ML17128A067

NWMI-2017-RAI-002, Rev. 0, Response to the U.S. Nuclear Regulatory Commission Northwest Medical Isotopes, LLC - Request for Additional Information Regarding Application for Construction Permit (TAC MF6138).

ML17128A067
Person / Time
Site:	Northwest Medical Isotopes
Issue date:	04/28/2017
From:	Northwest Medical Isotopes
To:	Office of Nuclear Reactor Regulation
References
NWMl-2017-RAl-003, TAC MF6138 NWMI-2017-RAI-002, Rev. 0
Download: ML17128A067 (33)

Text

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• NORTHWEST MEDICAL ISOTOPU ATTACHMENT 3 Northwest Medical Isotopes, LLC Response to the U.S. Nuclear Regulatory Commission Northwest Medical Isotopes, LLC -Request for Additional Information Regarding Application for Construction Permit (TAC No. MF6138) Docket No. 50-609 (Document No. NWMl-2017-RAl-002 , Rev. 0, April 2017) Public Version Information is being provided via hard copy

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• Response to the U.S. Nuclear Regulatory Commission Northwest Medical Isotopes, LLC -Request for Additional Information Regarding Application for Construction Permit (TAC No. MF6138) Prepared by: Docket No. 50-609 NWMl-2017-RAl-002, Rev. 0 April 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, OR 97330 This page intentionally left blank.

NWMl-2017-RAl-002 , Rev. 0 Response to the U.S. Nuclear Regulatory Commission Northwest Medical Isotopes, LLC -Request for Additional Information Regarding Application for Construction Permit (TAC No. MF6138) Docket No. 50-609 NWMl-2017-RAl-002, Rev. 0 Date Published:

April 28, 2017 Document Number. NWMl-2017-RAl-002 I Revision Number. 0 Title: Response to the U.S. Nuclear Regulatory Commission, Northwest Medical Isotopes, LLC -Request for Additional Information Regarding Application for Construction Permit (TAC No. MF6138), Docket No. 50-609 Approved by: Carolyn Haass Signature:

NWMl-2017-RAl-002 , Rev. 0 REVISION ffiSTORY Rev Date Reason for Revision Revised By 0 4/28/2017 Issued for submittal to the NRG N/A

..... ; .. NWMI :{::::;; -TWffllllllOICAlllllWQ TERMS Acronyms and Abbreviations 23s u ANS ANSI AOA CFR DOT DBE EOI HIC IROFS ISG MCNP Mos MURR NCS NRC NWMI OSTR PSAR QA RAI RPF SAR SR SSC U.S. USL uranium-235 American Nuclear Society American National Standards Institute area of applicability Code of Federal Regulations U.S. Department of Transportation design basis event end of irradiation high-integrity container items relied on for safety Interim Staff Guidance Monte Carlo N-Particle margin of subcriticality University of Missouri Research Reactor nuclear criticality safety U.S. Nuclear Regulatory Commission Northwest Medical Isotopes , LLC Oregon State University TRIGA Reactor preliminary safety analysis report quality assurance request for additional information radioisotope production facility safety analysis report safety-related structures , systems , and components United States upper subcritical limit NWM l-2017-RAl-002, Rev. 0 NWMl-2017-RAl-002 , Rev. 0 This page intentionally left blank. ii NWMl-2017-RAl-002 , Rev. 0 CHAPTER 3.0 -DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS Section 3.5 -Systems and Components

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• A . . .. RAI 3.5-10 Section 50.9 , "Co mpleteness and accuracy of information," of JO CFR Part 50 requires that informa tion maintain ed b y th e applicant b e complete and accurate in all mat er ial respects.

Th e definition/or items relied on/or safety (JROFS) is provided in JO CFR 7 0.4 , " Definitions." The definition s tat es: Items r e lied on for safety m ean structu r es, systems, e quipm ent, compo nent s, and activities of personnel that are relied on to prevent pot en tial accidents at a facility that could excee d the peiformance requirements in§ 70. 6J or to mitigat e their potential conse qu e nce s. This does not limit th e lic e ns ee from identifyin g additional structures, sys tems , e quipment , components, or a c tivities qf personnel (i.e., b eyond those in the minimum set necessary.for compliance with th e peiformance requirements) as items relied on/or safety. In respo ns e to RAJ 3.5-2 , NW.MI revised its PSAR. The NWMI PSAR , R evisio n A, Section 3.5.J .3, "Nucle ar Safety Classifications for Structures, Systems, and Components , "s tates the following classification/or struc tures , systems, and components (SSCs): Safety-related IROFS -SSCs identified through accident analyses as required to m ee t th e peiformance requirements of JO CFR 70.61 (see Table 3-2) Safety-related

-SSCs that pro vide r easo nabl e assurance that th e facility can be operated wi thout undue risk to th e health and safety of workers, th e public , and e nvironm ent, includ es SSCs to me e t J 0 CF R 20 normal release or exposure limit s Non-safety-related

-SSC s related to the production and d e liv ery of products or services that are not in th e above safety cla ssifications NW.Ml PSAR, R evisio n A, Section 3.5.J.2, "Classifications Definitions," defines safety-related (SR) as follows: Th e definitions us e d in classifications of SSCs include the followin g: Safety-related is a classification applied to it ems relied on to remain fanctional during or following a d esign basis event (DBE) to ensure the:

• Int egrity of the facility infrastructur e

• Capability to shut down the facility and maintain it in a safe shutdown co ndition

• Capability to prevent or miti ga t e the c on se qu ences of accidents that could result in potential offsite ex posures comparable to the applicabl e guideline exposures set forth in JO CFR 70.6J, " P eifo rman ce R equ ir e ment s, " as applicable Additional information is n ee ded to und ers tand the application of th ese d ef mitions to SSCs at th e pr opose d NWM!facility. How ever, it is not clear if whether all IROFS are classified as SR. Additionally, th ere appears to be an inconsistency b etwee n the definition o/SR in revised PSAR Section 3.5.J.2 and SR SSCs in revisedPSAR Section 3.5.1.3. SR in Section 3.5.1.2 u ses guidelines exposures set forth in JO CFR 70.6J and SR in Section 3.5.J.3 includ es SSCs to meet JO CFR Part 20 normal r e leas e or exp osur e limits. In response to RAJ 3.5-3 and RAJ G-4 (ADAMS Accession No. ML16344A053), it is the NRC staff's understanding that NWMI will follow the p eiforma n ce requirements in JO CFR 70.6J(b), JO CFR 70.6J(c), and JO CFR 70.6 J(d). RAI 3.5-10a Clarify if all enginee ring safety features and administrative controls IROFS ar e classified as SR. Yes , all engineering safety features and administrative controls items relied on for safety (IROFS) are c l assified as safety-related (SR). 1 of 10 NWMl-2017-RAl-00 2, R ev. 0 . . . . . Re uest for additional information RAI 3.5-10b With respect to the PSAR, R evision A , Sections 3.5.1.2 and 3.5.1.3, clarify if all SSCs that are used to comply with 1 0 CFR Part 20 dose criteria are classified as SR. Yes , all structures , systems , and components (SSC) that are used to comply with Title 10, Code of Federal Regulations (CFR), Part 20 , "Standards for Protection Against Radiation ," dose criteria are classified as SR. RAI 3.5-10c Provide additional information to resolve the apparent inconsistency between the definitions of SR in PSAR , Revision A, Section 3.5.1.2 and Section 3.5.1.3, as both of these definitions applies to SSCs but uses different exposure criteria. The definition of" safety-related" in Section 3.5.l.2 ofNWMI-2013-021, Construction Permit Application for Radioisotope Production Facility , is similar to the U.S. Nuclear Regu l atory Commission (NRC) definitions in 10 CFR 50.2, " Definitions

." In Section 3.5.1.3, Northwest Medical Isotopes, LLC (NWMI) has included dose guidance from both 10 CFR 70.61 , "Performance Requirements

," and 10 CFR 20 to differentiate between SR IROFS and other SR SSCs. RAI 3.5-10d PSAR, Revision A, Table 3-25, " Systems Safety and Seismic Classifications and Associated Quality Level Group, " classifies the Process Steam System as an I R OFS with a quality assurance (QA) level 2. Explain the application of QA Level 2 to a system that is classified as an IROFS There is an error in Table 3-25. Since part of the process steam system , the in-cell secondary steam loops , has criticality controls, the system should be assigned as QL-1. RAI 3.5-10e Confirm if NWMI plans to comply with the performance requirements in 10 CFR 70. 61 (b), JO CFR 70.6J(c), and 10 CF R 70.6J(d). Yes , the performance requirements of 10 CFR 70.6l(b), 10 CFR 70.61(c), and 10 CFR 70.61(d) apply. 2 of 10

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-0 02, R ev. 0 CHAPTER 6.0 -ENGINEERED SAFETY FEATURES Section 6.3 -Nuclear Criticality Safety in the Radioisotope Production Facility -. .. . . . . .. RAI 6.3-17 Section 6b.3 of the JSG Augmenting NUREG-1537, Part 2, states that the applicant should include a summary description of a documented, reviewed, and approved validation report (by NCS fanction and management) for each methodology that will be used to perform an NCS analysis.

The summary description of a reference manual or validation report should include the following:

a summary of the theory of the methodo l ogy that is sufficiently detailed and clear to be understood, including the method used to select the benchmark experiments, determine the bias and uncertainty in the bias, and determine the upper subcritical limit. The Validation Report (NWMI-2014-RPT-006, "MCNP 6.1 Validations with Continuous Energy ENDFIB-VII.J Cross-Sections," Rev. 0) includes a discussion of physical parameters in the area of applicability (AOA). Additional information is needed to determine if the range of parameters to be modeled to support the facility design are within the AOA and the justification of your margin of subcritica l ity of0.05. Specifically, address the following requests in regard to NWMJ's response to RAJ 6.3-12, dated November 28, 2016 (ADAMS Accession No. ML16344A053).

NWMl's response to RAJ 6.3-12b states tha t "Movingfrom higher to lower 235 U enrichment, the 238 U/1 35 U ratio decreases and the incident neutron energy spectrum softens due to decreased parasitic absorption of thermal neutrons from 238 U " The N R C staff notes that many factors can affect the neutron spectrum.

For examp l e, as enrichment increases, the dimensions of fzssile material containers and equipment will tend to reduce so as to ensure subcriticality.

This can result in more neutron leakage and a corresponding hardening of the neutron energy spectrum.

Additiona ll y, NWMI 's response states tha t "Fissions from fast neutrons increases, which results in a softening of the neutron spectrum," appears to be inconsis t ent. RAI 6.3-17a Provide additional information to address the various effects that can affect the neutron spectrum or, if a genera l case that increasing enrichment always results in spectral softening cannot be made, provide alternate justification for the margin of subcritica l ity. [Proprietary Information]

3 of 10 NWMl-20 17-RAl-002 , R ev. 0 Section 6.3 -Nuclear Criticality Safety in the Radioisotope Production Facility . .. . .. . . .. . . RAI 6.3-17b Justify the choice of benchmarks in the similar enrichment range to your proposed operations.

In particu l ar, justify not including benchmarks such as IEU-SOL-THERM-00 1 t hat are known to have a large negative bias. Exclusion of these benchmarks cou l d be non-conservative if design calculations are susceptible to similar biases. [Proprietary Information]

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-(Applies to RAls 9.7-3 through 9.7-5) NWMl-2017-RAl-002, Rev. 0 CHAPTER 9.0 -AUXILIARY SYSTEMS Section 9.7 -Other Auxiliary Systems Re uest for additional information Section 50.9, "Completeness and accuracy of information," of JO CFR Part 50 requires that information maintained by the applicant be complete and accurate in all material respects. NUREG-1537, Part 1, Chapter 9, "Auxiliary Systems ," states, in parl, that the applicant should provide sufficient information for alt auxiliary systems to sup[XJrt an understanding of the design and functions of the systems, with emphasis on those aspects that could affect the production facility and its safety features, radiation exposures , and the control or release of radioactive material.

The applicant should include the following information for each auxiliary system: (1) design bases (2) system description, including drawings and specifica tions of principal components and any specia l materials (3) operational analysis and safety function (4) instrumentation and control requirements not described in Chapter 7, "Instrumentation and Controls Systems," of the safety analysis reporl (SAR) (5) required technical specifications and their bases, including testing and surveillance.

NUREG-1537, Part 1 , Chapter 9, "Auxiliary Systems," also states, in parl that typical systems discussed in this chapter include the contro l and of radioactive waste and reusable radioactive components and that the applicant should describe the systems and show how they are designed to perform the design-basis.functions derived in Chapter 11, " Radiation Protection Program and Waste Management," of the SAR. NUREG-1537, Part 1, Chapter 9, Section 9.5, "Possession and Use of Byproduct, Source, and Special Nuclear Material, "states , in part , that the radiological design bases for handling radioactive materials and radioactive waste should be derived from Chapter 11 of the SAR. NUREG-153 7, Part 1, Chapter 9 , Section 9. 7, "Other Auxiliary Systems, " states, in parl, that th e applicant should demonstrate that the auxiliary system and any malfunction could not create conditions or events that could cause an unanalyzed production facility accident or the uncontrolled release of radioactive material beyond those analyzed in Chapter 13 of the SAR. NUREG-1537, Part 2, Chapter 9, Section 9.5, "Possessio n and Use of Byproduct, Source, and Special Nuclear Material, " states, in part, that the areas of review should include the provisions for controlling and disposing of radioactive wastes, including spec ial drains for liquids and chemicals, and air exhaust hoods for airborne materials , with design bases derived in Chapter 11 of th e SAR. NUREG-1537, Part 2, Chapter 9, Section 9. 7, "O th er Auxiliary Systems," Acceptance Criteria, states, in part that the design and functional description of the auxiliary system shou ld ensure that it [the system] conforms to the design bases. Section 9. 7 further states , in part, under "Evaluation Findings," that this section of the SAR shou ld contain sufficient information to support the following types of conclusions:

• The system has been designed to perform the.functions required by the design bases.

• No analyzed functions or malfunctions could initiate the uncontrolled release of radioactive material.

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• RA1 9. 7-3 NWMI PSAR Section 9. 7. 2, " Control and Storage of Radioactive Waste, " does not include sufficient information to determine whether the design as described in the PSAR will meet the design bases. Preliminary input flows, tank volumes, high dose waste concentrator flow , and a discussion of system operation and control were not provided in the PSAR. For example, the PSAR describes weekly target processing resulting in frequent additions of different acids and radionuclides to the single high-dose waste collection tank. PSAR Section 9. 7.2.2.1 provides no preliminary discussion of the processes used to add and mix sufficient NaOH to maintain the solution pH within the range required for high-dos e waste concentrator operation.

PSAR Section 9. 7.2.2.2 provides no preliminary information on water quality requirements for recycle or quantitative esti mat es of how much will be recycled and how much will flow to the lodose waste co ll ection tank. More information is needed to understand system fanction and operation for both the high-dose and low-dose waste systems. RAI 9. 7-3a Provide process flow diagrams of the radioactive waste systems, including inputs, storage, preparation for processing, sampling and analysis, and expected throughputs/frequency of processing to support target processing.

Show anticipated values of relevant parameters (e.g., nuclides, concentrations, chemical constituents, activity, volumes, flowrates, batch sizes, etc.) to enable evaluation of the design. Anticipated va lu es presented should reflect the maximum waste generation rates resulting from nominal operational processing capabi lity as presented in PSAR Section 4.1.2.1, "Process D esign Basis." NWMI-2013-021 , Section 9.7.2, was rewritten to address the items listed in RAI 9.7-3a (Attachment A). The operational processing capabilities a lign with NWMI-2013-021 , Section 4.1.2.1 , "Process Design Basis." RAI 9.7-3b Provide a discussion of the operation of the high-dose liquid waste handling system. This discussion should include sufficient information to understand how th e contents of th e high-dose was t e collection tank are samp l ed and controlled coincident with inputs from the off gas system, the molybdenum waste tanks, and the uranium recovery syste m. NWMI-2013-021 , Section 9.7.2 , was rewritten to address key elements of these requests for additiona l information (RAI). The input s to the high-dose waste collection tank are batch transfers. The inputs must be samp l ed before transferring to the non-criticality safe, high-dose waste collection tank. The high-dose collection tank can be "iso l ated" (i.e., no incoming waste , agitated, and sampled when needed). Similarly, the high-dose concentrate tank is filled in a b atch manner from the high-dose waste concentra tor , agitated, and sampled prior to transfer to the solidification process (see Attachment A). RAI 9.7-4 NWMI PSAR Chapter 3, Table 3-3 " Relevant US. Nuclear Regulato ry Commission Guidance ," does not list R egulatory Guide (RG) 1.143 , "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Pow er Plants". PSAR Section 3.1.7 "Codes and Standards," states the codes and standards li sted in Table 3-7 "Design Codes and Standards, " are design inputs but farther states , in part that th e specific requirements for that system produced by each applicable reference are identified in system design descriptions.

Although Table 3-7, lists ANSI I ANS-55.1 , "Solid Radioacti ve Waste Processing System for Light-Water Cooled R eactor Plants ," 1992 (R2009), ANSI I ANS-55.4 , "Gaseous Radioactive Waste Processing Systems for Light-Water R eactor Plants, " 1993 (R2007) and ANSI I ANS-55.6, " Liquid Radioacti ve Waste Pro cessing System for Light-Water Cooled Reactor Plants ," 1993 (R2007), these standards are not included inPSAR Section 9. 7.2. In response to RAJ 3.5-1 (ADAMS Accession No. ML16123Al 19), NWMI stated that specific design codes and standards will be finalized in the final design. 6of10

-NWMl-2017-RAl-002, Rev. 0 Section 9.7 -Other Auxiliary Systems Request for additional information Thus, NWMI has made no design commitments for waste systems and there is insufficient information on the design of the radioactive waste systems and the waste staging and shipping building to conclude definitively that the design of radioactive waste systems will comply with NRC regulatory requirements.

RAI 9.7-4a Verify if NWMI intends to useRG 1.143 and the associated ANSI/ANS standards as a design input for the radioactive waste system. If so , these guidance and standards should be listed in PSAR Sections 3.1. 7 and 9.7.2. NWMI-2013-021 , Table 3-3 , calls out Regulatory Guide 3.10 , Liquid Waste Treatment System Design Guide for Plutonium Processing and Fuel Fabrication Plants , as an appropriate design guide. As part of final design , NWMI will evaluate the need for use of Regulatory Guide 1.143, Design Guidance for Radioactive Waste Managem e nt Systems , Structures , and Components Installed in Light-Water-Cooled Nuclear Power Plants. RAI 9.7-4b With a single input tank (i.e., high dose waste collection tank) and output tank (i.e., waste concentrate collection tank) for the high-dose waste system and with frequent inputs to the collection tank from various waste sources, it is unclear how the output tank can be isolated and recirculated to assure a representative sample can be obtained and analyzed.

Provide additional information to explain how the radioactive waste system design will provide for sampling and analysis of the high dose waste collection tanks and/or waste concentrate collection tank NWMI-2013-021 , Section 9.7.2 was rewritten to address key elements of these RAis. The inputs to the high-dose waste collectio n tank are b atc h transfers.

The inputs must be sampled before transferring to the non-criticality safe , high-dose waste collection tank. The hi gh-dose collection tank can be "isolated" (i.e., no incoming waste, agitated , and sampled when needed. Similarly, the high-dose concentrate tank is filled in a batch manner from the high-dose waste concentrator , agitated, and sampled prior to transfer to the solidification process (see Attachment A). RAI 9. 7-5 PSAR Section 9. 7.2 provides a description of the radioactive waste contro l and storage systems. As stated in PSAR Section 9. 7.2. 1 , the waste handling system design basis provides:

• Lag storage capability for liquid waste , divided into high-dose and low dose source terms

• Safely transit solid waste from hot ce lls to a waste encapsulation area

• Capability to solidify low-dose liquid waste and load drums

• Capabi lity to solidify high-do se liquid waste and load high-integrity con tain ers (HICs)

• Capability to encapsulate solid was t e in drums

• Capability to handle and load a waste shipping cask with radiological waste drums or containers

• Storage space for Class A radioactive waste

• Storage space for decay of high-dose waste in HI Cs to meet DOT shipping requirements.

The radioisotope production facility (RPF) is designed for continuous sequentia l batch operation.

After startup and operations longer than the lon gest lag storage , systems will contain an equilibrium in-process radioactive material inventory.

PSAR Table 1-3 presents that inventory as 187 , 000 curies for the high-dose waste tanks. The concentration of radioactive material in the high dose waste concentrate tank will be e l evated by the volume reduction provided by the high-dose waste concentrator.

Thus radioactive material concentrations in HICs wi ll be significant and d ecay prior to shipment may be necessary to meet shipping requirements.

The RPF design includes a hidose waste decay cell with 15 locations for HI Cs or drum pallets. This is the only lo cation identified for storage of high-dose waste. The HICs, containing the same solidified high-dose liquid waste stream, should have simi lar radiation levels and require simi lar decay durations to be acceptable for shipment. The reserve storage locations for high-dose waste H!Cs to account for any delay in shipment of waste is reduced by the number of weeks required for decay in storage. 7of10 NWMl-2017-RAl

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• Request for additional information NUREG-1537, Part 2, Acceptance Criteria/or Section 1 1.2.1 , "Radioactive Waste Management Program ," states: "The program should be sufficiently flexible to accommodate changing radioactive waste loads , changing regulatory requirements, and changing environmental factors, and should remain effective in protecting the health and safety of the facility staff and the pub li c." NUREG-1537 does not contain guidance regarding what constitutes adequate storage to be "sufficiently flexible. " Waste production volumes presented in PSAR Table 11-6 indicate waste generation rates of between one and two HI Cs per week. At one HIC per week the maximum available decay in storage in the hi>:h-dose waste decay cell would be 15 weeks. Any de l ay in transport and/or disposal strategies assumed could reduce reserve storage capacity.

Fai l ure to main t ain reserve on-site storage capacity could significantly disrupt the as-designed, continuous operations RAI 9.7-Sa Provide add i tional informa t ion that qua n tifies the du r ation of required decay in storage in t h e dose waste decay eel/for waste representative of the "tota l equilibrium in-process inventory" before the dose rate on the HIC is sufficiently low to meet the shipping requirements in the proposed cask. Decay storage time for high-dose waste packages was estimated as part of a waste disposal cost evaluation documented in NWMI-2015-RPT-006 , Waste Disposal Cost Evaluation.

The decay time is based on waste radionuclide inventory and heat generation that complies with design criteria for the 1O-l60B cask. Evaluation results indicate waste generated during MURR and OSTR target processing must be decayed to greate r than 12 and 24 weeks after EOI , respectively , to comp l y with cask design criteria.

The waste radionuclide inventory is primari l y derived from uranium recovery and recycle operations.

Lag storage in the facility de l ays processing of uranium recovery solutions to greater than 3 weeks after EOI. Therefore , high-dose waste container l ag storage times range from 9 to 21 weeks for compliance with cask design criteria.

RAI 9.7-Sb Justify that the hi>:h-dose waste decay cell has adequate capacity to accommodate both waste bein>: decayed in storage and waste shipments that are delayed due to weather or other conditions t hat delay transport.

NWMI-2013-021 , Table 11-6 , presents the high dose liquid waste volume generated by the processing steps. As discussed in Section 9.7 , the high-dose liquid waste is feed to a concentrator that reduces the volume of liquid waste that will be solidified.

The flowsheet concentration factor for the high-dose concentrator is about 6, b ased on sodium nitrate so l u bili ty. At nomina l processing conditions, the RP F is processing 44 weeks of MURR targets and eight weeks of OSTR targets each year. MURR processing produces one RIC every four weeks; during OSTR target processing about one HIC is produced each week. So nominally 18 RICs are generated each year , and the longest decay storage time is approximately24 weeks (from EOI) for MURR waste. The RPF RIC storage area , as sized, allows for transportation disruptions and for the needed decay time. 8o f10 NWMl-2017-RAl-002, Rev. 0 REFERENCES 10 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register , as amended. 10 CFR 50 , "Domestic Licensing of Production and Utilization Facilities

," Code of Federal Regulations , Office of the Federal Register, as amended. 10 CFR 70 , " Domestic Licensing of Special Nuclear Material ," Code of Federal Regulations , Office of the Federal Register , as amended. ANSI/ ANS-55.1 , Solid Radioactive Waste Proc e ssing System for Light-Water Cooled R eactor Plants , American Nuclear Society , La Grange Park , Illinois , 1992 (R2009) ANSI/ANS-55.4 , Gaseous Radioa ctive Waste Processing Systems for Light-Water R eactor Plants , American Nuclea r Society , La Grange Park, Illinois , 1993 (R2007) ANSI/ANS-55

.6, Liquid Radioactive Waste Pro cessi ng System for Light-Water Cooled R eacto r Plants , American Nuc l ear Society, La Grange Park , Illinoi s, 1993 (R2007) NRC, 2012 , Final Interim Staff Guidance Augm e nting NUREG-1537 , "Guidelines for Preparing and R eviewing Applications for the Licensing of Non-Power Rea ctors," Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous R eactors, Docket Number: NRC-2011-0135 , U.S. Nuclear Regulatory Commission , Washington , D.C., October 17, 2012. NUREG-1537 , Guidelines for Preparing and R e viewing Appli c ations for th e Licensing of Non-Power Reactors -Format and Content, Part 1 , U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation , Washington, D.C., February 1996. NUREG-1537 , Guidelines for Preparing and R e viewing Applications for the Licensing of Non-Power R e actors: Standard R eview Plan and Acceptance Criteria, Part 2, U.S. Nuc l ear Regulatory Commission , Office of Nuclear Reactor Regulation , Washington , D.C., February 1996. NWMI-2013-021 , Construction Permit Application for Radioisotope Produ ction Facility , Rev. 0 , Northwest Medica l Isotopes , LLC , Corvallis , Oregon , June 29 , 2015. NWMI-2014-RPT-006 , MCNP 6.1 Validations wit h Continuous Energy ENDF I B-VII.l Cro ss Sections , Rev. 1 , Northwest Medical Iso top es, LLC , Corvallis, Oregon, March 2017. NWMI-20 l 5-RPT-006, Waste Disposal Cost Evaluation , Rev. 0 , Northwe st Medical Isotopes , LLC, Corva lli s, Oregon , March 2015. Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plant s, Rev. 2 , U.S. Nuclear Regulatory Commission, Washington, D.C., November 2001. Regulatory Guide 3.10, Liquid Waste Treatment Syst e m Design Guide for Plutonium Processing and Fuel Fabrication Plants, U.S. Nuclear Regulatory Commission , Washington , D.C., June 1973 (R2013). 9of10 NWMl-2017-RAl-002, Rev. 0 This page intentionally left blank. 10of10

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... NWMl-2017-RAl-002 , Rev. 0 Attachment A Section 9.7.2, "Control and Storage of Radioactive Waste" of NWMI-2013-021, Construction Permit Application for Radioisotope Production Facility A-i

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9.7.2 Control

and Storage of Radioactive Waste NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems The radioactive waste control and storage systems are designed to ensure that (1) any potential malfunctions do not ca u se accidents in the RPF or uncontrolled release of radioactivity , and (2) in th e event radioactive material is released by the operation of one of these systems, potential radiation exposures would not exceed the limits of 10 CFR 20 and remain consistent with the NWMI ALARA program. No function or malfunction of the auxiliary systems will interfere with or prevent safe shutdown of the RPF. 9.7.2.1 Design Basis The waste handling system design basis is provided in Chapter 3.5.2.7.6. 9.7.2.2 System Description To fulfill the design basis , th e 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 handling Waste handling Waste Staging and Shipping Buildin g (Class A storage) These functions are described in detail in the following subsections.

Figure 9-21 summarizes the weekly design b asis vo lumes and the average annua l weekly vo lum es of all waste handling process streams. The design basis vo lum e is based on eight Univers i ty of Missouri Research Reactor (MURR) targets and 30 Oregon State University (OSU) TR1GA 1 R eactor (OST R) targets per week to provide appropriately sized tanks. The annua l weekly average is based on proces sing eight MURR targets per week for 44 weeks per year and 30 OSTR targets per week for eight weeks p er year and is used in the sizing of the high-dose decay storage. 1 TRI GA (Training , R esearc h , I so top es, General Atomics) is a r egis t ered trademark of General Atomics , San D iego, California. 9-62

[P roprietary Inf ormatio n] Fig ure 9-20. Waste Ma n agement Process F l ow Di ag r a m a n d Process F l ow S t reams NWM l-20 1 3-021 , Rev. 1 Chap t er 9.0 -Au xili ary System s 9-03 NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems 9.7.2.2.1 High-Dose Liquid Waste Handling Figure 9-21 shows the location in the hot cell area where the high-dose liquid waste will be processed. High-dose liquid waste will be collected in the high-dose waste collection tank (shown in Figure 9-22), which will provide the needed handling capacity to match the volume of liquid waste generated by the upstream processes.

Chapter 4.0 provides descriptions of the high-dose liquid streams that will be directed to the collection tank. [Propri e tary Information]

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 wast e tank capacities. The process streams include: * *

• Caustic scrubber waste Oxidizing column waste NO x absorber waste Regeneration waste from Ti{h #1 IX 9-64 Raffinate/rinsate from #2 IX Raffinate/rinsate from #3 IX U IX waste

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• 0 NOITMWIST flftOICM.

ISCmlftl [Proprietary Inform at i on] NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxi liary Systems Figure 9-22. Simplified High-Dose Waste Handling Process Flow Diagram Tank ID WH-TK-100 WH-TK-240 Ta ble 9-5. High-Dose Waste Tank Capacities Tank capacity Description/purpose High-dose waste acc umul ation tank [Proprietary Information]

[Proprietary Informati on] High-dose concentrate accumulation tank (Proprietary Information]

[Proprietary Information]

9-65 NWMI *:.**.*.* . .......... . .* ! . ""'1IMUT ---lllYW'll NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Additions to the collection tank are in discrete , analyzed batches. Sodium hydroxide solution will be added as needed to neutralize any excess acidity. The neutralized liquid will be forwarded to the dose waste concentrator , where water is evaporated from the high-dose liquid , condensed , and directed to the condensate collection tank. The evaporator bottoms will be directed to a high-dose concentrate collection tank. Figure 9-23 shows the arrangement of the high-dose waste handling equipment.

A HIC will be transferred into the high-dose waste treatment hot cell through the HIC transfer drawer , and docked with the high-dose solidification mixer. Solidification agent will be transferred to the designated bin from 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 that may consist of an in-line mixer or a sacrificia l paddle within the HIC. After filling and mixing are comp l ete, the high-dose solidification mixer will b e disengaged, and the HIC lidded and prepared for transfer to the high-dose waste decay subsystem for storage. [Proprietary Information]

Figure 9-23. High-Dose Waste Treatment and Handling Equipment Arrangement 9.7.2.2.2 Low-Dose Liquid Waste Handling Figure 9-24 shows the location of the low-dose liquid waste collection tank. Low-dose condensate from the high-dose concentrator will be held in the condensate collection tank (Figure 9-25). Chapter 4.0 provides descriptions of the low-dose liquid streams that will be directed to the collection tank. The process stream volumes are summarized in Figure 9-20, a nd Table 9-6 provides the lo w-dose waste tank capacities.

Low-dose liquid received from other upstream processes, combined with the low-dose condensate not recycled , will be transferred to the low-dose waste collection tank where the contents of the tank will be 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 lo cated on the second floor (Figure 9-24). In these heated tanks , the liquid will be held at elevated temperatures (60°C [140°F]), and high rates of ventilation air will be passed through the tank. The heated tank contents , plus the high rate of ventilation , will evaporate excess water, reducing the volume of solid waste generated. Samples wi ll be collected and analyzed to ensure compliance with waste acceptance criteria.

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NWMl-2013-021 , Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-24. Low-Dose Liquid Waste Evaporation System Location 9-67

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• MJITitW{ST V£DttAL llOTOftS [Proprietary Inform a tion] NWMl-2013-021, Rev. 1 Chapter 9.0-Auxiliary Systems Figure 9-25. Low-Dose Liquid Waste Disposition Process Tank ID WT-TK-400 WH-TK-420 WH-TK-500 WH-TK-530 Table 9-6. Low-Dose Waste Tank Capacities Description/purpose Condensate tank for high-dose evaporator Low-dose waste accumulation tank Low-dose w a ste evaporation tank (LD-1) Low-dose evaporation tank (LD-2) 9-68 Tank capacity .. [Proprietary Information]

[Proprietary Information]

[Proprietary Information]

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NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The partially concentrated low-dose liquid waste will be transferred to the low-dose waste solidification area (Figure 9-26), where the waste will be metered into a drum that has been placed in the low-dose solidification hood (WH-EN-600). Solidification product vendor information indicates that a ratio of [Proprietary Information]

of solidification agent is sufficient to solidify [Proprietary Information]

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 bot cell. Figure 9-27 provides the flow diagram for the spent resin dewatering subsystem. This subsystem will transfer uranium recovery and recycle system spent IX resin slurry from the spent resin collection tanks located in the tank hot cell (Figure 9-28) to the dewatering filling head in the high-dose waste treatment hot cell (Figure 9-23). The dewatering filler head will remove liquid from the resin. Dry resin will be collected in a waste drum , and the liquid returned to the low-dose waste collection tank. The solid waste drum transfer drawer (WH-TP-810) (Figure 9-23) will be opened , and the high-dose waste handling crane will be used to lift the drum and place it in the solid waste drum scan for characterization.

After characterization is complete , the drum will be transferred by the high-dose waste handling crane from the solid waste drum scan feed conveyor and placed into a five-drum rack. As determined by characterization , the drum wiU either be held for decay storage or transferred to the hidose waste handling system for transfer to a shipping cask. 9-69

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• HO<<TIIWHT MUKM. tSOTift1 [Proprietary Information]

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

Figure 9-28. Spent Resin Collection Tanks Location 9-70

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• HOlmlWllT

.... "°""' 9.7.2.2.4 Solid Waste Encapsulation NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-29 provides the flow diagram for the solid waste encapsu l ation 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 mix.er (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. Th e solid waste drum transfer drawer will be opened, and the high-dose waste handling crane will be used to lift the drum and place it in the solid waste drum scan for characterization.

After characterization is complete , the drum will be transferred by the high-dose waste handling crane from the solid waste drum scan feed c onveyor 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 racks with high-dose source terms. After HICs or five-drum racks have been filled and lidded in the high-dose waste treatment hot cell , they will be transferred to the high-dose waste decay subsystem.

[Proprietary Information]

Figure 9-30. High-Dose Waste Decay Operational Flow Diagram 9-71

.......... : ..... NWMI .......... . * ....nMUT .... aotw(I NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The high-dose waste decay cell lift (WH-L-900) (Figure 9-31) will lower the HIC or five-drum rack into the high-dose waste decay cell , where the high-dose waste decay cell conveyor (WH-eN-900) wi ll transfer the me or five-drum rack to its decay storage position. The me or five-drum rack will remain in storage for a set amount time to a llow for short-lived radioisotopes in the waste to decay to lower levels. When the Hie or five-drum rack has decayed to an acceptable activity level, the high-dose waste decay cell conveyor (WH-eN-900) will transfer the Hie or five-drum rack to the high-dose waste decay cell lift , where the me 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 [Propri etary Information]

Figure 9-32. High Dose Waste Handling Operational Flow Diagram cask. When a me 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 me or two five-drum racks, from within the high-dose waste handling area , through the high-dose waste shipping transfer port , and into a shipping cask. 9-72 NWMI ..**.. . .. :.:. * * *

• NOITIMtSl WDICAL ll01VPlS 9.7.2.2.7 Waste Handling NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems The simp lifi ed operational flow diagram for the waste handlin g subsystem is shown in Figure 9-33. [Proprietary Information]

Figure 9-33. Waste Handling Flow Diagram The waste h andling 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 lo wboy trailer transporting an em pty DOT-approved cask (Figure 9-34). The truck, trailer , and shippi n g cask will enter the RPF to the waste management lo ading bay via an exterior faci lity high-bay door. The ship pin g cask will then be documented for material tracking an d accountability per the safeguards and security system requirements. Operators will u se the utility system's truck bay spray wand for any necessary wash-down of the truck , trailer , or shi pping cask wh il e lo cated in the waste management loading b ay. The operators will remove the shipping cask's upper impact limiter using the was t e ship ping overhead crane (WH-L-1100) (Figure 9-34). The upper impact limit er will b e plac ed in the designated impact limiter l anding zone and secured. Operators will unbolt the lid and pr e par e the DOT-approved shipping cask for l oading per the cask loading and unloading procedure.

At this po int, the truck, trailer , and shipping cask will ente r the waste loading area via a high-b ay door. Th e trailer containing the approved shipping cask will b e positioned belo w the high-dose waste sh ipping transfer port (WH-TP-1000) of the contaminated waste system. The truck will b e disconnected from the trailer and exit the RPF via the high-bay doors in which the ve hicle entered. All high-b ay doors will be verified as c lo sed a nd the shipping cask will then b e in position and ready for loading per the contaminated waste system procedures.

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NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems Figure 9-34. Waste Handling Equipment Arrangement After the DOT-approved cask has been loaded , the shipping cask will be separated from the high-dose waste shipping transfer port (WH-TP-1000). The truck will enter the RPF into the waste management loading bay via an exte rior facility high-bay door, and operators will use the utility system's truck bay overhead spray wand for any necessary wash-down of the truck while located in the waste management loading bay. The truck will then enter the waste loading area via a high-bay door. The truck will be connected to the trailer and exit to the waste loading area in the waste management loading bay. At this point , the facility process control and communications system will allow operators to replace the shipping cask's upper impact limiter using the waste shipping overhead crane (WH-L-1100).

The shipping cask will be documented for material tracking and accountability per the safeguards and security system requirements (Chapter 12.0). The truck , trailer , and shipping cask will exit the RPF through the high-ba y 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 b e docked to th e 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 materia l tracking and accountability per the safeguards and security system requirements. The pallet jack carrying an empty waste drum pallet will be transferred to the shipping and receiving airlock door , where the empty waste drums will enter the contaminated waste system for loading. After the waste drums have been loaded with liquid low-do se 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) will then enter the waste management loading bay via an exterior facility high-bay door. 9-74 NWMI .*.:.** ... * . ........ *. '

• rftN!nMm MCDCM. lWRlftS NWMl-2013-021, Rev. 1 Chapter 9.0 -Auxiliary Systems A waste shipping truck will also enter the waste management loading bay via an exterior facility high-bay door. Operators will open the high-bay door to the waste loading ar e a and use the forklift to load the waste drum pallet into the truck. The shipping truck will then be documented for material tracking and accountability per the safeguards and security system requir e m e nts. The truck containing the waste pallets will exit the RPF through the high-bay doors in which the vehicle entered. 9.7.2.2.8 Waste Staging and Shipping Building (Class A Storage) Th e Waste Staging and Shipping Building will be approximately

[Proprietary Information]

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

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

• IROFS RS-01 , "Hot Cell Liquid Confinement Boundary" IROFS RS-03, "Hot Cell Secondary Confinement Boundary" IROFS RS-04 , " Hot Cell Shielding Boundary" IROFS RS-08 , " Sample and Analysis of Low Dose Waste Tank Dose Rate Prior to Transfer Outside the Hot Cell Shielding Boundary" IROFS RS-10, "Active Radiation Monitoring and Isolation of Low Dose Waste Transfer" IROFS CS-14 , "Active Discharge Monitoring and Isolation" IROFS CS-15 , " Independent Active Discharge Monitoring and Isolation" IROFS CS-16 , "Sampling and Analysis of Uranium Mass or Concentration Prior to Discharge or Disposal" IROFS CS-17 , "Independent Sampling and Analysis 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 , " Gram Estimator Survey of Accessible Surfaces for Gamma Activity" IROFS CS-23 , "Non-Destructive Assay (NDA) of Items with Inaccessible Surfaces" IROFS CS-24 , "Independent NDA of Items with Inaccessible Surfaces" IROFS CS-25 , "Target Housing Weighing Prior to Disposal" IROFS CS-26 , "Active Discharge Monitoring and Isolation" )ROFS FS-02, "Overhead Cranes" Additional information on the analyses that identified these IROFS is provided in Chapter 13.0. 9.7.2.4 Instrumentation and Control Requirements Instrumentation and control requirements for the processes associated with the control and storage of radioactiv e waste are discussed 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. 9-75