ML17128A067
| 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) | |
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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
. *. ~ *.*! : . NORTHWEST MEDICAL ISOTOPES 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 April 2017 Prepared by:
Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, OR 97330
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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 IRevision 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:
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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 NWM l-2017-RAl-002, Rev. 0
~*.*~ * -TWffllllllOICAlllllWQ TERMS Acronyms and Abbreviations 23su uranium-235 ANS American Nuclear Society ANSI American National Standards Institute AOA area of applicability CFR Code of Federal Regulations DOT U.S. Department of Transportation DBE design basis event EOI end of irradiation HIC high-integrity container IROFS items relied on for safety ISG Interim Staff Guidance MCNP Monte Carlo N-Particle Mos margin of subcriticality MURR University of Missouri Research Reactor NCS nuclear criticality safety NRC U.S. Nuclear Regulatory Commission NWMI Northwest Medical Isotopes, LLC OSTR Oregon State University TRIGA Reactor PSAR preliminary safety analysis report QA quality assurance RAI request for additional information RPF radioisotope production facility SAR safety analysis report SR safety-related SSC structures, systems, and components U .S. United States USL upper subcritical limit
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NWMl-2017-RAl-002 , Rev. 0 CHAPTER 3.0 - DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS Section 3.5 - Systems and Components RAI 3.5-10
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- A Section 50.9, "Completeness and accuracy ofinformation," of JO CFR Part 50 requires that information maintained by the applicant be complete and accurate in all material respects.
The definition/or items relied on/or safety (JROFS) is provided in JO CFR 70.4, "Definitions. "
The definition states:
Items relied on for safety mean structures, systems, equipment, components, and activities of personnel that are relied on to prevent potential accidents at a facility that could exceed the peiformance requirements in § 70. 6J or to mitigate their potential consequences. This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities qfpersonnel (i.e., beyond those in the minimum set necessary.for compliance with the peiformance requirements) as items relied on/or safety.
In response to RAJ 3.5-2, NW.MI revised its PSAR. The NWMI PSAR, Revision A, Section 3.5.J .3, "Nuclear Safety Classifications for Structures, Systems, and Components, "states the following classification/or structures, systems, and components (SSCs):
Safety-related IROFS - SSCs identified through accident analyses as required to meet the peiformance requirements of JO CFR 70.61 (see Table 3-2)
Safety-related - SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of workers, the public, and environment, includes SSCs to meet J 0 CFR 20 normal release or exposure limits Non-safety-related - SSCs related to the production and delivery ofproducts or services that are not in the above safety classifications NW.Ml PSAR, Revision A, Section 3.5.J.2, "Classifications Definitions, " defines safety-related (SR) as follows:
The definitions used in classifications of SSCs include the following:
Safety-related is a classification applied to items relied on to remain fanctional during or following a design basis event (DBE) to ensure the:
- Integrity of the facility infrastructure
- Capability to shut down the facility and maintain it in a safe shutdown condition
- Capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the applicable guideline exposures set forth in JO CFR 70.6J, "Peiformance Requirements, " as applicable Additional information is needed to understand the application ofthese defmitions to SSCs at the proposed NWM!facility. However, it is not clear if whether all IROFS are classified as SR.
Additionally, there appears to be an inconsistency between 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 uses guidelines exposures set forth in JO CFR 70.6J and SR in Section 3.5.J.3 includes SSCs to meet JO CFR Part 20 normal release or exposure 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 eiformance requirements in JO CFR 70.6J(b), JO CFR 70.6J(c), and JO CFR 70.6J(d).
RAI 3.5-10a Clarify if all engineering safety features and administrative controls IROFS are classified as SR.
Yes, all engineering safety features and administrative controls items relied on for safety (IROFS) are classified as safety-related (SR).
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NWMl-2017-RAl-002, Rev. 0 Re uest for additional information RAI 3.5-10b With respect to the PSAR, Revision A, Sections 3.5.1 .2 and 3.5.1.3, clarify if all SSCs that are used to comply with 10 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 ofSR 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 Regulatory 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 IROFS 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 CFR 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.
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- .-.~ . NWMI NWMl-2017-RAl-002, Rev. 0
- ~~~;~* *TIMmMUtCALlllllPll 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 ofa reference manual or validation report should include the following: a summary of the theory of the methodology 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 ofphysical parameters in the area of applicability (AOA). Additional information is needed to determine if the range ofparameters to be modeled to support the facility design are within the AOA and the justification ofyour margin of subcriticality 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 that "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 NRC staff notes that many factors can affect the neutron spectrum. For example, as enrichment increases, the dimensions offzssile 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.
Additionally, NWMI 's response states that "Fissions from fast neutrons increases, which results in a softening of the neutron spectrum," appears to be inconsistent.
RAI 6.3-17a Provide additional information to address the various effects that can affect the neutron spectrum or, if a general case that increasing enrichment always results in spectral softening cannot be made, provide alternate justification for the margin ofsubcriticality.
[Proprietary Information]
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NWMl-2017-RAl-002, Rev. 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 particular, justify not including benchmarks such as IEU-SOL-THERM-001 that are known to have a large negative bias. Exclusion of these benchmarks could be non-conservative if design calculations are susceptible to similar biases.
[Proprietary Information]
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NWMl-2017-RAl-002, Rev. 0 CHAPTER 9.0 - AUXILIARY SYSTEMS Section 9.7 - Other Auxiliary Systems Re uest for additional information (Applies to Section 50.9, "Completeness and accuracy of information," of JO CFR Part 50 requires that RAls 9.7-3 information maintained by the applicant be complete and accurate in all material respects.
through 9.7-5) 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 specifications ofprincipal components and any special 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 control and stora~e 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 ofByproduct, 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 the 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 ofradioactive material beyond those analyzed in Chapter 13 of the SAR.
NUREG-1537, Part 2, Chapter 9, Section 9.5, "Possession and Use ofByproduct, 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 special drains for liquids and chemicals, and air exhaust hoods for airborne materials, with design bases derived in Chapter 11 of the SAR.
NUREG-1537, Part 2, Chapter 9, Section 9. 7, "Other Auxiliary Systems," Acceptance Criteria, states, in part that the design and functional description of the auxiliary system should 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 should 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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NWMl-2017-RAl-002, Rev. 0 Section 9.7 - Other Auxiliary Systems
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- RA1 9. 7-3 NWMI PSAR Section 9. 7. 2, "Control and Storage ofRadioactive 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-dose waste concentrator operation.
PSAR Section 9. 7.2.2.2 provides no preliminary information on water quality requirements for recycle or quantitative estimates ofhow much will be recycled and how much will flow to the low-dose waste collection 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 values presented should reflect the maximum waste generation rates resulting from nominal operational processing capability as presented in PSAR Section 4.1.2.1, "Process Design 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 align with NWMI-2013-021 , Section 4.1.2.1 , "Process Design Basis."
RAI 9.7-3b Provide a discussion ofthe operation ofthe high-dose liquid waste handling system. This discussion should include sufficient information to understand how the contents of the high-dose waste collection tank are sampled and controlled coincident with inputs from the offgas system, the molybdenum waste tanks, and the uranium recovery system.
NWMI-2013-021 , Section 9.7.2, was rewritten to address key elements of these requests for additional information (RAI). The inputs to the high-dose waste collection tank are batch transfers. The inputs must be sampled before transferring to the non-criticality safe, high-dose waste collection tank. The high-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 9.7-4 NWMI PSAR Chapter 3, Table 3-3 "Relevant US. Nuclear Regulatory Commission Guidance, "
does not list Regulatory Guide (RG) 1.143, "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants".
PSAR Section 3.1.7 "Codes and Standards," states the codes and standards listed in Table 3-7 "Design Codes and Standards, " are design inputs but farther states, in part that the specific requirements for that system produced by each applicable reference are identified in system design descriptions. Although Table 3-7, lists ANSIIANS-55.1, "Solid Radioactive Waste Processing System for Light-Water Cooled Reactor Plants," 1992 (R2009), ANSIIANS-55.4, "Gaseous Radioactive Waste Processing Systems for Light-Water Reactor Plants, " 1993 (R2007) and ANSIIANS-55.6, "Liquid Radioactive Waste Processing 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.
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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 ofradioactive 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 Management 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 tankfrom 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 collection tank are batch transfers. The inputs must be sampled before transferring to the non-criticality safe, high-dose waste collection tank. The high-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 control 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 cells to a waste encapsulation area
- Capability to solidify low-dose liquid waste and load drums
- Capability to solidify high-dose liquid waste and load high-integrity containers (HICs)
- Capability to encapsulate solid waste in drums
- Capability to handle and load a waste shipping cask with radiological waste drums or containers
- Storage space for Class A radioactive waste
The radioisotope production facility (RPF) is designedfor continuous sequential batch operation.
After startup and operations longer than the longest 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 elevated by the volume reduction provided by the high-dose waste concentrator. Thus radioactive material concentrations in HICs will be significant and decay prior to shipment may be necessary to meet shipping requirements. The RPF design includes a high-dose waste decay cell with 15 locations for HICs or drum pallets. This is the only location identified for storage of high-dose waste. The HICs, containing the same solidified high-dose liquid waste stream, should have similar radiation levels and require similar 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.
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NWMl-2017-RAl-002, Rev. 0 Request for additional information NUREG-1537, Part 2, Acceptance Criteria/or Section 11.2.1, "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 public."
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 HICs 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 delay in transport and/or disposal strategies assumed could reduce reserve storage capacity. Failure to maintain reserve on-site storage capacity could significantly disrupt the as-designed, continuous operations RAI 9.7-Sa Provide additional information that quantifies the duration of required decay in storage in the high-dose waste decay eel/for waste representative of the "total 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 greater than 12 and 24 weeks after EOI, respectively, to comply with cask design criteria. The waste radionuclide inventory is primarily derived from uranium recovery and recycle operations. Lag storage in the facility delays processing of uranium recovery solutions to greater than 3 weeks after EOI. Therefore, high-dose waste container lag 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 that 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, based on sodium nitrate solubility. At nominal processing conditions, the RPF 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.
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NWMl-2017-RAl-002, Rev. 0 REFERENCES 10 CFR 20, "Standards for Protection Against Radiation," Code ofFederal Regulations, Office of the Federal Register, as amended.
10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Code ofFederal 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 Processing System for Light-Water Cooled Reactor Plants ,
American Nuclear Society, La Grange Park, Illinois, 1992 (R2009)
ANSI/ANS-55.4, Gaseous Radioactive Waste Processing Systems for Light-Water Reactor Plants ,
American Nuclear Society, La Grange Park, Illinois,1993 (R2007)
ANSI/ANS-55 .6, Liquid Radioactive Waste Processing System for Light-Water Cooled Reactor Plants, American Nuclear Society, La Grange Park, Illinois, 1993 (R2007)
NRC, 2012, Final Interim Staff Guidance Augmenting NUREG-1537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors," Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors, Docket Number:
NRC-2011-0135, U.S. Nuclear Regulatory Commission, Washington, D.C., October 17, 2012.
NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors -Format and Content, Part 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February 1996.
NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing ofNon-Power Reactors: Standard Review Plan and Acceptance Criteria, Part 2, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., February 1996.
NWMI-2013-021 , Construction Permit Application for Radioisotope Production Facility, Rev. 0, Northwest Medical Isotopes, LLC, Corvallis, Oregon, June 29, 2015 .
NWMI-2014-RPT-006, MCNP 6.1 Validations with Continuous Energy ENDFIB-VII.l Cross Sections ,
Rev. 1, Northwest Medical Isotopes, LLC, Corvallis, Oregon, March 2017.
NWMI-20 l 5-RPT-006, Waste Disposal Cost Evaluation , Rev. 0, Northwest Medical Isotopes, LLC, Corvallis, Oregon, March 2015.
Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, Rev. 2, U.S . Nuclear Regulatory Commission, Washington, D.C., November 2001.
Regulatory Guide 3.10, Liquid Waste Treatment System Design Guide for Plutonium Processing and Fuel Fabrication Plants, U.S. Nuclear Regulatory Commission, Washington, D.C., June 1973 (R2013).
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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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- NWMI
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NWMl-2013-021 , Rev . 1 Chapter 9.0 - Auxiliary Systems 9.7.2 Control and Storage of Radioactive Waste The radioactive waste control and storage systems are designed to ensure that (1) any potential malfunctions do not cause accidents in the RPF or uncontrolled release of radioactivity, and (2) in the event radioactive material is released 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, the control and storage of radioactive waste will include the following functions:
High-dose liquid waste handling (collection, concentration, and solidification)
Low-dose liquid waste handling (collection, evaporation, recycle and solidification)
Spent resin dewatering Solid waste encapsulation High-dose waste decay High-dose waste handling Waste handling Waste Staging and Shipping Building (Class A storage)
These functions are described in detail in the following subsections.
Figure 9-21 summarizes the weekly design basis volumes and the average annual weekly volumes of all waste handling process streams. The design basis volume is based on eight University of Missouri Research Reactor (MURR) targets and 30 Oregon State University (OSU) TR1GA 1 Reactor (OSTR) targets per week to provide appropriately sized tanks. The annual weekly average is based on processing eight MURR targets per week for 44 weeks per year and 30 OSTR targets per week for eight weeks per year and is used in the sizing of the high-dose decay storage.
1 TRIGA (Training, Research , Isotopes, General Atomics) is a registered trademark of General Atomics, San Diego, California.
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[Proprietary Information]
Figure 9-20. Waste Management Process Flow Diagr a m and Process Flow St reams 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.
[Proprietary 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 waste tank capacities. The process streams include:
- Caustic scrubber waste Raffinate/rinsate from #2 IX
- Oxidizing column waste Raffinate/rinsate from #3 IX
- NOx absorber waste U IX waste Regeneration waste from Ti{h #1 IX 9-64
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- * * ~ 0 NOITMWIST flftOICM. ISCmlftl Chapter 9.0 -Auxiliary Systems
[Proprietary Information]
Figure 9-22. Simplified High-Dose Waste Handling Process Flow Diagram Table 9-5. High-Dose Waste Tank Capacities Tank capacity Tank ID Description/purpose WH-TK-100 High-dose waste accumulation tank [Proprietary Information] [Proprietary Information]
WH-TK-240 High-dose concentrate accumulation tank (Proprietary Information] [Proprietary Information]
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~* .*! . ""'1IMUT - - - lllYW'll 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 high-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. High-dose liquid waste concentrate from the waste concentrate collection tank and solidification agent will be metered into the HIC by the high-dose solidification mixer that may consist of an in-line mixer or a sacrificial paddle within the HIC. After filling and mixing are complete, the high-dose solidification mixer will be disengaged, and the HIC lidded and prepared for transfer to the high-dose waste decay subsystem for storage.
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Figure 9-23. High-Dose Waste Treatment and Handling Equipment Arrangement 9.7.2.2.2 Low-Dose Liquid Waste Handling Figure 9-24 shows the location of the low-dose liquid waste collection tank. Low-dose condensate from the high-dose concentrator will be held in the condensate collection tank (Figure 9-25). Chapter 4.0 provides descriptions of the low-dose liquid streams that will be directed to the collection tank. The process stream volumes are summarized in Figure 9-20, and Table 9-6 provides the low-dose waste tank capacities. Low-dose liquid received from other upstream processes, combined with the low-dose condensate not recycled, will be transferred to the low-dose waste collection tank where the contents of the tank will be analyzed and adjusted with sodium hydroxide (NaOH) to neutralize any residual acids.
Once neutralized, the low-dose waste will then be forwarded to the first of two evaporation tanks located on the second floor (Figure 9-24). In these heated tanks, the liquid will be held at elevated temperatures (60°C [140°F]), and high rates of ventilation air will be passed through the tank. The heated tank contents, plus the high rate of ventilation, will evaporate excess water, reducing the volume of solid waste generated. Samples will be collected and analyzed to ensure compliance with waste acceptance criteria.
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Figure 9-24. Low-Dose Liquid Waste Evaporation System Location 9-67
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Figure 9-25. Low-Dose Liquid Waste Disposition Process Table 9-6. Low-Dose Waste Tank Capacities Tank capacity Tank ID Description/purpose ..
WT-TK-400 Condensate tank for high-dose evaporator [Proprietary [Proprietary Information] Information]
WH-TK-420 Low-dose waste accumulation tank [Proprietary [Proprietary Information] Information]
WH-TK-500 Low-dose waste evaporation tank (LD-1) [Proprietary [Proprietary Information] Information]
WH-TK-530 Low-dose evaporation tank (LD-2) [Proprietary [Proprietary Information] 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.
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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 high-dose waste handling system for transfer to a shipping cask.
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Figure 9-27. Spent Resin Dewatering Operational Flow Diagram
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Figure 9-28. Spent Resin Collection Tanks Location 9-70
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9.7.2.2.4 Solid Waste Encapsulation Figure 9-29 provides the flow diagram for the solid waste encapsulation subsystem. Operators will enter the maintenance gallery and retrieve the solid waste drum cart from the waste collection port and transfer the drum cart into the high-dose waste treatment hot cell (Figure 9-23). The solid waste drum access port will be opened, and the solid waste encapsulation grout 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. The solid waste drum transfer drawer will be opened, and the high-dose waste handling crane will be used to lift the drum and place it in the solid waste drum scan for characterization. After characterization is complete, the drum will be transferred by the high-dose waste handling crane from the solid waste drum scan feed conveyor and placed into a five-drum rack. As determined by the drum's characterization, the drum will either be held for decay storage or transferred to the high-dose waste handling subsystem for transfer to a shipping cask.
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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.
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Figure 9-30. High-Dose Waste Decay Operational Flow Diagram 9-71
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. ~* .~ * ....nMUT . . . . aotw(I 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) will 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 allow 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.
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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 [Proprietary Information]
high-dose waste handling subsystem. This Figure 9-32. High Dose Waste Handling subsystem will provide the capability to remotely Operational Flow Diagram transfer high-dose waste containers into a shipping 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.
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Figure 9-33. Waste Handling Flow Diagram The waste handling subsystem will have multiple material handling capabilities. The liquid high-dose radiological waste and solid radiological waste handling will begin with the arrival of a truck and lowboy trailer transporting an empty DOT-approved cask (Figure 9-34). The truck, trailer, and shipping cask will enter the RPF to the waste management loading bay via an exterior facility high-bay door. The shipping cask will then be documented for material tracking and accountability per the safeguards and security system requirements. Operators will use the utility system's truck bay spray wand for any necessary wash-down of the truck, trailer, or shipping cask while located in the waste management loading bay.
The operators will remove the shipping cask's upper impact limiter using the waste shipping overhead crane (WH-L-1100) (Figure 9-34). The upper impact limiter will be placed in the designated impact limiter landing zone and secured. Operators will unbolt the lid and prepare the DOT-approved shipping cask for loading per the cask loading and unloading procedure. At this point, the truck, trailer, and shipping cask will enter the waste loading area via a high-bay door. The trailer containing the DOT-approved shipping cask will be positioned below the high-dose waste shipping transfer port (WH-TP-1000) of the contaminated waste system. The truck will be disconnected from the trailer and exit the RPF via the high-bay doors in which the vehicle entered. All high-bay doors will be verified as closed and the shipping cask will then be in position and ready for loading per the contaminated waste system procedures.
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Figure 9-34. Waste Handling Equipment Arrangement After the DOT-approved cask has been loaded, the shipping cask will be separated from the high-dose waste shipping transfer port (WH-TP-1000). The truck will enter the RPF into the waste management loading bay via an exterior facility high-bay door, and operators will use the utility system's truck bay overhead spray wand for any necessary wash-down of the truck while located in the waste management loading bay. The truck will then enter the waste loading area via a high-bay door. The truck will be connected to the trailer and exit to the waste loading area in the waste management loading bay. At this point, the facility process control and 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-bay doors in which the vehicle entered.
The liquid low-dose radiological waste handling process will begin with the arrival of a truck transporting the empty waste drum pallets to the fresh and unirradiated shipping and receiving area. The receiving area door will be opened, and the truck will be docked to the receiving bay, allowing for transfer of the waste drum pallets into the RPF. Pallet-loaded empty waste drums will be unloaded from the truck using the waste handling pallet jack (WH-PH-1100). All unloaded empty waste drum pallets will then be documented for material tracking and accountability per the safeguards and security system requirements.
The pallet jack carrying an empty waste drum pallet will be transferred to the shipping and receiving airlock door, where the empty waste drums will enter the contaminated waste system for loading.
After the waste drums have been loaded with liquid low-dose radiological waste and re-palletized, a pallet containing full waste drums will be transferred via the waste handling pallet jack (WH-PH-1100) from the shipping and receiving airlock door to the waste loading area. The waste handling forklift (WH-PH-1110) will then enter the waste management loading bay via an exterior facility high-bay door.
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- 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 area and use the forklift to load the waste drum pallet into the truck. The shipping truck will then be documented for material tracking and accountability per the safeguards and security system requirements. The truck containing the waste pallets will exit the RPF through the high-bay doors in which the vehicle entered.
9.7.2.2.8 Waste Staging and Shipping Building (Class A Storage)
The Waste Staging and Shipping Building will be approximately [Proprietary Information] and will provide additional waste storage and shipping preparation for Class A radioactive waste prior to disposal.
9.7.2.3 Operational Analysis and Safety Function Chapter 13.0, Section 13.2 evaluates the accident sequences that involve fissile solution or solid materials being introduced into systems not normally designed to process these solutions or solid materials. The waste handling system is not geometrically safe; therefore, a number of IROFS have been identified.
IROFS RS-01 , "Hot Cell Liquid 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 radioactive 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.
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