ML21095A217

Shine Medical Technologies, LLC, Revisions to Final Safety Analysis Report, Chapter 13, Rev. 4, Accident Analysis

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Chapter 13 - Accident Analysis Table of Contents CHAPTER 13 ACCIDENT ANALYSIS TABLE OF CONTENTS Section Title Page 13a2 IRRADIATION FACILITY ACCIDENT ANALYSIS .......................................... 13a2.1-1 13a2.1 ACCIDENT-INITIATING EVENTS AND SCENARIOS ................................... 13a2.1-6 13a2.1.1 IF MAXIMUM HYPOTHETICAL ACCIDENT .......................... 13a2.1-7 13a2.1.2 INSERTION OF EXCESS REACTIVITY ................................ 13a2.1-7 13a2.1.3 REDUCTION IN COOLING .................................................. 13a2.1-16 13a2.1.4 MISHANDLING OR MALFUNCTION OF TARGET SOLUTION ........................................................................... 13a2.1-19 13a2.1.5 LOSS OF OFF-SITE POWER .............................................. 13a2.1-23 13a2.1.6 EXTERNAL EVENTS ........................................................... 13a2.1-25 13a2.1.7 MISHANDLING OR MALFUNCTION OF EQUIPMENT ....... 13a2.1-30 13a2.1.8 LARGE UNDAMPED POWER OSCILLATIONS .................. 13a2.1-32 13a2.1.9 DETONATION AND DEFLAGRATION IN THE PRIMARY SYSTEM BOUNDARY ......................................................... 13a2.1-33 13a2.1.10 UNINTENDED EXOTHERMIC CHEMICAL REACTIONS OTHER THAN DETONATION .............................................. 13a2.1-36 13a2.1.11 SYSTEM INTERACTION EVENTS ...................................... 13a2.1-37 13a2.1.12 FACILITY-SPECIFIC EVENTS ............................................. 13a2.1-42 13a2.2 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 SHINE Medical Technologies 13-i Rev. 0

Chapter 13 - Accident Analysis Table of Contents CHAPTER 13 ACCIDENT ANALYSIS TABLE OF CONTENTS Section 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 ...................................... 13a2.2-23 13a2.2.12 FACILITY-SPECIFIC EVENTS ............................................. 13a2.2-25 13a3

SUMMARY

AND CONCLUSIONS .................................................................... 13a3-1 13a4 REFERENCES .................................................................................................. 13a4-1 13b RADIOISOTOPE PRODUCTION FACILITY ACCIDENT ANALYSES............... 13b.1-1 13b.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 13b.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.2-1 13b.2.3 EXTERNAL EVENTS ............................................................... 13b.2-1 13b.2.4 RPF CRITICAL EQUIPMENT MALFUNCTION ........................ 13b.2-2 13b.2.5 RPF INADVERTENT NUCLEAR CRITICALITY ....................... 13b.2-8 13b.2.6 RPF FIRE ................................................................................. 13b.2-9 SHINE 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

2.1-1 Hazard Types 2.1-2 Risk Matrix 2.1-3 Likelihood Category Definitions 2.1-4 Failure Frequency Index Numbers 2.1-5 Failure Probability Index Numbers 2.1-6 Duration Index Numbes 2.1-7 Consequence Category Definitions 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 Quantities of In-Process Hazardous Chemicals

.3-2 Hazardous Chemical Source Terms and Concentration Levels NE Medical Technologies 13-iv Rev. 1

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 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 A fire hazards analysis NE Medical Technologies 13-vi Rev. 1

onym/Abbreviation Definition EA failure modes and effects analyses feet gallons 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 interim staff guidance irradiation unit iodine and xenon purification and packaging effective neutron multiplication factor kilowatt NE Medical Technologies 13-vii Rev. 1

onym/Abbreviation Definition 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 A maximum hypothetical accident 99 molybdenum-99 NE Medical Technologies 13-viii Rev. 1

onym/Abbreviation Definition 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. 1

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. 1

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. 1

onym/Abbreviation Definition SS uninterruptible electrical power supply system SS uranium receipt and storage system C/ITS vacuum/impurity treatment subsystem vacuum transfer system atmospheric dispersion factor NE Medical Technologies 13-xii Rev. 1

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 (RPF).

ign basis accidents (DBAs) were identified using the following sources of information:

• NUREG-1537 (USNRC, 1996) and the Interim Staff Guidance (ISG) Augmenting NUREG-1537 (USNRC, 2012a);

• Process hazard analysis (PHA) method within the safety analysis; 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.

NE Safety Analysis (SSA) Methodology NE applies a SHINE-specific, risk-based methodology similar to the guidance described in REG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications (USNRC,

5) in the development of the detailed accident analysis. This methodology is applied to both IF and the RPF for consistency of the safety analysis for the entire SHINE facility.

SSA is a systematic analysis of facility processes used to identify facility hazards associated the processing and possession of licensed materials. The SSA has been performed for the pose of identifying relevant hazards, potential accident sequences and consequences, ipment and specific human actions credited for safety, and programmatic administrative trols necessary to ensure the availability and reliability of safety-related SSCs. This analysis NE Medical Technologies 13a2.1-1 Rev. 4

licability mal operation at the SHINE facility includes IF operations as well as chemical extraction and fication operations, target solution preparation and storage activities, and waste handling and obilization activities in the RPF.

SSA considers all modes of operation for potential process upsets and accident sequences.

subcritical assembly system (SCAS) for each IU is analyzed for each mode of operation (i.e.,

ution Removed, Startup, Irradiation, Post-Irradiation, Transfer to RPF). The associated target tion vessel (TSV) off-gas system (TOGS) operation is combined with the SCAS analysis as are tightly coupled systems. Since the tritium purification system (TPS) services all eight

, it is analyzed as a continuously operating integrated system. The operating modes for the include normal gas feed, recovery and purification, and TPS glovebox cleanup.

NE systems which operate in a batch mode are analyzed for active operation while ardous materials are present. The molybdenum extraction and purification system (MEPS) the iodine and xenon purification and packaging (IXP) system are either in use or idle. They therefore analyzed for normal extraction, purification, and packaging activities. Similarly, the et solution and preparation system (TSPS) and the uranium receipt and storage system SS) are analyzed for normal target solution preparation activities. The radioactive liquid te system (RLWS) and the radioactive liquid waste immobilization (RLWI) system are also lyzed for normal storage and processing of liquid wastes.

SSA considers maintenance activities as potential initiators for accident sequences including ntenance errors, improper system restoration, impacts on operating equipment, and fires.

se types of initiators were identified during the accident sequence development phase of the A.

-routine activities may include the repair or replacement of major components such as the tron drivers or the high voltage power supplies (HVPS). Accident sequences considered in SSA include heavy load drops on systems or components containing radiological material inadvertent exposure to neutrons.

iods of extended shutdown are not explicitly identified as a class of accident sequences in the A; however, SHINE systems are designed to achieve and maintain a safe condition for ological materials in extended storage.

hnical specifications require limiting conditions for operation (LCO) be met during the cified conditions of applicability. When an LCO is not met, the applicable actions specified in technical specifications are required to be completed.

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SSA is developed based on the following major steps:

• Identification and systematic evaluation of hazards at the facility

• Comprehensive identification of potential accident/event sequences that would result in unacceptable consequences, and the expected likelihoods of those sequences

• Assessment of radiological and chemical consequences for postulated accident sequences to demonstrate compliance with acceptable limits

• Identification and description of safety-related controls (i.e., structures, systems, equipment, components, or specific actions) that are relied on to limit or prevent potential accidents or mitigate their consequences

• Identification of programmatic administrative controls that ensure the availability and reliability of identified safety systems.

results of the SSA consist of postulated accident sequences for inclusion in this chapter.

includes a description of the accident sequences, potential consequences, controls credited revent or mitigate the accident sequence, and a summary of calculated dose consequences.

ard Identification and Evaluation ard identification is performed by identifying, for each process, radiological or chemical ards that have the potential for causing harm to the public, facility staff, or the environment.

includes physical process hazards (e.g., deflagration, fire, flooding) that could result in erse effects on licensed materials. Radiological hazards include radiation sources from the NE processes (e.g., neutron driver, TSV), fission products, activation products, and tritium.

ile material hazards are also considered for postulated criticality accidents. Chemical ards are identified that could affect licensed materials or the safe operation of the facility.

mical effects considered include flammable, reactive, oxidation, and chemical incompatibility cts.

hazard identification and evaluation is performed using standard hazard evaluation methods h as Hazard and Operability Analysis (HAZOP) and Failure Modes and Effects Analysis EA). The types of hazards identified for the SHINE facility are identified in Table 13a2.1-1.

ard evaluations are conducted to assess potential failures, causes, and consequences that vide a basis for the development of potential accident sequences. The output of the hazard luations are those failure-cause-consequences that have the potential for causing harm to the lic, facility staff, or the environment and the possible engineered or administrative controls may be applied for prevention or mitigation.

cess Hazard Analysis and Accident Sequence Development results of the hazard evaluations are used to inform the PHA and accident sequence elopment phase. The PHA uses the results of the hazard evaluations to develop accident uences in alignment with the accident sequence categories described in Section 13a2.1 for IF and Subsection 13b.1.2 for the RPF.

ident sequence development uses the risk index methodology based on risk index values cribed in NUREG-1520 (USNRC, 2015). Potential accident sequences are defined based on NE Medical Technologies 13a2.1-3 Rev. 4

rnal events, or combinations of these elements.

ernal event-induced accident sequences are treated on a site-wide basis. The external events A also includes fires and flooding from causes internal to the IF and the RPF. External event ating events that are considered include:

• external events such as earthquake, external flooding, external fires, high winds, and tornadoes;

• events that are external to the process being analyzed such as internal fires and internal flooding;

• deviations from normal process operations (credible abnormal events);

• failures of process components; and

• human errors that result in process upsets or failures.

ential consequences are also identified for each accident sequence as one or more of the wing:

• Radiological dose to the public or facility staff (i.e., control room operator)

• Chemical dose to the public or facility staff (i.e., control room operator and radiologically controlled area [RCA] worker)

• Criticality event

• No consequence of concern radiological consequence analysis is described in Section 13a2.2 for the IF and tion 13b.2 for the RPF. The chemical consequence analysis is described in Section 13b.3.

ident sequences that may result in a consequence of concern are first evaluated with no ineered or administrative controls applied, referred to as an uncontrolled accident uence. A total risk index number is determined based on an estimate for the likelihood of urrence and severity of consequences. For accident sequences with unacceptable risk ces, engineered and administrative controls are applied that reduce the likelihood of urrence and/or the severity of the consequences such that an acceptable risk level is ched. Acceptable risk levels for SHINE require that the postulated sequence is highly kely and/or the consequence severity is low. The final accident sequence is referred to as a ntrolled accident sequence. The credited engineered and administrative controls are tified as safety-related controls.

k Matrix Development risk matrix applied in the SSA is provided in Table 13a2.1-2. The risk matrix approach vides a method of determining the risk of various accident sequences based on a quantitative mate of the likelihood of occurrence and the severity of the consequences. The likelihood of urrence and the consequence severity for each uncontrolled accident sequence is estimated corresponding categories are assigned. The risk matrix then identifies those credible idents which have the potential to exceed the acceptable risk index values, and therefore uire engineered and/or administrative controls for prevention or mitigation. The risk index es are then reassessed after application of engineered or administrative controls that result n acceptable risk outcome.

NE Medical Technologies 13a2.1-4 Rev. 4

likelihood category definitions applied in the SSA are provided in Table 13a2.1-3. The ermination of the likelihood of occurrence consists of the initiating event frequency (e.g.,

mic event, process component failure, human error) and may be combined with an additional ponent failure or human error, including any recovery times (i.e., failure duration). In most es, the initiating events are represented by single events or single failures.

frequency of occurrence of an initiating event for an accident sequence is represented by a re frequency index number (FFIN). The FFINs applied in the SSA are provided in le 13a2.1-4. The bases for determining the FFIN for an accident sequence may include ence or the type of control.

determine the FFIN selected for an accident sequence initiator based on the type of control, eral factors are considered including:

• administrative (i.e., human error);

• type of component failure (i.e., active versus passive);

• degree of redundancy (i.e., single component, redundant component);

• design margin (e.g., design pressure versus nominal pressure); and

• other factors including degree of enhancement for administrative controls (e.g.,

independent verification and step sign-off).

e accident sequence is postulated to occur only if another condition or failure is present, an itional probability of component failure or condition is included in the evaluation. The failure bability index number (FPIN) represents this as a failure on demand, or as a probability that condition exists. This can be evaluated as a simple probability of failure on demand or roximated as the product of a failure rate and a recovery time, defined in this analysis as a ation index number (DIN). The quantitative characterization of the FPIN and DIN applied in SSA is provided in Table 13a2.1-5 and Table 13a2.1-6, respectively.

sequence Category Definitions consequence category definitions applied in the SSA are provided in Table 13a2.1-7.

merical limits for the radiological and chemical exposure effects are included in the definitions high and intermediate consequence for the public and facility staff. The low consequence gory is implicitly defined as resulting in consequences that are less than intermediate and et the SHINE safety criteria limits defined in Section 3.1.

ety-Related Controls accident sequences developed in the PHA phase identify the controls that are credited for vention and/or mitigation of accident sequences. The types of safety-related controls that are dited for prevention and/or mitigation of accident sequences are:

• Engineered controls (active or passive), identified as safety-related SSCs; and

• Specific administrative controls (e.g., procedural controls) ety-related controls that are credited for prevention and/or mitigation are identified for each ident scenario in Section 13a2.2 and Section 13b.2.

NE Medical Technologies 13a2.1-5 Rev. 4

not credited in accident sequences but provide additional margin for risk reduction.

rporation into the FSAR and Technical Specifications ident sequences developed in the SSA inform the accident analysis and determination of sequences of the limiting accident scenarios described in Section 13a2.2 for the IF and tion 13b.2 for the RPF.

safety-related SSCs that are required to be operable to meet the assumptions underlying the A are included within Section 3.0 of the technical specifications, Limiting Conditions for ration and Surveillance Requirements.

tion 4.0 of the technical specifications, Design Features, includes design features that are tified in the SSA. These are aspects of the facility design and other physical conditions (e.g.,

ance to the site boundary, building free volume) that are inputs or assumptions in the ological dose calculations that support the SSA dose consequence analysis.

SSA also identifies the programmatic administrative controls that are required to be lemented to ensure that safety-related SSCs will be capable of performing their intended ctions. Section 5.0 of the technical specifications, Administrative Controls, includes the grammatic administrative controls identified in the SSA (e.g., maintenance of safety-related Cs) and requires that those programs are established, implemented, and maintained.

tion 5.0 additionally requires the development and use of procedures that implement the cific administrative controls identified in the SSA. Section 5.0 also includes discussion of the figuration management program, which provides oversight and control of design information, ty information, and records of modifications that might impact the ability of safety-related Cs to perform their intended functions. The configuration management program also lists A-identified controls not otherwise included in Sections 3.0, 4.0, or 5.0 of the technical cifications that will be maintained under the configuration management program and will not modified as described in the technical specifications without prior NRC approval.

2.1 ACCIDENT-INITIATING EVENTS AND SCENARIOS DBAs identified in this section are credible accident scenarios that range from anticipated nts, such as a loss of electrical power, to events that are still credible, but considered unlikely ccur during the lifetime of the plant. The maximum hypothetical accident (MHA) is also ned to result in the bounding radiological consequences for the SHINE facility.

ed on the guidance provided in the ISG Augmenting NUREG-1537 (USNRC, 2012a), the wing accident categories were used to identify potential accident sequences:

• MHA (Subsection 13a2.1.1)

• Excess reactivity insertion (Subsection 13a2.1.2)

• 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)

NE Medical Technologies 13a2.1-6 Rev. 4

• 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 PHA process and are therefore considered within each event category.

2.1.1 MAXIMUM HYPOTHETICAL ACCIDENT guidance in NUREG-1537 (USNRC, 1996) describes the MHA as a postulated accident nario whose potential consequences are shown to exceed those of any credible accidents, that such a scenario need not be entirely credible. SHINE considers such a scenario to be a ond design basis accident (BDBA).

eu of identifying a BDBA scenario as the MHA for the SHINE facility, SHINE has chosen to tify a credible fission product-based DBA which bounds the radiological consequences to the lic of all credible fission product-based accident scenarios as the MHA for the SHINE facility.

MHA for the SHINE facility is identified as the failure of the TSV TOGS pressure boundary ulting in a release of off-gas into the TOGS cell. A general description of this scenario is vided in Subsection 13a2.1.7.2, Scenario 1. A detailed description of this scenario and an luation of the radiological consequences is provided in Subsection 13a2.2.7.

2.1.2 INSERTION OF EXCESS REACTIVITY excess reactivity insertion event during normal operations is identified as a potential initiating nt for accidents in the accident analysis. The potential for excess reactivity insertions during startup process and irradiation mode of the TSV was identified as scenarios to be evaluated.

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) 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 NE Medical Technologies 13a2.1-7 Rev. 4

the subcritical assembly, when it is not being driven by the neutron driver (such as in Mode 1 ode 2 during Driver Dropout), excess reactivity insertion could lead to inadvertent criticality unplanned 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 following postulated initiating events and scenarios that could lead to an excess reactivity rtion or power transient during operation were identified using the guidance in the ISG menting NUREG-1537 (USNRC, 2012a):

• Increase in the target solution density during operations (e.g., due to pressurization)

• Target solution temperature reduction during fill/startup (e.g., excessive cooldown)

• 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 and Mode 2 rations:

• 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).

• The TSV is operated in a subcritical state with a nominal keff of approximately [

]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.

NE Medical Technologies 13a2.1-8 Rev. 4

• 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.

ssure transients are described in Subsection 4a2.6.1.4 and limiting pressure transient lysis results are discussed in Subsection 4a2.6.3. Peak power during the pressure transients 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 NE Medical Technologies 13a2.1-9 Rev. 4

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 tion to the TSV dump tank results in safe shutdown of the target solution. No damage to the B occurs and there are no radiological consequences.

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 TPS control system or component failure during tup that injects excess tritium before the TSV is at operating temperature. The TRPS initiates U 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, [

]PROP/ECI. This prevents driver from producing excessive neutrons concurrent 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.

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]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.

ere were a breach between the TSV and PCLS or light water pool, cooling water could be ed to the target solution. Moderator addition could also occur due to failure of a TOGS 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.

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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 he TSV. No mechanisms, other than target solution adjustment, have been identified that ld change fissile material per volume of target solution outside the bounds of what has been 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.

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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 DID control allows the delayed neutron precursor population time to reach dy state. Subsection 7.4.4.1 provides a detailed description of the TRPS Fill Stop function.

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.

nario 7 - Realistic, Adverse Geometry Changes 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 NE Medical Technologies 13a2.1-13 Rev. 4

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 de 2. For a voiding of 100 percent, the reactivity changed by approximately

]PROP/ECI in Mode 1 and [ ]PROP/ECI in Mode 2. The reactivity impact from 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, process upsets that could lead to vertent introductions were evaluated. The inadvertent introduction of other materials into the could come from: (1) sources external to the PSB, or (2) sources internal to the PSB itself.

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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 potential sources of uranium solids entering the TSV and resulting in reactivity addition were luated: uranyl salt crystal buildup in the TSV or TOGS components and precipitation of nium solids.

first two postulated scenarios are a buildup of uranium-bearing salt crystals in the TSV (such "bathtub" ring) or in TOGS components. These salt crystals could become rewetted or 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 NE Medical Technologies 13a2.1-15 Rev. 4

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 tulated scenarios where the uranium concentration of the target solution could increase were luated. One identified scenario requiring control was the TOGS pressure control failure ing to excess evaporation. The other identified scenario requiring control was failure of GS 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.

nario 11 - Failure to Control Temperature during 1/M Measurement at Startup tulation that a failure in the PCLS occurs during the startup process results in high target tion temperature and errors in the 1/M measurements during the fill process. These errors ld be non-conservative and lead to an increase in reactivity during the fill process. This nario is prevented through the TRPS IU Cell Safety Actuation on high source range neutron or high PCLS temperature. Following the TRPS trip, the target solution dumps to the TSV p 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:

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• 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.

re are several failures that can result in the reduction in cooling in the SCAS:

• 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.

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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, no damage to the PSB occurs. The TSV dump tank is designed to maintain the target tion subcritical.

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.

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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.

eral potential initiating events were evaluated, including:

• 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.

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• 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:

nario 1a - Failure of the PSB Below the Level of the Light Water Pool 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.

protections in place to prevent a failure of the PSB below the level of the light water 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).

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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.

protections are present to help prevent a failure of the PSB between the TSV and PCLS, ch include:

• 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 primary 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 NE Medical Technologies 13a2.1-21 Rev. 4

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 Z1e IU cell radiation monitors in the RVZ1e exhaust duct, and initiating an IU Cell Safety uation including isolation of the PCLS isolation valves and the RVZ1e exhaust duct.

DID, the failure of the pressure boundary may first result in in-leakage and overflow into the dump tank, which is detected with the level detection in the TSV dump tank. The TRPS then es the PCLS isolation valves and RVZ1e isolation dampers, stopping potential transfer of et solution to the PCLS and reducing the source of water that could enter the PSB, and ating the ventilation exhaust from the IU cell.

itional DID measures are also in place to avoid a leak and detect leaks, which include:

• 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.

small amount of target solution that could diffuse into the PCLS cooling water after the ssure between the PCLS and the PSB is equalized, combined with the dilution of the leaked erial by the cooling water, minimizes the potential for criticality in the PCLS and dose to kers or the public.

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 vacuum transfer system (VTS) header, and into the VTS er tank. This failure potentially results in radiological exposures to workers or a criticality ident in non-favorable-geometry 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.

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.

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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.

2.1.5 LOSS OF OFF-SITE POWER 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 NE Medical Technologies 13a2.1-23 Rev. 4

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.

initial conditions and assumptions are summarized below:

• 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.

• Hydrogen concentration in TSV and TOGS is maintained within operating limits prior to the event.

• 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.

NE Medical Technologies 13a2.1-24 Rev. 4

• 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 PSB. Nitrogen gas is injected in the eight SCAS systems via a connection to the dump tank. The gas purges the PSB leaving through a vent connection from the TOGS to the process vessel vent system (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.

ddition to the above sequence of events in the IU, the following actions also occur ultaneously:

• 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.

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following potential external events were evaluated:

• 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.

ddition, seismic events assume that:

• 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.

• Seismic Category II SSCs are not required to remain functional in the event of a DBE.

further details of seismic design criteria refer to Section 3.4.

2.1.6.2 General Scenario Descriptions following discusses the external event scenarios which impact the 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 NE Medical Technologies 13a2.1-26 Rev. 4

nario 2 - Seismic Event causing PCLS Failure eismic event may cause the failure of the PCLS. A failure of PCLS in one or more IUs could ult in reduction in or excessive cooling, reactivity insertion, and potential criticality. Potential sequences 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 IUs results in a release of tritium in one or more IU cells.

ential consequences of multiple NDAS failures include radiological dose. The dose analysis servatively assumes the simultaneous failure of all eight NDAS to bound the maximum wable operating state in the IF.

mitigate the impact of such failures of the NDAS vacuum boundary, TPS target chamber ply pressure and TPS target chamber exhaust pressure instrumentation and ventilation ation mechanisms are used to confine released tritium. Accident consequences of this event discussed 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 NE Medical Technologies 13a2.1-27 Rev. 4

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 MEPS. The PVVS is assumed to resulting in a loss of sweep gas in the vacuum transfer tanks. Due to the lack of circulation, hydrogen concentration in the vacuum transfer tanks increases, approaching the deflagration

t. In this event, the N2PS dilutes the hydrogen gas concentration and prevents hydrogen agration.

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 PSB are discussed in section 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.

NE Medical Technologies 13a2.1-28 Rev. 4

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.

nario 11 - Heavy Snow or Ice due to Severe Winter Weather 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.

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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 main 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 A. Analysis of this event and other potential mishandling or malfunction of equipment events, luding detonation or deflagration of hydrogen within TOGS and other exothermic chemical ctions within the PSB, are included in this section.

• Events involving the mishandling or malfunction of target solution are discussed in Subsection 13a2.1.4.

• 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.

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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.

2.1.7.2 General Scenario Descriptions 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 NE Medical Technologies 13a2.1-31 Rev. 4

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.

TSV experiences power oscillations with reactivity variations within the target solution.

tron driver output can oscillate and lead to power variations as well, but driver output 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

-limiting 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:

• TSV TOGS failure results in variations in TSV gas pressure.

• 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 NE Medical Technologies 13a2.1-32 Rev. 4

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.

discussed in Subsection 4a2.6.1.4 the large negative temperature and void coefficients result stable TSV with self-limiting power oscillations under analyzed reactivity variations.

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 NE Medical Technologies 13a2.1-33 Rev. 4

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. Several potential scenarios that could result in the accumulation of hydrogen potential deflagration or detonation were evaluated. A deflagration or detonation accident ld occur if the TOGS fails, which could allow hydrogen to accumulate in the TSV headspace, p tank, or TOGS piping. Potential failures that have been identified include a loss of power, re of the TOGS blowers, blockage or restriction in the TOGS flow path, and water leakage the PSB that results in reduced sweep gas flow. Hydrogen could also accumulate if there is raded performance of the TOGS, such as reduced volumetric flow rate due to a partially-tructed 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 delayed fission product decay further increases hydrogen concentration. Hydrogen centration may increase above the LFL, which may cause a deflagration, but remains lower 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 NE Medical Technologies 13a2.1-34 Rev. 4

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, the criteria to transition from Mode 0 to Mode 1 include the TOGS mainstream flow g at or 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.

pressure safety limit of the PSB is greater than the maximum credible deflagration pressure does not fail due to a deflagration within the PSB.

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.

NE Medical Technologies 13a2.1-35 Rev. 4

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 scenario evaluated in this subsection is the uranium metal-water reaction in the neutron tiplier.

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.

neutron multiplier radionuclide inventory is developed assuming 30 years of continuous ration at 137.5 kW.

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.

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.

NE Medical Technologies 13a2.1-36 Rev. 4

Chapter 13 - Accident Analysis Accident-Initiating Events and Scenarios 13a2.1.10.3 Accident Consequences The accident consequences associated with unintended exothermic chemical reactions are discussed further in Subsection 13a2.2.10. The accident consequences associated with a tritium release from the TPS glovebox are discussed further in Subsection 13a2.2.12.

13a2.1.11 SYSTEM INTERACTION EVENTS This subsection discusses the effects of system interactions on the systems which contain radionuclide material. System interactions have the potential to cause damage that may lead to the release of these materials.

Three categories of system interactions between systems located within the IF and the RPF are considered in this analysis. These are: (1) functional interactions, (2) spatial interactions, and (3) human-intervention interactions.

Functional Interactions Functional interactions are interactions between systems or subsystems that result from a common interface. A functional interaction exists if the operation of one system can affect the performance of another system or subsystem. An adverse functional interaction exists when the operation and/or performance of an (initiating) system adversely affects the operation and/or performance of an SSC as it performs its safety-related function. Functional interaction events that are discussed in this section are those that may result from failures in support systems or other shared systems that could result in an adverse impact on the PSB.

PVVS is connected to the eight IUs via connections to TOGS. Accidents with PVVS failure are considered in Section 13b.2.

The functional interactions considered in this analysis are the following:

Loss of Off-Site Power The NPSS provides electrical power to SSCs in the IF and the RPF.

Reduction 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

- MEPS process coolers

- PVVS condensers

- NDAS cooling cabinets SHINE Medical Technologies 13a2.1-37 Rev. 4

s of ventilation

• The ventilation systems (RVZ1, RVZ2) are described in Section 9a2.1.

• Loss of RVZ1 flow may result in maloperation of multiple systems in the IF and RPF, such as the:

- TPS glovebox pressure control exhaust and the vacuum/impurity treatment subsystem (VAC/ITS) process vents

- 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 primary 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.

NE Medical Technologies 13a2.1-38 Rev. 4

following section discusses the system interactions that can occur at the main production lity. System interactions that are already analyzed in other parts of Chapter 13 are referenced hose subsections and not evaluated in this subsection. System interactions that are not cribed in other subsections are discussed below.

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 IU. Decay heat from the target solution is oved by the light water pool.

nario 2 - Loss of Normal Ventilation to Primary 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 primary 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 primary cooling room with potential for increased instrument curacies or failure. The consequences of an RVZ2 failure leading to equipment malfunction ult in TSV overcooling causing a reactivity insertion in the TSV. Excess reactivity additions are ussed further in Subsection 13a2.1.2.

NE Medical Technologies 13a2.1-39 Rev. 4

tem). The TRPS signal opens redundant TSV dump valves draining target solution to the TSV p tank and shuts down the IU. Decay heat from the target solution is removed by the light er pool system (LWPS).

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. The principle fire hazards in the IF are:

the HVPS used for the NDAS service cell, (2) hydrogen located in the TPS and within the B for each IU cell, and (3) the carbon filters in the RCA exhaust filter room in the mezzanine

a. Causes of fires include a catastrophic failure of the HVPS and maintenance activities uding 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.

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s caused by external events are discussed in Subsection 13a2.1.6.

thermic Chemical Reaction Scenarios thermic chemical reaction scenarios are discussed in Subsection 13a2.1.10.

rnal Flooding ential internal flooding scenarios in the IF have been evaluated.

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 DBE and DBA events. Penetrations through the t water pool liner are above the minimum water level.

oding caused by external events is discussed in Subsection 13a2.1.6.

amic Effects cess systems in the main production facility operate at low temperatures (i.e., generally less n 200°F [93°C], except for the TOGS hydrogen recombination components) and low ssures (i.e., less than 100 psig [689 kPa gauge]), which are not subject to dynamic effects as found in high energy systems. As needed, safety-related systems are protected from the amic effects related to equipment failure and external events. No consequences result from amic effects interactions in the main production facility.

man Intervention Interactions man interventions can cause adverse system interactions because of the single common trol room at the main production facility. Operators are able to control multiple systems within IF and the RPF from the control room. Operator errors may occur including performing trol operations on the wrong system, failing to perform required actions, or performing actions of sequence.

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 NE Medical Technologies 13a2.1-41 Rev. 4

ected upon return to service through post-maintenance testing. However, undetected errors y result in system failures at some later point in time.

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 eral accident scenarios that are unique to the IF and have the potential for inadvertent ation exposure to workers or members of the public were evaluated. Facility-specific accident narios are associated with the NDAS, the TPS, and potential IF damage resulting from heavy 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 IU cell radiation monitors 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 helium.

• 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-NDAS interface lines contain the maximum inventory of tritium gas.

NE Medical Technologies 13a2.1-42 Rev. 4

• 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 IF. These scenarios (1) inadvertent exposure to neutrons within the IU, (2) inadvertent exposure to neutrons in NDAS service cell (NSC), (3) catastrophic failure of the NDAS, and (4) an NDAS vacuum ndary 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.

NE Medical Technologies 13a2.1-43 Rev. 4

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 IF, which results in higher dose to workers and the public due to the ase of tritium.

accident scenario is mitigated by the primary confinement boundary pressure sensors in the target chamber supply lines and TPS target chamber exhaust lines and isolation valves on Z1e from the PCLS expansion tank. The primary confinement boundary is described in detail hapter 6. Multiple catastrophic failures of NDAS 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 IF. These scenarios (1) TPS piping failure due to deflagration, (2) release of tritium into the IF due to glovebox agration, (3) release of tritium to the facility stack, (4) excessive release of tritium from the m storage bed, and (5) release of tritium into the IF due to TPS-NDAS interface line chanical damage.

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 and secondary enclosure cleanup subsystem. The tritium confinement boundary is cribed in detail in Section 6a2.2. Isolation of the TPS room ventilation is also credited for gation.

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 13a2.1-44 Rev. 4

TPS gloveboxes are designed such that the minimum size prevents the possibility of ching the LFL for the quantity of available hydrogen. The TPS gloveboxes are also inerted helium which prevents the presence of oxygen. Based on the glovebox design and inert osphere, a deflagration in a glovebox is not considered credible and is not analyzed further.

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 a glovebox and a current misalignment of the VAC/ITS 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 TPS glovebox pressure control and VAC/ITS process vent exhaust to RVZ1e, which causes solation of the glovebox as part of the tritium confinement 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 TPS glovebox. The tritium release is confined within the tritium confinement boundary, which escribed in detail in Section 6a2.2.

protection in place to mitigate a release of tritium to the facility stack is the tritium monitor on TPS gloveboxes pressure control and VAC/ITS process vent exhaust, which causes an ation of the glovebox as part of the tritium confinement boundary.

Scenario 5 - Release of Tritium into the IF due to TPS-NDAS Interface Line Mechanical mage lease of tritium directly to the IF may be caused by mechanical damage to the TPS-NDAS rface lines which results in a leak of tritium to the IF. A release of tritium to the IF results in er dose to workers and to members of the public. The TPS-NDAS interface lines are in grade penetrations which reduces the likelihood of mechanical damage that results in a m leak and are protected from mechanical impact between the subgrade penetration and the gloveboxes.

majority of the length of the TPS-NDAS interface lines are routed in subgrade sleeves and therefore protected from mechanical damage from external impacts. A small length of the

-NDAS interface lines from the point at which they emerge from the subgrade sleeves in the room to the TPS glovebox isolation valves is protected from mechanical damage by external rds. Therefore, TPS-NDAS interface line mechanical damage is not credible.

NE Medical Technologies 13a2.1-45 Rev. 4

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 IF. These scenarios are a heavy load drop into an open IU cell, (2) a heavy load drop onto an in-service IU cell, and a heavy load drop onto TPS equipment.

vy Load Drop Scenario 1 - Heavy Load Drop into an Open IU Cell 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.

NE Medical Technologies 13a2.1-46 Rev. 4

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.

Scenario 2 and TPS Scenario 5 are not credible; therefore, accident consequences are not luated.

vy Load Drop vy load drop scenarios are not credible; therefore, accident consequences are not evaluated.

NE Medical Technologies 13a2.1-47 Rev. 4

Table 13a2.1 Hazard Types Hazard Type Hazards Fission products (in solution, aerosol, and off-gas), decay products, diological activation products, tritium, neutron, gamma Uranium oxide, uranium metal, uranyl sulfate (target solution),

sile uranyl peroxide, uranium salts Uranium, SF6 gas, SF6 decomposition products, fission and decay emical - Toxic products emical - Flammable/Explosive Hydrogen gas, oxygen gas, uranium metal emical - Reactivity Sulfuric acid, nitric acid, NaOH emical - Oxidizer Oxygen gas, hydrogen peroxide emical - Incompatibility Acids and bases emical - Asphyxiant Nitrogen gas, SF6 gas, clean agent for fire protection lagration/Detonation Hydrogen gas, oxygen gas h voltage Accelerator high voltage power supply h pressure Compressed gas cylinders (nitrogen, oxygen, helium), SF6 gas h temperature Accelerator ion beam, process heaters, hydrogen recombiners temperature Liquid nitrogen etic energy Ventilation and process steam blowers & fans Pressurized gas cylinders (nitrogen, oxygen, helium), SF6 pressure ential energy vessel Initiators (electrical equipment, maintenance), combustible rnal fire materials, hydrogen gas rnal flooding Process equipment, fire protection, cooling water systems Seismic, tornado, tornado generated missiles, severe weather, ernal events flooding (possible maximum precipitation), external fire, aircraft impact, industrial and transportation events (toxic gas, explosion)

NE Medical Technologies 13a2.1-48 Rev. 4

Chapter 13 - Accident Analysis Accident-Initiating Events and Scenarios Table 13a2.1 Risk Matrix Likelihood of Occurrence Severity of Likelihood Category 1 Likelihood Category 2 Likelihood Category 3 Consequences Highly Unlikely Unlikely Not Unlikely (1) (2) (3)

Consequence Category 3 Acceptable Unacceptable Unacceptable High 3 6 9 (3)

Consequence Category 2 Acceptable Unacceptable Unacceptable Intermediate 2 4 6 (2)

Consequence Category 1 Acceptable Acceptable Acceptable Low 1 2 3 (1)

SHINE Medical Technologies 13a2.1-49 Rev. 4

Table 13a2.1 Likelihood Category Definitions ikelihood Category Likelihood Index (T) Event Frequency Limit Risk Index Limits Less than 10-5 per Highly Unlikely 1 T -5 event, per year Between 10-4 and 10-5 Unlikely 2 -5 < T -4 per event, per year More than 10-4 per Not Unlikely 3 -4 < T event, per year NE Medical Technologies 13a2.1-50 Rev. 4

Table 13a2.1 Failure Frequency Index Numbers ailure Frequency Based on Based on Type of Control Comments ex Number (FFIN) Evidence External event If initiating event, no controls

-6 N/A with freq. < 10-6/yr needed.

For passive safe-by-design components or systems; failure is considered highly unlikely for robust passive engineered controls:

1. Whose dimensions fall within established single parameter limits or that can be shown by calculation to be subcritical Initiating event including the use of the approved

-5 N/A with freq. < 10-5/yr subcritical margin,

2. That have no credible failure mechanisms that could disrupt the credited design characteristics, and

3. Whose design characteristics are controlled so that the only potential means to effect a change that might result in a failure to function would be to implement a design change No failures in 1. Exceptionally robust passive engineered control (PEC), Rarely can be justified by 30 years for

2. Two independent active evidence. Further, most types of

-4 hundreds of engineered control (AECs), PECs, single control have been similar controls in or enhanced specific observed to fail.

industry administrative control (SAC)

No failures in 30 years for tens A single control with redundant

-3 None of similar controls parts, each a PEC or AEC in industry No failure of this

-2 type in the faclity A single PEC None in 30 years A few failures may 1. A single AEC

-1 occur during 2. Enhanced SAC None facility lifetime 3. Redundant SAC Failure occurs 0 None every 1 to 3 years A single SAC Several Frequent event, inadequate Not for controls, just initialing 1 occurrences per control events.

year Occurs every Very frequent event, inadequate Not for controls, just initialing 2 week or more control events.

often NE Medical Technologies 13a2.1-51 Rev. 4

Table 13a2.1 Failure Probability Index Numbers ailure Probability Probability of Failure Based on Type of Comments ex Number (FPIN) on Demand Control If initiating event, no

-6 10-6 N/A control needed.

1. Passive engineered control (PEC) with high design margin.

Can rarely be justified

2. Inherently safe by evidence. Most process.

-4 or -5 10 10-5 types of single controls

3. Two redundant controls more robust have been observed to than a simple AEC, fail.

PEC, or enhanced SAC.

1. Single PEC 10 10-4

-3 or -4 2. Single AEC with high None availability

1. Single AEC 10 10-3 2. Enhanced SAC

-2 or -3 None

3. SAC for routine planned operations A SAC that must be 10 10-2 performed in response 1- or -2 None to a rare unplanned demand.

NE Medical Technologies 13a2.1-52 Rev. 4

Table 13a2.1 Duration Index Numbers Duration Index Average Failure Duration in Years Comments Number (DIN) Duration 1 > 3 years 10 None 0 1 year 1 None Formal monitoring to

-1 1 month 0.1 justify indices < -1

-2 A few days 0.01 None

-3 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 10-3 None

-4 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 10-4 None

-5 5 minutes 10-5 None NE Medical Technologies 13a2.1-53 Rev. 4

Table 13a2.1 Consequence Category Definitions Consequence Category Facility Staff Off-Site Public RD > 25 rem RD > 100 rem High Consequence 3 30 milligrams sol U intake CD > PAC-3 CD > PAC-2 5 rem < RD 100 rem 1 rem < RD 25 rem Intermediate Consequence 2

PAC-2 < CD < PAC-3 PAC-1 < CD PAC-2 Accidents with lower Accidents with lower Low Consequence radiological and chemical radiological and chemical 1

exposures than those above exposures than those above NE Medical Technologies 13a2.1-54 Rev. 4

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.

iological consequences are determined for members of the public and workers (i.e., control m operators) and are provided in Table 13a3-1 and Table 13b.2-2. Radiological sequences to control room operators are determined for the duration of a postulated event, ounting for shift change outs, and demonstrate that SHINE Design Criterion 6 (Control room) et.

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 1 rem total effective dose equivalent (TEDE) for the duration of the accident, and

• Radiological consequences to workers do not exceed 5 rem TEDE during the accident.

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 Partitioning 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.

NE Medical Technologies 13a2.2-1 Rev. 4

ionuclide Transport Method transport of radioactive material for the accident analysis was quantitatively evaluated using ntrol volume and tracer method.

control volume method considers each part of the facility as a fixed volume that the material ee to disperse into. Dispersion within these volumes is assumed to be instantaneous. Each me is connected by one or more junctions which allows flow in one direction at a volumetric rate, either pressure-driven or constant. Counter-current flow, or flow back into the previous trol volume, is conservatively neglected. Flow from the irradiation facility (IF) to the oisotope production facility (RPF) or vice versa is not modeled; material present in the IF or F control volumes is assumed to exit the SHINE facility to the environment without further tion.

tracer method is a modeling tool that is representative of the kind of material being tracked.

ause an output of the tracer method is a fraction of material released, any material can be d as a tracer for any other kind of material as long as the tracers properties (density, molar ss, etc.) are used consistently throughout the scenario. For example, a gas may be used as a er for an airborne non-volatile because the output of the analysis is normalized to the amount aterial initially released. Therefore, the physical properties of the tracer itself are not ortant as long as they are applied consistently. In this calculation, iodine is used as a tracer odine, krypton for noble gases, xenon for non-volatiles, nitrogen or air for nitrogen, air for air, tritium for tritium.

each control volume, the amount and volume fraction of each tracer, in moles, is calculated g the density and molar weight of each tracer and the volume of the space. Material flowing of the control volume is calculated as the product of the volumetric flow rate multiplied by the me fraction of the tracer. Flow can be due to pressure-driven flow, barometric breathing, or a stant flow rate based on the design of the cell or glovebox.

w Between Control Volumes re are three types of flow between control volumes. The first is pressure-driven Poiseuille

, which is calculated using the following equation:

t flow t = p---------

k ere:

• flow is the volumetric flow rate at a given time (m3/s)

• p(t) is the pressure difference between the control volumes (Pa)

• k is the conversion factor from pressure to volumetric flow. This represents the tightness of the seal on the cell being pressurized (Pa-s/m3).

kind of flow is used in scenarios that model the pressurization of cells due to the nitrogen ge system (N2PS) actuation.

NE Medical Technologies 13a2.2-2 Rev. 4

design basis accident (DBA). These prescribed leak rates are some fraction of the volume hour, converted into a fraction of volume per second and multiplied by the volume of the cell onvert to m3/s.

third kind of flow is due to barometric breathing, which is the gas flow driven by cyclical nges in the atmospheric pressure. Barometric breathing is determined using meteorological a from the Southern Wisconsin Regional Airport. The barometric breathing rate is converted volume fraction per second which is multiplied by the volume of the cell to produce a metric flow rate in m3/s. This kind of flow is considered for all cells that are not pressurized do not have a designed leak rate.

ome cases, more than one flow type is modeled. For example, a transient may use a bination of pressure-driven flow and barometric breathing. This is due to the system initially g dominated by pressure-driven flow due to a gas release, but eventually achieves a ssure equilibrium between control volumes or between a control volume and the environment.

osition of Iodine and Non-Volatiles BA scenarios involving uranyl sulfate and leakage through the radiologically controlled area A), both iodine and non-volatile deposition surfaces are considered. Where possible, iodine non-volatile absorption coefficients are calculated using the following equation:

A

= d ---

V ere:

• is the absorption coefficient (s-1)

• vd is the settling velocity (m/s)

• A is the surface area that the radionuclide cloud encounters (m2)

• V is the volume of the gas (m3) the IF and RPF, the area that the gas encounters is only considered to be the floor area of IF or RPF. Because this eliminates the surface area of the walls and ceiling of the RCA, this conservative assumption. The free volume of the IF and RPF is used as the volume of the settling velocity is assumed to be 10-4 m/s, consistent with the dry conditions velocity for xy paint of 10-3 m/s. Desorption of the iodine back into the RCA or cell gas space is not sidered in this analysis.

iodine or non-volatile adsorption is modeled once the radionuclides exit the RCA.

e the absorption coefficient is determined, the removal of isotopes is calculated using the wing equation:

da = - a t dt ere a(t) is the moles of the corresponding tracer.

NE Medical Technologies 13a2.2-3 Rev. 4

ceptor activity fraction (RAF) represents the fraction of a tracer that is present in a control me at a specific time interval. The RAF for a control room operator at time bin j is therefore:

IA j O RAF CR j = ---------------j-a initial ere IAj is the integrated activity in the control room at time bin j, Oj is an occupancy factor for rators in the control room, and ainitial is the initial tracer moles released. The integrated vity is calculated for each time bin j using one second time steps as:

IA total t f = tt = 0 a t f

integrated moles for a given time bin j, defined by the initial time j1 and concluding time j2, is n calculated by the following equation:

IA j = IA total j 2 - IA total j 1 total receptor activity (RA) is then calculated by summing the products of IAj and Oj from the inning of the DBA to the end of the desired time period, dividing that value by the initial moles ased for that tracer, and multiplying the resultant RAF by the activities included in the narios MAR.

public RAF is calculated in a similar manner to the control room RAF, with the dispersion or (/Q)j replaces the control room occupancy factor.

IA j -----

Q j RAF p j = -------------------

a initial RAFp, j for a given time period is calculated by summing the calculated public RAFs from the inning of the DBA to the end of the desired time period and multiplying the summed RAFs by activities in the scenarios MAR.

determination of RAF to the public, 95th percentile site boundary time-dependent /Q values used. For the determination of RAF to the worker, 95th percentile control room time-endent /Q values are used. The maximum calculated value over all directions of the h percentile /Q was used for both receptor locations. The use of time-dependent /Q values onsistent with the methodology presented in Regulatory Guide 1.195, Methods and umptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-ter Nuclear Power Reactors (USNRC, 2003a). The environmental and meteorological ditions used to develop the atmospheric dispersion factors are discussed in Section 2.3.

ident Source Terms accident source terms for each accident scenario are consistent with the methodology as cribed in Section 3.2.5.2 of NUREG/CR-6410 (USNRC, 1998). However, the combined RAF described in the previous section accounts for the leak path factor, airborne release fraction, atmospheric dispersion factors. The cumulative RAF values represent the time-dependent age of radionuclides. The RAF values are calculated for the leakage from the source volume NE Medical Technologies 13a2.2-4 Rev. 4

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 factor (DCF) values for submersion for each radionuclide.

alation doses are calculated based on the respirable fraction, DCF for inhalation, and athing rate. The DCF values used in the analysis are taken from Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion tors (DCF) for Inhalation, Submersion, and Ingestion (EPA, 1988), and Federal Guidance ort No. 12, External Exposure to Radionuclides in Air, Water, and Soil (EPA, 1993).

rker dose was generally calculated over a 30-day interval. The scenario resulting in the ase of tritium from the tritium purification system (TPS) gloveboxes uses a 10-day release rval because it is expected that tritium recovery can be accomplished within this time frame.

rker dose also includes the control room occupancy factor used in the calculation of RAF.

rator action inside the facility is not required to stabilize accident conditions.

public dose was generally calculated over a 30-day interval at the site boundary. The nario resulting in the release of tritium from TPS gloveboxes uses a 10-day release interval ause it is expected that tritium recovery can be accomplished within this time frame. A ground ase was used as the release point.

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 NE Medical Technologies 13a2.2-5 Rev. 4

• Condensation was conservatively neglected in the radiation transport 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.

DCFs used in the analysis are well-recognized and are used without consideration of ertainty in the values.

ertainty in the /Q calculation was estimated by calculating the mean and standard deviation he 95th percentile values for the 16 sectors. The result of the estimation is +/- 25 percent.

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.

ARCON96 computer code was used to calculate the short-term atmospheric dispersion

) factors for an effluent release to the control room. ARCON96 is described in REG/CR-6331, Revision 1, Atmospheric Relative Concentrations in Building Wakes (USNRC, 7). The code was used as prescribed in Regulatory Guide 1.194, Atmospheric Relative centrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants NRC, 2003b). No additional validation was performed for the ARCON96 code.

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.

NE Medical Technologies 13a2.2-6 Rev. 4

described in Subsection 13a2.1.1, the postulated maximum hypothetical accident (MHA) for SHINE facility is a failure of the TSV off-gas system (TOGS) pressure boundary resulting in a ase of off-gas into the TOGS cell. A detailed description of this scenario and an evaluation of radiological consequences is provided in Subsection 13a2.2.7.

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.

2.2.2.3 Sequence of Events 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.

5. At [ ]PROP/ECI 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.

NE Medical Technologies 13a2.2-7 Rev. 4

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 2.2.2.4 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 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.

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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 path isolation due to inadvertent valve closure, loss of heat sink, or PCLS leakage due to a ng or component failure.

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 NE Medical Technologies 13a2.2-9 Rev. 4

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.

2.2.3.6 Radiological Consequences 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 DBA for mishandling or malfunction of target solution is a of the PSB integrity which results in a release of target solution into the IU cell. This scenario escribed 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.

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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 via the RVZ1e IU cell radiation monitors 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.

operator actions are taken or required to reach a stabilized condition or to mitigate dose sequences.

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 NE Medical Technologies 13a2.2-11 Rev. 4

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 RAF 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 urs, the iodine is dissolved in the water and partitioned according to the pH of the pool, the perature of the pool, and the pool and gas volumes. Deposition of iodine on the walls of the 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.

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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

• TOGS

• N2PS

• 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.

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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.

2.2.6.3 Sequence of Events 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 TPS target chamber supply pressure or high TPS target chamber exhaust 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.

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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

• TPS Train Isolation on high TPS target chamber supply pressure or high TPS target chamber exhaust 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.

r the initial IU cell pressurization has reached equilibrium, leakage between the IU cells and IF is driven primarily by barometric breathing. The leakage between the cells and the IF is not 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 RAF 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.

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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 via the RVZ1e IU cell radiation monitors 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 airborne radiation. A sufficient time delay is provided by the holdup volume in RVZ1e to prevent radioactive gases from exiting through RVZ1e prior to isolation.

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.

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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

• RVZ1e IU cell 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.

2.2.7.5 Radiation Source Terms 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.

RAF 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.

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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.

ety Controls

• 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.

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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.

3. TRPS detects the loss of flow and executes an IU Cell Safety Actuation and IU Cell Nitrogen Purge.

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.

NE Medical Technologies 13a2.2-19 Rev. 4

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 overflow lines to the TSV dump tank

• 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.

2.2.9.5 Radiological Source Terms 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.

NE Medical Technologies 13a2.2-20 Rev. 4

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.

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.

2. The water intrusion results in an exothermic uranium metal-water reaction, generating hydrogen.

3. The presence of [ ]PROP/ECI inhibits a potential deflagration.

4. Small amounts of hydrogen gas migrate into the PCLS and travel to the PCLS expansion tank, along with hydrogen normally generated in PCLS itself via radiolysis. The expansion 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 the RVZ1e IU cell radiation monitors 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).

sequences of events associated with unintended exothermic chemical reactions other than onation are further discussed in Subsection 13a2.1.10.2.

NE Medical Technologies 13a2.2-21 Rev. 4

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.

2.2.10.6 Radiological Consequences 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.

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.

NE Medical Technologies 13a2.2-22 Rev. 4

Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences Loss of Ventilation Postulated loss of ventilation scenarios do not result in radiological consequences based on the preventive controls described in Subsection 13a2.1.11.

Additional loss of ventilation scenarios are described in Subsection 13a2.2.5 and Subsection 13a2.2.6.

Safety Controls The safety controls credited for prevention of accidents which may cause radiological consequences 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 Spatial Interactions Fires Postulated fire scenarios in the IF are prevented by fire protection features identified in the fire hazards analysis (FHA), as described in Subsection 13a2.1.11.

Additional fire scenarios are discussed in Subsection 13a2.2.6 and Subsection 13a2.2.9.

Safety Controls The safety controls credited for prevention of accidents which may cause radiological consequences 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 Exothermic Chemical Reaction Exothermic chemical reaction scenarios are described in Subsection 13a2.1.10.

Internal Flooding Postulated internal flooding scenarios in the IF do not result in radiological consequences, as described in Subsection 13a2.1.11.

Dynamic Effects Dynamic effects are not present at the main production facility, as described in Subsection 13a2.1.11.

SHINE Medical Technologies 13a2.2-23 Rev. 4

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.

2.2.11.5 Radiation Source Terms 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 TPS. Three potential locations for the ase of tritium were analyzed to determine the dose consequences and necessary controls.

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 NE Medical Technologies 13a2.2-24 Rev. 4

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 TPS target chamber supply pressure or high TPS target chamber exhaust pressure actuates the primary confinement boundary isolation valves and an irradiation unit 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. 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)

• TPS Train Isolation on high TPS target chamber supply pressure or high TPS target chamber exhaust 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 RAF 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 Medical Technologies 13a2.2-25 Rev. 4

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.

2.2.12.2 Tritium Release into the Tritium Purification System Glovebox 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 seismic event that causes a break in two TPS trains and instantaneously releases the tritium inventory into their respective TPS gloveboxes.

2. For the first 20 seconds, tritium escapes from each of the gloveboxes to the IF through the glovebox pressure control exhaust process vent to RVZ1.

3. The glovebox ventilation shuts down after 20 seconds due to high TPS confinement A/B/C 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 high TPS confinement A/B/C 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.

NE Medical Technologies 13a2.2-26 Rev. 4

hange between each 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.

ety Controls safety controls credited for mitigation of this accident are:

• TPS room ventilation isolations

• Glovebox pressure control and VAC/ITS ventilation isolations

• TPS confinement A/B/C tritium monitors

• Tritium confinement boundary, as described in Section 6a2.2 ddition, TPS glovebox deflagration is prevented by:

• TPS glovebox gas space inerted with helium

• 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 200,000 curies of tritium from the TPS equipment in the TPS ebox.

accident source term development is discussed in Section 13a2.2. The RAF 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.

NE Medical Technologies 13a2.2-27 Rev. 4

Table 13a2.2 Summary of Radiation Transport Terms (Public)

Receptor Activity Fraction (RAF)

Nobles Iodine Non-volatiles Tritium Tritium cident Category (30-day) (30-day) (30-day) (10-day) (30-day) handling or Malfunction of Target Solution 1.30E-03 7.64E-05 1.16E-09 N/A N/A bsection 13a2.2.4) ernal Events (Subsection 13a2.2.6) N/A N/A N/A N/A 4.07E-04 handling or Malfunction of Equipment 1.41E-03 3.69E-04 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 4.07E-04

• Tritium Release into the Tritium Purification N/A N/A N/A 1.78E-04 N/A System Glovebox NE Medical Technologies 13a2.2-28 Rev. 4

Table 13a2.2 Summary of Radiation Transport Terms (Worker)

Receptor Activity Fraction (RAF) (30 days) cident Category Nobles Iodine Non-volatiles Tritium handling or Malfunction of Target Solution (Subsection 13a2.2.4) 8.55E-01 6.43E-02 7.45E-07 N/A ernal Events (Subsection 13a2.2.6) N/A N/A N/A 2.87E-01 handling or Malfunction of Equipment (Subsection 13a2.2.7) 9.92E-01 3.23E-01 0 N/A ility-Specific Events (Subsection 13a2.2.12)

• Tritium Release into an IU Cell N/A N/A N/A 2.87E-01

• Tritium Release into the Tritium Purification System Glovebox 1.08E-01 N/A N/A N/A (10 days)

NE Medical Technologies 13a2.2-29 Rev. 4

NE Medical Technologies 13a2.2-30 Rev. 4 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 1 rem 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. 3

Table 13a3 Irradiation Facility Accident Dose Consequences Public Worker Dose Dose TEDE TEDE cident Category (Bounding Scenario) (mrem) (mrem) 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 372 555 s of Off-Site Power (LOOP) (Subsection 13a2.2.5) No consequences ernal Events (Subsection 13a2.2.6) 292 588 handling or Malfunction of Equipment (Subsection 13a2.2.7) 727 1940 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 37 74

• Tritium Release into the Tritium Purification System Glove 798 1380 Box NE Medical Technologies 13a3-2 Rev. 3

A, 1988. Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion tors for Inhalation, Submersion, and Ingestion, Federal Guidance Report No. 11, U.S.

ironmental Protection Agency, 1988.

A, 1993. External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance ort No. 12, U.S. Environmental Protection Agency, 1993.

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, 1997. Atmospheric Relative Concentrations in Building Wakes, NUREG/CR-6331 ision 1, U.S. Nuclear Regulatory Commission, May 1997.

NRC, 1998. Nuclear Fuel Cycle Facility Accident Analysis Handbook, NUREG/CR-6410,

. Nuclear Regulatory Commission, March 1998.

NRC, 2003a. Methods and Assumptions for Evaluating Radiological Consequences of Design is Accidents at Light-Water Nuclear Power Reactors, Regulatory Guide 1.195, U.S. Nuclear ulatory Commission, May 2003.

NRC, 2003b. Atmospheric Relative Concentrations for Control Room Radiological Habitability essments at Nuclear Power Plants, Regulatory Guide 1.194, U.S. Nuclear Regulatory mmission, June 2003.

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.

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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.

NRC, 2015. Standard Review Plan for Fuel Cycle Facilities License Applications, REG-1520, Revision 2, U.S. Nuclear Regulatory Commission, June 2015.

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.

NL, 2011. ORIGEN-S: Depletion Module to Calculate Neutron Activation, Actinide nsmutation, Fission Product Generation, and Radiation Source Terms, Oak Ridge National oratory, Oak Ridge, Tennessee, 2011.

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.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 hazard identification and a process hazard lysis (PHA). The hazard identification process includes a review of potential radiological ards, chemical hazards, and other facility hazards that might be present.

cess that are conducted in the RPF fall into the following categories:

• 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 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 as an operation with hazardous chemicals. Because of the presence of uranium, the TSPS es a criticality, radiological, and chemical hazard.

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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 ansferred to the RLWS for disposal. The RDS involves operations with irradiated target tion processed for reuse or for waste disposal. Because of the presence of uranium, the RDS the potential for a criticality as well as radiological and chemical exposure hazards.

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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 also defined to result in bounding ological consequences for the SHINE facility.

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

• 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:

• MHA

• 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 sequences above the appropriate evaluation guidelines for worker or public exposure were n subject to control selection. Appropriate preventative or mitigative controls were identified to 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 NE Medical Technologies 13b.1-3 Rev. 3

.1.2.1 Maximum Hypothetical Accident MHA for the SHINE facility is identified in Subsection 13a2.1.1.

.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.

nario 1 - Spill of Target Solution in the Supercell (MEPS Column Misalignment) 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 (i.e., the RVZ1 supercell area 1-10 radiation monitors) that provide a signal to the ineered safety features actuation system (ESFAS) to isolate the affected cell and limit the ount of target solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped with bon filters, hot cell inlet (radiological ventilation zone 2 [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.

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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 (i.e., the RVZ1 supercell area 1-10 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. MEPS extraction pump akers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to t the amount of target solution introduced into the affected hot cell. This scenario is further cribed 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

, the RVZ1 supercell area 1-10 radiation monitors) that provide a signal to ESFAS to isolate affected cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and hot cell inlet Z2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant ation dampers. The resulting sequence of events for this scenario is analogous to the MEPS te spill described in Subsection 13b.2.4.2.

nario 4 - Spill of Target Solution in the Supercell (IXP Column Misalignment) 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 (i.e., the RVZ1 supercell area 1-10 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 outlet (RVZ1) ducts are equipped with carbon filters, hot cell inlet (RVZ2) and outlet Z1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers, 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 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 (i.e., the RVZ1 supercell a 1-10 radiation monitors) that provide a signal to ESFAS to isolate the affected cell and limit amount of target solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped carbon filters, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with FAS-controlled redundant isolation dampers, and ESFAS-controlled IXP extraction pump akers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to t the amount of target solution introduced into the affected hot cell. This scenario is further cribed 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 (i.e., the RVZ1 supercell a 1-10 radiation monitors) that provide a signal to ESFAS to isolate the affected cell and limit amount of target solution introduced into the cell, hot cell inlet (RVZ2) and outlet (RVZ1) tilation ducts are equipped with ESFAS-controlled redundant isolation dampers, hot cell et (RVZ1) ducts are equipped with carbon filters, and ESFAS-controlled IXP extraction pump akers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to t the amount of target solution introduced into the affected hot cell. This scenario is further cribed in Subsection 13b.2.4.1.

nario 7 - Spill of Iodine Solution in the Supercell (Overfill or Drop of Iodine Solution Bottle) 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

, the RVZ1 supercell area 1-10 radiation monitors) that provide a signal to ESFAS to isolate affected cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and hot cell inlet Z2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant ation dampers. The resulting sequence of events for this scenario is analogous to the MEPS te spill described in Subsection 13b.2.4.2.

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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 (i.e., the RVZ1 and RVZ2 RCA exhaust radiation monitors) that vide a signal to ESFAS to isolate the building ventilation supply and exhaust dampers on high ation. This scenario is further described in Subsection 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 (i.e., the RVZ1 and RVZ2 RCA exhaust radiation monitors) that provide a al to ESFAS to isolate the building ventilation supply and exhaust dampers on high radiation.

scenario is further described in Subsection 13b.2.4.4.

nario 11 - Spill of Target Solution in a Tank Vault (Hold Tank Deflagration) 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 (i.e., the RVZ1 and RVZ2 RCA exhaust radiation monitors) that provide a NE Medical Technologies 13b.1-7 Rev. 3

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 (i.e., the RVZ1 and RVZ2 RCA aust 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.2.4.4.

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 (i.e., the RVZ1 supercell area 1-10 radiation monitors) that vide a signal to ESFAS to isolate the affected cell, hot cell outlet (RVZ1) ducts are equipped carbon filters, and hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with FAS-controlled redundant isolation dampers. This scenario is further described in section 13b.2.4.2.

nario 14 - Target Solution Leaking out of the Supercell (MEPS Preheater Tube Leak) ak in the MEPS extraction column preheater allows target solution to enter the hot water

. Potential consequences of target solution leaking into the hot water loop, which is partially ted outside of the supercell, include radiological dose. To prevent the target solution from ulating through the water loop, conductivity instrumentation in the hot water loop detects et solution in-leakage and provides a signal to ESFAS to close the isolation valves on the ply and return of the hot water loop at the supercell boundary. This scenario was evaluated litatively and is not described 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, NE Medical Technologies 13b.1-8 Rev. 3

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 (i.e., the RVZ1 and RVZ2 RCA exhaust radiation nitors) that provide a signal to ESFAS to isolate the building ventilation supply and exhaust pers on high radiation. The resulting sequence of events for this scenario is analogous to the et solution leak in a pipe trench, which is further described 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 uence is prevented. The RLWI enclosure damage scenario is further described in section 13b.2.4.5.

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.

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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. Radiological consequences of criticality accidents not included in the accident analysis because preventative controls are used to ensure cality events are highly unlikely. Further discussion of the criticality safety bases for RPF cesses is included in Section 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.

nario 2 - PVVS Carbon Guard Bed Fire 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 NE Medical Technologies 13b.1-10 Rev. 3

NE Medical Technologies 13b.1-11 Rev. 3 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 rem 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 described in Subsection 13a2.1.1, the postulated maximum hypothetical accident (MHA) for SHINE facility is a failure of the target solution vessel (TSV) off-gas system (TOGS) pressure ndary resulting in a release of off-gas into the TOGS cell. A detailed description of this nario and an evaluation of the radiological consequences is provided in section 13a2.2.12.2.

.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.

.2.3 EXTERNAL EVENTS 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.

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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.

ating Event 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. RVZ1 supercell area 2/6/7 radiation monitors 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.

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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

• Radiological ventilation zone 1 (RVZ1) supercell area 2/6/7 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.

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 30 liters 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 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.

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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. RVZ1 supercell area 3/5/8/10 radiation monitors 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

• RVZ1 supercell area 3/5/8/10 radiation monitors

• Hot cell RVZ1 outlet carbon filters (radioiodine)

• Inlet (RVZ2) and outlet (RVZ1) ventilation isolation dampers

• ESFAS Supercell Isolation function 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.

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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.

4. A portion of the released material leaks through the process confinement boundary (trench 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 (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.

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initial MAR for this scenario is a batch of target solution from the IU at [ ]PROP/ECI PROP/ECI period removes more t-shutdown. The action of the TOGS during this [ ]

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 has been held for decay for [ ]PROP/ECI post-shutdown, which accounts for [

]PROP/ECI of hold time in the TSV dump tank and [ ]PROP/ECI of processing time.

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.

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• 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.

.2.4.5 Spill of Waste Solution in RLWI 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.

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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 present in the solution at shutdown. It is assumed that 35 percent of the post-shutdown ne inventory is released to the RLWI enclosure during the accident. Additionally, partitioning 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.

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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.

4. ESFAS isolates the affected carbon delay bed group using installed actuation valves.

Valve closure is assumed to occur within 30 seconds of detection for bounding 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 carbon monoxide detectors

• 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.

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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. 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.

.2.6.2 PVVS Carbon Guard Bed Fire 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.

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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 odine gas are used to describe the quantities of iodine in solution that move to the RPF.

moval of iodine by the TOGS zeolite beds are credited for all gases that are transported to the 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 (Sheet 1 of 2)

Receptor Activity cident Scenario Radionuclide Group Fraction (RAF) 1.00E-03 Public Nobles 6.56E-01 Worker 1.49E-06 Public ll of Target Solution in the Supercell Iodine 1.40E-03 Worker 1.38E-07 Public Non-Volatile 9.40E-05 Worker 1.04E-03 Public Nobles 6.96E-01 Worker 1.88E-06 Public ll of Eluate Solution in the Supercell Iodine 1.77E-03 Worker 1.52E-07 Public Non-Volatile 1.06E-04 Worker 1.27E-04 Public Nobles 7.71E-02 Worker ll of Target Solution in the RPF Pipe 2.49E-07 Public Iodine nch 2.26E-04 Worker 1.11E-08 Public Non-Volatile 6.71E-06 Worker 1.36E-04 Public Nobles 8.24E-02 Worker 1.61E-08 Public ll of Target Solution from a Tank Iodine 1.51E-05 Worker 1.18E-08 Public Non-Volatile 7.18E-06 Worker 5.66E-03 Public Nobles 6.69E+00 Worker 5.66E-03 Public ll of Waste Solution in RLWI Iodine 6.69E+00 Worker 1.13E-06 Public Non-Volatile 1.34E-03 Worker NE Medical Technologies 13b.2-12 Rev. 2

Receptor Activity cident Scenario Radionuclide Group Fraction (RAF) 5.66E-03 Public VS Carbon Guard Bed Fire Iodine 6.70E+00 Worker 1.50E-04 Public VS Carbon Delay Bed Fire Nobles 1.63E-01 Worker NE Medical Technologies 13b.2-13 Rev. 2

Table 13b.2 Radioisotope Production Facility Accident Dose Consequences Public Dose Worker Dose TEDE TEDE cident Scenario (mrem) (mrem) ll of Target Solution in the Supercell 42 76 ll of Eluate Solution in the Supercell 88 122 ll of Target Solution in the RPF Pipe Trench 22 40 ll of Target Solution from a Tank 24 42 ll of Waste Solution in RLWI 557 1880 VS Carbon Delay Bed Fire 532 40 VS Carbon Guard Bed Fire 546 1390 NE Medical Technologies 13b.2-14 Rev. 2

NE has evaluated the potential hazards of chemicals within the main production facility.

se include chemicals that are licensed materials or have licensed materials as precursor pounds, or substances that physically or chemically interact with licensed materials and that toxic, explosive, flammable, corrosive, or reactive to the extent that they endanger life or lth. These include substances that are comingled with licensed material or are produced by a ction with licensed material. These do not include substances prior to process addition to nsed materials or after process separation from licensed materials. The analysis is therefore nding for all hazardous chemicals produced from or comingled with licensed materials.

hazardous chemical consequence assessment is performed to demonstrate that potential sequences meet the SHINE Safety Criteria, as defined in Section 3.1, for the public and kers (i.e., a radiologically controlled area [RCA] worker and a control room operator). The ntory of in-process hazardous chemicals used at the SHINE facility, compiled by process tion and quantity, is provided in Table 13b.3-1.

mical Process Descriptions chemical processes used in the SHINE facility are described in Sections 4b.3, 4b.4, 9a2.2, 9b.7.

mical Accidents Description and Source Term Determination each of the hazardous chemicals identified in Table 13b.3-1, a release scenario is tulated. Each postulated scenario defines the material at risk (MAR) as the largest quantity sent in a single vessel or process location. The MAR may therefore be less than the ximum quantities identified in Table 13b.3-1 (e.g., the total waste stream may be subdivided multiple tanks). The chemical source term is then evaluated using the following hodology.

formula for determining the source term (ST), the amount of hazardous material made orne and respirable, of each chemical release is given by the following formula:

ST = MAR ARF RF DR LPF ere:

• MAR is the material at risk, the quantity of material potentially affected;

• ARF is the airborne release fraction;

• RF is the respiratory fraction;

• DR is the damage ratio, the portion of the MAR affected by the release scenario (conservatively assumed to be 1.0 for all scenarios); and

• LPF is the leak path factor, the proportion of airborne material that leaks out of a building or enclosure. A leak path factor of 0.1 is applied for scenarios that occur in confinements (i.e., supercell, gloveboxes, subgrade vaults) to model the confinement barrier for the spill locations. This represents a 10 percent vol/hr leak rate from confinements. This conservatively bounds leak rates determined through more detailed analyses in the radiological dose analyses for these confinements.

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1) Non-volatile chemicals (e.g., solids, liquids with low vapor pressures), and

2) Volatile chemicals (i.e., liquids with vapor pressures in excess of 10 Torr at 100°F).

non-volatile chemicals, the MAR is taken to be the largest quantity of the chemical present in ngle vessel or process location. Values for the ARF and RF are taken from the guidance in REG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook (USNRC, 1998).

volatile liquids, the MAR x ARF x RF product is replaced by the total mass released as ulated by the ALOHA (Areal Locations of Hazardous Atmospheres) computer code, sion 5.4.7.

account for uncertainty in the MAR quantities, a multiplier of 1.2 is applied to the calculated rce term.

MAR and source terms for each chemical release scenario are presented in Table 13b.3-2.

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 Protective Action eria (PAC) values.

onsequence analysis for the public and nearest residence was performed using the PAVAN Atmospheric Dispersion Program for Evaluating Design-Basis Accidental Releases of ioactive Materials from Nuclear Power Stations) computer code (USNRC, 1982). The mical exposure to the public and nearest residence is then calculated using the h percentile atmospheric dispersion factors (/Q) calculated using the PAVAN computer e.

model the chemical exposure to the worker, the source term is used to determine the amount ach chemical released into the facility atmosphere. For the RCA worker, the total source term ased into the facility is assumed to be well mixed within the building free volume (i.e.,

diation facility [IF] or radioisotope production facility [RPF]) to determine a chemical centration. For the control room operator, the same concentration is assumed to be released the facility roll-up door and is transported to the facility ventilation intake that services the trol room. ALOHA is then used to calculate the indoor concentration at the location of the tilation intake louver that services the control room.

ntitative exposure standards are selected to meet acceptable limits for public and worker lth and safety. The quantitative acceptance limits are taken from the PAC values (USDOE, 8), which correspond to the Acute Exposure Guideline Levels (AEGLs), Emergency ponse Planning Guidelines (ERPGs), or Temporary Emergency Exposure Limits (TEELs) es for such chemicals. Exceptions are applied to rhodium chloride, uranyl sulfate, and uranyl oxide, which do not have published PAC values. For these chemicals, acceptance limits were eloped using guidance from DOE-HDBK-1046-2016, Temporary Emergency Exposure Limits Chemicals: Method and Practice (USDOE, 2016).

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• Sulfuric acid: A spill from a subgrade liquid waste collection tank may potentially exceed the control room chemical consequence limit. The subgrade vault is credited as a safety-related control to limit the source term to maintain the peak control room concentration to less than the PAC-2 limit.

• Uranium oxide: A seismic event resulting in the failure or overturning of the uranium receipt and storage system (URSS) uranium oxide storage rack, causing multiple storage can failures. The uranium storage racks are seismically qualified to maintain their structure and position during a seismic event, which prevents the potential chemical exposure. The failure of a single can during transfer or handling operations does not result in chemical dose consequences which exceed acceptance limits.

• Uranium oxide: A spill of uranium oxide powder in the URSS glovebox or target solution preparation system (TSPS) glovebox causes a quantity of the powder to become airborne. The gloveboxes are seismically qualified to maintain their low leakage boundary during a seismic event, which limits the chemical exposure to workers to within acceptable limits.

acceptance limits established for chemical consequence are that the PAC-1 limit shall not be eeded for members of the public, and that PAC-2 limits shall not be exceeded for workers.

results in Table 13b.3-2 show that no chemical consequence exceeds PAC-1 limits at the boundary or the nearest residence, and no chemical consequence exceeds PAC-2 limits for worker.

mical Process Safety 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.

components credited for mitigation of the chemical dose consequences are:

• Confinement barriers (i.e., supercell, gloveboxes, subgrade vaults) are credited for those chemical spill scenarios that occur within a confinement structure as identified in Table 13b.3-2.

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Table 13b.3 Quantities of In-Process Hazardous Chemicals (Sheet 1 of 3)

Chemical Name Location Inventory (kg)

Subgrade Waste Tanks 4.48E-01 Alpha-Benzoin Oxime Supercell 1.00E-02 RLWI Tank 3.28E-03 Ammonium Hydroxide Supercell 2.17E-02 Subgrade Waste Tanks 1.81E+01 Ammonium Nitrate Supercell 4.96E-02 RLWI Tanks 1.32E-01 TSPS Room 6.20E-01 Target Solution Storage 2.79E+00

[ ]PROP/ECI Subgrade Waste Tanks 1.86E-01 Supercell 3.10E-01 RLWI Tank 2.70E-03 Subgrade Waste Tanks 1.92E-02 Hydrochloric Acid (38 wt.%) Supercell 1.00E-02 RLWI Tank 1.41E-04 TSPS Room 2.66E+00 Hydrogen Peroxide (30 wt.%)

Supercell 1.00E-02 Subgrade Waste Tanks 1.79E+01 Nitric Acid Supercell 9.49E-01 (70 wt.% in chemical storage)

RLWI Tank 1.31E-01 Subgrade Waste Tanks 7.47E-03 Potassium Supercell 1.00E-02 Hexachlororuthenate RLWI Tank 5.46E-05 NE Medical Technologies 13b.3-4 Rev. 1

Chemical Name Location Inventory (kg)

Subgrade Waste Tanks 4.73E-01 Potassium Permanganate Supercell 1.00E-02 RLWI Tank 3.46E-03 Subgrade Waste Tanks 8.96E-03 Rhodium Chloride Supercell 1.00E-02 RLWI Tank 6.55E-05 Silver Nitrate Supercell 1.00E-02 Sodium Hydroxide (50.5 wt.%) Supercell 5.17E-01 Sodium Iodide Supercell 1.00E-02 Subgrade Waste Tanks 3.12E+00 Sodium Sulfite (98 wt. %) Supercell 1.00E-01 RLWI Tank 2.28E-02 TSPS Room 9.67E+00 Target Solution Storage 5.46E+01 Sulfuric Acid Subgrade Waste Tanks 5.62E+02 Supercell 6.07E+00 RLWI Tank 4.49E+00 Uranium Metal URSS Room 6.20E+02 URSS Room 7.31E+02 Uranium Oxide TSPS Room 8.60E+00 Uranyl Peroxide TSPS Room 1.15E+01 NE Medical Technologies 13b.3-5 Rev. 1

Chemical Name Location Inventory (kg)

TSPS Room 1.91E+02 Target Solution Storage 8.58E+02 Uranyl Sulfate Subgrade Waste Tanks 7.26E+01 Supercell 9.54E+01 RLWI Tank 1.05E+00 NE Medical Technologies 13b.3-6 Rev. 1

Table 13b.3 Hazardous Chemical Source Terms and Concentration Levels (Sheet 1 of 2)

Control Room Hazardous Operator/ Site Boundary Nearest Chemical RCA Worker Concentration Residence (Release MAR Source Term PAC-1(a) PAC-2(a) PAC-3(a) Concentration (230 m) (788 m)

Location) (kg) (mg) (mg/m3) (mg/m3) (mg/m3) (mg/m3) (mg/m3) (mg/m3) lpha-Benzoin Oxime 0.0688 1.38 0.49 5.4 32 1.30E-03/7.68E-05 1.30E-05 8.50E-07 bgrade Waste Tanks) mmonium Hydroxide 0.1(b) 2490 13 140 840 1.53E+00/1.39E-01 2.89E-02 1.89E-03 (Supercell)

Ammonium Nitrate 2.77 55.37 6.7 73 440 5.23E-02/3.09E-03 5.22E-04 3.42E-05 bgrade Waste Tanks)

]PROP/ECI 0.744 0.76(c)/74.4(c) [ ]PROP/ECI [ ]PROP/ECI [ ]PROP/ECI 1.26E-02/3.15E-03 3.57E-05 2.34E-06 (TSPS Room)

Hydrochloric Acid 0.038(b) 1380 2.7 33 150 7.14E-01/7.71E-02 1.90E-02 1.24E-03 (Supercell)

Hydrogen Peroxide 3.2 1380 14 70 140 1.69E-01/7.71E-02 2.24E-03 1.47E-04 (TSPS Room)

Nitric Acid 2.7(d) 4820 0.41 62 240 2.55E+00/2.69E-01 7.91E-03 5.19E-04 ubgrade Waste Tank)

Potassium exachloro-ruthenate 0.012 0.24 0.5 2 20 2.27E-04/1.34E-05 2.26E-06 1.48E-07 (Supercell) assium Permanganate 0.0727 1.45 8.6 14 150 1.37E-03/8.12E-05 1.37E-05 8.99E-07 bgrade Waste Tanks)

Rhodium Chloride(e) 0.012 0.24 1.68 18.5 110 2.27E-04/1.34E-05 2.26E-06 1.48E-07 (Supercell)

Silver Nitrate 0.012 0.24 0.05 0.9 5 2.27E-04/1.34E-05 2.26E-06 1.48E-07 (Supercell)

Sodium Hydroxide 0.620 12.4 0.5 5 50 1.17E-02/6.93E-04 1.17E-04 7.67E-06 (Supercell)

Sodium Iodide 0.012 0.24 13 140 860 2.27E-04/1.34E-05 2.26E-06 1.48E-07 (Supercell)

NE Medical Technologies 13b.3-7 Rev. 1

Control Room Hazardous Operator/ Site Boundary Nearest Chemical RCA Worker Concentration Residence (Release MAR Source Term PAC-1(a) PAC-2(a) PAC-3(a) Concentration (230 m) (788 m)

Location) (kg) (mg) (mg/m3) (mg/m3) (mg/m3) (mg/m3) (mg/m3) (mg/m3)

Sodium Sulfite 0.478 9.55 11 120 710 9.02E-03/5.33E-04 9.01E-05 5.91E-06 bgrade Waste Tanks)

Sulfuric Acid 78.0 1560 0.2 8.7 160 1.47E+00/8.71E-02 1.47E-02 9.65E-04 bgrade Waste Tanks)

Uranium Metal(f) 7.8 0 0.6 5 30 0.00E+00/0.00E+00 0.00E+00 0.00E+00 (URSS Room)

Uranium Oxide 40.0 2400 0.68 10 30 2.27E+00/9.99E+00 2.26E-02 1.48E-03 (URSS Room)

Uranyl Peroxide(g) 6.84 1368 0.94 10.4 62 1.29E+00/5.70E+00 1.29E-02 8.46E-04 (TSPS Room)

Uranyl Sulfate(g) 191.2 235(c)/19120(c) 0.92 10.2 61 3.91E+00/1.07E+00 1.11E-02 7.25E-04 (TSPS Room)

a. Protective Action Criteria (PAC) values are based on the U.S. Department of Energys Protective Action Criteria Database (USDOE, 2018), unless otherwise specified.

b. MAR increased to the minimum mass that ALOHA can model for a puddle release.

c. The first source term value listed is for a two-minute release, while the second source term value corresponds to a full tank release. For each receptor, the source term value which yields the most conservative result is used.

d. Based on largest capacity subgrade waste tank.

e. PAC values were not identified for rhodium chloride in the PAC Database (USDOE, 2018). PAC values were developed from toxicity information found on the safety data sheet using the methodology from DOE-HDBK-1046-2016 (USDOE, 2016).

. Uranium metal is stored as solid pieces. Therefore, there is no hazard from dropping solid metal pieces.

g. PAC values were not identified for uranyl peroxide or uranyl sulfate in the PAC Database (USDOE, 2018). For uranium compounds, ACGIH STEL is 0.6 mg/m3, which is multiplied by a compound adjustment factor based on the methodology from DOE-HDBK-1046-2016 (USDOE, 2016) to obtain the TEEL-1 (PAC-1) value. PAC-2 and PAC-3 values were calculated based on the methodology from DOE-HDBK-1046-2016.

NE Medical Technologies 13b.3-8 Rev. 1

DOE, 2016. Temporary Emergency Exposure Limits for Chemicals: Methods and Practice, E-HDBK-1046-2016, U.S. Department of Energy, December 2016.

DOE, 2018. Chemicals of Concern and Associated Chemical Information, Protective Action eria (PAC) Tables Rev. 29a, U.S. Department of Energy, June 2018.

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, 1998. Nuclear Fuel Cycle Facility Accident Analysis Handbook, NUREG/CR-6410,

. Nuclear Regulatory Commission, March 1998.

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.

NE Medical Technologies 13b.4-1 Rev. 1