NRC r egu lation s for promoting high functional reliability , de s ign quality , and a sec ure development and operational environment for the u se of digital com puter s in the safety sys tem s of nucl ear power plants. Keywords:

Secure development an d operatio nal e n viro nm e nt (SDOE), co mput ers Regulatory Guide 1.53, Application of the Failure Criterion to Safety Systems

Description:

Provides methods acceptable to the NRC staff for satisfying NRC regulations with respect to the application of the single-failure criterion to the electrical power and I&C portions of nuclear power plant safety systems. Keywords: IEEE 379-2014, single-failure criterion Compliance:

Ensure that digital systems components that require CGD apply the guidance of this standard, as applicable Application:

Use for l&C system designs with comp uter s in related systems that make exte n s i ve use of advanced technology

Use for RPF d esig ns (that are expecte d to b e s i g nificantl y and functionally diff erent from c urr ent day process de s i g n s) with microproc ess ors , di g it a l sys tem s and displays, fiber optics , m u ltiplexing , and different i so l a ti o n techniques to achieve sufficient indep e nd e n ce and r ed undan cy Compliance:

Develop FPC sys t em and associated HMI u s in g thi s g uidan ce Exception:

The RPF i s no t considered a nucl ear power reactor but a production faci li ty. The fac ili ty w ill not ha ve a ll of the systems detailed in thi s standar d and guidance will be appl ied as appropriate.

Application:

Apply single-failure criterion to safety-related I&C systems

Apply to end-devices used by the FPC system that are identified as IROFS Compliance:

Evaluate FPC system, ESFs, and IROFS end-devices using this guidance 7-12 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (9 pages) Design criteria description*

Regulatory Guide I.97, Criteria for Accident Monitoring Instrum e ntation for N uclear Power Plant s Description

Pro v id es a m e thod th a t t he N RC s t a ff co n s id ers a ccept a bl e for u se in c ompl y in g with NR C regu l a t io ns w ith r es p ect t o sa ti sfy ing c r i t e ri a for acci d e nt m o nitorin g in s trum e n ta tion in nu c l ear pow e r plant s. Keywords: IEE E 4 97-20 I 0 , accid e nt m o nitorin g Regulatory Guide 5.71, Cyber Security Programs for Nuclear Facilities

==

Description:==

Provides an approach that the NRC staff deem s acceptable for complying with NRC regulations re g arding the protection of digital computers , communications sy s tems , and networks from a cyberattack , as defined by l 0 CFR 73. l. Keywords:

Cybersecurity , 10 CFR 73.54(a)(2), design basis threat *F ull r e fer e n ces p rovi d e d in S ec ti o n 7.7. CAAS CAM CF R CG D CO T S D CS ESF FPC HMI l&C I EEE c riti ca li ty acci d e nt a l arm syste m. c o ntinu o u s a ir monit o r. Co d e of Fe d e r al R egu l a ti ons. co mm e r cia l gra d e d e di cation. co mm e r cia l off-th e-s h e I f. di g it a l contro l sys t e m. e n gi n ee r e d safe t y fea tur e. fac il ity process co n tro l. hum a n-m ac hin e int e r face. in s trum e nt at i o n and co nt ro l. In s titute o f E l ec tri ca l a nd E l ec troni cs E n g in ee r s. Design bases as applied to RPF Application:

Use thi s gu idance for d eve lopm e nt of acc id e nt m o nitorin g for th e RP F Compliance:

D es i g n F P C sys t e m , CAAS, CAMs , a nd RAM s u s in g thi s gu id a n ce Exception:

Th e RP F i s not con s id e r e d a nu c l ear p owe r rea c tor but a production fa cili ty. Th e facilit y w ill not ha v e a ll o f th e sys t e m s d e t a iled in this s t a nd a rd and guid a n ce will be appli e d as a ppropri a t e. Application:

Use thi s guidance for development of c y bersecurity protections Compliance:

Design the FPC system and associated HMI based on this guidance IRO FS NR C PL C RAM RP F S DO E S IF S IL S I S SSC Y&Y it e m s r e li ed o n fo r safe t y. U.S. N u c l ear R eg ul a tory Co mmi ss ion. pro gr amm ab l e l og i c co nt ro ll er. ra di a ti o n a l arm m o n i t or. R a di o i so t o p e P ro du c ti o n Faci l ity. sec ur e d eve l op m e nt a nd operatio n a l e n v ir o nm ent. safe t y in strume nt e d fun c t io n. sa fety int egr i ty l eve l. sa f e t y in s tru me nt e d sys t e m. s tru c tur es , sys t e m s , a nd co mp o n e nt s. ve rifi ca ti o n a nd va lid a tion. Sp e cifi c r e quir e m e nt s w ill b e d eve l o ped durin g th e n e xt s t ages o f d es ign for th e Op e ratin g Licen s e A pplic a tion. The I&C d es ign will b e expand e d a nd a naly ze d to do c um e nt fulfillm e nt of th e de s ign c rit e ria a nd desi gn b as i s r e quirem e nt s for th e Op era tin g Lic e n se A pplic a tion. 7.2.3 System Description A s de s cribed in S ec tion 7.1 , th e RP F I&C s y s tem b as ic compon e nt s includ e th e FPC s y s t e m , ESF a ctuation system s, c ontrol con s ole a nd HMI di s pl a y in s trument s, a nd BMS. These s ystem s provide an interface for the operator to monitor and control tho s e s y s tem s. The FPC s y s tem will be a DCS that functions independ e ntl y. The it e m s relied on for sa fety (IROFS)/E SF safety function s will b e activated v ia hardwire (analo g) int e rlock s. 7-13

.

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

NDfllTlfWHT MEDICAL ISOTOPU NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.2.3.1 Facility Process Control System The FPC system controls and monitors the target fabrication system, hot cell area (e.g., Mo recovery and purification, uranium recovery and recycle system), process utility and support systems , and waste handling act i vities. The FPC system functions also include radiation monitoring , CAAS , HMls , safe shutdown control and initiation , supervisory information , and alarms. The BMS is a subsystem to the FPC system and monitors the faci lit y ve ntilati on system. The primary control location of the FPC system is in the control room. The contro l room FPC system operates with a standby redundant system structure. The standby workstations provide redundant hardware w ith identical PLC software systems as a utom atic backup control systems. The primary and backup PLC systems monitor each other. This backup control system minimizes the likelihood of downtime during Mo production processing.

7.2.3.2 Engineered Safety Feature Actuation Systems The operator will have direct visualizatio n of critica l values and the abi li ty to observe status of the features described in Table 7-13 (Section 7.4.1). The engineered safety feature actuation system dedicated displays will perform the following functions:

*

Static display -This display wi ll show critica l measurement values and perform the function of a n annunciator panel. This fixed display panel will not provide any interactive control functionality.

Alarm/event annunciator display panel -This panel will display any event or alarm that is defined for the process. The display will enab l e the operator to acknow l edge current events and a larm s , and wi ll provide a historical record of events. Dynamic interface display panel or HMI -This panel will enable the operator to perform tasks, change modes , enab l e/disable overrides , and other tasks that require o p erator input to a ll ow , perform , or modify a task or event. The set of displays wi ll be arranged in a workstat i on. This worksta tion will also include a keyboard and mouse that will be used to interface with the system. 7.2.3.3 Control Room/Human-Machine Interface Description The operator will have direct visualization of critica l va lu es and the ability to input control functions into the FPC system. The FPC system dedicated displays will perform the fo ll owing functions:

*

Static displa y -This display wi ll show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive contro l functionality. Alarm/event annunciator display panel -This panel wi ll display any event or alarm that is d efi n ed for the process. The display wi ll e nabl e the operator to acknow l edge current events and alarms , and w ill provide a historical record of events. Dynamic interface display panel or HMI -This panel will enable the operator to perform tasks , c han ge modes , ena bl e/disable overrides , an d other tasks that require operator input to a llow , perform , or modify a task or event. The set of displays will be arranged in a workstation.

This worksta tion will a l so include a keyboard and mouse that will be used to interface with the system. 7-14 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.2.3.4 Building Management System The BMS wi ll control the fac ili ty venti l ation system and receive indications from the fire protection, FPC , a nd process vessel ventilatio n systems. The primary purpose of t he BMS is to contro l the air balance of the facility vent il ation system and to sh ut down the fac ili ty ventilat ion system in the event of receiving an a l arm from the fire protection system or off-normal con diti ons indicated by the FPC. The operator will have direct visualization of critical values and the ab ili ty to input control functions into the BMS. The BMS dedicated displays wi ll perform the fo ll owing functions in the control room: * *

Static displa y -This display will show critica l measurement va lu es and perform the function of an annunciator panel. This fixed display pane l will not provide any interactive control functionality.

Alarm/event annunciator display panel -This panel wi ll di sp l ay any eve nt or alarm that is defined for the process. The display will enab l e the operator to acknowledge current eve nt s and a l arms, and will provide a historical record of events. Dynamic interface displa y panel or HMI -This panel w ill enable the operator to perform tasks , c h ange modes , enab l e/disable overrides , and other tasks that require operator input to a ll ow , perform , or modify a task or event. The set of displays will be arranged in a workstation. This workstation will also include a keyboa rd and mouse that will be used to interface with the system. 7.2.3.5 Fire Protection System The fire protection system will report the status of the fire protection e quipm ent to the central a l a rm station and the RPF contro l room with sufficient information to identify the genera l location a nd progress of a fire within the protected area boundaries.

Initiating devices for the fire detection and alarm subsystem, includ ing monitoring devices for the fire suppression su b system , will indi cate the presence of a fire within the facility.

Once an initiating device activates , signals will be sent to the fire alarm control panel. The fire alarm control panel will transmit signals to the central alarm stat i on and perform any anci ll ary functions.

As an example , signa l s from the fire control panel may initiate actions suc h as shutdown of the venti l at i on equipment or actuating the deluge valves. The fire protection system i s described in Chapter 9.0, Section 9.3. 7.2.3.6 Facility Communication Systems The RPF communication systems will re l ay information within the facil i ty during normal and emergency conditions.

The systems are designed to enable the RPF operator on duty to be in communication with the supervisor on duty, health physics staff, and other personnel required by the technical specificat ion s, 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 a ll operational areas and the control room. Faci li ty communications system is described in C h apter 9.0, Section 9.4. 7.2.3.7 Analytical Laboratory System The analytical l aboratory wi ll support the production of the Mo product and recycle of uranium. Sa mpl es from the process will be co ll ected, transported to the l a borat ory, a nd prepared in t h e l aboratory g l oveboxes and hoods , depending on the ana l ysis to be performed. The a n alytical l aboratory eq uipm ent wi ll be provided as vendor package units. Control room monitoring of the analytical laboratory wi ll be limit ed to the facility systems, including ventilation a nd r adiation monitorin g systems. Analytica l l aboratory system is described in Chapter 9.0, Section 9.7.3. 7-15 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.2.4 System Performance Analysis Th e RPF I&C sys t e m will moni to r th e proc esses a nd E SF s w h en r e quir e d. Th e IROF S w ill b e m a n age d b y th e F PC sys t e m. Th e FP C syste m w ill pro v id e th e ce ntr al d ec i s ion-m a kin g pro cesso r th a t eva lu a t es m o nit o r e d param eters fr o m th e va ri o u s pl a nt in st rum e nt a ti o n a nd fr om t h e ra di a ti on m o ni to rin g sys t e m s of t h e CA M s, CAAS , a nd RA Ms. The a n a l ys i s h e r e in di sc u sses safety as it r e l a t es t o th e I RO FS d es i gn c rit er ia a nd de s i gn b as i s. Pot e n t i a l va ri a ble s, condition s, o r o th e r it e m s th at w ill b e prob a ble s ubj ec t s of tec h nica l s p e cifi catio n s ass o c i ate d w ith th e RP F l&C sys t e m s are pro v id e d i n C h a pt er 1 4.0 , " T ec hni ca l S p ecifica tion s." 7.2.4.1 Facility Trip and Alarm Design Basis Th e d es ign ba s i s in fo rm a ti o n fo r t h e F P C s y s t e m trip fun ct i o n s i s b ase d o n t h e followin g t w o r e qu i r e ment s from Titl e I 0 , Cod e of F e d e r al R e gu l a ti ons , P a rt 70 (10 C FR 70), " Dom est i c Lic e n s in g o f S p ecia l Nucle a r M a t e ri a l." *

Double-contingenc y principle

-P rocess d es i gns s h o uld in co rp orate s u ffic i e n t fact or s of safety t o requir e at l eas t two un l ike l y , ind e p e nd e nt , a nd co n c urr e nt c h a n ges in pro cess co nditi o n s b efore a criti ca li ty acc id e nt i s p oss ible (b ase l i n e d es i g n c rit e ri a of 10 C FR 70.6 4 , " R e qu i r e ment s for N ew Facil i t i es or N ew Processes a t Ex i st in g Fac ili t i es ," p aragra ph [9]). Th e safety progra m will e n s u re th at eac h IRO FS wi ll b e ava il a bl e and r e li a bl e to perfo rm i ts intend e d fu n c ti o n w h e n nee d e d a nd in t h e co nt ex t of th e p e r fo rm ance r e quir e m ents o f thi s sectio n ( 10 CFR 70.6 1 , " P e r fo rm a n ce R e quir e m e nt s," p aragra ph [ e ]). Th e F PC s y s t em t ri p a nd a l a rm a nnun c i a tion a r e prot e cti ve functions a nd w ill b e p a rt of th e o vera ll p rotect ion a nd safety m o nit ori n g sys t e m s for th e RP F. The s p ec ific e qui p m e n t d es i gn b as i s for th e in stru m e nt a tion a nd e quipm e n t u se d for th e F P C system t ri p a nd a l a rmi ng fu n ctio n s i s d isc u sse d in Sectio n 7.2.2. T h e fo llowin g di sc u ss i on r e l a t es to t h e d es ign b as i s u se d for m o nit o ring s p ec ific s i gna l va lu es for RP F trips a nd a l a rm s , re quir e m e nt s for p e r fo rm a n ce , re quir e m e n ts fo r s p ec ifi c m o d es of o p era t ion o f th e RP F a nd t h e F P C system , a nd t h e ge n eral d es ign c ri te r ia not e d in T a bl e 7-1. 7.2.4.1.1 Safety Functions Corresponding Protective or Mitigative Actions for De s ign Basis Event s IEEE 603-2009 , IEEE Standard Criteria for Saf e ty Sy s tem s for N uclear Power Gen e rating Station s (Sections 4a and 4b). Th e res ul ts of th e int egra t e d safety a n a l ys i s (I SA) fo r th e RP F struc tur es , sys t ems , a n d com pon e n ts (SSC) a r e di sc u sse d in C h a p ter 1 3.0 , " Acci d e n t An a l ysis." Co nditi o n s t h at r e quir e m onitori n g and t h e s ub se qu e n t action t o b e take n are d escri b e d i n C h a p ter 13.0. 7.2.4.1.2 Variable Monitored to Control Prot e ctive or Mitigative Action IEEE 603-2009 (Section 4d). Th e li s t of va ri a bl es to b e m on it o r e d in th e RPF t o e limin ate o r r e du ce t h e ex p os ur e for th e opera tor w ill b e p rov id e d in t h e Op e ratin g L ice n se A ppli ca t io n. 7.2.4.1.3 Functional Degradation of Safet y S y stem Performance IEEE 603-2009 (Section 4h). Th ese d es ign r e quir e m e nts w ill b e fac tor e d in a nd w ill b e eva lu a t e d i n th e Op era tin g Lic e n si ng A ppli cat i o n. 7-16 NWMI ...*.. * *

NOITMWHT MEDtcAL ISOTOf'ES 7.2.4.2 Analysis NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.2.4.2.1 Facility Process Control System Trip Function Conformance to Applicable Criteria The FPC system will perform a trip as a protective function as part of the RPF safety analysis.

The associated design criteria are discussed in Sections 7.2.1 and 7.2.2. The following discussions relate to conformance to the criteria for the FPC system trip function. 7.2.4.2.2 General Functional Requirement Conformance IEEE 603-2009 (Section 5). The FPC system will initiate and contro l ESF activation and iso l ation , in a dditi on to the ab il ity of the ESF systems to perform the same , w h e n the system detects an off-normal event appropriate for activat ion. The FPC sys t em trips are discussed in Section 7.2.4. I. These monitored va l ues and s ub sequent trips are a result of the preliminary accident ana l ys i s in Chapter I 3 .0 and provide a means to mitigate or reduce the consequences from the design basis acc id ent to accepta bl e levels. 7.2.4.2.3 Requirements on B y passing Trip Functions Conformance IEEE 603-2009 (Sections 5.8 , 5.9 , 6.6, and 6.7). Trip override o r bypass is recognized as a design requirement.

Channel bypass will be a llow ed based on the nature of the signal. No c h annel bypass will be allowed w ith out a vis u a l indi cation o n the FPC sys t em display and recording the bypass eve nt in the hi storica l log. 7.2.4.2.4 Requirements on Setpoint Determination and Multiple Setpoint Conformance IEEE 603-2009 (Section 6.8). Table 7-1 discusses the criteria to be used for setpoi nt derivation. Setpoints wi ll be calculated in accordance with ISA-RP-67

.04.02 , Methodologies for the D e termination of Setpoints for Nuclear Safety-R elat e d Instrumentation. 7.2.4.2.5 Requirements for Completion of Trip Conformance IEEE 603-2009 (Section 5.2). The ESF and the interaction of a mitigative action going to comp l et ion wi ll be provided in the design. The F P C system will monitor for a co mpl ete trip of the ESF. Thi s information will be ava il ab le on the operator di splay for the FPC system and at the l oca l HMI terminals n ear the hot ce ll. An alarm/event annunciation will be di sp l ayed to the operator.

Sect i on 7.4.1 d escribes the activation of the ESF, a l arm/event strategy, and operator requirements to man u a ll y reset the system after a faci lity trip. 7.2.4.2.6 Requir e m e nts for Manual Control of Trip Conformance IEEE 603-2009 (Section 6.2). The FPC system wi ll h ave the ability to perform a manual activat i on of the ESF. Section 7.4.l describes the activat i on of the ESF, alarm/eve nt strategy, and operator r eq uir ements t o manually reset the system after a faci lit y trip. 7.2.4.3 Conclusion The I&C syste m s for the RPF w ill meet the stated design criteria and d esign basis requirements outlined in NUREG-1537, Guidelines for Preparing and R e viewing Applicatio n s for the Licensing of Non-Power R eactors -Format and Content. A crosswa lk of the I&C subsyste m s, a l ong with a cross-reference to specific design criteria , is presented in Table 7-2. 7-17

.; ... ;.*NWMI ...... ..* *.. .......... " "NOflll f WHTMlotCALISOTOflll NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages) Criteria* Design basis applicability Functional means IEEE 3 79

FPC syste m Single failure criterion

FPC sys tem di s pl ay IEEE 577 Reliability analysis criterion I EEE 603 Standard criteria safe t y system IEEE 384 Independence of Class lE equipment and circuits IEEE 323 Qualifying Class I E Equipment IEEE 344 Recommended practice for seismic qualification

FPC syste m IROFS end devices

ESFs

ESFs manual isolation

FPC system

FPC system display

FPC system IROFS end devices

ESFs

ESFs manual isolation

FPC system

F PC syste m di s pl ay

FPC syste m IROFS end d ev ice s

ESFs

ESFs manual i so l a tion

FPC system

FPC system display

FPC system IROFS end devices

ESFs

ESFs manual isolation

F P C syste m

FPC syste m di s pl ay

FPC sys t em IROF S end d evices

ESFs

ESFs manual i so lation

FPC system

FPC system display

FPC system IROFS end devices

ESFs

ESFs manual isolation IEEE 338

FPC sys tem Cri t e ria for the periodic

FPC sys tem di s pl ay survei ll a nce testing of

FPC sys t e m IROFS end d evices safety syste m s

ESFs

ESFs manual i so l a tion 7-18

Safety D CS pre ap pro ve d platform

R e dund ant independent isolation components

R e dundant o p erator interface workstations

R e dund ant se n sors

Alternative m a nual me a n s for ESF initiati on

Safety DCS pre-approved platform for an SIS Redundant in de pendent isolation components

Redundant operator interface workstations

Redundant sensors

Alternative manual means for ESF initiation See Section 7.3 for d eta il s.

IEEE 603 and IEEE 379 were used during development of the Construction Permit Application.

Additional details will be developed for the Operating Lic e nse Application.

Standard s upport s se lecti o n a nd qualifi catio n of e quipm e nt to be C la ss l E u se qu alified.

This standar d will b e reevaluated in th e Operatin g L i ce n se Application for a ppli ca bili ty.

Standard supports selection and qualification of equipment to be Class l E use qualified.

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

Standard su p po rt s se le ctio n of equipment; which r es ulted in the u se of ge n era l de sign c riteri a (prese nt e d in C h apte r 3.0) dur i n g development of th e Construction P e rmit Application.

S tand a rd wi ll be reevaluated in th e Operatin g License Application for a ppli ca bilit y.

NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages) Criteria* Design basis applicability Functional means IEEE 497

FPC system Criteria for accident

FPC system display monitoring instruments

FPC system IROFS end devices IEEE 7-4.3.2 Criteria for digital co mputer s in safety systems

ESFs

CAAS

RAMs

CAMs

FPC system

FPC system display

HMI displays IEEE 828

FPC system Configuration

FPC system display management in systems

HMI displays and software engineering IE EE 829 Software and system t es t documentation IEEE 1012 Criteria for software verification and validation I EEE 1028 Software re v iews a nd audits ANS-10.4 Verification and validation for safety software

FPC system

FPC system display

HMI di splays

FPC system

FPC system display

HMI displays

FPC system

FPC sys tem display

HMI di sp l ays

FPC system

FPC system display

HMI displays 7-19

Standard supports selection of accident monitoring equipment (e.g., radiation monitoring, annunciation), which resulted in the use of general design criteria (presented in Chapter 3.0) during development of the Construction Permit Application.

Standard will be reevaluated in the Operating License Application for applica bilit y.

Programming soft war e mu s t co mply with these cr it er ia a nd with the NWMI Software Quality Assurance Plan (prep a red durin g development of the Op erating License Application), which wi ll b e developed p er the d esign criteria outlined in C h a pter 3.0 a nd th i s standard.

Software and hardware u sed for the displays for the F P C system a nd HMI mu st a l so follo w guide lin es set fort h in this s tandard.

Standard will be r eeva lu a t ed i n th e Operating License Application for applicability.

Complies with IEEE 7-4.3.2 and the NWMI Software Quality Assurance Plan Standard will be reevaluated in the Operating License Application for applicability.

Co mpli es w ith IEEE 7-4.3.2 a nd th e NWMI Softwa re Quality Assurance Plan

Standard will b e r eeva lu a t ed in th e Operating License Application for a pplic a bilit y.

Complies with IEEE 7-4.3.2 and the NWMI Software Quality Assurance Plan

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

Co mpli es w i th IEEE 7-4.3.2 a nd th e NWMI So ftw a re Quality Assurance Plan

Standard w ill be r eeva lu a ted in the Operatin g License Applic a tion for a ppli ca bilit y.

Complies with IEEE 7-4.3.2 and the NWMI Software Quality Assurance Plan

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

NWM I ...**... * * *

NOllTMW'HT Mfotc:Al ISOTOl'U NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages) Criteria*

Design basis applicability Functional means

FP C sys t e m ANSI/I SA 67.04.01 S e tp o int s for nucl ea r safety-r e lated in s trum e nt s

FP C sys t e m IROF S e nd d ev i ces ANSI/ISA 84.00.01 , Part s 1, 2 , and 3 Functional safety: safety instrumented systems for the process industry sector

FPC s ystem

FPC system display

HMI displays NURE G-0700

FP C sys t e m Hu ma n-sys t e m

F P C sys t e m di s pl ay int e r fa c e d es i gn r eview

HMI di s pl ays gu id e lin es NUREG/CR-6463

FPC system Review guidelines on software languages for use in nuclear power plant safety systems NU REG/C R-60 9 0

FP C sys t e m PL C a nd applic a ti o n s in nu c l ea r re a ctor sys t e m s EPRI TR-l 06439

FPC system display Guideline on

HMI displays evaluation/acceptance of commercial grade digital equipment for nuclear safety applications 7-20

In c orpo ra t e d into o ve r a ll d es ign a nd th e C on s tru c t io n P e rmit A ppli ca tion.

Standard wi ll b e re eva lu a t e d in th e Op e r a ting L i ce n se A ppli ca ti o n fo r applicabilit

y.

Standard supports the design and development of non-safety-related systems that rely on safety , reliability , and functionality and was used during development of the Construction Permit Application.

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

St a nd a rd s upp o rt s th e de s i gn a nd d eve l o pm e n t of n o n-safety-re l a t e d sys t e m s th at p e rt ai n to c ontrol r o om a r ra n ge m e nt , s cr een d eve l o pm e nt s , a nd o p e rator int e r fa c e , a nd w as u se d durin g d ev elopm e nt of th e Co n s tru c tion P e rm it A pplic a tion.

St a nd a rd will b e r ee v a lu ate d in th e Op e rating L i ce n se A ppli ca ti o n for applicabili ty.

Standard supports the design , development , and review of safety-related software and was used during development of the Construction Permit Application.

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

Standard s upport s th e d es ign , dev e lopm e nt , and r ev i ew o f sa fet y-r e lat e d and nonsa fet y-r e l a t e d s oftw a re a nd was u s ed durin g d eve l o p me nt of th e Co n s tru c tion P e rmit Applic a t io n.

Sta nd a rd w ill b e r eeva lu ate d in th e Op e rating L i ce n s e A ppli cat ion for a ppli ca bi lity.

Standard supports the design , development , and review of safety-related systems that pertain to obtaining software or hardware for the FPC system , HMI displays , and data acquisition systems , and was used during development of the Construction Permit Application.

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

NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (4 pages) Criteria*

Design basis applicability Functional means R egu l atory Guide 1.1 52 C r iteri a for u se of computers in sa fet y systems

FPC system

FPC syste m display

HMI displays

FPC system

FPC system display Regulatory Guide 1.53 Single failure criterion evaluation for safety systems

FPC system IROFS end devices R egu l ato r y Guide 5.71 Cybersecuri t y programs for nucl ear fa c ilities

ESFs

ESFs manual isolation

FPC system

FPC system display

HMI display

Fu ll references are prov id e d in Section 7.7. CAAS criticality acci dent a l arm sys tem. CAM continuous air monitor. D CS digital contro l sys tem. ESF e n gineered s afety feature. FPC facility process co ntr o l. HMI hum an-mac hin e int erface. IR OFS NWM I PLC RAM S I S 7-21

Stan d ard supports the d es i gn and d eve l opment of redundant safe t y PL C platforms, FPC system r ed undant HMI works t atio n s, and operator interface workstations , and was u sed dur in g development of the Co n struct i on P ermit Ap pli catio n.

Sta nd ard w ill be reevaluated in the Operating License Application for ap plic ab ilit y.

Standard supports the design and development of high-integrity safety PLC s, redundant channels for ESFs , redundant operator interface workstations, redundant sensors, and alternative manual means for ESF initiation, and was u se d during development of the Construction Permit Application.

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

Cr it e ri a require th e d eve l opmen t of a d esign approach and impl eme nt ation for cy b ersecurity.

Stan d ard w ill be reevaluated in the Operating Lice n se Application for app li ca bilit y. item s relied on for safety. Northwest Medical I sotopes , LLC. programmable log i c contro ll er. radiat i on alarm monitor. safety in st rumented sys tem. 1 !

NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.3 PROCESS CONTROL SYSTEMS The process control systems for the RPF will include SNM preparation and ha n dling processes and radioisotope production processes. SNM preparation and handling processes i nc lude uranium recovery and recycle , and target fabrication. Radioisotope production processes include target receipt and disassembly, target dissolution, Mo recovery and purification, and waste handling. The RPF process contro l system includes interlocks (both hardwired

[ESF] and computer logic) to implement an automatic action on a parameter approaching or being outside its setting. Interlocks are defined as specific s et of conditions or parameters that need to be met for an activity to occur. An example of an interlock is the shutting down a pump on a tank high-level alarm signal or s witc hin g to a spare unit or process train based on a change in parameters (and corresponding a l arm). In additio n to interlocks , the RPF wi ll a l so implement a permissive philosophy that a ll ows HMI operations to be enabled o nc e the contro l room has confirmed the prerequisites conditions have been completed. Permissives differ from interlock s in that permissives require manua l approval via a switch (or similar) that must be satisfied for an acti v ity to occur. Interlocks are engineered features , a nd permissives are administrative features: The permis s i ve and interlocks wi ll be described in m o re detail in the Operating License Application.

The RPF process contro l wi ll be administered by the FPC system and is descri be d in Sect i on 7.2.3. The FPC system will p e rform the following hi gh-l eve l process functions. * *

Monitor the remote valve position for routing process fluid for int e r-equipment process fluid transfers

-For specific transfers identified by the operator , the FPC system wi ll provide a permissive to allow for the active pump in that circuit to be energized once the operator ha s m a nuall y configured the routing. Monitor and control inter-equipment process fluid transfers in the RPF -For transport requiring a pump , the FPC syste m will control the ability of the pump to be energized. For specific tran s fers, the FPC syste m will provide controlled fluid flow transfers based on a c loloop flow contro l. The operator will initialize the transfer of fluids. Other process fluid transfers , including:

Dissolved low-enriched uranium (LEU) solution to the Mo recovery and purification system Uranium sol ution to the uranium recovery and recycle system Liquid wastes to the waste handlin g system The l&C system for process utilities and support syste m s and for the ventilati o n systems wi ll be described in more detail in the Operating License Application.

The process systems described below provide for reliable co ntrol of the SNM prep a ration a nd handling process and the radioisotope production processes , and incl ud e: * * *

Range of operation of the sensor that is sufficient to cover the expected range of variat ion of the monitored variable during normal and transient process operation Reliable information about the status and magnitude of the proces s variab l e nece s sary for the full operating range of the radioisotope production and SNM recovery and recycle processes Reliable operation in the normal range of e n v ironm enta l co nditi ons anticipated within the facility Safe state during lo ss of electrical power Potential va riabl es , conditions, or other items that will be probable s ubj ects of technical specifications associated with the RPF process contro l systems are discussed in Chapter 14.0. 7-22 NWM I ...... ' *

NOflTMWHT MEOK:Al fSOTOf'll 7.3.1 Uranium Recovery and Rec y cle System NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems The uranium recovery an d recycle system will process raffinate from the Mo recovery and purification system for recycle to the target fabrication system. Two cycles of uranium purification will be included to separate uranium from unwanted fission products using ion exchange. The first ion exchange cycle will separate the bulk of the fission product contaminant mass from the uranium product. Product will ex it the ion exchange co lumn as a dilute uranium strea m that is concentrated to control the stored vo lum e of process so lution s. Urani um from the first cycle wi ll then be purified by a nearly identical seco nd cycle system to further reduce fission product contaminants to satisfy product criteria.

Eac h ion exc h ange system feed tank will include the capability of adding a reductant and modifying the feed chemical composition such that adequate separations are achieved, while minimizing uranium l osse s. Due to the variety of process activities performed during uranium recovery and recycle, the system description i s divided into the following s ub systems: * *

* * * * *

7.3.1.1 Impure uranium co ll ection Primary ion exchange Primary concentration Seco ndary ion exc h ange Seco nda ry conce ntration Uranium recycle Uranium decay and accountability Spent ion exchange resin Waste collection Design Criteria Design criteria for the uranium recovery and recycle I&C systems a r e described in Section 7.2. 7.3.1.2 Design Basis and Safety Requirements The design basis and safety requirements for the uranium recovery a nd recycle I&C systems are describ e d in Section 7.2. The ESFs for this sy s tem are listed in Chapter 6.0, " Engineered Safety Features." 7.3.1.3 System Description The uranium recovery and recycle I&C system will be defined in the Operating License Applicat ion. The strategy and associated parameters for the system are provided below. Pr eliminary proce ss seque nc es are provided in Chapter 4.0 to communicate the control s trategy for normal operations , which s ets the re q uirements for the process monitoring and control equipment , and the associated in strumentation. Normal operating functions will be performed remotely using the FPC system in the control room. Table 7-3 list s the anticipated control parameters , monitoring parameters , and primary control locati ons for each subsystem.

In addition , the implementation ofIROFS CS-14 , CS-15 , CS-20 , CS-27 , an d RS-JO interlocks for this system are under development.

Details of the control system (e.g., interlocks and permissive s ign als), nuclear a nd process in struments , contro l logic and e l ements, indication , ala rm, a nd control features will be developed for the Operating License Applicat ion. 7-23

.... :: .. NWMI ...... ........ *. ' " "NOmfWfST llfDtCAl ISOTOPU NWM l-2 013-021 , Re v. 3 Chapte r 7.0 -Inst ru men t ation and C ontrol Syste ms Ta bl e 7-3. Ur anium R e c ove r y and R ecy cl e Co n tr ol and M oni tori n g Par a m eters (2 pa ges) Subsystem Control parameters Primary control name (automatic/manual)

Monitoring parameters location Impure . Flowrate (A) . Density Contro l room uranium . Pump actuation (M) . Differentia l pressure collection . Pump motor speed (A) . Flowrate . Temperature (A) . Level . Valve actuation (AIM) . Pressure . Temperature . Valve position Primary ion . Flowrate (A) . Ana l yzer, uranium Control r oo m exc h ange Pu m p actuation (AIM) . D e n sity . Pu m p motor spee d (A) . D iffere nt ial pressure . Te mp e r ature (A) . F l owr a te . Va l ve ac tu atio n (AIM) . F l owra t e t ota l ize r . Leve l . Press ur e . Temperature . Va l ve p osition Primary . Density (A) . Ana l yzer, uranium Contro l room concentration Flowrate (A) . De n s i ty . Leve l (A) . Differential pressure . Pump actuation (AIM) . Flowrate . Pump motor speed (A) . Leve l . Temperature (A) . Pressure . Valve actuation (AIM) . Temperature . Va l ve position Seco nd ary i o n . Flowrate (A) . Analyzer, uranium Contro l room excha n ge Pu m p actua t io n (AIM) . D e n si t y . Pu m p m o to r spee d (A) . D iffe r entia l press ur e . Te mp e r a tur e (A) . F lo wra t e . Va l ve ac tu ation (AIM) . F lo wra t e tota li ze r . Leve l . Pressure . Tem p er a ture . Va l ve po sit i o n Secondary . Dens i ty (A) . Analyzer , uraniu m Contro l room concentrat i on . F l owrate (A) . Density . Level (A) . Differentia l pressure . Pump actuation (AIM) . Flowrate . Pump motor speed (A) . Leve l . Temperature (A) . Pressure . Valve actuation (AIM) . Temperature . Va l ve p osition U r ani um . F l o wra t e (A) . D ens it y Contr ol ro om r ecycle . Pum p actuatio n (AIM) . Di ffe r e n tia l press ur e . Pum p moto r s p ee d (A) . Flowra t e . Va l ve ac tu a t io n (AI M) . Leve l . Pr ess ur e . Te mp e ratur e . Va l ve po s it i on 7-24 NWMl-2013-0 21, R e v. 3 C h a p ter 7.0 -Ins trumen t at ion a n d Co n tr ol Sy s tems Tab l e 7-3. Uranium R ecovery a nd Recycle Co n t r o l and Mo n itoring Pa r ameters (2 pages) Subsystem Control parameters Primary control name (automatic/manual)

Monitoring parameters location Uranium d e ca y . Flowr a t e (A) . D e n s it y Co ntrol room a nd . Pump ac tuat io n (N M) . Di ffe r e nti al p res s u re a ccount a bilit y . Pump motor s p ee d (A) . Fl o wr a te . T e mp e ratur e (A) . Leve l . Va l ve a c tuati o n (N M) . P ress ur e . Te mp e ratur e . Va l ve po s iti o n Spent ion . Flowrate (A) . Analyzer , uranium Control room exchange resin . Pump actuation (NM) . Differential pre s sure . Pump motor speed (A) . Flowrate . Valve actuation (NM) . Level . Pressure . Valve position Was t e . Flowra t e (A) . D e n s it y Co n tro l room co ll ect i o n . P ump a ctuati o n (N M) . Di ffere ntial pr ess ur e . Pump motor spee d (A) . Flowra t e . Te mp era tur e (A) . Leve l . V al ve ac tuati o n (N M) . P ress ur e . Va l ve po s iti o n T a ble 7-4 pro v id es a pr e limina ry li sti n g of th e int e rlocks a nd p e rmi ss ive s i gna l s th a t ha ve b ee n id e ntifi e d. Th ese d ev ic es w ill b e furth er d eve l o p e d a nd d e t a il e d inform a ti o n w ill b e p rov id e d i n th e Op e ratin g Li ce n se A pp l i c ati o n. Table 7-4. Urani u m Re c ycle and Recover y S y s t em Inter l ocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock Impur e ur a nium co ll ec tion t a nk (UR-TK-JOO A) l ow-l eve l P LC N I A sw it c h (t yp i ca l of eig ht t a nk s) Impure uranium collection t a nk (UR-TK-lOOA) high-level PLC N I A switch (typical of eight tanks) Impur e ur a nium co ll ec t ion t a nk (UR-T K-IOOA) hi gh-P LC N I A t e mp e r a ture s w i tch (typ i cal o f ei g ht t a nk s) IX feed tank I (UR-TK-200) low-level s witch PLC N I A I X fee d ta nk I (UR-T K-2 00) hi g h-l eve l sw it c h PLC N I A IX feed tank 1 (UR-TK-200) h igh-temperature switch PLC N I A IX c ol u mn IA (U R-I X-2 40) hi g h-ur an ium a l a rm (AA H-252) P LC N I A IX column I A U so l ution fi l te r (UR-F-250) high-differential PLC N I A pressure a l arm I X col umn IA w as t e fi lt e r (U R-F-255) hi g h-diff e r e n t i a l P LC N I A pr ess ure a l a rm IX column lB (UR-IX-260) high-uranium alarm (AAH-272) PLC N I A I X column lB U s oluti o n fi l t e r (U R-F-2 70) h i g h-diff ere ntia l P LC N I A pr ess ure a l a rm 7-25

...**... * * *

NOITifWHT MEDICAl tSCJTOP(I NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock IX Column lB Waste Filter (UR-F-275) high-differential PLC N I A pre ss ure alarm Conce ntrator 1 feed tank (UR-TK-300) low-le vel switc h PLC N I A N I A N I A N I A N I A Concentrator 1 feed tank (UR-TK-300) high-level switch PLC Concentrator I (UR-Z-320) lo w-liquid l eve l alarm PLC Concentrator 1 (UR-Z-320) high-liquid level alarm PLC Co n ce ntrator 1 (UR-Z-320) d e mi ster hi g h-differenti a l pressure PLC a l a rm Concentrator 1 (UR-Z-320) conden s er high-differential PLC N I A pre ss ure alarm Concentrat or 1 (UR-Z-320) co nd enser hi g h-off gas temperature PLC N I A alarm Condensate sample tank IA (UR-TK-3 40) high-liquid level PLC N I A alarm Condensa t e samp l e tank I A (UR-TK-340) hi g h-ur an ium Hard-wired R e rout e co nd ensate tr a n sfe r to U R-TK-300 (pos ition V-396 , cl ose V-397) switc h (AE-35 6) Condensate delay tank I (UR-TK-370) high-liquid level a larm Co nd e n sa te sa mpl e tank I B (UR-TK-340) hi g h-liquid level a l arm Condensate sample tank lB (UR-TK-370) high-uranium sw itch (AE-386) I X feed tank 2A (UR-TK-4 00) l ow-l eve l sw itch IX feed tank 2A (UR-TK-400) high-level switch IX feed tank 2A (UR-TK-400) high-temperature switc h IX feed tank 2B (UR-TK-420) low-level switch IX fee d tank 2B (UR-TK-4 2 0) hi g h-l eve l sw itch IX feed tank 2B (UR-TK-420) high-temperature switch IX co lumn 2A (UR-IX-460) hi g h-ura nium alarm (AA H-472) IX column 2A U solution filter (UR-F-470) high-differential pressure alarm I X col umn 2A waste filter (U R-F-475) hi g h-differ e nti a l pr essure a larm IX column 2B (UR-IX-480) high-uranium alarm (AAH-492) 7-26 C lo se IX column eluent addition control va l ves (V-2 44 and V-264) PLC N I A PLC N I A Hard-wired Permissive to route condensate to WH-TK-420 (position V-496 , open V-397) Permissive to open IX column eluent addition control valves (V-244 and V-264) PLC N I A PLC NIA PLC N I A PLC N I A PL C N I A PLC N I A PLC N I A PLC N I A PL C N I A PLC N I A

.; .. NWMI ...... .. .. .

.. . * * *

NORTHWEST MfOK:Al ISOTO HS NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock IX column 2B U solution filter (UR-F-490) high-differential pressure alarm IX column 2B waste filter (UR-F-495) high-differential pressure alarm Concentrator 2 feed tank (UR-TK-500) low-level switch Concentrator 2 feed tank (UR-TK-500) high-level switch Concentrator 2 (UR-Z-520) low-liquid level alarm Concentrator 2 (UR-Z-520) high-liquid level alarm Concentrator 2 (UR-Z-520) demister high-differential pressure alarm Concentrator 2 (UR-Z-520) condenser high-differential pressure alarm Concentrator 2 (UR-Z-520) condenser high-offgas temperature alarm Condensate sample tank 2A (UR-TK-540) high-liquid level alarm Condensate sample tank 2A (UR-TK-540) high-uranium switch (AE-556) Condensate delay tank 2 (UR-TK-560) high-liquid level alarm Condensate sample tank 2B (UR-TK-570) high-liquid level alarm Condensate sample tank 2B (UR-TK-570) high-uranium switch (AE-586) Concentrate receiver tank (UR-TK-600) high-liquid level alarm Concentrate receiver tank (UR-TK-600) high-temperature alarm Product sample tank (UR-TK-620) high-liquid level alarm Product sample tank (UR-TK-620) high-temperature alarm Uranium rework tank (UR-TK-660) high-liquid level alarm Uranium rework tank (UR-TK-660) high-temperature alarm Uranium decay tank (UR-TK-700A) high-liquid level alarm (typical of 17 tanks) Uranium decay tank (UR-TK-700A) high-temperature alarm (typical of 17 tanks) 7-27 PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC Hard-wired PLC PLC Hard-wired PLC PLC PLC PLC PLC PLC PLC PLC N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A Reroute condensate transfer to UR-TK-500 (po s ition V-596 , close V-597) C lose IX column eluent addition control valves (V-464 and V-484) N I A N I A Permissive to route condensate to WH-TK-420 (position V-596 , open V-597) Permissive to open IX column eluent addition control valves (V-464 and V-484) N I A N I A N I A N I A N I A N I A N I A N I A NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Contro l Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock Uranium accountability tank (UR-TK-720) hi gh-liquid l evel a l arm Uranium accountability tank (UR-TK-720) high-temperature a l arm Spent resin tank A (UR-TK-820A) high-liquid l evel a l arm Spent resin tank A (UR-TK-820A) high-temperature alarm Spent resin tank B (UR-TK-820B) high-liquid le vel a larm Spent resin tank B (UR-TK-820B) high-temperature alarm Re s in transfer liquid tank (UR-TK-850) hi g h-liquid l eve l a l arm IX waste collection 1 tank (UR-TK-900) high-liquid level alarm IX waste co ll ection I tank (UR-TK-900) hi gh-t emperature a l arm IX waste collection 2 tank (UR-TK-920) high-liquid level alarm IX waste co ll ection 2 tank (UR-TK-920) hi gh-t emperature a l arm IX ion exc han ge. PL C programmable logi c co ntroller.

7.3.1.4 System Performance Ana l ysis and Conc lu sion PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A TBD to be d e termined. The system performance analysis and conclusion for eac h process system wi ll be provid ed in the Operating License App li cation. 7.3.2 Target Fabrication System The target fabrication system will produce LEU target s from fres h LEU mate ria l and recycled uranyl nitrate. The system will commence with the receipt of fresh LEU from the U.S. Department of Energy, and end with p ackagi n g new targets for s hipment to the univer s ity research reactor facilitie s. Due to the variety of process activities performed during target fa bricati on, the system de sc ription is divided into the following subsystems.

* * * * * * * *

Fresh uranium receipt and dissolution Nitrat e extractio n Ac id-d eficie nt uran y l nitrate (ADUN) concentration

[Proprietary Information]

Target fabrication waste Target assemb l y [Proprietary Informati on] New target handling 7-28 7.3.2.1 Design Criteria NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Design criteria for the target fabrication I&C systems are described in Section 7.2. 7.3.2.2 Design Basis and Safety Requirements The design basis and safety requirements for the target fabrication I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.2.3 System Description The target fabrication I&C system will be defined in the Operating License Application.

The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

Normal operating functions will be performed remotely using the FPC system HMI in the target fabrication area. Table 7-5 list s the anticipated control parameters , monitoring parameters, and primary control location for each subsystem.

In addition , the implementation of IROFS CS-14, CS-15, CS-20 , CS-27 , and RS-10 interlocks for this system are under development.

Details of the control system (e.g., interlocks and permissive signals), nucle ar and process instruments, control logic and e lements , indication, alarm , and control features will be developed for the Operating License Application. Table 7-5. Target Fabrication System Control and Mo nitorin g Parameters (2 pages) Subsystem name Fresh uranium receipt and . dissolution . (I 00-series tag numbers) . . . . . . Nitrate extraction . (200-series tag numbers) . . . . . .

. . ADUN concentrat ion . (300-series tag numbers) . . . . . . Control parameters (automatic/manual)

Current (A) Conductiv it y (A) Flow totalizer (A) Heater actuation (NM) Level (A) Pump actuation (NM) Temperature (A) Valve actuation (NM) Analyzer, pH (A) Contactor actuation (M) Flow totalizer (A) Flowrate (A) Level (A) Pump actuation (AIM) Pump motor speed (A) Temperature (A) Valve actuation (NM) Conductivity (A) Density (A) Flowrate (A) Leve l (A) Pump actuation (NM) Pump motor speed (A) Valve actuation (NM) 7-29 Monitoring parameters . Conduct i vity . Density . Differential pressure . Flowrate . Level . Pressure . Temperature . Analyzer, pH . Density . Differential pressure Flowrate . Level . Pressure . Pump motor speed . Temperature . Conductivity . Density . Flowrate . Leve l . Pressure . Temperature Primary control location Local Loca l Local

..*... * *

NOITHWHT lllDtcAl ISOTDPH NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-5. Target Fabrication System Control and Monitoring Parameters (2 pages) Subsystem name [Proprietary Information] . (400-series tag numbers) . . . . . [Proprietary Inform at ion] . (500-series ta g numb e r s) . . . . . [Proprietary Information] . (600-series tag numbers) . .

.

. . . T a r get fabrication waste . (700-series tag numbers) . . . . Target assembly [Proprietary Information]

New target handling Control parameters (automatic/manual)

Level (A) Pump actuation (AIM) Tank agitator actuation (AIM) Tank agitator speed (A) Temperature (A) Valve actuation (AIM) Flowrate (A) Pump actuation (AIM) Pump motor spee d (A) Temperature (A) Valve actuation (AIM) Vibrat ion dispersion asse mbl y actuation (M) Analyzer , hydrogen (A) Analyzer, oxygen (A) Flow totalizer (A) Level (A) Tank agitator speed (M) Temperature (A) Valve actuation (AIM) Flowrate (A) Level (A) Pump actuation (AIM) Pump motor spee d (A) Va l ve actuation (AIM) TBD TBD TBD ADUN acid-d eficie nt uran yl nitrate. TBD LEU = l ow-en ri c h ed uranium. Monitoring parameters . Flowrate . Level . Pressure . Temperature . Density . Differential pressure . Pre ss ure . Level . Temperature . Vibration . Analyzer , hydrogen . Analyzer, oxygen . Flowrate . Level . Pressure . Temperature . Density . Flowrate . Level . Pressure . Temperature TBD TBD TBD to be determined.

Primary control location Local Local Local Local Local Local Local Table 7-6 provide s a listing of the t a rget fabrication l&C system interlocks and permissive s ignals that h ave been identified. These devices will be further d eve loped a nd detailed information wi ll be pro v ided in the Operating License Application. 7-3 0 NWMl-2013-021 , Rev. 3 Chapter 7.0 -I nst rumentat ion and Contro l Systems Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages) Hard-wired or Interlock or permissive input PLC Safety interlock Dis s ol ve r column (TF-D-100) high-te mperature switch Uranium dissolution heat exchanger (TF-E-120) chilled water return high-conductivity switch Uranium dissolution heat exchanger (TF-E-1 20) lodifferential pressur e alarm Uranyl nitrate storage tank (TF-TK-200) level switch ADUN evaporator condenser (TF-E-350) chilled water return high-conductivity sw itch ADUN product heat exchanger (TF-E-360) differential pressure alarm ADUN product heat exc h a n ge r (TF-E-360) chilled wate r return high-conducti vity sw itch ADUN evaporator reboiler (TF-E-330) steam condensate high-conductivity switch ADUN storage tank (TF-T K-400) low-level switch ADUN storage tank (TF-TK-405) low-level switch ADUN storage tank (TF-TK-410) l ow-leve l switch ADUN storage tank (TF-TK-415) low-level switch ADUN s torage tank (TF-TK-400) hi g h-level switch ADUN storage tank (TF-TK-405) high-level switch ADUN st orage tank (TF-T K-401) hi g h-leve l switch ADUN storage tank (TF-TK-415) high-level switch [Propri e tary Inform atio n] (TF-TK-480) high-level swi tch [P r oprietary Information] (TF-C-500) high-temperature switch Silicone oi l heater (TF-E-550) outlet high-temperatur e swi tch [Proprietary Information] (TF-Z-660) high-temperature switch [Propriet ary Inform atio n] (TF-Z-661) high-t e mp erature sw itch [Proprietary Information] (TF-Z-662) high-temperature switch [Proprietary Inform at ion] (TF-Z-663) high-temperatur e switch [Proprietary Information] (TF-Z-660) door closed switch [Propriet ary Inform at ion] (TF-Z-661) door closed switc h 7-31 PLC Hard-wired PL C PLC Hard-wired PLC Hard-w ir e d Hard-wired PL C PLC PLC PLC PL C PLC PL C PLC PLC PLC Hard-wired Hard-wired Hard-wire d Hard-wired Hard-wired PLC PL C N/A Close chilled water return control valve (XV-12 2) on high conductivity N I A N I A C lo se chi ll e d water return control valve (HV-352) on hi g h conductivity N I A C l ose chilled water return control va l ve (HV-36 1) on hi g h conductivity Close steam condensate control valve (XV-333) on high conductivity N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A NIA N I A NIA N I A NIA N I A

NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-6. Target Fabrication System Interlocks and Permissive Signa l s (2 pages) Interlock or permissive input [Proprietary Information]

{TF-Z-662) door closed switch [Prop rietary In formatio n] {TF-Z-663) door closed sw it c h Reduction furnace offgas heat exchanger

{TF-E-670) outlet high-oxygen concentration R ed u ctio n furna ce offgas h eat excha n ge r {TF-E-670) outlet high-hydrogen co n centration Aqueous waste pencil tank {TF-TK-700) high-level alarm Aqueous waste p e n ci l tank {TF-TK-705) hi g h-l eve l a l a rm TCE tank (TF-TK-760) high-level s witch Target fabrication overflow tank {TF-TK-770) hi gh-hi gl evel sw it c h ADUN acid-deficient uranyl nitrate. PLC programmable logic contro ll er. Hard-wired or PLC Safety interlock PLC N IA PLC N I A PLC N IA PLC N I A PLC N IA PLC N I A PLC N IA PLC N I A TBD to be det e rmined. TCE tri c hl oroethy l ene. 7.3.2.4 System Performance Analysis and Conclusion The sys tem perform a nc e analysis and conclusion for each proc ess system wi ll b e pro v id ed in the Operating License Application.

7.3.3 Target Receipt and Disassembl y System The target receipt and di sassemb l y system will include th e delivery a nd r eceipt of the irradiated target cask, introduction of the irr a di ated targets into the hot cell, disassembly of the targets , a nd retrieval and transfer of the irradiated target material for processing. Thi s system wi ll feed the target dissolution syste m by the transfer of r ecove r ed irradiated target mat eria l through the dissolver I ho t ce ll (D S-EN-100) and dissolver 2 hot ce ll (DS-EN-200) isolation door int erfaces. Du e to the variety of ac ti v iti es performed durin g target r ecei pt a nd di sasse mbl y, the system de sc ription is di vi ded into the following s ub syste m s: * *

7.3.3.1 Cask r eceipt T a r get receipt T a rg e t di sasse mbl y Design Criteria D esign criteria for t h e tar get receipt a nd di sassem bl y I&C sys t e m s are de scri b e d in Sectio n 7.2. 7.3.3.2 Design Basis and Safety Req ui rements Th e design ba sis a nd safety r equire m e nt s for the target receipt a nd di sassem bl y I&C systems are d escri b e d in Section 7.2. The ESFs for this syste m a r e li sted in C hapt er 6.0. 7-32 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.3.3.3 System Description The target receipt and disassembl y I&C system wi ll be defined in t h e Operating License Application. The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the process monitoring and control equipment , and the associated in strumentation. Normal operating functions will be performed remotely using the FPC system HMI in the truck bay, cask preparation airlock , and the operating ga ll ery. Redundant control functions will be provided in the control room. In addition , the implementation ofIROFS CS-14 , CS-15 , CS-20 , CS-27 , and RS-10 interlocks for this sy s tem are under development.

Deta i l s of the contro l s ystem (e.g., interlocks and permissive signals), nuclear and process instruments , control logic and e l ements , indication , alarm, and control features will be developed for the Operating License Application. Pr i or to the start of disassembly operations , the following process control permissive signals will be required.

* *

Ventilation inside the hot cell is operable . Fission gas capture hood is on and functional.

Irradiated target material collectio n container is in position under the target c uttin g assem bly co ll ect ion bin. Waste drum transfer port is open and there is physical space to receive the waste target hardw are after disassembly and irradiated target material recovery.

The contro l parameters and monitoring parameters wi ll be defined during design development for the Operating License Application.

7.3.3.4 System Performance Analysis and Conclusion The system performance ana l ysis and conclusion for each proce s s system will be provided in the Operating License Application. 7.3.4 Target Dissolution System The target dissolution syste m process will receive the LEU target material from the target receipt and di s assemb ly system and dissolve the uranium and molybdenum-99 (99 Mo) in the solid irradiated target material in hot nitric acid. The concentrated uranyl nitrate solution will then be transferred to the Mo recovery and purification system for further processing.

The target dissolution process will b e operated in a [Proprietary Information]

transferred to a co ll ection container. The collection container wi ll move through the pass-through to a dissolver ba sket positioned over a dissolver , the target material will then be dissolved a nd the r es ultin g solution transferred to the Mo recovery and purification system. Target dissolution of irradiated LEU will result in gaseo u s fission products (iodine [I], krypton [Kr], and xenon [Xe]) with very high radiation fields. A primary function of the process offgas systems wi ll be to control release of these gases both inte rnal and external t o the facility. The dissolver off gas treatment system wi ll include the nitrogen oxide (NO x) tre a tment and fission gas treatment subsystems.

7-33

.. ;. NWMI *: :**::* ...*.. "NOftTHW£nMflMCALISOTOPU NWMl-2013-021, Rev. 3 Chapter 7.0 -I nstrumentation and Contro l Systems Due to the variety of process activities performed during target dissolution , the system description is divided into the following sub sys tems: * * * * *

7.3.4.1 Target dissolution 1 and target dissolution 2 NO x treatment 1 or NO x treatment 2 Pressure relief Primary fission gas treatment Secondary fission gas treatment Waste collection Design Criteria Design criteria for the target dis s olution I&C systems are described in Section 7.2. 7.3.4.2 Design Basis and Safety Requirements The design basis and safety requirements for the target dissolution I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.4.3 System Description The target dissolution I&C system will be defined in the Operating License Application. The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the process monitoring and control equipment , and the associated instrumentation. Loading of [Proprietary Information]

into the dis s olver will involve mechanic a l handling o f the transfer containers. Operators using remote in-cell cranes and manipulators will perform these functions. Other normal operating functions will be performed remotely using the FPC system H MI in the operating gallery. Redundant control functions will be provided in the control room. Table 7-7 lists the anticipated control parameters , monitoring parameters , and primary control locations for each subsys t em. Details of the control system (e.g., interlocks and permissive signals), control logic , indication , alarm , and control features will be defined in the Operating License Application. 7-34 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control S y stems Table 7-7. Target Dissolution System Control and Monitoring Parameters Subsystem name Ta r get dissolutio n 1 and 2 NOx treatment 1 o r 2 Pressure r e li ef Primary fission gas treatment Secondary fission gas treatment Waste collection . . . .

. . . . . . . . . . . . . . . . . . . Control parameters (automatic/manual)

Dissolver agitator actua ti on (AIM) Dissolver agitator spee d (A) Flowrate (A) Pump actuation (AIM) Pump motor speed (A) Temperature (A) Va l ve actuation (AIM) Flowrate (A) Pump actuation (AIM) Pump motor s peed (A) Temperature (A) Valve actuation (AIM) Pump actuation (AIM) Pump m otor speed (A) Temperature (A) Va l ve actuat i on (AIM) Temperature (A) Valve actuation (AIM) Va l ve actuation (AIM) Pump actuation (AIM) Pump motor Speed (A) Temperature (A) Valve act uation (AIM) NO x nitrogen oxide. Monitoring parameters . Di sso l ver ag it ato r speed . Flowrate . F l owrate totalizer . Level . Pressure . Radiation . Temperature . Va l ve position . Differential pressure . Flowrate . Flowrate totalizer . Level . Pres s ure . R a diation . Temperature . Valve po sition . Flowrate . Level . Pressure . Va l ve position . Differential pressure . Flowrate . Pre ss ure . Radiation . Temperature . Valve po sition . Differential pressure . F l owrate . Pressure . Radiation . Temperature . Va l ve position . Differential pressure . Flowrate . Level . Temp erature . Pre ss ure . Radiation . Valve position 7-35 Primary control location Operating ga ll ery Operating gallery Operati n g gallery Operating gallery Operating ga ll ery Operating ga ll ery

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

NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-8 provides a preliminary listing of the target di sso lution I&C system interlocks and permissive signals that ha ve been identified. In addition, the implementation ofIROFS CS-14, CS-15 , CS-20 , CS-27, and RS-10 interlocks for this system are under development.

These devices will be further de veloped and detailed information will be provided in the Operating License Application.

Table 7-8. Target Dissolution System Interlocks and Permissive Signals (2 pages) Hard-wired or Interlock or permissive input PLC Safety interlock Dissolver 1 (DS-D-100) high-liquid l eve l alarm PLC N I A Dissolver 1 (DS-D-100) low-liquid level alarm PLC N I A Dissolver 1 (DS-D-100) hi g h liquid temperature alarm PLC N I A Dissolver 1 Condenser (DS-E-130) high gas temperature alarm PLC N I A Dissolver 2 (DS-D-200) high-liquid l eve l a larm PLC N I A Dissolver 2 (DS-D-200) low-liquid level alarm PLC N I A Dissolver 2 (DS-D-200) high liquid temperature alarm PLC N I A Dissolver 2 condenser (DS-E-230) high gas temperature alarm PLC N I A Primary ca u st ic scrubber I (DS-C-310) high-liquid l eve l alarm PLC N I A Caustic scrubber I (DS-C-310) high gas temperature PLC N I A NO x oxidizer 1 (DS-C-340) high-liquid l eve l alarm PLC N I A NOx oxidizer 1 (DS-C-340) high gas temperature PLC N I A NO x absorber I (DS-C-370) high-liquid l evel a l arm PLC N I A NOx absorber 1 (DS-C-370) high gas temperature PLC N IA Primary ca u stic scrubber 2 (DS-C-410) hi gh-liquid level alarm PLC N I A Caustic sc rubber 2 (DS-C-410) high gas temperature PLC N I A NO x oxid i zer 2 (DS-C-440) high-liquid l eve l alarm PLC N I A NOx oxidizer 2 (DS-C-440) high gas temperature PLC N I A NO x absorber 2 (DS-C-470) high-liquid l eve l a l arm PLC N I A NOx absorber 2 (DS-C-470) high gas temperature PLC N I A Pressure relief tank (DS-TK-500) high-pressure alarm Hard-wired Opens va l ve to capture dissolver gases Pressure relief tank (DS-TK-500) high-liquid level alarm PLC N IA Pressure relief tank (DS-TK-500) low-liquid l evel alarm PLC N I A Dryer A (DS-E-610A) high gas temperature alarm PLC N I A Primary adso rb er A (DS-SB-620 A) high gas temperature a l arm PLC N I A Filter A (DS-F-630A) high-pres sure differential alarm PLC N I A Dryer B (DS-E-610B) high gas temperature alarm PLC N I A Primary adsorber B (DS-SB-620B) high gas temperature alarm PLC N I A Filter B (DS-F-6 30B) hi g h-pre ssure differential alarm PLC N I A Dryer C (DS-E-610C) high gas temperature alarm PLC N I A Primary adso rb er C (DS-SB-620 C) high gas temperature a l arm PLC N I A Filter C (DS-F-630C) high-pressure differential alarm PLC N I A 7-36

.... ;. NWMI .... ** .... .......... * * ' NOATifWEIT MEOtcAl ISOTOPlS NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-8. Target Dissolution System Interlock s and Permissive Signals (2 pages) Hard-wired or Interlock or permissive input PLC Safety interlock Secondary adsorber A (DS-SB-730A) high gas temperature alarm Secondary adsorber B (DS-SB-730B) high gas temperature a larm Secondary adsorber C (DS-SB-730C) high gas temperature a l arm Waste collection and sa mpling tank 1 (DS-TK-800) high-liquid l evel alarm Wa ste collect ion and sa mpling tank I (DS-TK-800) hi g h-liquid temperature a l arm Waste collection and sam pling tank 2 (DS-TK-820) high-liquid level alarm Wa s te collection and sa mpling tank 2 (DS-TK-820) high-liquids temperature a l arm N I A NO x not a ppli cab l e. = nit rogen ox id e. PL C 7.3.4.4 System Performance Analysis and Conclusion PL C PLC PL C PLC PLC PLC PL C N I A N I A N I A N I A N I A N I A N I A programmable l ogic controller.

The system performance ana l ysis a nd conclusion for each process system will be provided in the Operating License App li cation. 7.3.5 Molybdenum Recovery and Purification System The Mo recovery and purification sys tem wi ll receive the impure Mo/uranium so lution from the target dissolution system into feed tank IA a nd feed tank lB (MR-TK-100 an d MR-TK-140) located in the tank hot cell. The Mo/uranium so lu tion wi ll then be transferred to process hot cells and processed through three se parate ion excha ng e unit operations to achieve the d esired product criteria. A collection container holding the separated an d purified Mo product material will b e used for final chemica l adjustment and sampling for verification of batch acceptance.

The product will be sa mpl ed and weighed , placed in stainless stee l bottles w ith lid s applied and tightened , loaded into s hi e ld ed conta iner s, a nd then s hipp ed in an approved cask. Due to the variety of activities performed during Mo recovery and purification , the system description is divided into the following s ub systems: * * *

7.3.5.l Primary ion exchange Secondary ion exc han ge Tertiary ion exchange Mo product Design Criteria De sign criteria for the Mo recovery an d purification I&C systems are de scribed in Section 7.2. 7.3.5.2 Design Basis and Safety Requirements The design basi s and safety requirements for the Mo recovery and purification I&C systems are described in Section 7.2. The ESFs for this syste m are li sted in Chapter 6.0. 7-37 7.3.5.3 System D esc ripti o n NWMl-201 3-021 , R ev. 3 C hap t e r 7.0 -Ins tr u m en t ation a n d C ontrol Sy s te m s The Mo recovery and p u rification I&C system wi ll be defined in the Operating License App l ication. The strategy and associated parameters for the I&C system are provided below. Pre l iminary process sequences are provided in Chapter 4.0 to communicate the control strategy for n orma l operations , which sets the requ i rements fo r the process mon i toring and contro l equipment, and the associated i nstrumen tati on. Operators u sing remote in-ce ll manipu l ators will perform the product transfer and packagi ng func t ions. All other normal operating functions wi ll be performed remotely using the FPC system HMI in the operating ga ll ery. Red u ndant contro l functions wi ll be provided in the control room. Tab l e 7-9 lists the anticipated control parameters, monitoring parameters , and primary control locations for each subsystem.

In addition, the implementation ofIROFS CS-14 , CS-15, CS-20, CS-27, and RS-10 interlocks for this system are u n der development.

Detai l s of the contro l system (e.g., i nterlocks and permiss i ve signa l s), nuclear and process ins t rument s, contro l logic and e l ements, indication, alarm , and control features will be deve l oped for the Operating Lice n se Application.

Tab l e 7-9. Mo l y bd e num R ecovery a nd P urifi ca ti o n Sys t em C on t rol a nd M oni to ring P aram e t ers Subsystem name Primary ion exchange . . Secon d ary io n exchange . Tertiary io n exchange . Molyb d e n um p roduct . Control parameters (automatic/manual)

Temperature (A) Valve actuation (AIM) Pumps (M) Pumps (M) Actuate cap p i n g unit (M) . . . .

. . . . . . . . . . Monitoring parameters Density Flowrate Le ve l Temperature Pre ssure Radiation Valve position Temperature Densit y Flowrat e Level Pressure Temperature Weig h t Primary control location Operating gallery O p erating ga ll ery O p erating ga ll ery Operating ga ll ery Table 7-10 provides a p r e li minary l i sting of the Mo recovery and purification system interlocks and permissive signals that have b een ident i fied. These device s will be further deve l oped and detai l ed information wi ll be provided in the Operating License App l icat i on. 7-3 8 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-10. Molybdenum Recovery and Purification System Interlocks and Permissive Signals Interlock or permissive input Feed tank IA (MR-TK-100) high-liquid level alarm Feed tank IA (MR-TK-100) lo w-liquid l evel alarm Feed tank IA (MR-TK-100) high-temperature alarm Feed tank I A (MR-TK-100) high-pressure alarm Feed tank lB (MR-TK-140) high-liquid level alarm Feed tank lB (MR-TK-140) l ow-liquid level a l arm Feed tank lB (MR-TK-140) high-temperature alarm Feed tank IB (MR-TK-140) hi gh-pressure a l arm U solution collection tank (MR-TK-180) high-liquid level alarm U solution collection tank (MR-TK-180) low-liquid level a l arm U solution collection tank (MR-TK-180) high-pressure alarm Waste co ll ec tion tank (MR-TK-340) high-liquid l evel a l arm Waste collection tank (MR-TK-340) low-liquid level alarm Waste collection tank (MR-TK-340) high-pressure alarm N I A not applicab l e. U PLC = programmable l og i c c ontroller.

7.3.5.4 System Performance Analysis and Conclusion uranium. Hard-wired or PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC The system performance ana l ysis and conclusion for each process system will be provided in the Operating License App lication.

7.3.6 Waste Handling System N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A NIA N I A N I A N I A The waste handling system will consist of storage tanks for accumu latin g waste liquids and adjusting the waste composition, and the equipme nt n eeded for handling and encapsulating so lid waste. Liquid waste will be spli t into high-dose and low-dose streams by concentration.

The high-dose fraction w ill be further concentrated and adjusted.

Liquid waste will then be mixed with an adsorbent m ater i a l. The so lid waste streams w ill b e placed in a waste drum and encap s ulated b y adding a cement material to fill voids remaining within the drum. All high-dose waste streams will be held for decay and shipped to a disposal facility. Due to the variety of activit i es performed during waste handling , the system description is divided into the following subsystems

* * * * * * * *

High-dose liquid waste collection Low-do se liquid waste collection Low-dose waste evaporation High-dose liquid waste solidificatio n Low-dose liquid waste solidificatio n Spent resin dewatering Solid waste encaps ulation High-dose waste decay High-dose waste handlin g 7-39 7.3.6.1 Design Criteria NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems D esign criteria for the waste h and lin g I&C systems are d escribed in Section 7 .2. 7.3.6.2 Design Basis and Safety Requirements The design basis and safety requirements for the waste handlin g I&C systems a r e de scribed in Section 7.2. The ESFs for this system are li sted in C hapt er 6.0. 7.3.6.3 System Description The waste handlin g I&C syste m wi ll be defin ed in the Operating License Application. The strategy and associa ted param eters for the l&C sys tem are provided b e l ow. Preliminary process seq uenc es are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the proc ess monitoring and control e quipm ent, and the associ a ted instrumentation.

All normal operating functions for low-dos e liquid so lidific at ion will be controlled loc a ll y using HMi s in the low-dose waste room (Room Wl07). A lo ca l HMI di sp la y area will b e provided in thi s room for most waste handling operations. All normal operating function s for the hi g h-do se liquid wa s te so lidification , hi g h-do se waste decay , s p e nt resin d ewater in g, an d so lid waste handling hot cell operations will be controlled and/or monitored from the low-do se waste room (Room Wl07). Liqu id waste co ll ect ion and lo w-dose liquid waste evaporation operations wi ll be co ntroll e d from the RP F control room. Table 7-1 I li sts the anticipated co ntrol parameters , monitorin g p ara m eters , a nd primary contro l lo catio n s for eac h s ub syste m. In ad dition , the impl e m e ntation of IROF S CS-I 4, CS-15 , CS-20, CS-27 , and RS-I 0 int er locks for thi s syste m are und er d eve lopm ent. D eta il s of the control system (e.g., interlocks and permi ss i ve s ign a l s), nucl ea r a nd process in s trum e nt s , control lo g ic and elements , indicati o n , alarm, and control features will be d eve lop ed for the Operating License Application. 7-40 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-11. Waste Handling System Control and Monitoring Parameters Subsystem name High-dose liquid waste . collection High-dose liquid waste . solidification Low-dose liquid waste . collection . .

.

. Low-dose liquid waste . evaporatio n .

. . . Low-dose liquid waste . solidification . . . . Spent resin dewatering . Solid waste . encapsu l atio n High-dose waste decay High-dose waste h and lin g TBD = to be determined.

Control parameters (automatic/manual)

Valve position Valve po s ition Flowrate (A) Pump actuat ion (AIM) Pump motor speed (A) Temperature (A) Valve actuat ion (AIM) Flowrate (A) Pump actuation (AIM) Pump motor speed (A) Temperature (A) Valve actuation (AIM) F l owrate (A) Pump actuation (AIM) Pump moto r speed (A) Temperature (A) Valve actuation (AIM) Valve actuation (AIM) Actuate grout mixer (M) TBD TBD Monitoring parameters Primary control location . Density Contro l room . Differential pressure . Flowrate . F l owrate totalizer . Level . Temperature . Pressure . Radiation . Valve position . Density Low do se so lidification room . Differential Pressure . Flowrate . Flowrate totalizer . Level . Temperature . Pre ssure . Radiation . Valve Position . Density Contro l room . Differential pressure . F l owrate . F l owrate totalizer . Level . Temperature . Pressure . Valve position . Differential pressure Control room . Flowrate . Level . Temperature . Pressure . Valve position . Density Low dose so li dificatio n room . Differential pressure . Flowrate . F l owrate totalizer . Level . Temperature . Pressure . Va l ve position . Valve position Low dose solidification room . Pressure Low dose solidificatio n room TBD Low dos e solidification room TBD Low dose so l idification room 7-41

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NOftntWEST MEDtcAl ISOTOPE S NWMl-2013-021, Rev. 3 Chapter 7.0 -I nstrumen t ation and Control Systems Table 7-12 provides a preliminary listing of the waste handling system interlocks and permissive signals that have been identified.

These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-12. Waste Handling System Interlocks and Permissive Signals Hard-wired or Safety Interlock or permissive input PLC interlock High-dose waste collection tank (WH-TK-100) high-liquid level alarm PLC N I A High-dose waste collection tank (WH-TK-100) low-liquid level alarm PLC N I A High-dose waste collection tank (WH-TK-100) low-pressure alarm PLC N I A Hi g h-dose waste concentrator (WH-Z-200) high-liquid level alarm PLC N I A High-dose waste concentrator (WH-Z-200) low-liquid level alarm PLC N I A High-dose waste concentrator (WH-Z-200) demister high-differenti a l pressure PLC N I A alarm High-dose waste concentrator (WH-Z-200) condenser high-differential pressure PLC N I A alarm High-dose waste concentrator (WH-Z-200) condenser offgas hi g h-temperature PLC N I A alarm Low-dose waste collection tank (WH-TK-240) high-liquid level alarm PLC N I A Low-dose waste collection tank (WH-TK-240) low-liquid level alarm PLC N I A Low-dose waste collection tank (WH-TK-240) low-pressure alarm PLC N I A High-do se waste container offgas filter (WH-F-330) high-pressure differential PLC N I A alarm Condensate collection tank (WH-TK-400) high-liquid level a l arm PLC N I A Condensate collection tank (WH-TK-400) low-liquid level alarm PLC N I A Condensate collection tank (WH-TK-400) low-pressure alarm PLC N I A Low-dose waste collection tank (WH-TK-420) high-liquid level alarm PLC N I A Low-dose waste collection tank (WH-TK-420) low-liquid level alarm PLC N I A Low-dose waste collection tank (WH-TK-420) low-pressure alarm PLC N I A Low-dose waste evaporation tank I (WH-TK-500) high-liquid level alarm PLC N I A Low-dose waste evaporation tank 1 (WH-TK-500) low-liquid level alarm PLC N I A Low-dose waste evaporation tank 1 (WH-TK-500) low-pressure alarm PLC N I A Low-dose waste evaporation tank 2 (WH-TK-530) high-liquid le ve l alarm PLC N I A Low-dose waste evaporation tank 2 (WH-TK-530) low-liquid level alarm PLC N IA Low-dose waste evaporation tank 2 (WH-TK-530) low-pressure alarm PLC N I A Low-dose waste container offgas filter (WH-F-630) high-pressure differential PLC N I A alarm PL C = pro gramma ble lo gic contro ller. TBD = to be determined.

7-42

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NOITlfWUT MEDICAL ISOTGnS NWMl-2013-021 , Rev. 3 Chap ter 7.0 -Instrumentation and Control Systems 7.3.6.4 Sys tem Performance A nal ys is and Conclusion The syste m perform a nce analysis and conc lu sion for each process system will be provided in the Operating License Application. 7.3.7 Criticality Accident Alarm System The RPF wi ll u se a CAAS to monitor for a criticality and provide emergency notifications for evacuation.

7.3.7.1 Design Criteria Design criteria for the CAAS I&C systems are d esc ribed in Section 7.2. 7.3.7.2 Design Basis and Safety Requirements The de s ign basis and safety requirements for the CAAS l&C systems are de sc ribed in Section 7.2. 7.3.7.3 System Description The CAAS will be provide d as a vendor package wit h an integrated control syste m. The CAAS control HMI will be locat ed in the control room and will provide lo ca l alarms at the detector loc ations an d at the CAAS HMI. The FPC system will provide a l arm and s tatu s monitoring in th e contro l room. The facility-wide notification system configuration will be provided in the Operating License Application. The survei ll a n ce requirements for the CAAS system are described in Chapter 6.0. 7.3.7.4 System Performance Analysis and Conclusion Th e system performance a nal ysis for eac h process system will be provid e d in the Operating License Application. The overall I&C system performance a nalysis is discussed in Section 7.2. The CAAS wi ll provide for continuous monitoring , indi cation, and recording of neutron or gamma radiation l eve l s in areas where personnel m ay be pre se nt and wherever an accidenta l criticality event could result from operational proces ses. The CAAS wi ll be capable of detecting a criticality accide nt that produces an absorbed dose in soft tiss u e of 20 radiat ion absorbed dose (rad) of combined neutron or gamma radiation at an unshielded distance of 2 meters (m) from the reacting material within 1 minute (min), except for events occurring in areas not normally accessed by personnel and w h ere shielding provide s protection aga in st radiation generated from an accidental critica lity. Two detectors wi ll cover each area needing CAAS coverage. The control unit electronics will actuate lo ca l and remote alarm s. The loc ations of the detectors will be provided in the Operating License Application. The CAAS detector s wi ll provide local annunci a tion and remote annunciation in the control room to a larm when the radiation levels exceed estab li shed setpoints. Alarming CAAS monitors will communicate the l ocatio n of the criticality accident alarm to the FPC system. Diagrams of the CAAS and associated systems will be provided in the Operating License Application.

The uninterruptible power s uppl y (UPS) will provide emergency power to the CAAS during a lo ss of off-site power. The CAAS wi ll meet the criteria of 10 CF R 20.150 1 , " General ," and use the gui d a n ce provided by ANSI/ANS 8.3, Criti ca li ty Accident Alarm System, and Regulatory Guide 3.71 , N ucl e ar Criticality Safety Standards for Fuels and Mat e rial Facilities.

As a safety-related system, the CAAS will be designed to remain operational during d esign basis accidents , w hich are described in Chapter 13.0. 7-43 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS 7.4.1 System Description The ESFs are active or passive features designed to mitigate the consequences of accidents and to keep radiological exposures to workers, the public, and environment within acceptable values. Chapter 6.0 provides a description of the ESFs, including the accidents mitigated and SSC s used to provide the ESFs. The ESF systems will operate independently from the FPC systems as hard-wired controls. However, the ESFs will integrate into the FPC systems and provide a common point ofHMI , monitoring , and alarming at the control room and local HMI workstations.

Table 7-13 lists the ESFs that will require actuation by the I&C system. Moni t oring systems that are credited in the safety analysis are also included in the table. Table 7-13. Engineered Safety Feature Actuation or Monitoring Systems (2 pages) l&C SSCs providing Engineered safety feature IROFS Accident(s) mitigated engineered safety feature Primary offgas relief system RS-09 Dissolver offgas failure during Pressure relief device , pressure dissolution operation relief tank Active radiation monitoring RS-10 Transfer of high-dose process Radiation monitoring and and isolation of low-dose liquid outside the hot cell isolation system for low-dose waste transfer shielding boundary liquid transfers Cask local ventilation during RS-13 Target cladding leakage during Local capture ventilation closure lid removal and shipment system over closure lid during docking preparations lid removal Cask docking port enabler RS-15 Cask not engaged in the cask Sensor system controlling cask docking port prior to opening the docking port door operation docking port door Process vessel emergency FS-03 Hydrogen deflagration or Backup bottled nitrogen gas purge system detonation supply Active discharge monitoring CS-14 Accidental criticality To be provided in the Operating and isolation License Application Independent active discharge CS-15 Accidental criticality To be provided in the Operating monitoring and isolation License Application Evaporator or concentrator CS-20 Prevent nuclear criticality from Conductivity analyzer and condensate monitoring high-volume transfer to non-control valve geometrically favorable vessels in solutions with normally low fissile component concentrations 7-44

NWM I *

0 NOllTHWUT llUNCAl ISOTOl'lS NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Table 7-13. Engineered Safety Feature Actuation or Monitoring Systems (2 pages) l&C SSCs providing Engineered safety feature IROFS Accident(s) mitigated engineered safety feature Closed heating or cooling loop wit h monit or ing a nd alarm Dissolver offgas vacuum receiver or vacuum pump CS-27 TBD I&C IRO FS instrumentation and control. it ems relied o n for safety. 7.4.2 Annunciation and Display Accidental criticality Potential limiting control for operations; motive force for dissolver offgas C lo se d-loop , hi gh-volume h eat transfer fluid systems to pre vent nuclear criticality or transfer of high-dose material across s hi e lding bound ary in the event of a leak into the heat transfer fluid with norm a lly low fi ss il e compo nent concentra tion s Dissolver offgas vacuum receiver tanks, dissolver offgas vacuum pumps SSC TBD structures , systems , a nd compo n e nt s. to be determined. The actuation of an ESF will be displayed on the FPC system HMI and loc ally at the affected system with an audible alarm. The alarm annunciator display panel and the a l a rm or event display will s how the triggering event. Once actuated, the ESFs will require manu a l input from the operator to reset the ESF. Clearing the triggerin g event will be required.

7.4.3 System Performance Analysis Section 7.2.4 pro vi d es additional details on the ana ly s i s of syste m performance.

Potential varia ble s, conditions, or other items that will be probable su bject s of t echnica l specifications associated with the FPC syste m are provided in Chapter 14.0. 7-45 NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumen t ation and Control Systems 7.5 CONTROL CONSOLE AND DISPLAY INSTRUMENTS 7.5.1 Design Criteria Design criteria for the co ntrol room I&C systems are described in Section 7.2. 7.5.2 Design Basis and Safety Requirements The design basis and safety requirements for the contro l room I&C systems are described in Section 7.2. 7.5.3 System Description The control room will provide the majority of interfaces for the facility and process control systems, with overall process controls , monitoring, alanns, and acknow l edgement.

The contro l room will consis t of a properly sized and shaped control console with two or three operator interface stations or HMis (one being a dedicated engineering interface), a master PLC or distributed controlle r , and all related and necessary cabinetry and subcomponents (e.g., input/output boards, gateways, Et h ernet switc he s, power supplies, and UPS). This control system will be supported by a data hi g h way of se n si n g instrument signals in the faci lity process areas that will be gathered onto the highway throughout t he faci li ty by an Ethernet communication-based interface backbone and brought into the control room and onto the console displays. Dedicated controllers and human-machine monitoring int erfaces or stations for other equipment systems will also be in the contro l room. This eq uipm ent includes the facility crane, c l osed-circuit television system, CAAS, and rad iation monitoring system. A control panel for all facility on-site and off-site (if required) communications (e.g., telephone , intercom) will likely also be located there. The control room door into the facility wi ll be equipped with controlled access. The BMS will be primarily controlled and monitored from the control room. Utility systems with vendor packages a nd integrated co ntrol s will provide survei ll ance monitoring to the co ntrol room. The FPC system wi ll operate with a sync hroni zed hot standby redundant system structure for all hot cell processes.

Each hot cell process will be an independent subsystem having a local HMI with monitoring and control functions from the control room. Workstations for each system within the control room will be hot stan db y redundant.

The redundant stations will run software on identical PLC systems. The PLC systems will monitor each other. On loss of synchroniz in g signal from one system, the other system will continue with contro l and monitoring.

Process systems that w ill be primarily controlled in the control room include uranium recovery and recycle, target dissolution , and liquid waste handling.

The target receipt system will be controlled with local HMis in the irradiated target basket receipt bay or target cask preparation a irlock. Mo production process hot cell systems, including target disassembly and Mo recovery and purification , wi ll be controlled with local HMis in the hot cell operating ga ll ery. The hot cell processes will have monitoring and redundant contro l functions from the contro l room. The FPC subsystem for target fabricat i on processes will be contro ll ed with local HMis in the target fabricatio n area , with s ur veillance monitoring in the control room. Local HMis will be provided in Room Wl07, whic h hou ses equip ment for low-dose waste so lidific ation. Low-dose liquid waste will be piped in from the holding tanks in the utility area above Room W107 , and drums of so lidifi ed waste will be transported out by pallet jack. This lo cal HMI will be the primary control location for the hi g h-do se liquid waste so lidific ation, high-dose waste decay , spent resin dewatering, and solid waste handling hot ce ll operations.

7.5.4 System Performance Analysis and Conclusion The system performance analysis for each process system will be provided in the Operating License Applica tion. The overa ll I&C system perfonnance ana l ysi s and conclusions are provided in Section 7.2. 7-46 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems 7.6 RADIATION MONITORING SYSTEMS Th e radiation mon itoring systems will include CAMs, continuous monitoring at the exhaust stacks, process control in strume nt s, a nd personnel monitoring and dosimetry.

Process control instruments used to analyze for uranium conce ntrati ons are d escribed in each respective process system in Section 7.3. T he objective of the radiation monitoring system is to provide the RPF control room personnel wit h a continuous record and indication of ra di ation levels at se l ected locations where radioactive materials may be present, stored , handled , or inadvertently introduced. The syste m is a l so designed to ensure that there is accurate and reliable information concerning radiat ion safety as related to personnel safety. The de sign considera tion s for the rad iation monitoring system include the fo llowin g: *

Provision of information to RPF operators so that in the event of an accident resulting in a r e l ease of radioactive material , decisions on deployment of personnel can be properly made. Indication and recording in the control room of t h e gamma and airborne radiation levels in selected areas as a function oftime, and, if necessary , alarming to indicate any abnorma l radiation co ndition. These indicators aid in maintaining plant contami n ation levels as low as reasonably ac hie vab l e (ALARA) and in minimizing personne l exposure to radiation.

Provision of local a l arms and/or indicators posit i oned at key point s throughout the RPF w here a substantial increase in radiation l eve l s might be of immediate importance to personnel fre qu enting or working in the a r ea. Radiation Monitoring Locations RAMs wi ll b e lo cated in areas where personnel may be present and where radiation levels cou ld b ecome significant based on the fo ll owing considerations:

*

Occupancy status of the area, incl udin g time requirements of personnel in the area , the proximity to primary and seco nd ary rad ioacti ve sources , a nd shielding Potential for increase in the background radioactiv i ty level Desirability of survei ll ance of infrequently visited areas CAMs wi ll be loc ated in work areas w here ther e is a potential for a irborn e radioactivity. The CAMs will ha ve the capability to detect derived air co nc e ntr ations within a spec ifi e d time. 7.6.1 Design Criteria D esign crite ria for the radiat ion monitoring I&C systems are described in Section 7.2. 7.6.2 Design Basis and Safety Requirements The design ba sis and safety r equireme nt s for the radiat ion monitoring I&C systems are described in Section 7.2. Th e ESFs for this system are li s t ed in C h apter 6.0. 7.6.3 System Description Th e radiatio n safety monitoring system wi ll in clude CAMs, contin uou s monitors at the exhaust stac k s, and personnel monitoring a nd dosimetry.

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NOmfWEST M(DfCAl ISOT'Of'U NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems Thr ee basic types of personnel monitoring equipment will be used at the facility:

count rate meters (friskers), hand/foot monitors, and portal monitor s. All per sonne l whose duties require entry to restricted areas wi ll wear individual external dosimetry devices (e.g., passive dosimet e rs s uch as th ermo luminescent dosimeters that are sensitive to beta , gamma , and neutron radiation) from a Nat i onal Voluntary Laboratory Accreditation (NA VLAP)-certified vendor. Per sonne l monitoring a nd do s imetry is de scri bed in Chapter 11.0 , " Radiation Program and Waste Management." 7.6.3.1 Air Monitoring Continuous air monitors -CAM units will consist of a particulate mea s uring channel with a filter to capture particulate. Air will be drawn through the system by a pump assembly. The sam ple will be withdrawn from inside the a ppropriate area , room , or cell through an isokinetic nozzle with the sampling volume flow at a known fixed rate , so th a t the accumulation of radioactive particles can b e interpreted as a quantitative sa mple. After pa ssi ng through the nozzle , th e sample will be drawn throu g h tubing and through a fixed or moving filter tape before being di sc harged to the atmosphere.

The sam p lers also have a purging system for flushing the vo lume cell s urrounding the gas sample chamber with clean air for purpo ses of calibration and the removal of crust activity.

Replaceable liner s will be changed out periodically when contamination becomes excessive.

Flow regulating will e nsure that flow through the filters remains constant.

Each in s trument channel will include a detector , preamplifier , count rate meter , and pow e r s upply. Th e detector may be a scintillation counter or similar device having a gamma sens itive crystal , and a photo multiplier whose output pul ses are counted by the rate meter. Each readout module will be equipped with a li g ht that illuminates when the radiation level exceeds pre set limits. The se tpoint will be adjustable over the entire detection range. Pressing a button will cause the meter to indicate the alarm setpoint.

Visible alarms will be accompanied by a si multaneou s local a udible alarm with a n a larm light in t h e control room. A normally energized light will deenergize when there is a detector s ignal failure , circuit failure , pow er failure , or failure due to a disconnected cable. Power for the monitors that initiat es a sa fety signal will be provided from the UPS. Loss of power and signal failure will b e monitored for each detector.

CAMs will be provided with a check source. This check source wi ll simulate a radiation field and will be us ed as a convenient operational and gross calibration check of the detectors and readout equipment.

CAM calibration will include , where practical , exposures to the specific isotopes that the particular syste m monitors in the field. In st rument calibrations will be performed at prescribed frequencies. An electronic test signal and/or radioactive check so urce drift indication ma y also require CAM recalibration. Radiation area monitors -The RAM detector unit will be housed in an environmentally s uitable container that is mounted in a duct , on a wall, or other suitable s urface. The se nsitivity of each detector will be sufficient to have the a larm setpoint an order of m agni tud e high e r than the detection threshold.

Th e det ec tors are de signe d to be operational over a wide range of temp eratu re s. The de s i gn of the d etecto rs will m eet expected normal and abnormal environmental design conditions, as appropriate. Saturation will not be expected to adversely affect operation of the detector within its calibrated range. Sensors will be mounted as close as practic a l to the mo s t probable radiation so urces with no objects, per so ns , pillar s, a nd piping servi ng as s hielding. The sensors will also be mo u nted so as to minimize inaccuracies due to any directionality of th e detector.

Audible and visual alarm devices -When the radiation exceeds predetermined levels , a larms will actuate in the control room and at selected detector l ocations. 7-48 NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems The alarms will consist of the following capabilities:

* * "A lert li g h t" wi ll illuminate when the radiat ion l evel excee d s preset limits with an adj u sta ble setpoi nt "H i g h alarm red li g ht" will illuminate when rad i ation levels exceed a predetermined a larm setpoint "Fai lure alarm" will sound whe n either the power or the channe l's electronics fai l The visua l a larm s will be accompanied by a simultaneo u s audible a l a rm annunciator at the se l ected detector locations and in the control room. The annunc i ator windows for the monitors will be locat ed in the contro l room. The alarm can be manually reset w h e n the alarm conditions are corrected.

The lo cal a l arm horns and warning li g h ts will remain on until t h e radiation level is below the present leve l. Additional CAM requirements and locations are described in Chapte r 11.0. 7.6.3.2 Stack Release Monitoring The exhaus t stacks will be provided wi th continuous monitors for nobl e gases , particulate , and iodin e. The stack monitorin g system design basis i s to contin u ously monitor the radioactive stack releases.

Additiona l information wi ll be provided in the Operating License App lic ation. Airborne exposure pathway monitoring is described in Chapter 11.0. 7.6.4 System Performance Analysis and Conclusions Th e sys t e m perform ance a nal ys i s an d conclusions for eac h process sys t e m will b e pro vi d ed in the Operating License App li cat ion. The overall I&C system p erforma n ce analysis is pro v id ed in Section 7.2. 7-49 NWM I ...... HORTHWUTMfDtCAllSOTOPf.S

7.7 REFERENCES

NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems I 0 CFR 2 0.1501 , "Genera l ," Code of Federal R eg ulation s, Office of the Federal Regi ster, as amended. 10 CFR 70 , " Dome st ic Licensing of Special Nuclear Material ," Code of F e d e ral R egu l ations, Office of the Federal Register , as ame nded. I 0 CFR 7 0.61, " Performance Requirements," Code of F e d e ral R egu lati ons, Office of the Fe deral Regist er, a s amended. 10 CFR 70.64 , "Req uirem e nts for New Facilities or New Processes at Existing Facilities," Code of F e deral R egu lation s, Office of the Federal Register , as amended. 10 CFR 73.1 , " Purpose and Scope," Code of Federal R egu l ations, Office of the Federal Register, as amended. 10 CFR 73.54 , "Protection of Di g ital Computer and Communication Systems and Networks ," Code of F e deral R egu lations, Office of the Federal Regi ste r , as amended. ANS 10.4-2008 , V e rification and Va lidation of Non-Safety-Re lat e d Scientifi c and Engin eer ing Comput er Program s for th e Nuclear Indust ry , American Nation a l Standards Inst i tute , New York, New York , 2008. ANSVANS 8.3 , Criti c ality Accident A larm S ys t e m, American National Standards In st itut e/ American Nuclear Society , La Grange Park , Illinois , 1997 , R2003 , R2012. ANS V ISA 67.04.01-2006 , S e tp o int s for Nuclear Safety-R e lat e d Instrum e ntation , American National Standards Institute/International Society of Automation , Research Triangle Park , North Carolina , 2006 (R2011). ANSVISA 84.00.01-2004 Part 1 , Fun c tional Saf ety: Saf ety In s trum e nt ed S ys tem s for th e Proc e ss Indu stry Se c tor -Part 1: Fram ew ork , D e finiti o n s, Syst e m , Hardwar e and Softwar e R e quirements , American National Standards Institute/Internation a l Society of Automation , R esea rch Triangle Park , North Carolina , September 2004. ANSVISA 84.00.01-2004 Part 2, Fun c ti o nal Saf e ty: Saf ety In s trum e nt ed S ys t e m s for th e Proc ess Indu s try Sector-Part 2: Guid e lin es for th e Appli c ation of ANSI/ISA-84.00.01-2004 Part 1 (!EC 61511-1 Mod) -Informativ e, American National Standards Institute/Intern atio nal Society of Automation , Research Triangle Park , North Carolina , September 2004. ANS V ISA 84.00.01-2004 Part 3 , Fun c tional Saf ety: Saf ety In s trum e nt e d Systems for th e Proc e ss Indu stry Sector -Part 3: Guidan ce for the D e t e rmination of th e R e quir e d Saf ety Int e gri ty L eve l s -Informati ve, American National Standards Institute/International Society of Automation , Re sea rch Triangle Park , North Carolina , September 2004. EPRI TR-106439 , Guid e lin e on Evaluation and Accepta n ce of Comm e rcial Grade Digital Equipm e nt for Nuclear Saf ety App li c ations, Electric Power Re searc h Institute , Palo Alto, California, November 1996. IEC 61508 , Fun c tional Saf ety of El e ctrical/E l ec troni c/Programmable El ec tronic Saf ety-R e lat e d Syst e m s, Parts 1 -7 , International Electrotechnical Commission, Geneva, Switzerland , as amended. IEEE 7-4.3.2-2010 , IEEE Standard Crit e ria for Di g ital Computers in Safety S y st e m s of Nuclea r Po wer G e n e rating Station s, In st itute of Electrical and Electronics Engineers , Pi scataway, New Jer sey, 2010. IEEE 323-2003 , IEEE Standard for Qualifying C las s lE Equipment for Nuclear Pow e r Generating Station s, In st itute of Electrical and Electronics Engineers, Pi scataway, New Jer sey, 2003. 7-50 NWMI ...... *.* NOmfW£rT Ml:DICAI.

ISOTOPU NWMl-2013-021 , Rev. 3 Chapter 7.0 -Instrumentation and Control Systems IEEE 338-2012, IEEE Standard for Criteria for the P er iodi c Surveillance Testing of Nuclear Pow e r Generating Station Safety Systems, Institute of E l ectrica l a nd E l ectronics E n g in eers, Piscataway , New J ersey , 2012. IEEE 344-2004, IEEE R eco mmended Practic e for Seismic Qualification of Class 1 E Equipment for Nuclear Power Generating Stations, Institute of E l ectrica l a nd E l ectro nics E n gineers, Piscataway , New J ersey, 2 004. IEEE 379-2 014 , IEEE Standard Application of the Singl e-Failure Criterion to Nuclear Pow er Generating Station Safety Syst e ms , Institute of E l ectrica l and Electronics E n gineers, Pi scataway, New J ersey, 20 14. IEEE 384-2 00 8, IEEE Standa r d Criter i a for In dependence of Class J E Equipme nt a n d Circuits, In s titut e of E l ectrica l an d E l ectronics E n gi n eers, Piscataway, New Jersey , 2008. IEEE 497-2010, IEEE Standard Criteria for Accident Monitoring In strumentation for Nuclear Power Ge n erating Statio n s, Institute of E l ectrical an d E lectron ics E n gineers, Piscataway , New J ersey, 20 10. IEEE 577-2012 , IEEE Standard R e quir e m e nts for R e liabili ty Ana l ys is in the Design and Op e ration of Safety Systems for Nuclear Facilities, Institute of Electrica l a nd Electronics E n gineers, Piscataway , New J ersey , 2012. IEEE 603-20 0 9 , IEEE Sta ndar d Criteria for Safety Systems for Nuclea r Pow e r Generating Stations, Institute of Elect ric a l and Electron ic s Engineers, Pi scataway, New Jersey, 2009. IEEE 828-2012, IEEE Standard for Configuration Management in Systems and Software Engineering, Institute of Electr ic a l an d Electron i cs Engineers , Pi scataway , New Jersey , 2 01 2. IEEE 829-20 0 8, IEEE Standard for Software and System T est Documentation, Institute of Electr ical and E l ectro ni cs Engi ne ers, Piscataway, New J e rsey , 2 00 8. I EEE 1012-2012, IEEE Standard for System and Software Verification and Validation, In stitute of E l ectrica l and E l ectro ni cs Engi n eers , Piscataway , New Jersey , 2012. IEEE 1028-2008 , IEEE Sta nd ard for Software R eviews and Audits , In sti tut e of E l ec tri ca l and Electro n ics Eng in ee r s, Piscataway , New Jersey , 2008. I EEE STD 12207 , ISO I IEC I IEEE Standard for Systems and Software Engin ee ring -Softwar e Lif e Cycle Processes, Institute of E l ectrica l a nd E l ectronics E n gineers , Piscataway , New Jersey , 2008. IEEE STD 15 939, IEEE Standard Adoption of JSO/IEC 1 5939:2 00 7 Syst e m s and Softwar e Engineering Measurement Proc ess, In stitute of E l ectrica l an d E l ectronics Engi n eers , Piscataway , New J ersey, 2008. IS A-RP-6 7 .04.02 , M e thodologies for the D e t er mination of S e tpoints for Nucl e ar Safety-R e lat ed Instrumentation , Instrum ent Soc i e t y of America, R esearch Triangle Park , Nort h Caro lin a, 2 010. ISO/IEC/IEEE 15 288, Systems and Software Engineering

-System Life Cycl e Pro cesses, International Orga ni zation for Sta nd ardizatio n , Geneva , Sw it ze rl and, 20 15. ISO/IEC/IEEE STD 24765, Systems and Softwar e Engineer in g-Vocab ul ary , International Orga ni zatio n for S tanda rdizatio n , Geneva , Switze rland , 20 l 0. NUREG-0700, Human-System Interfa ce Design R ev i ew Guideline s, R ev. 2, U.S. Nuclear Regulatory Co mmi ssion, Office of Nuclear Rea ctor R egu lati o n , Washington , D.C., May 2002. 7-51

.; .. ;. NWMI ...*.. .. .. ........ *. *.* NOflllfW(n MEDtcAl ISOTOPU NWMl-2013-021, Rev. 3 Chapter 7.0 -Instrumentation and Control Systems NUREG-0711, Human Factors Engineering Program R eview Model, Rev. 3, U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety and Safeguards, Washington, D.C., November 2012. NUREG-0800, Standard R eview Plan for the Review of Safety Ana l ysis R eports for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards , Washington , D.C., as amended. NUREG-153 7, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Po wer R eactors -Format and Content, Part 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Re g ulation, Washington, D.C., February 1996. NUREG/CR-6090, The Programmable Logic Contro ll er and It s Application in Nuclear Reactor Systems , U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., September 1993. NUREG/CR-6463, R eview Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems, U.S. Nuclear Regulatory Commission, Washington, D.C., June 1996. NWMI-20 l 5-SAFETY-002 , Radi o i soto p e Production Facility Int egrated Safety Analysis Summary, R ev. 0 , Northwest M e dical I so topes , Corvallis, Oregon, 2015. Regulatory Guide 1.53, Application of the Single-Failure Criterion to Safety Systems, Rev. 2, U.S. Nuclear Regulatory Commission , Washington , D.C., June 2003. Regulatory Guide 1.97, Criteria for Accident Monitoring In stru m e ntati on for Nuclear Power Plants, Rev. 4 , U.S. Nuclear Re g ulatory Commission, Wa s hington , D.C., 2006. Regulatory Guide 1.152, Criteria for Use of Comp ut ers in Safety Systems of Nuclea r Power Plant s, Rev. 3, U.S. Nuclear Re gu latory Commission , Wa s hington , D.C., June 2011. Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Material Facilities, U.S. Nuclear Regulatory Commission , Office of Nuclear Regulatory Research , Washington, D.C., 2010. Regulatory Guide 5.71, Cyber Security Pr ograms for Nuclear Facilities, U.S. Nuclear R egu latory Commission, Washington, D.C., 2010. 7-52

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Chapter 8.0 -Electrical Power Systems Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021 , Rev. 3 September 2017 Northwest Medical Isotopes , LLC 815 NW g th Ave , Suite 256 Corvallis , OR 97330 This page intentionally left blank.

NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems Chapter 8.0 -Electrical Power Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 3 Date Published:

September 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 3 Title: Chapter 8.0 -Electrical Power Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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Rev D a t e 0 6/29/2015 1 6/26/2017 2 8/5/2017 3 9/5/2017 NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems REVISION HISTORY Reason fo r Re vision Rev is ed B y Initial Application Not re qu ired Incorporate changes based on resp o nses to NRC C Haass Requests for Add i tional Information Modifications based on ACRS comments C. Haass Incorporate final comments from NRC Staff and ACRS; C. Haass full document revision

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CONTENTS NWMl-2013-021, Rev. 3 Chapter 8.0 -Electrical Power Systems 8.0 ELECTRICAL POWER SYSTEMS ......................

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8-1 8.1 Normal E lectrical Power Systems ........................

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8-2 8.1.1 De s ign Ba sis of the Normal E lectric Power System ..........

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8-4 8.1.2 De s ign for Safe Shutdown ...................................................................................

8-5 8.1.3 Rang es of E l ec trical Po wer R e quired ..............

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8-5 8.1.4 Use of Substations De vote d Excl u sive l y to the Radioi soto pe Production Faci lit y ..................................................

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8-6 8.1.5 Spec ial Processing of Electrical Service ...................................

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8-6 8.1.6 De s ign and Performance Specification

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8-6 8.1. 7 Special Routing or Isolation

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8-6 8.1.8 Deviation s from National Codes ........................

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8-6 8.1.9 Technical Specifications

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8-6 8.2 Emergency Electrical Power Systems ...............

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8-7 8.2.1 Design Ba sis of the Emergency E le ctric Power System .......................

............... 8-8 8.2.2 Range s of Emergency Electrica l Po wer Required ...............................................

8-8 8.2.3 Power for Safety-Related Instruments

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.......... 8-8 8.2.4 Power for Effl uent , Proce ss, and Area Radiation Monitors .................................

8-8 8.2.5 Power for Ph ysi cal Security Control , Information , and Co mmuni catio n Systems ...............................................

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8-8 8.2.6 Power to Maintain Experimental Equipment in Safe Co ndition ..........................

8-8 8.2.7 Power for Active Confinement

/Containment Engineered Safety Feature E quipment an d Control Systems ............................

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.. 8-8 8.2.8 Power for Coolant Pump s or Systems ..........................

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8-9 8.2.9 Power for Emergency Cooling ..........................................

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..... 8-9 8.2.10 Power for E ngineered Safety Feature E quipment ....................

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.... 8-9 8.2.11 Power for Emerge nc y Lighting ............................................................................

8-9 8.2.12 Power for Instrumentation a nd Control Systems to Monitor Shutdown ..............

8-9 8.2. I 3 Technical Specifications

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8-9 8.3 Refer ences ....................

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........ 8-10 8-i Figure 8-1. Table 8-1. FIGURES NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems Radioisotope Production Facility E l ectrical One Line Di agram ..................................... 8-3 TABLES Summary of Radioi s oto pe Produ c tion Facility an d Ancillary Faci liti es E lectrical Loads (2 page s) .....................................................

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wotmfWEST MEDtCAl. tsOTOPD TERMS Acronyms and Abbreviations AEC active engineering contro l A TS automatic transfer switc h CAAS criticality accident alarm system NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems HY AC he ating , venti l ation , and air co ndit ioning IEEE Institute of Electrical and E l ectronics Engineer s IROFS item relied on for safety MCC motor control center NEP normal electrical power NFPA National Fire Protection Association NO x NWMI RPF SEP UPS Units gal hp hr Hz km kV kW L rru mrn sec v nitr ogen oxides Northwest Medical Isotopes , LLC Radioisotope Production Facilit y standby e l ec trical power uninterruptable power s upply ga llon hor sepower hour hert z kilometer kilovolt kilowatt liter mile minute second volt 8-iii

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NOflTHWEIT MEDICAi. ISOTort:S NWMl-2013-021, Rev. 3 Chapter 8.0 -Electrical Power Systems 8.0 ELECTRICAL POWER SYSTEMS This chapter provide s 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 supplies (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 systems will not damage the RPF or prevent safe RPF shutdown.

In addition, the RPF ha s a non-safety standby electrical power (SEP) system to reduce or eliminate process downtime due to electrical outages. A combination of UPSs and the SEP sys tem will provide emergency electrical power (defined in Section 8.2) to the RPF. Table 8-1 lists the RPF electrical load s, including the NEP system peak load s, which systems have UPSs, and t he loads for those systems supported by the SEP system. Table 8-1. Summary of Radioisotope Production Facility and Ancillary Facilities Electrical Loads (2 pages) Demand Target fabrication system Target receipt and disassembly system Target dissolution syste m Molybdenum recovery and purification system Uranium recovery and recycle system Was t e handling system Radiation monitoring an d CAAS systems Standby electrical power system General facility electrical power Process vessel ventilation system Facility ve nti l ation system Ventilation Zone I Ventilation Zone II/III Ventilation Zone IV Laboratory ventilation Supply air Fire protection system P l ant and instrument air system Gas s upply system Process chilled water system Normal electrical peak power load --125 16 8 30 40 40 54 30 40 10 13 25 34 5 7 N I A 173 232 40 54 67 90 215 288 295 396 38 51 49 66 0.8 I 60 83 0.8 280 375 8-1 Uninterruptable power No No No No No No Yes 3 No Yes 3 No No No No No No Yes 3 No No No Standby electrical peak power load --0 0 0 0 40 54 25 34 10 1 3 5 7 5 7 N I A N I A 101 13 5 40 54 67 90 215 288 295 396 10 13 49 66 o b o b 60 83 0.8 140 188 NW Ml-2013-0 21, Rev. 3 Chapter 8.0 -E l ectrical Power Systems Tab le 8-1. Summary of Radio i soto p e Productio n Facility and Ancillary Facilit i es Electrical Loads (2 pages) Normal electrical peak power load Uninterruptable power Standby electrical peak power load Demand Facility chilled water sys tem Facility heated water system Proces s stream system Dernineralize d water system Supply air system Chemical supply system Facility process control and communications systems Energy recovery Safeguards and security Administrative b u ilding Waste management building ---1 ,300 47 0.8 0.8 49 5 5 40 90 11 1, 743 63 66 7 7 54 121 15 No No No No No Yes No Yes No No

On l y parts of the system are provided with uninterruptable power supplies.

--0 0 0 0 0.8 1 0 0 49 66 5 7 0 0 40 54 18 24 3 4 b The fire detection and fire alarm subsystems will be provided by an uninterruptable power s upply with a 24-hr capacity. Chapter 9.0 provides additional det a il. CAAS = criticality accident a l ar m system N I A = not applicable. 8.1 NORMAL ELECTRICAL POWER SYSTEMS The NEP system will connect to electric utility power from the off-site utility transmission and distribution system at a point of common coupling.

This point of common coupling will be located near the property line on the NWMI site. The NEP distribution system will operate in a redundant electrical system topo l ogy from the utility transmission and distribution system to the 480 volt (V) service entrance switchgear that services the RPF electrica l distribution system and the de vices and equipment within the facility.

The RPF electrical distribution system is designed to support the safety functions protecting workers, the public , special nuclear material activities, and radioisotope production operation processes , as described in Chapter 4.0 , "Radioisotope Production Facility Descr i ption ," and to minimize the number of points where a failure in the RPF is a single point of power conveyance.

Figure 8-1 provides a preliminary electrical one-line diagrams for the electrical distribut io n topology. The electrical one-line diagrams will be updated after completion of the RPF final design and included in the Operating License Application.

Power will be provided to the NWMI site from an underground utility feed 0 to the pad-mounted switchgear l ocated outside of the RPF bui l ding. Power will then be routed underground from the switchgear to the Administrative Building f} and the RPF@. The underground feeders@) to the RPF will comprise two redundant full-capacity service laterals to the RPF. Each service lateral will support redundant fu ll-capacity service transformers 0 that will normally carry half the RPF load. Either of the RPF feeders can be opened and the tie breaker closed , as needed , allowing the other feeder to carry the entire RPF load. Any RPF loads requiring SEP will be provided power from the diesel generator when required 0. 8-2

[Proprietary In formatio n] Figure 8-1. Radioisotop e Production Facility Electrical One Line Diagram NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical P ower Systems 8-3 NWMl-2013-021, Rev. 3 Chapter 8.0 -Electric a l 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 una va ilable feeder main breaker and closing the tie breaker. Electrical distribution on the load side of the 480 V service entrance switchgear and the heating , ventilation, and air conditioning (HV AC) redundant loads will be serv iced from opposite sides of the switchgear through electrical equipment and feeders, including motor control centers (MCC), switchboards, and distribution panel boards. Equipment , systems , and devices designed with redundant or N+ 1 capability will be fed from opposite sides of the service entrance switchgear.

The planned loads on the MCC will be evaluated in the RPF final design to ensure the equipmen t is appropriately balanced. These loads will be provided in the Operating License Application. Systems requiring emergency electrical power in the event of the loss ofNEP will be serviced by an on-site diesel generator through the SEP system. Section 8.2 provides additional inform a tion on the SEP system. UPSs will be provided for selected systems for the RPF , as identified in Table 8-1. UPS systems include unit device , rack-mounted, and/or larger capacity cabinet units (a large battery room as part of the UPS system is not planned). These UPS systems will service loads requiring uninterruptable power on a short-t erm basis. The UPS systems will be backed up by the on-site die se l generator to extend the duration of power available to connected loads. The UPS systems lo cations on the electrical one-line diagram will be defined in the RPF final design and provided in the Operatin g License Application. Internal to the RPF and Administration Building , the NEP distribution system will service end user equipment and devices. Feeders , busing , overcurrent protection , devices , and equipment will provide the conveyance and conductor protection throughout the building.

Design of the electrica l distribution system includes recommended practices from the Institute of Electrical and Electronics Engineers (IEEE) 493 , R ec ommend e d Practice for the D es ign of R e liable Indu stria l and Comm e rcial Po wer Systems, and IEEE 379, Standard Application of the Single-Failure Criterion to Nuclear Pow er Generating Station Safety Systems. The electrica l distribution system topology will employ a redundant power conveyance system. The distribution system will include overcurrent protective devices, surge arresters, fusing , relays , and similar safety-related protective devices. These safety devices will conform to the requirements of the National Fire Protection Association (NFPA) 70 , National Electric Cod e, relevant IEEE standards and recommendations , and local codes and standards. 8.1.1 Design Basis of the Normal Electric Power System The NEP system design basis will provide sufficient and reliable electrical power to the RPF systems and components requiring electrical power for normal operations, inc ludin g the electrical requirements of the system , equipment, instrumentation , contro l , communication , and devices related to the safety functions and devices. 8-4 NWMl-2013-021, Rev. 3 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) or longer if identified as needed beyond 120 min in the final safety analysis.

The fire protection system will have a UPS that provides 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (hr) of uninterrupted power. If NEP service is reestablished within a determined timeframe (to be provided in the Operating License Application), the RPF will resume normal operation. Upon loss of normal power: * * * * *

Inlet bubble-tight isolation dampers within the Zone I ventilation system wi ll close , and the HY 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 emergency purge system will be activated for hydrogen concentration control in tank vapor spaces Uranium concentrator condensate transfer line valves will be automatically configured to return condensate to the feed tank due to residual heating or cooling potential for transfer of process fluids to waste tanks Equipment providing a motive force for process activities will cease, including:

Pumps performing liquid transfers of process solutions Pumps supporting operation of the steam and cooling utility heat transfer fluids Equipment supporting physical transfer of items (primarily cranes) 8.1.3 Ranges of Electrical Power Required The RPF power service wi ll be 480 V, 3-pha se, 120 amp , 60 hertz (Hz). The tota l 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 NWMl-2013-021, Rev. 3 Chapter 8.0 -Electrical Power Systems 8.1.4 Use of Substations Devoted Exclusively to the Radioisotope Production Facility The RPF will recei v e power from Columbia Water and Light through the Grindstone Substation.

Thi s 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 dist r ibution. 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 Lic e nse Application.

8.1.8 Deviations from National Codes The RPF electrical system will be designed to meet all required national codes and standards , as described in Chapter 3.0. 8.1.9 Technical Specifications As evaluated in Chapter 13.0, the RPF is designed to safely shut down without NEP for occupational safety and for protection of the public and environment.

The NEP system will not require a technical specification per the guidelines in Chapter 14.0 , " Technical Specification s." 8-6 8.2 EMERGENCY ELECTRICAL POWER SYSTEMS NWMl-2013-021, Rev. 3 Chapter 8.0 -Electrical Power Systems Emergency electrical power is defined by NUREG-1537 , Guidelines for Preparing and Reviewing Ap pli catio n s for the Licensing of Non-Pow er R eactors -Format a nd Content, as any temporary substitute for normal electrical serv ice. A combination of UPSs and the SEP system w ill provide emergency electrical power to the RPF , although only se lected UPS systems will ha ve a safety function.

A 1 , 000 kW ( 1 ,341 hp) diesel generator will pro vi de SEP. Figure 8-1 a lso pro vides the electrical distribution topology for the SEP system. Power from this generator will service the RPF through an automatic transfer sw itch (A TS). The normal power side of the A TS will be connected to the RPF serv ice entrance sw i tc hgear , with the load si de of the ATS to be connected to the sta ndby switchboard.

The SEP sys tem i s designed to s upport the safety functions during RPF operations to protect workers , the public , and environment.

The SEP system de sign include s recommended practices from IEEE 446 , R ecommended Practice for Emergency and Standby Power Systems for Indu stria l and Com m erc ial Applications, NFPA 110 , Standard for Emerg e nc y and Standby Po wer Syst e ms , IE EE 379 , and IEE E 493. The SEP sys tem will include overcurrent protective devices , surge arresters , fu s ing , rela ys, and similar safe t y-related protecti ve de v ice s. These safety d evices will conform to the requirements ofNFPA 70 , rele va nt IEEE standards and recommendations , and local codes and standards.

SEP will be available to the exhaust system throu g h a redundant electrical distribution topology.

Approximately half of the exhaust electrical l oad requiring s tandby will be connected to a n MCC , with the other half connected to a redundant MCC. The stan dby switchboard will service equipment and devices in the hot cell , contro l room , exhaust system ventilation system , and other loads requiring standby power. During switchover to the SEP , the loads will be sequenced to protect the generator and electrical equipment.

Feeders , bu sing, overcurrent protection , devices , and equipment will pro vide the conveyance and conductor protection throughout the building.

During normal operations, loads co nnected to the sta ndby switc hboard will be serv iced through the ATS with normal and facility electric power. In this way, any malfunctions of the SEP system during RPF operation with NEP will not interfere with normal RPF operations or prevent safe facility s hutdown. When the ATS senses a loss of normal power , the switc h will signa l the on-site diesel generator to start up. When the diesel generator voltage and frequency are withjn acceptab l e limits , the A TS will switch from the normal power so urce to the diesel generator power source. Load s connected to the standby switchboard will continue to be service d by the die se l generator until the normal power source return s. The ATS will sense the normal power source voltage and frequency.

Once the voltage and frequenc y are within acceptable limit s and after a prescribed dela y, the A TS wi ll switch from the die se l ge nerator po we r source to the normal power so ur ce. UPSs will be provided , as required.

The function of the UPSs i s to provide power to select load s 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 load s. The 1 , 000 kW (1,341 hp) diesel ge nerator will b e service d with a 3 , 785 liter (L) (1,000-gallon

[gal]) diesel tank. This capacity will enable the generator to operate for 11-14 hr , depending on actual loads , without requiring additional fuel. 8-7 8.2.1 Design Basis of the Emergency Electric Power System NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems The emergency e l ectrica l powe r system de sign basis i s to provide uninterrup te d po wer to instrumentation , contro l , comm un ication systems , a nd de v ic es required to su pp ort the safety functions protecting workers and the public , and to provide sufficient e l ectrica l power t o the RPF to ensure safe s hutd own in the event of loss ofNE P. The system design basis a l so provides SEP to opera t e select process-related equipment to limit the impacts of loss of NEP on RPF produ ction operations.

A dd itional information on the de sign b asis 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 th e RPF is 1 , 178.6 kW (1 , 585 hp). Table 8-1 lists the backup peak e l ectrica l power loads for di fferent lo ca ti ons and processes wit hin the RPF. 8.2.3 Power for Safety-Related Instruments Safety-relate d instrumentation will be provided with UPSs. The UPSs will provide power to safetyrelated instruments while the die se l generator starts a nd will provide service l oads requir i ng uninterruptable power on a short-term basi s. The diesel generator will maintain power until the normal power sys tem is operating within accepta bl e limits. 8.2.4 Power for Effluent, Process , and Area Radiation Monitors Effl u ent, process , and area radiation monitors will b e provided w ith the UPSs. T h e UPSs w ill pro vide service loads requiring unint errupta bl e power for up to 1 20 min , while the diesel ge n erator will main tain power until th e normal power system is operatin g within acceptable limi ts. 8.2.5 Power for Ph ys ical Security Control , Information, and Communication Sys t ems Physical security control , information , and communication syste ms will be provided with a UPS. The UPS provides service loads requiring uninterruptable power for up to 120 mi n , whi l e the diesel ge n erator wi ll maintain po wer until the normal power system is operating wit h in acceptab l e li mits. 8.2.6 Power to Maintain Experimental Equipment in Safe Condition T h ere 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 analysi s in Chapter 13.0 , the Zone I exha ust venti l at ion subsystems operations , e quipm ent , and com ponent s ensures the confinemen t of hazardous materials durin g normal and abnormal conditions , including natural phenomena , fires , and exp lo sions. After a loss of NEP , the Zo n e I exhaust ve ntil ation su bs ystem will au tomati ca lly place i t se lf into the passive mode , including inlet bubbl e-t ig ht isolation d ampers that cl ose to provide passive co nfin ement. 8-8 NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems The system will remain in this configuration until the voltage and frequenc y of power from the diesel generator are within acceptab le limit s. At that point , the system can be manually started and operated in a reduced venti lation mode with one operating group of HY AC fans a nd components. The Zo ne I exhaust ve ntilation subsystems are designed to function in a manner , whether operational or not , consiste nt with occupational safety and protection of workers , the publ i c, and environment.

Therefore , this system is not co n sidered an IROFS. 8.2.8 Power for Coolant Pumps or Systems Based on the analysis provided in Chapter 5.0 , "Coo lant Systems ," the coolant system i s 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 coo lant systems is not con s idered an IROFS. 8.2.9 Power for Emergency Cooling Based on the analysis provided in Chapter 5.0 , an emerge n cy coo lin g water system i s not required.

8.2.10 Power for Engineered Safety Feature Equipment Engineered safety features requiring power will be provided with UPSs. The UPSs wi ll provide service loads requiring uninterruptable power for up to 120 min. The diesel generator wi ll maintain power until the normal power system is operating within acceptable limit s. Additional information will be provided in the Operating License App lic ation. 8.2.11 Power for Emergency Lighting Power required for emergency li ghting w ill be provided b y UPSs. The UPSs wi ll provide serv ice load s requiring uninterruptable power for up to 120 min , wh il e the diesel generator wi ll maintain power until the normal power system is operating wit hin acceptab l e limit s. Additional informatio n will be provided in the Operating License App li cation. 8.2.12 Power for Instrumentation and Control Systems to Monitor Shutdown Pow er for instrumentation and control sys t ems used to monitor safe s hutdo wn wi ll be provided with UPSs. The UPSs will provide service load s requ iring uninterruptable power for up to 120 min , while the diesel generator will maintain power until the normal power system is operating within acceptable limit s. Additional information will be provided in the Operating License Application. 8.2.13 Technical Specifications As evaluated in Chapter 13.0 , th e RPF is designed to safe l y shut down without SEP consistent wit h occupational safety and protection of the publi c and the e nvironment.

The UPS systems, as required , are anticipated to be part of the technical specification for the system being supported.

The SEP syste m will not require a technical specification per the guidelines in C hapt er 14.0. 8-9

8.3 REFERENCES

NWMl-2013-021, Rev. 3 Chapter 8.0 -Electrica l Power Systems IEEE 379 , Standard Appli c ation of the Singl e-Failur e Crit e rion to Nuclear Power G e n e rating Station Safety Syst e ms , Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2014. IEEE 446 , R e comm e nd e d Pra c ti ce for Em e rg e n cy and Standby Power S y st e m s for Indu s trial and Comm e rcial Appli c ation s, Institute of Electrical and Electronics Engineers , Pi s cataway, New Jersey , 2014. IEEE 493 , R e comm e nd e d Pra c ti ce for th e D es i g n of R e liabl e Indu s trial and Comm e r ci al Pow er S ys t e m s (Gold Book), In s titute of Electrical and Electronic s Engineers , Piscataway , New Jersey , 2007. NFPA 70 , National El ec tri c al Co d e (N EC), National Fire Protection Association , Quinc y, Massachusetts , 2014. NFPA 110 , Standard for Em e rg e n cy and Standb y Pow er S ys t e m s, Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2014. NUREG-153 7, Guid e lin e s for Pr e paring and R e vi e wing Appli c ations for th e Li ce n s ing of Non-Pow e r R e actor s -Format and Content, Part 1 , U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation , Washington , D.C., February 1996. 8-10 NWM I ...... *

NOflTHWHT MEDICAl ISOTOPf.S NWMl-2013-021 , Rev. 3 Chapter 8.0 -Electrical Power Systems T hi s p age i n ten t io n a ll y l eft bl a nk. 8-11

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Chapter 9.0 -Auxiliary Systems Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021 , Rev. 3 September 2017 Northwest Medical Isotopes , LLC 815 NW gth Ave , Suite 256 Corvallis , OR 97330

.. ; .. NWMI ...... .. .... .......... * * , NORTlfWEST MEDICAL ISOTOPES This page intentionally left blank. NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Chapter 9.0 -Auxiliary Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 3 Date Published

September 5 , 2017 Document Number. NWMl-2013-021 I Revision Number. 3 Title: Chapter 9.0 -Auxiliary Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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NORTHWEST llEDICAl ISOTOPES T hi s p age inte n t i o n a ll y l eft bl a nk. NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Rev Date 0 6/29/2015 1 6/26/2017 2 N/A 3 9/5/20 1 7 REVISION HISTORY Reason for Revision Initial Application NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Revised By Not required Incorporate changes based on responses to NRC C. Haass Requests for Additional Information Incorporate final comments from NRC Staff and ACRS; C. Haass full document revision

.:;.-.; ... NWMI ..**..*.. *. .:-.* .. *.*.'* * *. *. * ' . NORTHWEST MEDICAL ISOTOPES T hi s pa ge int e nti o n a ll y l eft bl a nk. NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems

...... .; .. ;. NWMI ** ** NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems * * *

NORTHWEST MHUCAL ISOTOPES CONTENTS 9.0 RADIOISOTOPE PRODUCTION FACILITY AUXILIARY SYSTEMS ...................................

9-1 9 .1 Heating Ventilation and Air Conditioning Systems ...........................................................

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

9-2 9.1.2 System Description

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9-2 9.1.2.1 Confinement

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9-6 9 .1.2.2 Supply Air System ................................................................

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

9-10 9.1.2.4 Cleanroom Subsystem

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9-13 9. I .2.5 Physical Layout and Location ...........................................................

9-14 9. I .2.6 Principles of Operation

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

9-15 9.1.4 Instrumentation and Control Requirements

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9-16 9.1.5 Required Technical Specifications

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9-16 9 .2 Material Handling .......................

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9-I 7 9.3 Fire Protection Systems and Programs .............................................................................

9-18 9.3.1 Design Basis .......................................................................................................

9-18 9.3.2 System De scri ption ............................................................................................

9-18 9.3.2.1 Fire Suppression Subsystem

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9-19 9.3.2.2 Fire Detection and Alarm Subsystem

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9-20 9.3.2.3 Fire Extinguishers

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9-21 9.3.3 Operational Analysis and Safety Function .........................................................

9-21 9.3.3.1 Radioisotope Production Facility Fire Areas ......................

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9-22 9.3.3.2 Other Radioisotope Production Facility Systems ..............................

9-36 9.3.3.3 Architectural Features .......................................................................

9-36 9.3.4 Instrumentation and Control Requirements

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9-37 9.3.5 Required Technical Specifications

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9-38 9.4 Communication Systems ..................................................................................................

9-39 9.4.1 Design Ba sis ........................

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

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9-39 9 .4.3 Operational Analysis and Safety Function ....................

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9-39 9.4.4 Instrumentation and Control Requirements

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9-40 9 .4.5 Required Technical Specifications

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....... 9-40 9.5 Posse ssio n and Use of Byproduct, Source, and Special Nuclear Material..

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9-41 9.5.1 Design Ba sis ......................................................................

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.. 9-41 9.5.2 System De scriptio n ............................................................................................

9-4 I 9 .5.2.1 Special Nuclear Materials

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9-41 9.5.2.2 Byproduct Materials

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9-42 9.5.3 Operational Analysis and Safety Function .........................................................

9-42 9.5.4 Instrumentation and Control Requirements

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9-42 9.5.5 Required Technical Specifications

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9-42 9.6 Cover Gas Control in Closed Primary Coolant Systems ..................................................

9-43 9.6.1 Design Basis .................................................

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9-43 9.6.2 System Description

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9-43 9.6.3 Operational Analysis and Safety Function ...............................................

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9-43 9.6.4 Instrumentation and Control Requirements

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9-43 9.6.5 Required Technical Specifications

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9-43 9-i

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NOflTlfWEST MEDtCAL tsOTDPES 9.7 Oth e r A u x ili ary Syste m s ...............................

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... 9-4 4 9.7.1 U tili ty Syste m s ................................................................................

................... 9-4 4 9.7.1.1 D es i gn B as i s ...................

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9-44 9.7.1.2 Sys t em D esc r i p t i o n .......................................................

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......... 9-4 4 9.7.1.3 Op e rational Ana l ys is a nd Safe t y Fun ct i on ...........................

............. 9-6 0 9. 7. I .4 I n s t ru m e nt at i o n a nd Co n trol R e quir e m en t s ............................

.......... 9-6 1 9.7.1.5 R e quir ed T ec hni ca l S p ec i fica tion s ..............

...................................... 9-6 1 9.7.2 Co nt ro l a nd S tora ge of R a di oact i ve Was t e .........................................

............... 9-62 9.7.2.1 D es i gn B asis ..........

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9-62 9.7.2.2 System D escrip ti o n ...............

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9-62 9.7.2.3 Opera tion a l Ana l ys is a nd Safe ty F un ction ........................................

9-76 9.7.2.4 I n s trum e nt a ti o n a nd Co n trol R e quir e m en t s ....................

.................. 9-76 9.7.2.5 R e quir ed T e chni ca l S p ec ifi ca tion s .................................................... 9-7 6 9. 7.3 A n a l yt i cal La borator y ..................................................

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9-7 7 9.7.3.1 D es ign B as i s ....................................

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9-77 9.7.3.2 System D escrip ti o n .....................................

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..................... 9-77 9.7.3.3 O p era ti o n a l Ana l ys is a nd Safety F un ct i on ........................................ 9-78 9.7.3.4 In s trum e nt a t io n a nd Co ntr ol R e quir e m en t s .............

......................... 9-7 9 9.7.3.5 R e quir ed T ec hni ca l S p ecifica ti o n s .................................................... 9-7 9 9.7.4 C h e mi ca l S uppl y ................................................................................................ 9-79 9.7.4.1 D es i gn Ba s i s ...................................................................................... 9-79 9.7.4.2 Sys t em D esc r i pti o n .....................

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.. 9-79 9.7.4.3 Op e rational A n a l ys is a nd Safe t y Fun ct i o n ...................

..................... 9-86 9. 7.4.4 In s trum e ntati o n a nd C ontr ol R e quir e m e nts ...................................... 9-8 6 9. 7.4.5 R e quir ed T echn i ca l S p ec i fica tion s ...............................

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....... 9-8 6 9.8 R efe r ences ............................

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NORTHWEST MEDICAL ISOTOPES Figur e 9-1. Figur e 9-2. Figure 9-3. Figure 9-4. Figure 9-5. Figure 9-6. Figure 9-7. Figur e 9-8. Figure 9-9. Figure 9-10. Figure 9-11. Figure 9-12. Figur e 9-1 3. Figure 9-14. Figure 9-15. Figure 9-16. Figure 9-17. Figure 9-18. Figur e 9-19. Figur e 9-20. Figure 9-2 1. Figure 9-22. Figur e 9-23. F i gure 9-2 4. Figur e 9-25. Figure 9-26. Figure 9-27. Figur e 9-28. F igure 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 Leve l Confinement

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9-4 Upper Leve l Confineme nt. .........................................................

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9-5 Low er Leve l Confinement

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....... 9-6 Facility Ve ntil ation Syste m Diagram 1 ..................

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....... 9-8 Facility Ventilation System Diagram 2 ................

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9-9 Process F lo w Dia gra m for Proc ess Vesse l Venti l at ion Tr eat ment ......................

.......... 9-1 2 Life Safe t y Plan (First F l oor) ..........................................................

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........ 9-23 Life Safety Plan (Seco nd Floor) ..........

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............... 9-24 Second F loor Mechanical Utilit y Area .....................................................................

..... 9-45 Medium-Pressure Stea m System ............................

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.................... 9-46 Low-Pr ess ure Steam System .................

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

.................................. 9-50 Chilled Water System Critically Safe Hot Cell Loop ....................................................

9-51 Chilled Water System Target Fabrication Loop ............................................................ 9-52 Proces s Chi lled Water System ..........

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................. 9-53 Demineralized Water System ............

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

................................. 9-57 Nitro ge n and Helium Supp l y Sy ste m ............................................................................ 9-58 H y drogen a nd Ox yge n S uppl y System .......................................................................... 9-59 Waste Ma n agement Process Flow Diagram and Proces s Flow Streams .......................

9-64 High-Dose Liquid Waste Solidification Subsystem and Low-D ose Co ll ect ion Tank Loca tion ............

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9-65 Simplified High-Do se Waste Handling Proce ss Flow Diagram ....................................

9-66 Hi g h-Dose Wast e Treatment and Handling Equipme nt Arrangement

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9-6 7 Lo w-Do se Liquid Waste Evaporation System Location ................................................

9-68 Low-Dose Liquid Waste Disposition Proce ss ...............

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........ 9-69 Low-Dose Liquid Waste So lidi ficatio n Equipment Arrangement..

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9-70 Spent Resin Dewatering Operational Flow Diagram ...............

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

9-7 1 Solid Was t e Encap s ulation Operational F l ow Diagram ..............

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

9-72 High Dose Waste Deca y Ce ll Equipment Arrangement..

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9-73 High Do se Waste Hand lin g Operational Flow Diagram .................

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.... 9-73 Wa s t e Handling Flow Diagram ........................................

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......... 9-74 Wast e Han dlin g Equipment Arrangement

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.... 9-7 5 Anal yt ical Laboratory Layout .....................................

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............ 9-7 8 Chemica l S uppl y Room Eq uipm ent Layout ................

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9-80 Nitric Ac id Flow Diagram ................

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9-8 I Sodium Hydroxide Flow Diagram ............................

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.............. 9-83 Hydrogen Peroxide Flow Diagram ................................................................

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... 9-84 9-i ii

... ;.-.;* .. NWMI *::.**.*.*. ..... .. .. .. . * * * . NORTHWEST MEDICAL lSOTOPES NWM l-2 013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Table 9-1. Tab l e 9-2. Table 9-3. Table 9-4. Table 9-5. Tab l e 9-6. Table 9-7. Table 9-8. Table 9-9. Table 9-1 0. Table 9-11. TABLES Facility Areas and Respective Confinement Zones ......................................................... 9-3 Indications for Faci li ty Ventilation System Parameters

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

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

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9-60 High-Dose Waste Tank Capacities

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9-66 Low-Dose Waste Tank Capacities

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............. 9-69 Subsyste m 100 , Nitric Acid Tank Sizes ........................................................................

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

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..... 9-83 Subsystem 300, Reductant and Nitrogen Oxide Absorber Solutions Tank Sizes ..........

9-84 Subsystem 400 , Hydroge n Peroxide Tank Sizes ...........................................................

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

9-85 9-iv TERMS Acronyms and Abbreviations 89 Sr strontium-89 90 Sr strontium-90 99 Mo mol y bdenum-99 230 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-239 240 Pu plutonium-240 24 1 Am americium-241 ALARA as lo w as reasonabl y ac hi evab l e CF R Code of Federa l Regulations DBF design basis fire DOT U.S. Department of Transportation H 2 h y drogen g as HEGA hi g h-efficie nc y gas adso rpti on HEPA high-efficie ncy particulate a ir HIC high-integrity container HN0 3 nitric acid HV AC heatin g, venti l ation, and air conditioning IBC International Building Code NWMl-2013-021 , Rev. 3 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-enric hed uranium Mo mol y bdenum MURR University of Missouri Re sea rch Reactor NaOH sodium h y droxide NESHAP National Emission Standards for Hazardous Air Pollutant s NFPA National Fi re Protection Association NO x NRC NWMI OSTR osu PFHA RCA RPF SNM Ti0 2 u nitrogen oxide U.S. Nuclear Regul atory Commission Northwest Medical Isotopes , LLC Oregon State University TRJGA Reactor Oregon State University preliminary fire hazards analysis radiologically controlled area Radioisotope Production Faci lit y special nuclear material titanium dioxide uranium 9-v

.; .. ;*. NWMI ...... .. .... ..... .. .. .. * * * . NORTHWEST MEDICAL ISOTOPES U.S. U.S.C. VoIP Xe Units o c O f µ cm cm 2 ft ft 2 ft 3 ga l gmol hr m. in.2 kg L lb m M m z m 3 mm mm w wt% United States United States Code Voice over Internet Protocol xenon degre es Celsius de grees Fahrenheit micron centimeter square centimeter feet square feet cubic feet gallon gram-mo I hour inch square inch kilogram lit er pound meter molar square meter cubic meter minute millimeter watt weight percent 9-vi NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems NWM I ...... ' e * ! . NORTltWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 auxi li ary systems are important to the safe operation of the RPF and to protect the health and safety of workers, the public , and environment.

The chapter is organized in accordance with NUREG-1537, Guidelines for Pr eparing and R eviewing Applications for the Licensing of Non-Power Reactors -For mat and Content, as augmented by the Final Interim Sta.ff Guidance Augmenting NUREG-153 7 , "Guide lin es for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors ," Parts 1 and 2, for Licensing Radioisotop e Production Facilities and Aqueous Homogeneous Reactors (NRC, 2012). The RPF auxiliary syste ms include the following:

* * * *

Heating and ventilation, and air conditioning (HY AC) systems Fire protection systems Communication systems Possession and use of byproduct , source, and special nuclear material Cover gas control in the closed primary coolant system Other auxiliary systems, including utilit y systems, control and storage ofradioactive waste, analytical laboratory , and chemical supply For each auxiliary system, a description is provided of the system's capability to function as designed w ith out compromising RPF operations and to shut down the RPF during normal operations or under RPF accident conditions.

Each auxiliary system description includes:

* * *

9.1 Design ba s i s System de scriptio n Operational analysis and safety function Instrumentation and control requirements Required technical specifications and their bases , including testing an d s urveillanc e HEATING VENTILATION AND AIR CONDITIONING SYSTEMS The RPF HVAC system , a l so referred to as the facility ventilation system, i s designed to ensure that temperature , relative humidity , and air exchange rates are within the design-basis limit s for personnel and equipment and to ensure that all normal sources of airborne radioactive material are controlled so that occupational doses do not exceed the requirement s of T i tle 10 , Code of Federal R egu lation s, Part 20, "S tandards for Prot ectio n Against Radiation" (10 CFR 20). The system de sign i s consistent with NWMl's as low as reasonab l y achievable (ALARA) program. The RPF design features ensure that airflow and relative pressure will prevent inadvertent diffusion or other uncontrolled release of airborne radioactive material from the RPF. The facil it y i s also designed and operated to ensure that no uncontrolled relea se of airborne radioactive material to the unrestricted environment can occur. The analyses of system operations s how that planned releases of airborne radioactive material to the unrestricted environment wi ll 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 Waste Management," and a detailed airborne exposure analysis i s provided in Chapter 11, Section 11.1. I. I .2. 9-1

.. .. NWMI ...... .. .... ......... *.* * * * .' . NORTHW ES T ME DI CAL ISOTOPES 9.1.1 Design Basis NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems The facility ventilation sys tem is designed to provide confinement of hazardous chemical fumes and airborne radiological materials and conditioning of the RPF environment for facility per so nnel and equipment.

The de sign basis of the facility ventilation system and the proces s vessel ventilation sys tem is provided in Chapter 3.0 , " Design of Structures, Systems, and Components," Section 3.5.7.2; and the safety functions are provided in Chapter 6.0, "Eng ineered Safety Features," Section 6.2.1. l. 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 zo ne that is independent of the other ventilation zones. Zone IV wi ll be s lightly positi ve l y pressuri ze d w ith respect to the atmosphere.

Zone III will be the cleanest of the potentially contaminated areas, with each subsequent zo ne being more contaminated and having lower pressure s. 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 systems will maintain zone pressure differentials and containment:

(1) the Zone I exhaust system will service the hot cell , waste loading areas, target fabrication enclosures, and process vessel ventilation s ubsystems in Zone I; (2) the Zone Wiii exhaust system will se rvice exhaust flow needs from Zone II and Zone III in excess of flow cascaded to interior zones; and (3) a laborator y exhaust system will service fume hood s in the laboratory area. Supply air will be conditioned using filters , heater coi l s, 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 C F R 61 , " National Emission Standards for Hazardous Air Pollutants" (NESHAP), and applicable State J aw. A stack monitoring and sampling sys tem will be employed to demonstrate compliance with the stated regulatory requirements for exhaust. The RPF ventilation system will include the air s upply , proces s ventilation, and exhaust ai r systems and associated filters, fans, dampers , ducts , and control instrumentation.

The s upply air system will draw in and condition fresh air and distribute it throughout the facility. A portion of the supply air will enter the proces s venti lati on sys tem throu g h fume hoods , open-front enclosures, glove boxes , and hot cells, and will be removed with other exhaust air systems throu g h the stacks to the environment after being treated. The safety function s of the ve ntilation systems w ill serve to protect workers , t he public , and environment by maintaining confinement barriers in a multiple confinement barrier system. The RPF will typically be ventilated s uch that airflows travel from areas of lower potential for contamination to areas of higher potential.

The ve ntilation system functions will include temperature and air quali ty control to meet production and worker needs. 9-2 NWM I ...... .... *

NOlmfWEST MEDIC.AL ISOTOPlS 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 g loveboxes , HEPA 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. 3 Chapter 9.0 -Auxiliary Systems Table 9-1. Facility Areas and Respective Confinement Zones Area Hot cells (production)

Tank hot cell I Solid waste treatment hot cell I High dose waste solidification hot cell I Uranium decay and accountabi lity hot cell RIC vault I Analytical laboratory g lo veboxes R&D hot cell laboratory hot cells Target fabrication room and enclosures U t ility room Analytical l a boratory room a nd hood s R&D hot cell laboratory room and hoods Waste loading hot cell Maintenance gallery Manipulator maintenance room Exhaust filter room Airlocks" Irradiated target basket receipt bay Waste loading truck ba y Operating gallery and corridor E lectrical/mechanical supply room Chemical supply room Corridors Decontamination room Loading docks Waste management loading bay Irradiated target receipt truck bay Maintenance room Su p port staff areas I I II II II II II II II II II , III III III III III III III III IV IV IV IV IV a Confi nement zone of airlocks will be dependent on th e two adjacent zo ne s being connecte d. HIC high int egrity co nt a in er. R&D re sea rch and development. 9-3

.; ... ; .. NWMI ...... .. .. ........ *. ! * * * , NOKTMWEST MEmtAl ISOTOPES [Proprietary Information]

NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-1. Ground Level Confinement 9-4

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

NORTHWEST MEDICAL ISOTOPES [P ro p r i e t ary In fo rm a ti o n] Fi g ure 9-2. Upper Level Confin e ment 9-5 NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems

.*; .. ; ... *NWMI ...... ..* ... .... .. .. .. . *. NORTHWESTMEDICALISOTOPES 9.1.2.1 Confinement

[Proprietary Information]

Figure 9-3. Lower Level Confinement NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Confinement is an engineered safety feature of the HY 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 pro v ide a confinement system as simple and fail-safe as reasonably possible.

The process vessel ventilation system will serve a s the primary confinement pressure boundary and i s safety-related.

The Zone I exhaust subsystem is an engineered safety feature that (along with shielding) will create a secondar y 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 b y the zone boundaries, associated ventilation systems , and HEPA 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., loss of power). 9-6 NWMI .......... *. ........ . *. NORTHWEST MEDICAL ISOTOPES 9.1.2.2 Supply Air System NWMl-2013-021, Rev. 3 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 100 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 contro l valves will operate in conjunction with exhaust valves to control the pressure differential in each zone by maintaining a fixed difference between the total supp l y 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 supp l y air s y stem HV AC controls will operate through the building management system. 9-7

.:;.-.;* .. NWMI ..*...... *. : ........... . , *. NORTHWEfTMEDICA L ISOTOPfS [Proprietary Information]

NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-4. Facility Ventilation System Diagram 1 9-8

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

NORTHWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-5. Facility Ventilation System Diagram 2 9-9

.*; .. .NWMI ..**..*.. * . .............. . *. e * . . NORTlfWEST MEDICAL ISOTOPES 9.1.2.3 Exhaust Air System NWMl-2013-021, Rev. 3 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 s y stem 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 HEPA filters will prevent contamination spread in case of an upset condition that re s ults in positive pressurization of Zone I spaces with respect to Zone II spaces. The process vesse l 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 space s through HEPA filters. The exhaust air from each cell will pass through local HEPA 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 , HEPA 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 neg a tive 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 supply 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 NW MI * "NOlllTlfWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 could 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 Ventilation 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 l ocations that are routed to the iodine removal subsystem include the following:

* * * * [Proprietary Information]

[Proprietary Information]

9-11

..... NWMI ...... .. .. . .......... e * ! ' . NORTifWEST Mf.D ICM. ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 3 Cha p ter 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 ho t cell will include offgas from the target di sasse mbl y and the tar get dissolution offgas systems.

Target disassembly

-[Proprietary Information].

Target dissolution

-[Proprietary Information].

After the offgas filter will be the dissolver off gas system's vacuum pumps and tanks , then the stream will flow through the seco ndary fission gas adsorbers and into the process vessel vent header. 9-12

............ .. NWMI ......... * . . *. .". NCRTHWEST MEDICALISOTOPES NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Iodine removal unit for uranium , molybdenum, and waste accumulation tanks -Some of the liquid s in the hot ce ll w ill contain tellurium isotopes that generate iodine isotopes during decay. A portion of the iodine will remain in the dissolver solutio n. Although it is not likely that much of the iodin e will evolve into the offgas, these strea m s 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 ce ll vesse l s, including the Mo feed tanks, impure U collection and l ag storage tanks, U recovery waste tanks, and the liquid waste handlin g tanks. [Proprietary Information].

This offgas stream will flow into the process vessel vent header. General vessel vent system -This header system wi ll service the remaining vesse l s 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 vesse l vent header without ad ditional treatment.

High volume evaporative vent from waste handling -This system will service the three waste solidification unit operatio n s (low-dose liquid waste , hi gh-dose liquid waste , and so lid waste) and the low-do se evaporation tanks. The lo w-dose evaporatio n tanks will have hi gh flowrate and elevate d temperatures to a ll ow water to evaporate.

The h eader wi ll collect these humid air sweeps and dilute with additio nal air bleed to ensure that t h e evaporated water do es not conde n se in the ducting or pipes. This offgas stream will flow into the process vesse l vent header. Target fabrication vent -The target fa bric ation area venti l ation is required for confining:

(1) off gas from the dissolver and ot h er process vesse l , a nd (2) offgas from the calcination or reduction furnace systems , where h ydrogen will b e diluted w ith air to l ess th an the lower flammability lim it. This off gas s tr eam will flow into the process vesse l vent he ader. Process vessel vent iodine removal unit -The process vesse l vent IRU (VV-SB-520) syste m w ill consist of a so rb ent bed of charcoal or activated carbon to remove iodine from the vesse l ven t gases. The process vesse l ve nt IRU is part of an item relied on for safety (IROFS) RS-03, "Hot Ce ll Secondary Co nfin ement Boundary." C hapt er 6.0 , Section 6.2.1 , and C h apter 13.0, "Accident Analysis,'

' Section 13.2.2.8, provide a dditional detail on the safety function.

Process vessel vent filter -This treatment operation wi ll consist of HEPA filtrat ion and the exhaust fan and will flow to the Zone I exhaust syste m. 9.1.2.4 Cleanroom S ubsystem The Mo purification hot cell cleanroom s ub system is designed to provide filtered and conditione d air at an exc han ge rate to meet the sta ndard s of a n I SO 14644-1, "C l eanroo m s a nd Assoc iated Contro ll ed E n vironments-Part 1: Classification of Air Cleanliness,

C l ass 8 cleanroom.

T h e cleanroom wi ll be main ta in ed at a s li ghtly positive pressure relative to its s urroundin gs to ensure t hat unfilt ered air doe s not infiltrate the c leanr oom. Air in si d e the c le anroom w ill b e continua ll y recirculated thr o u gh a d e dicat ed filtration syste m to remove intern ally ge nerat ed contaminants.

Air will be 100 percent recirculated, with the only air exc han ge with the surroundings of the c l eanroom occurri n g through exfi ltr ation and mak eup air entering on the suction s ide of the fan. The cleanroo m a ir handling unit and filters will be locat ed inside the hot ce ll and , therefore, must be remotely maintainable.

Periodic cleanroo m certification testing will also need to b e performed remotely w ith permanently installed instrumentation.

9-13

.;.-.;*.NWMI ...... ..* .. . ..... .. .. .. * " "NORTHWEST MEDICAL ISOTOPES 9.1.2.5 Physical Layout and Location NWMl-2013-021, Rev. 3 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 HEPA 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 temperatu res and negative pressurization during normal operations. Supply air temperature from the air-handling uni ts 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 o v erall building pressures , the Zone I header pressure , and Zone II exhaust header pressures during normal operations.

Loca l room pressurization will be obtained by the airflow offset between supply and exhaust. Suppl y 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 airflow s are not e x ceeded. frequency drives on the exhaust fans wi ll be provided to maintain required exhaust flow s 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 wilJ 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 remain s fixed. Pressure and flow conditions for the process enclosures and laboratory ventilation will be manuaJJy controJJed using volume dampers and valves. Airflow control valves will be i nstaJJed in each room's main supply and exhaust ducts to maintain laboratory design space pressure. These valves will be located outside of the l aboratory modules. The Zone I exhaust system for each module will be adjusted manuaJJy 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 controlJed 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 enc l osure. The Zone II exhaust system is designed to maintain the Zone II enclosures at their requi re d pressure.

A balancing valve located in the exhaust duct of each enclosure will initially be partially close d. As the local filter of the enclosure loads up and a drop in pressure increases across the filter , the va l ve 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 enc lo sure. The enclosure's pressure drop reading will be calibrated to its acceptable face velocity range to monitor enclosure performance.

9-14

... ; .. ; .. NWMI ...... .. ... ..... .. .. .. . *. NORTHWEST MEOICAUSOTOPES NWMl-2013-021, Rev. 3 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 proces s enclosures, general exhaust, and to maintain the design temperatures in the laboratories. [Proprietary Information]

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

9.1.3 Operational Analysis and Safety Function Chapter 11.0 and Chapter 13.0 provide an analysis of normal and off-normal operation of the RPF HV AC system. Chapter 11.0, Section 11.1.1.1 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 repre sen ts the failure of a system designed to mitigate other accidents that lead to an airborne release of radionuclides in the form of particulates or gases. Systems that will mitigate these releases include the primar y confinement and primary offgas treatment system, which will capture particulates, absorb iodine, and absorb Xe and Kr and other gaseous radionuclides, to slow the release following decay to more stable i soto pes. In the target fabrication processes , uranium will be handled in physical forms that do not contribute to a dose rate factor in airborne releases. Uranium solutions will also be proce sse d in closed sys tems with filtered process ventilation systems to remove the s mall amounts of activity normally 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 designated as an IROFS (RS-03 , " Hot Cell Secondary Confinement Boundary"). The operations, equipment, and components of this system will ensure the confinement of hazardous materials during normal and abnormal conditions, including natural phenomena , fires , and explosions.

Components of the dissolver off gas subsystem and the process vessel ventilation system 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 eva lu ates 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 this accident, the exhaust stac k height was identified as an IROFS (FS-05, "Exha ust Stack Height").

This analysis is discussed in more detail in Chapter 13.0. This passive engineered control is designed and fabricated with a fixed height for safe release of gaseous effluents.

9-15

.. ; ... ; ... NWMI ...... ..* .... ......... *.* , *. ... * . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems 9.1.4 Instrumentation and Control Requirements Section 9.1.2.6 provides a general description of the operation of the RPF ventilation 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 sequence of operation will be developed and provided in the Operating License Application. 9.1.5 Required Technical Specifications The technical specifications associated with the ventilation system, if applicable, will be discussed in Chapter 14.0 , "Technical Specifications, as part of the Operating License Application. Table 9-2. Indications for Facility Ventilation System Parameters Parameter Equipment operating status Damper position status Exhaust header pressure Fan speed Filter differential pressures Equipment bearing vibration E quipment bearing temperatures HEPA filter unit air inlet temperature HEPA filter unit airflow rate First-stage HEP A inlet temperature Fan motor amperage Fan thermal overload Zone I header pressure Zone II header pressure Confinement z one pressure differentials 11mlm11M 1 1@fl(.]M ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ ,/ HEP A = high-efficienc y particulate air. 9-16

.:;.-.;* .. NWMI .*:.**.*.* . ........... :. . '. e * ! .' , NORTHWEST MEDICAL ISOTOPES 9.2 MATERIAL HANDLING NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems The RPF does not handle or store reactor fuel. Material h andling activities are di sc u ssed in Chapter 4.0 , "Radioisotope Production Facility Description," Sections 4.3 and 4.4 , and are analyzed in C h apte r 13 .0. 9-17

.; .. ;* .. NWMI ...... ..* .... ............ *. e * . NORTHWEST MEDICAL ISOTOPES 9.3 FIRE PROTECTION SYSTEMS AND PROGRAMS NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems The fire 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 h y drants) to the automatic sprinklers and standpipes in th e building.

9.3.1 Design Basis The fire protection s y stem design provides detection and suppression 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 fires 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 suppres s ion 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. All alarms (fire , supervisor y and trouble) will transmitted to the site central alarm station and RPF control room. Fire protection system components will have fail-safe features and audible/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 suppression subs y stem will include automatic sprinkler, HEPA filter plenum deluge water sprays , and portable fire extinguishers.

Water will be supplied from the exterior fire hydrant supply 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 i f required , the fire protection system can have a dedicated water storage facility on site. 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 Application.

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 ...... ..* .... ......... *. '. e * ! ' NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 sprink l er head, the links will melt or break (depending on type) and re l ease 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 Sprinkl e r Syst e ms. 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 environment.

Additional detailed information on the firewater runoff storage will be developed for the Operating License Application. The fire protection s y stem is divided into two major subsystems. The subsystems and components are categorized as follows: *

Fire suppression subsystem consisting of automatic sprink l ers, a HEPA filter plenum deluge, glovebox fire suppression , and fire hydrant s Fire detection and alarm subsystem consisting of: Control s (e.g., fire alarm control panel , subpanels , or devices used for control of devices) General area detection (e.g., room smoke and heat detectors , manual pull stations)

Duct smoke detection for non-nuclear venti l ation systems, g l ovebox heat detection HEPA filter plenum heat detection Fire suppression subsystem monitoring devices (e.g., waterflow switches, tamper switche s, 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 g lo veboxes 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 wi ll be provided by fire extinguishers with an appropriate classification (discussed further in Section 9.3.2.3). A 20.3 centimeter (cm) (8-inch [in.]) network of main piping (common l y called 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 contro l valves , check valves , waterflow switches , and a test/drain assembly for detection of waterflow and system maintenance. Pip in g from the risers will support automatic sprinklers located throughout the faci lit y. The automatic sprinkle r system is designed in accordance with NFPA 13. 9-19

.. ;.-.;* .. NWMI ...... .. .. . ' .... .... .. . * *. * * . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems The HEPA filter plenum deluge will be also supplied by the 20.3 cm (8-in.) pi p ing 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 act i vated 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 se nd a signal 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 HEP A filters. The heat detectors and deluge valve for a pa rticular p l enum 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, manuall y activated feature is designed to spray directly on the HEPA filters and is intended to only be used ifthe HEPA filter ignites. The manual feature will include a contro l 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 NFPA 24, Standard for the Installation of Private Fire Service Mains and Th eir Appurtenances, and the International Fire Code (IFC, 2012). The subsystem is designed to support fire flows of 5,680 L/min (1,500 gal/min) overall and at least 1,893 L/min (50 0 gal/min) at each fire hydrant. 9.3.2.2 Fire Detection and Alarm Subsystem The fire detection and alarm subsystem will pro v ide a range of fire detection capabi liti es and notification methods. The primary means of detection will be b y monitoring the fire suppression system devices , including flow switches that indicate release of water from automatic sprinklers or deluge va l ves, and tamper switches that supervise valve position. Smoke and heat detection will be provided in specific lo cations 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 b y NFP A 101 , Life Safety Code, and NFP A 72, National Fire Alar m Code, smo ke 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 severa l safety aspects. Smoke detectors will be provided in: * *

Non-nu clear ve ntilation systems, in accordance with NFPA 90A, Standard for the Installati on of Air-Conditioning and Ventilating Systems, 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 Nuc l ear 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 vent il ation system exhausts for both n otification of high temperatures and release of the automatic portion of the HEPA filter plenum deluge capabil it y. Control modules and relays will be integrated into the fire detection and alarm subsystem.

Control modules will provide signals for releasing the deluge va l ves for the HEPA filter plenum deluge capability, and control methods will be integrated for shutdown of non-safety HY AC systems. 9-20

.*; .. ;.NWMI ...... .. .... ..... .. .. .. *. NORTitWESTMEDICALISOTOPES NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Alarms received by the fire alarm control panel will be transmitted via a copper cable or fiber optic cable network to monitoring stations 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 rec eive its primary power supply from a dedicated circuit off of the normal building power. Internal batterie s will provide a secondary power source, with connection to the standby generator.

The batteries will be sized to provide 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (hr) of backup power, plus 10 min of alarm power, as required by NFPA stan dards. 9.3.2.3 Fire Extinguishers In addition to the automatic features of the fire sup pression subsyste m , manual response capabilities will be provided via fire extinguishers with an appropriate classification.

Fire exti nguishers will be located throughout the building , as requir ed b y NFPA 10, Standard for Portable Fire Exti ngui shers. Specific extinguisher type s, suc h as tho se for metal fires or particular chemicals, will be specified depending on the hazard. 9.3.3 Operational Analysis and Safety Function Chapter 13.0 identifies fire hazard s and evaluates adve rs e events and accident sequences. The criticality safety evaluations di sc ussed 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 high uranium proces s strea ms or the high-dose process streams. As part of this analysis, an emergency purge gas system was identified to prevent flammable concentration in proce ss vesse l head spaces. IROFS FS-03, " Process Vessel Eme rgency Pur ge System ," is discu ssed in Chapter 13.0, Section 13.2.7 , and in Chapter 6.0. Th e following summarizes NWMI-2013-039 , Preliminary Fire Hazards 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 situ a nd transient fire hazard s Determine the effec t s of a fire in any loc atio n in the RPF and the ability to safely s hut down the facility and/or minimize and control the release of radioactivity to the environment Specify m easures for fire prevention, dete ct ion , s uppre ss ion , and containment for eac h fire area housing s tructures , systems, and components that are important to safe ty , in accordance with U.S. Nuclear Regulator y Commission (NRC) guidelines and regulation s The PFHA assessed the fire hazards at the RPF , support facilities, and surrou nding project site. The analysis also assessed the fire safety criteria identified in NRC Regulatory Guide 1.189 , Fire Protection for Nuclear Pow er Plants. The PFHA provided a consequence evaluation of a design ba sis fire (DBF) scenario within each fire area, assuming the loss of automatic and manual fire suppression.

The PFHA also identified facility design features and fire ha zar d miti gating features for personnel safety and property protection commensurate with the NRC criteria.

9-21

... ; .. NWMI ...... .. .... ............ *. *. * ! .' . NORTHWEST MEDICAL ISOTOPES 9.3.3.1 Radioisotope Production Facility Fire Areas NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems The fire hazards , life safety 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 , respec t ively. 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 requirements Fire areas will typically be bounded by 2-hr fire-rated barriers to separate:

* * * * *

Processing areas and radioactive material storage areas from each other and adjacent areas Rooms with major concentrations of electrica l and mechanical equipment from adjacent areas Computer and control rooms from adjacent areas Maintenance shops from adjacent areas Combustible storage area s from adjacent areas Fan room s and plenum chambers from adjacent areas Office areas from moderate and high fire hazard areas In one case, two fire areas will be separated b y 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 Challenge Fire Walls, Fire Walls , and Fire Ba r rier Walls. Where fire-rated assemblies are partially or full y penetrated b y 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 assembly. 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 w i th the fire-rated barrier in which they are installed , and comply with the requirements ofNFPA 80, Standard/or Fire Doo rs and Oth e r Opening Prot ectives. 9-22

... ; ... NWMI ::::**:;; .. ...... *. " ".". NORTiiWEST MEDICALISOTOPES

[Proprietar y Information]

NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-7. Life Safety Plan (First Floor) 9-23

.:;.-.;* .. NWMI ..*...... * . ........... : . * * * ! ." , NORllfWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 3 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 Informati o n]. 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 a llowable 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 manually 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 ..... .. .. .. .

NORTHWESTMED I CALISOTOPES NWMl-2013-021, Rev. 3 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, provid ing a fire rating equivalent to the barriers.

The hot cell area could be exposed to a fire in an adjacent fire area when the large 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 by 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 fire 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 l ocated 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 systems 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. Manua l 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 ...... .. .... ...........

e ' . NOflTlfWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems * * *

Scenario 3 -A fire starts within the exhaust stack system that is caused by 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 starts 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 source caused by the operation of a semi-t ractor 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 battery , and small amounts of other combustible material.

A small amount of permanent combustibles , inc ludin g e l ectrical cab l es , polyethylene tarps , isopropyl alcoho l , vinyl, and trash bins , may also be present. These combustibles will be limited by administrative contro l s. The DBF scenario postulates that the entire contents of the fuel tanks 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 diameter will also ignite and be completely consumed (NWMI-2013-039). Consequences of an Automatic Fire Suppression Failure Failure of the automatic fire suppression system will cause a delay in respond i ng to a fire , resulting in the combustibles being completely consumed during the DBF. The adoption of administrative controls will limit combust ibl es 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 box 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 unlikely 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 i f the fire protection systems in this fire area wi ll 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 ...... ..* ... ......... *. * * * ! ' . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 (JOO ft), in accordance with IBC Section 1014.3. The exit access tra v el 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 I 018.4 , 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 wa s te shipping activities.

The primar y area of concern i s an open doorwa y 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 combustible s, 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 quickl y 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 a l so 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 l standpipe s y stem 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 NWM I ...... * * . NOllTHWfST MEDICAL JSOTOPES Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios NWMl-2013-021, Rev. 3 Cha p ter 9.0 -Auxiliary Systems 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 13.0]). 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 combustib l e loading of the area was considered low. The fire also consumes other transient combustibles located w i thin 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, whethe r 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 located adjacent to the productio n 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 NWMl-2013-021 , Rev. 3 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 l avatories, and several sma ll offices. The control room will be separated from the remainder of the adm ini stration and s upp ort area by 2-hr fire-rated barrier walls. The operations performed w ithin the administration and support areas will be consistent wit h office space uses. The occupant load of the administration and support area wi ll include non-production work staff. Life Safety Considerations The administration and support area is required to mee t IBC life safety criteria (ICC , 20 12) and will be provided wit h emergency li ghting, illuminated exit signs, automat ic sp rinkl ers , and an automatic and manually actuated fire a larm system wit h audible and v i sual indicating devices as necessary.

An accessib le means of egress will be provided to t h e area in accor danc e wit h the IBC. Exit access will be provided to the admin i strat ion and s upport area by one main exit at the front of the building and a secondary exit lo cated to the south side of the RPF. A break room w ill a l so be provided with an add i tional exit. The maximum distances to the exit access within the administratio n and support area will be within the following parameters.

The travel distance for the common path of egress travel for a mixed u se business (B) and assem bl y (A-2) occupancy eq uipp ed throughout with an automatic sprinkle r system will be 23 m (75 ft), in accordance with IBC Table 1014.3. The exit access travel distance for a fully sprinklered mixed-u se business (B) and assembly (A-2) occupancy wi ll be limited to 76 m (250 ft), in accordance with IBC Tab l e 101 6.2. Dead-ends in corri dor s will not exceed 6.1 m (20 ft), in accordance with IBC Sect ion 1018.4. Exposure Fire Potential/Potential for Fire Spread between Fire Areas The administration and s upport area w ill be separated from other fire areas of the RPF b y 3-hr fire-rated barriers.

Penetration s in the fire-rated barrier walls wi ll be protected with penetration sea ls , providing a fire rating eq ui va l ent to the barriers. Load-bearing structural elements are not required to be protected b y fire-resistive construction. To prevent a fire from s preading between areas , admi ni strative contro l s will be impl emented that dictate personnel procedure s and limit combust ibl es around access doors. Fire s pread between areas w ill be therefore mitigated b y personnel actions, limit ed com bu stibles, an d 3-h r fire-rated boundaries.

Fire Protection Features The administration and s upport area requires th e fo ll owing fire protection features to provide a dein-d epth approach to fire protection.

This approach wi ll result in a fire bein g quickly detected and suppressed, reducing fire-induced damage. * *

Automatic

-An au tom atic wet-pip e sprinkler system wi ll be installed throughout the admi nistration a nd 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 suppress ion will be provided withi n t h e administration a nd support area and consist of portable fire ext inguish ers a nd C l ass I sta ndpip e system ho se valves. Manual fire alarm pull statio n s will be provided a t the ex it s from the administration and s upp ort area. Passive -Pas sive fire protection will be provided in the form of fire-rated construction to protect the ad mini stration a nd support a r ea from other occupied areas of the facility.

9-29

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

NORTMWEST lllEDJCAL ISOTOPf.S Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems The DBF was determined to consist of ordinary combustibles (e.g. paper products 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 transient combustibles located within t he office and spreads to nearby cubicles.

The DBF would result in the complete combustion of the combustible materia l s 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 Su ppre ssion 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 wit hin 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 station or the automatic smoke or temperature detection systems. The DBF would be contained within the administrative and support area by 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 fai led and extinguish the fire using portable extinguishers or the ho se stream supported by the Class I standpipe system. The required response time of the fire department will be determined for the Operating License App li cation. Conclusion The above analysis and description show that the fire protection and life safety systems wit hin 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 thi s fire area is not expected to contain anything other than check sources for instrumentation, no releases to the public are expected to occur. Therefore, this system will likely 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 side of the RPF and will be a noncombustible enclosure that is considered a storage S-2 occupanc y area. The truck bay will be capable of accepting three semi-tractor trailers at the same time. Each truck bay will be se parated from the production area (cask unloading) by a 2-hr rated rollup door. The doors to the production area will 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 wil l be transported in approved containers.

The casks will reside on the heavy-duty tractor-trailer for delivery and removal from the RPF. The duty tractor-trailer will be present when the retractable doors are open to the production area. 9-30 NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Life Safety Considerations The irradiated target receipt and waste management truck bay areas will be required to meet IBC li fe 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 truck bays. The maximum distances to the exit access within the truck bay will be estab li shed 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 (I 00 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 I 016.2. 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-resistive construction , as indicated in the IBC (ICC , 20 1 2). 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 wi ll be implemented that dictate personnel procedures and limit combustibles around the interface access doors. Personnel actions, limited combustibles, and 2-hr fire-rated boundaries 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 pro t ection 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 wi ll be insta ll ed throughout the truck bay area . However , due to the l arge 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 bay area. Additional detailed information will be developed for the Operating License Application.

Manual -Manua l fire suppression will be provided within the truck bay area and consist of portable fire extinguishers and Class I standpipe system hose va l ves. 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 spi ll s within the area. The fuel traps and sloped floors wi ll a l so 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 NWM I ...... * *

NORTHWEST MEDICAL ISOTOPES Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios NWMl-2013-021 , Rev. 3 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 is caused by the ru p ture of the fuel tank and ignition of the unconfined (static) diesel spill. Scenario 4 -A fire starts on a di esel-dr iven forklift that is caused by the rupture of the fuel tank and ignition of the unconfined (static) diesel spi ll. The DBF for the truck bay consists of a die se l fuel spi ll and ignition from an unknown source caused by the operation of a diesel-powered sem i-tractor trailer (Scenarios 3 and 4). The truck i s assumed to have two 284 L (75-gal) diesel fuel tanks, along with 32 hard rubber tires, a batt ery, and small amo unt s of other combustib le materi al. The truck may also carry some combustibles on noncombustible pallets when supporting radiological material-handling operations. Administrative controls wi ll be used to limit temporary combustible items 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 approximate l y 2.6 m 2 (28 ft\ The results of the DBF were po stulated as the complete combustion of the combustible materials in the irradiated target receipt truck ba y area. No credit was taken for fire suppression act i vities. The DBF fire could result in the complete loss of function for the systems and/or equipment i n the area. Consequences of an Automatic Fire Suppression Failure Failure of the automatic fire s uppr essio n system will cause a delay in responding to a fire , resulting in the combustibles being comp l ete l y consumed durin g the DBF. The adoption of administrative controls will limit combustibles and minimize the s pread of fire. However , smoke and hot gases could damage equipment located withi n the truck bay area. In the event of a fire, the Co lumbia Fire Departm ent will be notified of a fire by either actuation of a m a nual fire alarm pull box s tation or the automatic smoke or temperature detection systems. The DB F wou ld be contained with in the truck bay area b y the 2-hr rated fire walls. If the automatic fire s uppres sio n system fails to operate , the fire department is expected to arrive we ll before t he 2-hr fire walls have fai l e d and extinguish the fire using portable exti n guishers or the ho se stream s uppo rted b y the Class I standp ipe system. The required response time of the fire department wi ll be det ermine d for the Operating License Application.

Conclusion The above analysis and description s how that the fire protection and life safety systems wit hin the truck bay are designed such that they wi ll function in a manner , whether operational or not , consistent with occupational safety and protection of the public and e nvironment.

Becau se the radioacti ve 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 likel y not be considered an IROFS. Additional detail ed information will be developed for the Operating License Application.

9-32 9.3.3.1.5 Laboratory Area NWMl-2013-021, Rev. 3 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 w ith 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 samp l es 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/s trontium-90 (89 Sr!9°Sr) Gamma energ y 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 audib l e 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 laboratory area will be separated 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 l arge 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 combust ible s, and 2-hr fire-rated boundaries.

9-33

... ; ... ; .. NWMI ...... .. .... ............ . * *. * .' . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Fire Protection Features 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 L icense 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 sprinkler s , 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 exi t 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

..

..... .*.: .. :.** NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems * * *

NOflTHWEST MEDICAL ISOTOPES 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 indu ced damage. * *

Automatic

-An automatic wet-pip e 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 wi ll be provided in the form of fire-rated construction to protect separation between fire areas. Isolation dampers will be provided in the inlet and out l et of each HEPA filter housing to prevent fire from spreadi n g to other fire areas. Fire Hazards, Ignition Sources, and Design Basis Fire Scenarios The following were considered DBF scenarios for the utility areas. * * *

Scenario 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.

Additiona l 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 "NWM I ...... *

NOllTKWEST MEDICAL ISOTOPES 9.3.3.2 Other Radioisotope Production Facility Systems 9.3.3.2.1 Facility Ventilation and Smoke Management NWMl-2013-021, Rev. 3 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 proces s. To assist in the confinement of airborne radioacti ve contamination, the RPF ventilation system is de s igned 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 Protection for Laboratories Using Chemica ls , and will be constructed to maintain fire ratings where ducting penetrate s fire-rated barriers , as appropriate.

The confinement ventilation systems will also include HEPA 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 th e damage caused by smoke. These strategies will be designed in accordance with NFPA 92, Standard for Smoke Control Systems. The smoke control methods for each fire area will be developed for the Operatin g 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-03). The HEPA filters are expected to contain low l evels 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 l arge filter face area will require automatic and manual sprinklers in the plenum housings and contaminated water collection or retention.

In addition, the HEPA filter housing s 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 s upporting the faci lit y 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 wil l be deve loped for the Operating License Application.

9.3.3.2.4 Security and Safeguard Components Security systems are discus se d in Chapter 12.0, "Co nduct 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 compliance with prescriptive criteria, as defined by the PFHA (NWMI-2013-039).

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...... ;*.NWMI ...... ..* .. . *.*.* .. *.*.* ' e * ' . NORTHWEST MEDICAL ISOTOPlS Types of Construction NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems All structures within the RPF complex confines will be constructed of Type IIB, noncombustible ma t erial, as defined by IBC Chapter 6 (ICC, 20 I 2). Additional detailed information will be developed for the Operating License App lication. 9.3.4 Instrumentation and Control Requirements The fire protection system will report the status of the fire protection equipment to the central alarm station and the RPF control room, with sufficient information to identify the ge neral location and progress of a fire within the protected area boundaries.

Initiating devices for the fire detection and alarm s ub system, which will include monitoring de vices for the fire su ppres s ion subsystem, will indicate the presence of a fire within the facility.

Once an initiating device activates, signa ls will be sen t to the fire alarm control panel. The fire alarm control panel will transmit s ignals to the central alarm station and perform any ancillary functions , such as shutting down the venti lation equipme nt or actuating the deluge valves. As required by NFPA I 01 and NFPA 72, smoke detection will be provided above the main fire alarm control panel and an y sub panel s necessary to perform control functions for the system. For ventilation unit s, smoke and heat detection will be provided in s upport of severa l safety aspects. Smoke d etectors will be provided in non-nuclear ventilation syste m s in accordance with NFPA 90A and the IFC. Smoke detectors will also be provided in air intake s to address smoke infiltration from wildland fires and fires in other facilities that might spread smoke to the surro undin g area. Smoke detection will be provided in ve ntilation systems servic ing potentially co ntamin ated zo ne s to s upport shut down and minimize the spread of contaminated smoke to other areas of the RPF. H eat d e tector s will be provided in the se vent ilation system exhausts for both notification of hi gh temperatures a nd release of the automatic portion of the HEPA filter pl enum deluge s ub system. Control modules and relays will be integrated into the fire detection and alarm subsystem to initiate reactions required for safety. Contro l module s wi ll provide s ignals for releasing of delu ge va lves on the HEPA filter plenum deluge s ub system. Control methods will a l so be inte grated for s hutd ow n of the HV AC sys tems. Shutdown of electrica l equipment or computers will also occur as de emed necessary by the design effort. Alarms received by the fire alarm contro l panel wi ll be t ransmitted v ia 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 sitewide infrastructure to the central a l arm stat ion. The alarm sta tions will provide data to the Columbia Fire Department for response. System Monitoring The fire protection system will be monitored b y the fire alarm contro l panel , wh ich will transmit signal s to the central alarm statio n via a digital alarm communicator transmitter and to the RPF control room. Command and control functions will be exclusively avai labl e at the fire alarm control panel. Localized monitorin g of the various fire pumps w ill occur at the respective pump controllers.

9-37

.:;.-.;* .. NWMI -*:.**.*.*. * ......... . . *. *. * ! ' . NORTHWEST MEDICAL ISOTOPES Control Capability and Locations NWMl-2013-021, Rev. 3 Chapter 9.0 -A u xiliary 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 HEPA 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 Maintenance of Water-Based Fire 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 available to general building personne l. Thus, there will be no ability for the system to be locally manipulated without proper authorization.

Additiona l information will be provided in Chapter 7.0 for the Operating License Application.

9.3.5 Required Technical Specificatio n s 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 ...... .. .. ........ !. NORTHWESTMEOICAUSOTOPES 9.4 COMMUNICATION SYSTEMS NWMl-2013-021, Rev. 3 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 complies with Electronic Industries All i ance 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 abilit y to make facility-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 legacy 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 comply with Telecommunications Industry Association and NFPA 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 sys t em. 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 *i: **:*:** ...... * ! * , NORTHWEST MEDICAL ISOTOPES 9.4.4 Instrumentation and Control Requirements NWMl-2013-021, Rev. 3 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 ...... * *

NORTffWEST MEDICAL ISOTOPlS NWMl-2013-021, Rev. 3 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

Personnel 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 procedures developed for the Operating License Application will ensure that only radioactive byproducts handled by the RPF are permitted, unless specifically authorized by the 10 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 (so lid , liqu id, or airborne) from the facilities can occur and that personnel exposures to radiation, including in gestio n 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 provided in Chapter 3.0. 9.5.2 System Description SNM is defined by Title I of the Ato mic Energy Act of 1954 (42 U.S.C. 2011 et seq.) as plutonium, uranium-233 (233 U), or uranium enriched in the isotopes 233 U or 235 U. The RPF wil l 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 b y exposure to radiation incident to the process of producing or using SNM. As defined by NRC regulations, byproduct material includes any radioact i ve material (except enriched uranium or plutonium) produced by a nuclear reactor. The RPF will handle byproduct material durin g the se paration 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 enr ich ed in the isotope 235 U. Source materials will not be present in the RPF. 9.5.2.1 Specia l 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 inventor y is listed in Chapter 4.0 , Table 4-1 , and the irradiated material area SNM in ve ntory 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 is 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 limitin g values in I 0 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 ..... .... .. .. .. ' e * ' . NORTHWEST MEDICAL ISOTOPES 9.5.2.2 Byproduct Materials NWMl-2013-021, Rev. 3 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 de s ign 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 dis solut ion -Chapter 4.0, Section 4.3.4

Molybdenum recovery and purification

-Chapter 4.0, Section 4.3.5

Uran ium r ecovery and recycle -Chapter 4 0 , Section 4.4.1

Waste handling-Chapter l I .0, Section 11.2 Chapter 4.0 and Section 9.7.2 provide description s of the design of spaces and e quipment to ensure that there is no uncontrolled release of radioactive materials (solid, liquid, or airborne) from the RPF and that personnel exposures to radiation, including inge st ion or inhalation , do not exceed limitin g values in 10 CPR 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 Ana l ysis and Safety Function The criticality safety of SNM i s discussed in Chapters 4.0 and 6.0, and the mat e rial control and accounting of SNM is discussed in Chapter 12.0 , Section 1 2.13. 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 In strumenta tion and control requirements for the processe s associated with the posses s io n and u se of byproduct materials and SNM are discussed in Chapter 7.0 and Chapter 1 2.0, Section 12.13. 9.5.5 Required Technical Specifications The technical specifications associated with the possession and use of byproduct material s and SNM, if applicable, will be di sc ussed in Chapter 14.0 as part of the Operating License Applicatio

n. 9-42 NWMl-2013-021 , Rev. 3 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 cooling water to numerous process tanks. The radiolytic decomposition of water withi n this system could result in the production of hydrogen (H 2) and oxygen mixtures.

This section provides a dis cussion of the cover gas control system within the pro cess coolant system. 9.6.1 Design Basis Information on the design basis of cover gas contro l in the c l osed primary coo l ant system (process chilled wa t er system) is provided in Chapter 3.0, Section 3.5.2.7. 9.6.2 System Description The process chilled water system is described in Section 9. 7 .1.2.2. The accumu lati on of comb ustibl e gases with in thi s system will be contro ll ed b y the "sweep" gas sys tem th at is described in Section 9.7.1.2.6.

Gases e ntrain ed in the chilled water system wi ll be released in the cooling water collection tanks. Hydrogen, which is the primary co mpon ent of evo lved combusti bl e gases, diffu ses very rap i dly and will be diluted by the airflow provided by the sweep gas flow. The plant a ir supply system (described in Section 9. 7.1.2.4) will provide low-flow [Proprietary Inform atio n] purge gases to Tanks TK-4 20 and TK-320. The process vessel ven t system wi ll collect the purge gas from each of the tanks and merge the co ll ected vent su b systems into t h e main faci lit y ve ntil ation system for treatment and filtration.

These systems will work together to prevent exp lo sive gas mixtures from d eve l opi n g. 9.6.3 Operational Analysis and Safety Function C h apter 13.0 eva lu ates the accident seq u e n ces that involve either com bu stible so lid s or liquids, or explosive gases, in close proximity to the high uranium process streams or the high-dose process s tr eams. T hi s analysis determined that i f the purge air system was not operational, a hydrogen-air co n ce ntr a ti on in se le cted tanks could rise above 25 percent of the lower exp l osive limit , and an ignition source co uld cause a d eflagratio n or detonation, r esulting in the r e l ease ofrad i onuc lide s into the air. The tanks associated wit h the coo lin g system are not anticipa t ed to require IROF S contro l s. 9.6.4 Instrumentation and Control Requirements Instrumentation a nd co ntrol r equi r eme nt s fo r the cover gas control in the closed primary coo l ant system are di scusse d in Chapter 7.0. 9.6.5 Required Technical Specifications The technical s p ecificatio n s associated with the cove r gas contro l in th e closed primary coo lant system, if applicable, wi ll b e discussed in Chapter 14.0 as part of t h e Operating License App lic ation. 9-43

.. ;.NWMI ...... .. .. . .... .. .. .. e

NOllTifWEST MEDfCAl lSOTOPES 9.7 OTHER AUXILIARY SYSTEMS NWMl-2013-021, Rev. 3 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 utiliti es Control and storage of radioactive waste (waste management)

Analytical laborator y Chemical s uppl y The followings subsections describe these auxiliary syste ms , including their de s ign basis , sys tem description , operational analysis and safety function , in st rumentation and control requirements , and t ec hnical specifications.

9.7.1 Utility Systems The utility systems will provide heating, coolin g, process water, compressed gases, instrument , motive force, and other functions to s upport uranium processing, waste handling , and ve ntilation. The utility systems will include the following subsystems:

* *

Process steam Process 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 i n the event radioactive material is released by the operation of one of these sys tems , potential radiation exposures would not exceed the limits of I 0 CFR 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 de s ign basis is provided in C hapt e r 3.0, Section 3.5.2.7. 9.7.1.2 System Description Figure 9-9 shows the seco nd floor mechanical utility area where the process steam, chilled process water, demineralized water , and plant or instrument supply air units will be housed. He lium , h y drogen, and oxygen will be pro vi ded b y bottl ed gases located near the point of use either in the laborator y area or the target fabrication area. Nitrogen will be provided b y a tube trailer for nitro ge n located outside of the laboratory area. 9-44

[Proprietary Information]

NWMl-2013-021 , Rev. 3 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 by a natural gas-fir ed boiler (ST-H-100). Low-p ressure 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

- - - ".NWMI ...... ** ** .......... ' e

NORTlfWEST MEDICAi. JSOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-10. Medium-Pressure Steam System 9-46

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

NORTHWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-11. Low-Pressure Steam System 9-47

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

NOIHffWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 proce s s chilled water loop that will cool the t hree 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 loop s will be cooled by the central chilled water s y stem through plate-and-frame heat exchangers. Several process 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 l oops. 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 system and the proces s fluid. The large-geometry secondary loop in the hot cell will meet the cooling demands where fissile material leaking through a heat exchang e r is not a credible event. The other cooling loops will be inherentl y criticality-safe b y geometr y, 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 exchanger 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 systems will provide heating and cooling media to the HY AC system. The facility chilled water s y stem (FCW) will supply the HY AC 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 HY AC air-handling unit cooling coi l s. There will be three equal-sized facility chillers located adjacent to the RPF: two in operation and one spare. The heating water system (HW) will supply the HY 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 HY 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

i*;h* NWMI ...... * * * , NOllTitWEST MEDICAL ISOTOPES This pa ge intentionall y l eft blank. 9-49 NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems

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

NOlmfWEST M£DfCAl ISOTOPES [Proprietar y Information]

NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-12. Chilled Water System Large Geometry Hot Cell Loop 9-50

.. ; ... ; ... NWMI .*;.**.*.* . .............. . *. * * ' . NORntWEST MEDICAL ISOTOPES [Proprietar y Inform a t i on] NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-13. Chilled Water System Critically Safe Hot Cell Loop 9-51

... NWMI ...... ..* .. . ..... .. .. .. * * . NOllTHWEST MH>>CAl tSOTOPlS [Proprietary Information]

NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-14. Chilled Water System Target Fabrication Loop 9-52

.:; .. NWM I .*:.**.*.* . ............. * * .. NOfllltWEST MUllCAl ISOTOPES [Propri etary Information]

NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-15. Process Chilled Water System 9-53

... NWMI ...... ..* .. . ........ * . . * *. NORTHWEST MEDICAUSOTOPES 9.7.1.2.3 Demineralized Water NWMl-2013-021 , Rev. 3 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 1 I 5). 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 several 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 liquids 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-1 lOA, 1 lOB , and IA-F-110).

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

The lead/lag configuration can supply reduced flow after a single compressor failure. The plant air receiver will 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-1 20) to make up the difference between peak demand and compressor capacity.

9.7.1.2.5 Gas Supply Gas supply of helium (Figure 9-I 8), hydrogen, and oxygen (Figure 9-19) will be supplied by 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 ft\ The nitrogen will be fed from the tube truck (GS-Z-100) to the chemical supply room where manifold piping will be used to distribute the gas. The primary use of nitrogen will be in the reducing furnaces during target fabrication.

Helium , hydrogen, and oxygen gas bottles will be located near the points of use. Gas supply pressures will be regulated to 1.7 kg/cm 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 supp l y when one bottle (or rack of bottles) is empty , and alert the operator when bottles need to be replaced.

Hydrogen and ox y gen 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 L/min (250 ft 3/min) air for each gas cabinet. 9-54

... ; .. ;* .. NWMI *********** .............. *. * * . NOflllfWEST MEDICAL ISOTOPES T h is p age i n te n t i ona ll y l eft bl ank. 9-55 NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems

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

NOflTltWEST Mf.DtCAL ISOTOP£S [Propri etary Information]

NWMl-2013-021, Rev. 3 Chap t er 9.0 -Auxiliary Systems Figure 9-16. Demineralized Water System 9-56

.:;.-.; .. NWMI :: i:**::-.* ...... '. NORTHWEST MEDICAL I SOTOPES [P ro p rieta r y In formatio n] Figure 9-17. Plant Air Sys tem NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems 9-57 NWMI .*:.**.*.* . . : .......... . *. * . NOimtWEST MEDICAL ISOTOPES [Propri et ar y Inform a ti o n] NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-18. Nitrogen and Helium Suppl y S y stem 9-58

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

NOllTlfWEST MEOtcAt. lSOTOfl£S

[P ro pr ieta r y In for m atio n] NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-19. Hydrogen and Ox y gen Supply S y stem 9-59

.. ;.-.;;. NWM I ...... ..* .. . ..... .. .. .. . '. *. * '

NOR TH WEST MED I CAL ISOTOPES 9. 7.1.2.6 Purge Gas NWMl-2013-021, Rev. 3 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. High purge Low purge Tab l e 9-3. Purge Gas Flows Flow rate [Propriet ary Informat ion] [Proprietary Information]

.. [Propri etary [Proprietary Information]

In formation]

[Proprietary

[Proprietary Information]

Information]

Units (basis) a NWM l-20 1 3-CALC-005 , Tank Air Bleed Est im at e, Re v. B , Northwest Medical I sotopes, LLC, Corvallis , Oregon , 2014. b NWM l-2 0 I 3-CALC-009 , Uranium Purification S ys tem Equipment Si z ing, R ev. B , Northwest Medi cal I so tope s , LLC , Co rv a lli s, Oregon, 2014. = h ydrogen ga s. u = uranium. 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 vessels. The tanks anticipated to require purge gas are listed in Table 9-4. Additional informatio n on the purge gas system will be developed for the Operating License Application.

Table 9-4. Tanks Requiring Purge Gas Tank number Tank number Tank name DS-D-100 Dissolver I UR-TK-120A Impure uranium collec t ion tank 2A DS-D-200 Dissolver 2 UR-TK-120B Impure uranium collec t ion tank 2B DS-TK-8 00 Waste co llection and sa mpling tank I 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 IA UR-TK-160A Impur e 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 I MR-TK-200 Feed tank 2 UR-TK-900 IX waste collection tank 1 MR-TK-340 Waste co ll ection tank UR-TK-920 IX waste collection tank 2 UR-TK-lOOA Impure uranium collection tank IA WH-TK-100 High-dose waste collection tank UR-TK-IOOB Impure uranium collection tank I B WH-TK-240 High-dose concentrate collection tank IX = ion exchange.

u uranium. 9.7.1.3 Operational Ana l ysis an d Safety F u nction Chapter 13.0 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 accident analysis associated with utilities addresses fissile solution leaks across a mechanical boundary between process vessels or backflows into a utility system. 9-6 0 NWMI ......

NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Defense-in-depth

-The tank and vessel walls will be made of corrosion-resistant materials and have wall thicknesses that are rated for long service with harsh acidic or basic chemicals. Items relied on for safety-Based on the analysis conducted in Chapter 13.0 ,Section I 3.2, the following IROFS are implemented.

* * *

IROFS CS-10 , "Closed Safe Geometry Heating/Cooling Loop with Monitoring and Alarm" IROFS CS-20, "Evaporator

/Concentrator Condensate Monitoring" IROFS CS-27 , " Closed Heating/Cooling Loop with Monitoring and Alarm" IROFS FS-03 , " Process Vessel Emergenc y Purge S y stem" IROFS CS-18 , "Backflow Prevention Device" The analyses that identified these IROFS and the associated s y stem 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 processes associated with the utility system are discussed in Chapter 7.0. 9.7.1.5 Required Technical Specifications The technical specifications associated with the utility system , if applicable, will be discussed in Chapter 14.0 as part of the Operating License Application.

9-61

... ; .. ; ... NWMI .*:.**.*.* . ............ * *: . NOflntWEST MEDICALISOTOPES 9.7.2 Control and Storage of Radioactive Waste NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems The radioactive waste control and storage systems are designed to ensure that ( 1) any potential malfunctions do not cause accidents in the RPF or uncontrolled release of radioactivity, and (2) in the event radioactive material is released b y the operation of one of these systems, po tential ra diation exposures would not exceed the limit s of 10 CFR 20 and remain consistent w ith the NWMI ALARA program. No function or malfunction of the auxiliary systems will interfere wi t h or preve nt safe shutdown of the RPF. 9.7.2.1 Design Basis The waste handling syste m design ba sis is provided in Chapter 3.5.2.7. 9. 7.2.2 System Description To fulfill the design basis, the control and storage of radioactive waste will include the following functions:

* * *

*

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

Low-dose liquid waste handling (collection, evaporation, recycle and solidification)

Spent resin dewatering Solid waste encapsulation High-dose waste decay High-dose waste handlin g Waste handlin g Waste Staging and Shipping Building (Class A storage) These functions are described in detail in the following subsections.

Figure 9-21 s ummari zes the weekly de s ign ba sis vo lumes and the average a nn ua l weekly vo lumes of all waste handling proces s streams. The design basis volume is based on eight University of Missouri Research Reactor (MURR) targets and 30 Oregon State University (OSU) TRI GA 1 Reactor (OSTR) targets per week to provide appropriately sized tanks. The annual weekly average is based on processing eight MURR targets per week for 44 weeks per year and 30 OSTR targets per week for eight weeks per year and is used in the s izing of the high-dose d ecay s torage. 1 TRIGA (Training , Research , I sotopes, General Atomics) is a registered trademark of General Atomic s, Sa n Diego , Ca liforni a. 9-62 NWMI ...... .. .. . ... .. : ... * * *

NOIITTM'EST MEDICAL ISOTOPES T hi s p age inte nti o n a ll y left bl a n k. 9-63 NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems

-::.**.*.*. NWM I ......... *.* .* NOllTNWESTMEDICAllSOTOPES

[Propri etary Inform a t io n] NWMl-2013-021 , Rev. 3 Chap t er 9.0 -Auxiliary Systems Figure 9-20. Waste Management Process Flow Diagram and Process Flow Streams 9-64 NWMl-2 013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems 9.7.2.2.1 High-Dose Liquid Waste Handling Figure 9-2 1 shows the location in the hot cell area where the high-dose liquid waste will be processed.

High-dose liquid waste w ill b e collected in the high-dose waste co ll ection tank (s ho wn in Figure 9-22), which w ill provide the needed h a ndling capacity to match the vo lume of liquid waste generated b y the upstr ea m processes.

C hapter 4.0 provides descriptions of the high-dose liquid streams that w ill b e dir ected to the collection tank. [Proprietary Inform ation] Figure 9-21. High-Dose Liquid Waste Solidification Subsystem and Low-Dose Collection Tank Location The process stream volumes are summarized in Figure 9-20 , and Table 9-5 provides the high-dose waste tank capacit ies. The process streams include: * * *

Caust ic scrubber waste Oxidizing column waste NO x absorber waste Regeneration waste from Ti0 2 #1 IX * *

9-65 Raffinate/rinsate from #2 IX Raffinate/rinsate from #3 IX U IX waste

.; .. ;.NWMI ...... : ... ...... * * . NOlmfWEST MEOICAl ISOTOPES [Propri e t a r y Inform a ti o n] NWMl-2013-021 , Rev. 3 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 Hi g h-do se wa s te a ccumulation tank High-dose concentrate accumulation tank 9-66 Tank capacity 5 , 050 1 , 270 19 , 000 4,800

.*; .. ; NWMI *:****.*.*.* ........... , . e * . NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 do se 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-do se 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 mi xer. Solidification agent will be transferred to the designated bin from the 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 th a t may consist of an in-line mixer or a sacrificial paddle within the HJC. After fillin g and mixin g are complete, the high-dose so lidification mi xe r will be disenga ge d , and the HIC lidded and prepared for transfer to the high-dose waste deca y subsystem for storage. [Proprietar y 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-dos e 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 pro vi des the low-dose waste tank capacities.

Low-d ose liquid received from other upstream proces ses, combined with the low-dose condensate not rec yc led , will be transferred to the low-dose waste collection tank where the contents of the tank will be analyzed 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 ven tilation air will be passed through the tank. The heated tank contents , plus the high rate of ventilation , will evaporate excess water, reducing the v olume of so lid waste generated. Samples will be collected and anal yzed to ensure compliance with waste acceptance criteria.

9-67

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

NORTlfWEST MEDICAL tsOTDPf.S

[Proprietary Information]

NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-24. Low-Dose Liquid Waste Evaporation System Location 9-68 NWM I ...... e * ! NOmfWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 3 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 e vaporator 4 , 300 16 , 250 WH-TK-420 Low-dose waste accumulation tank 5,900 22,300 WH-TK-500 Low-dose waste evaporation tank (LD-1) 5 , 900 22 , 300 WH-TK-530 Low-dose evaporation tank (LD-2) 2,600 9,800 9-69

. NWMI ...... ..* .. .. ' ... . .. .. .. NQllTlfWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 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 47 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 w i ll 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 was t e 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 NWM I ...... * *

NORTlfWEST MEDICAL ISOTOPES [Propri e t a r y Inform at ion] NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-27. Spent Resin Dewatering Operational Flow Diagram [Propri e t ary In fo rm a ti o n] Figure 9-28. Spent Resin Collection Tanks Location 9-71 NWMI ::::**::*.* ...... *. *. * ' . NORTHWEST MEDICAL ISOTOPES 9.7.2.2.4 Solid Waste Encapsulation NWMl-2013-021, Rev. 3 Chap t er 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 so l id 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 i s 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 lag storage capability for solidified liquid waste and the five-drum rac k s with high-dose source terms. After HICs or five-drum racks have been filled and lidded in the high-d ose 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 ...... ..* .. . ........... * * ' . NOIITTfWEST MEDICAL ISOTOPfS NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxi l iary 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 decay 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 [Propri etary Information]

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

.*;.-.;*.NWMI ...*.. ..* .... ............ *. * * . NORTHWEST MEDICAL lSOTOPfS 9.7.2.2.7 Waste Handling NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems The simplified operational flow diagram for the waste handling s ubsystem 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 lowbo y trailer transporting an empty DOT-approved cask (Figure 9-34). The truck , tra i ler , 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 design a ted impact limiter landing zone and secured. Operators will unbolt the lid and prepare the DOT-approved shipping cask for loading per the cask loading and unloading procedure. At this point, t he truck , trailer, and shipping cask will enter the waste loading area v ia a high-ba y door. The trailer containing the approved shipping cask will be positioned below the high-dose waste shipping 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-bay 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 NWM I ...... * * ' NORTHWEST MEDICAL lSOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 3 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 loadin g 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 lo ading bay. At this point , the facility process control and communications system will allow operators to replace the shipping cask's upper impact limiter using the waste shipping overhead crane (WH-L-1 I 00). 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 wi ll be transferred to the shipping and receiving airlock door, where the empty 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 drums will be transferred via the waste handling pallet jack (WH-PH-1100) from the shipping and receiving airlock door to the waste loading area. The waste handling forklift (WH-PH-1110) wi ll then enter the waste management loading bay via an exterior faci lit y high-bay door. 9-75

.. ; .NWMI *: :**:*:* ...*.. * * !' . NORTHWEST MEDICIJ. lSOTOPES NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems A waste shipping truck wi ll also enter the waste management loading ba y via an exterior facility high-ba y 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 document e d for material tracking and accountability per the safeg uards and security system requirements. The truck containing the waste pallets will exit the RPF through the high-ba y door s 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

[Proprieta ry Information]

and will provide additional waste storage and shipping preparation for Class A radioactive waste prior to di s po sa l. 9. 7.2.3 Operational Analysis and Safety Function Chapter 13.0 , Section 13.2 evaluates the accident se quences that involve fissile solution or solid material s being introduced into syste ms not normally designed to process these solutions or solid materials.

The waste handling system is not geometrically safe; therefore , a number of IROFS have been identified. * * * * * * * * * * * * * * * * *

IROFS RS-01 , " Hot Ce) I Liquid Confinement Boundary" IROFS RS-03 , " Hot Cell Secondary Confinement Boundary" IROFS RS-04 , " Hot Cell Shielding Boundar y" IROFS RS-08 , " Sample and Analysis of Low Dose Waste Tank Do se Rate Prior to Transfer Outside the Hot Cell Shielding Boundar y" IROFS RS-I 0, "Ac tive Radiation Monitoring and Isolation of Low Dose Wa ste 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 Analysi s ofU 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, "Gra m Estimator Survey of Accessible Surfaces for Gamma Activity" IROFS CS-23 , "N on-Destructive Assay (NOA) of Items with Inacces s ible Surfaces" IROFS CS-24 , " Independent NOA of Items with Inaccessible Surfaces" IROFS CS-25 , "Target Housing Weighin g Prior to Di sposa l" 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 processe s associated with the control and storage of radioactive waste are di sc ussed 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 ...... .. .. ......... *.*

e * . . NORTHWEST MEDICAL ISOTOPES NWMl-2013-0 21 , Rev. 3 Chapter 9.0 -Auxi l iary 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 b e performed.

9.7.3.1 Design Basis The RPF analytical laboratory de sign basis is to provide on-site anal ys is to support the production of 99 Mo product and fabrication of tar ge ts for irradiation.

This analysis will be used to determine (1) mass , 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.

Anal ysis 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 s pa ce will include the following:

* * * * *

Hoods to complete sample preparation , waste handlin g, and standards preparation Hoods for specialty in s trument s, including an IC P-MS and kinetic phosphorescence analyzer Glovebox for ICP-MS Gloveboxes for sample delivery and preparation prior to sample transfer to hood s Countertops for the gamma spectroscopy sys tem , low-energy photon spectroscopy, alpha s pectroscopy system, liquid scintillation system, and beta-counting syste m Storage for chemical and laboratory supplies Benchtop syste ms , such as balances , pH meters , and ion-chromatography The analytical laborator y layout is presented in Figure 9-35 and provides space for eight hoods, four gloveboxes , and two countertops. 9-77

... ; .. NWMI -::.**.-.*. .... .. .. .. , '. *. * * . NORTHWEST MEDICAL ISOTOPES [Proprietary Information]

NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-35. Analytical Laboratory Layout Analytical instrumentation will include the ICP-MS , kinetic phosphorescence analyzer , gamma energy analysis , alpha spectroscopy , liquid scintillation spectrometry , and gamma energy analysi s. 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 procedure s , reagents to use, equipment requirements, and disposal protocol for the sample residue material.

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

NORTHWEST MEDICAL ISOTOPES 9.7.3.4 Instrumentation and Control Requirements NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxi l iary Systems Instrumentation and control requir e ments 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 Operatin g License Application.

9. 7.4 Chemical Supply The chemical supply syste m will include tanks s upply i ng aqueous chemicals to the proce ss systems , flammable material storage cabinets used to segregate incompatible materials , and storage of chemical solids used in the process systems. 9.7.4.1 Des i gn Basis The chemical supply syste m design basis i s to provide chemical solutions mixed to the required concentrations that are used within the target fabrication, target dissolution , Mo recover y and purification , and waste mana geme nt systems. The system will provide nitric acid, NaOH, reductant and NO x absorber solutions , hydrogen peroxide, and fresh uranium IX resin. Additional information i s provided in Chapter 3.0 , Section 3.5.2.7. 9.7.4.2 System Description Figure 9-36 shows the layout of the chemical supply room within the RPF. Tanks are sized to provide support to the proce ss requirement

s. 9.7.4.2.1 S u bsystem 100, Nitric Acid Subsystem 100 will consist of five tanks that provide the following functions: * [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

[Proprietary Information]

9-79

.. ; .. ;. NWMI ...... ..* .. . ..... .. .. .. ' * ' NOlTHWtsT MEDICAL lS01lWES [Proprietary Information]

NWMl-2013-021 , Rev. 3 Chap t er 9.0 -Auxiliary Systems Figure 9-36. Chemical Supply Room Equipment Layout 9-80

.; .. ;. NWMI *::**::* ...... NORTHWEST MEOICALISOTOPES NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-37 provides the flow diagram for Subsystem 100 , and Table 9-7 provides a summary description of the tanks in this subsystem.

[Proprietary Information]

Figure 9-37. Nitric Acid Flow Diagram 9-81

.:;:.;* .. NWMI ..*...... * . ......... !:* * *.* ! .' . NORTHWEST MEDICALISOTOPES NWMl-2 0 13-021 , Rev. 3 C h ap t er 9.0 -Aux il i a ry Systems Ta ble 9-7. S ub sys t e m 1 0 0, N it r ic A cid Ta n k Sizes Tank number Chemical CS-TK-100 [Proprietary Information]

CS-TK-13 0 [Proprietary Information]

CS-TK-ISOA

[Proprietary Information]

CS-TK-150B [Proprietary Informa ti on] CS-TK-180A [Proprietary Information]

CS-TK-18 0 B [Pro p rietary Infor m at i on] CS-TK-300

[Proprietary Informat i on] CS-T K-320 [P rop ri etary Infor m at i on] CS-TK-600A/B/C/D [Proprietary Information]

HN0 3 = nitric acid. 9.7.4.2.2 S ub sys t e m 200, S odium Hy dr ox id e Working volume (L) [Proprietary lnfonnati o n] [Proprietary l n fonnation]

[Proprietary Lnfonnati o n] [Propr i etary lnfonnation]

[Propr i et a ry lnfonnati o n] [Pro p r i etary lnfonna ti on] [Propriet a ry Lnfonn a ti o n] [Pro p rietary lnfonnation]

[Proprietary lnfonn a ti o n] I X Total volume (L) [Prop r ietary lnfonnati o n] [Proprietary lnfonnation]

[Propr i et a ry Lnfonn a ti o n] [Pro p rietary Information]

[Proprietary lnfonnati o n] [Proprietary lnfonnation]

[Proprietary Lnfonnati o n] [Proprietary Information]

[Proprietary l nfonnati o n] i o n exchan g e. Su b system 200 w ill consist of three tanks t h at provi d e t h e fo ll owi n g functions:

* *

[Proprietary Information]

[Proprietary Information]

[Proprietary Informat i on] 9-82 Diameter (in.) 120 12 84 84 84 84 18 18

135 20 83 83 1 10 1 1 0 21 21

.. ;.::.; NWMI ..... .*.* .. *.*. e * . NOlllTlfWEST MEDICAL ISOTDf'ES NWMl-2013-021 , Rev. 3 Chapter 9.0 -Auxiliary Systems Figure 9-38 provid es the flow diagram for the NaO H s ub system, a nd Table 9-8 pro v id es a summary de sc ription of the tanks in this subsyste m. [Propri etary Inform atio n] Figure 9-38. Sodium Hydroxide Flow Diagram Table 9-8. Subsystem 200 , Sodium Hydroxide Tank Sizes ' Tank number Chemical ----CS-TK-200

[P roprietary Information]

[Proprietary

[Proprietary 84 96 In formation]

In formation]

CS-TK-230

[Proprietary Information]

[Proprietary

[P ro pri etary 18 18 in fo rm ation] In forma ti on] CS-TK-260 [Proprietary Inform atio n] [P rop ri etary [P rop ri et a ry 24 27 I nformation]

Information]

CS-TK-350

[Propri etary Information]

[Proprietary

[Proprietary 6 8 Inform at ion] Inform a ti on] Na OH sodi um hydroxide.

NO x nitro ge n oxide. 9-83

. ; .. ; NWMI " i:**::* ...... e * ' NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 3 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 summary description of the tanks in Subsystem 300. Table 9-9. Subsystem 300, Reductant and Nitrogen Oxide Absorber Solutions Tank Sizes Tank number Chemical CS-TK-300 [Propr i etary Information]

CS-TK-3 20 [Proprietary Information]

CS-TK-340

[Proprietary Information]

NO x nitrogen ox ide. 9.7.4.2.4 Subsystem 400, Hydrogen Peroxide S ub system 400 will provide the followi ng functions:

* [Proprietary Information]

[Proprietary Information]

Working volume (L) [Proprietary ln format i o n] [Proprietary Information]

[Proprietary In forma ti on] Total volume (L) [Proprietary ln format i on] [Proprietary Information]

[Proprietary In format i on] Diameter (in.) 18 18 6 [Proprietary I nfo rmation] Height (in.) 21 21 8 Figure 9-39 provides the flow di agram for the hydrogen peroxide subsystem.

The su b system will consist of one tank (CS-TK-400), which is s ummari ze d in Table 9-10. Figure 9-39. Hydrogen Peroxide Flow Diagram Tank number CS-TK-400 Table 9-10. Subsystem 400, Hydrogen Peroxide Tank Sizes Chemical Hydrogen peroxide Working volume (L) [Proprietary Information]

9-84 Total volume (L) [Proprietary Information]

Diameter (in.) 9 Height (in.) 12 NWMI *::**:*:*

NOllTlfWEST MEDICAUSOTOPES NWMl-2013-0 21, Rev. 3 Chapter 9.0 -Auxilia ry Systems 9.7.4.2.5 S u bsystem 600, Fresh Uranium Ion Exchange Resin Subsystem 600 will cons i st of four tanks (one tank to support each uranium IX column) that provide the following functions:

I I * * * [Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

Table 9-11 provides a summar y description of the tanks in Subsys tem 600. Tab l e 9-11. S u bsystem 600, Fresh Uran iu m Ion Exchange Resi n Tank Sizes Tank number Chemical CS-TK-600A

[Proprietary Information]

CS-TK-600B

[Proprietary Information]

CS-TK-600C [Proprietary In fo rmation] CS-TK-600D

[Proprietary Information]

I X = ion exchange.

Working volume (L) [Propri e t ary In format i on] [Proprietary In format i on] [P ropr i etary In fo rm a ti on] [Proprietary lnformation]

Total volume (L) [Propri e t ary In fo rm ation] [Proprietary In formation]

[Proprietary In forma ti on] [Proprietary Information]

Diameter (in.) 2 4 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 mak eup act i vity is s ummarized a s follows: * * * * * [Proprietary Information]

[Proprietary Information]

[Proprietar y Information]

[Proprietary Information]

Once resin has been prepared by fines removal and wash in g, the makeup tank will b e adjusted to contain a total vo lume of [Proprietary Information].

The makeup tank low-speed a g itator wi ll be s tarted to s u spend the resin inventory , and the va l ve opened to route the suspens ion to an IX column. 9-85

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

e *

NOllT1fWtsT MEDlCAl ISOTOPES 9. 7.4.3 Operational Analysis and Safety Function NWMl-2013-021, Rev. 3 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:

CS-18, " Backflow Prevention Device" CS-19, "Safe Geometr y Day Tanks" 9. 7.4.4 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

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9.8 REFERENCES

NWMl-2013-021, Rev. 3 Chapter 9.0 -Auxi l iary Systems 10 CFR 20, "Stan dards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended. 10 CFR 50, " Dome stic Licensing of Production and Utilization Facilities," Code of Federal Regulations, Office of the Federal Register, as amended. 40 CFR 61, " Nation a l Emission Standards for Hazardous Air Pollutants," Code of Federal R egu lati ons, Office of the Federal Regi ster, as amended. 42 U.S.C. 2011 et seq., "At omic Energy Act of 1954," United States Code, as amended. ICC , 2012, " International Building Code (IBC) and Commentary 2012," International Code Council, Falls Church, Virginia, 20 1 2. IFC, 2012, Int ernationa l Fire Code, Internation a l Code Council, Falls Church, Virginia, 2012. ISO 14644-1 , "C leanrooms and Associated Controlled Environments

-Part 1: Classification of Air Cleanliness," International Organization for Standardization, Geneva, Switzerland, 1999. NFPA 10, Standard for Portable Fire Extinguishers , National Fire Protection Association, Quincy, Massachusetts, 2013. NFPA 13, Standard for the Installation of Sprinkler Systems, National Fire Protection Association, Quincy, Massachusetts, 2013. NFPA 24, Standard for the In stallation of Private Fire Service Mains and Their Appurtenances, National Fire Protection Association, Quincy, Massachusetts, 2013. NFPA 25, Standard for the In spection, T es tin g , and Maintenance of Water-Based Fire Protection Systems , National Fire Protection Association, Quincy , Massachusetts, 2014. NFPA 45, Standard on Fire Protection for Laboratories Using C h e mical s, National Fire Protection Association, Quincy, Massachusetts, 2015. NFPA 72, Na tional Fire A larm and Signaling Code, National Fire Protection Association , Quincy , Massachusetts, 20 I 3. NFPA 80, Standard for Fire Doors and Other Opening Protectives , National Fire Protection Association, Quincy, Massachusetts , 2013. NFPA 90A, Standard for the In stallation of Air-Conditioning and Ventilating Systems, National Fire Protection Association , Quincy, Massachusetts, 2015. NFPA 92, Standard for Smoke Contro l Systems , National Fire Protection Assoc iation , Quincy , Massachusetts, 2015. NFPA 101, Life Safety Code, National Fire Prot ectio n Association, Quincy, Massachusetts, 2015. NFPA 221, Standard for High Challenge Fire Walls , Fire Walls , and Fire Barrier Walls, National Fire Protection Association, Quincy, Massachusetts, 2015. NRC , 2012, Final Int eri m Staff Guidance Augme ntin g NUREG-1537 , " Guidelines for Preparing and R eviewing App li cations for the Licensing of No n-Po wer R eac tor s ," Parts I and 2 , for Licensing Radioisotop e Production Facilities and Aqueous Homo ge n eous R eac tors, Dock et ID: NRC-2011-0135, U.S. Nuclear Regulator y Commission, Washington, D.C., October 30, 2012. 9-87 NWM I ...... * * . NOllTHWEST MEDtcAl ISOTOf'lS NWMl-2013-021, Rev. 3 Chap t er 9.0 -Auxiliary Systems NUREG-1537 , Guid e lin es for Pr e paring and R ev i ew ing Appli c ations for th e Li ce n s ing of N on-Pow e r R e actors -Format and Content , Part 1 , U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation , Washington, D.C., February 1996. NWMI-2013-039 , Pr e liminary Fir e Ha z ard s Anal y sis , Rev. C, Northwest Med i cal Isotopes , LLC, Corvallis, Oregon , 2015. NWMI-2013-CALC-005 , Tank A ir Bl ee d Estimat e, Rev. B , Northwest Medical I s otopes , LLC , Corvallis , Oregon , 2014. NWMI-2013-CALC-009 , Uranium Purification S y st e m Equipm e nt Si z ing , Rev. B , Northwest Medical Isotopes , LLC , Corvallis , Oregon , 2014. Regulatory Guide 1.189 , Fir e Prot e ction for N ucl e ar Po we r Plants, U.S. Nuclear Regulatory Commission, 2009. 9-88

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Chapter 10.0 -Experimental Facilities Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 3 September 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis , OR 97330 This page intentionally left blank.

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NORTHWEST MEDICAL ISOTOPES NWMl-2013-021 , Rev. 3 Chapter 10.0 -Experimental Facilities Chapter 10.0 -Experimental Facilities Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 3 Date Published:

September 5, 2017 Document Number. NWMl-2013-021 I Revision Number. 3 Title: Chapter 10.0 -Experimental Facilities Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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NORTlfWHTMEDtCIJ..ISOTOPE I NWMl-2013-021 , Rev. 3 Chapter 10.0 -Experimental Facilities This page intentionally left blank.

TERMS A.cronyms and Abbreviations CFR Code of Fe d era l Regulations NWMI Northwest Medical Isotopes , LLC RPF Radioisotope Production Faci l ity NWMl-2013-021 , Rev. 3 Chapter 10.0 -Experimental Facilit i es

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NOATWWUT MEDtCAL lSOTIM'U NWMl-2013-021 , Rev. 3 Chapter 10.0 -Experimental Facilities T h is p age inte n tio n a ll y l eft bl a nk NWMl-2013-021, Rev. 3 Chapter 10.0 -Experimental Facilities 10.0 EXPERIMENT AL FACILITIES This chapter of the Construction Permit Application addressing experimental facilities is not applicable to the Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF). Specifically , NWMI will not ha ve any laboratory-scale facilities designed or used for experimental or analytical purpo ses that relate to the proce ssing of irradiated materials containing special nuclear material per the definition of a production facility in Title 10 , Cod e of F e d e ral R egu lation s (CFR), Part 50.2 , "Definitions." 10-1

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tsOTOPU References NWMl-2013-021, Rev. 3 Chapter 10.0 -Experimental Facilities 10 CFR 50, " Domestic Licensing of Production and Utilization Facilities

," Code of F e d e ral R egu lation s, Office of the Federal Register , as amended. 10-2

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