ML19331A663
| ML19331A663 | |
| Person / Time | |
|---|---|
| Site: | SHINE Medical Technologies, 99902034 |
| Issue date: | 11/14/2019 |
| From: | SHINE Medical Technologies |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19331A832 | List: |
| References | |
| 2019-SMT-0119 | |
| Download: ML19331A663 (126) | |
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ACCIDENT ANALYSIS TABLE OF CONTENTS tion Title Page 2 IRRADIATION FACILITY ACCIDENT ANALYSIS .......................................... 13a2.1-1 2.1 ACCIDENT-INITIATING EVENTS AND SCENARIOS ................................... 13a2.1-1 13a2.1.1 IF MAXIMUM HYPOTHETICAL ACCIDENT .......................... 13a2.1-2 13a2.1.2 INSERTION OF EXCESS REACTIVITY ................................ 13a2.1-5 13a2.1.3 REDUCTION IN COOLING .................................................. 13a2.1-14 13a2.1.4 MISHANDLING OR MALFUNCTION OF TARGET SOLUTION ........................................................................... 13a2.1-16 13a2.1.5 LOSS OF OFF-SITE POWER .............................................. 13a2.1-21 13a2.1.6 EXTERNAL EVENTS ........................................................... 13a2.1-23 13a2.1.7 MISHANDLING OR MALFUNCTION OF EQUIPMENT ....... 13a2.1-27 13a2.1.8 LARGE UNDAMPED POWER OSCILLATIONS .................. 13a2.1-29 13a2.1.9 DETONATION AND DEFLAGRATION IN THE PRIMARY SYSTEM BOUNDARY ......................................................... 13a2.1-31 13a2.1.10 UNINTENDED EXOTHERMIC CHEMICAL REACTIONS OTHER THAN DETONATION .............................................. 13a2.1-33 13a2.1.11 SYSTEM INTERACTION EVENTS ...................................... 13a2.1-35 13a2.1.12 FACILITY-SPECIFIC EVENTS ............................................. 13a2.1-40 2.2 ACCIDENT ANALYSIS AND DETERMINATION OF CONSEQUENCES ...... 13a2.2-1 13a2.2.1 IF MAXIMUM HYPOTHETICAL ACCIDENT .......................... 13a2.2-6 13a2.2.2 INSERTION OF EXCESS REACTIVITY ................................ 13a2.2-8 13a2.2.3 REDUCTION IN COOLING .................................................. 13a2.2-10 13a2.2.4 MISHANDLING OR MALFUNCTION OF TARGET SOLUTION ........................................................................... 13a2.2-12 13a2.2.5 LOSS OF OFF-SITE POWER .............................................. 13a2.2-14 NE Medical Technologies 13-i Rev. 0
ACCIDENT ANALYSIS TABLE OF CONTENTS tion Title Page 13a2.2.6 EXTERNAL EVENTS ........................................................... 13a2.2-15 13a2.2.7 MISHANDLING OR MALFUNCTION OF EQUIPMENT ....... 13a2.2-17 13a2.2.8 LARGE UNDAMPED POWER OSCILLATION ..................... 13a2.2-19 13a2.2.9 DETONATION AND DEFLAGRATION IN THE PRIMARY SYSTEM BOUNDARY ......................................................... 13a2.2-20 13a2.2.10 UNINTENDED EXOTHERMIC CHEMICAL REACTIONS OTHER THAN DETONATION .............................................. 13a2.2-22 13a2.2.11 SYSTEM INTERACTION EVENTS ...................................... 13a2.2-24 13a2.2.12 FACILITY-SPECIFIC EVENTS ............................................. 13a2.2-26 3
SUMMARY
AND CONCLUSIONS .................................................................... 13a3-1 4 REFERENCES .................................................................................................. 13a4-1 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSES............... 13b.1-1
.1 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSIS METHODOLOGY ............................................................................................. 13b.1-1 13b.1.1 PROCESSES CONDUCTED OUTSIDE THE IRRADIATION FACILITY .......................................................... 13b.1-1 13b.1.2 ACCIDENT INITIATING EVENTS ............................................ 13b.1-3
.2 ANALYSES OF ACCIDENTS WITH RADIOLOGICAL CONSEQUENCES ..... 13b.2-1 13b.2.1 MAXIMUM HYPOTHETICAL ACCIDENT IN THE RPF ........... 13b.2-1 13b.2.2 LOSS OF ELECTRICAL POWER ............................................ 13b.2-2 13b.2.3 EXTERNAL EVENTS ............................................................... 13b.2-3 13b.2.4 RPF CRITICAL EQUIPMENT MALFUNCTION ........................ 13b.2-3 13b.2.5 RPF INADVERTENT NUCLEAR CRITICALITY ..................... 13b.2-10 13b.2.6 RPF FIRE ............................................................................... 13b.2-10 NE Medical Technologies 13-ii Rev. 0
ACCIDENT ANALYSIS TABLE OF CONTENTS tion Title Page
.3 ANALYSES OF ACCIDENTS WITH HAZARDOUS CHEMICALS ................... 13b.3-1
.4 REFERENCES ................................................................................................. 13b.4-1 NE Medical Technologies 13-iii Rev. 0
2.2-1 Summary of Radiation Transport Terms (Public) 2.2-2 Summary of Radiation Transport Terms (Worker) 3-1 Irradiation Facility Accident Dose Consequences
.2-1 Radiation Transport Factors
.2-2 Radioisotope Production Facility Accident Dose Consequences
.3-1 Inventory of Hazardous Chemicals
.3-2 Hazardous Chemical Source Terms and Concentration Levels NE Medical Technologies 13-iv Rev. 0
2.2-1 Radiological Consequence Assessment NE Medical Technologies 13-v Rev. 0
onym/Abbreviation Definition alternating current CL Atomic Energy of Canada Limited GLs Acute Exposure Guideline Levels HA Areal Locations of Hazardous Atmospheres F airborne release fraction S accelerator tritium interface system NM Best Estimate Neutronics Model MS continuous airborne monitoring system carbon monoxide A design basis accident E design basis earthquake F dose conversion factors damage ratio PGs Emergency Response Planning Guidelines FAS engineered safety features actuation system RS facility chemical reagent system NE Medical Technologies 13-vi Rev. 0
onym/Abbreviation Definition A fire hazards analysis EA failure modes and effects analyses feet gallons SS glovebox stripper system gallons per minute ZOP hazard and operability PS high voltage power supply intermediate bulk containers initiating event irradiation facility D Iodine Model for Containment Codes integrated safety analysis interim staff guidance irradiation unit iodine and xenon purification and packaging NE Medical Technologies 13-vii Rev. 0
onym/Abbreviation Definition effective neutron multiplication factor kilowatt liter (l/s) liters per second S quality control and analytical laboratories lower flammability limit IC Library of Iodine Reactions in Containment OP loss of off-site power leak path factor PS light water pool system meters per second R material at risk C motor control center NP Monte Carlo N-Particle Transport Code PS molybdenum extraction and purification system S molybdenum isotope packaging system NE Medical Technologies 13-viii Rev. 0
onym/Abbreviation Definition A maximum hypothetical accident 99 molybdenum-99 S nitrogen purge system S nuclear criticality safety AS neutron driver assembly system SS normal electrical power supply system C NDAS service cell HA Occupational Safety and Health Administration Pascal C Protective Action Criteria HS process chilled water system LS primary closed loop cooling system percent millirho A process hazard analysis S process integrated control system B primary system boundary NE Medical Technologies 13-ix Rev. 0
onym/Abbreviation Definition pounds per square inch absolute pounds per square inch gauge VS process vessel vent system A radiologically controlled area S radioactive drain system respiratory fraction 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 Z2r radiological ventilation zone 2 recirculation subsystem NE Medical Technologies 13-x Rev. 0
onym/Abbreviation Definition SS subcritical assembly support structure AS subcritical assembly system sulfur hexafluoride S standby generator system NE SHINE Medical Technologies M special nuclear material C system, structure, and component E total dose equivalent E total effective dose equivalent Ls Temporary Emergency Exposure Limits 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 NE Medical Technologies 13-xi Rev. 0
onym/Abbreviation Definition SS uninterruptible electrical power supply system SS uranium receipt and storage system vacuum transfer system atmospheric dispersion factor NE Medical Technologies 13-xii Rev. 0
purpose of this section is to identify the postulated initiating events and credible accidents form the design basis for the irradiation facility (IF), which includes the irradiation units (IUs) supporting systems. Section 13b identifies the postulated initiating events and credible idents within the radioisotope production facility.
ign basis accidents were identified using the following sources of information:
- NUREG-1537 (USNRC, 1996) and the Interim Staff Guidance Augmenting NUREG-1537 (USNRC, 2012a);
- Process hazard analysis method within the integrated safety analysis process; and
- Experience of the hazard analysis team.
h identified accident scenario was qualitatively evaluated for its potential chemical or ological consequences. For accident scenarios with potential consequences that could eed the appropriate evaluation guidelines for worker or public exposure, controls were lied to ensure that the scenario is prevented or that consequences are mitigated to within eptable limits. For accident scenarios which are not prevented, the radiological or chemical sequences were quantitatively evaluated to demonstrate the effectiveness of the selected gative controls or shown to be bounded by other quantitative analysis.
quantitative analysis includes:
- 1) Identification of the limiting initiating event, initial conditions, and boundary conditions.
- 2) Review of the sequence of events for functions and actions that change the course of the accident or mitigate the consequences.
- 3) Identification of damage to equipment or the facility that affects the consequences of the accident.
- 4) Review of the potential radiation source term and radiological consequences.
- 5) Identification of safety controls to prevent or mitigate the consequences of the accident.
results of these analyses are provided in Section 13a3. The analyses identify those safety-ted structures, systems, and components (SSCs) and engineered safety features for each ident, and demonstrate that the mitigated consequences do not exceed the radiological ident dose criteria, described in Section 13a2.2.
2.1 ACCIDENT-INITIATING EVENTS AND SCENARIOS design basis accidents (DBAs) identified in this section are credible accident scenarios that ge from anticipated events, such as a loss of electrical power, to events that are still credible, considered unlikely to occur during the lifetime of the plant. The irradiation facility (IF) ximum hypothetical accident (MHA) is also defined to result in the bounding radiological sequences for the IF.
ed on the guidance provided in the Interim Staff Guidance (ISG) Augmenting NUREG-1537 NRC, 2012a), the following accident categories were used to identify potential accident uences:
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- Reduction in cooling (Subsection 13a2.1.3)
- Mishandling or malfunction of target solution (Subsection 13a2.1.4)
- Loss of off-site power (LOOP) (Subsection 13a2.1.5)
- External events (Subsection 13a2.1.6)
- Mishandling or malfunction of equipment (Subsection 13a2.1.7)
- Large undamped power oscillations (Subsection 13a2.1.8)
- Detonation and deflagration in the primary system boundary (Subsection 13a2.1.9)
- Unintended exothermic chemical reactions other than detonation (Subsection 13a2.1.10)
- System interaction events (Subsection 13a2.1.11)
- Facility-specific events (Subsection 13a2.1.12) effects of losses of electrical power and operator errors were considered as initiating events in the scope of the process hazard analysis (PHA) process and are therefore considered in each event category.
2.1.1 IF MAXIMUM HYPOTHETICAL ACCIDENT ccordance with the guidance in the ISG Augmenting NUREG-1537 (USNRC, 2012a), a tulated fission-product release with radiological consequences that exceed those of any ident considered to be credible is analyzed.
IF MHA is an accident scenario defined to result in the most limiting consequences for the et solution and associated fission products in the IF. Although the MHA is an accident nario, it does not need to have a credible or defined initiating event or accident progression, ept as necessary to evaluate the consequences. The MHA itself is therefore not a DBA; ever, it is used as a metric for understanding radiological risk from the facility.
SHINE facility is divided into two major process areas: the IF and the radioisotope duction facility (RPF). The IF includes eight irradiation units (IUs) each containing, among er components, a subcritical assembly system (SCAS) (including the target solution vessel V] and TSV dump tank), light water pool system (LWPS), and the TSV off-gas system GS). The TSV, TOGS, TSV dump tank, and associated components make up the primary tem boundary (PSB). The RPF consists of several process areas that prepare target solution, act and purify the radioisotope products, and process waste streams. The major process tems include the uranium receipt and storage system (URSS), target solution preparation tem (TSPS), target solution staging system (TSSS), vacuum transfer system (VTS), process sel vent system (PVVS), radioactive liquid waste storage (RLWS) system, and the radioactive id waste immobilization (RLWI) system. The RPF also includes the supercell which is prised of several internal cells, including the molybdenum extraction areas, purification as, an iodine and xenon purification and packaging (IXP) cell, PVVS equipment, and kaging areas, that form one hot cell structure.
ISG Augmenting NUREG-1537, Part 1 (USNRC, 2012a) identifies several possible MHAs could be considered. The ISG Augmenting NUREG-1537, Part 2 (USNRC, 2012b),
tion 13a2.1, indicates that the MHA should release fission products from the uranium target tion (fuel). As such, SHINE has selected the IF MHA based on accidents involving the PSB.
idents that only involve tritium are not considered for the MHA. SHINE has established the NE Medical Technologies 13a2.1-2 Rev. 1
eral potential MHA scenarios were considered, including:
- Energetic dispersal of contents of the PSB with bypass of the light water pool scrubbing capacity,
- Failure of the TOGS pressure boundary and release of some or all of the TSV radioactive gases into the TOGS cell,
- Complete loss of target solution inventory (e.g., TSV break),
- Man-made external event that breaches the PSB of more than one IU, and
- Facility-wide external event that breaches various systems containing radioactive fluids.
main production facility is designed to withstand external events such as tornado, seismic, man-made external events. The structure protects the equipment inside its seismic envelope external events. With this protection, it is not credible for an external event such as an raft impact, tornado, flood, earthquake, or tornado missile to initiate an accident involving a ty-related SSC on one or multiple IUs within the structure.
tiple non-seismic SSCs within the structure can still be affected by and initiate an accident to seismic events. The neutron drivers are nonsafety-related components within the IUs. The tron drivers do not contain fission products and are not part of the PSB. The failure of multiple tron drivers during a seismic event is evaluated in Subsection 13a2.1.6.
integrated safety analysis (ISA) process did not reveal potential interaction events between that could lead to one accident propagating to another unit. Therefore, scenarios that involve tiple IUs are not analyzed further.
2.1.1.1 Identification of Causes, Initial Conditions, and Assumptions IF MHA is a failure of the PSB leading to a release of TSV radioactive gases into the TOGS
. The failure of the TOGS pressure boundary is assumed to cause a failure of the TOGS to tion, which initiates the nitrogen purge system (N2PS). The N2PS causes the pressure in the GS cell to increase as the nitrogen gas also leaks into the TOGS cell. The MHA assumes that normal N2PS flow path from the TOGS through the PVVS system is completely blocked, ulting in a higher pressurization of the TOGS cell than would occur for the credible design is accident. Therefore, the MHA is a pressurized release from the PSB, which sufficiently nds credible radiological releases.
selection of this event was determined to be not credible based on a number of factors:
- PSB piping and valves are fabricated and installed according to codes and standards appropriate to their application and safety classification.
- Corrosion allowances on the pressure vessel and piping wall thicknesses ensure that corrosion expected over component lifetime does not impact the pressure retaining capability of the pressure boundary.
- The N2PS flow path from TOGS to the vent release point is designed with redundant valves in parallel to ensure a flow path is available.
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initial conditions and assumptions used to analyze the IF MHA described above are:
- The material-at-risk for this scenario is 100 percent of the iodine and noble gases circulating through TOGS at the end of the bounding operating cycle, as described in Section 13a2.2. The TOGS zeolite bed removes a portion of the gaseous iodine during operation.
- Only one IU is affected by the event.
- The integrated primary confinement leakage is within specifications to limit the release of radioactive materials from the cells.
- No credit is taken for the TOGS cell filtration (radiological ventilation zone 1 recirculating subsystem [RVZ1r]), which provides normal cooling and filtration and may continue to function following the accident.
- No credit is taken for the TOGS blowers and zeolite bed once the leak has occurred, which normally function to scrub iodine out of the gases present in the PSB.
- The radiological ventilation zone 1 exhaust subsystem (RVZ1e) redundant ventilation isolation from the IU cell are closed on detection of high radiation in the RVZ1e duct. A sufficient time delay is provided by design to prevent significant radioactive gases from exiting through this path prior to isolation.
- The flow rate of compressed nitrogen from N2PS into the TOGS cell is assumed to be twice the flow rate required for hydrogen mitigation.
2.1.1.2 General Scenario Description ilure in the PSB in the TOGS cell causes radioactive gases to be released from the PSB into TOGS cell. The release of radioactive material to the IF is mitigated by the primary finement boundary and by the radiation monitoring of the primary closed loop cooling system LS) expansion tank exhaust, which provide a signal to the TSV reactivity protection system PS) to perform an IU Cell Safety Actuation and isolate the RVZ1e ventilation flow path from IU cell. The N2PS actuates and causes nitrogen to flow through the affected PSB and out of break in TOGS into the TOGS cell. The blockage of the PVVS flowpath results in a higher ssurization of the TOGS cell.
pressurized flow of nitrogen and radioactive gases from the PSB to the IF and the ironment results in a long-term release.
main facility ventilation system (i.e., radiological ventilation zone 2 [RVZ2]) is isolated within econds of detectable accident conditions. However, it is assumed that the release is a und release through unfiltered leakage pathways through the facility leak paths (e.g., door s). An unfiltered ground release is conservative.
ailed analysis of the IF MHA is provided in Subsection 13a2.2.1.
2.1.1.3 Accident Consequences accident consequences associated with the IF MHA are evaluated further in section 13a2.2.1.
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excess reactivity insertion event during normal operations is identified as a potential initiating nt for accidents in the accident analysis. The ISA process also identified the potential for ess reactivity insertion during the startup process and post-irradiation mode of the TSV as narios to be evaluated.
ee operating modes that have potential reactivity impacts were evaluated for the TSV:
- Mode 1 - Startup Mode: filling the TSV
- Mode 2 - Irradiation Mode: operating mode (neutron driver active)
- Mode 3 - Post-Irradiation Mode: TSV dump valves open ess reactivity insertion events can challenge the integrity of the PSB by causing increased er density, temperature, and pressure.
SCAS is designed to operate in a subcritical state without available excess reactivity.
ctors normally have engineered reactivity control mechanisms and load excess reactivity into core to accommodate power defect, fuel burnup, and uncertainty in keff. There are no ctivity control systems in the SHINE system. Analyzing the inadvertent withdrawal of the most ctive control element as performed for reactors is not possible. SHINE will not perform eriments with the IUs, so there are no reactivity effects from experiment malfunctions.
the subcritical assembly being driven by the neutron driver (such as in Mode 2), excess ctivity insertion (i.e., reactivity inserted beyond planned operations) has similar effects to ess reactivity insertions in a reactor, including increases in power, temperature, and gas eration. As substantial power can be generated even if reactivity remains subcritical in a en system, the effects of excess insertions of reactivity were considered in the safety lysis.
the subcritical assembly, when it is not being driven by the neutron driver (such as in Mode 1 ode 3 prior to dumping), excess reactivity insertion could lead to inadvertent criticality and lanned fission power generation, temperature increase, and gas generation.
assembly is designed to be in a subcritical condition during each mode of operation, with tiple safety controls to prevent or mitigate an excess reactivity insertion or inadvertent cality. The potential for an inadvertent criticality is greater during fill operations. However, as ussed in the following subsections, controls are in place to safely limit excess reactivity rtions.
dvertent criticality events outside the IF (i.e., in the RPF) are prevented by the nuclear cality safety program, as described in Section 6b.3.
2.1.2.1 Identification of Causes, Initial Conditions, and Assumptions ISA process and guidance in the ISG Augmenting NUREG-1537 (USNRC, 2012a) identified following postulated initiating events and scenarios that could lead to an excess reactivity rtion or power transient during operation:
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- Target solution temperature reduction during irradiation (e.g., excessive cooldown)
- High reactivity and power due to high neutron production at cold conditions
- Moderator addition due to cooling system malfunction (e.g., cooling water in-leakage)
- Additional target solution injection during fill/startup and irradiation operations
- Realistic, adverse geometry changes
- Reactivity insertion due to moderator lumping effects (e.g., voiding in the cooling system)
- Inadvertent introduction of other materials into the TSV (e.g., uranium solids introduction or precipitation of uranium from target solution)
- Concentration changes of the TSV target solution (e.g., through boiling or evaporation)
- Failure to control temperature during 1/M measurements at startup following initial conditions or assumptions are made with respect to the Mode 1, Mode 2, and de 3 operations:
- TSV is filled to an approximate keff of [ ]PROP/ECI at a cold startup temperature range of 59°F to 77°F (15°C to 25°C).
]PROP/ECI during steady-state irradiation operations. The TSV is designed to operate with the neutron driver in service with a source strength yielding a maximum value of 125 kilowatts (kW) power within the target solution.
- During irradiation, the TSV is designed to operate with a maximum average temperature below 176°F (80°C).
- The target solution has high negative temperature and void coefficients, as described in Section 4a2.6).
- The TRPS is designed to dump the TSV on high neutron flux level (source, wide range, and time-averaged) to protect the PSB.
- Redundant, fail-open TSV dump valves ensure target solution can be dumped and are cycled each irradiation cycle.
- The TRPS is designed to dump the TSV on high PCLS temperature, low PCLS temperature, and low PCLS flow.
2.1.2.2 General Scenario Descriptions general scenarios for each of the potential excess reactivity insertion events listed in section 13a2.1.2.1 are discussed in detail below.
nario 1 - Increase in the Target Solution Density During Operations TOGS regulates the pressure in the PSB. During irradiation operations, pressurization of the et solution could occur if there is a malfunction in TOGS.
ystem pressurization could also occur following a deflagration or detonation in the PSB due to rogen accumulation during or following irradiation operations. The causes of this event are cribed in Subsection 13a2.1.9. Related reactivity effects are considered in this section.
eased pressure in the TSV would cause the target solution to be compressed as void space reases. This would cause an increase in reactivity in the SHINE system. With a fixed neutron rce, the reactivity increase during irradiation operations leads to a power increase.
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alculated as less than 240 kW.
essurization is sustained, the event is terminated by the TRPS high time-averaged neutron trip, which de-energizes the neutron driver and opens the TSV dump valves. This trip results rapid reduction in the power generation in the TSV due to the loss of the neutron source, wed by the reactivity decrease from draining the target solution. The high time-averaged tron flux trip prevents damage to the PSB. No damage to the PSB occurs and there are no ological consequences.
nario 2 - Target Solution Temperature Reduction During Fill/Startup TSV is cooled by the PCLS. The PCLS is a closed loop that circulates cooling water [
]PROP/ECI past the TSV walls to remove heat erated in the target solution during normal irradiation. The light water pool provides passive ling of the TSV dump tank to remove heat generated during shutdown operations. The light er pool contains no dedicated cooling, and is cooled by contact with the PCLS-cooled ponents.
excessive cooldown could occur if the PCLS malfunctions and overcools the target solution in TSV, adding positive reactivity due to the negative temperature coefficient. An overcooling nt is prevented by the TRPS trip on low PCLS temperature.
ing fill/startup, the limiting scenario occurs when the TSV has been filled in Mode 1 to normal tup keff values. Then, the system is transitioned to Mode 2, and prior to accelerator rations, a failure of the PCLS occurs resulting in temperature decreasing in the TSV. The LS temperature decreases from 77°F to 59°F (25°C to 15°C) and results in a maximum ctivity insertion of [ ]PROP/ECI. This corresponds to a minimum volume of 4.2 percent, ch is less than the minimum volume margin to critical used during fill. The event is discussed ubsection 4a2.6.3. The reactivity increase is small and the system remains subcritical. No age to the PSB occurs and there are no radiological consequences.
ater temperature changes are prevented by the TRPS IU Cell Safety Actuation on high and PCLS temperatures.
nario 3 - Target Solution Temperature Reduction During Irradiation ing irradiation operations, the limiting target solution cooldown scenario occurs when the TSV perating normally at licensed power of 125 kW and then PCLS temperature instantaneously reases from 25°C to 15°C. Given the thermal mass of the PCLS, the instantaneous change is nservative approximation. The thermal mass of the TSV and target solution is slow to pond, allowing sufficient time for the TRPS IU Cell Safety Actuation on high time-averaged tron flux at 104 percent of licensed power. This drains the target solution to the TSV dump k, terminating the event.
ater PCLS temperature changes are prevented by the TRPS IU Cell Safety Actuation on high low PCLS temperature, resulting in a dump of the target solution and termination of the tron generation by the neutron driver assembly system (NDAS). The draining of the target NE Medical Technologies 13a2.1-7 Rev. 1
nario 4 - High Power Due to High Neutron Production and High Reactivity at Cold Conditions gh reactivity and power event can occur due to excess tritium injection into the NDAS during conditions. This can occur as a result of a tritium purification system (TPS) control system or ponent failure during startup that injects excess tritium before the TSV is at operating perature. The TRPS initiates an IU shutdown on high wide range neutron flux.
gh reactivity and power event can also occur if the NDAS neutron production drops to a lower than expected due to focusing issues, electrical arcing, or other malfunctions. This loss of tron source during irradiation results in a decrease in void fraction and a target solution ldown in the TSV. If the NDAS neutron production were to rapidly return to full output sequent to a loss of void fraction and cooldown, excessive power generation could occur that ld challenge target solution power density limits or PSB integrity.
prevent excessive power pulses at the start of the irradiation cycle, TPS permissives prevent sitioning from Mode 1 to Mode 2 until the [
]PROP/ECI. This prevents the driver from producing excessive neutrons current with high system reactivity.
prevent excessive power pulses during driver ramp-up as the target solution has not yet ched operating temperature, the rate of tritium concentration increase in the NDAS target mber is limited by the achievable flow rate of tritium from the TPS. This design characteristic assive and designed to prevent a TPS failure that could result in rapid tritium concentration ease in the target chamber.
prevent excessive power pulses during irradiation, the TRPS de-energizes the NDAS high age power supply (HVPS) redundant breakers on a driver dropout signal after
]PROP/ECI of low power range neutron flux in Mode 2 are detected. This prevents the er from producing excessive neutrons concurrent with high system reactivity.
described in Subsection 4a2.6.3, the cooldown and void loss during this event creates a ctivity insertion of up to [ ]PROP/ECI from loss of void and up to [ ]PROP/ECI in PROP/ECI
] from cooldown. The final keff of the system remains below the initial startup of the system. It is assumed the driver instantaneously returns to full output. The resulting k power density and return to equilibrium from the event is described in section 4a2.6.3.3.
TRPS Driver Dropout signal safely prevents power generation levels that would exceed et solution operating limits or damage the PSB.
her analysis is provided in Subsection 13a2.2.2.
nario 5 - Moderator Addition Due to Cooling System Malfunction PCLS is a closed loop that circulates cooling water [
]PROP/ECI past the TSV and neutron multiplier walls to remove heat erated in the TSV and neutron multiplier during normal irradiation and shutdown operations.
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denser demister unit or recombiner condenser unit, leading to radioisotope process facility ling system (RPCS) water ingress into the TSV.
er ingress into the TSV dilutes the target solution. A dilution event such as this would lower overall reactivity of the target solution due to the high hydrogen to uranium ratio in the target tion (target solution is over-moderated).
e break were to occur near the surface of the target solution or in TOGS, it is possible that the er could fill the upper space of the TSV between the solution level and the overflow lines. This ld create a reflector. The maximum potential reactivity effect of a reflector forming in this nner has been evaluated assuming no mixing. In Mode 1, the limiting event occurs after the has already been filled to normal startup keff values. The reactivity insertion is not significant ugh to drive the system to criticality. Excess neutron flux levels during Mode 1 are prevented he TRPS IU Cell Safety Actuation on high source range flux, resulting in a dump of the target tion to the TSV dump tank.
ode 2, the reactivity increase is prevented from resulting in excessive power generation by IU Cell Safety Actuation on high time-averaged neutron flux. In a TSV overflow condition, ess water and target solution drain to the TSV dump tank via overflow lines. The TRPS ates an IU Cell Safety Actuation and IU Cell Nitrogen Purge on dump tank low-high level in de 2. The water ingress could affect the proper functioning of TOGS by flooding sweep gas paths, but the nitrogen purge ensures that hydrogen gas concentrations remain within eptable limits. No damage to the PSB occurs and there are no radiological consequences.
nario 6 - Additional Target Solution Injection during Fill/Startup and Irradiation Operations ing Mode 2 operations, target solution injection from the target solution hold tank is not dible due to its isolation from the TSV using redundant isolation valves, and the fact that the is located higher than the target solution hold tank, thus preventing an accidental gravity-en transfer of target solution to the TSV during operation. Target solution in the TSV fill lift is drained back to the hold tank following the fill process. No damage to the PSB occurs and e are no radiological consequences.
ing fill/startup operations in Mode 1, excess fissile material is prevented from being added by eral controls. These controls are described in two primary groups: (1) physical design and mistry controls that prevent excess fissile material in the TSV, and (2) prevention of operator rs during the fill process that lead to excess fissile material. The limiting scenario is described wing the controls.
sical Design and Chemistry Controls first control is in the preparation of the target solution itself, where uranium enrichment is pendently verified by SHINE and concentration in the target solution is controlled to within uired accuracy levels. These controls ensure that the fissile material per volume of target tion is prepared within design calculation parameters.
uranium concentration of target solution is verified within acceptable range after preparation new batch and after making adjustments to an existing batch, prior to transferring the batch NE Medical Technologies 13a2.1-9 Rev. 1
ermined to be safe.
physical placement of the TSV above the target solution hold tank and TSV dump tank vents inadvertent draining of fissile material from these tanks to the TSV during the fill/startup rations.
ther control is the inherent limitations on fill rate. This limitation is due to the limited gravity-en head of the fill process combined with the high hydraulic resistance of the fill path.
vention of Operator Errors During the Fill Process ing the fill process, the operators use fill procedures following the 1/M measurement method containing hold points at certain volume levels to verify expected system behavior. The fill cedures limit the size of the solution addition steps the operators can use to one-half of the me to predicted critical. This reduces fill increments as keff increases until the desired critical multiplication is reached. These procedural controls are fundamentally similar to ctor startup processes that routinely and safely start up reactors.
ddition to the procedural controls, the TRPS stops inadvertent target solution injection during fill upon detection of high source range count rates in Mode 1. The TRPS initiates an IU Cell ety Actuation to close the target solution fill valves and opens the TSV dump valves upon ection of high count rates.
ough not credited, a manual trip by the operators also causes the TSV solution to transfer to TSV dump tank should an unsafe condition arise.
hermore, fill valve sequence controls in TRPS ensure proper neutron flux stabilization occurs ween solution addition steps. Above pre-determined neutron flux levels in Mode 1, the TRPS ts the time the fill valve can be open so that rate of target solution addition is controlled. This safety-related defense-in-depth control allows the delayed neutron precursor population time each steady state. Subsection 7.4.3.2 provides a detailed description of the TRPS Fill Stop tion.
analyzed event is an inadvertent target solution injection after the system has already been d to normal startup keff values. This could be caused by operator errors or malfunction of the cess integrated control system (PICS). An injection of target solution occurs at the maximum allowed by the fill valve sequencing. The TRPS trips the system on high source range flux, ning the target solution faster than it can be filled. Reactivity is rapidly decreased in the AS, terminating the event.
nsient analysis of this event is presented in Subsection 4a2.6.3. No damage to the PSB urs and there are no radiological consequences.
olution addition event not crediting the operation of the TRPS Fill Stop function has also been lyzed. The resulting power increase is small. No damage to the PSB occurs and there are no ological consequences.
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ause of the liquid nature of the target solution, the variability of TSV core geometry is sidered.
TSV, subcritical assembly support structure (SASS), TSV dump tank, piping, and associated p valves are of robust construction and are seismically-qualified. In addition, the PSB and SS are designed to withstand the pressures resulting from the maximum credible deflagration, significant geometry changes are prevented during that event.
ation is considered in Subsection 4a2.7.3 and discussed as having no significant reactivity ct.
sideration is given to the potential change in target solution spatial density caused by the ation and movement of voids. This can cause an insertion of reactivity event as voids form collapse, but does not lead to uncontrolled/undamped power oscillations (see section 4a2.6.1.4).
shing of the target solution due to seismic acceleration is the limiting event for geometry nges. This event has been analyzed by assuming a range of sloshing amplitudes, for the imum core volume and the nominal core. The effect generally results in negative reactivity cts in the TSV due to the geometry of the core. The sloshing distributes the core away from a e compact form, increasing neutron leakage. The event does not result in significant eases in power or challenging the safety limits of the system. No controls are needed to gate sloshing of the target solution. No damage to the PSB occurs and there are no ological consequences.
ution redistribution from vibrations is expected to be minimal and is bounded by sloshing.
nario 8 - Reactivity Insertion Due to Moderator Lumping Effects PCLS is a closed loop that circulates cooling water [
]PROP/ECI past the TSV walls to remove heat generated in the TSV during mal irradiation operations. The cooling system design and operating characteristics preclude ificant reactivity effects due to moderator changes in the subcritical assembly during ration. The PCLS is operated far from boiling conditions, and there is no scenario where ing occurs in the PCLS. The PCLS passes through straight-through vertical cooling channels, ch largely mitigate collection of voids and moderator lumping. Voids simply exit the top of the ling channels. PCLS contains an air separator to remove entrained air.
d formation within [ ]PROP/ECI changes the moderation profile in the TSV. A ulation of the expected reactivity changes due to voiding out the PCLS from nominal coolant perature and density to a fully-voided cooling system was performed. Voids were assumed to ur in the cooling channels around the TSV [ ]PROP/ECI. The ulation was performed at cold (Mode 1) startup conditions and hot (Mode 2) irradiation ditions.
ults of the calculation show that for the PCLS, there is a positive insertion of reactivity with orm voiding in the PCLS. The analysis shows that for a uniform voiding of 20 percent, ctivity changed by approximately [ ]PROP/ECI in Mode 1 and [ ]PROP/ECI in NE Medical Technologies 13a2.1-11 Rev. 1
ercent voiding is very small (i.e., approximately [
]PROP/ECI).
en the inherent design of the and the air-water separator in the PCLS, there is no significant ct from moderator lumping. Additional design features to prevent cooling channel voiding are ussed in Subsection 5a2.2.2. No damage to the PSB occurs and there are no radiological sequences.
nario 9 - Inadvertent Introduction of Other Materials into the Target Solution Vessel chemical control of the target solution is performed during the preparation and adjustment of solution in the RPF. Once the target solution is prepared for use in the TSV, there are no itional chemical control additives in the IF.
significant pH changes are expected during irradiation due to the stability of sulfuric acid er irradiation.
le other materials are not normally added to the TSV, the ISA process has evaluated process ets that could lead to inadvertent introductions. The inadvertent introduction of other erials into the TSV could come from: (1) sources external to the PSB, or (2) sources internal he PSB itself.
erial Entering the TSV from Sources External to the PSB TSV fill lines are isolated once the TSV is filled and ready for irradiation operations. There is need to add any chemicals to control the chemistry of the target solution during the irradiation e.
only systems that significantly interact with the SCAS during irradiation operations are the GS, NDAS, light water pool, and PCLS. The TOGS can adjust pressure and oxygen centrations through gas removal and additions to the PSB. These gas space changes have effect on reactivity beyond PSB pressure change reactivity effects, which are discussed in tion 4a2.6. The PCLS and light water pool are unable to add material to the TSV, except for er ingress scenarios, which are described in Subsection 13a2.1.2.2, Scenario 5.
arding the target solution itself, uranium solids used in the target solution preparation cess are prevented from reaching the TSV by a filter in the TSPS process.
er could potentially be introduced into the TSV through a leak from the PCLS, light water l, or from the RPCS-cooled components in TOGS. Dilution of the target solution in the TSV is ussed in Subsection 13a2.1.2.2, Scenario 5.
erial Entering the TSV from Sources Internal to the PSB ISA process has identified two potential sources of uranium solids entering the TSV and ulting in reactivity addition: uranyl salt crystal buildup in the TSV or TOGS components and cipitation of uranium solids.
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erwise dislodged and reenter the TSV. The buildup of salt crystals in the TSV is not expected to the high humidity of the TSV and the cold walls of the TSV. In addition, periodic inspection he TSV is performed which would allow for detection of salt crystal buildup.
lt crystals did accumulate, their release could lead to an unexpected reactivity increase due he increase in fissile material in the target solution. To quantify reactivity effects, it is tulated that a piece of deposited salt containing 100 grams of uranium is dislodged from the er TSV surfaces and falls into the target solution. The re-dissolution of the salt adds roximately [ ]PROP/ECI of reactivity to the system. This reactivity effect does not result gnificant consequences and does not lead to an inadvertent criticality. If additional salt pieces e to continue to enter the TSV, they could continue to re-dissolve and lead to further centration increases, and power could increase in the TSV. The TRPS would dump the target tion on high time-averaged neutron flux, terminating any reactivity increase. The TSV dump k is favorable geometry at any uranium concentration. No damage to the PSB occurs and e are no radiological consequences.
second postulated scenario is precipitation of uranium solids from the solution. Precipitation ranium solids due to uranyl peroxide formation is possible in aqueous reactors. In the SHINE tem, chemistry, power density, and temperature limits have been placed on the target tion as described in Subsection 4a2.6.3. Given these limits, no significant precipitation is ected. For transient events, precipitation has not been seen in transient operations of historic nyl sulfate systems. Therefore, the dump of the target solution by TRPS on high time-raged or wide range neutron flux occurs prior to significant precipitation developing in the et solution.
accumulation of small amounts of precipitation over many cycles has been considered. This ld lead to chemical effects on the TSV surface, which may have the potential to lead to a re of the PSB. A failure of the PSB is analyzed in Subsection 13a2.1.4.
nario 10 - Concentration of the TSV Target Solution ISA process identified postulated scenarios where the uranium concentration of the target tion could increase. One identified scenario requiring control was the TOGS pressure control re leading to excess evaporation. The other identified scenario requiring control was failure OGS to return condensate to the TSV.
GS pressure control could fail during irradiation operations and cause lower pressure (higher uum) in the TSV, which could increase solution evaporation and/or cause boiling. This could ult in increased uranium concentrations and a reactivity increase.
GS condensate return lines could clog, leading to increased holdup of condensate in TOGS or rsion of condensate to the TSV dump tank. Reduction of condensate return would lead to eased target solution uranium concentration and a reactivity increase.
pressure control failure scenario is prevented through redundant TOGS vacuum relief valves prevent excess vacuum in the PSB. Redundant relief valves protect the PSB from damage results in no radiological consequences. The reduction in condensate return scenario is vented through the TRPS IU Cell Safety Actuation on high power range neutron flux.
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ISA process postulates that a failure in the PCLS occurs during the startup process, which ults in high target solution temperature and errors in the 1/M measurements during the fill cess. These errors could be non-conservative and lead to an increase in reactivity during the rocess. This scenario is prevented through the TRPS IU Cell Safety Actuation on high source ge neutron flux or high PCLS temperature. Following the TRPS trip, the target solution dumps he TSV dump tank, decreasing reactivity and resulting in safe shutdown of the TSV.
ause each of these events has preventative measures in place, there are no radiological sequences.
2.1.2.3 Accident Consequences releases are expected to occur as a result of insertion of excess reactivity events described ve. However, additional discussion associated with the most limiting scenario (Scenario 4 -
h Reactivity and Power Due to High Neutron Production at Cold Conditions) is provided in section 13a2.2.2.
2.1.3 REDUCTION IN COOLING subsection discusses the reduction in cooling in the SCAS. The following components were luated:
- The neutron multiplier
- The TSV containing uranyl sulfate solution
- The TSV dump tank containing uranyl sulfate solution se components are cooled by the PCLS during irradiation operations to maintain a target tion average temperature of not more than 176°F (80°C) at 125 kW of heat generation in the
. PCLS rejects heat to the RPCS, which in turn is cooled by the process chilled water system HS). Because the PCLS, RPCS, and PCHS cooling pumps are driven by off-site power, a of coolant flow occurs due to power failure and could occur due to failure of a pump, vertent valve closure, or a pipe break.
oling loop circulation flow is lost, the target solution is dumped to the TSV dump tank. The t water pool removes decay heat from the TSV dump tank by passively absorbing the heat in pproximately 14,900 gallons (56,400 L) water volume.
2.1.3.1 Identification of Causes, Initial Conditions, and Assumptions SCAS is cooled by the PCLS and the light water pool. The PCLS is a closed loop that ulates cooling water through [ ]PROP/ECI PCLS ling water also flows around the TSV and neutron multiplier walls to remove heat generated e target solution and neutron multiplier during normal irradiation and shutdown operations.
tion 5a2.2 specifies that the PCLS is designed to remove 580,000 Btu/hr (170 kW).
light water pool provides a large heat capacity for passively rejecting heat from the TSV p tank during shutdown operations.
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- Loss of normal power
- Loss of or reduced cooling of PCLS due to:
- Flow blockage
- Pump malfunction
- Operator error
- Pipe break
- Valve closure
- Loss of RPCS or PCHS se failures create two possible scenarios for reduction in cooling evaluation:
- 1) Loss of normal power resulting in loss of PCLS and de-energized neutron driver.
- 2) Loss of PCLS cooling due to blockage, malfunction, or operator error (neutron driver remains operating).
nario 1 - Loss of Normal Power scenario results in a loss of coolant flow in the PCLS cooling loop. The neutron driver does function without off-site power, and therefore, the irradiation process is stopped. No further t is generated in the target solution with the exception of power from delayed neutrons and ay heat.
nario 2 - Loss of PCLS Cooling scenario assumes a loss of PCLS flow without a LOOP, resulting in continued operation of neutron driver. Full power continues to generate heat.
2.1.3.2 General Scenario Descriptions section 13a2.1.3.1 identifies two accident scenarios requiring evaluation of the temperature ponse of the light water pool and the target solution.
nario 1 - Loss of Normal Power loss of PCLS cooling flow is a result of a loss of normal power. The loss of neutron driver er terminates the neutron source production and reduces heat generated in the target tion prior to draining of the solution.
RPS signal initiates a Driver Dropout on low PCLS flow or high temperature, which opens the AS HVPS breakers, preventing a restart of neutron production. The TSV temperature ease prior to the TRPS dump of the target solution to the TSV dump tank introduces negative ctivity. The loss of PCLS flow also results in an IU Cell Safety Actuation after 180 seconds.
undant TSV dump valves open due to the TRPS actuation, draining the target solution to the dump tank located in the light water pool. The light water pool is the heat sink for removal of ay heat from the target solution in the TSV dump tank. Thermal analysis of the TSV has been ormed and shown that the target solution does not reach boiling conditions during this event, NE Medical Technologies 13a2.1-15 Rev. 1
nario 2 - Loss of PCLS Cooling ss of PCLS cooling with continued operation of the neutron driver is assumed. Failures that result in a loss of PCLS cooling are described in Subsection 13a2.1.3.1.
w PCLS flow or high temperature signal initiates a Driver Dropout, which causes the NDAS PS breakers to open. This terminates the irradiation process by the accelerator.
r a 180 second delay, an IU Cell Safety Actuation is initiated, opening the redundant TSV p valves and draining the target solution to the TSV dump tank. The light water pool is the t sink for removal of decay heat from the target solution in the TSV dump tank. Thermal lysis of the TSV has been performed and shown that the target solution does not reach ing conditions during this event, and no damage to the PSB occurs. The TSV dump tank is igned to maintain the target solution subcritical.
2.1.3.3 Accident Consequences accident consequences associated with reduction in cooling events are evaluated further in section 13a2.2.3.
2.1.4 MISHANDLING OR MALFUNCTION OF TARGET SOLUTION TSV uses a liquid target solution that generates fission products that are contained by the B. The PSB consists of the TSV, the TSV dump tank, the TOGS, and associated connected ng and piping components. The accidents involving the mishandling or malfunction of the et solution within the IF, including a failure of the PSB, are analyzed here. The accidents lving mishandling or malfunction of target solution within the RPF are analyzed in section 13b.2.4.
hin the boundaries of the IF, the target solution is contained in the TSV, the TSV dump tank, associated connected piping and piping components.
rtion of excess reactivity and inadvertent criticality events involving the target solution are ussed in Subsection 13a2.1.2.
2.1.4.1 Identification of Causes, Initial Conditions, and Assumptions handling and malfunction of target solution events fall broadly into two categories. The first is s or leaks that cause target solution to migrate into unintended locations. The second gory is changes in the physical or chemical form of the target solution that results in adverse cts. Within this category, three specific initiating events are considered: (1) failure to control of the solution, (2) failure to control solution temperature, and (3) failure to control solution ssure. These two categories of initiating events were used in the hazard evaluation process to rm the selection of appropriate initiating events for the SHINE system.
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- Spills or leakage from the TSV and process tanks
- Excessive cooling of target solution
- Precipitation of the target solution
- Failures of valves, piping, or tanks
- Failure to control pressure which initiates target solution boiling and impacts target solution concentration
- Operator errors or equipment failures resulting in inadvertently overflowing tanks or misdirecting flow ure to control pH of the target solution in the IF results in potential excessive corrosion and ssure boundary failure events, as described in this subsection, and potential for precipitation nts as described in Subsection 13a2.1.2. Failure to control temperature or pressure of target tion are also described in Subsection 13a2.1.2.
nts involving the failure to control pH during solution preparation or adjustment are discussed hapter 13b.
initial conditions and assumptions associated with mishandling or malfunction of target tion include:
- The PSB does not contain significant sources of pressure. Leakage between the PSB and the light water pool will normally flow from the pool to the PSB should a break occur.
- The primary confinement boundary isolates the PSB from the rest of the facility by robust walls, ceiling, and floor.
- Penetrations for piping, ducts and electrical cables, and shield plugs are sealed to limit the release of radioactive materials from the facility. Integrated leak rate from the primary confinement boundary (see Subsection 6a2.2.1) is less than that assumed in the dose analysis.
- The primary confinement is cooled by a recirculating air ventilation system. The primary confinement is ventilated to RVZ1e through the PCLS expansion tank.
- IF tanks and piping that have the potential to contain fissile material, except the TSV, are designed with passive measures that prevent an inadvertent criticality with the most reactive uranium concentration.
- Drains that lead from the pipe trenches and tank vaults are designed with a geometry that prevents an inadvertent criticality of the leaked target solution.
2.1.4.2 General Scenario Descriptions re are several types of scenarios that are identified as mishandling or malfunction of the et solution within the IF: (1) failure of the PSB below the level of the light water pool, failure of the TSV-to-PCLS pressure boundary resulting in in-leakage to the TSV, (3) failure of RPCS-to-PSB interface, (4) failure of the TSV-to-PCLS pressure boundary resulting in target tion leakage to the PCLS, (5) failure in the TOGS causes high pressure in the TSV during fill de, and (6) target solution leakage into a valve pit. Each of these scenarios and their potential ses is discussed below:
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ilure of a PSB component below the water line of the light water pool may be caused by essive corrosion of PSB components. This failure results in water in-leakage to the primary tem from the light water pool. The water in-leakage fills the dump tank, TSV, and TOGS with ixture of target solution and pool water. Potential consequences of the flooding of the PSB ude:
- An inadvertent criticality within the TOGS or,
- A deflagration of hydrogen gas in the TSV, TSV dump tank, or TOGS headspace due to insufficient sweep gas flow iticality in the TOGS could occur from target solution intrusion into the TOGS system.
icality in the TOGS is prevented by the favorable geometry of the TOGS components, as cribed in Section 4a2.8.
sequences related to hydrogen deflagrations are discussed in Subsection 13a2.1.9.
ense-in-depth protections in place to prevent a failure of the PSB below the level of the light er pool are:
- control of solution pH through target solution sampling in the TSPS and target solution hold tank;
- a 30-year corrosion allowance in the PSB component design; and
- chemistry monitoring of the PCLS to limit corrosion (see Section 5a2.5).
nario 1b - Failure of the PSB Resulting in Target Solution Leakage into the Light Water Pool ilure in the PSB below the light water pool surface may also result in target solution leakage the primary system to the light water pool. The target solution mixes with the pool water and le gases, volatile fission products, and particulates evolve into the IU cell gas space. Some of radionuclides would then leak through the primary confinement boundary into the building then into the environment. The consequences of leakage of target solution into the light er pool are mitigated by the primary confinement boundary, which keeps the doses within eptable levels. The dose consequences of this accident scenario are analyzed in section 13a2.2.4.
nario 2 - Failure of the TSV-to-PCLS Pressure Boundary Resulting in In-Leakage to the TSV ilure of the PSB between the TSV and the PCLS may be caused by excessive corrosion of B components. This failure generally results in water in-leakage to the primary system from PCLS. The water in-leakage fills the TSV dump tank, TSV, and TOGS with a mixture of target tion and PCLS water. Potential consequences of the flooding of the PSB include an vertent criticality within TOGS or deflagration of hydrogen gas in the TSV headspace or GS due to insufficient sweep gas flow. Criticality in the TOGS is prevented by the favorable metry of the TOGS components, as discussed in Section 4a2.8.
sequences related to hydrogen deflagrations are described in Subsection 13a2.2.9.
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- control of solution pH through target solution sampling in the target solution hold tank;
- a 30-year corrosion allowance in the PSB component design;
- chemistry monitoring of the PCLS to limit corrosion (see Section 5a2.5).
nario 3 - Failure of the RPCS-to-PSB Interface ilure of the RPCS pressure boundary in the TOGS may be caused by excessive corrosion of PSB in a TOGS condenser. This failure results in water in-leakage to the primary system the RPCS. The water in-leakage fills the TSV dump tank, TSV, and TOGS with a mixture of et solution and RPCS water. Potential consequences of the flooding of the PSB include an vertent criticality in TOGS or deflagration of hydrogen gas in the TSV headspace or TOGS to insufficient sweep gas flow. Criticality in the TOGS is prevented by the favorable geometry he TOGS components, as discussed in Section 4a2.8.
sequences related to hydrogen deflagrations are discussed in Subsection 13a2.1.9.
nario 4 - Failure of the TSV-to-PCLS Pressure Boundary Resulting in Target Solution kage to the PCLS kage from the primary system into the PCLS due to a failure of the PSB between the TSV the PCLS is an additional concern. This failure results in: (1) a potential release of target tion into the PCLS cooling room with potential for higher dose to workers or the public, or a criticality accident in PCLS equipment. Normally the PCLS is at higher pressure than the
, so water will flow from the PCLS into the TSV. However, once pressure equilibrium is blished, target solution could leak into the PCLS. The protections in place to prevent and gate a failure of the PSB between the TSV and PCLS are PCLS isolation supply and return es, radiation detection on the RVZ1e exhaust from the PCLS expansion tank, and redundant ation dampers on the RVZ1e exhaust from the PCLS expansion tank. Target solution leakage the PCLS will result in radioactive gases entering the PCLS expansion tank, flowing past the ation detection in the RVZ1e exhaust duct, and initiating an IU Cell Safety Actuation including ation of the PCLS isolation valves and the RVZ1e exhaust duct.
defense-in-depth, the failure of the pressure boundary may first result in in-leakage and rflow into the TSV dump tank, which is detected with the level detection in the TSV dump
. The TRPS then closes the PCLS isolation valves and RVZ1e isolation dampers, stopping ential transfer of target solution to the PCLS and reducing the source of water that could enter PSB, and isolating the ventilation exhaust from the IU cell.
itional defense-in-depth measures are also in place to avoid a leak and detect leaks, which ude:
- control of solution pH through target solution sampling in the target solution hold tank;
- chemistry controls of PCLS to limit corrosion (see Section 5a2.5); and
- conductivity instrumentation in PCLS, which detects intrusion of target solution.
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ause of the system characteristics and preventative controls in place, further analysis is not uired.
nario 5 - Failure in the TOGS Causes High Pressure in the TSV during Fill Mode ilure by the TOGS to control pressure, and a resulting pressure increase during TSV filling rations, may result in a backflow of target solution. Target solution may flow through the fill into the TSV fill lift tank, into the VTS header, and into the VTS buffer tank. This failure entially results in radiological exposures to workers or a criticality accident in non-favorable-metry components in the VTS.
protection in place for this scenario is the configuration of the TSV fill line to prevent ificant volume of target solution from backflowing from the TSV into the VTS lift tank. The fill line connects to the TSV with an air gap. The connection is located at the approximate ation of the TSV overflow lines. The fill line is sloped to allow it to drain after fill operations e occurred. Therefore, no significant volume of target solution will backflow from the TSV to VTS lift tank in the event of pressurization of the TSV.
ense-in-depth measures are also present to mitigate this scenario, which include:
- the VTS vacuum valve to lift tank closes from high liquid level in the lift tank, and
- a drain valve for the buffer tank opens and drains to RLWS if a high level in the buffer tank is detected.
ause of the system characteristics and preventative controls in place, further analysis is not uired.
nario 6 - Target Solution Leakage within a Valve Pit pe or valve failure in the valve pit may be caused by overpressurization due to thermal ansion of target solution in an isolated section of piping. This pipe or valve failure results in age of target solution from the system into the valve pit, which subsequently could result in:
ncreased worker or public dose, or (2) a criticality accident in the valve pit. The protections in e to mitigate the consequences of target solution leakage within a valve pit are: (1) drip pans drains to the radioactive drain system (RDS), which prevent accumulation of solution within valve pit and prevent criticality, and (2) valve pit shielding and confinement for fission ducts that could result from leakage, reducing potential dose to workers and the public.
ause this piping is potentially located in either the IF or the RPF, this event and associated e consequences is further analyzed in Chapter 13b.
2.1.4.3 Accident Consequences release of target solution from the PSB to the light water pool or connected process systems ults in potential radiological exposure to workers and the public. The accident consequences ociated with the mishandling or malfunction of target solution are evaluated further in section 13a2.2.4.
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OOP can occur for a variety of reasons related to the reliability and operation of the smission system, stress during peak grid load conditions, severe weather effects from high d, tornado, or ice and snowstorms, a seismic event, or equipment failure in the supplying station. It may also be a result of failure or malfunction of the facility normal electrical power ply system (NPSS) such as the facility transformers or switchgear. This may result in a partial omplete LOOP to the facility.
tial electrical power may also be lost resulting in partial system losses. System or equipment res due to partial losses of electrical power within the facility are discussed under other ident analysis sections (e.g., Subsection 13a2.1.3). The partial loss of power scenarios are sidered and bounded by the accident scenarios described in this section.
the purposes of this discussion, it is assumed that a complete loss of off-site alternating ent (AC) power occurs from causes that are external to the SHINE facility or common cause res in the NPSS. Consequences of a complete LOOP to the facility are presented in section 13a2.2.5.
2.1.5.1 Identification of Causes, Initial Conditions, and Assumptions electrical power systems that support the SHINE facility are described in detail in Chapter 8.
standby generator system (SGS) is a commercial grade natural gas generator that is used nonsafety functions at the SHINE facility. It is available as a normal back-up power supply for cted asset protection loads as discussed in Chapter 8 but is not credited as an emergency er source.
uninterruptible electrical power supply system (UPSS) provides two divisions of safety-ted emergency power to the SHINE facility. The facility equipment that is served by the UPSS escribed in Section 8a2.2. This system is capable of delivering required emergency power for required duration during normal and abnormal operation.
OOP may occur during any combination of operating modes within the IF and the RPF. Some ential causes of a LOOP are:
- Degradation (reliability) of the transmission system;
- Electrical grid stress during peak load conditions;
- Severe weather effects from high wind, tornado, ice or snowstorms;
- Seismic event;
- Equipment failure in the supplying substation; or
- Failure or malfunction of facility transformers or switchgear.
s of all off-site power bounds partial loss of power scenarios within the facility. Partial loss narios include: (1) a complete loss of one division of power, and (2) a loss of one individual or motor control center (MCC). The effect of partial loss of power scenarios are limited to e systems or processes affected whereas a total loss of electrical power affects all systems processes.
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- Eight TSVs are conservatively assumed to be in irradiation operations mode, with the maximum source term and decay heat levels.
- Irradiation power is assumed to be 137.5 kW, 10 percent above maximum operating power.
- Bulk target solution temperature in the TSV is at the limit of 176°F (80°C).
- Initial light water pool temperature is assumed to be 95°F (35°C).
- Complete loss of PCLS flow at time of initiating event.
- Light water pool level of not less than [ ]PROP/ECI below finished floor (water depth of approximately [ ]PROP/ECI), which provides sufficient passive heat sink to remove decay and residual heat from the target solution.
- UPSS is available providing sufficient battery capacity for essential loads for their required runtime as provided in Table 8a2.2-1.
- Resupply of N2PS occurs within three days following a LOOP.
2.1.5.2 General Scenario Description noted in Subsection 13a2.1.5.1, the worst-case scenario is a complete LOOP. Although the rruption of off-site power may be relatively brief, it is assumed for this analysis that off-site er remains unavailable for an extended period of time. This could potentially occur if the OP is due to severe weather or a seismic event that damages substation equipment or ociated transmission lines.
sequence of events for a long-term LOOP is as follows:
- The UPSS automatically maintains power to the 125 VDC UPSS buses A and B, supplying power to the equipment identified in Section 8a2.2.
- Each NDAS HVPS de-energizes and the associated irradiation processes stop.
- The TSV dump valves open, draining the uranyl sulfate solution in the operating TSVs to their respective TSV dump tanks, as designed.
- The PCLS loses power to its pumps. Forced convection cooling ceases and heat is removed by natural convection to the light water pool.
- Hydrogen generation continues to occur due to radiolysis from the decay of fission products. The TOGS blowers and recombiner heaters operate on UPSS power for at least five minutes. The blowers continue forced flow through the TOGS recombiners for a short period of time as hydrogen production levels decrease and bubbles leave the target solution.
- The N2PS begins passively injecting nitrogen gas into the primary system boundary.
Nitrogen gas is injected in the eight SCAS systems via a connection to the dump tank.
The gas purges the primary system boundary leaving through a vent connection from the TOGS to the PVVS header. The gas then passes through the PVVS carbon delay beds for removal of fission product gases before release to the environment. The nitrogen purge system has enough capacity for three days, after which the system is resupplied.
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- In the event that any transfer of uranyl sulfate solution is in progress, VTS transfer operations stop.
- Nitrogen gas sweeps RPF process tank and lift tank headspaces to dilute radiolytic hydrogen. Nitrogen from the N2PS is routed to the PVVS carbon beds for removal of fission product gases before release to the environment. The N2PS has enough capacity for three days, after which the system is resupplied.
- The UPSS supplies essential facility loads their required runtime as provided in Table 8a2.2-1. The 120 VAC UPSS buses automatically maintain power to essential instrumentation and equipment, including radiation monitoring systems.
2.1.5.3 Accident Consequences accident consequences associated with a LOOP are discussed further in section 13a2.2.5.
2.1.6 EXTERNAL EVENTS section discusses external events that impact the IF. This class of accident initiators esent natural or man-made events that occur outside the facility and have the potential to act facility SSCs. Scenario descriptions are provided in this section for the range of accident ators that were considered during the accident analysis.
2.1.6.1 Identification of Scenarios, Initial Conditions, and Assumptions ISA process considered the following potential external events:
- Seismic event affecting the IF and RPF (see Section 3.4).
- Severe weather events affecting the IF and RPF (see Section 3.2).
- Transportation accidents, including small aircraft crash into the IF or RPF (see Subsection 3.4.5), toxic gas releases (see Subsection 2.2.3), or explosions (see Subsection 2.2.3).
- External flooding affecting the IF and RPF (Subsection 2.4.2).
- External fires from natural sources (see Subsection 2.2.3).
initial conditions and assumptions associated with these external events include:
- Prior to an external event occurring, the facility is assumed to be running at nominal conditions.
- Unless otherwise noted, these scenarios only consider single failure mechanisms.
- Eight NDAS contain maximum tritium inventory of [
]PROP/ECI of tritium gas.
- The facility structure is designed to withstand credible external events including seismic events, severe weather effects, tornado generated missiles, and impact from aircraft.
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- SSCs, including their foundations and supports, that are required to perform their safety function(s) in the event of a design basis earthquake (DBE) are classified as Seismic Category I.
- SSCs that are co-located with a Seismic Category I SSC and that are required to maintain their structural integrity in the event of a DBE to prevent unacceptable interactions are classified as Seismic Category II.
further details of seismic design criteria refer to Section 3.4.
2.1.6.2 General Scenario Descriptions following discusses the external event scenarios identified in the ISA process which impact IF or RPF:
smic Events Affecting the IF and RPF nario 1 - Seismic Event causing TOGS Failure eismic event may cause the failure of the TOGS. A failure of TOGS in one or more IUs could ult in hydrogen deflagrations. Potential consequences of TOGS failure include radiological e.
prevent a TOGS failure from an earthquake the TOGS is seismically qualified. The UPSS vides the TOGS with emergency power if normal facility power is lost. The TOGS functions for ort time after the IU cell shutdown until the N2PS can purge TOGS and lower the centration of hydrogen, reducing the possibility of hydrogen deflagration. Based on this ussion, the TOGS does not fail during a seismic event and no further analysis is required.
nario 2 - Seismic Event causing PCLS Failure eismic event may cause the failure of the PCLS. A failure of PCLS in one or more irradiation s could result in reduction in or excessive cooling, reactivity insertion, and potential criticality.
ential consequences of PCLS failure include radiological dose.
prevent these conditions, redundant high power range neutron flux signals initiate a TRPS ation that opens the redundant TSV dump valves. A TRPS actuation is also initiated on a PCLS cooling water temperature or low PCLS cooling water flow. Reduction in cooling is ussed in Subsection 13a2.1.3. Reactivity insertions due to excessive cooling are discussed ubsection 13a2.1.2. Based on this discussion, the loss of PCLS does not result in a ological release and no further analysis is required.
nario 3 - Seismic Event Causing Multiple NDAS Failures eismic event may cause the failure of one or more NDAS units. A failure such as a NDAS uum boundary failure in one or more irradiation units results in a release of tritium in one or e IU cells. Potential consequences of multiple NDAS failures include radiological dose. The NE Medical Technologies 13a2.1-24 Rev. 1
mitigate the impact of such failures of the primary confinement boundary, accelerator tritium rface subsystem mixed-gas return line pressure instrumentation, and ventilation isolation chanisms are used to confine released tritium. Accident consequences of this event are ussed in Subsection 13a2.2.6.
eismic event may also cause the failure of TPS components located in the TPS glovebox. The ological consequences of a failure of the TPS components due to a seismic event is bounded he TPS failure due to deflagration, as described in Subsection 13a2.2.12.2.
nario 4 - Seismic Event Causing a Single NDAS Failure ilure of the NDAS in a single IU cell is discussed in Subsection 13a2.1.12.
nario 5 - Seismic Event Causing NDAS Tritium Feed Fault eismic event may cause the failure of the NDAS tritium feed. A NDAS tritium feed failure ults in tritium prematurely entering the NDAS which causes higher power density and ential uranium precipitation. Potential consequences of NDAS tritium feed failure include ological dose.
prevent these conditions, redundant high time-averaged power range neutron flux signals ate a TRPS actuation that opens the redundant TSV dump valves. In addition, the TPS is vided with a passive design feature to limit the flow rate of tritium into the NDAS target mber. In the event of NDAS tritium feed failure, the primary confinement boundary is used to tain such incidents. Excess reactivity insertions due to high neutron production at cold ditions are discussed in Subsection 13a2.1.2.2, Scenario 4. Based on this discussion, the m feed failure does not result in a radiological release and no further analysis is required.
nario 6 - Seismic Event Causing Light Water Pool Liner Failure eismic event may cause the failure or leak in the light water pool liner. A failure or leak in the t water pool liner could result in a loss of cooling water inventory and result in target solution t up. Potential consequences of the light water pool liner failure include radiological dose.
prevent a loss of pool cooling water, the light water pool liner is seismically-qualified, and etrations through the liner are located above the minimum pool water height to limit out-age below this level. Piping penetrations into the light water pool with the potential for oning below the minimum acceptable water level contain anti-siphon devices or other means revent inadvertent loss of pool water. Because of the limited volume of water available to
, anti-siphon design features, and the design leak rate of the penetration, no further analysis quired.
nario 7 - Seismic Event Causing PVVS/VTS Failure eismic event may cause the failure of the PVVS/VTS. The limiting postulated failure occurs ng target solution transfer from the TSV dump tank to the molybdenum extraction and fication system (MEPS). The PVVS is assumed to fail, resulting in a loss of sweep gas in the NE Medical Technologies 13a2.1-25 Rev. 1
rogen gas concentration and prevents hydrogen deflagration.
ddition, the loss of PVVS also results in the loss of the VTS, stopping the movement of target tion. The target solution remains in the lift tanks or drains back into the TSV dump tank. The anks and the TSV dump tanks are passively-cooled and geometrically-favorable tanks.
refore, there are no consequences resulting from this event.
radiological consequences of deflagrations within the primary system boundary are ussed in Subsection 13a2.1.9.
nario 8 - Seismic Event Causing Crane Failure eismic event may cause the failure of the IF crane. A failure of a crane during a heavy lift of a lt plug or neutron driver in the IF could result in the heavy load dropping onto the NDAS or AS components. Potential consequences of the crane failing include radiological dose.
prevent crane failure, the crane is a single failure proof design and has been seismically lified. Additional information on heavy load drops is provided in Subsection 13a2.1.12.
nario 9 - Seismic Event Causing Chemical Spill eismic event may cause uranium oxide powder to become airborne during target solution paration activities or may overturn a uranium storage rack causing multiple canisters to spill, ulting in a worker uptake of uranium oxide.
de handling operations occur within the TSPS and URSS gloveboxes, which are seismically lified and have installed filtered ventilation. The quantity of uranium used in handling rations is limited and is insufficient to cause chemical dose consequences that exceed the mical exposure criteria in the event of a single canister spill.
cussion of the consequences of an overturned uranium storage rack and additional ussion of accidents with chemical dose consequences is provided in Section 13b.3.
ere Weather Events Affecting the IF and RPF nario 10 - Tornado or High Winds Affecting the IF and RPF main production facility is designed to withstand credible wind and tornado loads, including siles, as described in Section 3.2 and Subsection 3.4.2.6, respectively.
rnado or high wind event may cause an N2PS tube failure. Potential consequences of a S failure include damage to the components containing radioactive materials.
educe the possibility of tube failure from a tornado or high wind event, the N2PS nitrogen e bank is located in a reinforced concrete vault protecting the cylinders from tornado missile act. No further analysis for this event is required.
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vy snow or ice accumulation due to severe weather may cause damage to the main duction facility structure or systems including a loss of the normal building ventilation path or a of the safety-related PVVS effluent release path, which can then lead to a deflagration in the lity. Severe weather may also disrupt the nitrogen gas resupply following a N2PS activation.
facility structure is designed to withstand heavy snow and ice loading to prevent damage.
exhaust point for the safety-related PVVS effluent path is designed to be above the design w accumulation level, and the ventilation system air intakes are above the potential snow drift ht. The N2PS system is supplied with enough nitrogen for three days of operation which is quate to allow a resupply of the nitrogen tanks. No chemical or radiological consequences ult from severe weather accident scenarios.
nsportation Accidents main production facility is designed to withstand credible aircraft impacts and transportation idents, as discussed in Subsection 3.4.5.
ernal Flooding Affecting the IF and RPF main production facility was evaluated for external flood events. The results of the evaluation w that external flood events do not have an impact on the IF or RPF, as described in section 2.4.2.
ernal Fires from Natural Sources main production facility was evaluated for the potential for external fires from natural causes.
results of the evaluation show that external fires from natural sources do not have an impact he production facility as described in Subsection 2.2.3.
2.1.6.3 Accident Consequences failure of eight NDAS pressure boundaries as a result of a seismic event (Scenario 3) results otential radiological exposure to workers and the public. The primary confinement boundaries credited to mitigate the consequences of this failure. The accident consequences associated this external event are discussed further in Subsection 13a2.2.6.
2.1.7 MISHANDLING OR MALFUNCTION OF EQUIPMENT waste gases from irradiation of the target solution are of two major types: the hydrogen and gen produced by radiolysis of water in the target solution, and radioactive fission product es. Mishandling or malfunction of equipment within the IU or TOGS cells has the potential to se leakage of these gases. Specifically, a failure of the TOGS portion of the PSB could allow ape of fission product gases or hydrogen into the primary confinement boundary and the ologically controlled area (RCA). Analysis of this event and other potential mishandling or function of equipment events, excluding detonation or deflagration of hydrogen within TOGS other exothermic chemical reactions within the PSB, are included in this section.
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- The detonation or deflagration of hydrogen within the TOGS or otherwise affecting the PSB is analyzed in Subsection 13a2.1.9.
- Other unintended exothermic chemical reactions within the PSB are analyzed in Subsection 13a2.1.10.
- The loss of vessels and line failures for systems within the RPF are analyzed in Subsection 13b.2.4.
neutron driver and TPS system failures within the IU cell could similarly result in releases of m and hydrogen.
- Events related to the neutron driver are analyzed in Subsection 13a2.1.6 and Subsection 13a2.1.12.
- Events related to the TPS are analyzed in Subsection 13a2.1.12.
2.1.7.1 Identification of Causes, Initial Conditions, and Assumptions ilure of the PSB resulting in a release of fission product gases may be caused by excessive osion of PSB components.
initial conditions and assumptions associated with mishandling or malfunction of equipment ude:
- Fission product gases (e.g., Kr, Xe, and halogens) produced during irradiation operations are retained within TOGS until the target solution batch irradiation cycle is completed.
- As the TOGS circulates sweep gas during the irradiation cycle, a portion of the iodine is removed by the zeolite beds, and hydrogen and oxygen are recombined by the catalytic recombiners, but no other gases are removed or purged.
- Each IU is operated on an irradiation cycle of 30 days with a minimum [ ]PROP/ECI residence for the target solution in the TSV dump tank following irradiation.
- The material-at-risk for these events is conservatively taken as the inventory at shutdown, at the end of the irradiation cycle, after [ ]PROP/ECI, with the safety-based assumptions listed in Table 11.1-1.
- The IUs are operated independently, so that an event on one TOGS does not affect another TOGS or IU cell.
- The TOGS cells are isolated from the rest of the facility by robust walls, ceiling, and floor.
The physical separation of individual TOGS prevents malfunctions in one TOGS from affecting the others.
- Penetrations for piping, ducts, and electrical cables are sealed to limit the release of radioactive materials from the confinement boundary.
- The TOGS cells are cooled by a recirculating air ventilation system and are isolated from all other facility ventilation systems. A single ventilation connection from the PCLS expansion tank to the RVZ1e subsystem is provided for hydrogen gas removal from the cooling systems, and is isolated on a high radiation signal in the ventilation duct.
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re are three scenarios identified as mishandling or malfunctions of equipment. Each of these narios and their potential causes is discussed below:
nario 1 - Failure of the TOGS Pressure Boundary Resulting in Release of Off-Gas into the GS Cell ilure of TOGS pressure boundary downstream of a TOGS blower may be caused by osion of a TOGS component. This failure results in a release into the TOGS cell of fission duct gases and hydrogen normally managed by the TOGS, resulting in increased worker and lic dose. Consequences of a release of fission product gases into the TOGS cell are ussed in Subsection 13a2.2.7. The protections in place to mitigate the consequences of a ase of fission product gases are maintenance, inspection, and testing of the PSB, and the ary confinement boundary, which is described in Section 6a2.2.
nario 2 - Failure of the TOGS Vacuum Tank TOGS vacuum tank is in the light water pool below the water line. A failure of the vacuum that results in flooding the TOGS vacuum tank results in a loss of vacuum and flooding in TOGS system on subsequent opening of the vacuum makeup valve. This scenario is sidered in Subsection 13a2.1.4, which discusses failures of the PSB below the level of the t water pool.
nario 3 - TOGS Vacuum Makeup Valve Fails Open TOGS vacuum makeup valve inadvertently opens because of a failure of the valve controller ue to operator error. A subsequent opening of the VTS to vacuum tank isolation valve results n elevated release of fission product gases and hydrogen to the VTS header and PVVS via vacuum tank. The elevated release of fission product gases and hydrogen does not exceed PVVS capability to remove the gases. The release is mitigated by the PVVS carbon delay s and the PVVS guard beds. The mitigation controls reduce the release to the level of normal f valve operation. Therefore, no further analysis is required.
2.1.7.3 Accident Consequences release of off-gas from the PSB to the IU or TOGS cell results in potential radiological osure to workers or the public. The primary confinement boundary is credited to mitigate the sequences of a release of target solution in the IU or TOGS cell. The accident consequences ociated with the mishandling or malfunction of equipment are evaluated as discussed above.
2.1.8 LARGE UNDAMPED POWER OSCILLATIONS ecommended by the ISG Augmenting NUREG-1537 (USNRC, 2012a), the TSV is evaluated arge undamped power oscillations as a potential event that could occur during irradiation ration due to reactivity variations in the target solution that lead to fluctuations in the neutron tiplication (keff) within the irradiated target solution. Large undamped power oscillations are er oscillations that grow over time due to positive reactivity feedback effects and challenge design limits of the subcritical assembly.
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illation amplitudes are limited due to the physical limitations of accelerator design. Coupled t response as a result of transients has been analyzed. Power oscillations that occur are self-ting as a result of the inherent design and safety characteristics associated with the TSV, rating parameters, and plant response to transients.
2.1.8.1 Identification of Causes, Initial Conditions, and Assumptions er oscillations may occur in the TSV as a result of normal anticipated reactivity variations in the target solution or neutron driver source output variations. The causes or scenarios for er oscillations include:
- Variations in the neutron driver voltage, current, tritium concentration, or other parameter results in variations in the fusion neutron production rate.
- Failure in the PCLS temperature control loop or RPCS supply results in temperature oscillations.
initial conditions and assumptions for this scenario are as follows:
- Negative temperature and void coefficients are within license limits discussed in Subsection 4a2.6.3.
- The TRPS neutron flux setpoints are within technical specification limits.
turbations from the TOGS can result in pressure changes in the TSV. During startup, diation, and shutdown operations, TOGS regulates gas pressures in the PSB to maintain ssures within the acceptable range. Increased gas pressures in the TSV reduce void fraction, ing to positive reactivity addition. Transient analysis of a complete void collapse is presented ubsection 4a2.6.3.3. Excessive TSV power oscillations from TOGS pressure oscillations are vented by redundant TRPS high neutron flux IU Cell Safety Actuation signals.
neutron driver has variability in neutron production rates due to normal variations in beam ent and focusing, voltages, and tritium gas concentrations. These variations lead to esponding variations in fission power in the SCAS. An evaluation of the neutron driver ced transient is presented in Subsection 4a2.6.1.4. The TRPS high wide range neutron flux ell Safety Actuation signal prevents excessive TSV power oscillations that challenge design ts should the neutron driver return to full power rapidly following a reduced power transient.
LS provides cooling water to the TSV, and therefore, temperature variations in the PCLS ctly lead to TSV temperature variations. PCLS provides constant cooling water inlet perature to the TSV within ranges described in Section 5a2.2. The target solution perature ranges are provided in Section 4a2.2. Subsection 4a2.6.1 evaluates PCLS perature variations and effects on the TSV. The TRPS high neutron flux IU Cell Safety uation signals prevent excessive TSV power oscillations from PCLS temperature variations.
re are no large, undamped power oscillations that result from PCLS operation.
er density limits for thermal-hydraulic and chemical stability of the target solution are cribed in Subsection 4a2.6.3. In addition, Subsection 4a2.6.3 discusses the limiting core figuration, which is the core configuration with the highest power density.
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2.1.8.2 General Scenario Description noted in Subsection 13a2.1.8.1, power oscillations may occur in the TSV during normal ration as a result of target solution reactivity or neutron driver source output variations.
ause of the TSV and interfacing system design and operating parameters, the reactivity ations are small at operating power, resulting in a very stable TSV with self-limiting power illations.
ge power oscillations that could potentially challenge design limits are prevented by TRPS oints on high neutron flux. No operator actions are required to damp power oscillations.
en a TRPS high neutron flux setpoint is exceeded, the neutron driver is automatically energized, the TSV dump tank valves automatically open, and the target solution is dumped force of gravity) into the favorable geometry TSV dump tank. No analyzed power oscillation nario results in damage to the PSB.
2.1.8.3 Accident Consequences itional discussion associated with large undamped power oscillations is provided in section 13a2.2.8.
2.1.9 DETONATION AND DEFLAGRATION IN THE PRIMARY SYSTEM BOUNDARY subsection discusses the effects of a hydrogen deflagration or detonation in the PSB.
diation of the target solution produces significant quantities of hydrogen and oxygen and ll quantities of fission products. The TOGS is the primary control for mitigating hazards ociated with the evolved gases. Functional requirements for the TOGS include maintaining concentration of hydrogen to less than the lower flammability limit (LFL), recombining the rogen and oxygen, and returning the recombined water back to the TSV. The TOGS tions largely as a closed loop during the irradiation process, with gas additions and removals eeded to maintain proper functioning. TOGS is purged as needed to the PVVS via the VTS.
pter 6 includes a discussion of the facility combustible gas management systems.
2.1.9.1 Identification of Causes, Initial Conditions, and Assumptions formation and release of hydrogen due to radiolytic decomposition is an inherent result of diation of water. The ISA process identified several potential scenarios that could result in the umulation of hydrogen and potential deflagration or detonation. A deflagration or detonation ident could occur if the TOGS fails, which could allow hydrogen to accumulate in the TSV dspace, dump tank, or TOGS piping. Potential failures that have been identified include a of power, failure of the TOGS blowers, blockage or restriction in the TOGS flow path, and er leakage into the PSB that results in reduced sweep gas flow. Hydrogen could also umulate if there is degraded performance of the TOGS, such as reduced volumetric flow rate to a partially-obstructed demister or reduced recombiner effectiveness.
n loss of TOGS function, hydrogen concentrations in the TSV headspace and TOGS rise.
r the neutron driver is shut down on loss of TOGS flow, the voids in the target solution apse and release hydrogen from the solution. This effect combined with continued radiolysis NE Medical Technologies 13a2.1-31 Rev. 1
n the concentrations needed to produce a hydrogen detonation.
initial conditions and assumptions associated with a deflagration of hydrogen gas are:
- A hydrogen deflagration produces a maximum overpressure condition of 65 pounds per square inch absolute (psia). Discussion of the design basis for the TOGS is found in Section 4a2.8.
- Each TSV is serviced by a dedicated and independent TOGS. In this section it is assumed a single TOGS fails, allowing hydrogen to accumulate in the TSV and TSV dump tank. Multiple TOGS failures resulting from a loss of normal power event are addressed in Subsection 13a2.1.5.
- The target solution is at steady-state conditions at 110 percent of the licensed power limit when the TOGS failure occurs. This is conservative since it implies the maximum hydrogen generation rate in the target solution.
2.1.9.2 General Scenario Descriptions eflagration could occur if the TOGS were to fail during the irradiation process. Irradiation of target solution generates significant quantities of hydrogen and oxygen. The LFL for rogen in the headspace is rapidly reached if the TOGS fails and the neutron driver continues perate.
nario 1 - TOGS Failure Resulting in Hydrogen Deflagration OGS failure may occur due to flow blockage or failure of the TOGS blowers, loss of sweep flow to the TSV dump tank or overfilling the TSV that causes a reduction of available dspace and sweep gas flow. The loss of TOGS functionality allows the hydrogen gas centration to increase in the headspace in the TSV and/or TSV dump tank. The hydrogen gas y ignite and cause a deflagration in the PSB. Failures of the PSB in the TOGS cell not related eflagration or detonation are considered in Subsection 13a2.1.7.
rfills of the TSV due to operator error are prevented by the redundant overflow lines to the dump tank and redundant TSV dump tank level sensors which initiate a dump of the TSV. In ition, a permissive interlock on the TSV fill valves prevents them from opening unless TOGS is above the low flow limit to ensure that TOGS is functioning prior to fill operations.
eakage of water from the light water pool, PCLS, or RPCS into the PSB may cause loss of dspace. In these events, the N2PS purges sweep gas through the PSB to prevent damage to PSB from excessive accumulation of hydrogen.
ogen purging of the PSB is actuated by TRPS on signals that indicate loss of TOGS ability to perly maintain PSB hydrogen concentrations: low-high and high-high dump tank level, low GS oxygen concentration, low TOGS mainstream loop flow, low TOGS dump tank flow, high peratures in the TOGS condenser demister, and ESFAS loss of electrical power. Nitrogen ging ensures that hydrogen concentrations remain below the level that could result in damage he PSB. The operation of the N2PS is described further in Chapter 6, and the design of the tem is described in Section 9b.6.
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nario 2 - PCLS Radiolysis Resulting in Hydrogen Deflagration er normal conditions, hydrogen gas generated in the PCLS is ventilated to the facility tilation system (RVZ1e). A failure of the ventilation system may result in increased hydrogen concentration in the PCLS expansion tank. The hydrogen may ignite and cause a agration or detonation in the PCLS expansion tank, resulting in a release of radioactive erial if the PSB is damaged.
ame arrestor on the PCLS expansion tank that vents to the primary confinement atmosphere vents potential ignition sources from causing a deflagration in the PCLS expansion tank. In event that a release of radioactive material did occur, then the release is mitigated by the ary confinement boundary. Radiation detection instruments on the RVZ1e duct generates an ell Safety Actuation and close redundant isolation valves to RVZ1e. The potential exposures this event are bounded by the release of target solution to the IU cell, which is discussed in section 13a2.2.4.
2.1.9.3 Accident Consequences ause detonations and deflagrations in the PSB do not result in the failure of the PSB, there no radiological consequences associated with these accident scenarios. Further discussion rovided in Subsection 13a2.2.9.
lysis of PSB failures below the light water pool is provided in Subsection 13a2.2.4.
2.1.10 UNINTENDED EXOTHERMIC CHEMICAL REACTIONS OTHER THAN DETONATION ntended exothermic chemical reactions other than detonation have been evaluated as ential initiating events as part of the accident analysis within the IF. This subsection examines ty aspects of exothermic chemical reactions that challenge the PSB integrity in the IF, other n hydrogen deflagrations or detonations. Hydrogen deflagrations and detonations are ressed in Subsection 13a2.1.9.
2.1.10.1 Identification of Causes, Initial Conditions, and Assumptions ISA process identified two potential scenarios that are evaluated in this subsection. The first e uranium metal-water reaction in the neutron multiplier, and the second is an ignition of the vated carbon bed in the TPS system.
nario 1 - Uranium Metal-Water Reaction in the Neutron Multiplier Assembly the uranium metal-water reaction, the IU is operating at normal irradiation conditions. The tron multiplier, as manufactured, is [
]PROP/ECI. The PCLS provides cooling to the TSV and the neutron multiplier and transfers es produced from radiolysis to the expansion tank.
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uranium metal-water reaction in the neutron multiplier may be caused by an event which aches the neutron multiplier cladding allowing water to come into direct contact with the nium metal. Possible causes include corrosion of the cladding, uranium metal-cladding raction due to radiation-induced growth, or other mechanical damage incurred during ntenance. The breach may occur at any time during the lifecycle of the neutron multiplier wing water intrusion over an extended period of time.
nario 2 - Ignition of the Activated Carbon Bed in the TPS the TPS carbon bed fire, the TPS is operating at its maximum normal capacity. There are no itional assumptions for Scenario 2.
tion of the activated carbon bed in the TPS may be caused by air intrusion into the TPS wing improper restoration of the system following maintenance.
2.1.10.2 General Scenario Descriptions nario 1 - Uranium Metal-Water Reaction in the Neutron Multiplier Assembly mall breach of the neutron multiplier cladding allows PCLS water into the cladding [
]PROP/ECI. The water intrusion results in an exothermic uranium metal-water reaction in neutron multiplier assembly. The reaction generates hydrogen gas inside the neutron tiplier cladding shell [ ]PROP/ECI. An accumulation of rogen gas could result in a deflagration under certain conditions. These conditions include icient oxygen concentration, an ignition source, or autoignition temperatures being reached.
his scenario, the hydrogen produced mixes with [
]PROP/ECI bits a potential deflagration. Therefore, a hydrogen deflagration in the neutron multiplier from event is considered unlikely.
rogen gas that migrates into the PCLS stream from the neutron multiplier leak accumulates e expansion tank, which is vented to the RVZ1e. Therefore, a hydrogen deflagration in the LS from uranium metal-water reactions is also unlikely.
nario 2 - Ignition of the Activated Carbon Bed in the TPS tion of the activated carbon bed in the TPS may occur due to improper system restoration wing maintenance. If air intrusion occurs in the TPS, the activated carbon bed in the impurity oval subsystems may ignite and cause a fire to spread through the TPS system and ebox. A failure of the TPS pressure boundary may result in a release of stored tritium into the ebox. The release of tritium is further evaluated in Subsection 13a2.1.12. The protections in e to prevent ignition of the activated carbon bed are the TPS glovebox inert atmosphere and ntenance to maintain the inert atmosphere. Therefore, the ignition of the activated carbon and spread of fire in the TPS glovebox is considered unlikely.
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accident consequences associated with unintended exothermic chemical reactions are ussed further in Subsection 13a2.2.10. The accident consequences associated with a tritium ase from the TPS glovebox are discussed further in Subsection 13a2.2.12.
2.1.11 SYSTEM INTERACTION EVENTS subsection discusses the effects of system interactions on the systems which contain onuclide material. System interactions have the potential to cause damage that may lead to release of these materials.
ee categories of system interactions between systems located within the IF and the RPF are sidered in this analysis. These are: (1) functional interactions, (2) spatial interactions, and human-intervention interactions.
ctional Interactions ctional interactions are interactions between systems or subsystems that result from a mon interface. A functional interaction exists if the operation of one system can affect the ormance of another system or subsystem. An adverse functional interaction exists when the ration and/or performance of an (initiating) system adversely affects the operation and/or ormance of an SSC as it performs its safety-related function. Functional interaction events are discussed in this section are those that may result from failures in support systems or er shared systems that could result in an adverse impact on the primary system boundary.
VS is connected to the eight IUs via connections to TOGS. Accidents with PVVS failure are sidered in Section 13b.2.
functional interactions considered in this analysis are the following:
s of Off-Site Power NPSS provides electrical power to SSCs in the IF and the RPF.
uction of cooling
- The RPCS is the common heat sink for the independent instances of PCLS, which are the primary cooling systems for each TSV. Each PCLS removes generated heat from its associated TSV during normal and shutdown operations. The generated heat is transferred to the RPCS via the PCLS heat exchangers. The RPCS is served by the PCHS, which exhausts heat to the environment.
- RPCS additionally provides cooling for several heat exchangers in the IF and the RPF, including:
- TOGS condenser-demisters
- TOGS recombiner condensers
- TSPS dissolution tank reflux condensers
- [ ]PROP/ECI
- PVVS condensers
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s of ventilation
- The ventilation systems (RVZ1, RVZ2) are described in Section 9a2.1.
- Loss of RVZ1 between the TPS glovebox stripper system (GBSS) hood in the IF and potentially contaminated areas and systems in the RPF, such as the:
- RLWI shielded enclosure,
- Individual cells of the supercell,
- URSS glovebox,
- TSPS gloveboxes, or
- Vent exhausts from the PCLS expansion tanks.
- Loss of RVZ2 to common areas of the IF and the RPF.
- Loss of ventilation to the PCLS cooling rooms tial Interactions tial interactions are interactions resulting from the presence of two or more systems in tions. Spatial interactions include a single event that could impact the operation of the cent systems, or the failure of one system that may impact the operation of another system.
spatial interactions considered include the effects of internal fires, internal flooding, chemical ases, and other dynamic failure effects.
man-Intervention Interactions man-intervention interactions are adverse system interactions caused by human errors in the F which can cause adverse system performance in the subcritical assembly during irradiation rations. Human errors are identified as potential causes for other accident sequences and not explicitly identified in this section. For example, human interactions or errors considered otential causes for accident sequences include:
- Failure to operate equipment when required
- Inappropriate operation of equipment
- Maintenance error affecting operating equipment
- Testing error affecting operating equipment man errors downstream in the RPF processes that are related to mixing or transfer of target tion are considered in Subsection 13b.2.5.
2.1.11.1 Identification of Causes, Initial Conditions, and Assumptions identification of causes of system interaction events are provided in the subsections in pter 13 as referenced below. There are no unique initial conditions or assumptions ociated with system interaction events.
2.1.11.2 General Scenario Descriptions following section discusses the system interactions that can occur at the SHINE facility.
tem interactions that are already analyzed in other parts of Chapter 13 are referenced to NE Medical Technologies 13a2.1-36 Rev. 1
ctional Interactions s of Off-Site Power OP events are described in Subsection 13a2.1.5.
uction of Cooling nts that could cause a reduction of cooling include PCHS or RPCS failure, LOOP, or external nts.
- Reduction in cooling due to PCHS or RPCS failure is described in Subsection 13a2.1.3.
- Reduction in cooling following a LOOP is described in Subsection 13a2.1.5.
- Reduction in cooling due to external events is described in Subsection 13a2.1.6.
s of Ventilation ss of ventilation could be caused by equipment failure, a LOOP, or external events.
nario 1 - Loss of Normal Ventilation to the IU or TOGS Cells ilure of RVZ1 may be caused by failure of a blower or cooler, including loss of cooling water.
ay also be caused by a failed-shut or mispositioned damper or other equipment failure. A loss ooling may cause instrumentation inaccuracies or failures which may lead to TOGS operation or loss of function. This can result in a potential deflagration and release of ological material.
protections in place to prevent a TOGS failure due to loss of ventilation are redundant and ironmentally qualified TOGS instrumentation (e.g., low flow) that initiates a TRPS signal if GS failures are detected. The TRPS signal opens redundant TSV dump valves draining target tion to the TSV dump tank and shuts down the irradiation unit. Decay heat from the target tion is removed by the light water pool.
nario 2 - Loss of Normal Ventilation to PCLS Cooling Rooms ilure of RVZ2 may be caused by failure of a blower or cooler, including loss of cooling water.
s of ventilation to individual PCLS cooling rooms may also be caused by a failed-shut or positioned damper. A failure of normal ventilation may lead to increased environmental peratures within the PCLS cooling room with potential for increased instrument inaccuracies ailure. The consequences of an RVZ2 failure leading to equipment malfunction result in TSV rcooling causing a reactivity insertion in the TSV. Excess reactivity additions are discussed her in Subsection 13a2.1.2.
protections in place to prevent TSV malfunctions related to ventilation failures are redundant and high PCLS temperature trip that initiates a TRPS signal (separate from the control tem). The TRPS signal opens redundant TSV dump valves draining target solution to the TSV NE Medical Technologies 13a2.1-37 Rev. 1
ed on the preventive controls the failure of normal ventilation does not have radiological sequences, and no further analysis is required.
s of ventilation due to a LOOP is described in Subsection 13a2.1.5.
s of ventilation due to external events is described in Subsection 13a2.1.6.
tial Interactions s
fire hazards analysis (FHA) evaluates the fire hazards and fire protection features for each area in the SHINE facility. The fire protection features in the IF rely on low combustible ing, fire detection, manual fire-fighting capabilities, and rated fire barriers to limit the potential ire initiation and spread within the IF. The fire protection program and the FHA are described ection 9a2.3.
ential fire scenarios in the IF have been evaluated in the ISA process. The principle fire ards in the IF are: (1) the HVPS used for the NDAS service cell, (2) hydrogen located in the and within the PSB for each IU cell, and (3) the carbon filters in the radiologically controlled a (RCA) exhaust filter room in the mezzanine area. Causes of fires include a catastrophic re of the HVPS and maintenance activities including hot work.
consequences of the fire scenarios are the potential release of radioactive materials, uding tritium. The release of tritium is evaluated in Subsection 13a2.1.12.
ioactive materials accumulated in the exhaust filter trains can also be released in the event of
- e. However the exhaust filter trains are monitored and alarmed for buildup and replaced.
refore, a significant release of radioactive material is not expected to occur.
itional effects of fire damage on other facility systems include potential loss of TOGS, PCLS, ventilation system functions. Loss of the TOGS is described in Subsection 13a2.1.4 and section 13a2.1.9. Loss of PCLS is described in Subsection 13a2.1.3 and section 13a2.1.5. Loss of ventilation systems is described in Subsection 13a2.1.11.2.
protections in place to prevent or mitigate the effects of a fire in the IF include the protection ures described above (i.e., detection, rated barriers, manual suppression). Strict inistrative controls on combustible materials and maintenance activities, including hot work also in place. For a fire involving the HVPS, a catchment pan to contain oil leakage or spray ts the potential spread of oil reducing the potential for fire spread from the HVPS. Therefore, a ase of radioactive material is not expected to occur.
piping failures resulting in deflagration are discussed in Subsection 13a2.1.12.3.
rogen deflagration within the PSB is discussed in Subsection 13a2.1.9.
s caused by external events are discussed in Subsection 13a2.1.6.
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thermic chemical reaction scenarios are discussed in Subsection 13a2.1.10.
rnal Flooding ential internal flooding scenarios in the IF have been evaluated in the ISA process.
re is no potential for widespread internal flooding within the IF. The primary sources of rnal flooding are cooling water systems (e.g., PCLS) located in the IF with limited volume and ssure. The primary consequence of a leak in these systems is a loss of cooling to ponents served by the system. Localized water leaks or spray are contained to the room in ch the system resides and would not result in widespread flooding.
flooding scenario unique to the IF is a leak in a light water pool that serves the IU. A leak in light water pool liner may result in leakage of water into the pipe trench and subgrade vaults oducing moderator around pipes and tanks containing uranyl sulfate solution. The nuclear cality analyses for the trench and vaults assumes bounding moderation conditions which udes full reflection. Therefore there is no consequence as a result of this scenario.
omplete drainage of a light water pool due to a large break would also result in a loss of dual heat removal capability from the SCAS. The light water pool liner is designed to remain ct during normal operation as well as during design basis earthquake and design basis ident events. Penetrations through the light water pool liner are above the minimum water
- l. The light water pool is also equipped with a leak chase system to detect leaks.
oding caused by external events is discussed in Subsection 13a2.1.6.
amic Effects cess systems in the SHINE facility operate at low temperatures (i.e., generally less than
°F [93°C], except for the TOGS hydrogen recombination components) and low pressures
, less than 100 psig [689 kPa gauge]), which are not subject to dynamic effects as are found igh energy systems. As needed, safety-related systems are protected from the dynamic cts related to equipment failure and external events. No consequences result from dynamic cts interactions in the SHINE facility.
man Intervention Interactions man interventions can cause adverse system interactions because of the single common trol room at the SHINE facility. Operators are able to control multiple systems within the IF the RPF from the control room. Operator errors may occur including performing control rations on the wrong system, failing to perform required actions, or performing actions out of uence.
ntenance is performed as a normal scheduled activity and as a response to emergent ipment problems. Maintenance may occur during all modes of operation, including while diation or processing activities are in progress. Errors that occur during maintenance activities cause failures in operating systems such as support systems. Maintenance errors may be NE Medical Technologies 13a2.1-39 Rev. 1
man intervention interactions as accident scenario initiating events are described in other tions in this chapter as applicable and are not evaluated further in this section.
2.1.11.3 Accident Consequences system interactions described in the preceding sections do not result in radiological sequences. Accident consequences resulting from system interactions that are referenced to er subsections in Chapter 13 are evaluated in those subsections.
her discussion regarding system interaction events described in this section is provided in section 13a2.2.11.
2.1.12 FACILITY-SPECIFIC EVENTS ISA process identified several accident scenarios that are unique to the SHINE facility and e the potential for inadvertent radiation exposure to workers or members of the public.
ility-specific accident scenarios are associated with the NDAS, the TPS, and potential age resulting from heavy load drops.
2.1.12.1 Identification of Causes, Initial Conditions, and Assumptions eral scenario descriptions for events involving the NDAS, TPS, and heavy load drop include ses of each scenario.
accident scenarios involving the NDAS, the following initial conditions and assumptions ly:
- The NDAS contains the bounding inventory of tritium gas for full power.
- The NDAS pressure vessel contains the maximum inventory of sulfur hexafluoride (SF6) gas.
- The primary confinement boundary for an affected IU cell is operable, including the RVZ1e radiation detection and isolation valves.
accident scenarios involving the TPS, the following initial conditions and assumptions apply:
- The TPS glovebox confinement is operable, including the confinement isolation valves.
- The glovebox atmosphere is inerted with nitrogen.
- Automatic isolation valves are installed in the system to isolate sections of the system to minimize system release.
- Leakage of tritium from the glovebox enclosure or the external piping is detected by the continuous airborne monitoring system (CAMS) or other leakage detection systems to provide alarms for facility personnel evacuation.
- The TPS header contains the maximum inventory of tritium gas NE Medical Technologies 13a2.1-40 Rev. 1
- An IU cell is in maintenance with the IU cell shielding plug removed and the TSV and NDAS empty, or
- An IU cell is in service with IU cell shielding plug in place.
2.1.12.2 General Scenario Descriptions tron Driver Assembly System Event Descriptions re are four scenarios that are specific to the operation of the NDAS in the SHINE facility.
se scenarios are: (1) inadvertent exposure to neutrons within the IU, (2) inadvertent exposure eutrons in the NDAS service cell (NSC), (3) catastrophic failure of the NDAS, and (4) an AS vacuum boundary failure.
AS Scenario 1 - Inadvertent Exposure to Neutrons within the IU dvertent exposure to neutrons may be caused by operation of a neutron driver while sonnel are in the IU cell, such as during maintenance or assembly/disassembly activities, vertent access to an IU cell during irradiation operations, or failure to properly install IU cell lding following access. An operator error which results in the neutron driver becoming rgized with nearby personnel or without adequate shielding results in a significant neutron e to workers. Operator error is the most likely cause of inadvertent exposure to neutrons in the IU.
tections in place to prevent the inadvertent operation of a neutron driver in the IU cell are the out/tagout of the HVPS, opening of electrical breakers for the HVPS, and a two-key interlock he NDAS control system. Also, the NDAS operating procedures require evacuation of terium and tritium from the NDAS and isolation of the deuterium and tritium supplies to the IU e in maintenance. Proper installation of IU cell shielding following IU cell access is verified ore operation of the neutron driver. The accelerator cannot produce significant neutron-ducing reactions without deuterium or tritium. The inadvertent exposure to neutrons within the s not credible due to the administrative controls and protections in place.
AS Scenario 2 - Inadvertent Exposure to Neutrons within the NSC dvertent exposure to neutrons may be caused by operation of a neutron driver while sonnel are in the NSC, such as during maintenance or assembly/disassembly activities. An rator error which results in the neutron driver becoming energized with nearby personnel or out adequate shielding results in a potential for significant neutron dose to workers. Operator r is the most likely cause of inadvertent exposure to neutrons within the NSC.
tections in place to prevent the inadvertent operation of a neutron driver in the NSC are NSC rating procedures, which include independent confirmation of room clearance prior to testing, job briefs, and postings; and a two-key interlock on the NDAS control system. Operating cedures also require control of deuterium gas during testing, to ensure deuterium gas plies are only available when use is planned. The inadvertent exposure to neutrons within the C is not credible due to the administrative controls and protections in place.
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astrophic failure of the NDAS may be caused by a failure of a ceramic component inside the tron driver. A leak or failure of the ceramic results in a loss of NDAS vacuum inside the SF6 ssure vessel and subsequent overpressure of the NDAS vacuum boundary causing failure.
ure of the vacuum boundary results in a leak of tritium and SF6 gas to the IU cell, which ses IU cell pressurization. Pressurization of the IU cell can cause increased leakage rates ween the IU cell and the irradiation facility (IF), which results in higher dose to workers and public due to the release of tritium.
accident scenario is mitigated by the primary confinement boundary pressure sensors in the AS mixed gas return line, and isolation valves on RVZ1e from the PCLS expansion tank. The ary confinement boundary is described in detail in Chapter 6. Multiple catastrophic failures of AS units are described in Subsection 13a2.1.6.
AS Scenario 4 - NDAS Vacuum Boundary Failure ease of tritium from the NDAS vacuum boundary may be caused by a weld or vacuum seal re or improper maintenance. A failure of the NDAS vacuum boundary results in a leak of m into the IU cell, causing higher dose to workers and to members of the public. The ident scenario is mitigated by the primary confinement boundary, which is described in detail hapter 6.
um Purification System Event Descriptions re are five scenarios that are specific to the operation of the TPS in the SHINE facility. These narios are: (1) TPS piping failure due to deflagration, (2) release of tritium into the IF due to ebox deflagration, (3) release of tritium to the facility stack, (4) excessive release of tritium the tritium storage bed, and (5) release of tritium into the IF due to TPS header mechanical age.
Scenario 1 - TPS Piping Failure due to Deflagration roper system restoration following maintenance allowing air intrusion into TPS piping or by air akage from the NDAS may result in a deflagration within the TPS piping that causes a piping re and a release of tritium gas into the TPS glovebox. The release of tritium gas into the TPS ebox results in higher dose to workers and to members of the public.
release of tritium is confined within the tritium confinement boundary, including the TPS ebox radiation monitors and GBSS ventilation isolation valves. The tritium confinement ndary is described in detail in Section 6a2.2. Isolation of the TPS room ventilation is also dited for mitigation.
Scenario 2 - Release of Tritium into the IF due to Glovebox Deflagration kage of TPS piping may lead to TPS glovebox failure caused by deflagration that causes the m confinement boundary to fail. TPS piping leakage may be the result of improper oration to operating conditions from maintenance or of liquid nitrogen ingress into the eous nitrogen lines which causes embrittlement and failure of the TPS piping. Failure of the NE Medical Technologies 13a2.1-42 Rev. 1
TPS glovebox is designed such that the minimum size prevents the possibility of reaching lower flammability limit for the quantity of available hydrogen. The TPS glovebox is also ted with nitrogen which prevents the presence of oxygen. Based on the glovebox design and t atmosphere, a deflagration in the glovebox is not considered credible and is not analyzed her.
Scenario 3 - Release of Tritium to the Facility Stack lease of tritium directly to the facility stack may be caused by improper restoration to rating conditions from maintenance which results in a leak of tritium into the glovebox and a current misalignment of the GBSS valves following maintenance. A release of tritium to the lity stack results in higher worker and public dose.
protection in place to mitigate a release of tritium to the facility stack is the tritium monitor on GBSS exhaust to RVZ1e, which causes an isolation of the glovebox as part of the tritium finement boundary.
Scenario 4 - Excessive Release of Tritium from the Tritium Storage Bed essive release of tritium from the tritium storage bed may be caused by failure of the storage heater control resulting in excessive heat input. Failure of the heater results in an excessive ntity of tritium added to the TPS system, resulting in overpressurization and release of tritium he TPS glovebox. The tritium release is confined within the tritium confinement boundary, ch is described in detail in Section 6a2.2.
protection in place to mitigate a release of tritium to the facility stack is the tritium monitor on GBSS exhaust, which causes an isolation of the glovebox as part of the tritium confinement ndary.
Scenario 5 - Release of Tritium into the IF due to TPS Header Mechanical Damage lease of tritium directly to the IF may be caused by mechanical damage to the TPS header ch results in a leak of tritium to the IF. A release of tritium to the IF results in higher dose to kers and to members of the public. The TPS header consists of jacketed piping which uces the likelihood of mechanical damage that results in a tritium leak.
vy Load Drop Scenario Descriptions h respect to the SHINE facility, a heavy load is defined as a load that, if dropped, may cause ological consequences that challenge the accident dose criteria described in Section 13a2.2.
re are three scenarios that are specific to heavy load drops in the SHINE facility. These narios are (1) a heavy load drop into an open IU cell, (2) a heavy load drop onto an in-service ell, and (3) a heavy load drop onto TPS equipment.
NE Medical Technologies 13a2.1-43 Rev. 1
ane mechanical failure or operator error during a lift may result in a heavy load drop into an n IU cell. The heavy load can damage the SCAS components and result in a release of oactive material.
NE has applied the applicable guidance from NUREG-0612, Control of Heavy Loads at lear Power Plants (USNRC, 1980), for control of heavy loads at the SHINE facility, as cribed in Subsection 9b.7.2. Therefore, a heavy load drop into an open IU cell is not credible.
vy Load Drop Scenario 2 - Heavy Load Drop onto an In-Service IU Cell.
ane mechanical failure or operator error during a lift may result in a heavy load drop onto an ervice IU cell. The heavy load can damage the IU cell plug which results in damage to SCAS ponents and result in a release of radioactive material.
NE has applied the applicable guidance from NUREG-0612, Control of Heavy Loads at lear Power Plants (USNRC, 1980), for control of heavy loads at the SHINE facility, as cribed in Subsection 9b.7.2. Therefore, a heavy load drop into an in-service IU cell is not dible.
vy Load Drop Scenario 3 - Heavy Load Drop onto TPS Equipment ane mechanical failure or operator error during a lift may result in a heavy load drop onto equipment. The heavy load can damage the equipment and result in a release of oactive material.
NE has applied the applicable guidance from NUREG-0612, Control of Heavy Loads at lear Power Plants (USNRC, 1980), for control of heavy loads at the SHINE facility, as cribed in Subsection 9b.7.2. Therefore, a heavy load drop onto TPS equipment is not dible.
2.1.12.3 Accident Consequences tron Driver Assembly System dose consequences of an NDAS failure are evaluated in Section 13a2.2.12.
um Purification System dose consequences of a release of tritium from TPS Scenario 1 are described in tion 13a2.2.12. This scenario bounds the dose consequences for the release of tritium from Scenario 3 and TPS Scenario 4. The dose consequences of TPS Scenario 5 is also cribed in Section 13a2.2.12.
vy Load Drop vy load drop scenarios are not credible; therefore, accident consequences are not evaluated.
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section describes the accident analysis for the limiting scenarios described in tion 13a2.1 and provides a determination of the radiological consequences. Chemical sequences are analyzed in Section 13b.3.
analyses in this section evaluate the applicable radiological consequences of these idents to demonstrate that the SHINE accident dose criteria are met. The SHINE accident e criteria are defined as follows:
- Radiological consequences to an individual located in the unrestricted area following the onset of a postulated accidental release of licensed material would not exceed 500 mrem total effective dose equivalent (TEDE) for the duration of the accident, and
iological Consequence Assessment Development radiological consequence assessment is a multi-step process. Figure 13a2.2-1 provides a phical representation of the process, which is further described in this section. The process lves: (1) calculation of radionuclide inventories, (2) definition of the accident-specific erials-at-risk (MAR), (3) transport methods of radionuclides, (4) development of accident rce terms, and (5) determination of radiological consequences.
ionuclide Inventories most accident scenarios, the MAR were derived from the target solution vessel (TSV) target tion inventory at the end of [ ]PROP/ECI of continuous 30-day irradiation cycles with a
]PROP/ECI downtime between cycles. The constant power level used for the analysis was
.5 kW, which is 110 percent of design operating power. The TSV inventory calculation udes effects from fission, transmutation, activation, and decay. The calculation contains time s from the start of irradiation through the end of the approximately [ ]PROP/ECI irradiation e and additional time steps that account for decay post-shutdown, as needed. [
]PROP/ECI was selected for the irradiation cycle based on the anticipated replacement period arget solution.
ident-Specific Materials-At-Risk accident scenarios involving the release of radionuclides produced in the target solution, a ion of the inventory was released based on various factors unique to each scenario. The ting inventory was selected based on the assumed start time for each scenario and was then itioned based on scenario specific nuclide removal mechanisms. For the source term ermination and the determination of resulting dose, the radionuclides are grouped into three ups: iodine, noble gases, and non-volatiles. The non-volatile group encompasses the onuclides which do not fall into the other groups.
scenarios involving the release of tritium, the available MAR was determined based on the ting operational values for the affected systems or components.
NE Medical Technologies 13a2.2-1 Rev. 1
transport of radioactive material for the accident analysis was quantitatively evaluated using e-step process. The result of the five-step method is a combination of the airborne release tion (ARF) and the leak path factor (LPF) into a single parameter (i.e., the LPF model) which then used to replace these two terms in the typical five-factor formula found in REG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook (USNRC, 1998).
Step One: Identify control volumes and leakage paths This step identifies the control volumes, leakage paths, and heat sinks that define the model geometry. The performance of this step establishes the geometric information for the LPF calculations including gas and liquid volumes, surface areas for deposition, flow path areas and characteristics that influence leakage, elevations, surface areas for liquid and air heat sinks, and system information for ventilation rates and filtering. The control volumes used in the analysis include the source volume, the building volume, and the environment. Leakage paths are treated as junctions between the control volumes.
Step Two: Quantify scenario progression source histories This step defines the specific physical phenomena for each scenario. This includes the accident initiator and source volume location to quantify the amount or rate at which materials and energy can be released. Where pertinent, this also includes the amounts or rates at which gases and aerosols are evolved. It also includes the initial conditions including gas temperature and pressure, liquid mass and temperature as appropriate, and initial released activity. Further rates of activity release are also specified for evolution of iodine and non-volatile evolution. The performance of this step quantifies initial sources and source rate histories for radionuclides and other mass and energy sources that drive material transport.
- Iodine partitioning was calculated using equations based on the definitions found in NUREG/CR-5950, Iodine Evolution and pH Control (USNRC, 1992), without modifications.
- Bursting bubble aerosols are treated using a linear relationship between entrainment coefficients and the volumetric flow rate with entrainment coefficients consistent with the small amount of dissolved materials present in the target solution.
- Spray and free-fall aerosols use a constant airborne release fraction found in NUREG/CR-6410 (USNRC,1998).
- Radiolysis was treated as either an instantaneous release of hydrogen or as entrainment with source generation based on decay power inputs for long-term problems.
Step Three: Quantify leakage rates between volumes In this step, leakage rates between volumes are defined in either closed-form expressions for input into the LPF model or in terms of time histories generated by the LPF model. Leakage rates are driven by pressure, gas density differences, and barometric breathing.
The gas pressure in each volume was determined using a combination of conservation laws for mass and energy, temperature-dependent specific heats, and the ideal gas law. In addition to intra-volume flows, the following heat and mass transfer rates are considered:
NE Medical Technologies 13a2.2-2 Rev. 1
for laminar and turbulent conditions
- Natural convection heat transfer between a liquid pool and the wall or floor using the heat sink temperature and standard correlations accounting for laminar and turbulent conditions
- Water pool evaporation using standard mass transfer correlations accounting for laminar and turbulent flow
- Heat sources to the gas and water pool to account for equipment
- Decay power in the gas and liquid based on the local activity Pressure-driven flow through each junction was calculated using the standard compressible flow equation for pressure-driven flow. Pressure-driven flow through shielding cover plug gaps was calculated using the equation for plane Poiseuille flow. In the absence of pressure-driven flow, counter-current flow was assumed to be induced and was calculated based on the difference in pressure from one side of the cover plug to the other.
Step Four: Quantify removal mechanisms This step establishes expressions for removal rates of radionuclides for use in the LPF model. Of particular interest are iodine adsorption, non-volatile aerosol deposition, and removal by filters along a flow path. Radioactive decay and build-in are not considered in this step.
- Removal by aerosol settling was evaluated using Stokes law and the calculated equilibrium particle size for a hygroscopic particle. The use of the equilibrium particle size is valid based on the time scale of the accident sequences.
- Consideration of removal by filters in the accident flow paths was done by assigning decontamination factors for noble gases, iodine, and aerosols to each filter.
- Removal due to barometric breathing was based on an analysis of the peak-to-peak magnitude of environmental pressure fluctuations. Eight years of pressure data were collected and analyzed to provide a bounding estimate of the barometric breathing flow rate for the facility.
Iodine removal through absorption was determined using models developed for Library of Iodine Reactions in Containment (LIRIC) and Iodine Model for Containment Codes (IMOD) codes written and validated by Atomic Energy of Canada Limited (AECL). Typical values for adsorption and desorption rate constants determined from experimental data were used in the analyses. Experimentally-determined saturation capacities were also used. Desorption was neglected because it has been shown to have little impact on equilibrium iodine concentrations since the rate constants for desorption are more than a factor of 1,000 lower than adsorption. The deposition velocity was estimated from correlations which serve as the basis for the models used in NUREG-75/014, Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants (USNRC, 1975) (WASH-1400).
Similarly, the diffusion coefficients were found using standard methods which form the basis for the values in WASH-1400.
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The final LPF model consists of a coupled set of differential equations that express the masses or mass fractions of radionuclides in each region as a function of time, including airborne and removed quantities.
resulting LPF model 30-day cumulative leakage factors and the 10-minute leakage factors provided in Table 13a2.2-1 and Table 13a2.2-2, respectively. These values represent the duct of the leak path factors and the airborne release fractions for each group of oisotopes.
ident Source Terms accident source terms for each accident scenario was then determined by the use of the
-factor formula as described in Section 3.2.5.2 of NUREG/CR-6410 (USNRC, 1998). The LPF del described above replaces the combined ARF x LPF term in the formula as it combines h of these parameters. The cumulative ARF x LPF values represent the time-dependent age of radionuclides. The ARF x LPF values are calculated for the leakage from the source me to the building for the worker dose (10 minutes), and the source volume to the ironment for the public dose (duration of the event).
respirable fraction (RF) and damage ratio (DR) are conservatively set at 1.0. Radioactive ay of the MAR for various times after the initiation of an accident was included in the source development. The MAR inventory was tabulated based on various decay times.
ermination of the accident source term was accomplished by linear interpolation between the ulated values, which is a conservative estimate.
atmospheric dispersion factor (/Q) was also applied for releases to the environment as ussed below.
iological Consequences radiological consequences for each accident are presented in terms of TEDE.
methodology uses external and internal radiation sources to calculate the effective external e equivalent and dose equivalent for external sources and committed effective dose ivalent and committed dose equivalent for internal sources. The TEDE and the total dose ivalent (TDE) are measures of the total body and organ doses respectively, received from rnal and internal radiation sources.
ernal doses are calculated for submersion in contaminated air for both the public and worker appropriate dose conversion factors (DCF) for submersion for each radionuclide. Inhalation es are calculated based on the respirable fraction, DCF for inhalation, and breathing rate.
rker dose was calculated based on a facility evacuation time of 10 minutes. The dose version factors used in the analysis are taken from ICRP Publication 119, Compendium of e Coefficients based on ICRP Publication 60 (ICRP, 2012) and Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, and Soil (EPA, 1993).
rker dose prior to initial evacuation has been evaluated. Immediate operator action inside the lity is not required to stabilize accident conditions.
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ause it is expected that tritium recovery can be accomplished within this time frame. The /Q es are calculated at the nearest point along the site boundary and at the nearest resident tion. The maximum calculated value over all directions of the 50th percentile /Q was used both receptor locations. A ground release was used as the release point.
environmental and meteorological conditions used to develop the atmospheric dispersion ors are discussed in Section 2.3.
servatism itional areas of conservatism included in the determination of radiological consequences ude:
- Conservative TSV power history and operational cycle: The TSV power history was derived from nearly continuous TSV operation over a [ ]PROP/ECI period at a power level that exceeds the design power level by ten percent. No credit was taken for medical isotope extraction activities that normally occur during the operation of the SHINE facility.
- Conservative statistical bounding of nuclide inventory: Due to inherent uncertainties in MCNP5, multiple unique sets of results were run through ORIGEN-S to determine the nuclide inventories. The nuclide inventories were analyzed such that a 95 percent confident 95th percentile upper bound was determined for each nuclide. These uncertainties on individual nuclides, 0 to 35 percent, were added to the safety basis inventory to account for the uncertainties inherent to the methods used.
- Conservative estimation of nuclide decay (linear interpolation in lieu of exponential decay): Analyses which account for the decay of nuclides between time steps use linear interpolation in lieu of exponential decay, which increases the available radionuclide inventory at the intervening points.
- Condensation was conservatively neglected in the LPF model.
ertainties ertainty in the radionuclide inventory was evaluated using statistical modeling to account for ertainties associated with the use of Monte Carlo N-Particle Transport Code (MCNP)
NL, 2011) in the SHINE Best Estimate Neutronics Model (BENM). The modeling produced a lide-dependent multiplication factor ranging from approximately 0 to 35 percent increase in nuclide inventory per nuclide. For the radionuclides which were increased, the average ease was approximately 2.5 percent, and the total estimated increase in inventory was roximately 1 percent. The unweighted uncertainty associated with the multiplication factors approximately 12 percent. Given that the majority of radionuclides either did not receive an ease or received an increase less than 10 percent and that the multiplication factor only eased the inventory this uncertainty is considered to be negligible.
ed on the results of the validation activities for the LPF model, described below, there is no itional uncertainty associated with the LPF model used in the analysis.
DCFs used in the analysis are well-recognized and are used without consideration of ertainty in the values.
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wever, SHINE conservatively uses the value for the highest sector.
of Computer Codes PAVAN computer code was used to calculate the short-term atmospheric dispersion (/Q) ors for an effluent release to the public. PAVAN is described in NUREG/CR-2858, PAVAN:
Atmospheric-Dispersion Program for Evaluating Design-Basis Accidental Releases of ioactive Materials from Nuclear Power Stations (USNRC, 1982). The code was used as scribed in Regulatory Guide 1.145, Revision 1, Atmospheric Dispersion Models for Potential ident Consequence Assessments at Nuclear Power Plants (USNRC, 1983). No additional dation was performed for the PAVAN code.
LPF model implementation was performed using Mathcad Version 15. The LPF model ulation was validated at three levels: (1) function tests, consisting of tests of individual tions for rate processes, heat transfer, mass transfer, and aerosol behavior; (2) integral s, which used the model for a simplified transient using one or more rate processes which is n compared against independent solutions to simplified differential equations; (3) use of a plified LPF model, which notes key leakage rates and integrates the equations for activity for h of the modeled regions assuming the leakage rates are constant. The results of the dation for the LPF model showed good agreement in each of the test categories and revealed ignificant sources of uncertainty due to the use of the model. The LIRIC and IMOD elations used in the LPF model have been extensively validated against experimental data.
additional validation was performed.
radionuclides included in the target solution inventory are determined using ORIGEN-S NL, 2011) with input from the SHINE BENM which provided the neutron flux and cross-tions for the data library used by ORIGEN-S. ORIGEN-S has been extensively validated for in calculating burnup in a variety of applications. No additional validation for ORIGEN-S was ormed. Additional discussion of the use of ORIGEN-S is provided in Section 4a2.6.
2.2.1 IF MAXIMUM HYPOTHETICAL ACCIDENT postulated maximum hypothetical accident (MHA) for the irradiation facility (IF) is a failure of TSV off-gas system (TOGS) pressure boundary leading to a release of TSV radioactive es into the TOGS confinement cell. The nitrogen gas purge system (N2PS) actuates, but the cess vessel vent system (PVVS) flow path is assumed to be completely blocked, causing a ximum pressurization of the irradiation unit (IU) cell. Therefore, the release of radioactive erial from the pressurized primary confinement boundary is hypothetical and bounds other ases of fission products from the IF.
2.2.1.1 Initial Conditions al conditions of the accident are described in Subsection 13a2.1.1.1.
2.2.1.2 Initiating Event initiating event for the MHA is the non-credible failure of the N2PS flow path through the VS system following a credible TOGS failure (described in Subsection 13a2.2.7). The MHA NE Medical Technologies 13a2.2-6 Rev. 1
2.2.1.3 Sequence of Events accident sequence proceeds as follows:
- 1. A failure of the primary system boundary (PSB) in the TOGS causes a release of noble gases and iodine into the TOGS cell.
- 2. The radioactive material is confined by the primary confinement boundary, which is described in Subsection 6a2.2.1.
- 3. The TSV reactivity protection system (TRPS) is assumed to actuate N2PS based on changes in TOGS flow. It is conservative to assume N2PS actuates, as it pressurizes and increases leak rates from the TOGS cell.
- 4. The N2PS flow path through the PVVS is assumed to be completely blocked, increasing the total pressurization of the TOGS cell.
- 5. The released TOGS gases and nitrogen gas pressurize the TOGS cell and begin to flow into the IF by pressure driven flow through leak paths in the primary confinement boundary.
- 6. The radioactive material is then dispersed throughout the IF and exits to the environment through building penetrations.
- 7. Detection of high radiation in the radiological ventilation zone 1 exhaust subsystem (RVZ1e) from the IU cell actuates redundant ventilation dampers and prevents the transport of radioactive material to the environment through this path. The assumed response time for RVZ1e ventilation is 20 seconds from detection of high radiation. A sufficient time delay is provided by design to prevent significant radioactive gases from exiting through this path prior to isolation.
- 8. The TRPS initiates an IU Cell Safety Actuation signal, which terminates irradiation operations and isolates the primary confinement boundary. The TRPS signal may be initiated by a TOGS failure or a RVZ1e high radiation signal.
- 9. The main facility ventilation system (i.e., radiological ventilation zone 2 [RVZ2]) is isolated by the engineered safety features actuation system (ESFAS) within 30 seconds of detectable accident conditions. Leakage to the environment continues through unfiltered leakage pathways.
- 10. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 11. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation area monitor alarms.
- 12. Pressurized flow due to nitrogen is considered to continue for the full accident duration of 30 days.
ety Controls safety controls credited to mitigate the release of radioactive material in this scenario are:
- Primary confinement boundary
- Ventilation radiation monitors
- Ventilation isolation mechanisms
- TRPS IU Cell Safety Actuation NE Medical Technologies 13a2.2-7 Rev. 1
TOGS pressure boundary leak only affects the TOGS operability. No other damage is sidered.
2.2.1.5 Radiation Source Terms initial MAR for the MHA is 100 percent of the noble gases and iodine present in the target tion at the end of [ ]PROP/ECI of continuous 30-day irradiation cycles with a
]PROP/ECI downtime between cycles. The power level used for the analysis is 137.5 kW, ch is 110 percent of design operating power. The complete inventory of radioactive gases in PSB is instantaneously transported into the TOGS cell at the beginning of the event uence.
accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.
2.2.1.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The public and worker dose consequences from the MHA are provided in le 13a3-1 and meet the acceptance criteria.
2.2.2 INSERTION OF EXCESS REACTIVITY discussed in Subsection 13a2.1.2.3, no releases are expected to occur as a result of insertion xcess reactivity events. There are no consequences to the workers or the public from excess ctivity events as discussed below. Accident consequences resulting from excess reactivity nts that reference other subsections are evaluated in those respective subsections.
2.2.2.1 Initial Conditions al conditions for insertion of excess reactivity events are described in Subsection 13a2.1.2.1.
2.2.2.2 Initiating Event section 13a2.1.2 identifies the postulated initiating events and scenarios with respect to an rtion of excess reactivity.
subcritical assembly is protected from excessive power with actuation signals from the PS on high flux in Mode 1 and Mode 2. When a power excursion occurs, the strong negative dback inherently reduces reactivity and power. However, during some transients the power eases to a level higher than the steady state maximum power level of 125 kW.
scenario that was found most limiting is the high power due to high neutron production and reactivity at cold conditions (Scenario 4 described in Subsection 13a2.1.2.2). This limiting nario adds substantial reactivity to the operating system and results in the highest calculated k power.
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limiting sequence of events is as follows:
- 1. The TSV is operating at steady state at the licensed power limit of 125 kW, with an operating keff of approximately [ ]PROP/ECI.
- 2. The accelerator ceases to produce neutrons because of an upset, dropping source neutron production to 0 percent. The TRPS detects the loss of neutron production and begins the [ ]PROP/ECI delay prior to initiating a Driver Dropout.
- 3. The primary closed loop cooling system (PCLS) continues to function and cool the target solution in the TSV.
- 4. As the system reduces in power, void leaves the solution, causing a reactivity increase of up to [ ]PROP/ECI. The keff after void loss is approximately [ ]PROP/ECI.
PROP/ECI
- 5. At [ ] after the loss of neutron production, target solution cooling results in a temperature decrease of less than 7°C. This results in a reactivity increase of up to [ ]PROP/ECI. The system remains subcritical.
- 6. The accelerator output is restored at [ ]PROP/ECI just prior to the TRPS Driver Dropout initiating. Power increases to a level that is greater than the steady state power before the upset occurred, with a peak power calculated at [ ]PROP/ECI. This power level would result in a TRPS IU Cell Safety Actuation on high wide range neutron flux.
ety Controls this most limiting scenario, the safety controls that prevent consequences of an insertion of ess reactivity event and ensure that damage to the PSB does not occur are:
- Low power range neutron flux
- TRPS Driver Dropout, resulting in redundant neutron driver assembly system (NDAS) high voltage power supply (HVPS) breakers opening
- Redundant HVPS breakers on neutron driver
- High wide range neutron flux
- TRPS IU Cell Safety Actuation on high wide range neutron flux itional safety controls that prevent consequences of other scenarios described in section 13a2.1.2.2 include:
- TRPS IU Cell Safety Actuation on the following parameters:
- High time-averaged neutron flux
- High wide range neutron flux
- High source range neutron flux
- Low PCLS temperature
- High PCLS temperature
- Low PCLS flow
- TRPS IU Cell Nitrogen Purge on the following parameters:
- Low-high TSV dump tank level
- High-high TSV dump tank level NE Medical Technologies 13a2.2-9 Rev. 1
TRPS is designed to end the event and place the target solution in a safe shutdown dition without the need for operator action. The TRPS prevents challenges to the integrity of PSB. No equipment damage results from the postulated insertion of excess reactivity event.
2.2.2.5 Radiation Source Terms ause the postulated insertion of excess reactivity events do not exceed any design limits or se damage to the PSB, there is no radiation source term.
2.2.2.6 Radiological Consequences ause the insertion of excess reactivity events do not exceed any design limits or cause age to the PSB, there are no radiological consequences to workers or the public.
2.2.3 REDUCTION IN COOLING section discusses the analysis and determination of consequences for the reduction in ling event.
2.2.3.1 Initial Conditions al conditions of the loss of or reduced PCLS flow event are:
- The TSV is operating normally at steady-state in Mode 2 (irradiation).
- 137.5 kW thermal power generated within the target solution.
- PCLS is providing greater than the minimum flow rate of [
]PROP/ECI.
- PCLS supply temperature is less than the maximum supply temperature of 77°F (25°C).
- The TSV was filled to the minimum cold fill volume of [
]PROP/ECI, which maximizes power density.
- Average target solution temperature of up to 175°F (80°C).
- Target solution decay heat based on end of life target solution conditions, which generates the highest decay heat.
2.2.3.2 Initiating Events section 13a2.1.3 describes two possible scenarios for the reduction in cooling accident:
oss of normal power, and (2) loss of PCLS flow. Of these two, the loss of or reduced PCLS is the limiting event, because in this scenario accelerator operation could continue with uced cooling flow. In the loss of normal power event, the accelerator is unable to function out continuous power, greatly reducing heat production. Target solution is drained to the TSV p tank after an allowable delay, and decay heat is removed by the light water pool.
scenario postulates a reduction in PCLS flow without a loss of off-site power (LOOP). The tron driver continues to operate. There are numerous possible causes that could result in loss otal flow or flow reduction in the PCLS, including failure of the power supply to the pump, p shaft lockup or failure, or operator error. Loss of PCLS cooling could also result because of NE Medical Technologies 13a2.2-10 Rev. 1
2.2.3.3 Sequence of Events scenario starts with the TSV in Mode 2, operating normally at full power. PCLS cooling flow duced, resulting in increased TSV temperature. Depending on the failure, PCLS supply perature may also increase. The neutron driver is de-energized by the TRPS Driver Dropout ow PCLS flow. For loss of PCLS cooling capability events due to high temperatures, the PS Driver Dropout would also actuate on high PCLS temperature.
low PCLS flow trip is a minimum of [ ]PROP/ECI, and the PCLS supply perature is a maximum of 77°F (25°C).
TRPS Driver Dropout opens the NDAS HVPS breakers, terminating neutron production.
r a 180 second delay, the TRPS initiates an IU Cell Safety Actuation on loss of PCLS flow (or PCLS temperature). The TSV dump valves open and the target solution is dumped to the dump tank. The light water pool provides passive cooling to the TSV dump tank for the oval of decay heat from the target solution.
ety Controls following safety controls prevent a reduction in cooling event and ensure that the target tion in the TSV does not boil:
- TRPS Driver Dropout on loss of PCLS flow and high PCLS temperature
- TRPS IU Cell Safety Actuation on low PCLS flow rate and high PCLS temperature
- Light water pool
- NDAS HVPS trip breakers
- Redundant TSV dump valves 2.2.3.4 Damage to Equipment TRPS is designed to end the event and place the target solution in a safe shutdown dition without the need for operator action. The TRPS also prevents challenges to the grity of the PSB. No equipment damage results from the postulated reduction in cooling nt.
2.2.3.5 Radiation Source Terms lyses show that if the PCLS supply temperature exceeds the operating limit of 77°F (25°C) or PCLS flow rate is below the operating limit of [ ]PROP/ECI, TRPS indicates an Cell Safety Actuation, the target solution is transferred to the TSV dump tank where it is sively cooled by the light water pool, and there is no boiling in the TSV or in the TSV dump k.
ause the postulated reduction in cooling events do not exceed any design limits or cause age to the PSB, there is no radiation source term.
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ause the postulated reduction in cooling events do not exceed any design limits or cause age to the PSB, there are no radiological consequences to workers or the public from a uction in cooling event.
2.2.4 MISHANDLING OR MALFUNCTION OF TARGET SOLUTION bounding scenario analyzed as a design basis accident (DBA) for mishandling or function of target solution is a loss of the PSB integrity which results in a release of target tion into the IU cell. This scenario is described in Subsection 13a2.1.4.2 as Scenario 1b.
2.2.4.1 Initial Conditions TSV is operating at 110 percent of its design power limit at the time of the initiating event.
itional initial accident conditions are described in Subsection 13a2.1.4.1.
2.2.4.2 Initiating Event accident sequence is initiated by a catastrophic loss of PSB integrity. Potential causes of the ating event are discussed in Subsection 13a2.1.4.1.
2.2.4.3 Sequence of Events assumed that the primary confinement boundary is intact and performs a mitigation function respect to radionuclide transport from the IU cell to the IF. The primary confinement ndary components are designed to maintain their integrity under postulated accident ditions and are maintained in accordance with the facility configuration management and ntenance requirements.
- 1. A failure of the PSB leads to mixing of irradiated target solution with the IU cell light water pool.
- 2. Radioactive material enters the gas space above the light water pool and is confined by the primary confinement boundary, which is described in Section 6a2.2.
- 3. Some radioactive material is transported into the IF through minor leakage paths around penetrations in the confinement boundary.
- 4. Detection of airborne radiation in RVZ1e actuates the primary confinement boundary isolation valves and an IU trip within 20 seconds of detection. A sufficient time delay is provided by the holdup volume in RVZ1e to prevent radioactive gases from exiting through RVZ1e prior to isolation.
- 5. The radioactive material is then dispersed throughout the IF and exits the facility to the environment through building penetrations.
- 6. Detection of high radiation in the RCA actuates ventilation dampers between the RCA and the environment and minimizes the transport of radioactive material to the environment.
- 7. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 8. Facility personnel evacuate the immediate area upon actuation of the radiation alarms.
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owing the failure of the PSB, it is assumed that the MAR is instantly well-mixed with the light er pool. Gases immediately evolve out of the pool and into the IU cell gas space. For the poses of the accident analysis, it is assumed that the N2PS is operating and causes ssurization of the IU cell. Radiation transport is driven by pressure-driven flow between the IU and the IF. Reduction in the MAR occurs during the release due to adsorption of iodine onto IU cell walls and other surfaces until equilibrium conditions are established. The majority of MAR is transported to the IF through leakage through the primary confinement boundary.
nsport to the environment occurs through leakage around penetrations in the RCA boundary.
ety Controls safety controls credited for mitigation of the dose consequences for this accident are:
- Primary confinement boundary
- Ventilation radiation monitors
- Ventilation isolation mechanisms
- Holdup volume in the RVZ1e 2.2.4.4 Damage to Equipment mical and radiological contamination may occur to systems within the IU cell. The tamination does not affect the safety function of the affected systems.
owing isolation of the primary confinement boundary, leakage between the IU cell and the IF riven primarily by pressure-driven flow caused by N2PS. The IU cell sealing is a significant tributor to the function of the primary confinement boundary and will maintain its function er accident conditions.
light water pool is required to act as a passive heat sink to remove decay heat from the diated target solution. The light water pool is constructed with a stainless steel liner ounded by concrete and maintains the light water pool water inventory and will not be cted by the release of target solution.
2.2.4.5 Radiation Source Terms initial MAR for this scenario is the TSV target solution inventory at the end of approximately
]PROP/ECI of continuous 30-day irradiation cycles with a [ ]PROP/ECI downtime ween cycles. The power level used for the analysis is 137.5 kW, which is 110 percent of ign operating power. The entire radionuclide inventory in the TSV is instantaneously released he light water pool and dispersed uniformly throughout the pool.
accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.
ne is partitioned by assuming that the iodine present is fully dissolved into the target solution r to the initiating event and none is present in the gas space of the TSV. Once the event NE Medical Technologies 13a2.2-13 Rev. 1
ell due to the temperature difference of the warm gas and the cold walls is also credited as a oval mechanism. As iodine is deposited on the cell walls, more iodine is evolved from the t water pool and into the gas space. The iodine partitioning determines the transport of tile iodine out of the pool.
me radionuclides deposited in the light water pool are released to the gas space as an aerosol n radiolytically-generated hydrogen gas bubbles burst at the pool surface. Radiolysis omes a long-term source of both non-volatiles and iodine.
e the MAR is released into the gas space of the IU cell there are several paths through which age could occur. The primary leak path will be around the IU cell plug perimeter. Potential paths are modeled as a single leakage junction to the IF.
2.2.4.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are provided in Table 13a3-1 and meet the ident dose criteria.
2.2.5 LOSS OF OFF-SITE POWER 2.2.5.1 Initial Conditions ility power is being supplied from off-site by the electric utility. The initial conditions for the OP event are further described in Subsection 13a2.1.5.1.
2.2.5.2 Initiating Event initiating event is a LOOP resulting in the loss of the normal electrical power supply tem (NPSS) function.
2.2.5.3 Sequence of Events sequence of events for an extended LOOP is described in Subsection 13a2.1.5.2.
facility combustible gas management system described in Chapter 6 is designed to function wing a LOOP to disperse the combustible gases that are generated by radiolysis. This tem removes combustible gases through the PVVS carbon beds and filters to the ironment, through the PVVS alternate vent path.
ety Controls safety controls credited for prevention of accidents resulting from a LOOP event are:
- Uninterruptible electrical power supply system (UPSS)
- NDAS HVPS breakers
- TSV dump valves
- Light water pool NE Medical Technologies 13a2.2-14 Rev. 1
- PVVS alternate vent path
- PVVS carbon guard and carbon delay beds 2.2.5.4 Damage to Equipment LOOP event does not result in any damage to equipment.
safety-related functions of the equipment supplied by the UPSS are uninterrupted; therefore, safety-related functions continue to be performed. Irradiation processes stop, and the target tion is drained from operating TSVs to their respective TSV dump tank. Decay heat is oved by natural convection to the light water pool. The combustible gas management system inates the risk of a release of radioactive material due to deflagration.
2.2.5.5 Radiation Source Terms ause the postulated LOOP event does not result in the loss of safety-functions of the ipment supplied by the UPSS, there is no radiological source term for this accident sequence.
2.2.5.6 Radiological Consequences ause the postulated LOOP event does not result in the loss of safety-functions of the ipment supplied by the UPSS, there are no radiological consequences for this accident uence.
2.2.6 EXTERNAL EVENTS facility structure is designed to withstand credible external events as described in section 13a2.1.6. Most of the analyzed accidents involving credible external events are vented by the facility structure or seismic qualification of affected SSCs. The only postulated ident scenario resulting in a radiological release involving an external seismic event is a m release from simultaneous failure of multiple NDAS units. The simultaneous release of m from all eight operating neutron driver assemblies is analyzed as a DBA. This scenario is cribed in Subsection 13a2.1.6 as Scenario 3. The consequences of this accident are lyzed in this subsection.
2.2.6.1 Initial Conditions initial conditions for external events are described in Subsection 13a2.1.6.1.
2.2.6.2 Initiating Event eismic event is the initiating event for a tritium release into multiple IU cells. All NDAS elerators experience vacuum boundary component failures and cause a pressurized release itium and SF6 gas into the eight IU cells simultaneously as a result of the seismic event.
initial accident conditions for each IU cell are the same to those accident conditions involving NDAS of a single IU cell, as described in Subsection 13a2.1.12.1.
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accident sequence proceeds as follows:
- 1. The initiating event is a seismic event that causes the simultaneous vacuum boundary component failure in all eight NDAS units, instantaneously releasing tritium and SF6 gas into the IU cells.
- 2. The IU cells become slightly pressurized due to the mass of released SF6 gas.
- 3. Some tritium is transported into the IF through penetrations in the confinement boundary.
- 4. Detection of high accelerator TPS interface system (ATIS) mixed gas return line pressure actuates the primary confinement boundary isolation valves and irradiation unit trips within 20 seconds of detection. A sufficient time delay is provided by the holdup volume in RVZ1e to prevent radioactive gases from exiting through RVZ1e prior to isolation.
- 5. Tritium migrates to the IF through the IU cell plugs and is released to the environment.
- 6. Detection of high radiation in the RCA actuates ventilation dampers between the RCA and the environment and minimizes the transport of radioactive material to the environment.
- 7. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 8. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation alarms.
iation transport is driven primarily by barometric breathing between the IU cell and the IF.
safety-related SSCs in the IU cell do not fail during a seismic event, but the NDAS and its rnal components are not safety-related and cannot be relied upon to remain intact following a ign basis earthquake.
operator actions are taken or required to reach a stabilized condition or to mitigate dose sequences.
ety Controls safety controls credited for mitigation of the dose consequences for this accident are:
- Primary confinement boundary
- IU Cell Safety Actuation on high ATIS mixed gas return line pressure
- Ventilation isolation mechanisms
- Holdup volume in the RVZ1e assumed that the primary confinement is intact and performs a mitigation function with pect to radionuclide transport from the IU cells to the IF. The primary confinement boundary ponents are designed to maintain their integrity under postulated accident conditions and are ntained in accordance with the facility configuration management and maintenance systems.
2.2.6.4 Damage to Equipment ure of the NDAS vacuum boundary does not cause subsequent damage to equipment. While NDAS vacuum boundary integrity is not seismically qualified to maintain integrity, the NDAS esigned to maintain structural integrity during and following a design basis earthquake.
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acted by the accident sequence.
2.2.6.5 Radiation Source Terms initial MAR for this scenario is a total of [ ]PROP/ECI of tritium from all of the tron driver assemblies.
accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.
2.2.6.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are provided in Table 13a3-1 and meet the ident dose criteria.
2.2.7 MISHANDLING OR MALFUNCTION OF EQUIPMENT bounding scenario analyzed for mishandling or malfunction of equipment events is a loss of PSB integrity which results in a release of off-gas into the TOGS cell. This scenario is cribed in Subsection 13a2.1.7.2 as Scenario 1.
2.2.7.1 Initial Conditions al accident conditions are described in Subsection 13a2.1.7.1.
2.2.7.2 Initiating Event accident sequence is initiated by a failure of the PSB in the TOGS within the TOGS cell. The se of the initiating event is discussed in Subsection 13a2.1.7.
2.2.7.3 Sequence of Events accident sequence proceeds as follows:
- 1. A failure of the PSB in the TOGS causes a release of noble gases and iodine into the TOGS cell.
- 2. The radioactive material is confined by the primary confinement boundary, which is described in Section 6a2.2.
- 3. Some radioactive material is transported into the IF through penetrations in the confinement boundary.
- 4. The radioactive material is then dispersed throughout the IF and exits to the environment through building penetrations.
- 5. Detection of high radiation in the RVZ1e ventilation from the IU cell actuates ventilation dampers and minimizes the transport of radioactive material to the environment. The assumed response time for RVZ1e ventilation is 20 seconds from detection of high NE Medical Technologies 13a2.2-17 Rev. 1
- 6. The TRPS initiates an IU Cell Safety Actuation signal which terminates irradiation operations and isolates the primary confinement boundary. The TRPS signal may be initiated by a TOGS failure or a RVZ1e high radiation signal. The N2PS actuates.
- 7. The main facility ventilation system (i.e., RVZ2) is isolated by the ESFAS within 30 seconds of detectable accident conditions. Leakage to the environment continues through unfiltered leakage pathways.
- 8. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 9. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation area monitor alarms.
ortion of the gaseous iodine is adsorbed onto the cell walls, while the majority of the available R is transported to the IF through pressure-driven flow caused by the N2PS and leakage ugh the primary confinement boundary. Transport to the environment occurs through etrations in the RCA boundary.
ety Controls safety controls credited for mitigation of the dose consequences for this accident are:
- Primary confinement boundary
- Ventilation radiation monitors
- Ventilation isolation mechanisms
- Holdup volume in the RVZ1e assumed that the primary confinement boundary is intact and performs a mitigation function respect to radionuclide transport from the TOGS cell to the IF. The primary confinement ndary components are designed to maintain their integrity under postulated accident ditions and are maintained in accordance with the facility configuration management and ntenance systems.
2.2.7.4 Damage to Equipment TOGS zeolite bed may continue to function following a failure of the TOGS but is not dited for source term reduction following the initiating event. Similarly, under normal operating ditions, the recirculating ventilation in the TOGS cell provides filtration which may reduce e consequences but is not credited to remain functional under accident conditions. These umed failures are conservative.
kage of the TOGS pressure boundary does not cause subsequent damage to equipment dited for safety.
owing isolation of the primary confinement boundary, leakage between the TOGS cell and IF is driven primarily by pressure-driven flow caused by the N2PS. The leakage paths ween the cell and the IF are not impacted by the accident sequence. The TOGS cell seals are gnificant contributor to the function of the primary confinement boundary and maintains its tion under accident conditions.
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initial MAR for this scenario is a fraction of the TSV target solution inventory described in tion 13a2.2. The initial MAR for this accident sequence is 100 percent of the noble gases and ne present in the TOGS gas space while it is operating. Non-volatiles are not included in this ident sequence because the system is designed as a gas-handling system.
LPF model values used in the source term development for the public and worker doses are vided in Table 13a2.2-1 and Table 13a2.2-2, respectively.
2.2.7.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are shown in Table 13a3-1 and meet the ident dose criteria.
2.2.8 LARGE UNDAMPED POWER OSCILLATION described in Subsection 13a2.1.8, power oscillations that occur in the subcritical assembly self-limiting as a result of the inherent design and safety characteristics of the subcritical embly, operating parameters, and plant response to transients. TRPS setpoints for high wide ge and high time-averaged neutron flux are set to actuate on high neutron flux before a large er oscillation occurs that challenges design limits. The IU Cell Safety Actuation results in the dump valves opening and target solution draining from the TSV to the TSV dump tank.
s, there are no consequences to workers or the public.
2.2.8.1 Initial Conditions al accident conditions are described in Subsection 13a2.1.8.1.
2.2.8.2 Initiating Event ential causes of power oscillations in the TSV are described in Subsection 13a2.1.8.1.
2.2.8.3 Sequence of Events accident sequence proceeds as follows:
- 1. An oscillation in power occurs as a result of one of the potential causes described in Subsection 13a2.1.8.1.
- 2. TSV reactivity oscillates due to the power oscillation but does not become undamped due to inherent design and safety characteristics of the TSV and operating parameters.
- 3. TRPS high neutron flux limits cause the IU to shutdown before a power oscillation challenges design limits.
- 4. The TSV dump tank valves automatically open and the target solution is dumped by force of gravity into the subcritical dump tank with favorable geometry, ending the event sequence.
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- The design and safety characteristics of the TSV to resist undamped power oscillations.
- IU Cell Safety Actuation initiated by TRPS
- TRPS high neutron flux trips
- TSV dump valves and TSV dump tank 2.2.8.4 Damage to Equipment damage to equipment occurs because power oscillations in the TSV are self-limiting and do become large undamped power oscillations. The TRPS high neutron flux limits halt power illations before they challenge design limits.
2.2.8.5 Radiation Source Terms ause power oscillations in the TSV are self-limiting and because the TRPS acts to prevents er levels that challenge design limits, there is no damage to the PSB, and therefore no ation source term.
2.2.8.6 Radiological Consequences ause large undamped power oscillations are shown to not occur and large power oscillations challenge design limits are halted before they occur, there are no radiological consequences orkers or the public.
2.2.9 DETONATION AND DEFLAGRATION IN THE PRIMARY SYSTEM BOUNDARY release of hydrogen and oxygen by radiolysis from the target solution during and after diation may lead to high concentrations of hydrogen, which may then result in detonation or agration within the PSB. Normally, the TOGS provides ventilation of the headspace above TSV to maintain hydrogen concentrations below the lower flammability limit (LFL). A failure of TOGS to perform its design function may result in conditions that could lead to a hydrogen onation or deflagration, as described in Subsection 13a2.1.9.
2.2.9.1 Initial Conditions rogen concentration in the TSV and TOGS prior to the initiating event is assumed to be at e percent. Additional initial conditions are described in Subsection 13a2.1.9.1.
2.2.9.2 Initiating Event ential initiating events are discussed in Subsection 13a2.1.9.1.
2.2.9.3 Sequence of Events accident sequence proceeds as follows:
- 1. A failure causes a single TOGS blower to fail, resulting in a complete loss of flow through the affected train and total loss of TSV dump tank flow.
- 2. The other TOGS blower continues to operate normally.
NE Medical Technologies 13a2.2-20 Rev. 1
- 4. The IU Cell Safety Actuation opens the TSV dump valves and NDAS HVPS breakers, terminating the irradiation process.
- 5. Hydrogen generation in the TSV and TSV dump tank continues due to radiolysis caused by delayed fission and decay radiation. Hydrogen evolution from solution occurs at an increased rate as solution voids collapse.
- 6. Within four seconds, N2PS is at full flow to the dump tank. Hydrogen and other gases are vented to PVVS through the combustible gas management system exhaust point. Gases pass through the PVVS carbon guard and carbon delay beds before being exhausted from the building at the safety-related exhaust point.
- 7. The remaining TOGS blower continues operation for a minimum of five minutes.
- 8. The combined action of the remaining TOGS blower and N2PS maintains the peak hydrogen concentration less than 7.7 percent; therefore, the peak pressure will not exceed the design pressure of the PSB if a deflagration occurs, and no radiological materials will be released. Detonations cannot occur because this peak concentration is less than the lower detonation limit.
- 9. As delayed fission and decay of short-lived radionuclides decline, the production and evolution of hydrogen declines following shutdown and draining of the TSV to the TSV dump tank. The N2PS continues to provide sweep gas diluting and removing any remaining hydrogen.
ety Controls safety controls credited for prevention of accidents which may cause detonation or agration in the PSB are:
- TOGS capable of maintaining hydrogen concentration within design limits, assuming the worst case single active failure following IU trip (see Subsection 4a2.8.6)
- TOGS low-flow trips (TRPS function)
- TOGS oxygen sensor which detect incipient degradation or failure
- TOGS demister high temperature trips (TRPS function), which detect incipient degradation or failure
- N2PS
- TSV fill line isolation valves mode-permissive interlock
- TSV dump tank level sensors (TRPS function)
- TSV dump tank low flow sensors (TRPS function)
- TSV target solution dump on dump tank level sensors (TRPS function)
- PCLS expansion tank flame arrestor
- Radiation detection in RVZ1e exit from PCLS expansion tank
- Isolation valves in RVZ1e exit from PCLS expansion tank 2.2.9.4 Damage to Equipment ydrogen deflagration occurs at the peak calculated concentration of 7.7 percent, the PSB ains intact. Damage to other primary system components internal to TOGS in the affected n may occur; however, such damage will not result in any release of radiological material.
NE Medical Technologies 13a2.2-21 Rev. 1
ause the PSB remains intact, there is no radiological source term for this accident sequence.
2.2.9.6 Radiological Consequences ause the PSB remains intact, there are no radiological consequences for this accident uence.
2.2.10 UNINTENDED EXOTHERMIC CHEMICAL REACTIONS OTHER THAN DETONATION discussed in Subsection 13a2.1.10, the potential for an unintended exothermic chemical ction within the IF is unlikely. Therefore, there is no radiological consequence to the workers he public.
ident scenario consequences associated with the release of tritium gas are discussed in section 13a2.2.12.
2.2.10.1 Initial Conditions al accident conditions are described in Subsection 13a2.1.10.1.
2.2.10.2 Initiating Event nario 1 - Uranium Metal-Water Reaction in the Neutron Multiplier Assembly initiating event is a small breach of the neutron multiplier cladding, allowing PCLS water into cladding [ ]PROP/ECI.
nario 2 - Ignition of the Activated Carbon Bed in the TPS initiating event is improper TPS system restoration following maintenance, allowing air usion into the impurity removal subsystem.
initiating events associated with unintended exothermic chemical reactions other than onation are further discussed in Subsection 13a2.1.10.1.
2.2.10.3 Sequence of Events nario 1 - Uranium Metal-Water Reaction in the Neutron Multiplier Assembly accident sequence proceeds as follows:
- 1. A small breach of the neutron multiplier cladding occurs, allowing PCLS water to enter the cladding [ ]PROP/ECI.
- 3. The presence of [ ]PROP/ECI inhibits a potential deflagration.
NE Medical Technologies 13a2.2-22 Rev. 1
tank is vented to RVZ1e to prevent hydrogen accumulation in that tank.
- 5. Small amounts of fission products from the multiplier migrate into the PCLS water. The presence of fission products in excess of normal operating levels is detected via in-line radiation monitoring installed in the exhaust of the PCLS expansion tank.
ety Controls
- The design of the neutron multiplier to inhibit deflagration is a safety control (including
[ ]PROP/ECI).
nario 2 - Ignition of the Activated Carbon Bed in the TPS accident sequence proceeds as follows:
- 1. The TPS is improperly restored after maintenance, and air is allowed to enter the TPS impurity removal subsystem.
- 2. The presence of air in the activated carbon bed causes an exothermic reaction.
- 3. The presence of an inert nitrogen atmosphere within the glovebox prevents any fire from damaging the TPS glovebox or causing a release.
ety Controls
- The presence of an inert nitrogen atmosphere and pre-operational and post-maintenance procedures to verify the atmosphere is within limits.
sequences of events associated with unintended exothermic chemical reactions other than onation are further discussed in Subsection 13a2.1.10.2.
2.2.10.4 Damage to Equipment discussed in Subsection 13a2.1.10, no damage beyond the initiating events is anticipated to ur as a result of unintended chemical reactions other than detonation.
2.2.10.5 Radiation Source Terms nario 1 - Uranium Metal-Water Reaction in the Neutron Multiplier Assembly ause a gross failure of the multiplier cladding is unlikely based on its design and a agration due to small leaks in the cladding is unlikely (as described in section 13a2.1.10.2), a uranium metal-water reaction in the neutron multiplier assembly does result in consequences to the worker or the public.
nario 2 - Ignition of the Activated Carbon Bed in the TPS radiation source term created by the release of tritium due to ignition of the activated carbon in the TPS is evaluated further in Subsection 13a2.2.12.
NE Medical Technologies 13a2.2-23 Rev. 1
nario 1 - Uranium metal-water reaction in the neutron multiplier assembly ause a gross failure of the multiplier cladding is unlikely based on its design and a agration due to small leaks in the cladding is unlikely, there are no radiological consequences he worker or the public from this event sequence.
nario 2 - Ignition of the activated carbon bed in the TPS radiological consequences resulting from the release of tritium due to ignition of the vated carbon bed in the TPS is evaluated further in Subsection 13a2.2.12.
2.2.11 SYSTEM INTERACTION EVENTS discussed in Subsection 13a2.1.11, no releases are expected to occur as a result of system raction events. There are no consequences to the workers or the public from system raction events, as discussed below. Accident consequences resulting from system ractions that are referenced to other subsections in Chapter 13 are evaluated in those sections.
2.2.11.1 Initial Conditions re are no unique initial conditions associated with system interaction events.
2.2.11.2 Initiating Event ential causes for system interaction events are described in Subsection 13a2.1.11.
2.2.11.3 Sequence of Events ctional Interactions s of Off-Site Power OP events are described in Subsection 13a2.2.5.
uction of Cooling uction of cooling events are described in Subsection 13a2.2.3, Subsection 13a2.2.5, and section 13a2.2.6.
s of Ventilation tulated loss of ventilation scenarios do not result in radiological consequences based on the ventive controls described in Subsection 13a2.1.11.
itional loss of ventilation scenarios are described in Subsection 13a2.2.5 and section 13a2.2.6.
NE Medical Technologies 13a2.2-24 Rev. 1
safety controls credited for prevention of accidents which may cause radiological sequences from a loss of ventilation are:
- Redundant and diverse TOGS instrumentation (e.g., low flow) that initiates a TRPS signal
- Redundant low and high PCLS temperature trip that initiates a TRPS signal tial Interactions s
tulated fire scenarios in the IF are prevented by fire protection features identified in the fire ards analysis (FHA), as described in Subsection 13a2.1.11.
itional fire scenarios are discussed in Subsection 13a2.2.6, Subsection 13a2.2.9, and section 13a2.2.12.3.
ety Controls safety controls credited for prevention of accidents which may cause radiological sequences from fires are:
- Fire protection measures: low combustible loading, fire detection, manual fire-fighting capabilities, and rated fire barriers to limit the potential for fire initiation and spread
- Administrative controls on maintenance and use of combustible materials
- Catchment pans for the high voltage power supplies thermic Chemical Reaction thermic chemical reaction scenarios are described in Subsection 13a2.1.10.
rnal Flooding tulated internal flooding scenarios in the IF do not result in radiological consequences, as cribed in Subsection 13a2.1.11.
amic Effects amic effects are not present at the SHINE facility, as described in Subsection 13a2.1.11.
man Intervention Interactions described in Subsection 13a2.1.11, human intervention interactions as accident scenario ating events are described in other sections in this chapter as applicable.
2.2.11.4 Damage to Equipment damage to equipment occurs due to system interaction events since the TRPS initiates an IU Safety Actuation or IU Cell Nitrogen Purge as needed prior to exceeding any design limits.
NE Medical Technologies 13a2.2-25 Rev. 1
ause the postulated system interactions do not exceed any design limits or cause damage to PSB, there is no radiation source term.
2.2.11.6 Radiological Consequences ause the postulated system interactions do not exceed any design limits or cause damage to PSB, there are no radiological consequences to workers or the public. Accident sequences resulting from system interactions that are referenced to other subsections in pter 13 are evaluated in those subsections.
2.2.12 FACILITY-SPECIFIC EVENTS majority of the evaluated facility-specific events do not have radiological consequences. The nts which do have radiological consequences are related to the release of tritium into the lity from the neutron driver assemblies or from the tritium purification system. Three potential tions for the release of tritium were analyzed to determine the dose consequences and essary controls. The results of the analysis are presented in this subsection.
2.2.12.1 Tritium Release into an IU Cell release of tritium from an operating neutron driver assembly is analyzed as a DBA. The nding scenario is described in Subsection 13a2.1.12.2 as NDAS Scenario 3, and the dose sequences are analyzed below.
2.2.12.1.1 Initial Conditions al conditions for facility-specific events are described in Subsection 13a2.1.12.1.
2.2.12.1.2 Initiating Event nternal NDAS vacuum boundary component fails and causes a pressurized release of tritium SF6 gas into the IU cell. Potential causes of the initiating event are discussed in section 13a2.1.12.2.
2.2.12.1.3 Sequence of Events assumed that the primary confinement is intact and performs a mitigation function with pect to radionuclide transport from the IU cell to the IF. The primary confinement is designed aintain its integrity under postulated accident conditions and is maintained in accordance the facility configuration management and maintenance programs.
- 1. The initiating event is a vacuum boundary component failure in the NDAS, which instantaneously releases tritium and SF6 gas into the IU cell.
- 2. The IU cell becomes slightly pressurized due to the mass of released SF6 gas.
- 3. Tritium is transported into the IF through penetrations in the confinement boundary.
- 4. Detection of high ATIS mixed gas return line pressure actuates the primary confinement boundary isolation valves and an irradiation unit trip within 20 seconds of detection. A NE Medical Technologies 13a2.2-26 Rev. 1
- 5. Tritium migrates to the IF through penetrations in the primary confinement boundary and is released to the environment.
- 6. Detection of high radiation in the RCA actuates ventilation dampers between the RCA and the environment and minimizes the transport of radioactive material to the environment.
- 7. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 8. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation area monitor alarms.
iation transport is primarily driven by barometric breathing between the IU cell and the IF.
ety Controls safety controls credited for mitigation of the dose consequences for this accident are:
- Primary confinement boundary (IU cell plugs and seals)
- IU Cell Safety Actuation on high ATIS mixed gas return line pressure
- IU cell ventilation isolations
- Holdup volume in the RVZ1e 2.2.12.1.4 Damage to Equipment ure of the NDAS vacuum boundary does not cause subsequent damage to equipment.
r the initial IU cell pressurization has reached equilibrium, leakage between the IU cells and IF is driven primarily by barometric breathing. The leakage paths between the cells and the IF not impacted by the accident sequence.
2.2.12.1.5 Radiation Source Terms initial MAR for this scenario is [ ]PROP/ECI of tritium from the neutron driver embly in the IU cell.
accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.
2.2.12.1.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2.
radiological consequences of this accident scenario are provided in Table 13a3-1 and meet accident dose criteria.
NE Medical Technologies 13a2.2-27 Rev. 1
lease of the tritium inventory from the TPS is analyzed as a DBA. This accident is described ubsection 13a2.1.12.3 as TPS Scenario 1. This analysis establishes bounding radiological ditions for a release of tritium due to a TPS process deflagration, release of tritium to the lity stack, and release of tritium from the tritium storage bed.
2.2.12.2.1 Initial Conditions al conditions for facility-specific events are described in Subsection 13a2.1.12.1.
2.2.12.2.2 Initiating Event event causes a break in the tritium piping and vessels such that the uncontrolled release of entire tritium in-process inventory occurs within the tritium confinement boundary. The tritium finement boundary is described in detail in Section 6a2.2. Potential causes of the initiating nt are discussed in Subsection 13a2.1.12.3.
2.2.12.2.3 Sequence of Events assumed that the tritium confinement boundary is intact and performs a mitigation function respect to radionuclide transport from the TPS to the IF. The tritium confinement boundary ponents are designed to maintain their integrity under postulated accident conditions and are ntained in accordance with the facility configuration management and maintenance grams.
- 1. The initiating event is a break in the tritium piping and vessels which instantaneously releases the entire tritium inventory of the TPS system into the TPS glovebox.
- 2. For the first 20 seconds, tritium escapes from the glovebox to the IF at 10 percent of the maximum GBSS flow rate.
- 3. The glovebox ventilation shuts down after 20 seconds due to the glovebox tritium monitors.
- 4. During the 30 seconds after the initiating event, the TPS room vents to the IF at an elevated rate due to the facility RVZ2 ventilation system.
- 5. The RVZ2 ventilation damper from the TPS room isolates after 30 seconds due to the glovebox tritium monitors.
- 6. The radioactive material is then dispersed throughout the IF and exits the facility to the environment through building penetrations.
- 7. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 8. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation area monitor alarms.
oughout the accident sequence, the leakage rate between the TPS glovebox and the TPS m is constant. After the TPS room ventilation is isolated, radiation transport is driven by air hange between the TPS glovebox and the IF. Transport to the environment occurs through A boundary leak paths. The accident duration used in this analysis is 10 days, after which it is umed that recovery actions will have occurred to stop further release and dispersion of oactive material.
NE Medical Technologies 13a2.2-28 Rev. 1
safety controls credited for mitigation of this accident are:
- TPS room ventilation isolations
- GBSS ventilation isolations
- TPS glovebox tritium radiation monitors
- Tritium confinement boundary, as described in Section 6a2.2 ddition, TPS glovebox deflagration is prevented by:
- TPS glovebox gas space inerted with nitrogen
- TSP glovebox minimum volume prevents deflagration conditions 2.2.12.2.4 Damage to Equipment ure of the TPS piping and vessels does not cause subsequent damage to other equipment.
2.2.12.2.5 Radiation Source Terms initial MAR for this scenario is 236,000 curies of tritium from the TPS equipment in the TPS ebox.
accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.
2.2.12.2.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The radiological consequences of this accident scenario are provided in le 13a3-1 and meet the accident dose criteria.
2.2.12.3 Tritium Release into the Irradiation Facility lease of tritium from a tritium header outside of confinement is analyzed as a DBA. This ident is described in Subsection 13a2.1.12.3 as TPS Scenario 5. This analysis establishes bounding radiological conditions for the direct release from the tritium header to the IF.
2.2.12.3.1 Initial Conditions al conditions for facility-specific events are described in Subsection 13a2.1.12.1.
2.2.12.3.2 Initiating Event event causes a break in the tritium header, releasing the header inventory directly to the IF.
NE Medical Technologies 13a2.2-29 Rev. 1
- 1. The initiating event is a break in the tritium header, which instantaneously releases the entire tritium inventory of the header into the IF.
- 2. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.
- 3. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation area monitor alarms.
- 4. The radioactive material is then dispersed throughout the IF and exits the facility to the environment through building penetrations.
ety Controls safety controls are credited for mitigation of the dose consequences for this accident.
2.2.12.3.4 Damage to Equipment ure of the TPS header does not cause subsequent damage to equipment.
2.2.12.3.5 Radiation Source Terms initial MAR for this scenario is 3000 curies of tritium from the tritium header.
accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.
2.2.12.3.6 Radiological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The radiological consequences of this accident scenario are provided in le 13a3-1 and meet the accident dose criteria.
NE Medical Technologies 13a2.2-30 Rev. 1
Table 13a2.2 Summary of Radiation Transport Terms (Public)
ARF x LPF Nobles Iodine Non-volatiles Tritium Tritium cident Category (30-day) (30-day) (30-day) (10-day) (30-day) ximum Hypothetical Accident (Subsection 13a2.2.1) 9.98E-01 9.98E-01 0 N/A N/A handling or Malfunction of Target Solution 9.98E-01 1.22E-01 8.39E-07 N/A N/A bsection 13a2.2.4) ernal Events (Subsection 13a2.2.6) N/A N/A N/A N/A 3.66E-01 handling or Malfunction of Equipment 9.98E-01 5.72E-01 0 N/A N/A bsection 13a2.2.7) ility-Specific Events (Subsection 13a2.2.12)
- Tritium Release into an IU Cell N/A N/A N/A N/A 3.66E-01
NE Medical Technologies 13a2.2-31 Rev. 1
Table 13a2.2 Summary of Radiation Transport Terms (Worker)
ARF x LPF (10-minute) cident Category Nobles Iodine Non-volatiles Tritium ximum Hypothetical Accident (Subsection 13a2.2.1) 1.19E-02 1.19E-02 0 N/A handling or Malfunction of Target Solution (Subsection 13a2.2.4) 8.24E-03 4.03E-05 9.69E-11 N/A ernal Events (Subsection 13a2.2.6) N/A N/A N/A 1.47E-01 handling or Malfunction of Equipment (Subsection 13a2.2.7) 1.19E-02 1.17E-02 0 N/A ility-Specific Events (Subsection 13a2.2.12)
- Tritium Release into an IU Cell N/A N/A N/A 1.47E-01
- Tritium Release into the Irradiation Facility (Header Release) N/A N/A N/A 1.00E+00 NE Medical Technologies 13a2.2-32 Rev. 1
NE Medical Technologies 13a2.2-33 Rev. 1 section presents the summary and conclusions for the accident analysis for the irradiation lity (IF).
following accident categories were addressed for the irradiation facility:
- Maximum hypothetical accident (MHA)
- Insertion of excess reactivity
- Reduction in cooling
- Mishandling or malfunction of target solution
- Loss of off-site power
- External events
- Mishandling or malfunction of equipment
- Large undamped power oscillations
- Detonation and deflagration affecting the primary system boundary
- Unintended exothermic chemical reactions other than detonation
- System interaction events
- Facility-specific events dose consequences of the bounding accident scenarios evaluated for each accident gory are provided in Table 13a3-1.
analyses in this section evaluated the applicable radiological consequences of these idents and demonstrated that an individual located in the unrestricted area following the onset postulated accidental release of licensed material would not receive a radiation dose in ess of 500 mrem total effective dose equivalent (TEDE) for the duration of the accident.
iological consequences to workers were also evaluated and shown to not exceed 5 rem DE during the accident.
NE has established the MHA based on the maximum consequence to the public. The MHA lf is not a DBA; however, it is used as a metric for understanding radiological risk from the lity.
NE Medical Technologies 13a3-1 Rev. 1
Table 13a3 Irradiation Facility Accident Dose Consequences Public Worker Dose Dose TEDE TEDE cident Category (Bounding Scenario) (mrem) (mrem) ximum Hypothetical Accident (Subsection 13a2.2.1)
- TOGS failure with complete PVVS blockage 366 4800 ertion of Excess Reactivity (Subsection 13a2.2.2) No consequences duction in Cooling (Subsection 13a2.2.3) No consequences handling or Malfunction of Target Solution (Subsection 13a2.2.4)
- Primary system boundary leak into an IU cell 65 1480 s of Off-Site Power (LOOP) (Subsection 13a2.2.5) No consequences ernal Events (Subsection 13a2.2.6) 106 4930 handling or Malfunction of Equipment (Subsection 13a2.2.7) 234 4760 ge Undamped Power Oscillations (Subsection 13a2.2.8) No consequences onation and Deflagration affecting the Primary System Boundary No consequences bsection 13a2.2.9) ntended Exothermic Chemical Reactions other than Detonation No consequences bsection 13a2.2.10) tem Interaction Events (Subsection 13a2.2.11) No consequences ility-Specific Events (Subsection 13a2.2.12)
- Tritium Release into an IU Cell 13 616
- Tritium Release into the Irradiation Facility (Header Release) 27 3140 NE Medical Technologies 13a3-2 Rev. 1
A, 1993. External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance ort No. 12, U.S. Environmental Protection Agency, 1993.
P, 2012. Compendium of Dose Coefficients based on ICRP Publication 60. ICRP lication 119. Ann. ICRP 41(Suppl.), 2012.
NRC, 1975. Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial lear Power Plants, NUREG-75/014 (WASH-1400), October 1975.
NRC, 1980. Control of Heavy Loads Nuclear Power Plants, NUREG-0612, U.S. Nuclear ulatory Commission, July 1980.
NRC, 1982. PAVAN: An Atmospheric-Dispersion Program for Evaluating Design-Basis idental Releases of Radioactive Materials from Nuclear Power Stations, NUREG/CR-2858,
. Nuclear Regulatory Commission, November 1982.
NRC, 1983. Atmospheric Dispersion Models for Potential Accident Consequence essments at Nuclear Power Plants Revision 1, Regulatory Guide 1.145, U.S. Nuclear ulatory Commission, February 1983.
NRC, 1992. Iodine Evolution and pH Control, NUREG/CR-5950, U.S. Nuclear Regulatory mmission, December 1992.
NRC, 1996. Guidelines for Preparing and Reviewing Applications for the Licensing of
-Power Reactors, Format and Content, NUREG-1537, Part 1, U.S. Nuclear Regulatory mmission, 1996.
NRC, 1998. Nuclear Fuel Cycle Facility Accident Analysis Handbook, NUREG/CR-6410,
. Nuclear Regulatory Commission, March 1998.
NRC, 2012a. Interim Staff Guidance Augmenting NUREG-1537, Part 1, "Guidelines for paring and Reviewing Applications for the Licensing of Non-Power Reactors: Format and tent," for Licensing Radioisotope Production Facilities and Aqueous Homogeneous ctors, Interim Staff Guidance Augmenting NUREG-1537, Part 1, U.S. Nuclear Regulatory mmission, 2012.
NRC, 2012b. Interim Staff Guidance Augmenting NUREG-1537, Part 2, "Guidelines for paring and Reviewing Applications for the Licensing of Non-Power Reactors: Standard iew Plan and Acceptance Criteria," for Licensing Radioisotope Production Facilities and eous Homogeneous Reactors, Interim Staff Guidance Augmenting NUREG-1537, Part 2,
. Nuclear Regulatory Commission, 2012.
NL, 2011. MCNP5-1.60 Release & Verification, LA-UR-11-00230, F.B. Brown, B.C.
drowski, J.S. Bull, M.A. Gonzales, N.A. Gibson, Los Alamos National Laboratory, Los mos, NM, 2011.
NE Medical Technologies 13a4-1 Rev. 0
oratory, Oak Ridge, Tennessee, 2011.
NE Medical Technologies 13a4-2 Rev. 0
.1 RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSIS METHODOLOGY accident analysis process for the radioisotope production facility (RPF) was conducted using same methodology as the accident analysis in the irradiation facility (IF), described in tion 13a2.1. The radiological consequences were evaluated using the same methodology cribed in Section 13a2.2 for the IF.
.1.1 PROCESSES CONDUCTED OUTSIDE THE IRRADIATION FACILITY production of molybdenum-99 (Mo-99) and other fission products occurs in the IF. After the diation of the target solution is completed, the solution is transferred from the IF to the RPF processed for radioisotope extraction and purification. Other processes occurring within the F include target solution processes for reuse, waste handling, and product packaging. These cesses that occur within the RPF are evaluated via an Integrated Safety Analysis (ISA) cess, involving hazard identification and a process hazard analysis (PHA). The hazard tification process includes a review of potential radiological hazards, chemical hazards, and er facility hazards that might be present.
cess that are conducted in the RPF and are analyzed in the ISA fall into the following gories:
- Operations with special nuclear material (SNM)
- Irradiated target solution processed for radioisotope extraction
- Irradiated target solution processed for reuse or for waste disposal
- Operations with unirradiated SNM
- Radiochemical operations
- Operations with hazardous chemicals operations involving SNM include the uranium receipt and storage system (URSS), target tion preparation system (TSPS), the molybdenum extraction and purification system PS), the iodine and xenon purification and packaging (IXP) system, the quality control and lytical testing laboratories (LABS), the target solution staging system (TSSS), the vacuum sfer system (VTS), the radioactive liquid waste storage (RLWS) system, the radioactive liquid te immobilization (RLWI) system, and the radioactive drain system (RDS). The operations do not involve SNM but pose a radiological or chemical hazard from radiochemical rations and operations with hazardous chemicals include the molybdenum isotope packaging tem (MIPS), the process vessel vent system (PVVS), and the facility chemical reagent system RS). Other systems in the RPF that do not have direct radiological or chemical hazards are luated in the ISA for impact on the systems listed above.
URSS receives, thermally oxidizes (if needed), repackages, and stores low-enriched nium prior to target solution preparation in the TSPS. The URSS is classified both as an ration with unirradiated SNM and as an operation with hazardous chemicals. Because of the sence of uranium, the URSS poses a criticality, radiological, and chemical hazard.
TSPS prepares low-enriched uranyl sulfate solution, which, once qualified for use, is rred to as target solution. The TSPS is classified both as an operation with unirradiated SNM NE Medical Technologies 13b.1-1 Rev. 1
MEPS separates the molybdenum from the irradiated target solution and purifies the ulting product. The extraction portion of the MEPS is an operation involving irradiated target tion processed for radioisotope extraction and contains significant quantities of uranium.
ause of the presence of uranium, the MEPS extraction process is analyzed for criticality ards. In addition, the extraction process involves radiological and chemical exposure ards. The purification portion of MEPS, as well as isotope packaging operations in MIPS, are sidered radiochemical operations, but pose a lesser hazard than extraction operations, ause these processes are physically separated from the extraction operations and involve ller quantities of radioactive material.
IXP system separates iodine from acidic solutions and purifies the resulting product. The aration operations handle irradiated target solution processed for radioisotope extraction and tain significant quantities of uranium. Because of the presence of uranium, the IXP process the potential for a criticality. In addition, the IXP has radiological and chemical exposure ards.
LABS are used to analyze samples of target solution, radioisotope products, and other cess fluids. The operations in the LABS involve small amounts of SNM, radiochemicals, and ardous chemicals. Because of the presence of uranium, the LABS are analyzed for criticality ards. In addition, the LABS involve radiological and chemical exposure hazards.
TSSS receives both irradiated target solution from the radioisotope extraction processes and radiated target solution from the TSPS. The TSSS allows for the target solution to be pled prior to reuse or disposal, and stages target solution for transfer to the IF or the waste tem. The system is categorized as irradiated target solution processed for reuse or waste osal. Because of the presence of uranium, the TSSS has the potential for a criticality as well adiological and chemical exposure hazards.
VTS serves as the transfer system for irradiated target solution between RPF tanks and for sfers between the RPF and the IF. The system also provides the capability to sample tank tents in the TSSS and the RLWS. The system performs operations involving irradiated target tion processed for reuse or waste disposal. Because of the presence of uranium, the VTS the potential for a criticality as well as radiological and chemical exposure hazards.
RLWS serves as a waste system for solutions resulting from the processing of licensed erial, and target solution batches or portions thereof that will no longer be used in facility cesses. The RLWS involves operations with irradiated target solution processed for waste osal. Because of the presence of uranium, the RLWS has the potential for a criticality. In ition, this process has radiological and chemical exposure hazards.
RLWI serves as a waste immobilization system for solutions received from the RLWS. The WI involves operations with irradiated target solution processed for waste disposal. Because he presence of uranium, the RLWI has the potential for a criticality. In addition, this process radiological and chemical exposure hazards.
RDS collects leakage and overflow of process fluids, including target solution, from process s and vessels and from hot cells. Fluids collected in the RDS can be returned to production NE Medical Technologies 13b.1-2 Rev. 1
the potential for a criticality as well as radiological and chemical exposure hazards.
PVVS handles the off-gas resulting from the processes of the IF and the RPF. The PVVS is sified as a radiochemical operation and poses a radiological hazard. This process contains onuclides removed from the off-gas.
FCRS stores and supplies reagents to the processes of the RPF. The FCRS is classified as operation with hazardous chemicals and poses a chemical hazard. The system contains no M or radionuclides.
.1.2 ACCIDENT INITIATING EVENTS design basis accidents (DBAs) identified in this section are initiating events (IEs) followed by dible accident scenarios that range from anticipated events, such as a loss of electrical power, vents that, while still credible, are considered unlikely to occur during the lifetime of the lity. The maximum hypothetical accident (MHA) is an accident scenario defined to result in worst-case (bounding) radiological consequences for the facility. Although the MHA is an ident scenario, it does not define a credible initiating event or accident progression, except for t is necessary to evaluate the consequences. Its purpose is to provide the most limiting sequence for the facility that bounds all credible DBAs.
As were identified using the following sources of information:
- IEs and accidents identified in the Interim Staff Guidance Augmenting NUREG-1537 (USNRC, 2012)
- Hazard and operability (HAZOP) studies, failure modes and effects analyses (FMEA),
and the PHA methods used as part of the ISA process
- Experience of the hazard analysis team DBA identification process resulted in a series of accident sequences that were then gorized into the following accident types:
- External Events
- Critical Equipment Malfunction (i.e., Malfunction or Mishandling of Equipment)
- Inadvertent Nuclear Criticality in the RPF
- RPF Fire
- Hazardous Chemical Accidents effects of a loss of off-site power (LOOP) and operator errors were considered as initiating nts within the scope of the PHA and were not classified as separate accident types.
litative evaluations are performed on the DBAs to further identify the bounding or limiting idents and scenarios, including the partial loss of systems or functions that could result in the est potential consequences. These evaluations are based on a review of identification of ses, the initial conditions, and assumptions for each accident.
ng the range of accident scenarios identified, each scenario was qualitatively evaluated for its ential chemical or radiological consequences. Scenarios that presented potential NE Medical Technologies 13b.1-3 Rev. 1
uce the overall risk of the evaluated scenarios to within acceptable limits. For accident uences that are not prevented and have mitigative controls applied, the radiological or mical consequences were quantitatively evaluated to demonstrate the effectiveness of the cted controls. The radiological consequences of accidents that were selected for additional luation are further evaluated in Section 13b.2. The accident analysis for chemical exposures rovided in Section 13b.3.
.1.2.1 Maximum Hypothetical Accident in the RPF MHA in the RPF is a fire in a carbon guard bed with degraded performance of the nstream carbon delay beds.
initiating event for this accident is ignition of transient combustibles or exothermic chemical ction in the bed resulting the formation of a hot spot and eventually a fire. Temperature cation normally detects an increase in temperature and the affected carbon guard bed is ated. The carbon guard bed releases its inventory to the downstream carbon delay beds ch are normally credited with adsorbing 99 percent of the released iodine. For the MHA, the bon delay beds are assumed to be operating at a reduced efficiency of 95 percent. Analysis onstrates that the carbon guard bed fire will not propagate to the carbon delay beds. This nario is described further in Subsection 13b.2.1.
.1.2.2 External Events external initiating events for the RPF that were evaluated include seismic events, tornados igh winds, small aircraft impacts, flooding, fires, and chemical releases. The SHINE main duction facility is designed to withstand credible external events, as described in section 13a2.1.6. External events were considered as potential IEs for a number of accident narios that fall within the other accident categories. The design basis seismic event results in ential chemical consequences, as described below and in Section 13b.3.
esign basis flooding event could result in potential flooding of internal vaults, trenches, and
, as well as the URSS and TSPS rooms. Flooding of the areas that contain fissile material uces the margin to criticality and challenges the double-contingency principle. Water intrusion these areas is minimized by sealed covers for the below-grade locations and by elevated m floors for the URSS and TSPS rooms. The local maximum probable precipitation event ulting in a 100-year flood will not exceed the first-floor entrance elevations, providing itional margin.
ernal event scenarios are further described in Subsection 13b.2.3.
.1.2.3 RPF Critical Equipment Malfunction ical equipment malfunctions in the RPF were evaluated as part of the accident analysis.
tiple scenarios were identified as having potential radiological consequences and were cted for additional evaluation. The identified scenarios are described below. For each nario, the controls that act to reduce the likelihood or consequences of the accident are listed.
scenarios that require mitigative controls, the radiological consequence assessments for ting exposures are presented in Subsection 13b.2.4.
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pill of target solution in the supercell has the potential to release radioactive gases, aerosol, particulates into the hot cell and ventilation system. Potential consequences of spilled target tion in the supercell include radiological dose. To mitigate the impact of spilled target tion, the following controls are applied: the supercell is designed as a confinement boundary, cell exhaust ventilation (radiological ventilation zone 1 [RVZ1]) is equipped with radiation nitors that provide a signal to the engineered safety features actuation system (ESFAS) to ate the affected cell and limit the amount of target solution introduced into the cell, hot cell et (RVZ1) ducts are equipped with carbon filters, hot cell inlet (radiological ventilation zone 2 Z2]) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant ation dampers, and ESFAS-controlled MEPS extraction pump breakers, VTS vacuum sfer pump breakers, and VTS vacuum break valves are provided to limit the amount of target tion introduced into the affected hot cell. This scenario is further described in section 13b.2.4.1.
nario 2 - Spill of Target Solution in the Supercell (MEPS Overpressurization) pill of target solution in the supercell caused by MEPS overpressurization has the potential to ase radioactive gases, aerosol, and particulates into the hot cell and ventilation system.
ential consequences of spilled target solution in the supercell include radiological dose. To vent deflagrations, which may cause overpressure events, the nitrogen purge system (N2PS) omatically actuates on a failure of PVVS and is relied on to dilute hydrogen concentrations in s and vessels in the RPF. Additionally, target solution extraction pumps are provided ssure relief mechanisms. To mitigate the impact of spilled target solution, the following trols are applied: the supercell is designed as a confinement boundary, hot cell exhaust tilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate affected cell and limit the amount of target solution introduced into the cell, hot cell outlet Z1) ducts are equipped with carbon filters, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ts are equipped with ESFAS-controlled redundant isolation dampers, and ESFAS-controlled.
PS extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum break es are provided to limit the amount of target solution introduced into the affected hot cell. This nario is further described in Subsection 13b.2.4.1.
nario 3 - Spill of Molybdenum Eluate Solution in the Supercell (Overfill or Drop of Rotovap k) pill of the molybdenum solution in the MEPS purification cell may result in the release of oactive gases, aerosol, and particulates into the hot cell and ventilation system. Potential sequences of spilled eluate solution in a hot cell include radiological dose. To mitigate the act of spilled eluate solution, the following controls are applied: the supercell is designed as a finement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors provide a signal to ESFAS to isolate the affected cell, hot cell outlet (RVZ1) ducts are ipped with carbon filters, and hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are ipped with ESFAS-controlled redundant isolation dampers. The resulting sequence of events his scenario is analogous to the MEPS eluate spill described in Subsection 13b.2.4.2.
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pill of target solution in the IXP extraction cell caused by IXP column misalignment has the ential to release radioactive gases, aerosol, and particulates into the hot cell and ventilation tem. Potential consequences of spilled target solution in supercell include radiological dose.
mitigate the impact of spilled target solution, the following controls are applied: the supercell is igned as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with ation monitors that provide a signal to ESFAS to isolate the affected cell and limit the amount arget solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped with carbon rs, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-trolled redundant isolation dampers, and ESFAS-controlled IXP extraction pump breakers, vacuum transfer pump breakers, and VTS vacuum break valves are provided to limit the ount of target solution introduced into the affected hot cell. This scenario is further described ubsection 13b.2.4.1.
nario 5 - Spill of Target Solution in the Supercell (IXP Overpressurization) pill of target solution in the IXP extraction cell caused by IXP column overpressurization has potential to release radioactive gases, aerosol, and particulates into the hot cell and tilation system. Potential consequences of spilled target solution in the supercell include ological dose. To prevent hydrogen deflagrations, which may cause overpressure events, the S automatically actuations on a failure of PVVS and is relied on to dilute hydrogen centrations in tanks and vessels in the RPF. Additionally, target solution extraction pumps are vided pressure relief mechanisms. To mitigate the impact of spilled target solution, the wing controls are applied: the supercell is designed as a confinement boundary, hot cell aust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to ate the affected cell and limit the amount of target solution introduced into the cell, hot cell et (RVZ1) ducts are equipped with carbon filters, hot cell inlet (RVZ2) and outlet (RVZ1) tilation ducts are equipped with ESFAS-controlled redundant isolation dampers, and ESFAS-trolled IXP extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum ak valves are provided to limit the amount of target solution introduced into the affected hot
. This scenario is further described in Subsection 13b.2.4.1.
nario 6 - Spill of Target Solution in the Supercell (Liquid Nitrogen Leak in IXP Hot Cell) quid nitrogen leak in the IXP hot cell may damage components in the supercell and result in a of target solution in the hot cell, with the potential to release radioactive gases, aerosol, and iculates into the supercell and ventilation system. Potential consequences of spilled target tion in the supercell include radiological dose. To mitigate the impact of spilled target tion, the following controls are applied: the supercell is designed as a confinement boundary, cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to FAS to isolate the affected cell and limit the amount of target solution introduced into the cell, cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled undant isolation dampers, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and FAS-controlled IXP extraction pump breakers, VTS vacuum transfer pump breakers, and VTS uum break valves are provided to limit the amount of target solution introduced into the cted hot cell. This scenario is further described in Subsection 13b.2.4.1.
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pill of iodine eluate solution in the IXP cell results in the release of radioactive gases, osols, and particulates into the hot cell and ventilation system. Potential consequences of ne solution spilling inside the IXP cell include radiological dose. To mitigate the impact of ed iodine solution, the following controls are applied: the supercell is designed as a finement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors provide a signal to ESFAS to isolate the affected cell, hot cell outlet (RVZ1) ducts are ipped with carbon filters, and hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are ipped with ESFAS-controlled redundant isolation dampers. The resulting sequence of events his scenario is analogous to the MEPS eluate spill described in Subsection 13b.2.4.2.
nario 8 - Spill of Target Solution in the Pipe Trench from a Single Pipe pill of target solution in the pipe trench results in the release of radioactive gases, aerosols, particulates into the pipe trench. Potential consequences of spilled target solution inside the trench include radiological dose. To mitigate the impact of spilled target solution, the wing controls are applied: the pipe trench is designed as a confinement boundary, RDS ns prevent the accumulation of target solution in the pipe trench, the RDS sump tank liquid ection sensor detects fluid in-leakage and provides a signal to ESFAS to stop any in-process sfers of solution within the facility via opening ESFAS-controlled VTS vacuum transfer pump akers and VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are ipped with radiation monitors that provide a signal to ESFAS to isolate the building ventilation ply and exhaust dampers on high radiation. This scenario is further described in section 13b.2.4.3.
nario 9 - Spill of Target Solution in the Pipe Trench from Multiple Pipes pill of target solution in the pipe trench results in the release of radioactive gases, aerosols, particulates into the hot cell and ventilation system. Potential consequences of spilled target tion in the pipe trench include radiological dose. To prevent the failure of multiple target tion-carrying pipes, the pipes are seismically qualified. This scenario is further described in section 13b.2.4.3.
nario 10 - Spill of Target Solution in a Tank Vault (Hold Tank Leak or Rupture) pill of target solution in a tank vault results in a release of radioactive gases, aerosols, and iculates into the tank vault. Potential consequences of target solution spilling in the tank vault ude radiological dose. To mitigate the impact of spilled target solution, the following controls applied: the tank vault is designed as a confinement boundary, RDS drains prevent the umulation of target solution in the tank vault, the RDS sump tank liquid detection sensor ects fluid in-leakage and provides a signal to ESFAS to stop any in-process transfers of tion within the facility via opening ESFAS-controlled VTS vacuum transfer pump breakers VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are equipped with ation monitors that provide a signal to ESFAS to isolate the building ventilation supply and aust dampers on high radiation. This scenario is further described in Subsection 13b.2.4.4.
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pill of target solution in a tank vault caused by a hold tank deflagration results a release of oactive gases, aerosols, and particulates into the tank vault. Potential consequences of et solution spilling in the tank vault include radiological dose. To prevent a deflagration in the tank, the N2PS automatically actuates on a failure of PVVS and is relied upon to dilute rogen concentrations. To mitigate the impact of spilled target solution, the following controls applied: the tank vault is designed as a confinement boundary, RDS drains prevent the umulation of target solution in the tank vault, the RDS sump tank liquid detection sensor ects fluid in-leakage and provides a signal to ESFAS to stop any in-process transfers of tion within the facility via opening ESFAS-controlled VTS vacuum transfer pump breakers VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are equipped with ation monitors that provide a signal to ESFAS to isolate the building ventilation supply and aust dampers on high radiation. This scenario is further described in Subsection 13b.2.4.4.
nario 12 - Spill of Target Solution in a Tank Vault (Seismic Event) pill of target solution in a tank vault caused by a seismic event results in a release of oactive gases, aerosols, and particulates into the tank vault. Potential consequences of et solution spilling in the tank vault include radiological dose. To prevent seismically caused age, the process tanks and piping are designed to withstand earthquakes. To mitigate the act of spilled target solution, the following controls are applied: the tank vault is designed as a finement boundary, RDS drains prevent the accumulation of target solution in the tank vault, RDS sump tank liquid detection sensor detects fluid in-leakage and provides a signal to FAS to stop any in-process transfers of solution within the facility via opening ESFAS-trolled VTS vacuum transfer pump breakers and VTS vacuum break valves, and the RVZ1 RVZ2 building exhausts are equipped with radiation monitors that provide a signal to ESFAS olate the building ventilation supply and exhaust dampers on high radiation. This scenario is her described in Subsection 13b.2.4.4.
nario 13 - Spill of Molybdenum Eluate in the Supercell (Deflagration) s of sweep gas flow from PVVS through the eluate tank in the supercell may result in a dup of hydrogen in the eluate tank and a subsequent deflagration. A spill of molybdenum te caused by a deflagration in the eluate tank results in the release radioactive gases, osols, and particulates into the hot cell. Potential consequences of spilled eluate solution in a cell include radiological dose. To prevent deflagrations in tanks and vessels in the RPF, the S automatically actuates upon a loss of PVVS and is relied upon to dilute hydrogen centrations. To mitigate the impact of spilled eluate solution, the following controls are lied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) quipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell, hot outlet (RVZ1) ducts are equipped with carbon filters, and hot cell inlet (RVZ2) and outlet Z1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers. This nario is further described in Subsection 13b.2.4.2.
nario 14 - Target Solution Leaking out of the Supercell (MEPS [ ]PROP/ECI k) ak in the MEPS extraction column [ ]PROP/ECI allows target solution to enter the
]PROP/ECI . Potential consequences of target solution leaking into the [
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rumentation in the [ ]PROP/ECI detects target solution in-leakage and provides a al to ESFAS to close the isolation valves on the supply and return of the [
]PROP/ECI at the supercell boundary. This scenario was evaluated qualitatively and is not cribed in Section 13b.2 because the accident sequence is prevented.
nario 15 - Extraction Column Three-Way Valve Misalignment ontroller or operator error resulting in a misaligned three-way valve causes target solution to towards the chemical skid, which is located outside of the supercell. Potential consequences is event include radiological dose. To prevent target solution from entering the chemical skid, eck valve in the chemical wash line prevents target solution backflow, and ESFAS nitoring of extraction three-way valve position causes the valves to de-energize and reposition never they are incorrectly aligned. This scenario was evaluated qualitatively and is not cribed in Section 13b.2 because the accident sequence is prevented.
nario 16 - Spill of Target Solution in a Valve Pit (Pipe Rupture or Leak) pill of target solution in a valve pit caused by a pipe rupture or leak results in a release of oactive gases, aerosols, and particulates into the valve pit. Potential consequences of spilled et solution in the valve pit include radiological dose. To mitigate the consequences of spilled et solution, the following controls are applied: the valve pit is designed as a confinement ndary, RDS drains prevent the accumulation of target solution in the valve pit, the RDS sump liquid detection sensor detects fluid in-leakage and provides a signal to ESFAS to stop any rocess transfers of solution within the facility via the opening ESFAS-controlled VTS vacuum sfer pump breakers and VTS vacuum break valves, and the RVZ1 and RVZ2 building austs are equipped with radiation monitors that provide a signal to ESFAS to isolate the ding ventilation supply and exhaust dampers on high radiation. The resulting sequence of nts for this scenario is analogous to the target solution leak in a pipe trench, which is further cribed in Subsection 13b.2.4.3.
nario 17 - Spill of Radioactive Liquid Waste in the RLWI Shielded Enclosure pe leak or rupture in the RLWI shielded enclosure results in a release of radioactive gases, osols, and particulates into the enclosure. Potential consequences of a pipe leak or rupture in RLWI shielded enclosure include radiological dose. To prevent unacceptable doses to kers, RLWS operating procedures provide limitations on concentration of waste solutions and uire a minimum holdup time in the blending tank prior to transfer to the RLWI enclosure. This nario is further described in Subsection 13b.2.4.5.
nario 18 - Heavy Load Drop onto RLWI Shielded Enclosure or Supercell ane failure or operator error resulting in a heavy load drop on the RLWI shielded enclosure or supercell causes damage to the affected structure and internal equipment. Potential sequences of a heavy load drop on the RLWI shielded enclosure or supercell include ological dose. To prevent a heavy load drop on the enclosure or the supercell, crane ration procedures include safe load paths to avoid the RLWI enclosure and supercell, and uire suspension of supercell and RLWI activities during a heavy lift. The supercell damage nario was evaluated qualitatively and is not described in Section 13b.2 because the accident NE Medical Technologies 13b.1-9 Rev. 1
nario 19 - Heavy Load Drop onto a Tank Vault or Pipe Trench Cover Block ane failure or operator error resulting in a heavy load drop on a tank vault or pipe trench er block causes a damage to the cover block and internal equipment. Potential sequences of a heavy load drop include radiological dose. To prevent damage to a cover k, the cover blocks have been designed to withstand a heavy load drop. This scenario was luated qualitatively and is not described in Section 13b.2 because the accident sequence is vented.
.1.2.4 RPF Inadvertent Nuclear Criticality lear criticality safety (NCS) in the RPF is accomplished through the use of criticality safety trols to prevent criticality during normal and abnormal conditions. Each process that involves use, handling, or storage of SNM is evaluated by the SHINE nuclear criticality safety staff er the requirements of the NCS program. The results of the criticality safety evaluations are rporated into the ISA. Radiological consequences of criticality accidents are not included in accident analysis because preventative controls are used to ensure criticality events are ly unlikely. Further discussion of the criticality safety bases for RPF processes is included in tion 6b.3.
.1.2.5 RPF Fire RPF was evaluated for internal fire risks based on the fire hazards analysis (FHA). The FHA uments the facility fire areas and each area was individually evaluated for fire risks. Internal lity fires are generally evaluated as an initiating event for the release of radioactive material are included in the scenarios evaluated in Section 13a2.1 and this section. Two unique narios are described below and evaluated in detail in Section 13b.2.
main production facility maintains a facility fire protection plan to reduce the risks of fires, as cribed in Section 9a2.3.
nario 1 - PVVS Carbon Delay Bed Fire upset or malfunction in the PVVS (high moisture or high temperature) results in ignition of the bon media in a delay bed. A fire in the carbon delay bed results in a release of the captured oactive material into the PVVS downstream of the delay bed and to the environment via the lity exhaust stack. A release to the environment results in radiological exposure to the public.
ease of radioactive material in excess of acceptable levels is prevented by the carbon delay carbon monoxide (CO) detectors, which provide a signal to ESFAS to close the PVVS bon delay bed isolation valves for the affected carbon delay bed group and bypass the cted group in the event of high CO concentration indicative of a fire in a bed. Releases to the F are further mitigated by the process confinement boundary (carbon delay bed vaults). This nario is further described in Subsection 13b.2.6.1.
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upset or malfunction in the PVVS (high moisture or high temperature) results in ignition of the bon media in a guard bed. A fire in the guard bed results in a release of the captured oactive material into the PVVS downstream of the guard bed, into the delay beds, and to the ironment via the facility exhaust stack. A release to the environment results in radiological osure to the public. Release of radioactive material in excess of acceptable levels is vented by the downstream carbon delay beds, which reduce or delay radioisotope release.
eases to the RPF are further mitigated by the supercell confinement boundary. This scenario rther described in Subsection 13b.2.6.2.
.1.2.6 RPF Chemical Accidents ential chemical exposures in the RPF were evaluated to identify chemical hazards and essary controls. The bounding inventories of chemicals used in the main production facility e identified and evaluated for exposure to workers and the public. Only exposure to uranium e presents a risk that exceeds the applicable evaluation criteria. This scenario is discussed her in Section 13b.3.
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eral design basis accidents described in Section 13b.1 result in a release of radioactive erials into or outside the controlled areas of the facility.
analyses in this section evaluate the applicable radiological consequences of these idents to demonstrate than an individual located in the unrestricted area following the onset of ostulated accidental release of licensed material would not receive a radiation dose in excess 00 mrem total effective dose equivalent (TEDE) for the duration of the accident.
iological consequences to workers are also evaluated and are shown to not exceed 5 rem DE during the accident.
.2.1 MAXIMUM HYPOTHETICAL ACCIDENT IN THE RPF maximum hypothetical accident (MHA) in the radioisotope production facility (RPF) is a fire carbon guard bed with degraded carbon delay bed efficiency. It is postulated that percent of the radionuclide inventory is released from the guard bed and flows downstream the carbon delay beds and is then released to the facility stack. The performance of the bon delay beds is assumed to be degraded to 95 percent. The automatically mitigated release a credible carbon guard bed fire is discussed in Subsection 13b.2.6.2.
al Conditions process vessel vent system (PVVS) is operating normally, with nominal flow through one bon guard bed.
affected carbon guard bed contains radioactive iodine from RPF process streams. The erial-at-risk (MAR) in this scenario is a combination of gases from eight irradiation units (IU),
various modifiers applied to account for decay and operational sequencing.
ating Event upset or malfunction in the PVVS results in high moisture or high temperature flow through carbon guard bed. The high moisture or high temperature results in ignition of the carbon rd bed absorber media. Potential initiating events for this scenario are discussed further in section 13b.1.2.1.
uence of Events
- 1. Ignition of one of the carbon guard beds occurs, resulting in an exothermic release of 100 percent of the stored radioactive material to the gas piping downstream of the guard bed.
- 2. The downstream carbon delay beds adsorb 95 percent of the radioactive material.
- 3. The radioactive material exiting the carbon delay beds is released to the environment through the PVVS and facility stack.
- 4. The maximum gas temperature entering the carbon delay beds is 130°C, which is insufficient to propagate the fire condition to the carbon delay bed. Temperature instrumentation in the carbon guard bed detects increased temperature and the carbon NE Medical Technologies 13b.2-1 Rev. 1
mage to Equipment occurrence of fire damages the affected carbon guard bed and eliminates its ability to ction. No other damage to the PVVS system or its components occurs.
nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.
leak path factor (LPF) model terms used in this accident are provided in Table 13b.2-1. For accident, the release of material from the guard bed is assumed to be instantaneous and is sported to the environment at an increased rate.
iation Source Terms initial MAR for this scenario is a portion of the iodine gas inventory evolved from target tion during normal operations. Development of the accident source term for this scenario is ussed further in Section 13a2.2.
iodine gas inventory is produced by fission and decay of fission products and continuously lved from the target solution and through the target solution vessel (TSV) off-gas system GS) during operations. Partitioning fractions for iodine gas are used to describe the ntities of iodine in solution that move to the RPF. Removal of iodine by the TOGS zeolite s are credited for all gases that are transported to the RPF. The MAR uses selected time rvals for the most recent purges (i.e., [ ]PROP/ECI) to ount for the operational sequencing of the combined eight IUs. The MAR assumes the bined iodine gas inventory produced by eight IUs over approximately [ ]PROP/ECI of diation with the most recent purges of [
]PROP/ECI. The iodine umulates in the carbon guard bed and decays.
iological Consequences radioactive material is contained in the PVVS system and does not result in worker osure. The radiological consequences of this accident scenario are determined as described ection 13a2.2. The results of the determination are provided in Table 13b.2-2.
.2.2 LOSS OF ELECTRICAL POWER s of off-site power (LOOP) was evaluated in the accident analysis as an initiating event for a ber of critical equipment malfunction scenarios. A facility-wide LOOP results in automatic ation of multiple facility engineered safety features, which act to ensure the risk associated radiological or chemical releases is reduced to within acceptable limits. The facility-wide OP does not result in system or component failures within the RPF that result in unacceptable ological or chemical consequences. The facility-wide LOOP is further discussed in section 13a2.1.5 and Subsection 13a2.2.5.
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eismic event was identified as an initiating event for several critical equipment malfunction idents. The accident analysis associated with these events is presented below.
ere weather was evaluated as an initiating event for the accident analysis. The main duction facility structure is designed to withstand credible severe weather conditions and vent damage to facility internal structures, systems, and components (SSCs). Based on the ign of the main production facility structure, the risk associated with the potential release of ological material or chemicals due to severe weather events is reduced to within acceptable s.
oding was evaluated as an initiating event for the accident analysis. The main production lity has internal flood control measures to prevent the intrusion of water into areas that would affected by the intrusion of water. The internal flood control measures are discussed below.
itionally, the local probable maximum precipitation event for the main production facility does exceed the first floor entrance elevations. Consequently, there are no radiological sequences associated with external flooding events.
ernal fires were evaluated as an initiating event for the accident analysis. Generally, external s do not have radiological consequences for the facility because safe shutdown conditions be achieved without operator actions. The likelihood of significant external fires is highly kely because the facility is located on open terrain with no nearby prairie or forest and there no natural gas lines that interact with the main production facility. The nearest natural gas line inates approximately 60 feet from the main production facility. Vehicle fires were also sidered. A vehicle fire at the loading dock presents a limited risk. The loading dock is igned to prevent combustible liquid spills from entering into the building, and the shipping/
eiving area is separated from the loading dock itself. An external fire from a vehicle in the ing dock would be locally contained and does not produce radiological consequences.
.2.4 RPF CRITICAL EQUIPMENT MALFUNCTION eral accident scenarios involve a release of radioactive solution into the supercell. Two types olutions are present in the supercell, irradiated target solution and product eluate solutions.
ls of these solution are analyzed to determine their radiological consequences.
rator errors were evaluated in the accident analysis as an initiating event for a number of cal equipment malfunctions. Operator errors and their effects are discussed in section 13b.1.2.
.2.4.1 Spill of Target Solution in the Supercell al Conditions he time of the initiating event, target solution is being pumped through the molybdenum action and purification system (MEPS) extraction cell. The target solution has decayed for
]PROP/ECI in the TSV dump tank prior to beginning the extraction process. The target tion irradiation assumptions are described in Section 13a2.2.
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event causes a break in the MEPS piping between the extraction pump discharge and the action column. The break downstream of the pump discharge causes spray and osolization of the target solution without any extraction of isotopes by the extraction column.
ential initiating events for this scenario and analogous scenarios for the iodine and xenon fication (IXP) system cell are discussed further in Subsection 13b.1.2.3; Scenarios 1, 2, 4, 5, 6.
uence of Events
- 1. A break in the MEPS piping between the extraction pump discharge and the extraction column occurs.
- 2. Aerosolized target solution sprays from the break into the hot cell, releasing radioactive material into the hot cell and causing the cell to become pressurized to the nominal pressure of the cell drain loop seal.
- 3. Radiation detectors in the hot cell exhaust ventilation detect high airborne radiation and cause the engineered safety features actuation system (ESFAS) to shut down the vacuum transfer system (VTS), shut down the extraction pump, and isolate the hot cell ventilation.
- 4. Leakage of radioactive material from the hot cell to the RPF and the environment through the ventilation dampers occurs, resulting in radiological consequences to workers and the public.
maximum volume of spilled target solution in this accident scenario is limited by the volume he vacuum lift tanks and installed piping of the MEPS. The ESFAS shutdown of the VTS vents additional target solution from entering the hot cell after high radiation has been ected. The analyzed volume of target solution for this scenario is 30 liters, which is servatively based on the volume of two vacuum lift tanks plus additional pipe volume.
controls credited for mitigation of the dose consequences for this accident are:
- Supercell confinement boundary
- Hot cell radiological ventilation zone 1 (RVZ1) radiation monitors
- Hot cell RVZ1 outlet carbon filters (radioiodine)
- Inlet (radiological ventilation zone 2 [RVZ2]) and outlet (RVZ1) ventilation isolation dampers
- MEPS or IXP extraction pump breakers
- VTS vacuum transfer pump breakers
- VTS vacuum break valves
- ESFAS Supercell Isolation function
- ESFAS VTS Safety Actuation function mage to Equipment leak of target solution in the supercell does not cause subsequent damage to equipment.
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methods used to calculate radioactive material transport are described in Section 13a2.2.
LPF model terms used in this accident are provided in Table 13b.2-1.
iation Source Terms initial MAR for this scenario is 30 liters of target solution from the IU at [ ]PROP/ECI t-shutdown. The action of the TOGS during this [ ]PROP/ECI period removes more n 67 percent of the iodine present in the solution at shutdown. It is conservatively assumed 35 percent of the post-shutdown iodine inventory is released to the supercell during the ident. Additionally, partitioning fractions are applied to the noble gases present in target tion. Development of the accident source term for this scenario is discussed further in tion 13a2.2.
iological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are shown in Table 13b.2-2.
.2.4.2 Spill of Eluate Solution in the Supercell al Conditions he time of the initiating event, eluate solution in the MEPS eluate tank is spilled onto the floor he hot cell, releasing radioactive material into the hot cell atmosphere.
ating Event event causes the failure of the MEPS eluate tank, which results in a spill of eluate solution.
ential initiating events for this scenario and analogous scenarios for the purification and IXP s are discussed further in Subsection 13b.1.2.3; Scenarios 3, 7, and 13.
uence of Events
- 1. A break in the MEPS eluate tank occurs.
- 2. Eluate solution spills from the tank into the hot cell, releasing radioactive material into the hot cell and causing the cell to become pressurized to the nominal pressure of the cell drain loop seal.
- 3. Radiation detectors in the hot cell exhaust ventilation detect high airborne radiation and cause ESFAS to isolate hot cell ventilation.
- 4. Leakage of radioactive material from the hot cell to the RPF and the environment through the ventilation dampers occurs, resulting in radiological consequences to workers and the public.
controls credited for mitigation of the dose consequences for this accident are:
- Supercell confinement boundary
- Hot cell RVZ1 radiation monitors
- Hot cell RVZ1 outlet carbon filters (radioiodine)
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mage to Equipment leak of target solution in the supercell does not cause subsequent damage to equipment.
nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.
LPF model terms used in this accident are provided in Table 13b.2-1.
iation Source Terms initial MAR for this scenario is the extraction column eluate, which contains radionuclides one entire target solution batch. The initial MAR is partitioned by the extraction column to duce the accident-specific MAR. Accident-specific partitioning factors are applied to the diated target solution batch as described in Section 13a2.2. Development of the accident rce term for this scenario is discussed further in Section 13a2.2.
iological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are provided in Table 13b.2-2.
.2.4.3 Spill of Target Solution in the RPF Pipe Trench al Conditions atch of irradiated target solution is being transferred within the RPF pipe trench. The target tion has been irradiated using the assumptions in Section 13a2.2 and has been held for ay in the TSV dump tank for [ ]PROP/ECI.
ating Event event causes a pipe containing target solution to break in the pipe trench. Multiple pipe res from a seismic event is considered to be highly unlikely because the pipes and their ports are seismically qualified. Therefore, the failure of multiple solution-containing pipes ld require the onset of a design basis earthquake concurrent with the failure of multiple sive, seismically-qualified components. Consequently, dose consequences for multiple pipe res are not evaluated. Potential initiating events for this scenario and the analogous scenario a spill in a valve pit are discussed further in Subsection 13b.1.2.3; Scenarios 8, 9, and 16.
uence of Events
- 1. A pipe containing target solution within the pipe trench breaks, spilling target solution into the trench.
- 2. The target solution collects on one of the three drip pans in the trench and drains to the radioactive drain system (RDS).
- 3. Radioactive material is released into the pipe trench atmosphere.
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to workers and the public.
controls credited for mitigation of the dose consequences for this accident are:
- Process confinement boundary (trench or pit cover and cover seal) itional controls described in Subsection 13b.1.2.3 are provided but not credited in the dose lysis.
mage to Equipment leak of target solution into the pipe trench does not cause further damage to equipment.
nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.
LPF model terms used in this accident are provided in Table 13b.2-1.
iation Source Terms initial MAR for this scenario is a batch of target solution from the IU at [ ]PROP/ECI PROP/ECI t-shutdown. The action of the TOGS during this [ ] period removes more n 67 percent of the iodine present in the solution at shutdown. It is conservatively assumed 35 percent of the post-shutdown iodine inventory is released to the pipe trench during the ident. Additionally, partitioning fractions are applied to the noble gases present in target tion. Development of the accident source term for this scenario is discussed further in tion 13a2.2.
iological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are provided in Table 13b.2-2.
.2.4.4 Spill of Target Solution from a Tank pill of target solution from any of the below-grade hold or storage tanks results in a release of et solution into the associated tank vault. Radionuclides from the target solution become orne and migrate into the RPF and the environment.
liquid waste blending tanks contain large volumes of dilute target solution that has already ergone extraction and processing. The accident analysis considers freshly-irradiated target tion that has not undergone processing and bounds the failure of the liquid waste blending k.
al Conditions ll batch of target solution is present in a target solution hold or storage tank at the time of the ating event. The target solution has been irradiated using the assumptions in Section 13a2.2 NE Medical Technologies 13b.2-7 Rev. 1
ating Event event causes a tank containing target solution to break and leak. Potential initiating events discussed further in Subsection 13b.1.2.3; Scenarios 10, 11, and 12.
uence of Events
- 1. A tank containing target solution breaks, spilling target solution into the tank vault.
- 2. The target solution collects on the drip pans in the vault and drains to the RDS.
- 3. Radioactive material is released into the pipe trench atmosphere
- 4. A portion of the released material leaks through the process confinement boundary (vault cover) into the RPF and the environment, resulting in radiological consequences to workers and the public.
controls credited for mitigation of the dose consequences for this accident are:
- Process confinement boundary (tank vault plugs and seals) itional controls described in Subsection 13b.1.2.3, including drainage of the solution out of vault via RDS, are provided but not credited in the dose analysis.
mage to Equipment leak of target solution into the tank vault does not cause further damage to equipment.
nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.
LPF model terms used in this accident are listed in Table 13b.2-1.
iation Source Terms initial MAR for this scenario is a full batch of target solution from the IU at
]PROP/ECI post-shutdown. The action of the TOGS during the [ ]PROP/ECI hold-period in the dump tank removes more than 67 percent of the iodine present in the solution at tdown. It is assumed that 35 percent of the post-shutdown iodine inventory is released to the vault during the accident. Additionally, partitioning fractions are applied to the noble gases sent in target solution. Development of the accident source term for this scenario is discussed her in Section 13a2.2.
iological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are shown in Table 13b.2-2.
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al Conditions 80-liter batch of waste solution (diluted target solution) is present in the radioactive liquid te immobilization (RLWI) system immobilization feed tank at the time of the initiating event.
volume of solution in this scenario is based on the volume of the immobilization feed tank a conservative scaling factor to account for the highest allowable concentration of onuclides. The waste solution has been irradiated using the assumptions in Section 13a2.2 has been held for decay for 35 days post-shutdown. The post-shutdown hold time is based he minimum hold time needed to reduce waste activity to within dose consequence limits and blishes an administrative control. Expected hold times for waste solution are significantly er than 35 days.
ating Event event causes the immobilization feed tank or RLWI system piping containing waste solution to ak and leak within the RLWI enclosure. Potential initiating events are discussed further in section 13b.1.2.3; Scenarios 17 and 18.
uence of Events
- 1. The immobilization feed tank breaks and spills waste solution into the RLWI enclosure.
- 2. The waste solution collects on the floor of the enclosure and leaks into the RPF and environment, resulting in radiological consequences to workers and the public.
controls credited for mitigation of the dose consequences for this accident are:
- Waste solution holdup times in the radioactive liquid waste storage (RLWS) system before processing in RLWI
- Concentration controls applied to waste solutions
- Heavy load drop controls described in Subsection 13b.1.2.3.
mage to Equipment leak of waste solution into the RLWI enclosure does not cause further damage to equipment.
nsport of Radioactive Material LPF and airborne release fraction (ARF) values used in this scenario are set at 1.0 instead sing the LPF model values described in Section 13a2.2. The LPF model terms used in this ident are provided in Table 13b.2-1.
iation Source Terms initial MAR for this scenario is 380 liters of waste solution at 35 days post-shutdown. The centration of radionuclides for the waste solution is determined by multiplication of the ratio of maximum uranium concentration permitted in the RLWI system to the nominal uranium centration of target solution. The action of the TOGS during the [ ]PROP/ECI period n the original target solution was held in the dump tank removes more than 67 percent of the NE Medical Technologies 13b.2-9 Rev. 1
tions are applied to the noble gases present in target solution. Development of the accident rce term for this scenario is discussed further in Section 13a2.2.
iological Consequences radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are shown in Table 13b.2-2.
.2.5 RPF INADVERTENT NUCLEAR CRITICALITY dvertent nuclear criticality events were evaluated in the accident analysis using the same hodology as non-criticality accidents. Nuclear criticality safety is achieved through the use of ventative controls throughout the RPF, which reduces the likelihood of a criticality accident to ly unlikely (or better). Preventative controls were selected based on nuclear criticality safety luations conducted under the facility nuclear criticality safety program. The nuclear criticality ty program and the criticality safety basis for RPF processes is described in Section 6b.3.
.2.6 RPF FIRE ility fires were evaluated in the accident analysis. Facility fire scenarios and their effects are ussed in Subsection 13b.1.2.5. Two facility fire scenarios were evaluated for radiological sequences.
.2.6.1 PVVS Carbon Delay Bed Fire al Conditions PVVS is operating normally, with nominal flow through a carbon delay bed.
affected carbon delay bed contains noble gases from RPF process streams. The MAR in scenario is a combination of gases from eight IUs with various modifiers applied to account decay and processing capacity of target solution batches in the supercell.
ating Event upset or malfunction in the PVVS results in high moisture or high temperature flow through carbon delay bed. The high moisture or high temperature results in ignition of the carbon y bed absorber media. Potential initiating events are discussed further in section 13b.1.2.5, Scenario 1.
uence of Events
- 1. Ignition of the carbon delay bed occurs, resulting in an exothermic release of stored radioactive material to the PVVS downstream of the delay bed.
- 2. Radioactive material is released to the environment through the PVVS and facility stack.
- 3. Incipient fire conditions are detected by the in-line carbon monoxide detectors, which send an actuation signal to the ESFAS.
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consequence determination.
- 5. Following valve closure, the gross release of radioactive material is stopped and the fire is extinguished. Leakage through the valve occurs at a diminished rate.
components credited for mitigation of the dose consequences for this accident are:
- PVVS carbon delay bed isolation valves
- ESFAS carbon delay bed isolation function mage to Equipment occurrence of fire damages the affected carbon delay bed and eliminates its ability to ction. No other damage to the PVVS system or its components occurs.
nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.
LPF model terms used in this accident are provided in Table 13b.2-1. For this accident, the ase of material for the first 30 seconds is assumed to be instantaneous and is transported to environment at an increased rate. Following isolation valve actuation, the transport occurs at duced rate.
iation Source Terms initial MAR for this scenario is a portion of the noble gas inventory evolved from target tion during normal operations. Development of the accident source term for this scenario is ussed further in Section 13a2.2.
noble gas inventory is produced by decay of fission products and continuously evolved from target solution and through the TOGS during operations. The MAR uses selected time rvals for the most recent purges (i.e., [ ]PROP/ECI) to ount for the processing capacity of target solution batches in the supercell for the combined t IU. The gases accumulate in the carbon delay bed and decay. The MAR assumes the bined noble gas inventory produced by eight IUs over approximately [ ]PROP/ECI of diation with the most recent purges of [
]PROP/ECI. Partitioning fractions noble gases are used to describe the quantities of noble gases in solution that move to the F to account for removal during movement of solution.
iological Consequences radioactive material is contained in the PVVS system and does not result in worker osure. The radiological consequences of this accident scenario are determined as described ection 13a2.2. The results of the determination are provided in Table 13b.2-2.
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al Conditions PVVS is operating normally, with nominal flow through a carbon guard bed.
affected carbon guard bed contains iodine from RPF process streams. The MAR in this nario is a combination of iodine from eight IUs with various modifiers applied to account for ay and processing capacity of target solution batches in the supercell.
ating Event upset or malfunction in the PVVS results in high moisture or high temperature flow through carbon guard bed. The high moisture or high temperature results in ignition of the carbon rd bed adsorber material. Potential initiating events are discussed further in tion 13b.1.2.5, Scenario 2.
uence of Events
- 1. Ignition of the carbon guard bed occurs, resulting in an exothermic release of stored radioactive material to the PVVS downstream of the guard bed.
- 2. Radioactive material is captured by the downstream carbon delay bed and filtered. One percent of the released radioactive material is released through PVVS and the facility stack to the environment.
component credited for mitigation of the dose consequences for this accident is:
- PVVS delay bed filtration mage to Equipment occurrence of fire damages the affected carbon guard bed and eliminates its ability to ction. No other damage to the PVVS system or its components occurs.
nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.
LPF model terms used in this accident are provided in Table 13b.2-1. For this accident, the rd bed inventory is assumed to be instantly transported to the delay bed. The delay bed is dited to reduce the release of material by 99 percent with no credit taken for carbon guard bed ation functions.
iation Source Terms initial MAR for this scenario is a portion of the iodine gas inventory evolved from target tion during normal operations. Development of the accident source term for this scenario is ussed further in Section 13a2.2.
iodine gas inventory is produced by fission and decay of fission products and continuously lved from the target solution and through the TOGS during operations. Partitioning fractions NE Medical Technologies 13b.2-12 Rev. 1
F. The MAR uses selected time intervals for the most recent purges (i.e., [
]PROP/ECI) to account for the operational sequencing of the combined eight IUs.
MAR assumes the combined iodine gas inventory produced by eight IUs over approximately
]PROP/ECI of irradiation with the most recent purges of [
]PROP/ECI. The iodine accumulates in the carbon guard bed and decays.
iological Consequences radioactive material is contained in the PVVS system and does not result in worker osure.
radiological consequences of this accident scenario are determined as described in tion 13a2.2. The results of the determination are provided in Table 13b.2-2.
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Table 13b.2 Radiation Transport Factors Radionuclide Group LPF Model cident Scenario (ARF x LPF) ximum Hypothetical Accident Iodine 1.0 8.12E-01 Public Nobles 1.80E-02 Worker 7.10E-03 Public ll of Target Solution in the Supercell Iodine 4.20E-03 Worker 3.46E-05 Public Non-Volatile 3.57E-06 Worker 8.12E-01 Public Nobles 1.80E-02 Worker 7.10E-03 Public ll of Eluate Solution in the Supercell Iodine 4.20E-03 Worker 3.46E-05 Public Non-Volatile 3.57E-06 Worker 1.73E-01 Public Nobles 3.85E-03 Worker ll of Target Solution in the RPF Pipe 1.20E-02 Public Iodine nch 3.77E-03 Worker 3.85E-06 Public Non-Volatile 9.22E-10 Worker 2.08E-01 Public Nobles 9.31E-03 Worker 1.60E-02 Public ll of Target Solution from a Tank Iodine 8.98E-03 Worker 5.11E-06 Public Non-Volatile 2.20E-09 Worker ll of Waste Solution in RLWI All 1.0 VS Carbon Guard Bed Fire Iodine 1.0 (Public Only)
VS Carbon Delay Bed Fire Nobles 4.80E-02 (Public Only)
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Table 13b.2 Radioisotope Production Facility Accident Dose Consequences Public Dose Worker Dose TEDE TEDE cident Scenario (mrem) (mrem) ximum Hypothetical Accident 403 No consequences ll of Target Solution in the Supercell 7 317 ll of Eluate Solution in the Supercell 14 401 ll of Target Solution in the RPF Pipe Trench 4 170 ll of Target Solution from a Tank 5 398 ll of Waste Solution in RLWI 13 861 VS Carbon Delay Bed Fire 39 No consequences VS Carbon Guard Bed Fire 81 No consequences NE Medical Technologies 13b.2-15 Rev. 1
potential hazards of the chemicals proposed to be used at the SHINE facility have been luated. The analysis has been performed for hazardous chemicals within the facility that ract with or are produced from licensed materials. The analysis also includes other toxic and ctive hazardous chemicals that are present in the SHINE facility but do not directly interact licensed materials. Therefore, the analysis is bounding for all hazardous chemicals that ract with or are produced from licensed materials. Safety-related or administrative controls e been developed only for those systems or processes where the hazardous chemical is duced from or otherwise associated with licensed materials. Consequence or chemical dose deling are evaluated using dispersion models and/or computer codes that conform to the hodologies described in NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis dbook (USNRC, 1998).
ntitative exposure standards are selected to meet acceptable limits for public and worker lth and safety. The quantitative acceptance limits are taken from the Protective Action Criteria C) values (USDOE, 2018), which correspond to the Acute Exposure Guideline Levels GLs), Emergency Response Planning Guidelines (ERPGs), or Temporary Emergency osure Limits (TEELs) values for such chemicals. Two exceptions are applied to rhodium ride and uranyl peroxide, which do not have published PAC values. For these chemicals, eptance values are assigned based on Occupational Safety and Health Administration HA) occupational exposure limits.
mical Accidents Description only chemical accident scenario with the potential to exceed established chemical exposure elines for workers is a seismic event resulting in the failure or overturning of the uranium eipt and storage system (URSS) uranium oxide storage rack, causing multiple storage can res. The uranium storage racks are seismically qualified to maintain their structure and ition during a seismic event, which prevents the potential chemical exposure.
failure of a single can during transfer or handling operations does not result in chemical dose sequences being exceeded.
mical Accident Consequences azardous chemical consequence assessment was performed to demonstrate that potential sequences are within acceptable limits. This assessment determines if the release of ardous chemicals from the SHINE facility could lead to exceeding the PAC values. The ntory of chemicals used at the SHINE facility is presented in Table 13b.3-1.
onsequence analysis for the public and nearest residence was performed using the ALOHA al Locations of Hazardous Atmospheres) computer code (USDOC, 2013). The material-at-(MAR) present for each chemical is identified in Table 13b.3-2. The MAR is assumed to be largest quantity of material that can be present for a single release event. In most cases, this qual to one-half of the chemical inventory listed in Table 13b.3-1, but limited by the capacity single storage container (i.e., a single container spill). One exception is the release of sulfur afluoride, in which the inventory from all eight neutron driver assembly system (NDAS) ssure vessels is assumed to be released as a result of a seismic event.
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amount of each chemical released into the facility atmosphere. The evaporation rate is ermined by setting the assumed wind speed to the minimum value allowed in ALOHA, which ulates the indoor air movement. The puddle area used for evaporation is modeled by using room dimensions that the chemical is stored in. The resulting concentration of a chemical ase within the facility is calculated as a homogenous mixture within the radioisotope duction facility (RPF) volume.
model the chemical exposure to the members of the public, the evaporation rates from the mical puddle are calculated in the same way as above for the worker dose. The evaporated mical is then dispersed using a 4.2 meters per second (m/s) mean wind speed, which is ed on meteorological data from the Southern Wisconsin Regional Airport. The resulting mical concentration for the site boundary and the nearest residence is determined. The mary of the chemical concentration calculation is provided in Table 13b.3-2. Note that two itional chemicals (hydrogen fluoride and sulfur dioxide) are included in Table 13b.3-2. These decomposition byproducts of the sulfur hexafluoride gas used in the NDAS and are not mally stored on-site.
chemical concentrations for workers and the nearest residence are below the PAC-1 values.
mical Process Controls components credited for prevention of the chemical dose consequences are:
- URSS uranium storage racks are seismically qualified to maintain their structure and position during seismic events.
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Table 13b.3 Inventory of Hazardous Chemicals (Sheet 1 of 2)
Location Inventory Storage Container Type and Capacity Chemical Name (kg) ha-Benzoin Oxime Chemical Storage 1 Two 250 gram (g) containers Liquid monia (28 wt.%) Chemical Storage 499 (Two 1000 liter [l] intermediate bulk containers [IBC])
]PROP/ECI Chemical Storage 10 Two 5 kilogram (kg) containers drochloric Acid (38 wt.%) Chemical Storage 4 Two one gallon containers Liquid drogen Peroxide (30 wt.%) Chemical Storage 668 (Two 1000 l IBC)
Liquid eral Oil Chemical Storage 360 (Two 55 gallon drums) lybdenum Trioxide Chemical Storage 1 Two one gallon containers ic Acid Liquid Chemical Storage 1979 wt.% in chemical storage) (Two 1000 l IBC) assium Chemical Storage 1 Two 250 g containers xachlororuthenate assium Permanganate Chemical Storage 1 Two 250 g containers odium Chloride Chemical Storage 1 Two 250 g containers er Nitrate Chemical Storage 1 Two 250 g containers dium Hydroxide (50.5 wt.%) Chemical Storage 1541 Two 1000 l IBC NE Medical Technologies 13b.3-3 Rev. 0
Location Inventory Storage Container Type and Capacity Chemical Name (kg) dium Iodide Chemical Storage 1 Two 250 g bottles dium Sulfite Chemical Storage 2 Two 250 g bottles Liquid furic Acid Chemical Storage 3599 (Two 1000 l IBC) fur Hexafluoride Irradiation unit cells (8) 1440 Pressurized Gas (NDAS pressure vessel)
Solid pieces nium Metal Uranium Storage 620 (Storage container maximum size: 7.8 kg)
Solid powder nium Oxide Uranium Storage 732 (Storage container maximum size: 5.04 kg)
Intermediate compound in nyl Peroxide Target Solution Preparation 43 uranyl sulfate dissolution tanks Liquid nyl Sulfate In Process 1058 (Target solution prep tank maximum [ ]PROP/ECI)
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Table 13b.3 Hazardous Chemical Source Terms and Concentration Levels (Sheet 1 of 2)
Hazardous Site Boundary Nearest Chemical/Release Airborne Source Term Worker Concentration Residence Mechanism MAR (kg) Fraction (kg) PAC-1 PAC-2 PAC-3 Concentration (230 m)(a) (788 m)(a) 0.49 5.4 32 0.0279 Alpha-Benzoin Oxime 0.25 2E-3 5.00E-4 Negligible Negligible mg/m3 mg/m3 mg/m3 mg/m3 Ammonia 249.5 1.0 249.5 30 ppm 160 ppm 1100 ppm 0.553 ppm 27.9 ppm 2.39 ppm (28 wt.%)
[ [ [ 0.558 0.0706
]PROP/ECI 5 2E-3 0.01 Negligible
]PROP/ECI ]PROP/ECI ]PROP/ECI mg/m3 mg/m3 Hydrochloric Acid 2 1.0 2 1.8 ppm 22 ppm 100 ppm 7.58E-3 ppm 0.155 ppm 0.0208 ppm (38 wt.%)
Hydrogen Peroxide 334.0 1.0 334.0 10 ppm 50 ppm 100 ppm 4.05E-3 ppm 0.0367 ppm Negligible (30 wt.%)
140 1500 8900 Mineral Oil 180 1.0 180 Negligible Negligible Negligible mg/m3 mg/m3 mg/m3 2.3 43 260 0.112 0.0141 Molybdenum Trioxide 1 2E-3 0.002 Negligible mg/m3 mg/m3 mg/m3 mg/m3 mg/m3 Nitric Acid (70 wt.% in 989.5 1.0 989.5 0.16 ppm 24 ppm 92 ppm 0.0157 ppm 0.0769 ppm 0.00858 ppm chemical storage)
Potassium 0.25 2E-3 5.00E-4 None Identified None Identified None Identified N/A N/A N/A exachloro-ruthenate(b) 8.6 14 150 0.0279 otassium Permanganate 0.25 2E-3 5.00E-4 Negligible Negligible mg/m3 mg/m3 mg/m3 mg/m3 0.1 0.0279 Rhodium Chloride(c) 0.25 2E-3 5.00E-4 None Identified None Identified Negligible Negligible mg/m3 mg/m3 0.047 0.9 5.4 0.0279 Silver Nitrate 0.25 2E-3 5.00E-4 Negligible Negligible mg/m3 mg/m3 mg/m3 mg/m3 Sodium Hydroxide 0.5 5 50 0.00287 770.5 4E-7/hr 3.082E-4 kg/hr Negligible Negligible (50.5 wt.%) mg/m3 mg/m3 mg/m3 mg/m3 NE Medical Technologies 13b.3-5 Rev. 0
Hazardous Site Boundary Nearest Chemical/Release Airborne Source Term Worker Concentration Residence Mechanism MAR (kg) Fraction (kg) PAC-1 PAC-2 PAC-3 Concentration (230 m)(a) (788 m)(a) 13 140 860 0.0279 Sodium Iodide 0.25 2E-3 5.00E-4 Negligible Negligible mg/m3 mg/m3 mg/m3 mg/m3 11 120 710 0.0279 Sodium Sulfite 0.25 2E-3 5.00E-4 Negligible Negligible mg/m3 mg/m3 mg/m3 mg/m3 0.20 8.7 160 4.48E-4 Sulfuric Acid 1799.5 1.0 1799.5 Negligible Negligible mg/m3 mg/m3 mg/m3 mg/m3 Sulfur Hexafluoride 1440 1.0 168.48 3000 ppm 33000 ppm 200000 ppm 9.41 ppm 3.31 ppm 0.368 ppm Hydrogen Fluoride(d) 0.3105 1.0 0.3105 1 ppm 24 ppm 44 ppm 0.0173 ppm 0.0445 ppm Negligible (d)
Sulfur Dioxide 1.45E-3 1.0 1.45E-3 0.2 ppm 0.75 ppm 30 ppm 8.10E-5 ppm Negligible Negligible 0.6 5 30 Uranium Metal(e) 7.8 0 Negligible Negligible Negligible Negligible mg/m3 mg/m3 mg/m3 0.71 10 50 0.558 0.0706 Uranium Oxide 5.04 2E-3 0.01 Negligible mg/m3 mg/m3 mg/m3 mg/m3 mg/m3 1.04E-3 1.60E-4 Uranyl Peroxide(f) 43 4E-7/hr 1.72E-5 None Identified None Identified Negligible Negligible mg/m3 mg/m3 4.0672E-4 kg/ 0.78 4.3 26 3.79E-3 Uranyl Sulfate(g) 1016.8 4E-7/hr Negligible Negligible hr mg/m3 mg/m3 mg/m3 mg/m3
- a. Consequence is taken as the outside concentration calculated by ALOHA. Negligible values are those determined by the software to not have significant concentration at the point selected.
- b. Based on the absence of PAC levels and significant health effects, further evaluation of potassium hexachlororuthenate was not required.
- c. PAC limits were not identified for rhodium chloride. An acceptance value of 0.1 mg/m3 is assigned based on acceptable OSHA occupational exposure limits from the MSDS for uranyl rhodium chloride.
- d. Hydrogen fluoride and sulfur dioxide are chemical byproducts and are not normally stored on site.
- e. Uranium metal is stored as solid pieces. Therefore, there is no hazard from dropping solid metal pieces.
- f. PAC limits were not identified for uranyl peroxide. An acceptance value of 1.04E-3 mg/m3 is assigned based on acceptable OSHA occupational exposure limits from the MSDS for uranyl peroxide.
- g. PAC limits were not identified for uranyl sulfate; therefore, values for a similar material uranyl fluoride are used. This is applicable since soluble uranium compounds pose similar health hazards.
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NRC, 2012. Final Interim Staff Guidance Augmenting NUREG-1537, Part 1, "Guidelines for paring and Reviewing Applications for the Licensing of Non-Power Reactors: Format and tent," for Licensing Radioisotope Production Facilities and Aqueous Homogeneous ctors, U.S. Nuclear Regulatory Commission, October 17, 2012.
NRC, 1998. Nuclear Fuel Cycle Facility Accident Analysis Handbook, NUREG/CR-6410,
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