ML21095A223

Shine Medical Technologies, LLC, Revisions to Final Safety Analysis Report, Chapter 11, Rev. 3, Radiation Protection Program and Waste Management

ML21095A223
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Site:	SHINE Medical Technologies
Issue date:	03/23/2021
From:	SHINE Medical Technologies
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RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT TABLE OF CONTENTS tion Title Page 1 RADIATION PROTECTION ............................................................................... 11.1-1 11.1.1 RADIATION SOURCES ................................................................... 11.1-1 11.1.2 RADIATION PROTECTION PROGRAM .......................................... 11.1-6 11.1.3 ALARA PROGRAM ........................................................................ 11.1-14 11.1.4 RADIATION MONITORING AND SURVEYING ............................. 11.1-18 11.1.5 RADIATION EXPOSURE CONTROL AND DOSIMETRY ............. 11.1-21 11.1.6 CONTAMINATION CONTROL EQUIPMENT AND FACILITY LAYOUT GENERAL DESIGN CONSIDERATIONS FOR 10 CFR 20.1406 ............................................................................. 11.1-26 11.1.7 ENVIRONMENTAL MONITORING ................................................ 11.1-28 2 RADIOACTIVE WASTE MANAGEMENT .......................................................... 11.2-1 11.2.1 RADIOACTIVE WASTE MANAGEMENT PROGRAM ..................... 11.2-1 11.2.2 RADIOACTIVE WASTE CONTROLS .............................................. 11.2-4 11.2.3 RELEASE OF RADIOACTIVE WASTE ............................................ 11.2-8 3 RESPIRATORY PROTECTION PROGRAM ..................................................... 11.3-1 4 REFERENCES ................................................................................................... 11.4-1 NE Medical Technologies 11-i Rev. 0

mber Title 1-1 Parameters Applicable to Target Solution Radionuclide Inventories 1-2 Nominal Versus Safety Basis Radionuclide Inventories in Target Solution 1-3 Irradiated Target Solution Activity for Select Radionuclides Pre-Extraction 1-4 Radiation Areas at the Main Production Facility 1-5 Airborne Radioactive Sources 1-6 Estimated Derived Air Concentrations 1-7 Key Parameters for Normal Yearly Release Calculation 1-8 Estimated Annual Releases from Normal and Maintenance Operations (Nuclides with Greater than 1 Ci Annual Release) 1-9 Liquid Radioactive Sources 1-10 Solid Radioactive Sources 1-11 Administrative Radiation Exposure Limits 1-12 Radiation Monitoring Equipment 1-13 Radiological Postings 1-14 Environmental Monitoring Locations 2-1 Estimated Annual Waste Stream Summary 2-2 Waste Methodology for Accelerator 2-3 Waste Methodology for Spent Columns 2-4 Waste Methodology for Process Glassware 2-5 Waste Methodology for Consolidated Liquids 2-6 Chemical Composition and Radiological Properties of Liquid Waste Streams NE Medical Technologies 11-ii Rev. 1

mber Title 1-1 Probable Radiation Area Designations Within the SHINE RCA, Ground Floor Level 1-2 Estimated Derived Air Concentrations, Ground Floor Level 1-3 Radiation Protection Organization 1-4 Environmental Dosimeter Locations NE Medical Technologies 11-iii Rev. 0

onym/Abbreviation Definition RA as low as reasonably achievable SI American National Standards Institute AS criticality accident alarm system M continuous air monitor S continuous air sampler BEM carbon delay bed effluent monitor DE committed effective dose equivalent MP Community Environmental Monitoring Program O Chief Executive Officer curies r curies per year centimeter O Chief Operating Officer ground level deposition factor C derived air concentration T U.S. Department of Transportation NE Medical Technologies 11-iv Rev. 0

onym/Abbreviation Definition

/100 cm2 disintegrations per minute per 100 square centimeters O data quality objectives A U.S. Environmental Protection Agency r cubic feet per year PA high efficiency particulate air hour A high radiation area 1 iodine-131 irradiation facility irradiation unit iodine and xenon purification and packaging kilometers krypton kilowatt low-enriched uranium lower level of detection NE Medical Technologies 11-v Rev. 0

onym/Abbreviation Definition low level waste low specific activity liquid scintillation counter PS light water pool system RLAP Multi-Agency Radiological Laboratory Analytical Protocols Manual maximum exposed individual PS molybdenum extraction and purification system LW mixed low level waste molybdenum 99 molybdenum-99 m millirem m/hr millirem per hour m/yr millirem per year v millisievert 6 nitrogen-16 AS neutron driver assembly system NE Medical Technologies 11-vi Rev. 0

onym/Abbreviation Definition DS neutron flux detection system LS primary closed loop cooling system NL Pacific Northwest National Laboratory E personal protective equipment B primary system boundary VS process vessel vent system radiation area M radiation area monitor A radiologically controlled area RA Resource Conservation and Recovery Act MP radiological environmental monitoring program WI radioactive liquid waste immobilization WS radioactive liquid waste storage F radioisotope production facility C Radiation Safety Committee CC Radiation Safety Information Computational Center NE Medical Technologies 11-vii Rev. 0

onym/Abbreviation Definition Z1 radiological ventilation zone 1 Z2 radiological ventilation zone 2 P radiation work permit SS subcritical assembly support structure AS subcritical assembly system M stack release monitor C system, structure, and component sievert P toxicity characteristic leaching procedure E total effective dose equivalent GS TSV off-gas system tritium purification system S target solution preparation system S target solution storage system target solution vessel 35 uranium-235 NE Medical Technologies 11-viii Rev. 0

onym/Abbreviation Definition SS uranium receipt and storage system RA very high radiation area vacuum transfer system C waste acceptance criteria S Waste Control Specialists annual average relative atmospheric concentration xenon NE Medical Technologies 11-ix Rev. 0

RADIATION PROTECTION

.1 RADIATION SOURCES SHINE facility is designed to generate molybdenum-99 (Mo-99) for use as a medical ope. The process of producing Mo-99 involves irradiating a uranyl sulfate target solution with eutron source in a subcritical assembly to cause fission. Irradiation of the target solution ates Mo-99 along with other radioactive fission and activation products. When the irradiation e is complete, the radioactive materials are transferred to various locations in the facility to plete the separation and purification processes. This section identifies sources of radiation radioactive materials received, used, or generated in the facility; sources and the nature of orne, liquid or solid radioactive materials; and the type of radiation emitted (alpha, beta, ma, and neutron).

lysis has been performed that quantifies the radionuclide inventory for normal operations in main production facility. The highest radionuclide inventory for one target solution batch ts in the target solution vessel (TSV) at the end of the irradiation cycle. As the target solution rocessed in the facility for Mo-99 and other medical isotope extraction, solution adjustments, waste handling, radiation sources are transferred within the facility by means of pipes in lded trenches.

re are two scenarios with assumptions listed in Table 11.1-1: nominal and safety basis. The inal parameter values or ranges are the best estimate operating conditions for full power ration of the facility. The safety basis parameter values define the bounding radionuclide ntory relative to the TSV, TSV dump tank, and supercell.

safety basis inventories throughout the facility are generated by using the limiting values for h parameter to maximize the individual inventories. This includes using bounding values for ment partitioning during the extraction process. This approach of maximizing inventories at h location results in an overall facility fission product inventory that is greater than originally erated in the irradiation process. This ensures that the individual safety basis inventories are nding when being used to calculate releases for the safety analysis but makes them uitable for use in analyzing normal operations.

ration of the TSV results in the production of radioactive fission products and actinides dominantly through neutron capture in uranium. Table 11.1-2 provides a summary of the ults for total activity in curies (Ci) from actinides and fission products contained within each batch of target solution after [ ]PROP/ECI of irradiation, [ ]PROP/ECI inal cycles or [ ] PROP/ECI safety basis cycles. The at shutdown values represent the vity contained within the target solution immediately after shutdown of the neutron driver. The

-extraction values are the target solution activity when it is ready to be transferred from the dump tank in the irradiation unit (IU) cell to one of the supercells in the radioisotope duction facility (RPF) to begin the molybdenum extraction process. This represents the ximum expected activity for a target solution batch as it is processed through the RPF. For the inal inventory, the post extraction values are the activity remaining in the target solution wing extraction of Mo and other elements according to best estimate partitioning fractions.

the safety basis inventory, only noble gases were removed during extraction, at bounding

) element partitioning fractions.

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ease Fractions (USNRC, 1986) for the nominal and safety basis radionuclide inventories after

]PROP/ECI of irradiation and the subsequent decay time in the TSV dump tank.

his time, it is ready to be pumped into the supercell to begin the molybdenum extraction and on product removal processes. The cycle and decay times used for the radionuclide ntory generation are listed in Table 11.1-1.

NE uses the following radiation area designations, as defined in 10 CFR 20, including sideration for neutron and gamma dose rates:

• Unrestricted Area means an area to which access is neither limited nor controlled by SHINE. This would be the area beyond the site boundary.

• Radiation Areas (RAs) are those accessible areas in which radiation levels could result in an individual receiving a dose equivalent in excess of 5 millirem (mrem) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (hr) at 30 centimeters from the radiation source or from any surface that the radiation penetrates.

• High Radiation Areas (HRAs) are those accessible areas in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 100 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters from the radiation source or from any surface that the radiation penetrates.

• Very High Radiation Areas (VHRAs) are those accessible areas in which radiation levels from radiation sources external to the body could result in an individual receiving an absorbed dose in excess of 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter from the radiation source or 1 meter from any surface that the radiation penetrates.

SHINE facility is designed and constructed so that the measurable dose rate in the estricted area due to activities at the plant are less than the limits of 10 CFR 20.1301(a)(2).

radiation shielding is designed to ensure that during normal operation internal facility ation dose rates are consistent with as low as reasonably achievable (ALARA) radiological ctices required by 10 CFR 20. The goal for the normal operations dose rate for normally upied locations in the facility is 0.25 mrem/hr at 30 centimeters from the surface of the lding. Radiation levels may rise above the 0.25 mrem/hr level during some operations such ank transfers. At full-power operation of the eight units, portions of the normally occupied a in IF and RPF exceed the 0.25 mrem/hr goal but remain below 5 mrem/hr, except in small tions above the pipe trench during solution transfers. These dose rates were calculated using maximum specified shield plug gap sizes, minimum density shielding materials, and the inal inventories for full power operation.

bulation of normally and transient-occupied areas, dose rates, and designations is provided able 11.1-4. Figure 11.1-1 provides the probable radiation area designations, above grade, in the radiologically controlled area (RCA) at the main production facility.

cedures for transient access to shielded vaults, cells, and rooms ensure doses are ntained ALARA by addressing the following:

• job planning,

• radiation protection coverage,

• survey techniques and frequencies, NE Medical Technologies 11.1-2 Rev. 3

• frequency for updating radiation work permits or their equivalent, and

• placement of measuring and alarming dosimeters.

elded vaults, cells, and rooms designated as high radiation areas or very high radiation areas enoted in Figure 11.1-1 are not normally occupied when those conditions exist.

ministrative procedures address the management oversight and specific control measures ded for entry into high radiation areas and very high radiation areas, if it is ever necessary to

o. The procedures include the process for gaining entry to these areas, such as the control distribution of keys.

ical transient access for maintenance or other necessary work to the shielded vaults, cells, rooms that are usually high radiation areas or very high radiation areas is normally ormed after dose rates have been reduced to at least the level of a radiation area. This is e by removing the radioactive materials or changing the conditions (such as shutting down accelerator in an IU cell), using temporary shielding, and waiting for sufficient decay.

or radiation sources in the facility originate in the target solution. At the end of the TSV diation cycle, irradiated target solution is transferred to one of the three extraction cells for cessing. Off-gas that is purged from the primary system boundary (PSB) is sent to the cess vessel vent system (PVVS), where it travels through carbon guard beds and a series of bon delay beds to allow for capture of iodine and decay of short-lived noble gas nuclides ore being released through the facility exhaust stack. Facility special nuclear material (SNM) ntories are tabulated in Table 4b.4-1.

three sections below describe the major radiation sources in the facility. Other radiological rces in the facility are bounded by the fission product source coming from the TSV described ubsection 11.1.1.2.

.1.1 Airborne Radioactive Sources ioactive sources that could become airborne at the main production facility are primarily m and radioactive gases produced as a byproduct of the Mo-99 production process. The tems handling gaseous radioactive materials include the tritium purification system (TPS) and TSV off-gas system (TOGS), both located in the irradiation facility (IF) area; and the PVVS vacuum transfer system (VTS) located in the RPF. These airborne radioactive materials are tained within closed systems consisting of piping components and tanks. Table 11.1-5 vides information on the various locations, types, and expected dose rates from gaseous oactive sources, as well as isotopic activity inventories applicable to the associated expected e rate calculations.

on-41 is produced in the IU cells during irradiation. Due to the low flow rate out of the primary finement boundary to radiological ventilation zone 1 (RVZ1), most argon-41 decays prior to g released. Approximately 0.02 curies per year (Ci/yr) of argon-41 are released to the ironment through the facility stack.

ogen-16 is produced within the primary cooling loop and the light water pool. Dose rates from e sources are mitigated by delay tanks and biological shielding that limits radiation dose to upied areas adjacent to the shielding.

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ect workers from sources of airborne radioactivity during normal operation and minimize ker exposure during maintenance activities, keeping with the ALARA principles outlined in CFR 20.

ough most process gas systems within the facility are maintained below atmospheric ssure, some leakage of process gases is expected due to the difference in partial pressure ween the system and the surrounding environment. A conservative best estimate of airborne ases due to normal operation and maintenance was performed to estimate derived air centrations (DACs) for the facility.

kage from process systems was estimated based on the number of components and fittings, ievable leak tightness per fitting, permeation through equipment, and partial pressures of orne radionuclides. For processes in hot cells that require routine disconnection of ponents (e.g., extraction columns) special fittings are used to minimize process leakage.

effects of the confinement systems are incorporated into the analysis. The results of the luation, broken down into particulates, halogens, noble gases, and tritium, are provided in le 11.1-6. These values provide a conservative best estimate of the facility DACs.

ure 11.1-2 provides the DAC zoning map for the facility, using the following definitions:

• Zone 1 (< 1.0 DAC);

• Zone 2 (1.0 - 10 DAC); and

• Zone 3 (> 10 DAC).

eous activity from the TSV and process operations is routed through the PVVS which udes carbon delay beds to allow for airborne radionuclides to decay to low enough levels h that normal releases are below the 10 CFR 20 limits. Additional airborne release pathways RVZ1 ventilation of the facility hot cells, flow out of the primary confinement boundary to Z1, and radiological ventilation zone 2 (RVZ2) ventilation of any leakage to the general area terial evaluated for the DAC). These additional pathways do not pass through the carbon y beds but do contain filters as described in Subsection 9a2.1.1. Table 11.1-7 lists key ameters used in the normal release calculation. Tritium releases that are treated by TPS are ligible in comparison to tritium releases to the general area due to maintenance and leakage are not included in Table 11.1-7 or Table 11.1-8.

ual off-site doses due to the normal operation of the SHINE facility have been calculated g the computer code GENII2 (PNNL, 2012). The GENII2 computer code was developed for Environmental Protection Agency (EPA) by Pacific Northwest National Laboratory (PNNL) is distributed by the Radiation Safety Information Computational Center (RSICC). Annual rage relative atmospheric concentration (/Q) values were determined using the methodology egulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of eous Effluents in Routine Releases from Light-Water-Cooled Reactors (USNRC, 1977) with meteorological data in Section 2.3. The /Q values for the maximally exposed individual I), which is the nearest point on the site boundary, and the nearest full-time resident are E-5 sec/m3 and 5.3E-6 sec/m3, respectively.

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ase pathways described above: PVVS, hot cells, primary confinement boundary, and erial leaked to the general area. The dominant source term is the process gases released ugh PVVS. Only nuclides with greater than 1 Ci/yr released are included in the table.

dose analysis considered the estimated release of airborne radionuclides provided in le 11.1-8 and exposure to off-site individuals through direct exposure and potential ironmental pathways, such as leafy vegetable ingestion, meat ingestion, and milk ingestion.

analysis considered variations in consumption and other parameters by age group. The mated annual doses at the MEI and the nearest resident are 4.6 mrem total effective dose ivalent (TEDE) and 0.3 mrem TEDE, respectively, which are less than the limit in 10 CFR 20.

culational methodologies related to accidental releases of airborne radioactive sources are ussed in Chapter 13.

.1.2 Liquid Radioactive Sources re are numerous locations within the main production facility where the presence of oactive liquids results in a source of radiation. These sources (except for as noted below) are ved from the irradiated uranyl sulfate target solution as it is being processed through the lity. The first exception is the primary cooling water, which carries nitrogen-16 and other vation products as it is pumped through the primary closed loop cooling system (PCLS). The ond exception is the production of low-activity fresh uranyl sulfate target solution. These oactive materials are contained within closed systems consisting of piping components and s.

ddition, there are two locations where tritium is expected to collect due to operation of the tron driver assembly system (NDAS). These are the light water pool and the oil used in the AS pumps. The small quantities of tritium released into the IU cell by permeation through and age from the NDAS components is expected to be converted to tritiated water and slowly ease the tritium concentration in the pool water. The oil used in the NDAS pumps is in direct tact with the tritium in the accelerator, causing it to become contaminated with tritium over

. Table 11.1-9 provides information on the various locations, types, and expected doses from id radioactive sources, as well as isotopic activity inventories applicable to the associated ected dose rate calculations.

id radioactive wastes generated at the facility are generally solidified and shipped to a osal facility. Table 11.2-1 contains a list of liquid radioactive waste generated at the facility uding the annual quantities and disposal destinations. Radioactive liquid discharges from the n production facility to the sanitary sewer are made in accordance with 10 CFR 20.2003 and CFR 20.2007. See Section 11.2 for additional information on liquid discharges from the RCA.

.1.3 Solid Radioactive Sources d radioactive sources exist in several locations in the SHINE facility. Fresh, low enriched nium is received at the facility in the form of uranium metal or uranium oxide that has been ched to a nominal 19.75 percent by weight in uranium-235 (U-235). If uranium metal is eived, it is converted to uranium oxide and then to a liquid uranyl sulfate solution. Other solid NE Medical Technologies 11.1-5 Rev. 3

natural uranium neutron multiplier is located in the subcritical assembly. The uranium racts with the neutron flux producing both activation products and fission products that are ined within the metal structure.

ddition, metal components in the IU cell are activated and components of the TOGS contain oactive material. The subcritical multiplication sources for the subcritical assemblies are also ted in the IU cell.

se solid radioactive sources are contained within IU cells, shielded cells, hot cells, or paration areas within the RCA of the facility. Table 11.1-10 provides information on the major d radioactive sources including their location and activity, as well as isotopic activity ntories applicable to the associated expected dose rate calculations. The radionuclide ntory in the solid waste system is a function of the TSV system operation.

t of solid radioactive wastes including annual quantities and disposal destinations is provided able 11.2-1.

posal of solid radioactive waste with respect to storage, monitoring, and management is ussed in Section 11.2.

.1.4 Technical Specifications tain material in this section provides information that is used in the technical specifications.

includes limiting conditions for operation, setpoints, design features, and means for omplishing surveillances. In addition, significant material is also applicable to, and may be renced by, the bases that are described in the technical specifications.

.2 RADIATION PROTECTION PROGRAM radiation protection program protects the radiological health and safety of workers and mbers of the public and complies with the regulatory requirements in 10 CFR 19, 20, and 70.

.2.1 Commitment to Radiation Protection Program Implementation NE has established a radiation protection program with the specific purpose of protecting the ological health and safety of workers and members of the public. The objectives of the gram are to prevent acute radiation injuries (non-stochastic or deterministic effects) and to t the potential risks of probabilistic (stochastic) effects (which may result from chronic osure) to acceptable levels. The SHINE radiation protection program was developed and is lemented commensurate with the risks posed by a medical isotope facility. The program tains the SHINE management policy statement to maintain occupational and public radiation osures ALARA.

radiation protection program meets the requirements of 10 CFR 20, Subpart B, Radiation tection Programs, and is consistent with the guidance provided in Regulatory Guide 8.2, ision 1, Administrative Practices in Radiation Surveys and Monitoring (USNRC, 2011), and SI/ANS 15.11-2016, Radiation Protection at Research Reactor Facilities (ANSI/ANS, 2016).

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RA. The radiation protection program content and implementation are reviewed at least ually as required by 10 CFR 20.1101(c).

radiation protection program includes written procedures, periodic assessments of work ctices and internal/external doses received, work plans, and the personnel and equipment uired to implement the ALARA goal. Protection of plant personnel requires (a) surveillance of control over the radiation exposure of personnel and (b) maintaining the exposure of sonnel not only within permissible limits, but also within ALARA philosophy and exposure ls.

NEs administrative personnel exposure limits for radiation workers are set below the limits cified in 10 CFR 20. This provides assurance that regulatory radiation exposure limits are not eeded and that the ALARA principle is emphasized. Administrative exposure limits are vided in Table 11.1-11.

radiation exposure policy and control measures for personnel are established in accordance requirements of 10 CFR 20 and the guidance in the following regulatory guides:

• Regulatory Guide 8.10, Revision 2, Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as Is Reasonably Achievable (USNRC, 2016)

• Regulatory Guide 8.13, Revision 3, Instruction Concerning Prenatal Radiation Exposure (USNRC, 1999)

• Regulatory Guide 8.29, Revision 1, Instruction Concerning Risks from Occupational Radiation Exposure (USNRC, 1996)

SHINE corrective action process is implemented if (1) personnel dose monitoring results or sonnel contamination levels exceed the administrative personnel limits; (2) if an incident ults in airborne occupational exposures exceeding the administrative limits; or (3) the dose ts in 10 CFR 20 are exceeded.

rmation developed from reportable occurrences is tracked in the corrective action program is used to improve radiation protection practices, decreasing the probability of similar dents.

.2.1.1 Responsibilities of Key Program Personnel key personnel responsible for implementing the radiation protection program are shown in ure 11.1-3 and are discussed below. Chapter 12 discusses the SHINE organization and ponsibilities of key management personnel in further detail.

ef Executive Officer Chief Executive Officer (CEO) is responsible for the overall management and leadership of company.

NE Medical Technologies 11.1-7 Rev. 3

Chief Operating Officer (COO) reports to the CEO and is responsible for overall company rations.

e President Regulatory Affairs & Quality Vice President Regulatory Affairs & Quality reports to the CEO and is responsible for nsing and quality activities.

lity Manager Quality Manager reports to the Vice President Regulatory Affairs & Quality and is ponsible for assuring compliance with regulatory requirements and procedures.

nt Manager Plant Manager is responsible for operation of the facility, including the protection of sonnel from radiation exposure resulting from facility operations and materials, and for pliance with applicable NRC regulations and the facility license. The Plant Manager ignates the authority to approve procedures related to personnel radiation protection to the iation Protection Manager in accordance with the guidance provided in ANSI/ANS-15.1-2007 SI/ANS, 2007). The Plant Manager reports to the COO.

iation Protection Manager Radiation Protection Manager is responsible for implementing the radiation protection gram. The Radiation Protection Manager reports directly to the Plant Manager, independent facility operations. The Radiation Protection Manager has direct access to executive nagement for matters involving radiation protection. The Radiation Protection Manager and ation protection personnel are responsible for:

• Establishing the radiation protection program.

• Generating and maintaining procedures associated with the program.

• Ensuring that ALARA is incorporated into procedures and practiced by personnel, including stopping work when unsafe practices are identified.

• Ensuring the efficacy of the program is reviewed and audited for compliance with NRC and other governmental regulations and applicable regulatory guides.

• Modifying the program based upon experience, facility history, regulatory updates, and changes to guidance documents.

• Adequately staffing the Radiation Protection Department to implement the radiation protection program.

• Ensuring that the occupational radiation exposure dose limits of 10 CFR 20 are not exceeded under normal operations.

• Ensuring administrative radiation dose limits are not exceeded without prior approval from the Radiation Safety Committee.

• Establishing and maintaining an ALARA program.

• Demonstrating, where practical, familiarity and reasoning associated with improvements in ALARA principles and practices, including modifications that were considered and implemented.

NE Medical Technologies 11.1-8 Rev. 3

• Establishing and maintaining a Radioactive Waste Management Program.

• Monitoring worker doses, both internal and external.

• Assuring that the proper radiation protection instrumentation, equipment, and supplies are available at workplaces, in good working order, and are used properly.

• Ensuring calibration and quality assurance of health physics associated radiological instrumentation.

• Establishing and maintaining a radiation safety training program for personnel working in radiologically controlled areas.

• Posting restricted areas and, within these areas, posting radiological areas, as required by the radiation protection program (e.g., airborne radioactivity area, high radiation area, contamination area).

• Informing management of any radiation protection concerns.

rations Manager Operations Manager is responsible for operating the facility safely and in accordance with lity procedures so that effluents released to the environment and exposures to the public and site personnel meet the limits specified in applicable regulations, procedures and guidance uments.

site Personnel site personnel are responsible for performing their work activities in a safe manner. SHINE established policies, procedures and practices to ensure that personnel can work safely in facility. The policies, procedures and practices implement rules and regulations intended to ure workers and the public are protected from specific hazards encountered at the facility.

sonnel whose duties require (1) working with radioactive material, (2) entering restricted as, (3) controlling facility operations that could affect effluent releases, or (4) directing the vities of others, are trained such that they understand and effectively carry out their ponsibilities.

.2.1.2 Radiation Protection Program Staffing and Qualifications radiation protection program staff is assigned responsibility for implementation of the ation protection program functions; therefore, only suitably trained radiation protection sonnel are employed at the facility. The radiation protection staff includes, at a minimum, a iation Protection Manager and radiation control technicians.

ff selection and qualification are addressed in Chapter 12. The Radiation Protection Manager ction and qualification is consistent with the requirements for a Level 2 position. Radiation trol technicians are considered Other Technical Personnel, as described in section 12.1.4.

ficient resources in terms of staffing and equipment are provided to implement an effective ation protection program.

NE Medical Technologies 11.1-9 Rev. 3

radiation protection program is independent of facility operations. This independence ures that the radiation protection program maintains its objectivity and is focused only on lementing sound radiation protection principles necessary to achieve occupational doses and es to members of the public that are ALARA.

.2.1.4 Radiation Safety Committee adiation Safety Committee (RSC) is established to maintain a high standard of radiation ection during facility operations. The RSC oversees activities at the SHINE facility to protect sonnel from unnecessary radiation exposure, prevent contamination of natural resources, and nsure compliance with state and federal regulations governing the possession, use, and osal of radioactive materials. The RSC meets periodically, but at least annually, to monitor lity radiological performance and ALARA implementation, review proposed changes to the ation protection program, identify trends, and set ALARA policy and goals for the facility. The C reviews the results of audits and regulatory inspections, worker suggestions, reportable urrences, and exposure incidents. The RSC assesses changes to the facility for the effect on radiation protection program and the license.

Radiation Protection Manager chairs the RSC. The RSC Charter defines the purposes, tions, responsibility, composition, qualifications, quorum, meeting frequency, and reporting uirements of the RSC.

.2.1.5 Commitment to Written Radiation Protection Procedures iation protection procedures are prepared, reviewed and approved to carry out activities ted to the radiation protection program. Procedures are used to control radiation protection vities in order to ensure that the activities are carried out in a safe, effective and consistent nner. Radiation protection procedures are reviewed and revised as necessary by the iation Protection Manager or designee to incorporate facility or operational changes.

iation protection procedures provide direction for the following activities:

• Facility radiation monitoring, including surveys, personnel monitoring, and sampling and analysis of solid, liquid and gaseous wastes processed or released from the facility

• Calibration of area radiation monitors, facility air monitors, laboratory radiation detection systems, personnel radiation monitors and portable instruments

• Access control, radiological posting, and monitoring of radiological work activities

• Radioactive materials handling and shipment

• Contamination control

• Control of exposures and ALARA implementation

• Control of instrument alarm setpoints

• Administration of the radiation work permit (RWP) process iation protection procedures undergo technical verification and review to ensure compliance regulatory requirements, applicable license conditions and the radiation protection program, well as conformance with industry standard practices, as applicable. Radiation protection cedures are reviewed at least once every three years in accordance with the guidance in NE Medical Technologies 11.1-10 Rev. 3

rk performed in radiologically controlled areas is performed in accordance with the RWP cess. The RWP specifies radiological controls for intended work activities and provides ten authorization for entry into and work within Radiation Areas, High Radiation Areas, Very h Radiation Areas, Contamination Areas and Airborne Radioactivity Areas. The RWP informs kers of area radiological conditions and entry requirements and provides a mechanism to te worker exposure to specific work activities. The procedures controlling RWPs are sistent with the guidance provided in Regulatory Guide 8.10 (USNRC, 2016).

.2.1.6 Commitment to Radiation Protection Training design and implementation of the radiation protection training program complies with the uirements of 10 CFR 19.12. Records are maintained in accordance with 10 CFR 20, part L.

development and implementation of the radiation protection training program is consistent the guidance provided in the following regulatory guidance documents:

• Regulatory Guide 8.10 - Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable (USNRC, 2016)

• Regulatory Guide 8.13 - Instructions Concerning Prenatal Radiation Exposure (USNRC, 1999)

• Regulatory Guide 8.29 - Instructions Concerning Risks from Occupational Radiation (USNRC, 1996)

• ASTM E1168 Radiological Protection Training for Nuclear Facility Workers (ASTM, 2013).

viduals who require unescorted access into restricted areas (as defined in section 11.1.5.1.1) receive training that is commensurate with the radiological hazard to ch they may be exposed. Non-facility visitors and fire or emergency responders requiring ess to restricted areas are provided with trained escorts who have received radiation ection training.

level of radiation protection training provided is based on the potential radiological health s associated with an employee's work responsibilities and incorporates the provisions of CFR 19.12. In accordance with 10 CFR 19.12, any individual working at the facility who is y to receive in a year a dose in excess of 100 mrem (1 millisievert [mSv]) is:

• Kept informed of the storage, transfer, or use of radioactive material.

• Instructed in the health protection problems associated with exposure to radiation and radioactive material, in precautions or procedures to minimize exposure, and in the purposes and functions of protective devices employed.

• Provided with access to and training on the use of personal protective equipment (PPE).

• Required to observe, to the extent within the worker's control, the applicable provisions of the NRC regulations and licenses for the protection of personnel from exposure to radiation and radioactive material.

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exposure to radiation and radioactive material.

• Instructed in the appropriate response to warnings made in the event of any unusual occurrence or malfunction that may involve exposure to radiation and radioactive material.

• Advised of the various notifications and reports to individuals that a worker may request in accordance with 10 CFR 19.13.

rkers who perform or supervise the shipment of radioactive materials are trained and qualified ccordance with 49 CFR 172, Subpart H, in accordance with 10 CFR 71.5.

radiation protection training program takes into consideration a worker's normally assigned k activities. Abnormal situations involving exposure to radiation and radioactive material, that reasonably be expected to occur during the life of the facility, are also evaluated and factored the training. The extent of these instructions is commensurate with the potential radiological lth protection problems present in the workplace.

raining of personnel previously trained is performed for radiological, chemical, industrial, and cality safety at least annually. The retraining program also includes procedure changes and ating and changes in required skills. Changes to training are implemented, when required, to incidents potentially compromising safety or if changes are made to the facility or cesses.

ords of training are maintained in accordance with the SHINE records management system.

ility training programs are established in accordance with Subsection 12.1.4. The radiation ection sections of the training program are evaluated at least annually. The program content viewed to ensure it remains current and adequate to ensure worker safety.

.2.1.7 Radiation Safety Audits iation safety audits are conducted, at a minimum, on an annual basis for the purpose of ewing all functional elements of the radiation protection program to meet the requirement of CFR 20.1101(c). The audit activity is led by a member of the Review and Audit Committee, or er designated independent individual, with the knowledge and experience to perform the vity. The audits provide sufficient information to assess:

• Compliance with NRC regulations

• Compliance with the terms and conditions of the license

• Occupational doses and doses to members of the public for ALARA compliance

• Maintenance of radiation protection program required records iciencies identified during the audit are addressed through the corrective action program. The ults of the radiation safety audits are provided to the Radiation Safety Committee, the COO the CEO for review. Section 12.2 provides additional details of audit activities.

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iation protection records are used for developing trend analysis, for keeping staff and nagement informed regarding radiation protection matters, and for reporting to regulatory ncies. In addition, the records are used to formulate action based on data obtained (such as vey or sample results), including historical trends.

ccordance with 10 CFR 20, Subpart L, the following records are retained until termination of facility operating license:

• Records documenting provisions of the radiation protection program

[10 CFR 20.2102(b)].

• Results of surveys to determine individual dose from external sources

[10 CFR 20.2103(b)(1)].

• Results of measurements and calculations used to determine individual intakes of radioactive material used in the assessment of internal dose [10 CFR 20.2103(b)(2)].

• Results of air sampling, surveys and bioassays required pursuant to 10 CFR 20.1703(c)(1) and (2) for the Respiratory Protection Program

[10 CFR 20.2103(b)(3)].

• Results of measurements and calculations used to evaluate release of radioactive effluents to the environment [10 CFR 20.2103(b)(4)].

• NRC Form 4, Cumulative Occupational Dose History [10 CFR 20.2104(f)].

• Planned Special Exposure documentation [10 CFR 20.2105(b)].

• Dose received by all individuals for whom monitoring was required pursuant to 10 CFR 20.1502, and records of doses received during planned special exposures, accidents and emergency conditions. Dose to an embryo/fetus are maintained with record of dose to the declared pregnant woman. [10 CFR 20.2106(a) through (f)].

• Declaration of pregnancy [10 CFR 20.2106(e)].

• Compliance with dose limit for individual members of the public [10 CFR 20.2107(b)].

• Disposal of licensed materials and disposal by burial in soil [10 CFR 20.2108(b)].

ccordance with 10 CFR 20, Subpart L, the following records are retained for three years:

• Records of audits and reviews of the radiation protection program [10 CFR 20.2102(b)].

• Records of surveys and calibrations required by 10 CFR 20.1501, Surveys and Monitoring, and 20.1906(b), Receiving and Opening Packages [10 CFR 20.2103(a)].

• Records used in preparing NRC Form 4 [10 CFR 20.2104(f)].

ccordance with 10 CFR 20.2110, records will be legible throughout the retention period. The ord may be an original, or reproduced copy or microform provided it is authenticated by horized personnel and the microform is capable of producing a clear copy throughout the ntion period. Records may be stored in electronic media with the capability for producing ble, accurate and complete records during the required retention period. Records, such as rs, drawings and specifications, include all pertinent information, such as stamps, initials and atures.

.2.1.9 Technical Specifications vities related to the administration and audit of the radiation protection are contained in the lity technical specifications.

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section 11.1.2.1 states the facility's commitment to the implementation of an ALARA gram. The objective of the program is to make every reasonable effort to maintain exposure adiation as far below the dose limits of 10 CFR 20.1201 and 10 CFR 20.1301 as is practical.

design and implementation of the ALARA program is consistent with the guidance provided egulatory Guides 8.2 (USNRC, 2011), 8.13 (USNRC, 1999), and 8.29 (USNRC, 1996). The ration of the facility is consistent with the guidance provided in Regulatory Guide 8.10 NRC, 2016).

ual doses to individual personnel are maintained ALARA. In addition, the annual collective e to personnel (i.e., the sum of annual individual doses, expressed in person-sievert [Sv] or son-rem) is maintained ALARA. The dose equivalent to an embryo/fetus of a declared gnant worker is maintained at or below the limit in 10 CFR 20.1208.

radiation protection program is written and implemented to ensure that it is comprehensive effective. The written program documents policies that are implemented to ensure the RA goal is met. Procedures are written so that they incorporate the ALARA philosophy into routine operations and ensure that exposures are consistent with administrative dose limits.

discussed in Subsection 11.1.5, radiological zones/areas are established within the facility.

establishment of these zones supports the ALARA commitment by minimizing the spread of tamination and reducing exposure of personnel to radiation.

cific goals of the ALARA program include maintaining occupational exposures and ironmental releases as far below regulatory limits as is reasonably achievable. The ALARA cept is also incorporated into the design of the facility. The plant is divided into radiation es with radiation levels that are consistent with the access requirements for those areas.

as where on-site personnel spend significant amounts of time are designed to maintain the est dose rates reasonably achievable.

Radiation Protection Manager is responsible for implementing the ALARA program and uring that adequate resources are committed to make the program effective. The Radiation tection Manager prepares an annual ALARA program evaluation report. The report reviews radiological exposure and effluent release data for trends, including ALARA dose goals, results of audits and inspections, (3) use, maintenance, and surveillance of equipment used exposure and effluent control, and (4) other issues that may influence the effectiveness of the ation protection/ALARA programs. The effectiveness of the ALARA program is reviewed by RSC. The RSC sets the ALARA goals for the facility and reviews new activities to ensure RA principles are considered. Efforts for improving the effectiveness of equipment used for ent and exposure control are also evaluated by the RSC. Any resulting recommendations the committee reviews and evaluations are documented in RSC meeting minutes. The mittee's recommendations are dispositioned in the facilitys corrective action process.

.3.1 ALARA Program Considerations SHINE facility is designed to maximize the incorporation of good engineering practices and ons learned to accomplish ALARA objectives.

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RA principles were applied during the design of the SHINE facility, consistent with the ommendations in Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational iation Exposures at Nuclear Power Stations Will be As Low As Is Reasonably Achievable NRC, 1978).

ign considerations for maintaining personnel external doses ALARA include the following:

• Materials of construction

• Radioactive material processing, storage, and disposal facilities

• Radiation monitoring systems

• Facility layout for personnel traffic and equipment maintainability and accessibility

• Systems and devices to control access to High Radiation Areas and Very High Radiation Areas

• Utilizing the ALARA concepts of time, distance and shielding. For example:

- Design work stations to minimize time operators need to be in radiation fields to perform work

- Locate equipment that require access at a maximum distance from radiation sources or provide remote equipment operation, where practicable

- Incorporate shielding, where appropriate, to achieve the design condition of 0.25 mrem/hr at 12 inches [30 cm] from the shielding surface ign considerations for preventing personnel contamination and minimizing the spread of tamination within the facility include the following:

• Ventilation and filter systems

• Confinement to keep contamination ALARA

• Enclosures to prevent the spread of contamination

• Materials of construction to facilitate decontamination

• Facility layout, with emphasis on personnel and material movement patterns following design considerations are used to control radioactive effluent releases:

• Control of airborne effluents by incorporating confinement, radioactive gaseous waste system disposal capabilities, and exhaust system features

• Control of liquid effluents to ensure radioactive materials in excess of the limits are not released

• Use of radioactivity monitoring systems to monitor radioactive effluents original facility design concepts to maintain exposures ALARA are presented in section 11.1.3.2.

.3.1.2 Operation Policies activities conducted by management personnel who have plant operational responsibility for ation protection are addressed in Subsection 11.1.2. These activities are consistent with the ommendations of Regulatory Guide 8.10 (USNRC, 2016).

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ility design considerations for maintaining personnel exposures ALARA are presented in the wing paragraphs. The basic management philosophy guiding the SHINE facility design to ntain radiation exposures ALARA includes:

• Designing structures, systems and components such that radioactive material, to the greatest extent practical, is remotely handled and isolated from on-site personnel by shielded compartments and hot cells.

• Designing structures, systems and components for reliability and maintainability, thereby reducing the maintenance requirements on radioactive components.

• Designing structures, systems and components to reduce the radiation fields and control streaming, thereby reducing radiation exposure during operation, maintenance, and inspection activities.

• Designing structures, systems and components to reduce access, repair and removal times, thereby reducing the time spent in radiation fields during operation, maintenance, and inspection.

• Designing structures, systems and components to accommodate remote and semi-remote operation, maintenance and inspection, thereby reducing the time spent in radiation fields.

.3.2.1 General Design Considerations for ALARA Exposures eral design considerations and methods to maintain in-plant radiation exposures ALARA sistent with the recommendations of Regulatory Guide 8.8 (USNRC, 1978) have two ctives:

• Minimizing the necessity for access to and personnel time spent in radiation areas.

• Minimizing radiation levels in routinely occupied plant areas in the vicinity of plant equipment expected to require personnel attention.

following operations are considered during the equipment and facility design to maintain osures ALARA:

• Normal operation.

• Maintenance and repairs.

• In-service inspection and calibrations.

• Other anticipated operational occurrences.

• Decommissioning.

mples of features that assist in maintaining exposures ALARA include:

• Design provisions for maintenance of the PCLS and light water pool chemistry conditions, such that corrosion and resulting activation product source terms are minimized.

• Features to allow draining, flushing, and decontamination of equipment and piping.

• Shielding for personnel protection during maintenance or repairs and during decommissioning.

• Means and adequate space for the use of movable shielding.

• Separation of more highly radioactive equipment from less radioactive equipment and separate shielded compartments for adjacent items of radioactive equipment.

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• Means and adequate space for the use of remote operations, maintenance, and inspection equipment.

• Separating clean areas from potentially contaminated ones.

.3.2.2 Equipment Design Considerations for ALARA Exposures ipment design considerations to minimize the necessity for, and amount of, time spent in a ation area include:

• Reliability, availability, maintainability, inspectability, constructability, and other design features of equipment, components, and materials to reduce or eliminate the need for repair or preventive maintenance.

• Design features to facilitate ease of maintenance or repair, including ease of disassembly and modularization of components for replacement or removal to a lower radiation area for repair or disposal.

• Capabilities to remotely or mechanically operate, repair, service, monitor, or inspect equipment.

• Consideration of redundancy of equipment or components to reduce the need for immediate repair when radiation levels may be high and when there is no feasible method available to reduce radiation levels.

• Capabilities for equipment to be operated from accessible areas both during normal and abnormal operating conditions.

ipment design considerations directed toward minimizing radiation levels near equipment or ponents requiring personnel access include:

• Selection of materials that minimize the creation of radioactive contamination.

• Equipment and piping designs that minimize the accumulation of radioactive materials (e.g., the use of buttwelding fittings and minimizing the number of fittings reduces radiation accumulation at the seams and welds).

• Provisions for draining, flushing, or, if necessary, remote cleaning or decontamination of equipment containing radioactive materials.

• Design to limit leaks or control the fluid that does leak. This includes the use of hermetically sealed valves and directing leakage via drip pans and piping.

• Provisions for isolating equipment from radioactive process fluids.

.3.2.3 Facility Layout Design Considerations for ALARA Exposures ility layout design considerations to minimize the amount of personnel time spent in a ation area include the following:

• Locating equipment, instruments, and sampling stations that require routine maintenance, calibration, operation, or inspection, to promote ease of access and minimize occupancy time in radiation areas.

• Laying out plant areas to allow remote or mechanical operation, service, monitoring, or inspection of contaminated equipment.

• Providing, where practicable, for movement of equipment or components requiring service to a lower radiation area.

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• Separating radiation sources and occupied areas, where practicable.

• Redundant components requiring periodic maintenance that are a source of radiation are located in separate compartments, where practicable, to allow maintenance of one component while the other component is in operation.

• Highly radioactive passive components with minimal maintenance requirements are located in shielded enclosures and are provided with access via shielded openings or removable blocks.

• Providing means and adequate space for using movable shielding when required.

• Designing of the plant layout so that access to a given radiation zone does not require passing through a higher radiation zone.

• Locating equipment, instruments, and sampling sites in the lowest practicable radiation zone.

• Providing control panels to permit remote operation of essential instrumentation and controls from the lowest radiation zone practicable.

• Providing means to control contamination by maintaining ventilation air flow patterns from areas of lower radioactivity to areas of higher radioactivity.

• Providing means to facilitate decontamination of potentially contaminated areas.

.4 RADIATION MONITORING AND SURVEYING

.4.1 Radiation Monitoring nventory of calibrated radiation detection and measurement instruments is maintained to orm functions such as radiation surveys, contamination surveys, package surveys, sealed rce leak tests, air sampling measurements, effluent release measurements, and dose rate asurements. Radiation monitoring equipment, their function and location is shown in le 11.1-12 and is discussed below.

a. Personnel Monitors Personnel who enter radiologically restricted areas (as defined in Subsection 11.1.5.1) are required to wear personnel monitoring devices. In addition, personnel are required to monitor themselves prior to exiting restricted areas which may have the potential for contamination.

b. Continuous Air Monitors Continuous air monitors (CAMs) provide indication of the airborne activity levels in the restricted areas of the facility. Alarms are used to provide early warning of unanticipated increases in airborne radioactivity levels. Procedures provide detailed instructions for using and determining CAM alarm setpoints. When deemed necessary, portable air samplers may be used to collect a sample on filter paper for subsequent analysis in the laboratory.

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Tritium is monitored at specific locations where airborne tritium may be present and present a potential hazard to individuals. Tritium monitoring is accomplished using fixed continuous instruments for room air sampling and ventilation duct sampling.

d. Gaseous Effluent Monitoring The stack release monitor (SRM) on the facility effluent stack and the carbon delay bed effluent monitor (CDBEM) must be capable of:

• Continuous monitoring of radioactive stack releases for noble gases.

• Generating real time data for control room display and recording.

• Allowing periodic collection of filters to allow for laboratory analysis for particulate and iodine.

The SRM provides continuous on-line sampling of releases of gaseous effluents from the facility to demonstrate that releases are within the regulatory limits. The CDBEM is provided to monitor the safety-related alternate release path.

e. Detection and Monitoring of Radioactivity in Liquid Systems and Liquid Effluents There are no piped radioactive liquid effluent discharges from the facility; therefore, there are no installed liquid effluent monitors. However, infrequent liquid effluent releases are collected, sampled, and verified to meet the criteria for release provided in 10 CFR 20.2003 and 10 CFR 20.2007 prior to discharge to the sanitary sewer.

Closed loop process cooling water systems are monitored (through sampling or installed instrumentation) to detect leakage between process fluids and cooling water due to failure in a heat exchanger or other system boundary component.

f. Radiation Area Monitors Radiation area monitors (RAMs) provide radiation monitoring and alarms to alert personnel and the control room of radiation levels that are in excess of normal background levels. RAMs are located in areas to monitor the environment for radioactivity during normal operations, operational occurrences and postulated accidents. Procedures provide detailed instructions for determining and employing alarm set points for RAMs.

RAMs may be provided in High Radiation Areas in order to provide a remote readout. If a RAM is not provided in a particular High Radiation Area, then portable instruments are required by the RWP to measure dose rates when personnel access the area.

g. Control Point Monitoring Monitor stations are located at the access points for restricted areas. Monitors are provided to detect radioactive contamination of personnel. Monitoring station locations NE Medical Technologies 11.1-19 Rev. 3

Monitoring equipment used at the facility access points are shown in Table 11.1-12.

h. Criticality Monitoring Criticality monitoring in the main production facility is provided by the criticality accident alarm system (CAAS). This system is described in Subsection 6b.3.3.

iation monitoring systems, their functions, and their interfaces with the engineered safety ures in the facility are described in Section 7.7.

.4.1.1 Calibration and Maintenance of Radiation Monitoring Equipment cedures are prepared for each of the radiation monitoring instruments used and specify the uency and method of calibration. Radiation monitoring equipment is calibrated before being into use and after any maintenance or repair that may affect instrument performance.

bration of portable radiological monitoring equipment used to document radiological survey ults is performed in accordance with ANSI N323AB-2013, American National Standard for iation Protection Instrumentation Test and Calibration, Portable Survey Instruments SI/ANS, 2014).

iation monitoring equipment is calibrated in accordance with manufacturer ommendations.

ntenance and repair of radiation protection instrumentation is performed in accordance with roved procedures and instrument manufacturer recommendations.

.4.1.2 Operational Tests of Radiation Monitoring Equipment ration and response tests of radiation monitoring, counting, and air sampling instruments are ormed by personnel trained in the use of the instrument and following approved procedures.

se tests are consistent with the manufacturers recommendations and applicable regulatory uirements. Operation and response tests are conducted at a frequency consistent with stry practices and is addressed in detailed instructions.

.4.2 Radiation Surveys iation surveys are conducted for two purposes: (1) to ascertain radiation levels, centrations of radioactive materials, and potential radiological hazards that could be present e facility; and (2) to detect releases of radioactive material from facility equipment and rations.

assure compliance with the requirements of 10 CFR 20, Subpart C, there are written cedures for the radiation survey and monitoring programs. The radiation survey and nitoring programs assure compliance with the requirements of 10 CFR 20, Subpart F, part C, Subpart L, and Subpart M.

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• Regulatory Guide 8.2, Guide for Administrative Practice in Radiation Monitoring (USNRC, 2011)

• Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure Data (USNRC, 2018)

• Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program (USNRC, 1993)

• Regulatory Guide 8.24, Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication (USNRC, 2012) (applicable to target solution preparation processes)

• Regulatory Guide 8.34, Monitoring Criteria and Methods to Calculate Occupational Radiation Doses (USNRC, 1992)

• ANSI N323AB-2013, American National Standard for Radiation Protection Instrumentation Test and Calibration, Portable Survey Instruments (ANSI/ANS, 2014) cedures include sampling protocol and data analysis methods. Equipment selection is based he type of radiation being monitored.

vey procedures also specify the frequency of measurements and record keeping and orting requirements. Survey records include:

• Radiation dose rate survey results

• Surface contamination survey results

• Airborne radioactivity survey results

.4.3 Technical Specifications tain material in this section provides information that is used in the technical specifications.

includes limiting conditions for operation, setpoints, design features, and means for omplishing surveillances. In addition, significant material is also applicable to, and may be renced by, the bases that are described in the technical specifications.

.5 RADIATION EXPOSURE CONTROL AND DOSIMETRY

.5.1 Controlled Area sistent with 10 CFR 20.1003, the controlled area is the area outside of a restricted area but de the site boundary, access to which can be limited by SHINE for any reason. At the SHINE lity, this is the area within the site boundary, and is referred to as the owner controlled area, escribed in Subsection 2.1.1.2.

ility visitors include delivery people, tour guests, and service personnel who are transient upants of the controlled area. Area monitoring demonstrates compliance with public dose ts for such visitors.

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iological zones with varied definitions and span of control have been designated for the lity site and areas surrounding the facility site. The purpose of these zones is to (1) control the ead of contamination, (2) control personnel access to avoid unnecessary exposure of sonnel to radiation, and (3) control access to radioactive sources present in the facility. Public ess to radiological areas is restricted as detailed in this section and as directed by facility nagement. Areas where personnel spend substantial amounts of time are designed to imize the exposure received when routine tasks are performed, in accordance with the RA principle.

following definitions are provided to describe how the radiation protection program is lemented to protect workers and the general public on the site:

a. Unrestricted Area NRC regulation 10 CFR 20.1003 defines an unrestricted area as an area for which access is neither limited nor controlled by the licensee. The area adjacent to the facility site is an unrestricted area. This area can be accessed by members of the public or by facility personnel. The unrestricted area is governed by the limits in 10 CFR 20.1301. The TEDE to individual members of the public from the licensed operation may not exceed 1 mSv (100 mrem) in a year (exclusive of background radiation). The dose in any unrestricted area from external sources may not exceed 0.02 mSv (2 mrem) in any one hour.

b. Restricted Area 10 CFR 20.1003 defines a restricted area as an area where access is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Access to and egress from a restricted area at the facility site is through a radiation protection control point. Monitoring equipment is located at these control points.

Most restricted areas are located within the physical structure of the main production facility and locations in the material staging building where radioactive material is normally stored. Radioactive material may be temporarily stored in outdoor areas during transfer between areas. These temporary areas may require that a restricted area be established with the controls described in this section.

c. Radiologically Controlled Area The RCA is a restricted area. The RCA is an area within the restricted area posted for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Only individuals who have successfully completed training in radiation protection procedures are permitted to access this area without escort by trained personnel.

itional radiological areas may exist within the restricted area. The areas may be temporary or manent. The areas are posted to inform workers of the potential hazard in the area and to NE Medical Technologies 11.1-22 Rev. 3

iation areas and expected dose rates are shown in Table 11.1-4.

.5.2 Access and Egress Control NE establishes and implements an access control program that ensures that (a) signs, ls, and other access controls are properly posted and operative, (b) restricted areas are blished to prevent the spread of contamination and are identified with appropriate signs, and step-off pads, change facilities, protective clothing facilities, and personnel monitoring ruments are provided in sufficient quantities and locations, as necessary.

sonnel access to high radiation areas is controlled to prevent unplanned radiation exposures.

sonnel access is controlled through administrative methods, including procedures and RWPs.

ve and passive safety features are provided to control access to high radiation areas in ordance with 10 CFR 20.1601. These safety features include:

• Neutron driver service cell personnel access door interlocks de-energize the accelerator to reduce the level of radiation upon personnel entry (defense-in-depth design attribute),

and accelerator key switches prevent activation of the accelerator while personnel are present.

• Hot cells requiring periodic/routine entry where there is potential for excessive personnel exposures are equipped with door interlocks to prevent the hot cell door from being opened when the evaluated hazard exists (e.g., excessive radiation field, target solution transfer occurring in cell).

• The neutron driver service cell and hot cells are equipped with audible and visual warnings so that an individual attempting to enter the High Radiation Area and the supervisor of the activity are made aware of the entry or are controlled by locked entry with positive access controls over each individual entry, consistent with 10 CFR 20.1601(a).

• High radiation areas are radiologically shielded and isolated from access to individuals by the use of engineered physical barriers. These include structural shield blocks and/or locked shield doors, consistent with 10 CFR 20.1601(a)(3).

ess to and egress from the restricted area is through one of the monitor stations at the ricted area boundary. Access to and egress from each Radiation Area, High Radiation Area, taminated Area or Airborne Radioactivity Area within the restricted area may also be vidually controlled. A monitor (frisker), step-off pad, and container for any discarded ective clothing may be provided at the egress point from certain of these areas to prevent the ead of contamination.

.5.3 Posting for Radiation Protection Awareness iological postings are clearly identified by physical means such as placarding or boundary king in accordance with 10 CFR 20.1902.

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sonnel working in areas that are classified as airborne radioactivity areas or contaminated as must wear appropriate PPE. If the areas containing the surface contamination can be ated from adjacent work areas via a barrier such that dispersible material is not likely to be sferred beyond the area of contamination, personnel working in the adjacent area are not uired to wear PPE. Areas requiring PPE are posted at each of their entry points. The radiation ker training program provides instruction to personnel on the proper use of PPE.

iation protection management and associated technical staff are responsible for determining need for PPE in each work area and documenting the PPE requirements on the applicable P. For areas with removable contamination from beta/gamma emitters or uranium above 00 disintegrations per minute per 100 square centimeters (dpm/100 cm2) or from alpha tters other than uranium above 20 dpm/cm2 PPE is required. PPE includes coveralls, gloves, e covers, and rubber boots. Guidance for selecting and using PPE is provided in the facility ation protection program.

respiratory protection program is described in Section 11.3.

.5.5 Personnel Monitoring for External Exposures ernal exposures are received primarily from the fission products produced in the target tion. Other potential sources of exposure include neutrons (e.g., from operational neutron ers), activation products, and tritium gas. The nuclides of radiological significance are tified above in Section 11.1.

sonnel whose duties require them to enter restricted areas wear individual external dosimetry ices that are sensitive to beta, gamma and neutron radiation. Personnel handling licensed rces and working around radioactive materials outside restricted areas (e.g.,

sportation-related surveys) wear individual external dosimetry devices that are sensitive to a, gamma and neutron radiation. Any individual entering a High Radiation Area or Very High iation Area wears personal dosimetry, and supplemental dosimetry with dose and dose rate m capability.

sonal dosimetry shall be worn in a manner consistent with the manufacturers directions.

ernal dosimetry devices are evaluated at least quarterly, or soon after participation in high-e evolutions, to ascertain external exposures. Administrative limits on radiation exposure are d in Table 11.1-11. The administrative limits are reflective of ALARA principles.

stigation levels are set at 25 percent of the annual administrative limit for any workers upational dose received during a calendar quarter. An investigation is performed and umented to determine what types of activities may have contributed to the worker's external osure. The investigation may include, but is not limited to, procedural reviews, efficiency ies of the ventilation system, uranium storage protocol, and work practices.

time an administrative limit is exceeded, the Radiation Protection Manager is informed. The iation Protection Manager is responsible for determining the need for and recommending stigations or corrective actions to the responsible manager(s). Copies of the Radiation tection Manager's recommendations are provided to the RSC.

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.5.6 Determination of Internal Exposures purposes of assessing dose used to determine compliance with occupational dose ivalent limits, SHINE shall, when required under 10 CFR 20.1502, take suitable and timely asurements of one of the following:

1. Concentrations of radioactive materials in air in work areas.

2. Quantities of radionuclides in the body.

3. Quantities of radionuclides excreted from the body.

4. Combinations of these measurements.

ess respiratory protective equipment is used, as provided in 10 CFR 20.1703, or the essment of intake is based on bioassays, SHINE shall assume that an individual inhales oactive material at the airborne concentration in which the individual is present.

radiation protection program includes detailed methodology for determination of internal osures.

.5.7 Evaluation and Record of Doses vidual worker occupational dose is assessed on a quarterly basis and is performed more uently when reasonable suspicion exists regarding an abnormal exposure. External imetry devices are processed and evaluated by a provider accredited by the National untary Laboratory Accreditation Program.

• Procedures for the evaluation and summation of doses are based on guidance contained in Regulatory Guides 8.7 (USNRC, 2018) and 8.34 (USNRC, 1992).

ords are maintained of doses received by all individuals for whom monitoring is required er 10 CFR 20.1502, in accordance with 10 CFR 20.2106. The records include the following, pplicable:

• The deep-dose equivalent to the whole body, lens dose equivalent, shallow-dose equivalent to the skin and shallow-dose equivalent to the extremities;

• The estimated intake of radionuclides;

• The committed effective dose equivalent (CEDE) assigned to the intake of radionuclides;

• The specific information used to calculate the CEDE under 10 CFR 20.1204(a) and (c),

when required by 10 CFR 20.1502;

• The TEDE, when required by 10 CFR 20.1202; and

• The total of the deep-dose equivalent and the committed dose to the organ receiving the highest total dose.

also Subsection 11.1.2.1.8 for retained individual dose evaluation records.

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NE may authorize an adult worker to receive (non-emergency) doses, in addition to and ounted for separately, from the doses received under the limits specified in CFR 20.1206(e), provided that each of the requirements of 10 CFR 20.1206(a), (b), (c),

(d) are met.

NE maintains records of the conduct of a planned special exposure and submit a written ort as required by 10 CFR 20.1206(f). In accordance with 10 CFR 20.2105, the record of a ned special exposure includes:

• The exceptional circumstances requiring the use of a planned special exposure.

• The name of the management official who authorized the planned special exposure and a copy of the signed authorization.

• What actions were necessary.

• Why the actions were necessary.

• How doses were maintained ALARA.

• What individual and collective doses were expected to result, and the doses actually received in the planned special exposure.

.6 CONTAMINATION CONTROL EQUIPMENT AND FACILITY LAYOUT GENERAL DESIGN CONSIDERATIONS FOR 10 CFR 20.1406 tamination control is part of the radiation protection program described in Subsection 11.1.2.

sonnel receiving Radiation Worker Training are instructed on the sources, detection and trol of radioactive contamination. Procedures provide instruction for identifying and controlling tamination. Records of contamination events are entered into the corrective action process, ewed by the RSC, and maintained as records, as applicable, in accordance with the radiation ection program requirements described in Subsection 11.1.2.

eral equipment and facility layout design considerations to prevent the spread of tamination in the facility and to the environment and to facilitate eventual decommissioning in ordance with 10 CFR 20.1406 include the features discussed in the following subsections.

.6.1 Shielded Compartments and Hot Cells cess equipment containing significant radioactive material is located within shielded partments or hot cells.

cess equipment which does not require local operator interaction during production, such as neutron driver assembly and the subcritical assembly, is located in shielded compartments cess is provided via shielded openings as required). Where operator intervention is required ng processing activities, for example molybdenum extraction and purification, the equipment cated in shielded hot cells and the operator is provided with a means for remote viewing and nipulation of components.

se shielded compartments and shielded hot cells are provided to facilitate confinement, ation, and collection of potential liquid spills to minimize the spread of contamination to the lity and the environment. With the exception of the below grade confinement, these shielded NE Medical Technologies 11.1-26 Rev. 3

.6.2 Piping ere shielding is required, radioactive piping is located inside shielded compartments or hot

s. For transfers between hot cells the piping is located in shielded pipe trenches which vide for liquid and airborne confinement and detection of leakage. Inspection ports are vided to allow for visual inspection of piping. Use of embedded piping is minimized to facilitate ection and detect leakage.

.6.3 Light Water Pool light water pool which provides shielding and cooling for the subcritical assembly system AS) is designed with leak detection to prevent unidentified leakage to the facility and the ironment.

.6.4 Process Tanks cess tanks are seismically supported and are located in seismically designed concrete vaults are designed to prevent unidentified leakage to the facility and the environment.

.6.5 Monitoring and Controlled Entry and Egress to Restricted Area ess to and egress from these areas is strictly controlled via administrative procedures and sive confinement structure design.

sonnel access and egress is controlled by Radiation Protection personnel, equipment and cedures. Prior to entry, personnel must don appropriate PPE to minimize the potential for sical contamination of the worker and the subsequent spread of contamination beyond the ricted area. This PPE is either removed and disposed of or monitored for contamination prior elease from the restricted area. Personnel must then pass through appropriate portal nitoring equipment prior to egress from the restricted area.

entially contaminated materials removed from the restricted area (for example, production erial, tools, disposed equipment, various process and maintenance consumables) are veyed and released, when appropriate, following radiation protection program implementing cedures. Disposal of contaminated materials is performed in accordance with radioactive te management program implementing procedures (see Section 11.2).

tricted areas in the main production facility are provided with fixed CAMs to detect the ential spread of airborne contamination within the restricted area. Additionally, RAMs are in e to detect potential increases in background radiation levels.

iation protection personnel routinely perform radiation and contamination assessments of essible areas within restricted areas. Special surveys are performed, prior to entry, if access quired to normally unoccupied areas.

NE Medical Technologies 11.1-27 Rev. 3

.7.1 Environmental Monitoring Program NE maintains a radiological environmental monitoring program (REMP) as required by CFR 20.1302. The REMP is used to verify the effectiveness of facility measures which are d to control the release of radioactive material and to verify that measurable concentrations of oactive materials and levels of radiation are not higher than expected based on effluent asurements and modeling of the environmental exposure pathways.

dance provided in Regulatory Guide 4.1, Radiological Environmental Monitoring for Nuclear er Plants (USNRC, 2009) and Table 3.12-1 of NUREG-1301, Offsite Dose Calculation nual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors NRC, 1991), was considered when developing the REMP for the SHINE facility. In addition, REMP was developed using the data quality objectives (DQO) process which is a scientific tematic planning method. The DQOs were developed according to the U.S. Environmental tection Agency (EPA) Guidance on Systematic Planning Using the Data Quality Objectives cess (EPA, 2006).

ironmental monitoring is conducted at potential receptor locations. Details of the REMP are sented in the following sections.

.7.2 Effluent Release Pathways orne effluents from the facility include noble gases, iodine and other halogens, particulates, tritium. The following pathways represent plausible public exposure scenarios from airborne ents:

• Direct radiation exposure pathway monitored using dosimeters.

• Inhalation pathway monitored using continuous air samples.

• Ingestion exposure pathway.

re are no routine radioactive liquid effluent discharges from the RCA. Radioactive liquid harges from the SHINE facility to the sanitary sewer are infrequent and made in accordance 10 CFR 20.2003 and 10 CFR 20.2007. There are no piped liquid effluent pathways from the A to the sanitary sewer. Sampling is used to determine suitability for release. See tion 11.2 for additional information on liquid discharges from the RCA.

.7.2.1 Direct Radiation Monitoring ct exposure to gamma and beta emitting radionuclides released through the stack of the n production facility is monitored and measured at receptor locations using environmental imeters. The dosimeters measure direct radiation from radiation sources contained within the NE main production facility, from sources within the material staging building, from oactivity in the airborne effluent, and from deposition of airborne radioactivity onto the und.

escription of dosimeter locations and the rationale for locations are provided in Table 11.1-14.

imeter locations are shown on Figure 11.1-4. Table 3.12-1 of NUREG-1301 (USNRC, 1991) ommends 40 dosimeter locations (i.e., an inner ring and an outer ring of dosimeters with one NE Medical Technologies 11.1-28 Rev. 3

gnificant distance from the facility such that it represents a background dose. Considering the of the SHINE facility and the low power level of the SHINE subcritical IUs, 24 dosimeter tions are specified. These dosimeters are located in order to provide annual direct dose rmation at on-site locations which are expected to have occupancy and at property line tions which ensure all directions are monitored. The property line locations include the ction of the theoretical MEI and the direction of the nearest occupied structure. At least one tion includes a paired dosimeter so that data quality can be determined. Three of the imeters are stationed off site at special interest areas and one dosimeter is located a ificant distance from the facility to represent background dose.

imeter values are calculated using the reports from the laboratory providing results.

kground radiation is subtracted from the dosimeter results. The background radiation values those established during the baseline environmental survey which obtained baseline imeter readings at each dosimeter location.

.7.2.2 Iodine and Particulate Monitoring for Releases via Airborne Pathway orne effluent releases from the SHINE facility contribute to off-site doses. Air monitoring ects iodine or particulate releases from the SHINE facility. Noble gas and tritium asurements are not included in the REMP. Noble gas and tritium measurements are ormed by the radiation protection program.

ironmental airborne sampling is performed to identify and quantify particulates and oiodine in airborne effluents. Regulatory Position C.3.b of Regulatory Guide 4.1 (USNRC,

9) indicates that airborne sampling should always be included in the environmental nitoring programs for nuclear power plants since the airborne effluent pathway exists at all

s. Since the SHINE facility includes airborne effluent releases and radioactivity in the airborne ent can result in measurable off-site doses and since there is a potential for a portion of the e to be attributable to radioactive iodine and airborne particulate radioactivity releases, the MP includes airborne sampling.

.7.2.2.1 Air Sampling Locations DQO process and the guidance provided in Table 3.12-1 of NUREG-1301 (USNRC, 1991) e used to establish locations for airborne sample acquisition, sampling frequency, and type of ple analysis. Continuous air sample locations are specified in accordance with guidance vided in Table 3.12-1 of NUREG-1301 (USNRC, 1991). The continuous air sampling is ormed using continuous air samplers (CAS) which include a radioiodine canister for ne-131 (I-131) analysis and a particulate sampler which is analyzed for gross beta oactivity.

r CAS locations (CAS 2 - CAS 5) are near the facility property line in the north, south, east west direction sectors co-located with ED1, ED9, ED5, and ED13 (refer to Figure 11.1-4),

pectively, to ensure all directions are monitored. The north and east direction sectors (with pect to the SHINE facility vent stack) have the highest calculated annual ground level osition factor (D/Q) values (CAS 2 and CAS 4). There is also a control CAS (CAS 1) located fficient distance from the SHINE facility to provide background information for airborne NE Medical Technologies 11.1-29 Rev. 3

osition factor, D/Q. This CAS requirement is combined with the air sample location at the site ndary location in the north direction (refer to Table 11.1-14). A description of air sample tions and the rationale for air sample locations are provided in Table 11.1-14.

air sampling data is used to validate the effluent monitoring and dose compliance data sets.

ults are compared to the radionuclide-specific values provided in 10 CFR 20, Appendix B. A

-of-the fractions approach is used wherein the isotopic values measured are compared with r associated limits in 10 CFR 20, Appendix B. This allows the calculation of dose due to ne and particulate activities and includes both inhalation dose and cloud immersion dose.

kground subtraction is based on results of the baseline environmental survey, thus providing cation-specific and statistically valid means to subtract background.

.7.2.3 Ingestion Pathway (Biota Monitoring)

REG-1301 (USNRC, 1991) suggests sampling of various biological media as a means to rectly assess doses due to particulate and iodine ingestion. This type of monitoring may ude sampling of soils, broad leafed plants, fish, meat, or milk. Nuclear power plants have long nitored this pathway and have seen neither appreciable dose nor upward trending of osition. Since the SHINE source term is expected to be several orders of magnitude lower n that of a nuclear power plant and particulate and iodine radionuclides are not normally ected to be present in measurable quantities within airborne effluent releases from the NE facility, biota monitoring is not routinely included in the REMP. The air sampling and undwater monitoring included in the REMP are sufficient to meet the objectives of the REMP.

.7.2.4 Groundwater Monitoring re are no surface water features on the SHINE site, nor are there any surface waters ediately adjacent to the SHINE site. Surface waters of the rivers in the vicinity of the SHINE (e.g., the Rock River and its tributaries) are not expected to accumulate detectable levels of oactivity. SHINE does not utilize surface water in any production process, and SHINE does discharge to surface water; therefore, surface water is not considered to be a route of osure for ingestion or direct radiation for the SHINE facility. As such, surface water sampling ot included in the REMP. Similarly, marine life in the rivers is not expected to accumulate ectable levels of radioactivity and thus sampling of fish or other marine creatures for the stion pathway is not included in the REMP.

asured local water table elevations for the site identify the groundwater gradient and indicate the groundwater flow is to the west and to the south. The nearest drinking water source is a located approximately a third of a mile (0.54 km) to the northwest of the facility.

re are four test wells within the property boundary for the SHINE facility that were used for nitoring groundwater in support of a hydrological assessment of the site. One test well is ted north, one south, one east, and one west of the SHINE main production facility. Although e are no defined liquid effluent release pathways and the groundwater is not expected to be taminated due to operation of the SHINE facility, the test wells to the west and the south are pled for the presence of radionuclide contaminants. Sampling is in accordance with the ommendations in Table 3.12-1 of NUREG-1301 (USNRC, 1991) (i.e., quarterly with gamma NE Medical Technologies 11.1-30 Rev. 3

.7.3 Community Environmental Monitoring Program ddition to the monitoring that is performed by the REMP to meet regulatory requirements, NE has a Community Environmental Monitoring Program (CEMP). The CEMP includes ntary environmental monitoring based on public or SHINE interests that are not regulatory in ure.

.7.4 Preoperational Baseline Monitoring operational monitoring, beginning approximately two years prior to anticipated licensed vity, serves to provide baseline data for evaluating the impact of operation of the SHINE lity. The collection of samples and analysis of data follow the sampling and analyses edule specified in Subsection 11.1.7.5 and continue into the operational phase of facility ration. The preoperational monitoring is conducted so that the preoperational radiological ditions are understood in sufficient detail to allow future reasonable, direct comparison with a collected after licensed operation of the facility.

.7.5 Sampling and Analysis following frequencies are used; however, alterations may be made based upon data and ds, and the justification of any such alterations are described in the Annual Report. If sample nalysis frequencies are reduced, the changes are not to reduce the overall effectiveness of REMP.

• Air sample filters - monthly, or more frequently if required by dust loading on media

• Environmental dosimeters - quarterly

• Groundwater test wells - quarterly mple analysis employs analytical techniques so that an appropriate analytical sensitivity (e.g.,

iori Lower Level of Detection [LLD]) is achieved. SHINE may also use the analytical detection sitivities as determined based on the Multi-Agency Radiological Laboratory Analytical tocols Manual (MARLAP). Deviations from the a priori analytical sensitivity levels due to rference from other radionuclides or other factors are evaluated and documented. SHINE orts analytical sensitivity capabilities of the REMP in the Annual Report.

ccordance with Regulatory Guide 4.1 (USNRC, 2009), Revision 2, analyses for carbon-14 in ironmental media are not required since the facility produced component is a small fraction of naturally occurring carbon-14.

.7.6 Environmental Monitoring Program Procedures ironmental surveys conducted in support of the REMP are performed in accordance with lity implementing procedures. Document control measures are employed to ensure that nges to the REMP or implementing procedures are reviewed for adequacy, approved by horized personnel and are distributed to and used at the appropriate locations throughout the lity.

NE Medical Technologies 11.1-31 Rev. 3

Annual Report is provided to the NRC in accordance with ANSI/ANS 15.1-2007 (ANSI/ANS, 7). The Annual Report provides summarized results of environmental surveys performed ide the facility.

.7.8 Records, Periodic Review and Corrective Actions ords of off-site environmental surveys are retained in accordance with the SHINE records nagement program for the lifetime of the facility.

annual environmental monitoring program review is conducted to examine the adequacy and ctiveness of the REMP to achieve its objectives. The program review evaluates the need to and (or reduce) the environmental monitoring program given the results of the environmental a and trends in environmental radioactivity. Any reductions shall be thoroughly evaluated and ified, given that environmental data indicating the absence of facility-related radioactivity are ortant. The review confirms exposure pathways and sampling media and validates that the cipal radionuclides being discharged are the same nuclides being analyzed in the ironmental program.

adverse trends or anomalies identified during the conduct of the program, during Annual ort preparation, or during periodic reviews, are entered into the facility corrective action gram for disposition.

NE Medical Technologies 11.1-32 Rev. 3

Table 11.1 Parameters Applicable to Target Solution Radionuclide Inventories Parameter Nominal Values Safety Basis Values wer 125 kW 137.5 kW diation Time 5.5 days 30 days al Time Between Irradiations [ ]PROP [ ]PROP/ECI mber of Cycles [ ]PROP/ECI [ ]PROP/ECI ment Partitioning (Extraction) Nominal None ween Cycles ment Partitioning (Extraction) on Nominal Bounding (noble gases only) al Cycle V Dump Tank Decay Time [ ]PROP/ECI [ ]PROP/ECI percell Extraction Time [ ]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-33 Rev. 3

Table 11.1 Nominal Versus Safety Basis Radionuclide Inventories in Target Solution Actinide Activity (Ci) Fission Product Activity (Ci)

Case At Shutdown Pre-Extraction Post-Extraction At Shutdown Pre-Extraction Post-Extraction minal Values(a) [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI Safety Basis [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI Values(b)

Difference 13 percent 20 percent 28 percent 20 percent 70 percent 170 percent

a. Pre-Extraction: [ ]PROP/ECI post-shutdown; Post-Extraction: [ ]PROP/ECI post-shutdown

b. Pre-Extraction: [ ]PROP/ECI post-shutdown; Post-Extraction: [ ]PROP/ECI post-shutdown NE Medical Technologies 11.1-34 Rev. 3

ble 11.1 Irradiated Target Solution Activity for Select Radionuclides Pre-Extraction (Sheet 1 of 3)

Radionuclide Nominal Activity (Curies) Safety Basis Activity (Curies)

Kr-85 [ ]PROP/ECI [ ]PROP/ECI Kr-85m [ ]PROP/ECI [ ]PROP/ECI Kr-87 [ ]PROP/ECI [ ]PROP/ECI Kr-88 [ ]PROP/ECI [ ]PROP/ECI Rb-86 [ ]PROP/ECI [ ]PROP/ECI Sr-89 [ ]PROP/ECI [ ]PROP/ECI Sr-90 [ ]PROP/ECI [ ]PROP/ECI Sr-91 [ ]PROP/ECI [ ]PROP/ECI Sr-92 [ ]PROP/ECI [ ]PROP/ECI Y-90 [ ]PROP/ECI [ ]PROP/ECI Y-91 [ ]PROP/ECI [ ]PROP/ECI Y-92 [ ]PROP/ECI [ ]PROP/ECI Y-93 [ ]PROP/ECI [ ]PROP/ECI Zr-95 [ ]PROP/ECI [ ]PROP/ECI Zr-97 [ ]PROP/ECI [ ]PROP/ECI Nb-95 [ ]PROP/ECI [ ]PROP/ECI Mo-99 [ ]PROP/ECI [ ]PROP/ECI Tc-99m [ ]PROP/ECI [ ]PROP/ECI Ru-103 [ ]PROP/ECI [ ]PROP/ECI Ru-105 [ ]PROP/ECI [ ]PROP/ECI Ru-106 [ ]PROP/ECI [ ]PROP/ECI Rh-105 [ ]PROP/ECI [ ]PROP/ECI Sb-127 [ ]PROP/ECI [ ]PROP/ECI Sb-129 [ ]PROP/ECI [ ]PROP/ECI Te-127 [ ]PROP/ECI [ ]PROP/ECI Te-127m [ ]PROP/ECI [ ]PROP/ECI Te-129 [ ]PROP/ECI [ ]PROP/ECI Te-129m [ ]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-35 Rev. 3

Radionuclide Nominal Activity (Curies) Safety Basis Activity (Curies)

Te-131m [ ]PROP/ECI [ ]PROP/ECI Te-132 [ ]PROP/ECI [ ]PROP/ECI I-131 [ ]PROP/ECI [ ]PROP/ECI I-132 [ ]PROP/ECI [ ]PROP/ECI I-133 [ ]PROP/ECI [ ]PROP/ECI I-134 [ ]PROP/ECI [ ]PROP/ECI I-135 [ ]PROP/ECI [ ]PROP/ECI Xe-131m [ ]PROP/ECI [ ]PROP/ECI Xe-133 [ ]PROP/ECI [ ]PROP/ECI Xe-133m [ ]PROP/ECI [ ]PROP/ECI Xe-135 [ ]PROP/ECI [ ]PROP/ECI Xe-135m [ ]PROP/ECI [ ]PROP/ECI Xe-138 [ ]PROP/ECI [ ]PROP/ECI Cs-134 [ ]PROP/ECI [ ]PROP/ECI Cs-136 [ ]PROP/ECI [ ]PROP/ECI Cs-137 [ ]PROP/ECI [ ]PROP/ECI Ba-139 [ ]PROP/ECI [ ]PROP/ECI Ba-140 [ ]PROP/ECI [ ]PROP/ECI La-140 [ ]PROP/ECI [ ]PROP/ECI La-141 [ ]PROP/ECI [ ]PROP/ECI La-142 [ ]PROP/ECI [ ]PROP/ECI Ce-141 [ ]PROP/ECI [ ]PROP/ECI Ce-143 [ ]PROP/ECI [ ]PROP/ECI Ce-144 [ ]PROP/ECI [ ]PROP/ECI Pr-143 [ ]PROP/ECI [ ]PROP/ECI Nd-147 [ ]PROP/ECI [ ]PROP/ECI Np-239 [ ]PROP/ECI [ ]PROP/ECI Pu-238 [ ]PROP/ECI [ ]PROP/ECI Pu-239 [ ]PROP/ECI [ ]PROP/ECI Pu-240 [ ]PROP/ECI [ ]PROP/ECI Pu-241 [ ]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-36 Rev. 3

Radionuclide Nominal Activity (Curies) Safety Basis Activity (Curies)

Am-241 [ ]PROP/ECI [ ]PROP/ECI Cm-242 [ ]PROP/ECI [ ]PROP/ECI Cm-244 [ ]PROP/ECI [ ]PROP/ECI Rb-88 [ ]PROP/ECI [ ]PROP/ECI Y-91m [ ]PROP/ECI [ ]PROP/ECI Nb-97m [ ]PROP/ECI [ ]PROP/ECI Nb-97 [ ]PROP/ECI [ ]PROP/ECI Rh-103m [ ]PROP/ECI [ ]PROP/ECI Rh-105m [ ]PROP/ECI [ ]PROP/ECI Rh-106 [ ]PROP/ECI [ ]PROP/ECI Ba-136m [ ]PROP/ECI [ ]PROP/ECI Ba-137m [ ]PROP/ECI [ ]PROP/ECI Pr-144 [ ]PROP/ECI [ ]PROP/ECI Pr-144m [ ]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-37 Rev. 3

Table 11.1 Radiation Areas at the Main Production Facility Area Dose Rate Designation mally occupied areas within RCA room 5 mrem/hr Normally occupied area S service cell without elerator operation ells, hot cells, and other lded vaults; cells; and rooms -

erial not present or accelerator n operation, after sufficient ay period ve RPF trench during solution > 5 mrem/hr but Radiation Area sfers 100 mrem/hr (transient occupation) ary cooling rooms during ration eneral area during accelerator ration in NDAS service cell ells, hot cells, and other lded vaults; cells; and rooms -

erial present or accelerator in > 100 mrem/hr (High Radiation High Radiation Area or Very High ration or shutdown without Area) or Radiation Area (rarely occupied, cient decay period > 500 rad/hr (Very High Radiation per ALARA controls)

Area)

S service cell with accelerator ration NE Medical Technologies 11.1-38 Rev. 3

Table 11.1 Airborne Radioactive Sources (Sheet 1 of 4)

Estimated Exterior Major Maximum Dose Rate System Component Location Sources Activity (Ci) (mrem/hr)(a)

Tritium purification S TPS gloveboxes H-3 300,000(b) < 0.25 system Driver vacuum AS IU cell H-3 [ ]PROP/ECI(c) < 0.25 hardware Off-gas piping, zeolite GS TOGS shielded cell I, Kr, Xe 120,000(d) < 0.25 beds IU cell atmosphere and Ar-41: 1E-05 Z1 IU cell Ar-41 and N-16 N/A PCLS N-16: 10(d)

I, Kr, Xe, and Z1 Supercell atmosphere Supercell gloveboxes 3 < 0.2 particulates Pipe trenches, valve pits, VS and VTS PVVS and VTS piping I, Kr, Xe 25,000(d) <1 and PVVS hot cell

a. Dose contribution from listed source in normally occupied area, includes direct dose at 30 cm from the exterior of the shielding surface and contributions from the derived air concentration.

b. Includes inventory in NDAS units.

c. H-3 activity is per NDAS unit.

d. Value is per irradiation unit (IU).

NE Medical Technologies 11.1-39 Rev. 3

GS, Off-gas Piping, Zeolite Beds, TOGS Shielded Cell (Conservative Best Estimate Activity)

Isotope Activity (Ci)

I-124 [ ]PROP/ECI I-125 [ ]PROP/ECI I-126 [ ]PROP/ECI I-129 [ ]PROP/ECI I-130 [ ]PROP/ECI I-131 [ ]PROP/ECI I-133 [ ]PROP/ECI Kr-81 [ ]PROP/ECI Kr-83m [ ]PROP/ECI Kr-85 [ ]PROP/ECI Kr-85m [ ]PROP/ECI Kr-87 [ ]PROP/ECI Kr-88 [ ]PROP/ECI Xe-127 [ ]PROP/ECI Xe-131m [ ]PROP/ECI Xe-133 [ ]PROP/ECI Xe-133m [ ]PROP/ECI Xe-135 [ ]PROP/ECI Xe-135m [ ]PROP/ECI Xe-138 [ ]PROP/ECI NE Medical Technologies 11.1-40 Rev. 3

RVZ1, Supercell Atmosphere, Supercell Gloveboxes (Conservative Best Estimate Activity)

Extraction Cell PVVS Cell Isotope Activity (Ci) Isotope Activity (Ci)

Br-83 [ ]PROP/ECI I-123 [ ]PROP/ECI I-130 [ ]PROP/ECI I-124 [ ]PROP/ECI I-131 [ ]PROP/ECI I-125 [ ]PROP/ECI I-132 [ ]PROP/ECI I-126 [ ]PROP/ECI I-133 [ ]PROP/ECI I-129 [ ]PROP/ECI I-134 [ ]PROP/ECI I-130 [ ]PROP/ECI I-135 [ ]PROP/ECI I-131 [ ]PROP/ECI Mo-99 [ ]PROP/ECI I-132 [ ]PROP/ECI Tc-99m [ ]PROP/ECI I-132m [ ]PROP/ECI Xe-133 [ ]PROP/ECI I-133 [ ]PROP/ECI Xe-133m [ ]PROP/ECI I-134 [ ]PROP/ECI Xe-135 [ ]PROP/ECI I-135 [ ]PROP/ECI Xe-135m [ ]PROP/ECI Kr-81 [ ]PROP/ECI Kr-83m [ ]PROP/ECI Purification Cell Kr-85 [ ]PROP/ECI Isotope Activity(Ci) Kr-85m [ ]PROP/ECI I-130 [ ]PROP/ECI Kr-87 [ ]PROP/ECI I-131 [ ]PROP/ECI Kr-88 [ ]PROP/ECI I-132 [ ]PROP/ECI Xe-127 [ ]PROP/ECI I-133 [ ]PROP/ECI Xe-131m [ ]PROP/ECI I-134 [ ]PROP/ECI Xe-133 [ ]PROP/ECI I-135 [ ]PROP/ECI Xe-133m [ ]PROP/ECI Mo-99 [ ]PROP/ECI Xe-135 [ ]PROP/ECI Tc-99m [ ]PROP/ECI Xe-135m [ ]PROP/ECI Xe-133 [ ]PROP/ECI Xe-138 [ ]PROP/ECI Xe-133m [ ]PROP/ECI Xe-135 [ ]PROP/ECI Xe-135m [ ]PROP/ECI NE Medical Technologies 11.1-41 Rev. 3

PVVS and VTS, PVVS and VTS Piping, Pipe Trenches, Valve Pits, and PVVS Hot Cell (Conservative Best Estimate Activity)

Isotope Activity (Ci)

I-123 [ ]PROP/ECI I-124 [ ]PROP/ECI I-125 [ ]PROP/ECI I-126 [ ]PROP/ECI I-129 [ ]PROP/ECI I-130 [ ]PROP/ECI I-131 [ ))PROP/ECI I-132 [ ]PROP/ECI 1-132m [ ]PROP/ECI I-133 [ ]PROP/ECI I-133m [ ]PROP/ECI I-134 [ ]PROP/ECI I-135 [ ]PROP/ECI Kr-81 [ ]PROP/ECI Kr-83m [ ]PROP/ECI Kr-85 [ ]PROP/ECI Kr-85m [ ]PROP/ECI Kr-87 [ ]PROP/ECI Kr-88 [ ]PROP/ECI Xe-122 [ ]PROP/ECI Xe-123 [ ]PROP/ECI Xe-127 [ ]PROP/ECI Xe-131m [ ]PROP/ECI Xe-133 [ ]PROP/ECI Xe-133m [ ]PROP/ECI Xe-135 [ ]PROP/ECI Xe-135m [ ]PROP/ECI Xe-138 [ ]PROP/ECI NE Medical Technologies 11.1-42 Rev. 3

Table 11.1 Estimated Derived Air Concentrations Source Description Location Particulate Halogen Noble Gas Tritium Total mary System Boundary IF General Area - 0.4% 0.1% - 0.4%

TPS Room - - - 1.4% 1.4%

IF General Area,

- - - 3.2% 3.2%

ium Systems Normal Operation IF General Area,

- - - 5.2% 5.2%

Maintenance ow-Grade Vaults RPF General Area - 0.1% 0.0% - 0.1%

PVVS Hot Cell - 12% 1.9% - 14%

VS Hot Cell RPF General Area - 0.0% 0.0% - 0.0%

Extraction Hot Cell 13% > 10 DAC 76% 0.0% > 10 DAC raction Hot Cell RPF General Area 0.0% 2.1% 0.0% 0.0% 2.1%

Purification Hot Cell 38% > 10 DAC 220% 0.0% > 10 DAC ification Hot Cell RPF General Area 0.0% 4.2% 0.0% 0.0% 4.2%

General Area Total - 0.4% 0.1% 8.3% 8.8%

F General Area Total 0.0% 6.4% 0.0% - 6.4%

NE Medical Technologies 11.1-43 Rev. 3

Table 11.1 Key Parameters for Normal Yearly Release Calculation Primary Confinement Parameter PVVS Pathway Hot Cells Boundary General Area mary Nuclide Kr, Xe, I, Kr, Xe, Ar 41, ntory Kr, Xe, I Kr, Xe, I, H-3 particulates N-16 stituents e of Radiation Beta and Gamma Beta and Gamma Beta and Gamma Beta and Gamma tted al Curies 9.7E+05 320 430 130 ary stituents Kr, Xe Kr, Xe Kr, Xe Kr, Xe, H-3 eased e of Radiation Beta and Gamma Beta and Gamma Beta and Gamma Beta and Gamma tted al Curies 9.2E+03 16 8.2 1 30 ay Time 1.7 days (Kr) dited for None 1 minute None 40 days (Xe) ay Carbon Filter on Carbon Guard Hot Cell RVZ1 Bed Carbon Filter on Carbon Filter on ne Removal Exhaust Facility RVZ1 Facility RVZ1 hanisms Carbon Filter on Carbon Delay Exhaust Exhaust Facility RVZ1 Beds Exhaust NE Medical Technologies 11.1-44 Rev. 3

able 11.1 Estimated Annual Releases from Normal and Maintenance Operations (Nuclides with Greater than 1 Ci Annual Release)

Radionuclide Annual Release (Ci)

Kr-83m 5.9E+00 Kr-85 1.2E+02 Kr-85m 5.0E+01 Kr-88 2.1E+00 Xe-131m 1.3E+03 Xe-133 7.8E+03 Xe-133m 1.1E+00 Xe-135 6.0E+00 Xe-135m 1.0E+01 H-3 1.3E+02 NE Medical Technologies 11.1-45 Rev. 3

Table 11.1 Liquid Radioactive Sources (Sheet 1 of 8)

Exterior Major Estimated Maximum Dose Rate System(a) Component(a) Location(a) Sources Activity (Ci) (mrem/hr)(d)

Target solution, Target Solution PS U-234, U-235, U-238 3 N/A unirradiated Preparation Area U-235 Fission Target solution in TSV AS IU cell (Neutrons and [ ]PROP/ECI(b) < 0.25 (operating)

Photons)

Target solution in TSV, AS TSV dump tank IU cell (see Table 11.1-3) [ ]PROP/ECI(b) < 0.03 (shutdown)

Water in the light water PS IU cell H-3 30(b) N/A pool AS Oil in NDAS pumps IU cell H-3 2000(b) N/A Primary cooling water IU cell and primary LS N-16 7.5(b) <2 in pump and piping cooling room Target solution in PS pump, extraction Supercell (see Table 11.1-3) [ ]PROP/ECI(c) <5 column, and lift tanks Mo eluate in Mo eluate PS Supercell Mo, [ ]PROP/ECI [ ]PROP/ECI(c) <3 hold tank PS Mo-99 product Supercell Mo-99, Tc-99 [ ]PROP/ECI(c) < 0.2 Target solution in target SS Tank vault (see Table 11.1-3) [ ]PROP/ECI(b) < 0.25 solution hold tank NE Medical Technologies 11.1-46 Rev. 3

Exterior Major Estimated Maximum Dose Rate System(a) Component(a) Location(a) Sources Activity (Ci) (mrem/hr)(d)

[

Liquid waste in annular WS Tank vault ]PROP/ECI and 3.8E+04 < 0.1 waste tank other fission products

[

Liquid waste in RLWS WS Tank vault ]PROP/ECI and 5.7E+04 < 0.1 collection tank other fission products

a. Physical and chemical properties of process solutions, special nuclear material inventories, and descriptions of the systems can be found in Chapter 4.

b. Value is per irradiation unit (IU).

c. Value is per cycle.

d. For normally-occupied areas.

NE Medical Technologies 11.1-47 Rev. 3

SCAS, Target Solution in TSV and TSV Dump Tank (Shutdown), IU Cell, and SCAS, Target Solution in TSV (Operating), IU Cell (Conservative Best Estimate Activity) otope Activity (Ci) Isotope Activity (Ci) Isotope Activity (Ci) g-109m [ ]PROP/ECI Nd-149 [ ]PROP/ECI Sn-128 [ ]PROP/ECI a-137m [ ]PROP/ECI Np-239 [ ]PROP/ECI Sr-89 [ ]PROP/ECI a-139 [ ]PROP/ECI P-32 [ ]PROP/ECI Sr-90 [ ]PROP/ECI a-140 [. ]PROP/ECI Pd-109 [ ]PROP/ECI Sr-91 [ ]PROP/ECI a-141 [ ]PROP/ECI Pm-147 [ ]PROP/ECI Sr-92 [ ]PROP/ECI Br-83 [ ]PROP/ECI Pm-149 [ ]PROP/ECI Tc-101 [ ]PROP/ECI Br-84 [ ]PROP/ECI Pm-151 [ ]PROP/ECI Tc-104 [ ]PROP/ECI e-141 [ ]PROP/ECI Pr-143 [ ]PROP/ECI Tc-99m [ ]PROP/ECI e-143 [ ]PROP/ECI Pr-144 [ ]PROP/ECI Te-127 [ ]PROP/ECI e-144 [ ]PROP/ECI Pr-145 [ ]PROP/ECI Te-129 [ ]PROP/ECI s-137 [ ]PROP/ECI Pr-146 [ ]PROP/ECI Te-131 [ ]PROP/ECI s-138 [ ]PROP/ECI Rb-88 [ ]PROP/ECI Te-131m [ ]PROP/ECI I-131 [ ]PROP/ECI Rb-89 [ ]PROP/ECI Te-132 [ ]PROP/ECI I-132 [ ]PROP/ECI Rh-103m [ ]PROP/ECI Te-133 [ ]PROP/ECI I-133 [ ]PROP/ECI Rh-105 [ ]PROP/ECI Te-133m [ ]PROP/ECI

-133m [ ]PROP/ECI Rh-105m [ ]PROP/ECI Te-134 [ ]PROP/ECI I-134 [ ]PROP/ECI Rh-106 [ ]PROP/ECI U-237 [ ]PROP/ECI I-135 [ ]PROP/ECI Rh-107 [ ]PROP/ECI U-239 [ ]PROP/ECI r-83m [ ]PROP/ECI Ru-103 [ ]PROP/ECI Xe-133 [ ]PROP/ECI r-85m [ ]PROP/ECI Ru-105 [ ]PROP/ECI Xe-133m [ ]PROP/ECI Kr-87 [ ]PROP/ECI Ru-106 [ ]PROP/ECI Xe-135 [ ]PROP/ECI Kr-88 [ ]PROP/ECI Sb-127 [ ]PROP/ECI Xe-135m [ ]PROP/ECI a-140 [ ]PROP/ECI Sb-128m [ ]PROP/ECI Xe-138 [ ]PROP/ECI a-141 [ ]PROP/ECI Sb-129 [ ]PROP/ECI Y-90 [ ]PROP/ECI a-142 [ ]PROP/ECI Sb-130 [ ]PROP/ECI Y-91 [ ]PROP/ECI a-143 [ ]PROP/ECI Sb-131 [ ]PROP/ECI Y-91m [ ]PROP/ECI o-101 [ ]PROP/ECI Se-81 [ ]PROP/ECI Y-92 [ ]PROP/ECI Mo-99 [ ]PROP/ECI Se-83 [ ]PROP/ECI Y-93 [ ]PROP/ECI Nb-97 [ ]PROP/ECI Sm-153 [ ]PROP/ECI Y-94 [ ]PROP/ECI b-97m [ ]PROP/ECI Sm-155 [ ]PROP/ECI Zr-95 [ ]PROP/ECI b-98m [ ]PROP/ECI Sn-127 [ ]PROP/ECI Zr-97 [ ]PROP/ECI d-147 [ ]PROP/ECI NE Medical Technologies 11.1-48 Rev. 3

PCLS, Primary Cooling Water in Pump and Piping, IU Cell and Primary Cooling Room (Conservative Best Estimate Activity)

Isotope Activity (Ci)

Ag-108 6.39E-04 Ag-110 1.25E-03 Ag-110m 6.14E-05 C-14 4.18E-05 Cl-38 1.01E-05 Co-60 4.64E-05 Co-60m 2.54E-04 Cr-51 7.22E-06 Cu-64 1.74E-04 Cu-66 3.54E-05 H-3 3.71E-03 N-16 4.98E-02 Na-24 7.98E-05 S-35 3.07E-05 NE Medical Technologies 11.1-49 Rev. 3

MEPS, Mo Eluate in Mo Eluate Hold Tank, Supercell (Conservative Best Estimate Activity)

Isotope Activity (Ci) Isotope Activity (Ci)

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-50 Rev. 3

TSSS, Target Solution in Target Solution Hold Tank, Tank Vault (Conservative Best Estimate Activity)

Isotope Activity (Ci) Isotope Activity (Ci)

Ag-109m [ ]PROP/ECI Pr-143 [ ]PROP/ECI Ag-111 [ ]PROP/ECI Pr-144 [ ]PROP/ECI Ag-112 [ ]PROP/ECI Pr-144m [ ]PROP/ECI Ba-137m [ ]PROP/ECI Pr-145 [ ]PROP/ECI Ba-139 [ ]PROP/ECI Rb-88 [ ]PROP/ECI Ba-140 [ ]PROP/ECI Rh-103m [ ]PROP/ECI Br-83 [ ]PROP/ECI Rh-105 [ ]PROP/ECI Cd-115 [ ]PROP/ECI Rh-105m [ ]PROP/ECI Ce-141 [ ]PROP/ECI Rh-106 [ ]PROP/ECI Ce-143 [ ]PROP/ECI Ru-103 [ ]PROP/ECI Ce-144 [ ]PROP/ECI Ru-105 [ ]PROP/ECI Cs-137 [ ]PROP/ECI Ru-106 [ ]PROP/ECI I-131 [ ]PROP/ECI Sb-129 [ ]PROP/ECI I-132 [ ]PROP/ECI Sm-153 [ ]PROP/ECI I-133 [ ]PROP/ECI Sm-156 [ ]PROP/ECI I-134 [ ]PROP/ECI Sn-121 [ ]PROP/ECI I-135 [ ]PROP/ECI Sn-127 [ ]PROP/ECI In-115m [ ]PROP/ECI Sr-89 [ ]PROP/ECI La-140 [ ]PROP/ECI Sr-90 [ ]PROP/ECI La-141 [ ]PROP/ECI Sr-91 [ ]PROP/ECI La-142 [ ]PROP/ECI Sr-92 [ ]PROP/ECI Nb-97 [ ]PROP/ECI Tc-99m [ ]PROP/ECI Nb-97m [ ]PROP/ECI Te-132 [ ]PROP/ECI Nd-147 [ ]PROP/ECI U-237 [ ]PROP/ECI Nd-149 [ ]PROP/ECI Y-90 [ ]PROP/ECI Np-239 [ ]PROP/ECI Y-91 [ ]PROP/ECI P-32 [ ]PROP/ECI Y-91m [ ]PROP/ECI Pd-109 [ ]PROP/ECI Y-92 [ ]PROP/ECI Pd-112 [ ]PROP/ECI Y-93 [ ]PROP/ECI Pm-147 [ ]PROP/ECI Zr-95 [ ]PROP/ECI Pm-149 [ ]PROP/ECI Zr-97 [ ]PROP/ECI Pm-151 [ ]PROP/ECI NE Medical Technologies 11.1-51 Rev. 3

RLWS, Liquid Waste in Annular Waste Tank, Tank Vault (Conservative Best Estimate Activity)

Isotope Activity (Ci) Isotope Activity (Ci)

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-52 Rev. 3

RLWS, Liquid Waste in RLWS Collection Tank, Tank Vault (Conservative Best Estimate Activity)

Isotope Activity (Ci) Isotope Activity (Ci)

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [4 ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-53 Rev. 3

Table 11.1 Solid Radioactive Sources (Sheet 1 of 5)

Estimated Exterior Major Maximum Dose Rate System(a) Component(a) Location Sources Activity (Ci) (mrem/hr)

AS Neutron Driver IU Cell Activation Products 300(b) N/A Rb, Cs, Ba, Sr, Y, La, and GS TOGS Components IU Cell and TOGS Cell 5.6E+04(b) < 0.25 Ce Neutron Multiplier, Activation and Fission AS IU Cell 1.5E+05(b) N/A SASS Products Spent Extraction [

PS Supercell [ ]PROP/ECI 2.6E+04(c) <5

]PROP/ECI Supercell and Solid PS Glassware [ ]PROP/ECI 100(c) N/A Waste Drum Storage Target Solution Fresh Uranium Metal PS and URSS Preparation and Storage U-234, U-235, U-238 3 N/A and Uranium Oxide Areas Liquid Waste Activation and Fission WI Solidified Waste Drum 125(d) < 0.25 Solidification Cell Products d Radwaste Spent Filters Supercell Iodine 400 <1 Subcritical Alpha-neutron Source AS IU Cell [ ]SRI N/A Multiplication Source (PuBe or AmBe)

a. Descriptions of the systems and their physical characteristics can be found in Chapter 4.

b. Value is per irradiation unit (IU).

c. Value is per cycle.

d. Value is per drum.

NE Medical Technologies 11.1-54 Rev. 3

TOGS, TOGS Components, IU Cell and TOGS Cell (Conservative Best Estimate Activity)

Isotope Activity (Ci)

Ba-137m 4.00E+01 Ba-139 3.61E+03 Ba-140 1.09E+03 Ba-141 1.09E+02 Ce-141 7.90E+01 Cs-137 4.22E+01 Cs-138 6.76E+03 La-140 1.01E+03 La-141 1.07E+02 La-142 2.70E+01 Rb-88 3.86E+03 Rb-89 4.48E+03 Sr-89 2.38E+03 Sr-90 2.37E+01 Sr-91 1.10E+03 Sr-92 1.57E+02 Y-90 2.30E+01 Y-91 5.59E+02 Y-91m 6.72E+02 Y-92 1.56E+02 Y-93 3.10E+01 NE Medical Technologies 11.1-55 Rev. 3

MEPS, Spent Columns, Supercell (Conservative Best Estimate Activity)

Isotope Activity (Ci) Isotope Activity (Ci)

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI NE Medical Technologies 11.1-56 Rev. 3

RLWI, Solidified Waste Drum, Liquid Waste Solidification Cell (Conservative Best Estimate Activity)

Isotope Activity (Ci)

Ba-137m 3.80E+00 Ce-141 7.59E-01 Ce-144 3.67E+01 Cs-137 4.01E+00 Eu-155 1.32E-01 H-3 1.20E-01 Nb-95 2.48E-01 Pm-147 1.01E+01 Pr-144 3.67E+01 Pr-144m 3.51E-01 Rh-103m 2.04E+00 Rh-106 4.04E+00 Ru-103 2.07E+00 Ru-106 4.04E+00 S-35 1.08E-01 Sm-151 8.99E-02 Sr-89 1.24E+00 Sr-90 5.30E-01 U-234 3.86E-02 Y-90 5.30E-01 Y-91 1.71E+01 Zr-95 1.23E-01 NE Medical Technologies 11.1-57 Rev. 3

Solid Radwaste, Spent Filters, Supercell (Conservative Best Estimate Activity)

Isotope Activity (Ci)

I-131 3.82E-01 I-132 5.13E-02 I-133 1.42E-01 I-135 2.02E-02 Mo-99 4.62E-04 Tc-99m 4.10E-04 NE Medical Technologies 11.1-58 Rev. 3

Table 11.1 Administrative Radiation Exposure Limits SHINE 10 CFR 20 Limit Administrative Type of Dose (rem/year) Limit (rem/year) more limiting of:

Total effective dose equivalent to whole body, or 5 2 Sum of deep-dose equivalent and committed dose 50 20 equivalent to any organ or tissue other than lens of eye dose equivalent to lens of eye 15 6 llow-dose equivalent to skin of the whole body or any 50 20 emity lared Pregnant Worker e to embryo/fetus during the entire pregnancy: taken as sum of the deep-dose equivalent to the woman and the 0.5 rem per 0.5 rem per e to the embryo/fetus from radionuclides in the embryo/ gestation period gestation period s and the woman vidual Members of the Public l effective dose equivalent 0.1 0.1 NE Medical Technologies 11.1-59 Rev. 3

Table 11.1 Radiation Monitoring Equipment Radiation Monitoring Instrument Type(a)(b) Location Function Radiation Survey Instruments able dose rate - neutron Various Routine and job coverage surveys able dose rate - beta/gamma Various Routine and job coverage surveys kers Various egress points within the RCA Ensure effective control of the spread of contamination sonnel contamination monitors Egress points from RCA Verify effectiveness of contamination controls Laboratory/Benchtop Instruments id Scintillation Counter (LSC) Counting room Tritium and low energy beta-emitting radionuclide sample analysis Background Sample Counter - alpha/beta Counting room Count smears and air samples ma Spectroscopy Counting room Various gamma-emitting radionuclide sample analyses Air Sampling and Monitoring sonnel Lapel Sampler Various as specified by procedure or RWP Representative air monitoring during work; internal dose assignment Samplers Various as specified by procedure or RWP Airborne radioactivity concentration measurement tinuous Alpha / Beta Air Monitor (CAM) Areas where airborne contamination may be present, Early detection of unanticipated increases in airborne as specified by procedure radioactivity concentration tinuous Tritium Air Monitor See Table 7.7-3 See Table 7.7-3 Radiation Area Monitors iation Area Monitors (RAM) See Table 7.7-2 See Table 7.7-2 Radiological Effluent Monitor k Release Monitor Located in the main production facility stack Direct exposure to gamma and beta emitting radionuclides released through the stack of the SHINE main production facility is monitored and measured rcoal Delay Bed Effluent Monitor Located at the outlet of the process vessel vent Monitor to trend the performance of the charcoal system (PVVS) charcoal delay beds delay beds. Ensure the PVVS effluent stream is monitored if the safety-related effluent release point is in use.

a. See Table 7.7-1 for safety-related process radiation monitors.

b. See Table 11.1-14 for Environmental Monitoring equipment and locations.

NE Medical Technologies 11.1-60 Rev. 3

Table 11.1 Radiological Postings Posting Requirement Accessible area in which radiation levels could result in an CAUTION individual receiving in excess of 5 mrem in one hour 30 cm RADIATION AREA from the radiation source or surface that the radiation penetrates.

CAUTION HIGH RADIATION AREA Accessible area in which radiation levels could result in an individual receiving in excess of 100 mrem in one hour 30 cm or from the radiation source or surface that the radiation penetrates.

DANGER HIGH RADIATION AREA Accessible area in which radiation levels could result in an GRAVE DANGER individual receiving an absorbed dose in excess of 500 rads VERY HIGH RADIATION AREA in one hour at one meter from a radiation source or from any surface that the radiation penetrates.

CAUTION AIRBORNE Licensed airborne radioactive materials in a room, enclosure, RADIOACTIVITY AREA or area exists in concentrations exceeding the derived air concentrations specified in 10 CFR 20, Appendix B, Table I, or or when an individual present in the area without respiratory protective equipment could exceed, during the hours an DANGER AIRBORNE individual is present in a week, an intake of 0.6% of the RADIOACTIVITY AREA annual limit on intake or 12 DAC-hours.

An area where removable contamination levels are above CAUTION 20 dpm/100 cm2 of alpha activity or 1,000 dpm/100 cm2 CONTAMINATION AREA beta/gamma activity.

CAUTION RADIOACTIVE MATERIAL(S)

Areas or rooms in which there is use of, or stored, an amount or of licensed radioactive material exceeding 10 times the quantity of material in Appendix C to 10 CFR 20.

DANGER RADIOACTIVE MATERIAL(S)

NE Medical Technologies 11.1-61 Rev. 3

Table 11.1 Environmental Monitoring Locations nitoring Type Location Rationale Groundwater Sampling st Well The groundwater gradient is to the west and the south Test well located directly west of

-GW4A and thus any groundwater contamination is likely to flow the SHINE facility mpling to the west and to the south.

st Well The groundwater gradient is to the west and the south Test well located directly south of

-GW2A and thus any groundwater contamination is likely to flow the SHINE facility mpling to the west and to the south.

Environmental Dosimeters One in each of the 16 compass directions from the site 1 - 16 Site Boundary center.

Outside the main production facility but within the site boundary One in each of the four cardinal directions surrounding 17 - 20 (ED 17 north, ED 18 east, ED 19 the main production facility.

south, ED 20 west)

Rock County Christian Elementary School (ED 21)

Jackson Elementary School Special interest areas (e.g., population centers, nearby 21 - 23 (ED 22) residences or schools).

University of Wisconsin - Rock County (ED 23)

To serve as a control (i.e., located a significant distance ED 24 Kennedy Elementary School from the facility such that is represents a background dose).

Air Samplers Control air sampler located a sufficient distance from the Sampler Off-site location, co-located with SHINE facility such that airborne samples are AS 1) ED 24 unaffected by airborne effluent releases from the facility.

This direction has high ground level deposition factor (D/Q) and is in the direction of Janesville. Since the Close to property line, north of the Sampler community of Janesville is relatively close to the site main production facility, AS 2) boundary, this air sampler location is credited with co-located with ED 1 satisfying two of the conditions for air sample location recommendations in Table 3.12-1 of NUREG-1301.

Close to property line, east of the Sampler This direction has high D/Q and is in the direction of main production facility, AS 3) dairy production and the horse pasture.

co-located with ED 5 Close to property line, west of the Sampler main production facility, This location ensures all directions are monitored.

AS 4) co-located with ED 9 Close to property line, south of the Sampler This location is in the direction of the nearest occupied main production facility, AS 5) structure.

co-located with ED 13 NE Medical Technologies 11.1-62 Rev. 3

NE Medical Technologies 11.1-63 Rev. 3 NE Medical Technologies 11.1-64 Rev. 3 NE Medical Technologies 11.1-65 Rev. 3 NE Medical Technologies 11.1-66 Rev. 3 NE produces medical isotopes by the fission of low enriched uranium (LEU) driven by elerator-produced neutrons. Several irradiation and processing steps create liquid, gaseous, olid radioactive waste materials. This section describes the management program, controls, disposal pathways established to ensure proper identification, classification, control, cessing (as required), and packaging, for each anticipated radioactive waste stream erated by the SHINE facility. SHINE is committed to comply with all applicable local and onal regulations for managing radioactive wastes.

NE will comply with the following federal regulations related to radioactive wastes:

• 10 CFR 20, Standards for Protection Against Radiation

• 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste

• 10 CFR 71, Packaging and Transportation of Radioactive Material

• 40 CFR, Chapter I, Subchapter F, Radiation Protection Programs

• 40 CFR, Chapter I, Subchapter I, Solid Wastes

• 49 CFR, Chapter I, Subchapter C, Hazardous Materials Regulations NE is regulated by the NRC. The State of Wisconsin regulates radioactive waste once it es the SHINE facility and is transported. SHINE complies with Wisconsin regulations relating he transportation and disposal of hazardous waste per Wisconsin Administrative Code pter NR 662. The State of Wisconsin implements the U.S. Department of Transportation T) radioactive waste transportation regulations.

ioactive wastes are prepared for shipment in approved shipping containers and shipped site using common or contract carriers in compliance with DOT regulations (49 CFR) and CFR 20, 10 CFR 61 and 10 CFR 71, as applicable.

NE complies with the waste acceptance criteria (WAC) of the selected licensed disposal lities, including any local or state regulations specified in those criteria. The State of consin is in the Midwest Interstate Low-Level Radioactive Waste Compact. Waste disposal s available for this compact include:

• EnergySolutions in Clive, UT

• Waste Control Specialists (WCS) in Andrews, TX tion 11.1 describes the program and procedures for controlling and assessing radioactive osures associated with radioactive sources, including radioactive waste streams.

.1 RADIOACTIVE WASTE MANAGEMENT PROGRAM Radioactive Waste Management Program is coordinated with the Radiation Protection gram under the Plant Manager. The goal of the Radioactive Waste Management Program is inimize waste generation, minimize exposure of personnel, and to protect the general public environment. The authority, duties, and responsibilities of personnel in the waste nagement organization are prescribed in the Radioactive Waste Management Program ument.

NE Medical Technologies 11.2-1 Rev. 2

Plant Manager reports to the Chief Operating Officer. The Plant Manager has overall ponsibility for the safe operation of the SHINE facility and is responsible for ensuring the ection of personnel from radiation exposure resulting from processing, handling and storing oactive material and waste. The Plant Managers responsibilities are to:

• Assign responsibility and delegates commensurate authority to implement the Radioactive Waste Management Program.

• Provide waste management staff appropriate to the scope of operations and experienced in waste management operations.

• Ensure that the waste management self-assessment program is implemented.

• Ensure compliance with applicable federal and state regulations, and facility license conditions.

• Approve changes to the facility Process Control Program.

.1.2 Radiation Protection Manager Radiation Protection Manager reports to the Plant Manager. The Radiation Protection nager is responsible for establishing and maintaining the Radioactive Waste Management gram. The Radiation Protection Department maintains organizational independence from the rations Department. The Radiation Protection Manager and Radiation Protection staff ponsibilities are to:

• Develop waste management procedures for the processing, packaging and shipment of radioactive waste from the facility.

• Ensure that the concept of ALARA is incorporated into the Radioactive Waste Management Program procedures and is practiced by personnel.

• Process radioactive waste generated at the facility.

• Provide technical input to the design of equipment and processes.

• Perform radiological analysis tasks supporting the Radioactive Waste Management Program.

• Provide technical input to the Radioactive Waste Management Program training program.

• Maintain contractual relationships with waste disposal sites, waste processing facilities, and radioactive waste carriers.

• Maintain working knowledge of waste disposal acceptance criteria, regulations, standards and guides.

• Conduct self-assessments of radioactive waste management practices and compliance with procedures.

.1.3 Training Manager Training Manager reports to the Plant Manager and is responsible for implementation of the ioactive Waste Management Program training as described in the Radiation Protection gram. The Training Manager has the following responsibilities:

• Develops the waste management training and qualification program in accordance with facility procedures and ensuring compliance with 49 CFR 172, Subpart H, Training.

• Provides training to personnel commensurate with the radiological waste hazard to which they may be exposed.

NE Medical Technologies 11.2-2 Rev. 2

• Evaluates the waste management and qualification training program periodically.

Reviews program content to ensure it remains current and adequate to ensure worker safety.

.1.4 Quality Manager Quality Manager reports to the Vice President Regulatory Affairs & Quality. The Quality nager has the following responsibilities:

• Review and audit facility radioactive waste handing, storing and shipping activities in accordance with the Quality Assurance Program Description to verify compliance with facility procedures, applicable federal and state regulations and applicable regulatory guides.

.1.5 Shipping Personnel viduals who perform the duties of shipping radioactive waste, are trained in accordance with CFR 172, Subpart H, Training.

.1.6 Radioactive Waste Management Procedures ioactive Waste Management Program implementing procedures are developed to provide ction for efficient and safe conduct of waste operations. The procedures include applicable trols and limits significant to the waste management operation. The procedures include:

• Waste minimization and pollution prevention, including process controls to minimize generation of waste and separation of radioactive waste and nonradioactive waste to reduce volumes of radioactive wastes.

• Radiological characterization and waste classification.

• Operating and process controls with parameters for processing wastes.

• Verification of compliance with disposal and processor site WAC.

• Preparation of radioactive waste for shipment, including preparation of manifests and notifications, and measures for security on site and during transport.

• Container specifications, selection, packaging wastes, inspections, vehicle inspections, and proper loading and shoring of shipments.

• Marking, labeling and placarding requirements.

• Radioactive materials and contamination survey requirements and limits for shipment on public highways.

• Waste disposal recordkeeping.

• Interim waste storage controls and recordkeeping.

Radiological Waste Management Program and implementing procedures are developed and trolled in accordance with SHINEs document control requirements.

.1.7 Record Keeping and Document Controls ords are developed and retained in accordance with the requirements specified in the iation Protection Program (see Subsection 11.1.2.1.8), the SHINE Document Control NE Medical Technologies 11.2-3 Rev. 2

.1.8 Waste Management Audits ility radioactive waste management audits are conducted, at a minimum, on an annual basis ccordance with 10 CFR 20.1101(c) for the purpose of reviewing the functional and safety ments of the radioactive waste management program. The audits also evaluate programmatic rts to minimize production of radioactive wastes. The audit activity is led by the Review and it Committee (see Section 12.2) as a subset of the Radiation Protection Program audit and results are sent to executive management. Any deficiencies identified by the audit are ressed by the corrective action process.

.1.9 Technical Specifications iables, conditions, or other items that may be subjects of a technical specification associated radioactive waste management are contained in the facility Technical Specifications.

.2 RADIOACTIVE WASTE CONTROLS ioactive waste is generally considered to be any item or substance which is no longer of use he facility and which contains radioactivity above the established natural background oactivity. The wastes generated by the SHINE facility are not spent nuclear fuel, high-level te, or byproduct material as defined in paragraphs (2), (3) and (4) of the definition of roduct Material set forth in 10 CFR 20.1003. Therefore, the radioactive wastes generated by SHINE facility are all classified as low level waste (LLW). The LLW generated by the SHINE lity during operation is expected to be classified as Class A, Class B or Class C waste. The tron multipliers are designed for the life of the facility and will be disposed of as greater-than ss C (GTCC) waste during decommissioning.

the purposes of transportation, packaged wastes may be categorized as low specific activity A), requiring Type A packaging, or requiring Type B packaging.

the purposes of both transportation and operational ALARA, wastes may be categorized as er contact handled or remote handled. The upper limit for remote handled waste dose rates is ned based on payload limits for the specific shielded transportation casks used and on WAC he intended disposal site.

iation Protection Program requirements and the ALARA Program (see Section 11.1) apply to oactive waste management, including, but not limited to, control of materials, monitoring and eys, radiologically controlled area (RCA) access control, contamination control and personnel itoring. ALARA goals and implementation are detailed in Subsection 11.1.3.

material staging building is used for interim storage of wastes for decay and for preparation for ment. Wastes are not stored for more than five years. The material staging building design luated the shielding provided by the building to ensure 10 CFR 20 site dose limits are met and RA principles are followed.

NE Medical Technologies 11.2-4 Rev. 2

oactive wastes.

.2.1 Radioactive Waste Minimization ste minimization and pollution prevention are key elements of the Radiological Waste nagement Program. Implementing procedures (see Subsection 11.2.1.6) address:

a. Responsibilities for waste minimization and pollution prevention.

b. Employee training and education on general environmental activities and hazards regarding the facility, operations, pollution prevention, waste minimization requirements, goals and accomplishments.

c. Setting goals for reducing the volume or radioactivity in each waste stream.

d. Sorting and compaction to reduce the volume of solid waste.

e. Segregation of nonradiological and radiological wastes to reduce the volume of radiological waste due to contamination.

f. Process controls that minimize generation of wastes.

g. Periodic assessments to identify opportunities to reduce or eliminate the generation of wastes.

h. Recognition of employees for efforts to improve waste minimization and environmental conditions.

.2.2 Waste Stream Sources ste management operations occur in the main production facility and the material staging ding (see Figure 1.3-1 and Figure 1.3-3). At least 5,600 square feet (ft2) of the material ing building is for temporary storage to allow for decay. As allowed by the waste drum ign, building design, and programmatic controls (e.g., inspection requirements), drums may tored in multiple layers. Equipment and associated features for containment and/or kaging, storage, and disposal of solid, liquid, and gaseous radioactive waste are discussed in section 9b.7.3, Subsection 9b.7.4, and Subsection 9b.7.5.

nges to the facility will be performed in accordance with 10 CFR 50.59, Changes, Tests and eriments, and will be assessed for their impact on radioactive waste sources or nagement, as applicable.

le 11.2-1 summarizes the facility waste streams, characteristics, generation rates, and ment categories. The waste streams and typical waste classifications are described in the wing subsections.

.2.2.1 Uranium Receipt and Storage System ste generated by uranium receipt and storage includes used cannisters in which new uranium al and uranium oxide are received. The used cannisters are processed as Class A waste, if returned to the supplier. The uranium receipt and storage system (URSS) utilizes gloveboxes high efficiency particulate air (HEPA) filters in the air supply and return lines. The spent PA filters are Class A waste.

NE Medical Technologies 11.2-5 Rev. 2

target solution preparation process may generate waste in the form of spent filters from uranyl sulfate dissolution tanks, if not cleaned and reused, and spent HEPA filters from vebox air supply and return lines. The spent filters are Class A waste.

.2.2.3 Irradiation Unit rradiation unit (IU) consists of a subcritical assembly system (SCAS) coupled with a neutron er assembly system (NDAS). The IU components become activated during their service life.

AS major components are designed for the life of the facility and are not anticipated waste ams. Spent NDAS components are Class A waste. Contaminated oil from the NDAS vacuum ps is Class B waste.

.2.2.4 TSV Off-Gas System target solution vessel (TSV) off-gas system (TOGS) removes radiolysis and fission duct gases from the TSV during irradiation operation and from the TSV dump tank during l down operation. There are a total of eight independent TOGS, one for each IU.

TOGS contains skid-mounted equipment that includes recombiner beds, demisters, and lite beds. Skid replacement occurs infrequently. Skids containing recombiner beds and misters are treated with an acid flush and processed as Class A or B waste. Zeolite beds are igned for the life of the facility, however, if replaced more frequently and processed arately from the remainder of the skid components, the zeolite beds are expected to be ss B or Class C waste.

.2.2.5 Molybdenum Extraction and Purification System molybdenum extraction and purification system (MEPS) separates molybdenum from an diated uranyl sulfate target solution. The molybdenum is then concentrated and purified into a ium molybdate solution. The MEPS is located within a series of hot cells. Waste generated the MEPS includes spent molybdenum extraction columns, [

]PROP/ECI , and purification glassware. MEPS liquid wastes are processed by the oactive liquid waste immobilization (RLWI) system.

nt extraction columns [ ]PROP/ECI are stored in a hot cell, then sferred to the drum storage bore holes for decay, and ultimately disposed as Class B or C te.

glassware used in this process is not expected to contain significant quantities of long-lived onuclides and is Class A waste.

ective ion stripping associated with MEPS column washes occurs in the RLWI system. The nt selective ion stripping columns are disposed as Class B or Class C waste.

.2.2.6 Process Vessel Vent System process vessel vent system (PVVS) removes radioactive particulates, iodine, and noble es that are generated within the radioisotope production facility (RPF) and primary system NE Medical Technologies 11.2-6 Rev. 2

nt carbon guard beds are Class A or Class B waste. Condensate from PVVS can be blended other waste streams and processed by RLWI.

.2.2.7 Iodine and Xenon Purification and Packaging System iodine and xenon purification and packaging (IXP) system separates the iodine fission ducts from the uranyl sulfate target solution or from [ ]PROP/ECI. The system generates spent iodine recovery, [

]PROP/ECI.

ne recovery, [ ]PROP/ECI will be regularly changed out and Class B or Class C waste.

.2.2.8 Hot Cells cells contain HEPA and carbon filter combinations on the air supply and return lines. Spent PA and carbon filters are Class A waste.

.2.2.9 Primary Closed Loop Cooling System primary closed loop cooling system (PCLS) has potential for radioactive contamination due inor leakage from the PSB and activation products. Contamination would collect on the LS filters and deionizer resins. PCLS filters could become contaminated with radionuclides to activation of corrosion particles as the water passes through the TSV, however, corrosion e stainless steel components is expected to be small. The spent PCLS filters are expected to Class A waste. PCLS deionizer resins are contained in disposable deionizer units. The tanks designed for complete replacement without removal of the ion exchange resins in the tanks.

disposable tanks are Class A waste.

.2.2.10 Light Water Pool System light water pool has potential for radioactive contamination due to minor leakage from the B and activation products. Any contamination would collect on the filters and deionizer resins d to cleanup the light water pool. Similar to the PCLS, the deionizer resins are contained in osable deionizer units and are expected to be Class A waste. Spent filters are expected to be ss A waste.

.2.2.11 Radioactive Liquid Waste ioactive liquid waste streams include waste liquids from:

• MEPS

• IXP system

• PVVS NE Medical Technologies 11.2-7 Rev. 2

• Laboratory liquid waste liquid waste streams are shown in Table 11.2-1.

id waste streams are collected in uranium liquid waste and radioactive liquid waste collection s, consolidated in liquid waste blending tanks and treated for disposal using the RLWI tem. The quantity and size of the tanks are managed to maximize decay time and provide a er for upset conditions. Each uranium liquid waste tank has at least [ ]PROP/ECI acity and the liquid waste collection and blending tanks each have at least 600 gallons acity. Hold times for decay are based on minimizing dose rates to workers during the obilization process. Solidified liquid waste is expected to be Class A.

chemical composition and relative radiological inventory of liquid waste streams is presented able 11.2-6.

.2.2.12 Radioactive Gaseous Waste orne radioactive sources are present in the tritium purification system (TPS), PVVS, TOGS, uum transfer system (VTS), and the NDAS. Airborne radioactive sources and release are ressed in Subsection 11.1.1.1 and Table 11.1-5.

RCA ventilation systems generate spent prefilters, HEPA filters and carbon filters that are ss A generated solid waste.

.2.3 Technical Specifications iables, conditions, or other items that may be subjects of a technical specification associated radioactive waste controls are contained in the facility Technical Specifications.

.3 RELEASE OF RADIOACTIVE WASTE ease, for the purposes of this subsection, means that wastes are processed and packaged as uired to meet the WAC of an established, licensed LLW disposal facility. Processing may be prised of one or more of several operations, including compaction, solidification with an ropriate solidification agent, adsorption onto a solid medium (e.g., elemental iodine onto vated carbon filters), interim storage for decay of radionuclides, consolidated handling and cessing, extraction and consolidation of radionuclides by segregation, and mixing (possibly more than one waste stream) so that the bulk volume of waste is readily disposable.

iation monitoring of effluent waste streams is described in Section 7.7. Radiation monitoring uirements are also described in the Radiation Protection Program. The Radiation Protection gram is described in detail in Subsection 11.1.2.

id effluent is not routinely discharged from the RCA. Radioactive liquid discharges from the NE facility to the sanitary sewer are infrequent and made in accordance with CFR 20.2003 and 10 CFR 20.2007. There are no piped liquid effluent pathways from the RCA he sanitary sewer. Liquids collected for discharge from the RCA are sampled and analyzed NE Medical Technologies 11.2-8 Rev. 2

er. Liquid discharge volumes are estimated to be less than 40 gallons weekly.

le 11.2-1 shows the anticipated waste generation, classifications, shipment types, and ected disposal sites for the identified waste streams. Final determinations of waste sification and management will be made in accordance with the Radioactive Waste nagement Program implementing procedures.

.3.1 Solid Wastes subsections below discuss the methodology for the eventual release of the major solid tes generated by the SHINE facility. Processing requirements are in accordance with the eiving facilitys WAC and will be modified as needed to reflect any change in the disposal site WAC.

.3.1.1 Irradiation Units d waste streams associated with the IUs are the NDAS activated components. The NDAS is prised of an accelerator section, pumping section, roots stack, and target chamber embly. The target chamber assembly is expected to be Class A waste and the WAC specified EnergySolutions will apply. The accelerator stage, pumping stage and roots stack are sidered oversize and must meet specific WAC applicable to oversize components.

le 11.2-2 displays the typical methodology associated with disassembly and processing of waste stream.

.3.1.2 Spent Columns nt molybdenum extraction columns, [ ]PROP/ECI, and IXP recovery, PROP/ECI

] will be held in hot cells for decay, then consolidated supercell export waste drums prior to disposal.

columns are removed from the process lines using quick-disconnect style inlet and outlet nectors specifically designed for use with remote manipulators in hot cell environments.

iation and wear-resistant seals and automatically closing valves built into the connectors vide leak tightness to minimize or prevent leakage.

r removing a spent column from the originating process, it is stored in a hot cell for sufficient to allow short-lived fission products to decay. After several columns have decayed, they are sported out of the cell in one transfer to reduce personnel exposure and the number of sfer operations. The number of columns transferred is limited based on export waste drum acity. The export waste drum is shielded to ensure personnel doses are maintained ALARA within procedure limits during the transfer. The estimated dose rate for an extraction column, e time of process removal is approximately 9500 rem/hr at 3 feet unshielded. The peak dose drops to approximately 580 rem/hr at 3 feet unshielded after storage in the hot cell.

en a set of columns are to be transferred out of the hot cell, they are remotely loaded into an ort waste drum within a shielded cask. Dose rates from the cask and contamination levels are firmed to be within limits, then the cask is remotely transported to a bore hole for interim NE Medical Technologies 11.2-9 Rev. 2

en a shipment of columns is to be prepared, the export waste drum is retracted using the ote-controlled grappler and placed into a shielded cask and the cask is transported to an a for loading into a Type B shipping container.

spent columns are expected to be Type B or C generated waste and have no specified time uirement in storage. The spent columns are stored in order to consolidate shipments to imize handling for ALARA and to consolidate the columns to reduce disposal volumes.

uirements for this waste stream are presented in Table 11.2-3.

ective stripping columns are contained within the RLWI system. When a column is removed service it is dewatered and processed for disposal as Type B or C waste.

.3.1.3 Process Glassware nt molybdenum purification glassware is remotely handled to move the glassware from the cell to an export waste drum. The glassware may be crushed in the waste drum using a otely controlled compactor and transported to the material staging building in a shielded sport cask. Requirements for this waste stream are presented in Table 11.2-4.

.3.1.4 Zeolite Beds silver coated zeolite beds are a component of the TOGS and are provided to remove iodine the sweep gas. Toxicity characteristic leaching procedure (TCLP) would result in the sification of this waste as Resource Conservation and Recovery Act (RCRA) waste; ever, the waste is also radioactive and as such may be a mixed low level waste (MLLW). The te classification for this material is a function of both the efficiency of the zeolite beds and the nge out frequency of the beds. The design goal is for the beds to last the lifetime of the lity; however, this waste stream is assumed to be replaced every five years. The zeolite bed the potential to be Class B or Class C waste.

.3.1.5 Recombiner Beds, Demister and Component Replacement waste stream is associated with the TOGS. This waste stream is based on infrequent acement of the TOGS skids. Acid flushing of the skid components (excluding the zeolite s) will be performed prior to disposal. Cs-137 and Sr-90 are expected to dominate the waste sification. Remote handling and packaging may be required due to considerable dose rates ected should replacement be required. This waste stream is Class A or Class B waste.

.3.1.6 PCLS and LWPS Deionizer Units PCLS and LWPS deionizer resins are contained in disposable deionizer units. The spent s are dewatered and disposed as Class A generated waste.

NE Medical Technologies 11.2-10 Rev. 2

eral waste streams are solidified on site to meet DOT criteria and disposal site WAC, as cribed in Subsection 11.2.2. The consolidated liquid waste stream (post-treatment) is enable for disposal as Class A waste at EnergySolutions.

.3.2.1 Consolidated Liquids ioactive liquid waste and estimated generated volumes are provided in Table 11.2-1.

nium liquid wastes and other radioactive liquid wastes are collected and processed arately, then blended prior to solidification. Uranium liquid wastes may consist of ybdenum extraction column acid wash, extraction column water wash, iodine recovery mn [ ]PROP/ECI, VTS knockout pot contents, spent target solution, or ontamination waste. Radioactive liquid waste may consist of [

]PROP/ECI, purification PROP/ECI te, [ ] , or PVVS densate. Blending of wastes is performed without exceeding the maximum uranium centration applicable to the receiving disposal site. Certain fissile material may be exempted er 10 CFR 71.15.

waste stream process includes removal of radionuclides, radioactive decay, pH adjustment, ding of uranium and radioactive liquid wastes, and solidification in 55-gallon drums using a dification agent.

anticipated disposal site for the solidified liquid waste is EnergySolutions.

uirements for this waste stream are presented in Table 11.2-5.

.3.3 Gaseous Waste Streams orne radioactive sources are identified in Subsection 11.1.1 and Table 11.1-5. The RCA tilation system filtering and exhaust stack discharge is described in Subsection 9a2.1.1. The aust stack location is shown on Figure 1.3-1. The stack release monitor provides continuous nitoring of radioactive noble gas stack releases and a means to sample and measure the k air for particulate, iodine, and tritium concentration to ensure compliance with gaseous ent regulatory limits. The estimate of annual release of radionuclides is provided in le 11.1-8. The effect of releases on the surrounding environment is addressed by the ironmental Monitoring Program described in Subsection 11.1.7.

NE Medical Technologies 11.2-11 Rev. 2

Table 11.2 Estimated Annual Waste Stream Summary (Sheet 1 of 2)

As As As Class as Generated Generated Disposed Shipment Description Matrix Generated Amount Units (ft3) Type Destination(a)

PS Extraction Columns [ [

B or C [ ]PROP/ECI ft3/yr 270 Type B WCS

]PROP/ECI ]PROP/ECI ctive Ion Stripping Columns [ ]PROP/ECI B or C 72 ft3/yr 72 Type B WCS Separation Columns [ ]PROP/ECI B or C [ ]PROP/ECI ft3/yr 35 Type B WCS Type A or PS Deionizer Units Resin A 48 ft3/yr 80 EnergySolutions LSA Type A or S Deionizer Units Resin A 48 ft3/yr 80 EnergySolutions LSA Type A or nium Canisters Solid A 2.0(b) ft3/yr 3.3 EnergySolutions LSA Type A or AS Accelerator Subassembly Solid A [ ]PROP/ECI ft3/yr 13,600 EnergySolutions LSA Type A or AS Target Chamber Subassembly Solid A [ ]PROP/ECI ft3/yr 1330 EnergySolutions LSA Type A, B, or EnergySolutions or GS Skids Solid A or B 922 ft3/yr 1540 LSA WCS GS Zeolite Beds Solid B or C 0.64 ft3/yr 1 Type B WCS Type A or PS Filters Solid A 1.6 ft3/yr 2.7 EnergySolutions LSA Type A or S Filters Solid A 1.6 ft3/yr 2.7 EnergySolutions LSA S, URSS, PVVS, Hot Cell, RVZ1, RVZ2, RLWI Type A or Solid A 182 ft3/yr 142 EnergySolutions A Filters LSA Type A or Cell, RVZ1, RVZ2 Charcoal Filters Solid A 32 ft3/yr 54 EnergySolutions LSA Type A or S Uranyl Sulfate Solution Filters Solid A 0.35(c) ft3/yr 0.58 EnergySolutions LSA NE Medical Technologies 11.2-12 Rev. 2

As As As Class as Generated Generated Disposed Shipment Description Matrix Generated Amount Units (ft3) Type Destination(a)

EnergySolutions or S Carbon Guard Bed Solid A or B 0.48 ft3/yr 0.81 Type B WCS Type A or PS Glassware Solid A 208 ft3/yr 347 EnergySolutions LSA Type A or s A Trash Solid A 400(d) ft3/yr 677 EnergySolutions LSA Type A or taminated Oil Oil B 2 ft3/yr 3.3 WCS LSA action Column Acid Wash Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI action Column Water Wash Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI ne Recovery Column [ ]PROP/ECI Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI nt Target Solution Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI Type A or 2,599(f)(g) EnergySolutions LSA uum Transfer System Knockout Pot Liquid(e) A 14 gal/yr iological Laboratory Waste Liquid(e) A 275 gal/yr (e) ontamination Waste Liquid A 2,768 gal/yr ichem Purification Waste & Rotary Evaporator Liquid(e) A 82 gal/yr densate

]PROP/ECI Liquid(e) A [ ]PROP/ECI [ ]PROP/ECI S Condenser Condensate Liquid(e) A 701 gal/yr

a. Waste destination may be subject to change.

b. Uranium metal and/or uranium oxide cannisters may be returned to the supplier in lieu of disposition as solid waste.

c. TSPS uranyl sulfate dissolution tank filter elements may not become a waste stream if reconditioned and reused.

d. Class A trash is exclusive of other solid wastes identified in the table.

e. Liquid waste streams may be reused or may be combined and treated as a homogenous influent waste stream and solidified together.

f. As shipped volume of liquid waste streams is in the form of a uniform solidified matrix using a solidification agent.

g. 25 percent margin has been added to volume of solidified liquid shipped waste.

NE Medical Technologies 11.2-13 Rev. 2

Table 11.2 Waste Methodology for Accelerator Requirement Basis assemble irradiation unit Operational requirement.

parate accelerator section, mping section, and roots stack m the target chamber embly).

ermine if free liquid is present Required to meet WAC maximum free liquids requirement absorb liquids, if present. of 1 percent. This is particularly applicable to drift tubes and target chamber section waste.

ke waste characterization Waste must be characterized in the manner appropriate asurements. and in conformance with the procedures of the destination to which it will be sent.

vide capability to load oversized Ensure capability to maneuver radioactive oversize debris.

ris into cargo container.

vide storage, waste Items meeting the "standard debris" definition are shipped regation, consolidation and in a roll-off. One roll-off may be continuously stored in the kaging capacity. material staging building. Oversized items (non-standard debris) are shipped in a cargo container. One cargo container may be continuously on-site.

void space (if required) in Required to meet WAC requirement to minimize void ordance with the WAC. space.

NE Medical Technologies 11.2-14 Rev. 2

Table 11.2 Waste Methodology for Spent Columns(a)

Requirement Basis d spent columns in hot cell for a period of Spent columns are highly radioactive when ay sufficient to allow short-lived fission removed from active service. Hold time is for ducts to decay. decay and consolidated processing.

mote transfer from hot cell to export waste Maintain worker dose ALARA.

m.

vide safe, shielded storage outside of hot Protected on-site storage until a full shipment

. of spent columns is prepared for disposal.

vide management controls to ensure proper Since multiple columns can be held in each hot d time is applied to spent columns. cell post service, it is necessary to ensure each column has been held for a sufficient time to meet radiological dose requirements during handling prior to being transferred.

ermine if free liquid is present and absorb Required to meet WAC maximum free liquids ids, if present. requirement of 1 percent.

void space (if required) in accordance with Required to meet WAC requirement to WAC. minimize void space.

Applicable to spent molybdenum extraction columns [ ]PROP/ECI and IXP recovery, [ ]PROP/ECI.

NE Medical Technologies 11.2-15 Rev. 2

Table 11.2 Waste Methodology for Process Glassware Requirement Basis mote transfer from hot cell to export waste Maintain worker dose ALARA.

m.

ear sample glassware. Waste characterization to confirm disposal site and applicable WAC.

ssware is compacted. Glassware can be compacted for efficient packaging and transportation.

ermine if free liquid is present and absorb Required to meet WAC maximum free liquids ids, if present. requirement of 1 percent.

void space (if required) in accordance with Required to meet WAC requirement to WAC. minimize void space.

NE Medical Technologies 11.2-16 Rev. 2

Table 11.2 Waste Methodology for Consolidated Liquids Requirement Basis lect uranium liquid waste and non-uranium Process separately prior to blending.

id wastes separately.

ply hold time to uranium liquid waste. Radioactive decay to achieve solidification product suitable for LSA or Type A packaging.

nsolidate uranium and non-uranium liquid Liquid waste consolidation and processing.

stes into blending tanks.

mple blended wastes after mixing. A representative sample is required to verify maximum uranium concentration is not exceeded and for accurate waste characterization prior to solidification.

idify waste. Use of a solidification agent to ensure final waste form meets requirements. Required to meet WAC maximum free liquids requirement for solidified waste forms (0.5 percent by volume).

it void space. WAC requirement to minimize void space.

ablish dedicated area in the material Solidified waste may require decay post-ging building for decay or shipment processing to meet DOT limits.

solidation.

intain records relative to drums in the Drums may be held to decay to DOT limits.

rage area.

NE Medical Technologies 11.2-17 Rev. 2

Table 11.2 Chemical Composition and Radiological Properties of Liquid Waste Streams (Sheet 1 of 2)

Qualitative Chemical Composition Estimated Annual Radiological Radiological Description (wt/wt) Volume Inventory(1) Properties

[ ]PROP/ECI raction Column [ ]PROP/ECI

[ ]PROP/ECI d Wash [ ]PROP/ECI PROP/ECI

[ ]

>99% H2O raction Column trace H2SO4

[ ]PROP/ECI ter Wash trace UO2SO4

[ ]PROP/ECI

[ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI PROP/ECI PROP/ECI Most fission

] [ ]

products pass

[ ]PROP/ECI through separation

[ ]PROP/ECI [ ]PROP/ECI columns, though PROP/ECI PROP/ECI some are

] [ ] Medium expected to be

[ ]PROP/ECI retained on the

[ ]PROP/ECI columns and then

[ ]PROP/ECI

]PROP/ECI be removed with column washes.

[ ]PROP/ECI ne Recovery [ ]PROP/ECI

[ ]PROP/ECI umn Washes [ ]PROP/ECI PROP/ECI

[ ]

[ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI

]PROP/ECI [ ]PROP/ECI

[ ]PROP/ECI

[ ]PROP/ECI [ ]PROP/ECI PROP/ECI

] [ ]PROP/ECI Fission products remaining in the

[ ]PROP/ECI target solution nt Target [ ]PROP/ECI after useful lifetime

[ ]PROP/ECI High ution [ ]PROP/ECI contribute to a PROP/ECI

[ ] relatively high radiological inventory.

NE Medical Technologies 11.2-18 Rev. 2

(Sheet 2 of 2)

Qualitative Chemical Composition Estimated Annual Radiological Radiological Description (wt/wt) Volume Inventory(1) Properties Liquid collected in uum Transfer the Knockout pot tem Knockout 100% H2O 14 gal. Low generally consists of condensed water vapor.

Laboratory waste 99% H2O is expected to iological 1.0% H2SO4 consist of small 280 gal. Low oratory Waste trace UO2SO4 sample volumes of

[ ]PROP/ECI highly-diluted process fluids.

98% H2O Dependent on ontamination 1.0% H2SO4 2,800 gal. Varies decontamination ste 1.0% UO2SO4 needs.

[ ]PROP/ECI 98% H2O 1.1% NH4OH 0.88% HNO3 trace HCl tichem Fission products trace K3RuCl6 ification Liquid remaining after trace KMnO4 ste including 82 gal. Low majority removed trace -benzoin oxime ary Evaporator in prior MEPS (ABO) densate processing steps.

trace MoO2(ABO)2 trace MoO2 trace NaNO3 trace RhCl3 Process vessel cess Vessel vent system t System condensate 100% H2O 700 gal. Low denser generally consists densate of condensed water vapor.

Radiological inventory relative to other liquid waste streams.

NE Medical Technologies 11.2-19 Rev. 2

ccordance with 10 CFR 20, Subpart H, the respiratory protection program:

• Incorporates process and engineering controls, pursuant to 10 CFR 20.1701, to control the concentration of radioactive material in the air. The design of heating, ventilation, and air conditioning systems is described in Section 9a2.1.

• Implements other controls, pursuant to 10 CFR 20.1702, when it is not practical to apply process or engineering controls to control the concentrations of radioactive material in the air to values below those that define an airborne radioactivity area. Consistent with the as low as reasonably achievable (ALARA) program described in Section 11.1, the respiratory protection program implements increased monitoring and limiting intakes by controlling access, limiting exposure times, and using respiratory protection equipment.

• Implements controls, pursuant to 10 CFR 20.1703, for the use of individual respiratory protection equipment to limit the intake of radioactive material. The respiratory protection program includes evaluation of potential hazards and estimated doses by performing surveys, bioassays, air sampling, or other means as necessary. The program provides protection of personnel from airborne concentrations exceeding the limits of Appendix B to 10 CFR 20 and ensures that respiratory equipment is tested and certified, including testing of respirators for operability before usage. The program ensures that written procedures specify the selection, fitting, issuance, maintenance, testing, training of personnel, monitoring, medical evaluations, and recordkeeping for individual respiratory protection equipment and for specifying when such equipment is to be used. Procedures for the use of individual respiratory protection equipment are revised as applicable when making changes to processes, facility, or equipment. Records are maintained for the respiratory protection program, including training in respirator use and maintenance.

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NRC, 1978. Information Relevant to Ensuring that Occupational Radiation Exposures at lear Power Stations Will be As Low As Is Reasonably Achievable, Regulatory Guide 8.8, ision 3, U.S. Nuclear Regulatory Commission, June 1978.

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NRC, 1992. Monitoring Criteria and Methods to Calculate Occupational Radiation Doses, ulatory Guide 8.34, Revision 0, U.S. Nuclear Regulatory Commission, July 1992.

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NRC, 1999. Instruction Concerning Prenatal Radiation Exposure, Regulatory Guide 8.13, ision 3, U.S. Nuclear Regulatory Commission, June 1999.

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NRC, 2011. Administrative Practices in Radiation Surveys and Monitoring, Regulatory de 8.2, Revision 1, U.S. Nuclear Regulatory Commission, May 2011.

NRC, 2012. Health Physics Surveys During Enriched Uranium-235 Processing and Fuel rication, Regulatory Guide 8.24, Revision 2, U.S. Nuclear Regulatory Commission, June 2.

NRC, 2016. Operating Philosophy for Maintaining Occupational Radiation Exposures As Low s Reasonably Achievable, Regulatory Guide 8.10, Revision 2, U.S. Nuclear Regulatory mmission, August 2016.

NRC, 2018. Instructions for Recording and Reporting Occupational Radiation Exposure Data, ulatory Guide 8.7, Revision 4, U.S. Nuclear Regulatory Commission, April 2018.

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