ML22249A132
| ML22249A132 | |
| Person / Time | |
|---|---|
| Site: | SHINE Medical Technologies |
| Issue date: | 08/31/2022 |
| From: | SHINE Technologies, SHINE Health. Illuminated |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22249A148 | List: |
| References | |
| 2022-SMT-0077 | |
| Download: ML22249A132 (70) | |
Text
ENGINEERED SAFETY FEATURES TABLE OF CONTENTS tion Title Page IRRADIATION FACILITY ENGINEERED SAFETY FEATURES ..................... 6a2.1-1
.1
SUMMARY
DESCRIPTION ............................................................................. 6a2.1-1
.2 DETAILED DESCRIPTIONS ............................................................................ 6a2.2-1 6a2.2.1 CONFINEMENT ............................................................................. 6a2.2-1 6a2.2.2 COMBUSTIBLE GAS MANAGEMENT .......................................... 6a2.2-3
.3 NUCLEAR CRITICALITY SAFETY ................................................................... 6a2.3-1 6a2.3.1 CRITICALITY SAFETY CONTROLS .............................................. 6a2.3-1 6a2.3.2 CRITICALITY ACCIDENT ALARM SYSTEM .................................. 6a2.3-2
.4 REFERENCES ................................................................................................. 6a2.4-1 RADIOISOTOPE PRODUCTION FACILITY ENGINEERED SAFETY FEATURES ........................................................................................................ 6b.1-1 1
SUMMARY
DESCRIPTION ............................................................................... 6b.1-1 2 DETAILED DESCRIPTIONS .............................................................................. 6b.2-1 6b.2.1 CONFINEMENT ............................................................................... 6b.2-1 6b.2.2 PROCESS VESSEL VENT ISOLATION .......................................... 6b.2-3 6b.2.3 COMBUSTIBLE GAS MANAGEMENT ............................................ 6b.2-3 6b.2.4 CHEMICAL PROTECTION .............................................................. 6b.2-4 3 NUCLEAR CRITICALITY SAFETY .................................................................... 6b.3-1 6b.3.1 NUCLEAR CRITICALITY SAFETY PROGRAM ............................... 6b.3-1 6b.3.2 CRITICALITY SAFETY CONTROLS................................................. 6b.3-8 NE Medical Technologies 6-i Rev. 2
ENGINEERED SAFETY FEATURES TABLE OF CONTENTS tion Title Page 6b.3.3 CRITICALITY ACCIDENT ALARM SYSTEM ................................. 6b.3-22 6b.3.4 TECHNICAL SPECIFICATIONS .................................................... 6b.3-23 4 REFERENCES ................................................................................................... 6b.4-1 NE Medical Technologies 6-ii Rev. 2
.1-1 Summary of Engineered Safety Features and Design Basis Accidents Mitigated
.1-2 Comparison of Unmitigated and Mitigated Radiological Doses for Select Irradiation Facility DBAs 1-1 Summary of Engineered Safety Features and Design Basis Accidents Mitigated 1-2 Comparison of Unmitigated and Mitigated Radiological Doses for Select Radioisotope Production Facility DBAs 3-1 Summary of Benchmarks Selected for the SHINE Validation Report 3-2 Area of Applicability Summary NE Medical Technologies 6-iii Rev. 1
.1-1 Irradiation Facility Engineered Safety Features Block Diagram
.2-1 Primary Confinement Boundary
.2-2 Tritium Confinement Boundary
.2-3 Irradiation Facility Combustible Gas Management Functional Block Diagram 1-1 Radioisotope Production Facility Engineered Safety Features Block Diagram 2-1 Supercell Confinement Boundary 2-2 Below Grade Confinement Boundary 2-3 RPF Combustible Gas Management Functional Block Diagram 3-1 Target Solution Staging System Overview 3-2 Radioactive Liquid Waste System Overview 3-3 Molybdenum Extraction and Purification System Overview 3-4 Target Solution Preparation System Overview 3-5 Vacuum Transfer System Overview 3-6 Uranium Receipt and Storage System Overview 3-7 Radioactive Drain System Overview 3-8 Radioactive Liquid Waste Immobilization System Overview NE Medical Technologies 6-iv Rev. 0
onym/Abbreviation Definition S American Nuclear Society SI American National Standards Institute AS criticality accident alarm system P criticality safety program A design basis accident P double contingency principle engineered safety feature FAS engineered safety features actuation system RS facility chemical reagent system O fissionable material operation L grams of uranium per liter PA high efficiency particulate air AC heating, ventilation, and air conditioning S irradiation cell biological shield irradiation facility NE Medical Technologies 6-v Rev. 1
onym/Abbreviation Definition irradiation unit iodine and xenon purification and packaging S quality control and analytical testing laboratories lower flammability limit C minimum accident of concern PS molybdenum extraction and purification system S nitrogen purge system S nuclear criticality safety SE nuclear criticality safety evaluation AS neutron driver assembly system C protective action criteria LS primary closed loop cooling system S production facility biological shield B primary system boundary VS process vessel vent system NE Medical Technologies 6-vi Rev. 1
onym/Abbreviation Definition S radioactive drain system roentgen equivalent man Regulatory Guide WI radioactive liquid waste immobilization WS radioactive liquid waste storage CS radioisotope process facility cooling system F radioisotope production facility Z1 radiological ventilation zone 1 Z1e radiological ventilation zone 1 exhaust subsystem Z1r radiological ventilation zone 1 recirculation subsystem Z2 radiological ventilation zone 2 AS subcritical assembly system M special nuclear material WP solid radioactive waste packaging C structure, system, and component NE Medical Technologies 6-vii Rev. 1
onym/Abbreviation Definition GS TSV off-gas system tritium purification system PS TSV reactivity protection system S target solution preparation system S target solution staging system target solution vessel SS uninterruptible electrical power supply system SS uranium receipt and storage system vacuum transfer system NE Medical Technologies 6-viii Rev. 1
.1
SUMMARY
DESCRIPTION section provides a summary of the engineered safety features (ESFs) installed in the diation facility (IF). Table 6a2.1-1 contains a summary of the ESFs and the IF design basis idents (DBAs) they are designed to mitigate. Table 6a2.1-2 provides unmitigated and gated doses for the public and the worker, with one DBA selected per confinement system, to onstrate the mitigative effects of the confinements. The same methods described in tion 13a2.2 were used to calculate the unmitigated doses, but with a leak path factor of 1 for h the worker and public. A block diagram for the IF ESFs is provided as Figure 6a2.1-1. This k diagram shows the location and basic function of the structures, systems, and components Cs) providing the ESFs in the IF portion of the main production facility.
finement Systems finement systems are provided for protection against the potential release of radioactive erial to the IF and the environment during normal conditions of operation and during and after As. Passive confinement is performed by physical barriers such as concrete or steel ndaries, sealed access plugs, and sealed doors. The confinement systems provide active ation of penetrations during and after certain DBAs that include process piping and heating, tilation, and air conditioning (HVAC) systems penetrating confinement boundaries. The IF s two confinement systems: (1) the primary confinement barrier for the irradiation unit (IU) s, target solution vessel (TSV) off-gas system (TOGS) shielded cells, and the IU cell and GS cell HVAC enclosures; and (2) the tritium confinement barrier for the tritium purification tem (TPS). A detailed description of these confinement systems is provided in section 6a2.2.1.
accidents for which IF confinement systems are credited are described in detail in tion 13a2.1 and listed in Table 6a2.1-1. The accident sequences in the IF which require finement are related to the release of irradiated target solution, radioactive off-gas from GS, or the release of tritium from the TPS.
IF confinement systems remain operational during and following any of the DBAs, including mic events and loss of off-site power. Active components which comprise portions of the finement boundary are designed to fail safe on a loss of control or actuating power and ntain the integrity of the confinement boundary.
sting of the automatic isolation valves included in the confinement boundaries is provided in tion 7.4 and Section 7.5.
mbustible Gas Management combustible gas management systems perform mitigation functions for the primary system ndary (PSB). The combustible gas management system uses the nitrogen purge system PS), PSB piping, and the process vessel vent system (PVVS) to establish an inert gas flow ugh the IUs.
of the functions of the TOGS is to maintain PSB hydrogen concentrations below values ch could result in a hydrogen explosion overpressure capable of rupturing the PSB during NE Medical Technologies 6a2.1-1 Rev. 4
long-term hydrogen gas mitigation during and after an accident, or if TOGS is unavailable, N2PS provides sweep gas to dilute hydrogen within the TSV headspace, TSV dump tank, TOGS piping and maintain the hydrogen gas concentration. The N2PS is described further ubsection 6a2.2.2, and a detailed description is provided in Subsection 9b.6.2.
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Table 6a2.1 Summary of Engineered Safety Features and Design Basis Accidents Mitigated redited Engineered Safety Detailed Description Feature (ESF) Irradiation Facility Design Basis Accidents Mitigated by ESF Subsection Mishandling or Malfunction of Target Solution (Subsection 13a2.1.4)
External Events (Subsection 13a2.1.6) mary Confinement Boundary 6a2.2.1.1 Mishandling or Malfunction of Equipment (Subsection 13a2.1.7)
Facility-Specific Events (Subsection 13a2.1.12)
External Events (Subsection 13a2.1.6)
Unintended Exothermic Chemical Reactions Other Than Detonation tium Confinement Boundary 6a2.2.1.2 (Subsection 13a2.1.10)
Facility-Specific Events (Subsection 13a2.1.12)
Mishandling or Malfunction of Target Solution (Subsection 13a2.1.4)
Loss of Off-Site Power (Subsection 13a2.1.5) mbustible Gas Management Mishandling or Malfunction of Equipment (Subsection 13a2.1.7) 6a2.2.2 Detonation and Deflagration in the Primary System Boundary (Subsection 13a2.1.9)
Insertion of Excess Reactivity (Subsection 13a2.1.2)
Reduction in Cooling (Subsection 13a2.1.3) ne N/A Large Undamped Power Oscillations (Subsection 13a2.1.8)
System Interaction Events (Subsection 13a2.1.11)
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Table 6a2.1 Comparison of Unmitigated and Mitigated Radiological Doses for Select Irradiation Facility DBAs Unmitigated Public Dose (rem) Mitigated Public Dose (rem)
Worker Worker Public Worker Public Worker Limiting Limiting TEDE TEDE TEDE TEDE Representative DBA Organ Organ handling or Malfunction of Target Solution 5.3E+01 3.7E+01 8.6E+02 4.5E-01 1.2E+00 2.3E+01 imary Confinement Boundary - IU Cell) handling or Malfunction of Equipment 5.3E+01 3.7E+01 8.6E+02 7.3E-01 1.9E+00 4.2E+01 imary Confinement Boundary - TOGS Cell) cility-Specific Events 2.5E+01 8.6E+01 8.6E+01 8.0E-01 1.4E+00 1.4E+00 itium Confinement Boundary)
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section provides the details of the design, initiation, and operation of engineered safety ures (ESFs) that are provided to mitigate design basis accidents (DBAs) in the irradiation lity (IF). The IF DBAs, the ESFs required to mitigate the DBAs, and the location of the bases hese determinations are listed in Table 6a2.1-1.
.2.1 CONFINEMENT confinement systems are designed to limit the release of radioactive material to occupied or ontrolled areas during and after DBAs to mitigate the consequences to facility staff, the lic, and the environment. The principal objective of the confinement systems is to protect site personnel, the public, and the environment. The second objective is to minimize the nce on administrative or active engineering controls to provide a confinement system that is imple and fail-safe as reasonably possible. See Figure 6a2.1-1 for an overview of the ctures, systems, and components (SSCs) that provide IF confinement safety functions.
.2.1.1 Primary Confinement Boundary primary confinement boundary consists predominantly of the irradiation unit (IU) cell, the et solution vessel (TSV) off-gas system (TOGS) shielded cell, and the IU cell and TOGS cell ting, ventilation, and air conditioning (HVAC) enclosures. The IU and TOGS shielded cells equipped with removable shield plugs which allow entry into the confined area. The primary finement boundary is primarily passive, and the boundary for each IU is independent from the er IUs. In the event of a DBA that results in a release within the primary confinement ndary, radioactive material is confined primarily by the structural components of the boundary process isolation valves which actuate to isolate the confinement. Gaskets and other
-structural features are used, as necessary, to provide sealing where separate structural ponents meet (e.g., shield plugs). Portions of the confinement are included as part of the diation cell biological shield (ICBS) and their shielding functions are described in tion 4a2.5.
IU cell portion of the primary confinement boundary holds the TSV, TSV dump tank, portions he TOGS, portions of the primary closed loop cooling system (PCLS), associated primary tem boundary (PSB) piping, the light water pool, and the neutron driver. The balance of the GS is located in the TOGS shielded cell. The TSV, TSV dump tank, TOGS, and primary tem piping comprise the PSB which contains the target solution, fission products, and off-gas roducts associated with the irradiation process. The neutron driver is independent from the B and contains an inventory of tritium gas. Figure 6a2.2-1 provides a block diagram of the ary confinement boundary.
umber of process systems penetrate the primary confinement boundary as shown on ure 6a2.2-1. Each piping system capable of excessive leakage that penetrates the primary finement boundary is equipped with one or more isolation valves which serve as active finement components except for the N2PS supply and PVVS connections, which may remain n to provide combustible gas mitigation. Actuation of the isolation valves is controlled by the reactivity protection system (TRPS). A detailed description of the TRPS is provided in tion 7.4.
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s (i.e., radiological ventilation zone 1 recirculating subsystem [RVZ1r]) circulate and cool the within the IU cell and the TOGS cell. Each subsystem is equipped with a cooling coil and high iency particulate air (HEPA) and carbon filters to remove contaminants in the circulated air.
cooling coil is supplied by the radioisotope process facility cooling system (RPCS). The ed loop ventilation units are entirely located in the primary cooling rooms. There are no mally-open external connections between the RVZ1r subsystem and the main RVZ1 system.
etailed discussion of RVZ1r is provided in Section 9a2.1.
PCLS expansion tank has a connection to radiological ventilation zone 1 exhaust subsystem Z1e) which provides a vent path for radiolysis gases produced in the PCLS and light water l, to avoid the buildup of hydrogen gas. The PCLS expansion tank is located in the IU cell but ws air from the TOGS cell atmosphere. A small line connecting the IU cell and TOGS cell ospheres creates a flow path from the IU cell, into the TOGS cell, and out through the PCLS ansion tank to RVZ1e. This flow path normally maintains the cells at a slightly negative ssure. The connection to RVZ1e is equipped with redundant dampers or valves that close on nfinement actuation signal, isolating the cells from RVZ1. A detailed discussion of RVZ1e is vided in Section 9a2.1.
complete listing of variables within the TRPS that can cause the initiation of an IU Cell Safety uation is provided in Subsection 7.4.3.1. The parameters indicating a release of radioactive erial into the primary confinement boundary are high RVZ1e IU cell radiation (indicating a ase of fission products), high tritium purification system (TPS) target chamber supply ssure, and high TPS target chamber exhaust pressure (indicating a release from the neutron er assembly system [NDAS]).
owing an IU Cell Safety Actuation, PSB and primary confinement boundary isolation valves sition to their deenergized (safe) states. The normal flow of materials passes through the zzanine RVZ1 exhaust filter bank before being released to the environment. RVZ filtration is credited in the accident analysis. If sufficient radioactive material reaches the radiation nitors in the RVZ1 exhaust duct, the engineered safety features actuation system (ESFAS) isolate the RVZ building supply and exhaust.
owing cell isolation, three mechanisms by which the primary confinement boundary hanges air with the IF are considered in the accident analysis: pressure-driven flow, counter-ent flow, and barometric breathing. The facility accident analysis models the combined effect hese mechanisms as a minor outflow of radioactive material from the primary confinement ndary directly to the IF and then to the environment under accident conditions. The evaluated ident sequences for which the primary confinement boundary is necessary are listed in le 6a2.1-1 and discussed further in Chapter 13a2.
requirements for the ICBS and TRPS needed for system operability, periodic surveillance, oints, and other specific requirements needed to ensure the functionality of the primary finement boundary are located in the technical specifications.
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tions of the TPS serve as the tritium confinement boundary. The TPS is described in detail in tion 9a2.7. A functional block diagram of the tritium confinement is provided in ure 6a2.2-2.
um in the IF is confined using active and passive features of the TPS. The TPS gloveboxes secondary enclosure cleanup subsystems are credited passive confinement barriers. The gloveboxes enclose TPS process equipment. The process equipment of the secondary losure cleanup subsystem is a credited passive confinement barrier. The TPS gloveboxes maintained at negative pressure relative to the TPS room and have a helium atmosphere.
TPS gloveboxes provide confinement in the event of a breach in the TPS process equipment results in a release of tritium from the isotope separation process equipment.
TPS gloveboxes include isolation valves on the helium supply, the glovebox pressure control aust, and the vacuum/impurity treatment subsystem process vents.
TPS has isolation valves on the process connections to the NDAS target chamber supply exhaust lines. The TPS-NDAS interface lines themselves are part of the credited tritium finement boundary up to the interface with the primary confinement boundary.
en the isolation valves for a process line or glovebox close, the spread of radioactive material mited to the glovebox plus the small amount between the glovebox and its isolation valves.
liquid nitrogen supply and exhaust lines are credited to remain intact during a DBA and the rnal interface between the gloveboxes and nitrogen lines serves as a passive section of the m confinement boundary.
n detection of high TPS exhaust to facility stack tritium concentration or high TPS glovebox m concentration, the ESFAS automatically initiates a TPS isolation. The active components uired to function to maintain the confinement barrier are transitioned to their deenergized e) state by the ESFAS. A description of the ESFAS and a complete listing of the active ponents that transition state with a TPS isolation are provided in Section 7.5.
he event of a break in the process piping within the TPS glovebox, the release of tritium from glovebox is uncontrolled for up to 20 seconds until the isolation valves close. Long-term age and permeation of the confinement barrier result in migration of tritium out of the finement and into the TPS room, IF, and environment. The facility accident analysis siders the effect of this air exchange in its evaluation of radiological consequences. The luated accident sequences for which the tritium confinement boundary is necessary are listed able 6a2.1-1 and further discussed in Chapter 13a2.
requirements for the TPS and ESFAS needed for system operability, periodic surveillance, oints, and other specific requirements needed to ensure the functionality of the tritium finement boundary are located in the technical specifications.
.2.2 COMBUSTIBLE GAS MANAGEMENT rogen gas is produced by radiolysis in the target solution during and after irradiation. During mal operation the concentration of hydrogen gas is monitored and maintained below the er flammability limit (LFL) using the TOGS. The management of combustible gases during NE Medical Technologies 6a2.2-3 Rev. 5
tem, which uses the N2PS, PSB piping, and the process vessel vent system (PVVS) to blish an inert gas flow through the IUs.
principal objective of the combustible gas management system is to prevent the conditions uired for a hydrogen deflagration within the PSB that results in an explosion overpressure eeding the pressure safety limit of the PSB.
N2PS provides back-up nitrogen sweep gas to each IU upon a loss of power or loss of mal sweep gas flow to maintain hydrogen concentrations in these systems below the values ch could result in a hydrogen explosion overpressure capable of rupturing the PSB. A tional block diagram of the combustible gas management system is provided in ure 6a2.2-3.
h pressure nitrogen gas is stored in pressurized vessels which are located in an above-grade forced concrete structure adjacent to the main production facility. On a loss of power or eipt of an appropriate TRPS or ESFAS actuation signal, solenoid-operated isolation valves on nitrogen discharge manifold open and supply nitrogen to the IU cell supply header. The ogen is regulated to a lower pressure and supplied to each TSV dump tank (as necessary) flows through the TSV dump tank, the TSV, and the TOGS equipment and piping before g discharged to the PVVS. The nitrogen flows through the PVVS guard, delay beds, and PA filter before being discharged to the environment via a safety-related vent path. The ogen purge system is described in detail in Section 9b6.2.
complete listing of variables within the TRPS that can cause the initiation of an IU Cell ogen Purge is provided in Subsection 7.4.3.1. These variables indicate a loss of flow or ability ecombine hydrogen by the TOGS. Upon initiation of an IU Cell Nitrogen Purge, active ponents required to function to establish and maintain the N2PS flow path are transitioned to r deenergized (safe) state by the TRPS and the ESFAS. Descriptions of the TRPS and FAS are provided in Sections 7.4 and 7.5, respectively.
ure of the TOGS to manage the combustible gases generated by the subcritical assembly potentially result in a deflagration within the PSB. Hydrogen deflagration within the PSB is an ating event and accident analyzed in Chapter 13a2. The accident sequences for which the bustible gas management system is necessary are listed in Table 6a2.1-1 and discussed in pter 13a2.
capacity of the system is sufficient to provide at least three days of flow to maintain the rogen concentration within acceptable limits with additional margin. The system is flow-nced to ensure that sufficient nitrogen is provided to maintain hydrogen concentrations within eptable limits.
requirements for the TRPS, ESFAS, and N2PS systems needed for system operability, odic surveillance, setpoints, and other specific requirements needed to ensure the tionality of the combustible gas management system are located in the technical cifications.
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NDAS Secondary Enclosure Cleanup Ion Source Supply Supply NDAS Secondary Target Chamber Enclosure Cleanup Supply Return Vacuum/Impurity Target Chamber Treatment System Exhaust PVVS Irradiation Unit Cell TOGS Cell RPCS Supply NDAS Cooling Supply (x2) Neutron TOGS TOGS RPCS Return Driver Components Components NDAS Cooling Assembly Return (x2)
TOGS Gas Supply Air Inlet RVZ1e TOGS Vacuum Subcritical Supply Assembly PCLS Supply RVZ1r RVZ1r PCLS Components PCLS Return N2PS Nitrogen RPCS Supply Supply Cooling Water RPCS Return TSV Fill Process Boundary RVZ1e Makeup Air Confinement Dump Tank Boundary Return NE Medical Technologies 6a2.2-5 Rev. 5
NE Medical Technologies 6a2.2-6 Rev. 5 Nit r ogen IF Nit r ogen Irr adiat ion PVVS PVVS Delay PVVS HEPA PVVS Alt ernate Gas Header Unit s (8) Guar d Beds Beds Filter Ven t Pat h St or age NE Medical Technologies 6a2.2-7 Rev. 5
NE maintains a nuclear criticality safety program (CSP) that complies with applicable erican National Standards Institute/American Nuclear Society (ANSI/ANS) standards, as orsed by Regulatory Guide (RG) 3.71, Revision 3, Nuclear Criticality Safety Standards for ls and Material Facilities (USNRC, 2018). A description of the CSP is provided in tion 6b.3.
, handling, and storage of fissile material in the irradiation facility (IF) is evaluated in ordance with the CSP, with the exception of the target solution vessel (TSV).
.3.1 CRITICALITY SAFETY CONTROLS eria used to select controls and the use of controlled parameters are described in tion 6b.3.2.
.3.1.1 Subcritical Assembly System etailed description of the subcritical assembly system (SCAS) is provided in Section 4a2.2.
system is designed to maintain fissile material in a subcritical state during irradiation and to ly store the target solution following irradiation in the TSV dump tank.
icality Safety Basis nuclear criticality safety evaluation (NCSE) for the SCAS shows that the evaluated sections he process will remain subcritical under normal and credible abnormal conditions. The TSV is igned to operate at a higher keff for the production of medical isotopes and is not considered art of the NCSE. The effects of reactivity changes in the SCAS are provided in sections 4a2.6.3.3 and 4a2.6.3.4.
remaining portions of the SCAS are safe-by-design. The TSV dump tank is shown to remain er the upper subcritical limit under the most reactive credible conditions of concentration, ection, and corrosion. Piping which contains fissile solutions between the TSV and the TSV p tank is shown to be within the evaluated single parameters limits.
.3.1.2 Target Solution Vessel Off-Gas System etailed description of the TSV off-gas system (TOGS) is provided in Section 4a2.8. The major ponents of the system are condenser demisters, a zeolite bed, blowers, hydrogen ombiners, recombiner condensers, a recombiner demister, and a vacuum tank. Components OGS are located in the irradiation unit (IU) cell and the adjacent TOGS cell. Components in IU cell are the vacuum tank, condenser demisters, recombiner demister, and associated ng. The remaining components are arranged on a skid in the TOGS cell.
system is designed to maintain the hydrogen concentration in the primary system boundary w the lower flammability limit by circulating gas from the TSV during irradiation and from the dump tank during cool-down through its demisters, zeolite bed, and recombiner. The TOGS rates at slightly negative pressure. Under normal conditions, the system does not contain ificant quantities of fissile material.
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NCSE for the TOGS shows that the entire system will remain subcritical under normal and dible abnormal conditions.
er abnormal conditions, it is credible that significant quantities of fissile material enter the GS. Each of the individual components located in the IU cell and the skid arrangement of ponents in the TOGS cell has favorable geometry under the most reactive credible ditions.
itional criticality safety considerations of the TOGS are provided in Subsection 4a2.8.5.1.
.3.2 CRITICALITY ACCIDENT ALARM SYSTEM IF utilizes a criticality accident alarm system (CAAS) to detect a criticality event in the areas hich special nuclear material is used, handled, or stored outside of the IU cells. Coverage of cial nuclear material storage in the TSV dump tanks and interconnecting piping is provided by neutron flux detection system (NFDS) and level instrumentation in the TSV dump tank, which vides indication of abnormal conditions in the IU cells.
escription of the CAAS is provided in Subsection 6b.3.3.
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NRC, 2018. Nuclear Criticality Safety Standards for Fuels and Material Facilities, Regulatory de 3.71, Revision 3, 2018.
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1
SUMMARY
DESCRIPTION section provides a summary of the engineered safety features (ESFs) installed in the oisotope production facility (RPF). Table 6b.1-1 contains a summary of the ESFs and the F design basis accidents (DBAs) they are designed to mitigate. Table 6b.1-2 provides itigated and mitigated doses for the public and the worker, with one DBA selected per finement system, to demonstrate the mitigative effects of the confinements. The same hods described in Section 13a2.2 were used to calculate the unmitigated doses, but with a path factor of 1 for both the worker and public. A block diagram for the RPF ESFs is vided as Figure 6b.1-1. This block diagram shows the location and basic function of the cture, system and components (SSCs) providing the ESFs in the RPF portion of the main duction facility.
finement Systems finement systems provide active and passive protection against the potential release of oactive material to the environment during normal conditions of operations and during and r a DBA. Passive confinement is performed by physical barriers such as concrete or steel ndaries, sealed access plugs, and sealed doors. The confinement systems provide active ation of penetrations that include process piping and heating, ventilation, and air conditioning AC) systems penetrating confinement boundaries during and after certain DBAs. The cess confinement boundary includes two areas: (1) the supercell confinement, which includes extraction, purification, and packaging hot cells, the iodine and xenon purification and kaging cell, and the process vessel ventilation system (PVVS) hot cell; and (2) the below de confinement, which confines the PVVS delay beds, the target solution hold, storage, and te tanks, the pipe trench and valve pits, and the waste processing tanks. A detailed cription of the confinement systems is provided in Subsection 6b.2.1.
accidents for which confinement is credited are described in detail in Section 13b.1 and d in Table 6b.1-1. The accident sequences in the RPF which require confinement are related he release of radioactive liquids and gases from irradiated target solution, waste streams, or cessing streams.
RPF confinement systems remain operational during and following any of the DBAs, uding seismic events and loss of off-site power. Active components which comprise portions he confinement boundaries are designed to fail safe on a loss of actuating power and ntain the integrity of the confinement boundaries.
sting of the automatic isolation valves included in the confinement boundaries is provided in tion 7.4 and Section 7.5.
cess Vessel Ventilation System Isolation PVVS is equipped with isolation valves that actuate to confine and extinguish fires, which y occur in the PVVS carbon guard beds or carbon delay beds. These isolation functions are cribed in detail in Subsection 6b.2.2. The PVVS is described in detail in Section 9b.6.
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combustible gas management system performs mitigation functions for the RPF systems components that may potentially contain hydrogen gas from radiolysis. The PVVS maintains hydrogen concentration in these areas below the lower flammability limit (LFL) during normal rating conditions. The PVVS is described in detail in Section 9b.6.
hydrogen gas mitigation during and after an accident, or if the PVVS is unavailable, the ogen purge system (N2PS) provides sweep gas to dilute the RPF tanks to maintain the rogen concentration below the LFL. The N2PS is described further in Subsection 6b.2.3, and etailed description is provided in Section 9b.6.
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Table 6b.1 Summary of Engineered Safety Features and Design Basis Accidents Mitigated Credited Engineered Safety Radioisotope Production Facility Design Basis Accidents Detailed Description Feature (ESF) Mitigated by ESF Subsection percell Confinement Critical Equipment Malfunction (Subsection 13b.2.4) 6b.2.1.1 low Grade Confinement Critical Equipment Malfunction (Subsection 13b.2.4) 6b.2.1.2 ocess Vessel Ventilation Isolation Radioisotope Production Facility Fire (Subsection 13b.2.6) 6b.2.2 Loss of Electrical Power (Subsection 13b.2.2) mbustible Gas Management 6b.2.3 Critical Equipment Malfunction (Subsection 13b.2.4) ne External Events (Subsection 13b.2.3) N/A NE Medical Technologies 6b.1-3 Rev. 4
Table 6b.1 Comparison of Unmitigated and Mitigated Radiological Doses for Select Radioisotope Production Facility DBAs Unmitigated Public Dose (rem) Mitigated Public Dose (rem)
Worker Worker Public Worker Public Worker Limiting Limiting TEDE TEDE TEDE TEDE Representative DBA Organ Organ tical Equipment Malfunction 8.0E+00 1.7E+01 2.5E+02 4.2E-02 7.6E-02 5.2E-01 ocess Confinement Boundary - Supercell) tical Equipment Malfunction 8.0E+00 1.7E+01 2.4E+02 2.4E-02 4.2E-02 2.9E-01 ocess Confinement Boundary - Below Grade)
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ESFAS Confinement Isolation Signal Active I & C Systems Supercell Ventilation Isolation (Supply and Exhaust)
Supercell - RPF General Area N2PS Valves Facility-Wide Hot cell Isolation Valves Delay Bed Isolation Valves Supercell - RPF General Area Delay Bed Vault PVVS Safety Exhaust Valves Facility Mezzanine VTS Safety Actuation RVZ1 Isolation Dampers Supercell - RPF General Area (RCA Boundary)
Facility Mezzanine Active Components N2PS Piping Production Facility Biological Facility-wide Supercell Confinement Shield Components Supercell - RPF General Area RPF PVVS Carbon Beds and Piping Supercell and Delay Bed Vaults Passive Components Combustible Gas Process Confinement PVVS Process Isolation Management Boundary NE Medical Technologies 6b.1-5 Rev. 4
section provides the details of the design, initiation, and operation of engineered safety ures (ESFs) that are provided to mitigate the design basis accidents (DBAs) in the oisotope production facility (RPF). The RPF DBAs, the ESFs required to mitigate the DBAs, the location of the bases for these determinations are listed in Table 6b.1-1.
2.1 CONFINEMENT confinement systems are designed to limit the release of radioactive material to uncontrolled as during and after DBAs to mitigate the consequences to workers, the public, and the ironment. The principal objective of the confinement systems is to protect on-site personnel, public, and the environment. The second objective is to minimize the reliance on inistrative or active engineering controls to provide a confinement system that is as simple fail-safe as reasonably possible. Figure 6b.1-1 provides an overview of the structures, tems, and components that provide RPF confinement safety functions.
sting of the automatic isolation valves included in the confinement boundaries is in tion 7.5.
2.1.1 Supercell Confinement supercell is a set of hot cells in which isotope extraction, purification, and packaging is ormed, and gaseous waste is handled. The supercell provides shielding and confinement to ect the workers, members of the public, and the environment by confining the airborne oactive materials during normal operation and in the event of a release. The supercell udes features to allow the import of target solution, consumables, and process equipment; sfer between adjacent cells; and export of final products, waste, spent process equipment, samples for analysis in the laboratory. The export features of the supercell are integrated the confinement boundary to allow export operations while maintaining confinement. The ercell is described in detail in Section 4b.2.
ure 6b.2-1 provides a block diagram of the supercell confinement boundary. The process port loop represents the MEPS hot water loop.
hot cells are fitted with stainless steel boxes for confinement of materials and process ipment. The radiological ventilation zone 1 (RVZ1) draws air through each individual finement box, drawing air from the general RPF area, to maintain negative pressure inside confinement, minimizing release of radiological material to the facility. Filters and carbon orbers on the ventilation inlets and outlets control release of radioactive material to workers the public. RVZ1 is described in Section 9a2.1.
supercell ventilation exhaust ductwork is fitted with radiation monitoring instrumentation to ect off-normal releases to the confinement boxes. Upon indication of a release exceeding oints, isolation dampers or valves on both the inlet and outlet ducts isolate the hot cells from ventilation system. Additionally, the actuation signal closes isolation valves on the ybdenum extraction and purification system (MEPS) heating loops and conducts a vacuum sfer system (VTS) safety actuation. As part of VTS safety actuation, connections to the ercell from the facility chemical reagent system (FCRS) skid isolate, closing the MEPS and ne and xenon purification and packaging (IXP) supply valves as described in NE Medical Technologies 6b.2-1 Rev. 4
he ESFAS is provided in Section 7.5.
taminated air is confined to the supercell by the confinement boxes, the ventilation exhaust pers or valves, and the process isolation valves.
facility accident analysis considers the effect of air exchange from the confinement to the eral areas in its evaluation of radiological consequences. This outflow of radioactive material the confined area to the RPF and the environment is based on the leak rate of the supercell.
fficient radioactive material reaches the radiation monitors in the RVZ1 exhaust duct, ESFAS isolate the RVZ building supply and exhaust. The evaluated accident sequence for which the ercell is necessary is listed in Table 6b.1-1 and discussed further in Section 13b.2.
requirements needed for supercell confinement system operability, periodic surveillance, oints, and other specific requirements needed to ensure the functionality of the supercell are ted in technical specifications.
2.1.2 Below Grade Confinement below grade confinement provides a barrier to protect workers, members of the public, and environment by reducing radiation exposure. The below grade confinement includes the RPF vaults, valve pits, pipe trench, and carbon delay bed vault. Portions of the below grade finement are identified as part of the production facility biological shield (PFBS), which is cribed in detail in Section 4b.2.
ure 6b.2-2 provides a block diagram of the below grade confinement.
he event of a DBA that results in a release within the process confinement boundary, oactive material is confined primarily by the structural components of the boundary. Gaskets other non-structural features are used, as necessary, to provide sealing where components et (e.g., shield plugs and inspection ports). Each vault is equipped with a concrete cover plug icated in multiple sections with one or more inspection ports which allow remote inspection of confined areas without personnel access. Each valve pit is equipped with a concrete cover fabricated in multiple sections with one inspection port. The pipe trench is equipped with crete cover plugs fabricated in multiple sections with some having inspection ports. The pipe ch, vaults, and valve pits with equipment containing fissile material are equipped with drip s and drains to the radioactive drain system (RDS).
below grade confinement is primarily passive. Most process piping that passes through the finement boundary is entering or exiting another confinement boundary. Process piping for iliary systems entering the boundary from outside confinement is provided with appropriate nual or automatic isolation capabilities. The confinement boundary includes cover plugs and ection ports for access to the confined areas. Contaminated air is confined to the vaults, e pits, and pipe trench.
facility accident analysis considers the effect of air exchange from the confinement to the eral areas in its evaluation of radiological consequences. Three mechanisms by which the cess confinement boundary exchanges air with the RPF are considered: pressure-driven
, counter-current flow, and barometric breathing. The combined effect of these mechanisms NE Medical Technologies 6b.2-2 Rev. 4
Z1 exhaust duct, ESFAS will isolate the RVZ building supply and exhaust. The evaluated ident sequence for which the process confinement boundary is necessary is listed in le 6b.1-1 and discussed further in Section 13b.2.
requirements needed for process confinement boundary system operability, periodic veillance, setpoints, and other specific requirements needed to ensure the functionality of the cess confinement boundary are located in technical specifications.
2.2 PROCESS VESSEL VENT ISOLATION process vessel vent system (PVVS) captures or provides holdup for radioactive particulates, ne, and noble gases generated within the RPF and primary system boundary. The system ws air from the process vessels through a series of processing components which remove the oactive components by condensation, acid adsorption, mechanical filtration with high-iency particulate air (HEPA) filters, and adsorption in carbon beds. Two sets of carbon beds used; the guard beds located in a below-grade, shielded vault, and the delay beds located in carbon delay bed vault.
s may occur in the carbon guard and delay beds which result in the release of radioactive erial into the downstream PVVS system, which leads to the facility ventilation system and the ironment. The PVVS guard and delay beds are equipped with isolation valves that isolate the cted guard bed or delay bed from the system and extinguish the fire. The isolation valves also ve to prevent the release of radioactive material to the environment. Delay beds 1, 2, and 3 equipped with sensors to detect fires which provide indication to ESFAS. The isolation valves he affected delay bed close upon receipt of the actuation signal from ESFAS, preventing fire pagation to the downstream beds. The redundancy in the beds and the ability to isolate vidual beds allows the PVVS to continue to operate following an isolation.
evaluated accident sequence for which the PVVS isolation is necessary is listed in le 6b.1-1 and discussed further in Section 13b.2.
requirements to be specified in the technical specifications for system operability, periodic veillance, setpoints, and other specific requirements needed to ensure the functionality of the VS isolations are located in Section 7.5 and Section 9.6, which describes the ESFAS and the VS, respectively.
2.3 COMBUSTIBLE GAS MANAGEMENT rogen gas is produced by radiolysis in the target solution during and after irradiation. During mal operation, the PVVS removes radiolytic hydrogen and radioactive gases generated within RPF and primary system boundary. The PVVS is described in detail in Section 9b.6. If PVVS omes unavailable, the buildup of hydrogen gas is limited using the combustible gas nagement system, which uses the nitrogen purge system (N2PS), process system piping, and PVVS to establish an inert gas flow through the process vessels.
principle objective of the combustible gas management system is to prevent the conditions uired for a hydrogen deflagration in the gas spaces in the RPF process tanks.
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ram of the combustible gas management system is shown in Figure 6b.2-3.
h pressure nitrogen gas is stored in pressurized vessels which are located in an above-grade forced concrete structure adjacent to the main production facility. On a loss of power or eipt of an ESFAS actuation signal, isolation valves on the radiological ventilation zone 2 Z2) air supply to PVVS shut and isolation valves on the N2PS discharge manifold open, asing nitrogen into the RPF N2PS distribution piping. The nitrogen gas flows through the RPF ipment and into the PVVS process piping. On loss of power, the discharged gases bypass passive filtration skid equipment (i.e., the PVVS condenser, PVVS reheater, acid adsorber, PA filter, and guard bed) before being discharged to the alternate vent path in the PVVS and stack. On receipt of an ESFAS actuation signal other than loss of power, the discharged es flow through the PVVS passive filtration skid equipment and the passive filtration skid ipment bypass line before being discharged to the alternate vent path in the PVVS and the
- k. The N2PS is described in detail in Section 9b.6.
complete listing of variables within the ESFAS that can cause the initiation of an RPF ogen Purge is provided in Subsection 7.5.3.1. These variables indicate a loss of flow. The ve components required to function to initiate the RPF Nitrogen Purge are actuated by the FAS. A detailed description of the ESFAS is provided in Section 7.5.
combustible gas management system prevents deflagrations and detonations in RPF cess tanks which could lead to a tank or pipe failure and cause a target solution spill inside process confinement boundary. The accident sequences for which the combustible gas nagement system is necessary are listed in Table 6b.1-1 and discussed in Chapter 13a2.
requirements needed for PVVS system operability, periodic surveillance, setpoints, and er specific requirements needed to ensure the functionality of the combustible gas nagement system are located in technical specifications.
2.4 CHEMICAL PROTECTION chemical dose analysis is provided in Section 13b.3 and has shown that no potential mical release exceeds the established acceptance limits. As described in Section 13b.3, finement barriers (i.e., supercell, gloveboxes, subgrade vaults) are credited for mitigation of mical dose consequences. The URSS uranium storage racks are seismically qualified to ntain their structure and position during seismic events to limit the material at risk for uranium e accidents.
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VZ2 Ventilation (Supply) RVZ1 Ventilation (Exhaust)
RS Reagent Skid Extraction Process rocess Support Purification Packaging Vessel Vent and IXP Adjacent Loop Inlet Hot Cells Hot Cells System Hot Hot Cells Confinement Cell rocess Support Process Loop Outlet Boundary Supercell Confinement Boundary Confinement Boundary Vacuum Transfer Cell Drains System Process Process Process Process Vessel Vent Vessel Vent Piping Piping System System Below Grade Confinement Boundary NE Medical Technologies 6b.2-5 Rev. 4
N2PS TSPS Valve Pits Pipe Trench Tank Vaults Carbon Delay Bed Vaults Process Tanks PVVS Delay Process Valves Process Piping PVVS Beds PVVS Guard Beds Confinement Irradiation Hot Cells Units Primary Confinement Confinement Boundary Supercell Confinement Boundary NE Medical Technologies 6b.2-6 Rev. 4
Nit r ogen RPF PVVS PVVS PVVS RPF Nit rogen PVVS Alt ernate Gas Pro cess Guar d Delay HEPA Header Ven t Pat h St or age Tanks Beds Beds Filter NE Medical Technologies 6b.2-7 Rev. 4
NE maintains a nuclear criticality safety program (CSP) that complies with applicable erican National Standards Institute/American Nuclear Society (ANSI/ANS) standards, as orsed by Regulatory Guide (RG) 3.71, Revision 3, Nuclear Criticality Safety Standards for ls and Material Facilities (USNRC, 2018). The CSP meets the following criticality safety uirements of 10 CFR 70:
- The criticality accident requirements of 10 CFR 70.24;
- The criticality reporting requirements of 10 CFR 70.50, 10 CFR 70.52, and the SHINE-specific reporting requirements which meet the intent of 10 CFR 70, Appendix A, as described in the technical specifications;
- Application of 10 CFR 70.61(b) to criticality accidents, considering such accidents as high-consequence events; and
- Application of 10 CFR 70.61(d), ensuring that nuclear processes are subcritical under normal and credible abnormal conditions, including use of an approved margin of subcriticality and the use of preventative controls as the primary means of protection.
3.1 NUCLEAR CRITICALITY SAFETY PROGRAM CSP is administered through a written nuclear criticality safety (NCS) policy and program cription, with an additional program description for NCS training and qualification. The CSP is cuted by qualified NCS staff using written procedures. The program description and written cedures are formally controlled through the SHINE document control procedure.
goal of the CSP is to ensure that workers, the public, and the environment are protected the consequences of a nuclear criticality event. In order to accomplish this goal, all cticable measures are implemented to prevent an inadvertent criticality from occurring. The P also contains provisions necessary to mitigate the consequences (i.e., criticality accident m system [CAAS] and emergency response activities) should an inadvertent criticality occur.
3.1.1 Nuclear Criticality Safety Program Organization SHINE Chief Executive Officer holds overall responsibility for the CSP. The Safety Analysis nager is the Responsible Manager for the CSP and may delegate administrative authority to NCS Lead.
NE facility management holds the following responsibilities with respect to the CSP:
- Formulate and maintain the NCS policy and ensure that personnel involved in fissionable material operations (FMOs) are informed of the policy.
- Assign responsibility and delegate commensurate authority to implement the criticality safety policy and program.
- Ensure that everyone, regardless of position, is made aware of their responsibilities for implementing the requirements of the CSP.
- Ensure that appropriately trained and qualified NCS staff are available to provide technical guidance appropriate for the FMOs performed at the SHINE facility.
- Establish and maintain a training and qualification program for NCS staff.
- Establish a method to monitor the CSP.
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- Establish and maintain a configuration management program that identifies and controls changes to facility, equipment, and processes important to NCS.
- Establish a process for developing, reviewing, supplementing, and revising operating procedures important to NCS.
- Require that activities involving fissile material are conducted using approved written procedures and for situations for which existing procedures are inadequate or do not exist, require personnel to take no action until the NCS staff has evaluated the situation and provided recovery instructions.
- Encourage the use of stop-work authority and reporting of defective conditions.
ervisors responsible for FMOs hold the following responsibilities with respect to the CSP:
- Accept responsibility for the safety of operations under their control.
- Be knowledgeable in those aspects of NCS relevant to operations under their control.
- Ensure that NCS training is provided to the personnel under their supervision.
- Personnel under their supervision must understand procedures, limits, controls, and other NCS considerations such that personnel can be expected to perform their functions without undue risk.
- Maintain records of training activities and verification of personnel understanding.
- Develop or participate in the development of procedures applicable to the operations under their control. Maintain these procedures to reflect changes in operations as a continuous supervisory responsibility.
- Verify compliance with NCS specifications for new or modified equipment before its use.
Verification may be based on inspection reports or other features of the quality assurance program.
- Be responsible for the inspection, testing, and maintenance of engineered controls.
- Require conformance with good safety practices, including unambiguous identification of fissile materials and good housekeeping.
S staff hold the following responsibilities with respect to the CSP:
- Provide technical guidance for the design of equipment and processes and for the development of operating procedures.
- Maintain familiarity with current developments in NCS standards and guides and other nuclear criticality information.
- Maintain familiarity with operations within the SHINE facility requiring NCS controls. This shall be accomplished by individual staff members maintaining familiarity with operations for which they provide guidance.
- Assist supervisors, on request, in training personnel.
- Participate in the development of the NCS training program.
- Review facility non-conformances that have the potential to impact NCS and provide appropriate response recommendations to violations or deficiencies.
NE's NCS staff consists of an NCS Lead and one or more NCS Engineers, at least one of m shall be qualified at the Senior level, and any number of individuals identified as NCS NE Medical Technologies 6b.3-2 Rev. 5
lysts, whose function is to perform and document NCS calculations in support of NCS luation (NCSE) development. NCS staff are kept administratively separate from operations to extent practicable.
3.1.2 Nuclear Criticality Safety Staff Qualifications minimum qualification entry requirements for NCS staff are:
NCS Analyst: Baccalaureate degree in science or engineering from an accredited college or university, or at least five years of directly applicable experience, or an equivalent combination of education and experience.
NCS Engineer: Same as for an NCS Analyst Senior NCS Engineer: Current qualifications as an NCS Engineer, plus three years of experience as an NCS Engineer S qualifications use a tiered approach, with three qualification levels for NCS Staff and cific functional area qualifications for Fissile Material Handlers. The specific training uirements are taken from ANSI/ANS-8.26-2007 (R2016), Criticality Safety Engineer Training Qualification Program (ANSI/ANS, 2007a). SHINE uses qualification cards to record an vidual's progress towards qualification. Qualification cards list the necessary knowledge and ormance requirements for NCS staff and provide a record of completion for qualification vities. Assignment of personnel for qualification is made by an engineering manager.
ntenance of qualifications is required for NCS staff.
lifications granted by external organizations may be recognized based on verification and pletion of SHINE facility-specific portions of the appropriate qualification card. Experience in S may be used to exempt individual training and qualification requirements. Where erience is used for exemptions, appropriate documentation is attached to the qualification d and retained. Facility familiarity and walk-through requirements may not be exempted and required in addition to recognition of externally-completed qualifications. Maintenance of lifications is required for NCS staff.
3.1.3 Use of National Consensus Standards CSP commits to the requirements of the following national consensus standards, subject to clarifications and exceptions identified in RG 3.71, with certain SHINE-specific limitations cribed below:
Standards endorsed without clarifications or exceptions by the Nuclear Regulatory Commission (NRC) in RG 3.71:
- ANSI/ANS-8.6-1983 (R2017), Safety in Conducting Subcritical Neutron-Multiplication Measurements In Situ (ANSI/ANS, 1983)
- ANSI/ANS-8.7-1998 (R2017), Nuclear Criticality Safety in the Storage of Fissile Materials (ANSI/ANS, 1998)
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- ANSI/ANS-8.20-1991 (R2015), Nuclear Criticality Safety Training (ANSI/ANS, 1991)
- ANSI/ANS-8.22-1997 (R2016), Nuclear Criticality Safety Based on Limiting and Controlling Moderators (ANSI/ANS, 1997a)
- ANSI/ANS-8.26-2007 (R2016), Criticality Safety Engineer Training and Qualification Program Standards endorsed in RG 3.71 with clarifications or exceptions:
- ANSI/ANS-8.1-2014, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors (ANSI/ANS, 2014b)
The clarification applied to this standard is related to subcritical limits for plutonium isotopes and is not applicable to the SHINE facility.
- ANSI/ANS-8.3-1997 (R2017), Criticality Accident Alarm System (ANSI/ANS, 1997b)
The clarifications and exceptions applied to this standard are applicable to the SHINE facility.
- ANSI/ANS-8.23-2007 (R2012), Nuclear Criticality Accident Emergency Planning and Response (ANSI/ANS, 2007b)
The clarification applied to this standard is applicable to the SHINE facility.
- ANSI/ANS-8.24-2017, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations (ANSI/ANS, 2017)
The clarifications applied to this standard are applicable to the SHINE facility.
The following ANSI/ANS Series 8 Standards are not used by the SHINE CSP. For each standard, the basis for non-implementation is provided:
- ANSI/ANS-8.5-1996 (R2017), Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material.
Borosilicate-glass Raschig rings are not used in the SHINE facility.
- ANSI/ANS-8.10-2015, Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement (ANSI/ANS, 2015).
SHINE does not apply the criteria provided in this standard for determining the adequacy of shielding and confinement.
- ANSI/ANS-8.12-1987 (R2016), Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors.
Plutonium is not used as a fuel component at SHINE. Only small quantities are present due to burnup.
- ANSI/ANS-8.14-2004 (R2016), Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors.
SHINE does not use soluble neutron absorbers for control of criticality.
- ANSI/ANS-8.15-2014, Nuclear Criticality Control of Selected Actinide Nuclides.
SHINE does not conduct operations with non-negligible quantities of the selected actinide nuclides.
- ANSI/ANS-8.17-2004 (R2014), Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
SHINE does not handle, store, or transport LWR fuel rods or units.
- ANSI/ANS-8.21-1995 (R2011), Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors (ANSI/ANS, 1995)
SHINE does not use fixed neutron absorbers for control of criticality.
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3.1.4 Nuclear Criticality Safety Evaluations SEs are conducted for each FMO to ensure that under normal and credible abnormal ditions, all nuclear processes remain subcritical with an approved margin of subcriticality for ty. An FMO is any process or system that has the potential to contain more than 250 g of
-exempt fissile material. This limit is selected based on one-half of the single parameter mass t for uranium-233 identified in ANSI/ANS-8.1-2014. For the purposes of application of this
, all fissionable isotopes in the process or system are considered to be fissile.
mpt fissile material is defined as special nuclear material (SNM) that meets the requirements classification as fissile nuclear material as specified in 10 CFR 71.15. The limits specified in CFR 71.15 are derived for use in nuclear material transport and long-term storage and are eptably conservative. When 10 CFR 71.15 is invoked to exempt a process or system, the SE must show that there are no credible means of changing the physical composition or figuration of the material.
S limits are derived based on assuming optimum or most-reactive credible parameter values ss specific controls are implemented to limit parameters to a particular range. If less-than-mum values are used, the basis for use is included in the NCSE. Operating limits which take cess variability and uncertainty into account are used to ensure NCS limits are unlikely to be eeded. Controls used to enforce safety and operating limits are specified in the NCSEs.
NCSEs are conducted using appropriate hazard evaluation techniques, including "What-if,"
at-if Checklist," and Event Tree Analysis, to determine potential scenarios which could result n inadvertent criticality event. Process hazards evaluations are referenced to identify itional potential scenarios that have been determined to have potential criticality safety lications (e.g. chemical safety, fire, radiological events). The identified scenarios are ened based on a qualitative determination of likelihood and those events which are deemed e credible are evaluated for appropriate control selection. For the purposes of NCSEs, cality events are always considered to be "high" consequence, with a strict emphasis on ction of controls to prevent criticality. Where the double contingency principle (DCP) is ployed, the NCSE contains a description of its implementation.
NCS limits used in the evaluations are derived from industry-accepted and peer-reviewed rences, including ANS standards; from hand calculations using industry-accepted and peer-ewed techniques, such as solid-angle or surface density calculation; or from computational hods. In cases where hand calculations are used, each technique is used consistent with any tations.
3.1.5 Computational System Validation ere computational methods are employed, the computational system is verified and validated g the guidance in NUREG/CR-6698 (USNRC, 2001).
ritten validation report for the computational systems used for NCS calculations is umented and maintained in accordance with the SHINE document control process. The dation process was performed using Monte Carlo n-Particle (MCNP) software, NE Medical Technologies 6b.3-5 Rev. 5
ducted following any changes to the hardware or operating system.
validation report uses benchmarks from the Handbook of the International Criticality Safety chmark Evaluation Project (ICBEP). Benchmarks were selected for evaluation based on their ilarity to the SHINE solution system, as no plant-specific benchmark experiments are ilable. The fissile material, enrichment, chemical form, range of concentration, and reflector erials were considered in the selection of benchmarks. The selected benchmarks series, ber of cases selected from each benchmark series, and a description of each physical tem is provided in Table 6b.3-1. A summary of the area of applicability covered by the dation report is provided in Table 6b.3-2.
bias and bias uncertainty were calculated using the methodology described in REG/CR-6698. The benchmark data were tested using a modified Shapiro-Wilk test for mality and were determined to be normally distributed. A single-sided tolerance limit approach used to determine the bias uncertainty. The upper subcritical limit is the difference between y and the sum of the bias (zero, because a positive bias was determined), the bias ertainty, and the subcritical margin.
margin of subcriticality used for SHINE solution processes is 0.06. A subcritical margin of 5 was conservatively selected based on the quantity and quality of the selected benchmarks.
additional subcritical margin of 0.01 is applied to provide additional conservatism to account he limited number of experimental benchmarks specific to uranyl sulfate systems. NCSEs ure that the evaluated processes fall within the range of the validated computational system.
validation range may be extended beyond the range of the benchmark data using additional critical margin or bias trending analysis to ensure that the existing subcritical margin is ropriate. Where extrapolation or wide interpolations are used to extend the validation range, recommendations of NUREG/CR-6698 are used. When a positive bias is encountered, it is to 0 for the purposes of calculating subcritical limits, and data outliers are only rejected based nconsistency with known physical behavior; statistical rejection methods for outliers are not
- d. NCS limits are selected to incorporate appropriate margins to protect against uncertainty in cess variables and to prevent a limit being accidently exceeded. Allowances for uncertainty in methods, data, and bias are included in the selected limits. Studies are conducted to elate the effects of changing one controlled parameter on other controlled parameters, such o evaluate compliance with the DCP.
S program documentation, evaluations, and calculations are maintained in accordance with SHINE records management system. Equipment characteristics relied on to maintain NCS ts are identified as NCS controls and are maintained by the SHINE configuration nagement system.
cess or design changes that could affect NCS limits or controls are evaluated using the lity change process requirements of 10 CFR 50.59. Prior to implementing the change, the SE is reviewed and updated if needed to determine that the entire process will be subcritical er both normal and credible accident scenarios.
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upport of SHINE's CSP, a two-tiered NCS training program is established and maintained.
first-tier training program includes the Program Content identified in ANSI/ANS-8.20-1991 015), and is directed toward those who manage, work in, or work near areas where the ential exists for a criticality accident. The second-tier training is specific to NCS staff. NCS f training meets the requirements identified in ANSI/ANS-8.26-2007 (R2016). Both tiers of S training include procedural compliance, stop-work authority, response to criticality alarms, reporting of defective conditions.
3.1.7 Criticality Safety Program Oversight rations are reviewed at least annually to verify that procedures are being followed and that cess conditions have not been altered to affect the NCSE. NCS staff conduct walkthroughs of lity processes and procedures as part of the annual operational review. These reviews are ducted, in consultation with operating personnel, by individuals who are knowledgeable in S and who, to the extent practicable, are not immediately responsible for the operation, and documented. Active procedures are reviewed periodically by supervisors.
NCS Lead schedules and coordinates routine NCS oversight activities:
- NCS staff conduct and participate in routine audits of NCS practices, including compliance with procedures.
- NCS staff examine reports of procedural violations and other deficiencies for possible improvement of safety practices and procedural requirements. Findings are reported to management.
- NCS staff periodically review NCSEs to determine their continued applicability and validity. This should include a review of elements of the evaluation such as scope, assumptions, normal conditions, credible abnormal conditions, controls, and limits.
Annual reviews of NCSEs and calculations are conducted, with each evaluation and calculation being reviewed at least once every three years.
- At least every three years, an audit of the overall effectiveness of the CSP is performed.
Management participates actively in this activity.
ipment and procedures needed for NCS controls are clearly identified. Activities involving le material are conducted using written and approved procedures. For situations in which roved procedures are inadequate or do not exist, personnel are required to take no action l the NCS staff has evaluated the situation and provided recovery instructions. Procedures supplemented by appropriate material labeling and postings, specifying material identification limits on parameters, in areas, operations, workstations, and storage locations subject to cedural controls. Equipment and procedures are maintained as part of the facility nagement measures.
3.1.8 Criticality Safety Nonconformances adequacy of engineered and administrative NCS controls is routinely assessed as part of the NE facility audits and inspections. Deviations from procedures and unintended alterations in cess conditions that affect NCS are promptly reported to management using the corrective on program, investigated promptly, corrected as appropriate, and documented. Action to ect such deviations or alterations is taken in accordance with procedural requirements and NE Medical Technologies 6b.3-7 Rev. 5
ntained in the corrective action program.
n the loss of double contingency protection, operations are suspended and processes dered safe until double contingency protection can be restored. Adequacy of the affected trols is subsequently assessed as part of the corrective actions.
S events are reported to the NRC in accordance with the reporting requirements of CFR 70.50, 10 CFR 70.52, and the SHINE-specific reporting requirements which meet the nt of 10 CFR Part 70, Appendix A, as described in the technical specifications.
3.1.8.1 Planned Response to Criticality Accidents CAAS is described in Subsection 6b.3.3.
NE maintains an emergency plan which includes the planned response to criticality idents. The emergency plan contains information on the provision of personnel accident imeters in areas that require the CAAS and arrangements for on-site decontamination of sonnel and the transport and medical treatment of exposed individuals. The SHINE ergency plan is further described in Section 12.7.
3.1.8.2 Criticality Safety Event Reporting ility procedures include provisions for rapid evaluation of the significance of NCS events, uding immediate notifications of facility NCS staff and the assessment of events with respect he loss or degradation of double contingency protection.
significance and reportability of NCS events is based on the loss or degradation of NCS trols and not on the event sequence with respect to whether or not limits were exceeded.
n NCS event cannot be affirmatively determined to not require a one-hour report within one r, it is reported as an event requiring a one-hour report.
3.2 CRITICALITY SAFETY CONTROLS eral failure of a single NCS control which maintains two or more controlled parameters is sidered a single process upset when determining whether the DCP is met.
sive engineered geometry controls are the most preferred type of NCS controls. Otherwise, preferred hierarchy of NCS controls is (1) passive engineered, (2) active engineered, enhanced administrative, and (4) administrative. Use of explicit NCS controls is preferred to nce on the natural and credible course of events. Generally, control on two independent cality parameters is preferred over multiple controls on a single parameter. If redundant trols on a single parameter are used, a preference is given to diverse means of control on parameter.
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controlled parameters used in the CSP are mass, moderation, enrichment, geometry, me, concentration, interaction, physicochemical form, reflection, heterogeneous effects, sity, and process variables. Where these parameters are used to control the criticality risk, following guidance is implemented.
eral:
- When a single-parameter limit is used, all other parameters are evaluated at their optimum or most reactive credible values.
- When process variables can affect the normal or most reactive credible values of parameters, controls to maintain them within specified ranges are established.
- When measurement of a parameter is needed, instrumentation subject to the facility management measures is used.
- When criticality control is based on measuring a single parameter, independent means of measurement are used.
- Limits on controlled parameters are established, taking any tolerances and uncertainty into account.
s:
- When mass limits are derived for a material that is assumed to have a given weight percent of SNM, determinations of mass are based on either (1) weighing the material and assuming that the entire mass is SNM, or (2) conducting physical measurements to establish the actual weight percent of SNM in the material.
- When the dimensions of equipment or containers with a fixed geometry are used to limit the mass of SNM, a conservative process density is used to calculate the resulting mass.
- When over-batching of SNM is credible, the largest mass resulting from a single failure is shown to be subcritical.
deration:
- Physical structures are the preferred means of preventing ingress of moderators.
- Moderation-controlled areas are used to exclude moderator from areas of the SHINE facility.
- Moderation-controlled areas are conspicuously marked, and administrative controls are established to prevent the introduction of moderators.
- Firefighting procedures for use in moderation-controlled areas are evaluated in NCSEs.
Restrictions on the use of moderating firefighting agents are included in procedures and training. The effects of fire and the activation of fire suppression systems is evaluated.
ichment:
- A facility-wide maximum authorized enrichment is used, and the most limiting enrichment is applied to all material.
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- Before beginning operations, in response to changes in operations, and at periodic intervals, all dimensions relied on in demonstrating subcriticality are verified. Relevant dimensions and material properties are maintained by the facility's configuration management program.
- Means of losing geometry control are evaluated and controls are established as needed if they are credible.
- Neutron interaction with other SNM-bearing equipment is considered as part of the demonstration of subcriticality, unless individual units meet the criteria for being considered neutronically isolated.
sity:
- The general criteria listed above are applied.
ume:
- Fixed geometry is used to restrict the volume of SNM. Limiting material to part of a larger geometry using active level probes and overflow lines is also used.
- The maximum subcritical volume is evaluated using the most reactive credible geometry, optimum moderation, and full water reflection.
centration:
- Controls are established to limit concentration of SNM unless the process has been demonstrated to be subcritical at optimum concentration.
- When using a tank containing concentration-controlled solution, the tank is kept closed and locked to prevent unauthorized introduction of precipitating agents.
- Precautions are taken to preclude the inadvertent introduction of precipitating agents.
- Transfers to unfavorable geometry tanks containing concentration-controlled solutions will only be authorized based on dual independent sampling and/or in-line monitoring. No single error may result in transfer of concentrated solution to a tank with unfavorable geometry.
- Process variables that can affect the solubility of fissile solutions are controlled and monitored. The need to ensure homogeneity of the solution is assessed in the NCSEs.
raction:
- To maintain physical separation between units, engineered controls are used. If engineered controls are not feasible, administrative controls with visual aids are used.
- The structural integrity of spacers, storage racks, etc. is sufficient to ensure subcriticality under normal and credible abnormal conditions, including seismic events.
- Engineered devices that are movable are inspected periodically for deformation.
sicochemical Form:
- Explicit controls are established to limit material composition to particular forms.
- Both in-situ changes in the physicochemical form and the migration of material between process areas are considered in evaluating credible abnormal conditions.
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lection:
- In determining the subcritical limits for an individual unit, the wall thickness and all adjacent reflecting materials are considered in setting up the criticality model.
- Criteria are established and documented in the NCSEs for determining when materials are sufficiently far away to be neglected in the criticality model.
- When reflection is not controlled, full reflection is represented by 12 inches of tight-fitting water or 24 inches of tight-fitting concrete.
- Minimum reflection conditions equivalent to a 1-inch tight-fitting water reflector are assumed to account for personnel and other transient incidental reflectors not explicitly included with fixed reflectors in the model.
- When less-than-full reflection conditions are assumed in calculations, controls to limit reflection around individual units are established. Rigid barriers are preferred.
- When evaluating arrays of units, the most reactive combination of interstitial moderation and exterior array reflection is considered and documented in the NCSE and/or calculation.
erogeneity Effects:
- Methods of causing a fissile material to become inhomogeneous are evaluated in NCSEs and controls are established as necessary. If heterogeneity is considered credible, its effect is evaluated in criticality calculations.
- Assumptions that can affect the physical scale of heterogeneity are based on observed physical characteristics of the material; process variables that can affect the scale of heterogeneity are controlled.
cess Variables:
- Process variables relied on to control or monitor other controlled parameters are identified as controls in criticality safety evaluations; sufficient management measures are applied to ensure that the associated controlled parameter limit is not exceeded.
- The associated controlled parameter is explicitly identified and the correlation of process variables to the associated parameter is established by experiment or plant-specific measurements.
3.2.1 Target Solution Staging System target solution staging system (TSSS) is the set of tanks and associated piping used to vide staging and storage of target solution in the radioisotope production facility (RPF). A cess overview is provided in Figure 6b.3-1.
system consists of eight target solution hold tanks and two target solution storage tanks ch receive target solution from the target solution preparation system (TSPS), the iodine and on purification and packaging (IXP), or the molybdenum extraction and purification system PS). Each tank is connected to the vacuum transfer system (VTS) which allows transfer in the system and to other connected systems. The tanks in the system are geometrically rable annular tanks and are in individual below grade vaults equipped with floor drains to the NE Medical Technologies 6b.3-11 Rev. 5
icality Safety Basis NCSE for the TSSS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
er normal conditions, the system is safe-by-design. The pipe and valve sizes and ngements within the system are individually within the evaluated single-parameter limits on metry. Groups of piping have been evaluated and shown to be subcritical given worst-case ditions of concentration, reflection, and corrosion. The tanks in the system have an annular ign that will remain subcritical under the most reactive conditions of concentration, reflection, corrosion. Tanks are equipped with redundant overflows and tank vault drip trays are ipped with adequately sized drains in the event of a tank overflow or leak of target solution.
3.2.2 Radioactive Liquid Waste Storage System radioactive liquid waste storage (RLWS) system collects, stores, blends, conditions, and ers liquid wastes to the radioactive liquid waste immobilization (RLWI) system. A process rview is provided in Figure 6b.3-2.
system consists of two uranium liquid waste tanks, the first of which receive potentially high centration (greater than 25 grams of uranium per liter [gU/L]) uranium-bearing wastes from VTS, MEPS, or IXP. Wastes from VTS come from other upstream sources, such as TSSS.
nominal uranium concentrations from the MEPS and IXP washes are less than 25 gU/L.
h concentration is only expected when a target solution batch is disposed of.
uranium liquid waste tanks are of the same geometrically-favorable design as similar tanks e TSSS and are contained in individual below grade vaults. The uranium liquid waste tanks connected in series to preclude inadvertent direct transfers to the non-favorable-geometry id waste blending tanks.
r radioactive liquid waste tanks are large volume, non-favorable-geometry tanks which eive and store negligible concentration (less than 1 gU/L) wastes from the process vessel t system (PVVS), MEPS, and IXP.
ht liquid waste blending tanks are large volume, non-favorable-geometry tanks which store concentration (less than 25 gU/L) wastes. These tanks receive low concentration wastes the second uranium liquid waste tank and negligible concentration waste from the upstream oactive liquid waste tank.
radioactive liquid waste tanks and liquid waste blending tanks are equipped with dedicated pling equipment which is used to draw liquid samples of the tank contents to determine nium concentration and pH. Samples from the uranium liquid waste tanks are obtained using VTS and analyzed in the quality control and analytical testing laboratories (LABS).
normal process for receiving high concentration wastes proceeds as follows. First, the high centration wastes are moved from an upstream system into the first uranium liquid waste
. When the waste is desired to be transferred to the waste immobilization system, it is first NE Medical Technologies 6b.3-12 Rev. 5
ansferred to the second uranium liquid waste tank and re-sampled. If the sampling conditions met, the low concentration waste is then transferred by vacuum to the liquid waste blending
- k. The liquid waste blending tank may be further down-blended with negligible concentration tes from the radioactive liquid waste tanks to meet downstream waste disposal specifications e RLWI system.
icality Safety Basis NCSE for the RLWS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
er normal conditions, portions of the system are safe-by-design. The pipe and valve sizes arrangements within the system are individually within the evaluated single-parameter limits geometry. Groups of piping have been evaluated and shown to be subcritical given worst-e conditions of concentration, reflection, and corrosion. The uranium liquid waste tanks have annular design that will remain subcritical under worst-case conditions of concentration, ection, and corrosion. The tanks are equipped with dual overflows and the tank vault drip tray quipped with an adequately sized drain in the event of an overflow or leak from the tank.
radioactive liquid waste tanks and the liquid waste blending tanks are not safe-by-design require application of the DCP to prevent criticality accidents. The concentration limit for e tanks is significantly less than the single-parameter limit for uranium concentration.
undant in-series controls on concentration are relied upon to meet the DCP. The sampling transfer processes consist of multiple independent sampling and authorization steps.
ore mixing begins, the first uranium liquid waste tank is isolated from its inputs. The second nium liquid waste tank is isolated from the first tank and from the downstream liquid waste ding tanks. Before sampling, the tank is mixed well to ensure the sample is representative of contents of the tank. A sample is drawn into the sample tank and an operator takes a sample proceeds to test this sample using a prescribed sampling method. The solution in the sample is then returned to the first tank. Results of the sample are sent by the operator to the control m supervisor, who confirms the results are acceptable and authorizes the contents of the first nium liquid waste tank to be transferred to the second uranium liquid waste tank.
n successful transfer to the second uranium liquid waste tank, the tank is isolated from the tank until the completion of the transfer process to the liquid waste blending tanks. Before pling, the second tank is mixed well to ensure the sample is representative of the contents of tank. A sample is drawn into the sample tank and a different operator takes a sample and ceeds to test this sample using a prescribed sampling method, different from the previous ple. Once the operator has finished, they relay the results of the sample to the control room ervisor. The supervisor reviews the tests, confirms the results are acceptable, and authorizes transfer to the liquid waste blending tanks.
3.2.3 Molybdenum Extraction and Purification System MEPS extracts and purifies molybdenum from irradiated target solution. A process overview rovided in Figure 6b.3-3.
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reagents used in the system are contained on a chemical reagents skid located outside the cell.
ing the extraction process, target solution is lifted into the vacuum transfer tanks in the action cell and pumped through a regenerative and non-regenerative heat exchanger and the action column. The extraction column is an adsorption media column which separates out the ybdenum from the target solution. The target solution is returned to the TSSS after extraction e-use, though it may be sent to the RLWS if desired. A series of acid and water washes to RLWS are used to flush the extraction process lines following target solution to remove any dual target solution from the lines. After the wash, the three-way valves in the system are ositioned to allow sodium hydroxide to flow through the extraction column and release the orbed molybdenum into the eluate hold tank.
icality Safety Basis NCSE for the MEPS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
er normal conditions, portions of the system containing fissile target solution are safe-by-ign. The pipe, heat exchanger, valve, tank, column, and pump sizes are individually within the luated single-parameter limits on geometry and/or volume. Groups of piping and other ponents have been evaluated and shown to be subcritical given worst-case conditions of centration, reflection, and corrosion with minimum edge-to-edge separation and minimum aration between adjacent extraction cells. The extraction cell is equipped with a drain to RDS target solution through-cell transfer pipes are double walled, with the outer wall draining to S as well. The cell is fully enclosed to minimize the intrusion of moderating liquids.
molybdenum eluate hold tank is not safe-by-design and requires application of the DCP to vent criticality accidents. A three-way valve design prevents inadvertent transfer of target tion to the eluate tank. Additionally, an isolation valve is administratively closed to prevent vertent transfer if the three-way valve fails.
dvertent transfer of target solution to the facility chemical reagent system (FCRS) requires lication of the DCP to prevent criticality accidents. A three-way valve design prevents flow of et solution toward the FCRS reagent vessels. An isolation valve is installed between the RS and upper vacuum lift tanks that is administratively closed during target solution cessing, and a check-valve also exists to prevent inadvertent flow of target solution to the gent vessels.
cipitation due to the inadvertent addition of caustic reagents requires application of the DCP revent criticality accidents. The volume of caustic reagents and the sequence of column hes is administratively controlled to prevent potential precipitate formation. Additionally, a mn frit filter prevents downstream transfer of any potential solid precipitates.
3.2.4 Target Solution Preparation System TSPS produces uranyl sulfate solution, referred to as target solution, from uranium oxide der. The uranium oxide powder is dissolved in sulfuric acid to produce uranyl sulfate.
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uranium oxide powder is manually transferred from the uranium receipt and storage system SS) to the TSPS glovebox. The powder is stored and handled in sealed cans which are ned inside the glovebox. The oxide powder is then metered and poured into the dissolution
- s. The dissolution tank is then charged with hydrogen peroxide (if used) and sulfuric acid in uence to produce the final uranyl sulfate product. The tanks are heated during the process to ure proper dissolution. The tanks themselves are favorable geometry vessels with a trolled diameter to protect against potential criticality.
e the dissolution process is complete, the tank contents are pumped through a filter into the et solution preparation tank and can then be transferred into the TSSS. The target solution paration tank is a favorable-geometry annular tank like those found in the TSSS and RLWS.
ause the dissolution process evolves heat and water vapor, the off-gas from the process s through a reflux condenser which condenses the vapor and returns it to the dissolution
. The reflux condenser is cooled by the radioisotope process cooling system (RPCS). The ebox and reflux condenser are vented to the facility radiological ventilation system.
icality Safety Basis NCSE for the TSPS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
TSPS is subject to two sets of criticality safety limits. Portions of the system contain oxide der in both dry and wet (partially-dissolved) conditions, and the remainder of the system tains uranyl sulfate. The uranium concentration in the uranyl sulfate may be higher in this tem than in the rest of the facility due to the nature of the process.
er normal process conditions, the mass of uranium oxide is controlled to less than the mally-moderated, fully-reflected critical mass for uranium oxide of oxide per canister, and a single oxide canister is permitted in the glovebox at any given time. High efficiency iculate air (HEPA) filters are favorable geometry within the single parameter limit and alled on the glovebox to prevent significant buildup of oxide powder outside of the glovebox n downstream ventilation ductwork. Visual surveillance is performed to identify any spills of le material or introduction of moderators.
TSPS room moderator exclusion features (e.g., non-hydrogenous fire protection, elevated r) and glovebox itself are designed to preclude the intrusion of significant amounts of derator. Therefore, the glovebox will remain safely subcritical under normal process ditions. The mass limit also protects the dissolution process in the dissolution tanks, though are designed with favorable geometry even for the most reactive combination of uranium e and water.
wnstream of the dissolution tanks are pipes, transfer pumps, and filters, which are favorable metry within the single parameter limit. The target solution preparation tank is favorable metry including corrosion allowances and optimum concentration of solution. Interaction ween components is controlled with minimum separation distances and a cage around the olution tanks.
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ation valves on cooling and ventilation lines. There is a check-valve on the return side of the ux condenser cooling line to prevent backflow of cooling water into the dissolution tanks, and reflux condenser is favorable geometry within the single parameter limits. Additionally, a er-tight plug is inserted into the powder chute after oxide powder introduction into the olution tanks.
ition of moderator during maintenance activities requires application of the DCP to prevent cality accidents. Maintenance activities are administratively controlled, and independently fied, to ensure fissile material is removed prior to maintenance activities, and that all derating materials are removed prior to re-starting operations.
mplete dissolution and transfer of solids downstream of the dissolution tanks requires lication of the DCP to prevent criticality accidents. The dissolution procedure is inistratively controlled, with supervisory oversight, to ensure the appropriate sequencing and me of reagents is followed to ensure complete dissolution. Reagent tanks have unique nectors and limited volume to prevent inadvertent reagent addition. Additionally, downstream rable geometry filters remove potential solids in the target solution.
3.2.5 Vacuum Transfer System VTS is an interconnected series of pipes and vacuum lift tanks which facilitate the transfer of et solution throughout the facility. A process overview is provided in Figure 6b.3-5.
lift tanks are capable of drawing solution from the TSSS, RLWS, subcritical assembly tem (SCAS), and the RDS for various purposes and supply solution to the TSSS, RLWS, WI, SCAS, and MEPS. The tanks are supplied with vacuum through associated vacuum ps and valves which regulate and maintain vacuum pressure throughout the system.
uum is broken in the lift tanks by venting the tank through a three-way valve which isolates vacuum header and allows inflow from radiological ventilation zone 2 (RVZ2). Breaking uum in a lift tank allows gravity drain of its contents to the desired destination in one of the nected systems. Note that two-way transfers are not possible for the MEPS, RLWI, and RDS.
can only supply to MEPS and RLWI, and it can only remove target solution from the RDS.
icality Safety Basis NCSE for the VTS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
VTS components which contain target solution are designed with favorable geometry for the st reactive concentration. The components individually have geometry within the evaluated le parameter limits for target solution. In cases where favorable geometry components are in ximity to each other, the interaction between the components is evaluated and controlled.
VTS components are designed to prevent leaks of solution. Vaults or hot cells containing the tanks or associated piping are equipped with drip trays and adequately sized drains that n to RDS. The vacuum buffer tank is equipped with a demister that separates potentially ained liquid in the vapor, which prevents transfer of target solution to downstream ponents.
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tion to non-favorable geometry components within the VTS. The vacuum headers are ipped with liquid detection that stops transfers upon detection of liquid. Additionally, a
-check valve is located between the vacuum lift tanks and the vacuum buffer tank (VTS ckout pot) to prevent high level transfer of solution to the vacuum buffer tank.
3.2.6 Process Vessel Vent System PVVS is an off-gas management system for the process equipment which contains oactive liquids with the potential for excessive hydrogen production in the IXP system, PS, RLWI, RLWS, TSSS, and VTS. The PVVS also periodically accepts gas from the target tion vessel (TSV) off-gas system (TOGS). The PVVS supplies ventilation flow and receives oactive gas from the tanks and other equipment in these systems and processes it through a es of filters, delay beds, and blowers before it is released from the facility stack. The system s not normally contain significant fissile material.
icality Safety Basis NCSE for the PVVS shows that the entire process will remain subcritical under normal and dible abnormal conditions. There are no identified criticality safety controls for the PVVS.
dvertent transfer of target solution into the PVVS is prevented in upstream systems.
3.2.7 Uranium Receipt and Storage System URSS receives and stores enriched uranium oxide and metal and converts uranium metal oxide for use in the TSPS. A process overview is provided in Figure 6b.3-6.
vities for the receipt and measurements of uranium and the conversion from metal to oxide ur inside the URSS glovebox. Upon receipt, the convenience cans are removed from the ping container and imported into the glovebox for measurement and repackaging into metal xide storage cans, as appropriate. Once the metal or oxide cans are appropriately loaded, are moved to the appropriate storage rack. For conversion activities, a metal can is moved the storage rack to the glovebox where it is converted using specified time and temperature straints to the appropriate uranium oxide. The oxide is then measured, and an oxide can is ed with the product which is then transferred to the oxide storage rack. Oxide may also be sferred to the TSPS for processing into solution.
icality Safety Basis NCSE for the URSS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
eipt and handling of shipping containers which contain uranium is in accordance with the roved safety analysis for packaging associated with each container. Areas in which intact ping containers are stored are controlled by limiting the aggregate criticality safety index for storage area. Administrative controls are used to ensure the criticality safety index limits are exceeded.
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blish double contingency protection for the system. For a criticality to occur under normal ditions, a non-credible quantity of metal or oxide would need to be introduced into the system ass limits would need to be exceeded concurrent with the introduction of a significant ntity of moderator. Moderator controls and the glovebox itself prevent the uncontrolled usion of moderators into areas containing exposed fissile material.
oduction of high-enrichment uranium requires application of the DCP to prevent criticality idents. Upon receipt of uranium, examination of the supplier certification is used to confirm condition of received material prior to import of material to the glovebox. Confirmation of erial form and enrichment by sample analysis are used to ensure that appropriate limits are lied.
umulation of excess mass requires application of the DCP to prevent criticality accidents. The ss of uranium in- and out-of-storage is administratively controlled. Material contained within led shipping containers, the glovebox, and the storage racks is considered to be in-storage is subject to specific limits for each of these areas. Material out-of-storage is administratively ted to a value significantly below the single-parameter subcritical limit. Controls on the use transport of moderators within the room are used to prevent the interaction of material out-of-age with moderating materials. HEPA filters, which are favorable geometry within single ameter limits, prevent the accumulation of oxide outside of the glovebox or in downstream tilation. Holdup of fissile material in the process is controlled in the glovebox and furnace by king mass and periodic cleanout of the glovebox and furnace based on the throughput of nium. Cleanout of fissile material holdup is independently verified prior to restarting rations. During maintenance activities, fissile material is removed prior to maintenance and derators are removed prior to restarting operations. Confirmation of fissile material and derator removal is performed under supervisory oversight.
mplete oxidation of metal requires application of the DCP to prevent criticality accidents. The ace oxidation steps are administratively controlled to ensure adequate oxidation.
itionally, sample analysis following oxidation verifies oxide powder content and moisture tent of the oxide. Operators visually confirm that only uranium oxide is added to an oxide ister.
URSS oxide storage rack and metal storage rack are favorable geometry and maintain the ropriate storage cell size. The maximum number of storage cells is significantly below the wable number of storage cells based on the mass per storage canister. The mass in each age canister is administratively controlled. Movement of fissile material out-of-storage is ntained at an appropriate separation distance to other fissile material in storage to prevent avorable interaction.
3.2.8 Radioactive Drain System RDS collects overflows and leakage of target solution from systems in the RPF and directs it wo favorable-geometry tanks in below grade vaults. A process overview is provided in ure 6b.3-7.
system is comprised of drip pans, piping, and collection tanks. The collection tanks are mally maintained empty and are equipped with instrumentation to alert personnel of an NE Medical Technologies 6b.3-18 Rev. 5
pled, and withdrawn through the VTS to the TSSS or RLWS as appropriate.
icality Safety Basis NCSE for the RDS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
er normal process conditions, the RDS does not contain fissile material. Leakage or overflow rget solution to the RDS is considered an abnormal condition for the facility but is considered normal condition for the purpose of the criticality safety evaluation for the system. The RDS tanks and piping are favorable geometry for the most reactive concentration of target tion and are safe-by-design. The vacuum lift tanks within the RDS are favorable geometry in the single parameter limits. The hold tanks are equipped with overflow lines, and RDS ns are adequately sized to prevent buildup of solution in the vault drip tray. Drip trays are also ed toward the drain lines. Interaction is controlled between components with minimum aration distances between components and between vaults.
cipitation of solids requires application of the DCP to prevent criticality accidents. The hold s are equipped with level instrumentation to detect a leak of solution transferred to RDS.
itionally, administrative controls ensure that, upon a leak, normal operations stop, the leaked tion is sampled, and appropriate recovery actions are performed.
3.2.9 Radioactive Liquid Waste Immobilization System RLWI system receives radioactive liquid waste from the RLWS and mixes it with solidifying nts to stabilize and solidify the liquid waste in drums. The drums are then moved into storage eventually to long-term disposal. A process overview is provided in Figure 6b.3-8.
ste with a uranium concentration capable of meeting the waste acceptance and storage uirements enters the system from the RLWS liquid waste blending tanks and into the obilization feed tank by drawing a vacuum on the immobilization feed tank. When the waste ady to be immobilized, it is pumped from the immobilization feed tank by the liquid waste m fill pump and into a radioactive liquid waste drum pre-loaded with solidification agents. The tents of the radioactive liquid waste drum are solidified and after adequate cure time, the dified waste drum is remotely loaded into a shielded drum for transport to the material staging ding.
icality Safety Basis NCSE for the RLWI system shows that the entire process will remain subcritical under mal and credible abnormal conditions.
er normal process conditions, the incoming feed stream from RLWS contains low centrations of fissile material and is significantly below the single parameter limit for uranium centration in solution. The operational limits on uranium concentration for the input stream driven by waste acceptance requirements and are even lower than the allowable limits for cality safety.
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vity. A barrel which meets the waste acceptance limits meets the criticality safety limits.
mple analysis of solution transferred to RLWI is performed and compared to previous sample ults and verify uranium concentration is within the established limits. The proper amount of dification agents is added to a barrel and weighed prior to transfer of uranium-bearing solution he barrel to ensure waste acceptance limits are satisfied for downstream storage of the waste els within the material staging building.
raction between barrels is controlled by limiting the number of barrels present within the obilization skid.
cipitation of uranium requires application of the DCP to prevent criticality accidents. Reagent sels have unique nozzle connections to prevent inadvertent transfer of reagents, and the me of the vessels is limited. Process lines are sloped, and equipment are equipped with ns to prevent holdup of fissile material. Additionally, solutions transferred to the RLWI system ergo dual, independent sample analysis to verify the pH of the solution is within limits prior to sferring the solution.
3.2.10 Laboratories LABS receive, store, and process liquid and solid analytical samples of oxides, metals, and diated and unirradiated target solution.
laboratory is controlled by an overall limit on mass which is significantly below the subcritical t on mass for uranium-235 and is subcritical under all conditions.
icality Safety Basis NCSE for the LABS shows that the entire process will remain subcritical under normal and dible abnormal conditions.
LABS system is administratively controlled to ensure the combined total uranium mass is ificantly below the subcritical mass for uranium-235.
3.2.11 Material Staging Building material staging building exists to process, characterize, and store byproduct material and M, used in the production of medical isotopes. The material staging building provides a tion for the packaged radioactive material to decay until it can be transported to an off-site l disposal location. The material staging building will mostly store standard-sized 55-gallon ms containing cured, solidified waste. Other forms of radioactive waste are stored in the erial staging building (e.g., used neutron drivers, glassware). Transient processing of column te drums, including encapsulation and loading into appropriate shipping containers, also urs in the material staging building. Column waste drums are not stored in the material ing building.
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NCSE for the material staging building shows that the entire process will remain subcritical er normal and credible abnormal conditions.
material stored in the material staging building is comprised entirely of exempt fissile erial. Solidified liquid waste meets the requirements for exemption from classification as le material under 10 CFR 71.15(c). As-generated solid waste meets the requirements for mption from classification as fissile material under 10 CFR 71.15(a). Because the SNM ed in the material staging building is exempt fissile material, there is no credible means by ch a criticality would occur.
nsient processing of column waste drums requires application of the DCP. The total non-mpt fissile material in a fissile material area is limited to 600 grams. Additionally, packages non-exempt fissile material are limited to a maximum fissile material content of 50 grams.
3.2.12 Iodine Extraction and Purification System IXP is designed to separate iodine from irradiated uranyl sulfate target solution [
]PROP/ECI. The iodine is then purified a sodium hydroxide solution. Xenon is collected from [
]PROP/ECI. The IXP is in a hot cell.
operating line of the IXP is part of the RPF.
icality Safety Basis NCSE for the IXP shows that the entire process will remain subcritical under normal and dible abnormal conditions.
piping and equipment in the IXP containing target solution is favorable geometry within the le parameter limits. The IXP cell is equipped with a drain to RDS that is adequately sized to vent buildup of solution in the cell.
inadvertent transfer of target solution to the IXP eluate tank requires application of the DCP revent criticality accidents. A three-way valve is designed to prevent transfer of target solution he eluate tank during extraction processing. Additionally, an isolation valve located between three-way valve and eluate tank is administratively closed during processing of target tion.
vention of target solution backflow into the FCRS requires application of the DCP to prevent cality accidents. A check valve is installed to prevent the flow of solution upstream to FCRS.
itionally, an isolation valve located between the check valve and the FCRS is administratively ed during processing of target solution.
cipitation due to inadvertent addition of caustic reagents requires application of the DCP to vent criticality accidents. The IXP is equipped with unique nozzle hookups for each reagent to vent improper FCRS hookups. Additionally, the wash sequence of the column is inistratively controlled to prevent precipitation.
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SHINE facility provides a CAAS to detect a criticality event in the areas in which non-exempt ntities of fissile material greater than the limits identified in 10 CFR 70.24(a) are used, dled, or stored outside the TSVs. The criticality accident alarm system at the SHINE facility is igned to meet the requirements of 10 CFR 70.24, and conforms to the requirements in SI/ANS-8.3-1997 (R2017), as endorsed by RG 3.71.
CAAS consists of detectors located throughout the main production facility at locations ignated to provide sufficient coverage of areas in which SNM is used, handled, and stored.
3.3.1 Minimum Accident of Concern minimum accident of concern (MAC) for the SHINE facility is developed based on a critical ere of 20 percent enriched uranyl sulfate solution. This system is representative of the ority of operations conducted within the SHINE facility. Process accidents involving solutions also statistically more likely to occur, based on available historical data.
ector placement is determined by neutron transport analysis using the MAC. The transport lysis converts the neutron and gamma spectrum of the MAC to a point source which is used a computer model of the facility structure, shielding, and intervening equipment to determine ropriate detector placements and detection thresholds. The detection thresholds are based he requirements of 10 CFR 70.24 and the detector response to neutron radiation. Selection eutron detectors and neutron transport analysis are appropriate for the SHINE facility ause the facility contains multiple sources of gamma radiation which could interfere with the ration of the CAAS in a way that would result in an unacceptable number of false alarms.
3.3.2 Criticality Accident Alarm System Design CAAS will energize visible and audible alarms in the affected area of the main production lity and in the facility control room if a criticality accident occurs. Mandatory evacuation areas determined and clearly marked with evacuation routes for areas in which personnel would eive a dose exceeding 12 rads (0.12 grays) in free air. Evacuation routes are selected to ure personnel are evacuated away from areas with potentially higher dose during a criticality ident.
CAAS detectors are arranged so that each area outside of the irradiation unit cells in which cial nuclear material is used, handled, or stored within the main production facility receives erage from at least three detectors, which allows a single detector to be taken out of service maintenance without impact to the operability of the system. Under normal conditions, the ector logic requires that two detectors are needed to trigger an alarm condition, which imizes the potential for false actuations of the alarm. Protection against latent detector res during maintenance conditions is achieved by locking in an alarm signal from any ectors which are out of service for maintenance, which reduces the detection requirement to a le detection within the affected zones.
CAAS employs a logic unit, located in the facility control room, which contains redundant m logic to ensure that a latent failure in the logic unit does not preclude an alarm when ded. Electrical power is normally supplied by the facility normal electrical power supply tem (NPSS), with a backup connection to the uninterruptible electrical power supply system NE Medical Technologies 6b.3-22 Rev. 5
icient time to secure the movement of fissile material before loss of alarm system coverage.
table instruments may be used to provide equivalent coverage in rare circumstances.
luation and deployment of portable instrumentation is managed on a case-by-case basis.
CAAS is designed to be resistant from anticipated adverse effects such as a fire, explosion, osive atmosphere, seismic shock, or other adverse conditions that do not result in evacuation he entire facility. The system is designed to preclude false alarms due to system failure and tains sufficient fault detection to alert operators as needed during failures.
maintenance or other conditions which would disable multiple detectors or the logic unit, the wing compensatory measures are implemented to ensure an equivalent level of safety:
- Temporary criticality detection equipment with audible alarms will be used for personnel remaining in or entering the affected area, and
- Personnel access to the affected area will be limited to essential activities.
se compensatory measures are specific to the affected area of the main production facility provide a time allowance to restore the system to full operation in lieu of immediate process tdown.
3.4 TECHNICAL SPECIFICATIONS controls required to maintain the criticality safety basis are contained in the SHINE technical cifications.
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Table 6b.3 Summary of Benchmarks Selected for the SHINE Validation Report Benchmark Series Cases Description of Physical Systems U-SOL-THERM-003 9 10.06% enriched uranyl nitrate, un-reflected U-SOL-THERM-002 13 30.45% enriched uranyl fluoride, water-reflected and un-reflected U-SOL-THERM-003 46 30.3% uranyl fluoride, water-reflected and un-reflected U-SOL-THERM-004 1 14.7% uranyl sulfate, reflected by beryllium oxide U-SOL-THERM-004 7 9.97% enriched uranyl nitrate, water-reflected U-SOL-THERM-007 5 9.97% enriched uranyl nitrate, un-reflected U-SOL-THERM-008 4 9.97% enriched uranyl nitrate, concrete-reflected U-SOL-THERM-016 7 9.97% enriched uranyl nitrate, water-reflected U-SOL-THERM-017 6 9.97% enriched uranyl nitrate, un-reflected U-SOL-THERM-018 6 9.97% enriched uranyl nitrate, concrete-reflected U-SOL-THERM-020 4 9.97% enriched uranyl nitrate, water-reflected U-SOL-THERM-021 4 9.97% enriched uranyl nitrate, un-reflected U-SOL-THERM-023 9 9.97% enriched uranyl nitrate, un-reflected U-SOL-THERM-025 7 9.97% enriched uranyl nitrate, concrete-reflected NE Medical Technologies 6b.3-24 Rev. 5
Table 6b.3 Area of Applicability Summary Parameter Area of Applicability sile Material and Composition Uranyl Sulfate Uranyl Nitrate Uranyl Fluoride emical Form Solution rage Neutron Energy Causing Fission (ANECF) (MeV) 0.004-0.064 ichment (wt. %) 10-30.5 lector Materials None Water Graphite Beryllium Oxide Concrete nium Concentration (g-U/L) 52.8-960 35U Ratio 75-1610 NE Medical Technologies 6b.3-25 Rev. 5
NE Medical Technologies 6b.3-26 Rev. 5 NE Medical Technologies 6b.3-27 Rev. 5 NE Medical Technologies 6b.3-28 Rev. 5 NE Medical Technologies 6b.3-29 Rev. 5 RCA VTS Vacuum Ven tilat ion Eq uipm ent Zone 2 Pro cess Vessel Ven tilat ion Molybd enum Ext r act ion Vacuum Lif t an d Tanks Pur if ication Syst em Target Radioact ive Su bcrit ical Solut ion Radioact ive Liqu id Waste Assembly St aging Drain Syst em St or age Syst em Syst em NE Medical Technologies 6b.3-30 Rev. 5
Receive Shipping Package Remove Canisters Oxide Metal Uranium Handling Glovebox Import Metal Import Oxide Thermal Oxidation Canister Canister Repackage to Metal Repackage to Oxide Storage Canister Storage Canister Export Metal Export Oxide Storage Canister Storage Canister Storage in Racks Storage in Racks NE Medical Technologies 6b.3-31 Rev. 5
Molybd enum Target Iodine and Radioact ive Ext r act ion Vacuum Solut ion Xenon Liqu id Waste an d Tr ansfer St aging Pur if ication St or age Pur if ication Syst em Syst em Syst em Syst em Tank Tank Vault, Pipe Su per cell Overf low Trench, and Valve Pit Drains Drains Lines RDS Hold Tanks NE Medical Technologies 6b.3-32 Rev. 5
Vacuum Tr ansfer Syst em Radioact ive Imm obilization Liqu id Waste Liqu id Waste Pum p Drum Fill Head Feed Tank Drum Syst em NE Medical Technologies 6b.3-33 Rev. 5
SI/ANS, 1983. Safety in Conducting Subcritical Neutron-Multiplication Measurements In Situ, SI/ANS-8.6-1983 (R2017), American National Standards Institute/American Nuclear Society, 3.
SI/ANS, 1991. Nuclear Criticality Safety Training ANSI/ANS-8.20-1991 (R2015), American ional Standards Institute/American Nuclear Society, 1991.
SI/ANS, 1995. Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors SI/ANS-8.21-1995 (R2011), American National Standards Institute/American Nuclear iety, 1995.
SI/ANS, 1997a. Nuclear Criticality Safety Based on Limiting and Controlling Moderators SI/ANS-8.22-1997 (R2016), American National Standards Institute/American Nuclear iety, 1997.
SI/ANS, 1997b. Criticality Accident Alarm System ANSI/ANS-8.3-1997 (R2017), American ional Standards Institute/American Nuclear Society, 1997.
SI/ANS, 1998. Nuclear Criticality Safety in the Storage of Fissile Materials, SI/ANS-8.7-1998 (R2017), American National Standards Institute/American Nuclear Society, 8.
SI/ANS, 2007a. Criticality Safety Engineer Training and Qualification Program SI/ANS-8.26-2007 (R2016), American National Standards Institute/American Nuclear iety, 2007.
SI/ANS, 2007b. Nuclear Criticality Accident Emergency Planning and Response SI/ANS-8.23-2007 (R2012), American National Standards Institute/American Nuclear iety, 2007.
SI/ANS, 2014a. Administrative Practices for Nuclear Criticality Safety ANSI/ANS-8.19-2014, erican National Standards Institute/American Nuclear Society, 2014.
SI/ANS, 2014b. Nuclear Criticality Safety in Operations with Fissionable Materials Outside ctors ANSI/ANS-8.1-2014, American National Standards Institute/American Nuclear Society, 4.
SI/ANS, 2015. Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and finement ANSI/ANS-8.10-2015, American National Standards Institute/American Nuclear iety, 2015.
SI/ANS, 2017. Validation of Neutron Transport Methods for Nuclear Criticality Safety culations ANSI/ANS-8.24-2017, American National Standards Institute/American Nuclear iety, 2017.
NRC, 2001. Guide for Validation of Nuclear Criticality Safety Methodology. NUREG/CR-6698, 1.
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