NWMI-2015-CSE-008, 'NWMI Preliminary Criticality Safety ... · kg L lb m M mz mol wt% [Proprietary...

NWMl-2016-RAl-004, Rev. 0

Attachment C

NWMI-2015-CSE-008, NWMI Preliminary Criticality Safety Evaluation: Hot Cell Uranium Purification (Recovery and Recycle) (Rev. A) (Public Version)

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NWMl-2016-RAl-004, Rev. 0

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Report No:

Revision No:

Report Cover Sheet

NWMl-2015-CSE-008

A

Report Title: NWMI Preliminary Criticality Safety Evaluation: Hot Cell Uranium Purification (Recovery and Recycle)

Project Title: NWMI Radioisotope Production Facility

Status: ~ In Process D Final

Contains assumptions and/or inputs that require verification? 0Yes ~No

Approvals

. · >Revi~w (R) aria · · · . Approval (A)? · i;

- ·-' . : '' ·oate

• • ~ ' ' ' ,. ", ' r • ,; ' ' •

Originator A S. J. Wachtel 6/15/15

Project Manager RIA C. Haass 6/29/15

Registered Professional Engineer's Stamp (if required) ~ N/A

Revision History

Revision· i:>escriptio~ .. .. , , '·

. Affected Pages .. ',;.· ·' •. v':

A 6/29/15 Initial Issue All

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e. • @• • • NORTHWEST MEDICAL ISOTOPES • •

NWMI Preliminary Criticality Safety Evaluation:

Prepared by:

Hot Cell Uranium Purification

NWMl-2015-CSE-008, Rev. A June 2016

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

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NWMl-2015-CSE-008 Rev.A

NWMI Preliminary Criticality Safety Evaluation: Hot Cell Uranium Purification

NWMl-2015-CSE-008, Rev. A

Date Published: June 29, 2015

Document Number: NWM 1-2015-CSE-008 I Revision Number. A

Title: NWMI Preliminary Criticality Safety Evaluation: Hot Cell Uranium Purification

Approved by: Carolyn Haass Signature:

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CONTENTS

1.0 IN'fRODUCTION ....................................................................................................................... 1 1.1 Scope ................................................................................................................................ l

2.0 NORMAL OPERATING CONDIDONS ..................................................................................... 2 2.1 Process Materials .............................................................................................................. 2

2.1.1 Uranyl Nitrate Solution ....................................................................................... 2 2.1.2 Nitric Acid .......................................................................................................... 2 2.1.3 Ferrous Sulfamate ............................................................................................... 2 2.1.4 Sulfarnic Acid ..................................................................................................... 2 2.1.5 Ion Exchange Media ........................................................................................... 2 2.1.6 Chilled Water ...................................................................................................... 3 2.1.7 Plant Stearn ......................................................................................................... 3 2.1.8 Plant Gases ......................................................................................................... 3 2.1.9 Filter Media ........................................................................................................ 3

2.2 Facility Description ........................................................................................................... 4 2.3 Process Description ........................................................................................................... 6

2.3.1 Impure Uranium Collection ................................................................................. 6 2.3.2 Uranyl Nitrate Feed Preparation or Blending for Ion Exchange #1 ....................... 6 2.3.3 Ion Exchange #1 .. : .............................................................................................. 6 2.3.4 Evaporation or Uranium Concentrator Feed Tank #1 ........................................... 7 2.3.5 Uranyl Nitrate Feed Preparation or Blending for Ion Exchange #2 ....................... 8 2.3.6 Ion Exchange #2 ............................................................................... : ................. 8 2.3. 7 Evaporation or Concentration #2 ......................................................................... 8 2.3.8 Recycled Uranium Collection Tanks ................................................................... 9 ' 2.3.9 Uranium Decay Holdup ....................................................................................... 9 2.3.10 Condensate Tanks #1 .......................................................................................... 9 2.3.11 Ion Exchange Waste Collection Tanks ............................................................... 10 2.3.12 UraniumReworkTank ...................................................................................... 10 2.3.13 Resin Handling System ..................................................................................... 10

2.3.13.1 Spent Resin ....................................................................................... 11 2.3.13.2 New Resin ....................................... · ................................................. 11

2.4 Equipment Description .................................................................................................... 12 2.4.1 Uranyl Nitrate Feed Solution Lag Storage ......................................................... 12 2.4.2 Uranyl Nitrate Feed Preparation or Blending for Ion Exchange (Stages 1

and 2) ................................................................................................................ 12 2.4.3 Ion Exchange Columns (Stages 1and2) ............................................................ 13 2.4.4 Concentrator Feed Tanks (Stages 1and2) ......................................................... 13 2.4.5 Concentrator System (Stages 1and2) ............................................................... 13 2.4.6 Purified Product Uranyl Nitrate Tanks ............................................................... 14 2.4.7 Decay or Lag Storage Tanks .............................................................................. 14 2.4.8 Ion Exchange Liquid Waste Collection Tanks ................................................... 14 2.4.9 Condensate Tanks (Stages 1and2) .................................................................... 14 2.4.10 Spent Resin Tanks ............................................................................................. 15 2.4.11 Fresh Resin Tank .............................................................................................. 16 2.4.12 Pumps ............................................................................................................... 17

2.5 Evaluation of Normal Conditions .................................................................................... 17

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3.0 NUCLEAR CRITICALITY HAZARD IDENTIFICATION ....................................................... 18 3.1 HazardidentificationMethod .......................................................................................... 18 3.2 Hazard Identification Results .......................................................................................... 18

4.0 CONTINGENCY ANALYSIS ................................................................................................... 22 4.1 Double Contingent Scenarios .......................................................................................... 22

4.1.1 Overfill of a Tank Results in Fissile Solution in the O:ffgas Treatment System .............................................................................................................. 22 4.1.1.1 Description ....................................................................................... 22 4.1.1.2 Initiating Event ................................................................................. 22 4.1.1.3 Primary Contingency ........................................................................ 22 4.1.1.4 Secondary Contingency ..................................................................... 23 4.1.1.5 Common Mode Failure Potential ....................................................... 23 4.1.1.6 Summary .......................................................................................... 23

4.1.2 Fissile Solution Leaks into Steam or Chilled Water Systems .............................. 23 4.1.2.1 Description ....................................................................................... 23 4 .1.2.2 Initiating Event ................................................................................. 24 4.1.2.3 Primary Contingency ........................................................................ 24 4.1.2.4 Secondary Contingency .................................................................... 24 4.1.2.5 Common Mode Failure Potential ....................................................... 24 4.1.2.6 Summary .......................................................................................... 24

4.1.3 Fissile Solution Forced into the Fresh Resin System .......................................... 25 4.1.3.l Description ....................................................................................... 25 4.1.3.2 Initiating Event ................................................................................. 25 4.1.3.3 Primary Contingency ........................................................................ 25 4.1.3.4 Secondary Contingency .................................................................... 25 4.1.3.5 Common Mode Failure Potential.. ..................................................... 26 4.1.3.6 Summary .......................................................................................... 26

4.1.4 Fissile Solution Forced into Process Gas System ............................................... 26 4.1.4.1 Description ....................................................................................... 26 4.1.4.2 Initiating Event ................................................................................. 26 4.1.4.3 Primary Contingency ........................................................................ 26 4.1.4.4 Secondary Contingency .................................................................... 27 4.1.4.5 Common Mode Failure Potential.. ..................................................... 27 4.1.4.6 Summary .......................................................................................... 27

4.2 Not Credible/Criticality Not Credible Scenarios .............................................................. 27 4.2.1 Fissile Solution Leaks from Vessels .................................................................. 27 4.2.2 System Tanks are Larger than Intended ............................................................. 28 4.2.3 Process Vessels Dislodged from Mounts and Move or Fall Closer

Together. ........................................................................................................... 29 4.2.4 Fissile Solution Forced into Chemical Reagent or Water Supply Systems .......... 29 4.2.5 Incompatible Resin is Installed .......................................................................... 30 4.2.6 Excessive Uranium in Ion Exchange Liquid Waste, Condensate, or Spent

Resin Collection Tanks ...................................................................................... 30 4.2.7 Excessive Concentration in System Tanks ......................................................... 31 4.2.8 Hot Cell Floods ................................................................................................. 32

4.3 Nuclear Criticality Safety Parameters .............................................................................. 33 4.4 Defense Table ................................................................................................................. 33

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5.0 CRITICALITY ACCIDENT ALARM SYSTEM ....................................................................... 35

6.0 DOUBLE CONTINGENCY CONTROLS ................................................................................. 36 6.1 Passive Design Features .................................................................................................. 36 6.2 Active Engineered Features ............................................................................................. 38 6.3 Administrative Controls .................................................................................................. 38 6.4 General Requirements ..................................................................................................... 3 8

7.0 LIST OF ASSUMPTIONS AND OPEN ITEMS ........................................................................ 39 7 .1 Assumptions ................................................................................................................... 39 7.2 Open Items ..................................................................................................................... 39

8.0 CONCLUSION .......................................................................................................................... 40

9.0 REFERENCES .......................................................................................................................... 41

Figure 1-1.

Figure 2-1.

Figure 2-2.

Figure 2-3.

Figure 2-4.

Figure 2-5.

Figure 2-6.

Figure 2-7.

Figure 2-8

Table 3-1.

Table 4-1.

Table 4-2.

FIGURES

[Proprietary Information] ................................................................................................ 1



[Proprietary Information] .............................................................................................. 12





[Proprietary Information] ...................................................................................... ." ....... 16

TABLES

Uranium Purification System What-If Upset Analysis (4 pages) ................................... 18

Nuclear Criticality Safety Parameters ........................................................................... 33

Summary of Primary and Secondary Contingencies (2 pages) ....................................... 33

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TERMS

Acronyms and Abbreviations 99Mo molybdenum-99 99Tc technetium-99 235U uranium-235

AAC AC AEF CSE DBE

HAZOP

Mo NWMI O.D. P&ID PDF PFD QA QRA RPF SPL u UN

USL Units oc cm cm2

cm3

ft ft2 g gmol lll.

in.2 kg L lb m M mz mol wt%

[Proprietary Information] augmented administrative control administrative control active engineered feature criticality safety evaluation design basis event [Proprietary Information] hazard and operability nitric acid sulfamic acid ion exchange low-enriched uranium molybdenum Northwest Medical Isotopes, LLC outside diameter piping and instrumentation diagram passive design feature process flow diagram quality assurance qualitative risk analysis radioisotope production facility single parameter limit uramum uranyl nitrate [Proprietary Information] upper subcritical limit

degrees Celsius centimeter square centimeter cubic centimeter feet square feet gram gram-mol inch square inch kilogram liter pound meter molar square meter mole weight percent

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1.0 INTRODUCTION


The Northwest Medical Isotopes, LLC (NWMI) Radioisotope Production Facility (RPF) is being designed to chemically extract and purify molybdenum-99 (99Mo) from irradiated low-enriched uranium (LEU) reactor targets. The 99Mo decays to technetium-99 (99Tc) for use in medical procedures. The uranium (U) recovery and recycle system will take the liquid effluents from the molybdenum (Mo) recovery process and selectively separate and recover uranium for recycle from the effluent stream. [Proprietary Information] (evaluated in NWMI-2015-CSE-005, NWMI Preliminary Criticality Safety Evaluation: Target Fabrication Uranium Solution Processes). The liquid wastes generated from the U recovery and recycle system will be collected and treated for cementation, decay holding, and eventual disposal (evaluated in NWMI-2015-CSE-009, NWMI Preliminary Criticality Safety Evaluation: Liquid Waste Processing).

1.1 SCOPE

This criticality safety evaluation (CSE) analyzes the activities and equipment associated with the chemical processes that purify the uranyl nitrate (UN) solution after Mo extraction, including various staging tanks, ion exchange (IX), and concentration (NWMl-2013-034, Uranium Recovery and Recycle Process Descriptions, PFD and P&ID). Figure 1-1 provides an overview of the [Proprietary Information] and depicts the process flow.

Criticality safety evaluation of liquid waste processing and the ventilation and off gas treatment system downstream from the process tanks are not within the scope of this evaluation (evaluated in NWMI-2015-CSE-009 and NWMI-2015-CSE-Ol l, NWMI Preliminary Criticality Safety Evaluation: Offgas and Ventilation, respectively).

[Proprietary Information] [Proprietary Information]

Figure 1-1. [Proprietary Information]

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2.0 NORMAL OPERATING CONDITIONS


This section discusses the NWMI RPF materials, facility, equipment, and normal operating conditions related to uranium purification.

2.1 PROCESS MATERIALS

2.1.1 Uranyl Nitrate Solution

UN will be the uranium form in the recovery process. UN originates from the Mo recovery process and carries the various fission products from target irradiation that must be removed to reuse the uranium. [Proprietary Information]

The maximum uranium enrichment allowed by the RPF license for [Proprietary Information] uranium-235 {235U).

2.1.2 Nitric Acid

Nitric acid (HN03) will be used as the carrier agent for dissolved uranium and added to achieve the appropriate process chemistry. Plant supply will be acquired at [Proprietary Information], but will be diluted per individual process needs. Nitric acid will be used, as necessary, to adjust the UN concentration, flush solution between transfers, and as the eluent and regeneration washes for the IX columns.

Nitric acid will be supplied at different molarities from facility bulk supply tanks in the chemical makeup room.

2.1.3 Ferrous Sulfamate

[Proprietary Information]. This solution will be supplied to the IX feed tanks by pipe from the chemical makeup room outside the hot cell and circulated to mix the solution. Ferrous sulfamate will be supplied to the RPF as a solid and dissolved in water in the chemical makeup area to provide solution for delivery to the hot cell.

2.1.4 Sulfamic Acid

[Proprietary Information]. This solution will be supplied to both IX feed tanks by pipe from the chemical makeup room outside the hot cell and circulated to mix the solution. Sulfamic acid will be supplied to the RPF as a solid and dissolved in water in the chemical makeup area to provide solution for delivery to the hot cell.

2.1.5 Ion Exchange Media

[Proprietary Information] 1 [Proprietary Information].

1 [Proprietary Information]

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2.1.6 Chilled Water


The process chilled water system will be a central process chilled water loop that cools the three secondary loops:

• [Proprietary Information] • [Proprietary Information] • [Proprietary Information]

The central process chilled water loop will rely on three air-cooled chillers, each sized to accommodate [Proprietary Information] of the process cooling demands. The secondary loops will be cooled by the central chilled water system 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 loops.

The chilled water system will operate with cascading pressure differentials. The central system will operate at the highest pressure, and the secondary loops will operate at a pressure between the central system and the process fluid. The large-geometry secondary loop in the hot cell will meet the cooling demands where fissile material leaking through a heat exchanger is not a credible event. The other cooling loops will be inherently criticality-safe by geometry, so active controls will not be required to keep fissile material out of the chilled water return. At each process cooling demand where fissile material may be present, conductivity sensors will monitor the chilled water return to detect heat exchanger leaks.

2.1.7 Plant Steam

The process steam system will be divided into a medium-pressure central heating loop and a low-pressure secondary loop within the hot cell. Medium pressure steam will be generated by a natural gas-fired boiler. Low-pressure steam in the secondary loop will be generated by medium-pressure steam in a shelland-tube heat exchanger. Medium-pressure steam will be at least [Proprietary Information], to provide an adequate temperature differential to generate [Proprietary Information] steam for the low-pressure steam loop.

The process steam system will include criticality safe vessels that interface with the uranium-bearing process vessels. Steam condensate will be routed to the hot cell low-dose liquid waste system for recycle.

2.1.8 Plant Gases

Plant air and/or nitrogen may be used as motive force for pumps or for system purge. Air and nitrogen will be supplied from the plant feeds, through dryers and/or oil dryers upstream.

2.1.9 Filter Media

[Proprietary Information]

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2.2 FACILITY DESCRIPTION

The U recovery and recycle process, including the UN feed from Mo recovery staging or decay tanks, will be housed completely within




the RPF tank hot cell (Figure 2-1 ). The process equipment will be installed on removable skids of safegeometry vessels, with related process equipment generally grouped or co-located (Figure 2-2). The skids are designed to be removable from the hot cell and sized to limit the corresponding size of the overhead cover blocks that must be removed to allow the skids to be lifted out of the hot cell.

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The RPF tank hot cell will be the primary radiation shield protecting operators from the strong radiation emitted from the fission products in the process solutions. [Proprietary Information]. The ceiling will be equipped with cover blocks that can be lifted off to allow access to specific equipment skids within the hot cell.

The tanks and solution process vessels will typically be constructed on supporting framework skids that are anchored to the floor in designated positions. The skids and associated structural frames will provide seismically qualified support to the fissile-bearing equipment.

The room floor will be concrete and with a stainless steel liner to prevent chemical and radiological contamination migration into the concrete. The floor will be verified flat without drains or collection points to prevent significant pooling and collection of fissile solution to an unsafe depth.

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The area is assumed to be equipped with overhead wet-pipe fire suppression sprinklers for this analysis. The process water, steam, and reagent (e.g., nitric acid, ferrous sulfamate) supply piping will run into the hot cell, with the UN transfer piping from tank system to tank system. Process gas lines will also serve the hot cell.

2.3 PROCESS DESCRIPTION

Activities associated with the uranium purification process include the following steps (NWMI-2013-034):

1. Lag storage of feed solution from Mo recovery 2. Solution preparation and Stage 1 IX 3. Concentration or evaporation Stage 1 4. Solution preparation and Stage 2 IX 5. Concentration or evaporation Stage 2 6. Lag storage for decay 7. IX resin handling 8. Liquid waste handling

2.3.1 Impure Uranium Collection

Feed to the U recovery and recycle system will consist of U-bearing solutions generated by the first cycle of the Mo purification system and accumulated in the impure U collection tanks. This series of vessels will provide a lag storage capability between the Mo purification and the U recovery system equipment. The lag storage time will help reduce the [Proprietary Information]. The U-bearing solution will have a nominal composition of [Proprietary Information].

The collection tank bank will be divided into four pairs of tanks to allow for separated collection and staging of at least [Proprietary Information] of impure UN from the Mo recovery process. The tanks will allow for [Proprietary Information] of decay time before the contents are transferred into the uranium recovery feed tank. The vessel contents will be maintained at a nominal temperature of [Proprietary Information] by cooling jackets while residing in the lag storage tanks to remove radiolytic decay heat.

2.3.2 Uranyl Nitrate Feed Preparation or Blending for Ion Exchange #1

UN feed from the impure uranium collection or lag storage tanks will be transferred in batches to the IX feed tank # 1 to allow the feed to be chemically prepared for the IX process. Solution from the impure uranium collection tanks will be adjusted to a composition of [Proprietary Information]. Reductant will also be added to each feed batch, [Proprietary Information]. Blending will be effected by a circulation pump.

2.3.3 Ion Exchange #1

The uranium recovery columns will operate by passing a sequence of solutions through the IX media. Column effluents will be routed to different vessels during a process cycle, depending on the ions present in the effluent. The column cycle operations include the following:

• Loading cycle - Adjusted solution from the IX feed tanks will be fed to the uranium purification column during the loading cycle to capture uranium in the liquid phase on the IX media, allowing contaminants (e.g., fission products and plutonium) to pass through the column. [Proprietary Information]. Column effluent during the loading cycle will contain a small fraction of the feed uranium and most of the contaminants. The column effluent will be routed to the IX waste collection tanks during the loading cycle. [Proprietary Information].

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NWM 1-2015-CSE-008 Rev.A

•

•

•

Pre-elution rinse cycle - Once the loading cycle is complete, the uranium purification column feed will switch to a solution containing [Proprietary Information]to flush the residual loading cycle feed solution from the column liquid holdup. Effluent from the uranium purification column will be routed to the IX waste collection tanks during the pre-elution rinse cycle, because liquid holdup in the column is considered a solution potentially contaminated with uranium. [Proprietary Information].

Elution cycle - Once the pre-elution rinse cycle is complete, the uranium purification column feed will be switched to a solution containing [Proprietary Information] that transfers UN from the media to the liquid phase passing through the column. Effluent from the uranium purification column will be routed to the uranium concentrator feed tank # 1 during the elution cycle. The selected eluent volume will be sufficient to flush any desorbed UN from the column liquid holdup by the time the elution cycle is complete. [Proprietary Information].

Regeneration cycle - The regeneration cycle will prepare the uranium purification media to perform a new loading cycle by replacing the liquid phase with a solution composition similar to the adjusted impure uranium feed solution. The column feed will be switched to a solution containing [Proprietary Information], which will be used to displace any residual liquid holdup that may be present at approximately [Proprietary Information]. Effluent from the uranium purification column will be routed to the IX waste collection tanks during this cycle, and the effluent composition can be characterized as a solution that is on the order of [Proprietary . Information].

The projected flows and volumes are based on a two-column system operating in parallel, with a [Proprietary Information].

[Proprietary Information], which is used as the column operating conditions. Temperature control will be provided for column feed streams and not on the IX column (no cooling jacket on column).

2.3.4 Evaporation or Uranium Concentrator Feed Tank #1

Uranium-bearing solutions in IX column effluents during the elution cycle will be concentrated as the solutions are generated to control the stored volume of process solutions. Eluate from IX column # 1 will be routed to the U concentrator feed tank # 1. This vessel will provide an interface between the column and concentrator that controls the concentrator feed rate. The capability to add water to the concentrator feed tank will enable control of the concentrate acid concentration.

Uranium-bearing solution for purification will originate from elution oflX column #1, with a solution composition of [Proprietary Information]. Overhead vapors from the concentrator will be routed to a condenser.

The evaporator boiler will be heated by steam from the house supply and the offgas condenser cooled by the house chilled water system. Condensate from the system will be sent to the condensate collection tank for sampling and eventual processing for disposal (evaluated in NWMI-2015-CSE-009). [Proprietary Information].

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2.3.5 Uranyl Nitrate Feed Preparation or Blending for Ion Exchange #2


Concentrate from uranium concentrator/condenser #1 will be collected in one of two tanks that are used to alternate between collecting concentrate and feeding to IX column #2. After collecting a batch of concentrate, the solution will be prepared for feeding the IX column by adding a reductant to modify the valence state of plutonium remaining in the solution. The reductant will be based on addition of ferrous sulfa.mate, stabilized by sulfamic acid. [Proprietary Information].

A majority of the radionuclides will be separated from the U-bearing solution by IX column #1, and radiolytic decay heat will not be significant in this vessel. However, a cooling jacket will be required to control temperature at the [Proprietary Information], as concentrate from concentrator/ condenser # 1 is accumulated in a feed adjustment tank.

2.3.6 Ion Exchange #2

The IX column #2 operation will be similar to IX column # 1 and consist of a sequence of solutions that pass through the IX media.

• Loading • • •

Pre-elution rinse Elution Regeneration

The column effluent will be routed to the IX waste collection tanks during the loading and pre-elution rinse cycles, and the composition is projected to contain trace uranium and trace concentrations of plutonium. Effluent from the uranium purification column will be routed to the uranium concentrator feed tank #2 during the elution cycle. The selected eluent volume will be sufficient to flush any desorbed UN from the column liquid holdup by the time the elution cycle is complete. [Proprietary Information].

Column sizing for IX column #2 was assumed to be identical to IX Column # 1. [Proprietary Information].

2.3. 7 Evaporation or Concentration #2

Eluate from IX column #2 will be routed to the U concentrator feed tank #2. This vessel will provide an interface between the column and concentrator that allows control of the concentrator feed rate. The capability to add water to the concentrator feed tank will enable control of the concentrate acid concentration.

Uranium concentrator/condenser #2 will be similar to uranium concentrator/condenser #1 [Proprietary Information]. The concentrate will be transferred to the recycled uranium collection and adjustment tanks.

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Overhead vapors from the concentrator will be routed to a [Proprietary Information]. Condensate from the condenser is predicted to be characterized as a [Proprietary Information].

The evaporator boiler will be heated by steam from the house supply, and the o:ffgas condenser cooled by the house chilled water system. Condensate from the system will be sent to the condensate collection tank for sampling and eventua.I processing for disposal (evaluated in NWMI-2015-CSE-009). Product UN will be cooled through an in-line heat exchanger as the UN is moved to the Stage 2 IX feed tanks.

2.3.8 Recycled Uranium Collection Tanks

The recycled uranium collection tanks will provide lag storage capability between the uranium purification and uranium decay holdup tanks. The solution entering the vessels will originate as concentrate from uranium concentrator/condenser #2. [Proprietary Information].

Three individual tanks will be provided for recycled uranium product collection. The recycled uranium collection tanks fulfill three primary functions:

•

•

Concentrate receiver tank - Will accumulate recycled uranium batches generated by uranium concentrator #2and provide holdup of uranium solution as generated by the concentrator to create solution batches that can be periodically transferred to the product sample tank.

Product sample tank - Will verify that the recycle uranium complies with product specifications; provides a vessel for sampling an accumulated batch of concentrate from uranium concentrator #2. The sample vessel will provide a location for sampler installation, and a holdup location while the uranium product batch sample is analyzed. The tank will be equipped to divert the sampled solution to a rework tank, if sample analysis indicates that the product batch does not comply with product specifications.

• Recycled uranium transfer send tank - Will perform accountability measurements on uranium crossing a facility licensing boundary; provides a vessel for performing measurement of the uranium mass that is transferred between the uranium purification and target fabrication systems.

A [Proprietary Information] is currently specified for solution stored in the recycled uranium collection tanks, and cooling jackets will be included to cool concentrate stored in the product sample and recycle uranium transfer send tanks.

2.3.9 Uranium Decay Holdup

The uranium solution product from the uranium recycle system must be held within a shielded bank of tanks to allow for [Proprietary Information] (Section 2.3.1).

2.3.10 Condensate Tanks #1

Condensate tanks # 1 will provide an interface point for monitoring condensate generated by uranium concentrator/condenser # 1 prior to transfer to the liquid waste handling system. Equipment in the uranium purification system will be of geometrically favorable design for criticality control, while it is anticipated that the waste handling system equipment will use an alternate criticality control philosophy (e.g., mass control). The condensate tanks will provide a location for verifying that solutions comply with waste handling criticality control requirements.

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Condensate generated by eluate concentration will represent a [Proprietary fuformation] [Proprietary fuformation]. Therefore, online monitoring will be employed for condensate transfers to the liquid waste handling system. Continuous monitoring of the uranium concentration in the condensate tank system will be provided by a sample loop to a uranium concentration detector (e.g., fluorimeter; equipment TBD in final design). Circulation to the detector will be operated at flow rates that allow sample tanks to approximate a continuous, stirred tank flow pattern. The detector will control the routing of transfers out of the condensate tank system; solution transfers out of the condensate system will be routed to waste handling (nonfavorable geometry; evaluated in NWMI-2015-CSE-009) as long as condensate uranium concentrations comply with criticality control requirements. [Proprietary fuformation]. Operation in this recycle mode will continue until the off-normal conditions causing the high uranium condensate concentrations are corrected.

[Proprietary fuformation].

2.3.11 Ion Exchange Waste Collection Tanks

The IX waste collection tanks will provide lag storage capability for waste streams generated by the uranium purification IX columns. The tanks will act as an interface between the uranium purification and waste handling systems as a transition between vessels where criticality control is maintained by geometry and vessels in waste handling that use an alternate criticality control strategy (e.g., mass control). The solutions entering this vessel will originate as effluent from the uranium purification IX columns during the loading, pre-elution rinse, and regeneration cycles. [Proprietary fuformation].

[Proprietary fuformation] (evaluated in NWMI-2015-CSE-009). Monitoring will be limited to uranium content for criticality control purposes. Other monitoring for waste disposal purposes will be performed in the waste handling system.

2.3.12 Uranium Rework Tank

The uranium rework tank will provide the capability to divert out-of-specification recycled uranium detected in the product sample tank to the uranium IX adjustment tanks in the second-cycle uranium purification system. This recycled uranium will be prepared as feed to IX column #2.

2.3.13 Resin Handling System

The IX resin beds will periodically require replacement, as most resin gradually degrades due to exposure to chemicals and radiation. The degradation will reduce the resin uranium capacity and the loading cycle volume (decreasing the process throughput rate) or decrease the effectiveness of uranium separation from unwanted fission products. The resin replacement vessels will support removal of spent resin from an IX column and the addition of fresh resin to a column after spent resin has been removed. The resin replacement vessels will be a combination of tanks located inside and outside of the hot cell to minimize space allocation in the hot cell. The resin replacement vessels will include spent resin collection tanks and a transfer liquid storage tank located inside the hot cell. Fresh resin makeup tanks will be located outside the hot cell.

10

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2.3.13.1 Spent Resin


The spent resin collection tanks will be provided to support removal of spent resin from the IX columns and sampling resin prior to the transfer of resin to the waste handling system for disposal. The spent resin collection tanks are designed with geometrically favorable dimensions to control the potential for criticality. Sampling or monitoring of the spent resin uranium content will be required prior to transfer to the waste handling system, where vessels are not expected to be designed to dimensions that control criticality by geometry. [Proprietary Information]. The spent resin collection tank operation will be supported by a resin transfer liquid tank to manage liquids in the resin slurry during transfers.

The spent resin exchange operational sequence, described in detail in NWMI-2013-034, includes:

• • • • •

Fluidize the resin bed using backflow of water Pump the resin slurry to the resin collection tank Sample the resin via in-line monitoring for uranium content Fluidize the resin and pump to the waste handling process (evaluated in NWMI-2015-CSE-009) Transfer any excess liquid wastes to the IX waste collection tanks

2.3.13.2 New Resin

The fresh resin makeup tanks will support preparation of fresh resin for addition to an IX column after spent resin has been removed. The fresh resin makeup tanks will be located outside the hot cell an~ will not contain materials that have been contacted with uranium or fission products. Therefore, these vessels are not designed using dimensions to control the potential for criticality.

The new resin loading operational sequence, described in detail in NWMI-2013-034, includes:

• • • • • •

Manually add new resin to a makeup tank and fluidize Filter the resin slurry to remove fmes Wash the resin Pump the fluidized, washed resin to an IX column in the hot cell Perform "regeneration" cycle to prepare resin for first use Transfer any excess liquid wastes to the IX waste collection tanks

11

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. •. !~.~!.'. NORTHWESTtJIEOICALISOTOPES

2.4 EQUIPMENT DESCRIPTION


Typical tanks will be constructed using [Proprietary Information] multiple individual safe-geometry, spaced F" 2 3 [P · t I fi f ] risers Or columns that are interconnected to igure - " roprie ary n orma IOn form a single operational tank unit that is filled and drained using piping common to the bank of risers. [Proprietary Information] (see Figure 2-3).

The tank risers will be mounted together on a supporting framework and skid, anchored to the room floor, and seismically designed to withstand the RPF DBE. Individual tank skids in the hot cell will be equipped with remote-operable, quick-connect fittings and anchor points and lifting points to allow the skids to be lifted out of the hot cell for maintenance or replacement.

2.4.1 Uranyl Nitrate Feed Solution Lag Storage

The lag storage tanks will be typical pencil tank risers mounted on skids. [Proprietary Information] will be provided for separate lag collection for three weeks of Mo recovery operations. The tanks will be cooled by chilled water jackets to remove residual radiolytic heating. Connections will be provided for solution receipt from Mo recovery, circulation of the tank contents, and transfer out of the system, and for the cooling water supply and return connections. The tanks will be vented to the process vessel vent offgas treatment system.

2.4.2 Uranyl Nitrate Feed Preparation or Blending for Ion Exchange (Stages 1 and 2)

The IX feed preparation tanks will be typical skid-mounted risers. Chemical and water connections will be included in the skid connections to allow for process chemistry solution adjustments. Typical connections will be for solution receipt from lag storage, circulation of the tank contents, and transfer out of the system. The tanks will be vented to the process vessel vent off gas treatment system.

12

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. ~ !~.:~:. NORTHWESTMED!CAl ISOTOPES

2.4.3 Ion Exchange Columns (Stages 1 and 2)

The IX columns are currently envisioned to have a [Proprietary Information] for criticality control, with the IX media supported on a screen to form a resin bed (Figure 2-4). An upper screen will be included in the column to restrain the resin within a fixed portion of the column. Inlet and outlet






piping connections will communicate with the resin section of the column to allow periodic bed replacement using slurry transfer of the resin. The current concept will be based on providing a configuration with two of the columns that operate in parallel for each of the IX cycles.

Liquid phase will pass through the column in a down-flow such that feed for a particular column cycle will enter at the top of the column, and cycle effluents leave the column from the bottom. The column is anticipated to include a rupture disk-type safety pressure relief assembly as part of the column design. Pressure relief capabilities are typically required when using organic resins in a nitric acid system.

Each IX column will be mounted on a skid similar to typical tank skids, with piped solution connections to allow the process cycles and resin exchanges. The columns will each be vented to the process vessel vent offgas treatment system.

2.4.4 Concentrator Feed Tanks (Stages 1 and 2)

The concentrator feed tanks will be typical pencil tank risers mounted on skids. Connections will be provided for solution receipt from the IX column stages, circulation of the tank contents, and transfer out of the system. The tanks will be vented to the process vessel vent off gas treatment system.

2.4.5 Concentrator System (Stages 1and2)

The concentrator systems will be identical sets of evaporators and condensate holding tanks. Each stage train will include an evaporative reboiler with a steam-heated boiler, pencil tank reservoir, offgas cooler/condenser, and product cooler heat exchanger. The tank components will be safe-geometry pencil tanks with [Proprietary Information]. The system equipment will be mounted on a supporting framework and skid, anchored to the room floor, and seismically designed to withstand the RPF DBE (Figure 2-5).

13

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The boiler will be a steam-heated heat exchanger, [Proprietary Information]. The condenser and product cooler will both be [Proprietary Information]. The condenser will be mounted above the boiler, with the product cooler mounted beside the reservoir column.

A small [Proprietary Information] centrifugal pump will transfer the UN to the next IX feed (to Stage 2) or the purified product tanks (from Stage 2).

2.4.6 Purified Product Uranyl Nitrate Tanks

The UN from the second-stage evaporator will be staged in a bank of tank risers for sampling and eventual transfer to decay storage. The tanks will be safe-geometry pencil tanks with [Proprietary Information]. The tanks will be mounted to supporting frameworks on skids, anchored to the room floor, and seismically designed to withstand the RPF DBE.

2.4. 7 Decay or Lag Storage Tanks

This bank of [Proprietary Information] will serve as a lag storage area to allow recycled uranium from the irradiated targets a nominal [Proprietary Information]. The tanks will be arranged in a [Proprietary Information] mounted on a supporting framework and skid, anchored to the room floor, and seismically designed to withstand the RPF DBE.

2.4.8 Ion Exchange Liquid Waste Collection Tanks

Both IX systems will use the same banks of liquid waste collection tanks, as will the resin exchange system. The waste collection tanks will be divided into two separate banks to allow one bank to be online to receive liquid waste while the other is sampled and waiting for analysis results to allow transfer to the liquid waste processing system (evaluated in NWMI-2015-CSE-009).

Both banks of tanks will include multiple risers that are mounted on typical frame skids, with transfer connections and water flush connections. The tanks will be vented to the process vessel vent offgas treatment system.

2.4.9 Condensate Tanks (Stages 1 and 2)

The condensate collection tanks for each concentrator stage will provide continuous uranium content monitoring of the effiuent condensate as the solution is passed to the liquid waste processing system. Due to the volume of condensate generated and the space constraints of the tank hot cell, the condensate will not be collected and sampled; continuous monitoring will protect the downstream non-safe geometry process vessels in the liquid waste processing system (evaluated in NWMI-2015-CSE-009).

Each condensate collection stage will consist of [Proprietary Information] (Figure 2-6). Continuous monitoring of the uranium concentration in condensate sample tank #IA will be provided by a sample loop to a uranium concentration detector (e.g., fluorimeter). Circulation to the detector will be operated at flow rates that allow sample tanks to approximate a continuous, stirred tank flow pattern. The detector on Tank #IA will control the routing of transfers out of condensate sample tank #lB. Solution transfers out of condensate sample tank # lB will be routed to waste handling as long as condensate uranium concentrations comply with criticality control requirements.

14

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The tank vessels will be mounted together on a supporting framework and skid, anchored to the room floor, and seismically designed to withstand the RPF DBE. The tanks will be vented to the process vessel vent offgas treatment system.

2.4.10 Spent Resin Tanks

The spent resin collection tanks will support removal of spent resin from the IX columns and sampling resin prior to transfer of resin to the waste handling system for disposal. The spent resin collection tanks will be geometrically favorable, spaced risers that will include vessels for resin holding and for liquid associated with the resin transfers (fluidizing the resin). Sampling or monitoring of the spent resin uranium content will be required prior to transfer to the waste handling system, where vessels are not expected to be designed to dimensions that control criticality by geometry. [Proprietary Information]. The spent resin collection tank operation will be supported by a resin transfer liquid tank to manage liquids in the resin slurry during transfers (Figure 2-7).

15

.:··

.: ...... ~:.NWMI .·:.••.-.·.-..............

. ~ ~ ~--~! ," ' NORTHWEST MEDICAL ISOTOPES




The tank vessels will be mounted on separate supporting frameworks and skids, anchored to the room floor, and seismically designed to withstand the RPF DBE. The tanks will be vented to the process vessel vent offgas treatment system.

2.4.11 Fresh Resin Tank

The fresh resin tank will be an [Proprietary [Proprietary Information] Information] located outside the hot cell Figure 2-8 [Proprietary Information] and is not intended to hold any radioactive materials, including uranium (Figure 2-8). The tank will be equipped with an agitator and taps for water and fresh nitric acid additions to fluidize the resin for filtering, transfer, and flushing.

While not intended for service with radioactive materials, the tank will not be vented to the facility ventilation system and open to the room. The tank will be protected against backflow from the IX columns by its elevated location above the IX columns and an interlocked pair of three-way valves with an air break between them (double block-and-bleed).

16

.: . .-.~:. NWMI ............ : ..• • .. •.•:. . : !~~~~: ~ NORTHWISTMEO!CM.ISOTOPES


The valves will be closed until a fresh resin transfer is needed and the subject IX system is configured to receive the new resin (i.e., IX pumps that could push uranium-bearing solution back to the fresh resin tank are deenergized).

2.4.12 Pumps

Various small centrifugal, metering, peristaltic, and air-diaphragm pumps will be used throughout the uranium purification system to move solutions (or resin) between vessels. The pumps in the uranium purification system will have a [Proprietary Information]. Each pump will be mounted on a skid equipped with remote operable connections, [Proprietary Information].

2.5 EVALUATION OF NORMAL CONDITIONS

Under normal conditions, the uranium recovery system will contain UN solution in the impure feed tanks, IX feed tanks and columns, concentrator feed tanks and evaporators, and product uranium collection and possibly rework tanks. The liquid waste tanks, condensers, and condensate monitoring tanks should not contain significant uranium. [Proprietary Information].

As shown in the normal condition analysis models of the hot cell in NWMI-2015-CRITCALC-006, Tank Hot Cell (cases tankpit3_500 [-1500]), the uranium purification system vessels in the hot cell are modeled as containing [Proprietary Information]. ·

Therefore, under normal operating conditions and within the limits set by the controls listed in Section 6.0, the uranium purification and recycle system tanks and equipment will remain safely subcritical.

17

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3.0 NUCLEAR CRITICALITY HAZARD IDENTIFICATION

3.1 HAZARD IDENTIFICATION METHOD

A what-if evaluation was performed to identify upsets and associated consequences related to criticality safety for the various UN solution operations in the RPF uranium purification system.

The general what-if methodology was modified to include additional information focused on the criticality safety aspects of the operations:

• • • • •

Identification of criticality-related parameters Potential or hypothetical causes Consequences of the upset Presumed safeguards that will likely be required or implemented Analysis modeling consideration and/or needs

The analysis was compared against the hazard and operability studies (HAZOP) in NWMI-2015-SAFETY-001, NWMI Radioisotope Production Facility Preliminary Hazards Analysis.

3.2 HAZARD IDENTIFICATION RESULTS

The what-if analysis is summarized in Table 3-1.

Table 3-1. Uranium Purification System What-If Upset Analysis (4 pages)

What-if?


Cause(s)

i • [Proprietary Information]

Consequences

i [Proprietary /Information]

18

Safeguards Analysis

! [Proprietary 'Information] i

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Table 3-1.

What-if?


Uranium Purification System What-If Upset Analysis (4 pages)

Cause(s) Consequences Safeguards Analysis

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; • [Proprietary Information]

l [Proprietary ! • [Proprietary Information] l [Proprietary ;Information]. ! jinformation]

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'= "'"•"·''""~"~•'•''">··•·•'"'•"" ,,,,_:;..-..~;,., •,-_,.:;....;,-!•-~;..·r-Jr • --·····--.,.,, .-c..,7,,:_-L,, _".._.,•:.--.; ,•,

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, [Pr~prl~bzy-----~-F- '[P~~p~t;y·w~~~ti~~]--; [:Proprietary

finformation] 1 ;Information]

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L.::. .. 2 ..... .. : .. " .. :. .... «"~-:.: . 2, ... , .. ... <. ·.---- ....... ·=· ""'··-" ..... :... ..... _ .... ..: ....... ~ [Proprietary Information]

i • [Proprietary Information]

19

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[Proprietary Information] ; [Proprietary [Information]

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Table 3-1.

What-if?




Uranium Purification System What-If Upset Analysis (4 pages)

Cause(s) Consequences

20

Safeguards Analysis

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Table 3-1. Uranium Purification System What-If Upset Analysis (4 pages)

What-if? Cause(s) Consequences Safeguards Analysis

,c~e~ ·c'l4: , .. -i>-~ram~te~; co"U:tentratio.n ..... :- ·· , .. . . ; ' ·· •. .... ·' : .. ;j I ' ' ", •v,' "·' , ' ' ,, ,; ' ' •· ',, "" , 1 ' ·:·- l

~~-:L~[.-[P;;;p;i~<-:;:.l~~_;fy~;__,e,~it,'i;~:~~~~]~~~~-;:._,~ .. ;J Information] Information] [Information] I 'Information]

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t=:~=;:1£~~~~~"-2~~~-j-~-~-,~c.~~~-·· '"'~~-E~t5~~:::"2] [Proprietary : • [Proprietary i [Proprietary l · [Proprietary Information] ; [Proprietary Information] ! Information] (Information] I !Information]

l ' ! I ; ' \ i

i • NWMI-2015-CRITCALC-001, Single Parameter Subcritical Limits for 20 wt% 235U - Uranium Metal, Uranium Oxide,

and Homogenous Water Mixtures, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. b NWMI-2015-CRITCALC-006, Tank Hot Cell, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. c NWMI-2015-SAFETY-004, Quantitative Risk Analysis of Process Upsets Associated with Passive Engineering Controls

Leading to Accidental Criticality Accident Sequences, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. d NWMI-2015-SAFETY-005, Quantitative Risk Analysis of Criticality Accident Sequences that Involve Uranium Entering

a System Not Intendedfor Uranium Service, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. e NWMI-2015-SAFETY-Ol 1, Quantitative Risk Analysis of Natural Phenomenon and Man-Made Events on Safety

Features and Items Relied on for Safety, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. f NWMI-2015-SAFETY-009, Quantitative RiskAnalysis of Administratively Controlled Enrichment, Mass, Container

Volume, and Interaction Limit Process Upsets Leading to Accidental Criticality Accident Sequences, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

CSE O.D. IX QA

g Criticality in a system outside of the uranium purification is addressed in separate CSEs for the affected systems.

criticality safety evaluation. SPL single parameter limit outside diameter. U uranium. ion exchange. UN uranyl nitrate. quality assurance.

21

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4.0 CONTINGENCY ANALYSIS

4.1 DOUBLE CONTINGENT SCENARIOS


Those scenarios from the what-if study that are considered to be double contingent against occurrence or against a resulting criticality are discussed in the following subsections. The referenced qualitative risk analyses (QRA) provide additional information, including event tree and/or fault tree analyses of the criticality scenarios. ·

4.1.1 Overfill of a Tank Results in Fissile Solution in the Offgas Treatment System

Scenario: C3

4.1.1.1 Description

The UN processes vessels and support tanks are designed to be of safe geometry to prevent the fissile solution in the system from achieving criticality. The system, however, will be connected to other systems, including the o:ffgas treatment system that employs vessels oflarger, geometrically unfavorable sizes. Because of this, fissile solution must be prevented from migrating due to a mishap to the offgas treatment equipment.

4.1.1.2 Initiating Event

The tanks will be connected by hard pipe to makeup water and nitric acid supply systems through the UN process system and common vent headers, which contain enough volume to cause the tanks to overflow. This event could be initiated by operator error (failure to actuate closure of a makeup solution valve) or by a mechanical failure of the same valve(s) (either valve seat leak or valve failure to close).

4.1.1.3 Primary Contingency

The primary barrier to an overflow reaching an unfavorable geometry in the ventilation train will be a physical overflow drain built into each local vent header that empties overflow solution in the vent header to a safe geometry (TBD in detail design.- either pencil tanks or the room floor) NWMI-2015-SAFETY-005, Quantitative Risk Analysis of Criticality Accident Sequences that Involve Uranium Entering a System Not Intended for Uranium Service, evaluates the potential for failure of this physical feature and shows the failure to be unlikely.

Criticality Safety Controls Credited [Proprietary Information]

•

•

•




22

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4.1.1.4 Secondary Contingency


The secondary barrier to an overflow reaching an unfavorable geometry in the ventilation train will be a second physical overflow drain built into each local vent header that empties overflow solution in the vent header to a safe geometry (TBD in detail design- either pencil tanks or the room floor). NWMI-2015-SAFETY-005 evaluates the potential for failure of this physical feature and shows the failure to be unlikely. The primary and secondary overflow features are required to be independent.

Criticality Safety Controls Credited

[Proprietary Information]:

•

•

•

4.1.1.5




Common Mode Failure Potential

The facility will be equipped with independence and redundancy in the overflow physical features. Because the physical features are built-in and redundantly installed, there is no common mode failure potential for the two features.

4.1.1.6 Summary

Fissile solution must be prevented from reaching the unfavorable-geometry o:ffgas treatment vessels. To ensure that an overfill of the UN system does not result in such an occurrence, redundant and independent overflow solution diversions will divert any liquid that enters the ductwork to a safe geometry prior to reaching unfavorable geometry ductwork or equipment. The function of the diversion features will be reinforced by management measures such as periodic inspections. Double contingency for this scenario is demonstrated by the two independent physical features that divert overflow solution to safe geometry before the solution can enter unfavorable-geometry ductwork or equipment.

Therefore, compliance with the double-contingency principle is demonstrated.

4.1.2 Fissile Solution Leaks into Steam or Chilled Water Systems

Scenario: C5

4.1.2.1 Description

Some U recycle system tanks, evaporators, and associated condensers are heated or cooled by external systems supplying either steam or chilled water to the outside of the vessels. The steam and chilled water systems employ unfavorable-geometry vessels, so the fissile solution must not reach those systems to prevent a criticality. Due to the physical interface, fissile solution could potentially migrate into the steam or chilled water system through a leak in the inner vessel wall and into the jacket or coil.

23

.:~~~~;. NWM I ··•········ . ........... ;. .........

• • ••• • • NO.RTHYJEST MEDJCAI. ISOTOPES



A potential criticality in the steam or chilled water system could be initiated by corrosion or the presence of a fabrication flaw of a vessel wall such that fissile solution can pass into a surrounding jacket, coil, or past a heat exchanger tube.


The process vessels are designed to withstand the intended process environment, including resistance to corrosion by the concentrated and heated nitric acid process chemistry. Nuclear facilities are subject to thorough quality assurance (QA) and inspection of the equipment at procurement and during installation prior to startup. NWMl-2015-SAFETY-004, Quantitative Risk Analysis of Process Upsets Associated with Passive Engineering Controls Leading to Accidental Criticality Accident Sequences, evaluates the potential for a leak through from the process vessel to the heating or cooling system and shows the upset to be unlikely.


[Proprietary Information] [Proprietary Information].


The steam and chilled water systems will interface with a number of process vessels that contain fissile solution, any of which could leak and allow fissile solution to migrate to an unfavorable geometry vessel in the steam or chilled water supplies. To prevent significant transfer of solution to either of these systems, each will be equipped with an intermediate loop within the hot cell to prevent migration ofU mass and high-dose radionuclides to the steam or water systems outside the hot cell. Heat exchangers between the main supply system and the hot cell loop will provide a second physical barrier that must also leak before U can enter the unfavorable-geometry vessels of the steam or chilled water supplies.






4.1.2.5 Common Mode Failure Potential

There are no common mode failures identified for these two contingencies.

4.1.2.6 Summary

Fissile solution must be prevented from reaching the unfavorable-geometry steam system and chilled water system vessels. The design and construction of the UN process vessels and the intermediate heat exchangers between the hot cell loops and the main supplies combine to provide double-contingency against a criticality in the steam or chilled water systems.


24

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4.1.3 Fissile Solution Forced into the Fresh Resin System

Scenario: C7

4.1.3.1 Description


The IX columns will be serviced by connections to external fresh IX resin supplies. Because the IX columns use hard-piped connections from unfavorable-geometry makeup tanks outside the hot cell, a potential flow path exists for uranium-bearing solution to find its way back up into these supply tanks from the IX columns.


Upset conditions could be due to U solution supply valve misalignments (misdirected flows) or pressurization of the IX column(s).


Each fresh resin supply line to an IX column will be equipped with a double block-and-bleed valve set that is configured with an unbroken flow path between the resin supply and associated IX column only when the double-block valves are aligned to supply resin. The bleed between the block valves will drain to a safe-geometry seal pot and the floor of the hot cell. The valves will be physically connected together and to the actuator such that both are opened or closed with the other. The valves will fail-closed to isolate the resin supply tank on loss of plant power or air. Failure of an augmented administrative control (lock and valve alignment) with management measures is shown to be unlikely in NWMI-2015-SAFETY-009, Quantitative Risk Analysis of Administratively Controlled Enrichment, Mass, Container Volume, and Interaction Limit Process Upsets Leading to Accidental Criticality Accident Sequences.



•

•

•


[Proprietary Information] .

[Proprietary Information] .



Each fresh resin supply line to an IX column will be equipped with a paddle blank that is inserted at a flange connection outside the hot cell when fresh resin is not being supplied to the IX column. The blank will prevent passing any solution to the non-favorable-geometry tank. '

The blank will incorporate a lock that is controlled by a supervisor to provide a second administrative check of the intended system configuration before the blank is "released" for removal to allow passing of the resin to the IX column. Failure of an administrative control with management measures is shown to be unlikely in NWMI-2015-SAFETY-009.

25

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• [Proprietary Information]




Common mode failure potential will exist for these two contingencies if a manager or supervisor instructs operators to open the supply line (and unlock the blank) while U solution remains in an IX column (process not properly prepared for resin transfer). Operating procedures and management measures (conduct of operations) will provide checks and a process sequence to ensure that the two physical measures are not inadvertently defeated.

4.1.3.6 Summary

Fissile solution must be prevented from reaching the unfavorable-geometry resin supply system vessels. Incorporation of a double block-and-bleed valve set and a blank in the supply line when not supplying resin will combine to provide double contingency against a criticality in the steam or chilled water systems.


4.1.4 Fissile Solution Forced into Process Gas System

Scenario: C8

4.1.4.1 Description

Some U recycle system tanks will have air or nitrogen connections to the tank or immediate piping. The gas supply systems will employ unfavorable-geometry vessels, requiring that fissile solution never reaches those systems to prevent a criticality. Due to the physical interface, fissile solution could potentially migrate into one of the systems.


A potential criticality into a process gas system could be initiated by a pressure transient or a valve. misalignment combination that causes overfilling of a tank and subsequent pressurization of the tank.


The process vessels will be continuously vented to prevent overpressurization and to remove any radiolytically produced hydrogen. NWMI-2015-SAFETY-005 evaluates the potential for a ventilation failure, with resulting overpressurization, and shows the upset to be unlikely.



26

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. : ~~.~!;. NORTHW!STMEDICAllSOTOPES



Where gases are added to the vented vessel below the overfill level (purge gases and reaction gases), an overloop seal system will be provided to prevent backup into the gas addition system if the tank vent clogs at the same time when loss of gas pressure occurs. This passive engineering control will prevent accidental nuclear criticality by precluding fissile solution from backflowing into the gaseous distribution system where a non-safe geometry might exist. NWMI-2015-SAFETY-005 evaluates the potential for failure of the gas pressure and overloop seals at the time of a system overpressurization and shows the upset to be unlikely.




There are no common mode failures identified for these two contingencies.

4.1.4.6 Summary

Fissile solution must be prevented from reaching the unfavorable-geometry gas supply system vessels. Continuous process tank venting to prevent overpressurization along with installation of gas line overloop seals will combine to provide double contingency against a criticality in a gas supply.


4.2 NOT CREDIBLE/CRITICALITY NOT CREDIBLE SCENARIOS

The scenarios from the what-if study that are considered to be not credible to occur or not credible to result in a criticality are discussed in the following subsections.

4.2.1 Fissile Solution Leaks from Vessels

Scenario: Cl

Mechanisms such as tank corrosion, seal corrosion, seal misplacement, valve not seated, or similar, could result in process solution leaks from any of the tanks to the floor of the hot cell. Given the relatively small volume of the tank systems considered and with the hot cell floor in excess [Proprietary Information], as analyzed in NWMI-2015-CRITCALC-001, Single Parameter Subcritical Limits for 20 wt% 235U - Uranium Metal, Uranium Oxide, and Homogenous Water Mixtures. To bound potential interaction effects with tanks above the spill on the floor, a model case in NWMI-2015-CRITCALC-006 showed that [Proprietary Information]. Therefore, this mode of upset is judged not credible to result in a criticality accident.



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4.2.2 System Tanks are· Larger than Intended

Scenario: C2


The hot cell uranium-bearing tanks, as stated in NWMl-2013-034, will be designed for each riser to be [Proprietary Information]. The design sizes are safe for the intended solutions (normal conditions analyzed in NWMI-2015-CRITCALC-006; evaluated in Section 2.5). However, a number of upset conditions could cause the tank risers to be of different geometry than intended:

• • •

Installation error (including procurement error) Bulging due to pressurization Corrosion of the tank walls

Installation error: A base premise and assumption of the safety analysis is that the facility is built as intended. Nuclear facilities are subject to thorough QA and inspection of equipment at procurement and during installation prior to startup. NWMI-2015-SAFETY-004 evaluates the potential for improper procurement and installation of safe-geometry intended equipment. Administrative controls and management measures will be in place to prevent improper installation. The slight manufacturing tolerance variance will be bounded by the conservative analysis ofNWMI-2015-CRITCALC-006, which analyzed (and showed safe) tanks filled [Proprietary Information]. Therefore, this mode of upset is judged not credible to result in a criticality accident.

Bulging due to pressurization of process vessels: The system tank vessels are designed to withstand the RPF operating conditions, including the pressures exerted by the water and nitric acid liquid feeds, the chilling or heating jackets on the outside of some of the vessels, and potential gas generation in the solutions. The tanks associated with the [Proprietary Information] will be vented continuously; · chemically generated gases is accounted for in the system design, with the offgas removed such that it cannot pressurize the tank vessels. A slight heating or pressure variance will be bounded by the conservative analysis ofNWMI-2015-CRITCALC-006, which analyzed (and showed safe) tanks filled [Proprietary Information]. Therefore, this mode of upset is judged not credible to result in a criticality accident.

Corrosion of the tank walls: The process vessels are designed to withstand the intended process environment, including resistance to corrosion by the concentrated and heated nitric acid process chemistry. The slight wall erosion or corrosion will be bounded by the conservative analysis of NWMI-2015-CRITCALC-006, which analyzed (and showed safe) tanks filled [Proprietary Information]. Therefore, this mode of upset is judged not credible to result in a criticality accident.




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Conditions analyzed include the following features:

•

•

•





4.2.3 Process Vessels Dislodged from Mounts and Move or Fall Closer Together

Scenario: C4

A base premise and assumption of the safety analysis is that the RPF is built as intended and that the equipment is resistant to damage due to the design basis event (DBE). Nuclear facilities are subject to QA and inspection of equipment at procurement and during installation prior to startup. NWMI-2015-SAFETY-004 evaluates the potential for relocation of safe-geometry intended equipment and concludes that the frequency of the facility DBE and design measures are sufficient to prevent a criticality. Administrative controls and management measures will be in place to prevent errant installation. Therefore, this upset is judged not credible to result in a criticality accident.






4.2.4 Fissile Solution Forced into Chemical Reagent or Water Supply Systems

Scenario: C6

The U recycle system will be serviced by several connections to external systems such as chemical reagents or reactants and process water supplies. Because these are hard-piped connections from outside the hot cell to various process vessels, uranium-bearing solution could potentially find its way back up into any of these supply systems, which include unfavorable-geometry vessels of various sizes. Upset conditions could be due to valve misalignments (misdirected flows) or pressurization of the process vessels.

Criticality in each of these systems will be prevented by incorporation of safe-geometry intermediate day tanks in the liquid systems that are physically isolated from any larger-geometry tanks with an air break, such that backflow of uranium to an unsafe geometry is physically impossible. The day tanks will be constructed with the same geometry and spacing constraints as the U recycle process tanks, which are shown to be safe for all process solutions in NWMI-2015-CRITCALC-006. Therefore, this mode of upset is judged not credible to result in a criticality accident.

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4.2.5 Incompatible Resin is Installed

Scenario: C9


Chemical incompatibility of resin loaded with UN could result in an uncontrolled chemical reaction and pressure buildup in an IX column ifthe wrong resin is inadvertently procured and installed. The resulting pressure or explosion could then result in uranium solution in an uncontrolled geometry.

The wrong resin could be installed only if the wrong material is purchased, passes receipt procedures, and is made available for loading. A knowledgeable operator should recognize the wrong resin, but no credit is taken for that error recovery mechanism. Rather, the ion columns are designed with a pressure relief path to the hot cell floor to minimize the potential for explosive overpressure. The floor of the hot cell will be very large relative to the UN solution volume at risk and provide a flat surface with no significant collection points other than a safe-geometry sump. Therefore, a solution spill from an overpressurized IX column will remain safely subcritical as the solution cannot physically collect to [Proprietary Information] analyzed in NWMI-2015-CRITCALC-OO 1. Therefore, this mode of upset is judged not credible to result in a criticality accident.










4.2.6 Excessive Uranium in Ion Exchange Liquid Waste, Condensate, or Spent Resin Collection Tanks

Scenarios: ClO, Cll, C12

The liquid waste, condensate, and spent resin collection tanks will be of favorable-geometry and installed similarly to the uranium process vessels. The collection tanks are addressed in the NWMI-2015-CRITCALC-006 model analysis. With the conservative modeled range of concentrations, the analysis easily bounds any credible larger than intended uranium concentration in these support systems. The models remained safely subcritical, demonstrating that an improper transfer of uranium solution to these support systems will not achieve criticality. Therefore, this mode of upset is judged not credible to result in a criticality accident.

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•

•

•

4.2.7




Excessive Concentration in System Tanks

Scenario: Cl3, Cl4


The uranium recycle system is designed to intentionally recover and concentrate uranium in acid solution. If a temperature problem occurs (radiolytic heating or concentrator heating upset), the uranium concentration could potentially exceed expected U concentrations, including the precipitation of uranium from the process solution. High concentration, including potential precipitation, has been considered in the conservative analysis ofNWMI-2015-CRITCALC-006, which analyzed (and showed safe) tanks filled [Proprietary Information]. Therefore, this mode of upset is judged not credible to result in a criticality accident.









• [Proprietary Information]

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•

4.2.8



Hot Cell Floods

Scenarios: C15


The hot cell will be equipped with waterlines for process purposes and fire suppression. Due to these waterlines and the site vulnerability to a flood from the surrounding area, the room could potentially experience pooled or deep water flooding such that the process equipment could be inundated.

NWMI-2015-CRITCALC-006 analyzes the hot cell with different degrees of water flooding, from reduced density water representing sprinkler activation up to completely flooded with water. The results of the calculations showed that even for the worst-case concentration of uranium mixture in the arrays pf tanks, the system remains safely subcritical. A worse water reflection condition cannot be physically achieved. Therefore, this upset is judged not credible to result in a criticality accident.







•

•

•

•





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4.3 NUCLEAR CRITICALITY SAFETY PARAMETERS

Table 4-1 lists the nuclear criticality safety parameters.

Table 4-1. Nuclear Criticality Safety Parameters

Nuclear parameter Controlled? (Y/N) Basis

Section in which failure is analyzed

Mass 1 [Proprietary Information] i [Proprietary Information]

[~~:::!~~JL~~~~~i::!~:~~:~JC~'."'':~~~~~:~·~:5~1~:~~LiJi~~~J~~~:~;:t:;~:~z~;f2~~~~·1 Moderation ; [Proprietary Information] l [Proprietary Information] '

[t~:::::~i:'Jf ~:~~~:~~~!:~r1:~][I}~·!~~~1~17~~~~~(_~-EJ[~IJ~;1~~k41~~f~~~f I~1tl51 Volume· .. i [Proprietary Information] l . [Proprietary Information]. ·. ;

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L:~~~~~~ii:Ir~1~~g~~1~ifili~~<i~£tI.~:·~:.lffe.~en~ilii!Iif0~~~9Ec~:J·c:=~rI:~~~:z::·:.~;::::;~;;:3::.:cJ Enrichment ! [Proprietary Information] l [Proprietary Information] ·

• NWMI-2015-CRITCALC-006, Tank Hot Cell, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015. 235U = uranium-235. U = uranium.

4.4 DEFENSE TABLE

Table 4-2 summarizes the primary and secondary contingencies (i.e., defenses) that demonstrate each credible, unlikely process upset complies with the double-contingency principle.

Table 4-2.

Scenario

Summary of Primary and Secondary Contingencies (2 pages)

Primary Contingency Secondary Contingency

.. :' ,'ii \' ...


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Table 4-2. Summary of Primary and Secondary Contingencies (2 pages)

Scenario Primary Contingency Secondary Contingency -~ ~~~ce~-- -.'..:.-·;.,- "~· ::: ,,--,,- · 7: ·.:··-:·'I· < .. -:~·;·~.~:· ~~~(,~ -·· ,--· ;k~~-= ... :::·· .. ~/--:--"':-· ·.· ···:···-r.-·"'.:""" "·- --::' - -::: ~-~·-r. :':"~-- M -; .... ·-: ~-··· .. -1,:-·7 ':;'

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5.0 CRITICALITY ACCIDENT ALARM SYSTEM


The process areas discussed in this CSE will be within the detector and alarm coverage of the criticality accident alarm system (the criticality accident alarm system coverage evaluation will be performed after final design is complete and prior to facility startup).

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6.0 DOUBLE CONTINGENCY CONTROLS

6.1 PASSIVE DESIGN FEATURES


Basis: [Proprietary Information]













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6.2 ACTIVE ENGINEERED FEATURES



6.3 ADMINISTRATIVE CONTROLS


•

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6.4 GENERAL REQUIREMENTS


1. The facility license requires the facility structure and process equipment to be designed and installed to withstand a DBE.

Basis: Seismic considerations are a fundamental design aspect of the RPF.

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7.0 LIST OF ASSUMPTIONS AND OPEN ITEMS

7.1 ASSUMPTIONS

The assumption used in this CSE includes the following:

1. [Proprietary Information]


7.2 OPENITEMS

Open items include the following:


a. [Proprietary Information] b. [Proprietary Information] c. [Proprietary Information] d. [Proprietary Information] e. [Proprietary Information] f. [Proprietary Information] g. [Proprietary Information]



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8.0 CONCLUSION


The analysis presented in Section 4.0, with incorporation of the limits and controls presented in Section 6.0, demonstrate that the RPF U recovery and recycle systems and equipment meet the requirements of the double-contingency principle. As shown by the upset identification and analysis, uranium recovery operations in the hot cell will remain safely subcritical for all normal and credible abnormal conditions.

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9.0 REFERENCES


NWMI-2013-034, Uranium Recovery and Recycle Process Descriptions, PFD and P&ID, Rev. B, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2014.

NWMI-2015-CRITCALC-OO 1, Single Parameter Subcritical Limits for 20 wt% 235U - Uranium Metal, Uranium Oxide, andHomogenous Water Mixtures, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-CRITCALC-006, Tank Hot Cell, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-CSE-005, NWMI Preliminary Criticality Safety Evaluation: Target Fabrication Uranium Solution Processes, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-CSE-009, NWMI Preliminary Criticality Safety Evaluation: Liquid Waste Processing, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-CSE-01 l, NWMI Preliminary Criticality Safety Evaluation: Ojfgas and Ventilation, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-SAFETY-001, NWMI Radioisotope Production Facility Preliminary Hazards Analysis, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-SAFETY-004, Quantitative Risk Analysis of Process Upsets Associated with Passive Engineering Controls Leading to Accidental Criticality Accident Sequences, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-SAFETY-005, Quantitative Risk Analysis of Criticality Accident Sequences that Involve Uranium Entering a System Not Intended for Uranium Service, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-SAFETY-009, Quantitative Risk Analysis of Administratively Controlled Enrichment, Mass, Container Volume, and Interaction Limit Process Upsets Leading to Accidental Criticality Accident Sequences, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

NWMI-2015-SAFETY-011, Quantitative Risk Analysis of Natural Phenomenon and Man-Made Events on Safety Features and Items Relied on for Safety, Rev. A, Northwest Medical Isotopes, LLC, Corvallis, Oregon, 2015.

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