ML19331A659

Shine Medical Technologies, LLC Supplement 1 to Final Safety Analysis Report, Change Summary

ML19331A659
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Site:	SHINE Medical Technologies, 99902034
Issue date:	11/14/2019
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ENCLOSURE 2 SHINE MEDICAL TECHNOLOGIES, LLC SHINE MEDICAL TECHNOLOGIES, LLC OPERATING LICENSE APPLICATION SUPPLEMENT NO. 1 FINAL SAFETY ANALYSIS REPORT CHANGE

SUMMARY

PUBLIC VERSION Summary Description of Changes FSAR Impacts Figure 1.3-1 Identify Figure 1.3-1 as security-related information in accordance with Regulatory Issue Summary (RIS) 2005-31.

Section 4a2.5 Update the irradiation unit (IU) cell cover plug thickness to reflect the current design. The change in the cover plug thickness does not impact the structural, seismic, or radiological analyses as described in the Final Safety Analysis Report (FSAR).

Section 9b.7 Relabel "boron injection" line to "reagent addition" line in Figure 9b.7-4 Figure 9b.7-4, and include description of reagent addition line to reflect the current design. Boron injection is not required and is not described in the FSAR.

Table 3.1-1 Remove process vessel vent system (PVVS) function of mitigating Section 7.5 hydrogen concentration in the radioactive liquid waste immobilization Table 7.5-2 (RLWI) system to reflect the current design. Conforming changes Figure 7.5-1 include the removal of the nitrogen purge system (N2PS) interface Section 9b.6 with the RLWI system, and the addition of PVVS isolation at the Section 9b.7 RLWI interface. RLWI hydrogen concentration levels remain below Table 9b.7-1 the lower flammability limit without mitigation.

Figure 9b.7-1 Additionally, the RLWI interface with the facility nitrogen handling system (FNHS) was removed from Table 9b.7-1 to reflect the current design. The FNHS interface is not required and not described in the FSAR.

Page 1 of 3

Summary Description of Changes FSAR Impacts Figure 6b.1-1 Update PVVS carbon guard bed and carbon delay bed configuration Section 6b.2 to reflect the current design. The primary configuration changes Section 7.5 include updates to the carbon guard bed bypass, configuring carbon Table 7.5-1 delay beds within groups, and updates to the carbon guard bed and Table 7.5-2 carbon delay bed isolation functions. This system configuration Figure 7.5-1 change reduces engineered safety feature actuation system Figure 7.5-3 (deleted)

(ESFAS) interfaces with PVVS while maintaining process functions Section 9a2.3 which support prevention or mitigation of potential accident Section 9b.6 scenarios, as described in Chapter 13.

Table 9b.6-1 Additionally, Figure 9b.6-1 was simplified to remove Figure 9b.6-1 nonsafety-related instrumentation and controls, and to remove Section 13b.1 duplication of process flow provided in Figures 4b.4-2 and 4b.4-3. Section 13b.2 Section 7.5 Provide clarification for ESFAS single failure design criteria discussion. Clarification only.

Section 6b.3 Consolidate ESFAS instrumentation supporting Vacuum Transfer Section 7.5 System (VTS) Safety Actuation to the VTS vacuum header liquid Table 7.5-1 detection instruments. Additionally, the VTS vacuum headers are Figure 7.5-1 divided based on the liquids being transferred. These changes Figure 7.5-3 reflect the current design. The changes reduce ESFAS interfaces (previously Figure 7.5-4) with VTS while maintaining process functions which support Section 9b.2 prevention or mitigation of potential accident scenarios, as described Figure 9b.2-1 in Chapter 13.

Section 6b.2 Update carbon filters on the radiological ventilation zone 1 (RVZ1)

Section 9a2.1 supercell exhaust from nonsafety-related to safety-related to reflect Section 13b.1 the current design. The carbon filters are credited with a reduction in Section 13b.2 radioiodine from the supercell exhaust prior to RVZ1 isolation.

Section 4b.1 Update the uranium receipt and storage system (URSS) process Section 4b.4 flow, system description, and criticality safety discussion to reflect a Table 4b.4-5 revised nuclear criticality safety evaluation (NCSE) for URSS and Table 4b.4-6 the current design. The revised Criticality Safety Basis in Section 6b.3 Subsection 6b.3.2.7 supports this change.

Figure 6b.3-6 Section 13a2.2 Update the accident duration from 10 days to 30 days for several Table 13a2.2-1 tritium release scenarios. The 30-day accident duration is more conservative as it provides additional margin to accident recovery.

The accident duration for the scenario associated with release of tritium from the TPS glovebox remains unchanged because it is expected that tritium recovery can be accomplished within 10 days for this scenario.

Section 6a2.2 Update the glovebox stripper system (GBSS) isolation time from Section 13a2.2 10 seconds to 20 seconds to reflect the current design. The 20 second isolation time is more conservative.

Page 2 of 3

Summary Description of Changes FSAR Impacts Section 5a2.2 Update accident analysis to credit RVZ1 holdup volume within Table 5a2.2-2 ventilation exhaust of the primary closed loop cooling system (PCLS)

Figure 5a2.2-1 expansion tank to reflect the current design. While the holdup Section 5a2.7 volume had previously been credited in some scenarios, this change Section 9a2.1 expands the applicability to additional scenarios and describes the Figure 9a2.1-3 system configuration which supports the holdup volume. The system Section 13a2.2 configuration is described in the revisions to Chapter 5 and Chapter 9, and the applicability to the accident analysis is described in the revision to Chapter 13.

Figure 4a2.2-2 Remove discussion of the facility compressed air system (FCAS) to Table 4a2.8-2 reflect the current design. FCAS is not required. The previously Table 5a2.2-3 described functions of the FCAS are replaced by facility nitrogen handling system (FNHS).

Section 3.1 Remove ambiguous language associated with safety classifications Table 3.1-3 of systems, structures, and components throughout the FSAR.

Section 3.4 Clarification only.

Section 4a2.7 Section 9a2.3 Section 13b.2 Update accident scenario involving a spill of target solution in the radioisotope production facility (RPF) pipe trench to reflect the current design. The scenario is updated to remove radioactive drains system (RDS) liquid detection as a credited control (including a conforming update of the radiation source term discussion). RDS liquid detection is not credited to support prevention or mitigation of this accident scenario, as described in Chapter 13.

Figure 6b.3-4 Correct Figure 6b.3-4 to reflect the current design (i.e., target solution preparation system (TSPS) demister vented by RVZ1),

consistent with process flow diagram provided in Figure 4b.4-3.

Section 13a2.1 Correct the defense in depth measures described in Subsection 13a2.1.4, Scenario 5, to reflect the current design, consistent with the process flow diagram provided in Figure 9b.2-1.

Section 4b.4 Update the isolation function description of the radioisotope process Table 5a2.3-3 facility cooling system (RPCS) supplying the TSPS reflux condenser Section 7.5 to reflect the current design. The update supports compliance with Table 7.5-2 the ESFAS single failure criteria described in Subsection 7.5.2.4.

Figure 7.5-1 Table 6a2.1-2 Update the airborne release fraction (ARF) x leak path factor (LPF)

Table 6b.1-2 tables and resulting dose consequence tables to reflect the current Table 13a2.2-1 design. The updates to these tables are supported by the FSAR Table 13a2.2-2 changes described throughout this change summary and by the Table 13a3-1 Integrated Safety Analysis (ISA) Summary revision.

Table 13b.2-2 A markup of the FSAR changes is provided as Attachment 1.

Page 3 of 3

ENCLOSURE 2 ATTACHMENT 1 SHINE MEDICAL TECHNOLOGIES, LLC SHINE MEDICAL TECHNOLOGIES, LLC OPERATING LICENSE APPLICATION SUPPLEMENT NO. 1 FINAL SAFETY ANALYSIS REPORT CHANGE

SUMMARY

PUBLIC VERSION FINAL SAFETY ANALYSIS REPORT MARKUP 187 pages follow

Security-Related Information - Withheld under 10 CFR 2.390(d)

Chapter 1 - The Facility General Description of the Facility Figure 1.3 Production Facility Building General Arrangement SHINE Medical Technologies 1.3-6 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Design Criteria CHAPTER 3 - DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 DESIGN CRITERIA Structures, systems, and components (SSCs) present in the SHINE facility are identified in Tables 3.1-1 and 3.1-2, including the applicable FSAR section(s) which describe each SSC and the applicable SHINE design criteria. Design criteria derived from external codes, guides, and standards specific to the design, construction, or inspection of SSCs are included in the applicable FSAR section describing those SSCs. For each SSC, the FSAR section identifies location, function, modes of operation, and type of actuation for specific SSCs, as applicable.

Nuclear Safety Functional Classification and Relative Importance to Safety Safety-related SSCs at SHINE are those physical SSCs whose intended functions are to prevent accidents that could cause undue risk to health and safety of workers and the public; and to control or mitigate the consequences of such accidents.

Acceptable risk is achieved by ensuring that events are highly unlikely or by reducing consequences less than the SHINE safety criteria. The SHINE safety criteria are:

• An acute worker dose of five rem or greater total effective dose equivalent (TEDE).

• An acute dose of 500 millirem or greater TEDE to any individual located outside the owner controlled area.

• An intake of 30 milligrams or greater of uranium in a soluble form by any individual located outside the owner controlled area.

• An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that could lead to irreversible or other serious, long-lasting health effects to a worker or could cause mild transient health effects to any individual located outside the owner controlled area.

• Criticality in the radioisotope production facility (RPF).

• Loss of capability to reach safe shutdown conditions.

Some SSCs are nonsafety-related but perform functions that impact safety-related SSCs. These nonsafety-related SSCs have design basis requirements necessary to prevent unfavorable interactions with safety-related SSCs due to failure of the nonsafety-related SSCs.

Safety-related SSCs are identified in Table 3.1-1 and nonsafety-related SSCs are identified in Table 3.1-2.

SHINE Design Criteria The SHINE facility uses design criteria to ensure that the SSCs within the facility demonstrate adequate protection against the hazards present. The design criteria are selected to cover:

• The complete range of irradiation facility and radioisotope production facility operating conditions.

• The response of SSCs to anticipated transients and potential accidents.

• Design features for safety-related SSCs including redundancy, environmental qualification, and seismic qualification.

• Inspection, testing, and maintenance of safety-related SSCs.

SHINE Medical Technologies 3.1-1 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Design Criteria Table 3.1 Safety-Related Structures, Systems, and Components (Sheet 1 of 2)

Structure, System, or Component (SSC) Acronym Section Applicable Design Criteria Engineered safety features 7.1.3 ESFAS 13-19, 37-39 actuation system 7.5 Facility structure FSTR 3.4.2 2, 6 4a2.1 Irradiation cell biological shield ICBS 29-36 4a2.5 Iodine and xenon purification 4b.1.3 IXP 9, 33, 36-37, 39 and packaging 4b.3.1 4a2.1 Light water pool system LWPS 25, 29-32, 36 4a2.4.2 Molybdenum extraction and 4b.1.3 MEPS 27, 33, 36, 37, 39 purification system 4b.3 Normal electrical power supply NPSS 8a2.1 27, 28 system 4a2.1 Neutron flux detection system NFDS 7.1.7 13-19 7.8 6b.2.3 Nitrogen purge system N2PS 39 9b.6.2 Primary closed loop cooling 4a2.1 PCLS 9, 12, 21, 29, 33 system 5a2.2 4b.1.3 Process vessel vent system PVVS 35, 39 9b.6.1 Production facility biological PFBS 4b.2 29-32, 36 shield Radioactive drain system RDS 9b.7.6 36, 37 Radioactive liquid waste RLWI 9b.7.3 35-398 immobilization Radioactive liquid waste 4b.1.3 RLWS 35-36, 38-39 storage 9b.7.4 RVZ1 Radiological ventilation RVZ2 9a2.1 29, 30, 32-36 zones 1, 2, and 3 RVZ3 4a2.1 Subcritical assembly system SCAS 9-11, 20, 22-25, 29-34, 36, 39 4a2.2 SHINE Medical Technologies 3.1-3 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Design Criteria Table 3.1 SHINE Design Criteria (Sheet 1 of 11)

Generally-Applicable Design Criteria Criterion 1 - Quality standards and records Safety-related structures, systems, and components (SSCs) important to safety are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they are identified and evaluated to determine their applicability, adequacy, and sufficiency and are supplemented or modified as necessary to ensure a quality product in keeping with the required safety function.

A quality assurance program is established and implemented in order to provide adequate assurance that these SSCs satisfactorily perform their safety functions.

Appropriate records of the design, fabrication, erection and testing of safety-related SSCs important to safety are maintained by or under the control of SHINE throughout the life of the facility.

Criterion 2 - Natural phenomena hazards The facility structure supports and protects safety-related SSCs and is designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches as necessary to prevent the loss of capability of safety-related SSCs to perform their safety functions.

Safety-related SSCs are designed to withstand the effects of earthquakes without loss of capability to perform their safety functions.

Criterion 3 - Fire protection Safety-related SSCs are designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials are used wherever practical throughout the facility, particularly in locations such as confinement boundaries and the control room.

Fire detection and suppression systems of appropriate capacity and capability are provided and designed to minimize the adverse effects of fires on safety-related SSCs. Firefighting systems are designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these SSCs.

SHINE Medical Technologies 3.1-7 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Design Criteria Table 3.1 SHINE Design Criteria (Sheet 7 of 11)

Criterion 24 - Inspection of primary system boundary The primary system boundary design includes provisions for in-service inspection to ensure structural and leak tight integrity, and an appropriate material surveillance program for the primary system boundary.

Criterion 25 - Residual heat removal The light water pool is provided to remove residual heat. The system safety function is to transfer fission product decay heat and other residual heat from the target solution vessel dump tank at a rate such that target solution design limits and the primary system boundary design limits are not exceeded.

Criterion 26 - Cooling water The radioisotope process facility cooling system and process chilled water system are provided to transfer heat from structures, systems, and componentssafety-related SSCs important to safety to the environment, which serves as the ultimate heat sink.

SHINE Medical Technologies 3.1-13 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Design Criteria Table 3.1 SHINE Design Criteria (Sheet 8 of 11)

Electric Power Systems Design Criteria Criterion 27 - Electric power systems An on-site electric power system and an off-site electric power system are provided to permit functioning of structures, systems, and components important to safetysafety-related SSCs. The safety functions are to provide sufficient capacity and capability to assure that:

1) target solution design limits and primary system boundary design limits are not exceeded as a result of anticipated transients, and

2) confinement integrity and other vital functions are maintained in the event of postulated accidents.

The on-site uninterruptible electric power supply and distribution system has sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure.

Provisions are included to minimize the probability of losing electric power from the uninterruptible power supply as a result of or coincident with, the loss of power from the off-site electric power system.

Criterion 28 - Inspection and testing of electric power systems The safety-related electric power systems important to safety are designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems are designed with a capability to test periodically:

1) the operability and functional performance of the components of the systems, such as on-site power sources, relays, switches, and buses; and

2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the on-site and off-site power supplies.

SHINE Medical Technologies 3.1-14 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Design Criteria Table 3.1 SHINE Design Criteria (Sheet 11 of 11)

Criterion 36 -Target solution storage and handling and radioactivity control The target solution storage and handling, radioactive waste, and other systems that contain radioactivity are designed to assure adequate safety under normal and postulated accident conditions. These systems are designed with:

1) capability to permit appropriate periodic inspection and testing of safety-related components important to safety,

2) suitable shielding for radiation protection,

3) appropriate confinement and filtering systems, and

4) residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal.

Criterion 37 - Criticality control in the radioisotope production facility Criticality in the radioisotope production facility is prevented by physical systems or processes and the use of administrative controls.

Use of geometrically safe configurations is preferred. Control of criticality adheres to the double contingency principle.

A criticality accident alarm system to detect and alert facility personnel of an inadvertent criticality is provided.

Criterion 38 - Monitoring radioactivity releases Means are provided for monitoring the primary confinement boundary, hot cell, and glovebox atmospheres to detect potential leakage of gaseous or other airborne radioactive material. Potential effluent discharge paths and the plant environs are monitored for radioactivity that may be released from normal operations, including anticipated transients, and from postulated accidents.

Criterion 39 - Hydrogen mitigation Systems to control the buildup of hydrogen that is released into the primary system boundary and tanks or other volumes that contain fission products and produce significant quantities of hydrogen are provided to ensure that the integrity of the system and confinement boundaries are maintained.

SHINE Medical Technologies 3.1-17 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage to accidental eccentricity. The torsional moment is taken equal to the story shear at the elevation and in the direction of interest times a moment arm equal to 5 percent of the building dimension.

The torsional moment is distributed to the building shear walls based on the relative rigidity of the walls in plane. The loads are applied statically and account for variability in the load direction.

3.4.2.6.5 Structural Analysis Model A three-dimensional finite element model of the SHINE facility structure was created using the computer program SAP2000 (version 17.2) to represent the mass and stiffness of the major structural elements, equipment, and components of the FSTR. The model utilizes shell elements to represent slabs and walls, and frame elements to represent columns and beams. Elements are modeled at the geometric centerline of the structural member they represent with the following exceptions:

• The below grade and mezzanine slabs are modeled at their actual top-of-slab elevation.

• Minor adjustments are made to the dimensions and locations of wall openings to maximize mesh regularity in the model.

• Roof truss locations are adjusted to align with the roof shell element mesh.

The adjustments described above are intended to maintain mesh regularity to the extent possible.

3.4.2.6.6 Structural Analysis Results Concrete walls and slabs in the SHINE facility are designed for axial, flexural, and shear loads per provisions of ACI 349-13 (ACI, 2014) considering all applicable design basis load combinations. Walls and slabs are modeled in SAP2000 using shell elements. To determine the longitudinal and transverse reinforcement required within a wall or slab, the design is performed on an element basis. Using resultant forces obtained from SAP2000 model data, the element is designed as a reinforced concrete section per ACI 349-13 (ACI, 2014). The required area of steel is determined for combined axial and flexural loads, in-plane shear loads, and out-of-plane shear loads. Using these results, reinforcement size and spacing is specified.

3.4.3 SEISMIC CLASSIFICATION AND QUALIFICATION This subsection discusses the methods by which the SHINE facility SSCs are classified and qualified to ensure functional integrity.

3.4.3.1 Seismic Classification Facility SSCs, including their foundations and supports, that must perform safety function(s) after an SSE are designated as Seismic Category I. Safety-related SSCs are classified as Seismic Category I.

SSCs that are co-located with a Seismic Category I SSC and must maintain structural integrity in the event of an SSE to prevent unacceptable interactions with a Seismic Category I SSC, but are not required to remain functional, are designated as Seismic Category II.

The seismic classifications of SSCs important to safety are shown in Table 3.4-1.

SHINE Medical Technologies 3.4-10 Rev. 1

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage performance of the active components during the tests. For acceptability, the components shall demonstrate their ability to perform their intended safety functions when subjected to all applicable loads.

3.4.3.2.3 Comparison with Existing Databases ISRS are used to develop RRS for comparison with existing response from a database. The candidate equipment must be similar to equipment in the existing seismic experience databases.

3.4.3.2.4 Combined Methods of Qualification Based on the available information, component complexity, and functional requirements, the above mentioned analytical and test methods may be combined in various sequence and content to achieve seismic qualification of the subject components.

3.4.4 SEISMIC INSTRUMENTATION Seismic instrumentation is not required under Section IV(a)(4) of Appendix S to 10 CFR 50 or Section VI(a)(3) of Appendix A to 10 CFR 100 because the SHINE facility is not a nuclear power plant. However, the facility has nonsafety-related seismic instrumentation to record accelerations experienced at the site during a seismic event.

The seismic instrumentation establishes the acceptability of continued operation of the plant following a seismic event. This system provides acceleration time histories or response spectra experienced at the facility to assist in verifying that safety-related SSCs important to safety at the SHINE facility can continue to perform their safety functions.

Seismic monitoring is performed by the process integrated control system (PICS), which is described in Section 7.3. Indication of a seismic event results in an alarm in the facility control room.

3.4.5 SEISMIC ENVELOPE DESIGN FOR EXTERNAL HAZARDS 3.4.5.1 AIRCRAFT IMPACT ANALYSIS The safety-related structures at the SHINE facility are evaluated for aircraft impact loading resulting from small aircraft which frequent the Southern Wisconsin Regional Airport (SWRA).

The analysis consists of a global impact response analysis and a local impact response analysis.

The global impact response analysis is performed using the energy balance method, consistent with U.S. Department of Energy (DOE) Standard DOE-STD-3014-2006 (DOE, 2006). The permissible ductility limit for reinforced concrete elements is in accordance with Appendix F of ACI 349-13 (ACI, 2014). The permissible ductility limit for truss members is determined from Chapter NB of ANSI/AISC N690-12 (ANSI/AISC, 2012). The calculated values are then used to create the appropriate elastic or elastic-plastic load deflection curves. From these curves, the available energy absorption capacity of the structure at the critical impact locations is determined. The Challenger 605 was selected as the critical aircraft for the global impact analysis based on a study of the airport operations data. The Challenger 605 is evaluated as a design basis aircraft impact. The probabilistic distributions of horizontal and vertical velocity of impact are determined from Attachment E of Lawrence Livermore National SHINE Medical Technologies 3.4-13 Rev. 1

onym/Abbreviation Definition AS facility compressed air system RS facility chemical reagent system A fire hazard analysis

z. fluid ounce HS facility nitrogen handling system feet per second feet gram m3 grams per cubic centimeter grams per liter grams per second gallon gallons per minute L grams of uranium per liter NE Medical Technologies 4-xii Rev. 1

NE Medical Technologies 4a2.2-19 Rev. 0

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Subcritical Assembly Figure 4a2.2 Primary System Interfaces During Irradiation Unit Operation

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SHINE Medical Technologies 4a2.2-19 Rev. 1

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Irradiation Facility Biological Shield The thickness of the walls of the IU cell shielding varies from approximately 4.0 feet (ft.)

(1.2 meters [m]) to 5.8 ft. (1.8 m), the walls of the TOGS shielded cell shielding vary from approximately 4.0 ft. (1.2 m) to 6.0 ft. (1.8 m), and the walls of the primary cooling room shielding vary from approximately 0.7 ft. (0.2 m) to 1 ft. (0.3 m). The IU cell cover plug thickness is approximately 5.04.3 ft. (1.53 m), the TOGS cover plug thickness is approximately 6.0 ft. (1.8 m),

and the primary cooling room cover plug thickness is approximately 1.0 ft. (0.3 m).

Concrete shielding is of standard density (nominally 140 pounds per cubic foot [lb/ft3])

(2.24 grams per cubic centimeter [g/cm3]) concrete, and shield thicknesses result in general dose rates on the external surface of the shielding of less than 1.0 millirem per hour (mrem/hr).

Local hot spots (e.g., penetrations, interfaces) will be measured as part of the shielding test program and will be managed appropriately according to the Radiation Protection Program (see Section 11.1). See Figure 4a2.5-1 for a general depiction of the ICBS.

The primary cooling room shield doors are carbon steel and have an approximate thickness of 3 inches (in.) (8 centimeters [cm]).

4a2.5.2.3 Loss of Shield Integrity The biological shield walls and supporting structures are designed and constructed to remain intact during normal operations as well as during and following design basis accidents. A loss of shield integrity is not credible given the seismic design and robust nature of the IU and TOGS cells.

4a2.5.2.4 Unrestricted Environment Based on the design and construction of the biological shield walls, the neutron flux to soils surrounding the biological shield walls, in the unrestricted environment, is estimated to be less than 100 n/cm2-s. Thus, the neutron activation of groundwater and soils surrounding the biological shield is expected to be insignificant.

4a2.5.3 SHIELD MATERIALS The ICBS concrete shielding uses two distinct materials in different configurations to assemble the biological shield and meet the radiation exposure goals defined in Chapter 11. The materials that make up the concrete shielding use an engineered concrete mix with carbon steel reinforcing bars. Standard concrete is used with no special additives for shielding purposes. In the shielding analyses, individual rebar is not modeled. Instead a homogenization of rebar and concrete is used when rebar is included in the modeling. Conservative assumptions are used to define the overall shielding properties of the concrete and rebar, and secondary radiation production is considered in the analysis.

4a2.5.3.1 Shielding Calculations Calculations are performed with the software package MCNP (Monte Carlo N-Particle Transport Code). MCNP is developed and validated by Los Alamos National Laboratory (LANL) and distributed by the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL). MCNP uses a Monte Carlo based particle (neutrons and photons) transport method to generate a set of particle tracks through a model of the facility geometry (LANL, 2011). The Monte Carlo method generates a statistical set of results for individual SHINE Medical Technologies 4a2.5-2 Rev. 1

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Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Thermal Hydraulic Design

[ ]PROP/ECI, the flow rate to CC3 is approximately [

]PROP/ECI.

The cooling water supply to the lower plenum of the SASS is at a pressure of up to approximately 20 pounds per square inch gage (psig) (138 kilopascal [kPa]) and a temperature of 59°F to 77°F (15°C to 25°C). The maximum PCLS cooling water temperature of 77°F (25°C) is used in the calculations presented in Subsection 4a2.7.5, below. This temperature limit is protected by the TSV reactivity protection system (TRPS) IU Cell Safety Actuation setpoints.

The TSV headspace pressure is maintained slightly below atmospheric pressure. The pressure over the target solution in the TSV is normally between -2 psig (-14 kPa) and 0 psig (0 kPa).

The cooling water pressure difference across the TSV within the SASS is less than approximately 7 psid (48 kPa), which accounts for entrance and exit pressure losses and frictional pressure losses in the cooling channels. See Table 5a2.2-1 for the PCLS parameters.

As the primary cooling water is maintained far from boiling even at atmospheric pressures, pressure profiles of the cooling water in the flow channels are not important to safetyfor the heat transfer characteristics of the system. Total cooling water flow rate and inlet temperature are principal variables of importance for heat transfer. These variables are monitored by the TRPS.

4a2.7.3.2 Chemical Effects Related to Heat Transfer The TSV is constructed of 347 stainless steel. The target solution is chemically compatible with this alloy of stainless steel. When 347 stainless steel is placed in a uranyl-sulfate solution at temperatures up to 212°F (100°C), the steel retains its metallic luster, and only after long periods of time does it develop a very thin tarnish film (Lane, 1958).

Plating out of chemicals on the TSV surfaces is not expected in the operating temperature range of the SHINE process. Plating out of chemicals onto surfaces can occur via two mechanisms: a layer of non-volatile material can be left on surfaces when water is removed by boiling or vaporizing, or a layer of material can form when soluble components are electro-chemically reduced to a non-soluble state. The TSV is maintained at a nominal 120°F (50°C) during irradiation, which is well below the boiling point of water, even at a pressure slightly below atmospheric. No plating out of chemicals is expected from boiling because no boiling will occur in the TSV. Evaporation of and collection of solid salts on the TSV walls at the liquid surface is postulated; however, this does not affect the heat transfer as this will be above the liquid surface.

Salts that are formed are expected to re-dissolve if they are rewetted due to differing solution heights between runs. There are multiple factors minimizing the opportunity for fission product ions to be reduced at the surface of the TSV. A stable, passive, non-porous oxide layer is quickly formed on the surface of the stainless steel. This will minimize plating out on the surface of the TSV. Therefore, plating out of chemicals on the TSV surfaces is not expected.

Potential precipitates are not expected to have significant effects on heat transfer in the TSV.

Small amounts of precipitates could form in the target solution, as discussed in Section 4a2.2.

However, the heat transfer surfaces are vertical, which reduces collection of settled precipitates.

Suspended precipitates will also be separated from the target solution through the normal filtration of the molybdenum-99 extraction column during normal isotope extraction processes.

Finally, precipitate masses are low, as indicated in Table 4a2.2-1, relative to masses that would be expected to affect heat transfer.

SHINE Medical Technologies 4a2.7-5 Rev. 1

System Interface Description cility mpressed The FCANHS supplies TOGS with compressed air or nitrogen to allow nitrogen pressure regulation in TOGS.

ndling system ANHS) cility chemical The FCRS supplies TOGS with oxygen gas to ensure hydrogen gent system recombination capability.

RS) cess vessel The PVVS accepts TOGS pressure relief gases. The PVVS also accepts t system sweep gas from TOGS during nitrogen purge system (N2PS) operation.

VS)

NE Medical Technologies 4a2.8-11 Rev. 1

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Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Facility and Process Description present a health (toxicity) risk through inhalation or other ingestion pathways. Exposure to uranium oxide and uranium metal areis controlled through storage in sealed containers, seismically -qualified storage racks and transfer carts, and confinement in the TSPS glovebox while being transferred from open containers. Airborne droplets that may emanate from the solution are controlled by sweep gas routed through a reflux condenser and exhausted to radiological ventilation zone 1 (RVZ1).

Other hazardous chemicals present in the TSPS room include sulfuric acid, hydrogen peroxide, and [ ]PROP/ECI. These reagents are used in the target solution preparation and are not produced from licensed materials.

A detailed description of the TSPS is provided in Section 4b.4.

4b.1.3.2 Molybdenum Extraction and Purification System 4b.1.3.2.1 Process Functions

• Separate Mo from irradiated uranyl sulfate solution (target solution).

• [

]PROP/ECI

• [ ]PROP/ECI

• Concentrate the product solution volume such that it may be processed by the purification subsystem.

• Reduce the acid and base vapor load of the gases going to the process vessel ventilation system.

• Purify the molybdenum-99 (Mo-99) product to within the limits described in the customer supply agreements.

4b.1.3.2.2 Safety Functions

• Prevent inadvertent criticality through design of equipment in accordance with the criticality safety evaluation.

• Ensure confinement boundaries are maintained during normal conditions and during and following design basis events to ensure dose consequences during accidents are within acceptance criteria.

• Ensure the process system boundary integrity is maintained during normal conditions and during and following design basis events to prevent uncontrolled release of radioactivity.

4b.1.3.2.3 Process Description The MEPS extracts the Mo from the irradiated uranyl sulfate. The Mo is concentrated and purified to the final Mo product form and is sampled for product specification compliance prior to transfer to the MIPS.

MEPS operations are performed in two distinct phases:

1. Mo extraction

2. Mo purification SHINE Medical Technologies 4b.1-4 Rev. 1

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Facility and Process Description 4b.1.3.5 Uranium Receipt and Storage System 4b.1.3.5.1 Process Functions

• Provide receipt and repackaging of uranium (metal or oxide) received from a supplier

• Convert uranium metal to uranium oxide

• Provide storage for uranium oxide prior to use in the TSPS

• Provide storage for uranium metal

• Provide for sampling to measure uranium enrichment 4b.1.3.5.2 Safety Functions

• Prevent inadvertent criticality through design of equipment in accordance with the criticality safety evaluation.

• Prevent exposure to uranium that would exceed allowable uptake limits through confinement of uranium during the uranium receipt, repackaging, conversion, and storage processes.

4b.1.3.5.3 Process Description The URSS processes include receipt, repackaging, storage, and conversion of uranium received from a supplier. Upon receipt of a uranium shipment, inner containers from the supplier shipping container are imported to the URSS glovebox and repackaged. If the material received is uranium metal, it is converted to uranium oxide using a furnace. If the material received is uranium oxide, further processing is not required. The uranium oxide is then weighed and repackaged into storage canisters. Once transfer to storage canisters is complete, the canisters are removed from the URSS glovebox and stored in the uranium oxide storage rack. Transfer of the storage canisters to the storage rack and to the TSPS is accomplished using the URSS transfer cart.

The SNM within the URSS system is LEU. This is present in the form of uranium oxide and uranium metal. The maximum inventory of LEU in the system is 770 kg. This is divided between the URSS glovebox, supplier shipping containers, furnace, SHINE storage containers, and the storage rack. A description of provisions for criticality control in the URSS is contained in Subsection 6b.3.2.7.

The chemical hazards present from licensed materials in the URSS room include uranium oxide powder and uranium metal. The uranium can present a health (toxicity) risk through inhalation or other ingestion pathways. Exposure to uranium oxide and uranium metal areis controlled through storage in sealed containers, seismically -qualified storage racks and transfer carts, and confinement in the URSS glovebox while being transferred from open containers. A detailed description of the URSS is provided in Section 4b.4.

4b.1.3.6 Process Vessel Vent System 4b.1.3.6.1 Process and Safety Functions Section 9b.6 describes the process and safety functions for the PVVS.

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Chapter 4 - Irradiation Unit and Special Nuclear Material Radioisotope Production Facility Description Processing and Storage is stored in a favorable configuration for criticality safety within the shipping containers and is stored in those containers in accordance with packaging limitations for use. Operators verify supplied documentation and survey the shipping container and contents for damage.

URSS aAdministrative, criticality- safety controls limit the handling of uranium metal from the transportshipping container to the URSS glovebox, as described in Section 6b.3.

Within the URSS glovebox, the uranium metal contents of the shipping container are sampled and verified for form and enrichment. Uranium metal is then repackaged into a uranium metal storage canister. The mass of uranium metal repackaged into a uranium metal storage canister is limited for criticality safety. The mass of SNM in the glovebox is limited for criticality safety. The uranium metal storage canister is then transferred to the uranium metal storage rack from the URSS glovebox using the transfer cart prior to further processing. Criticality safety controls in the URSS are further described in Subsection 6b.3.2.7.

4b.4.2.1.3 Uranium Metal Storage Uranium metal storage canisters are stored in the uranium metal storage rack within the [

]SRI. The uranium metal storage rack holds uranium metal storage canisters in a favorable configuration for criticality safety. The uranium metal storage rack is designed as an array of storage cells. Each storage cell may hold one uranium metal storage canister and the cubic dimensions for each storage cell are maintained. Criticality safety controls in the URSS are further described in Subsection 6b.3.2.7.

Uranium storage canisters are transferred individually from the uranium metal storage rack using the transfer cart to the URSS glovebox for conversion to uranium oxide. Detailed descriptions of the metal to oxide conversion can be found in Subsection 4b.4.2.1.6.

4b.4.2.1.4 Uranium Oxide Receipt Shipments of uranium oxide are received in licensed shipping containers. Shipping containers are manually transferred to the [ ]SRI within the RCA. Uranium oxide is stored in a favorable configuration for criticality safety within the shipping containers and is stored in those containers in accordance with packaging limitations for use. Operators verify supplied documentation and survey the shipping container and contents for damage.

URSS administrative, criticality safety controls limit the handling of uranium oxide from the shipping container to the URSS glovebox, as described in Section 6b.3.

Within the URSS glovebox, the uranium oxide contents are sampled and verified for form, enrichment, and moisture content. Uranium oxide is then repackaged into a uranium oxide storage canister. The mass of uranium oxide repackaged into a uranium oxide storage canister is limited for criticality safety. The mass of SNM in the glovebox is limited for criticality safety. The uranium oxide storage canister is then transferred to the uranium oxide storage rack from the URSS glovebox using the transfer cart prior to further processing. Criticality safety controls in the URSS are further described in Subsection 6b.3.2.7.

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Chapter 4 - Irradiation Unit and Special Nuclear Material Radioisotope Production Facility Description Processing and Storage 4b.4.2.1.5 Uranium Oxide Storage Uranium oxide storage canisters are stored in the uranium oxide storage rack within the [

]SRI. The uranium oxide storage rack holds uranium oxide storage canisters in a favorable configuration for criticality safety. The uranium oxide storage rack is designed as an array of storage cells. Each storage cell may hold one uranium oxide storage canister and the cubic dimensions for each storage cell are maintained. Criticality safety controls in the URSS are further described in Subsection 6b.3.2.7.

Uranium oxide storage canisters are transferred individually from the uranium oxide storage rack to the TSPS glovebox, located in the [ ]SRI, using the transfer cart for conversion to uranyl sulfate. A detailed description of the TSPS is provided in Subsection 4b.4.2.2.

4b.4.2.1.6 Uranium Metal to Oxide Conversion Uranium metal is converted to uranium oxide thermally by an oxidation furnace within the URSS glovebox. Uranium metal storage canisters are transferred individually from the uranium metal storage rack to the URSS glovebox using the transfer cart. Uranium metal storage canisters are imported into the URSS glovebox, and contents are transferred to the furnace.

The oxidation process results in the uranium metal being converted into a powder. To prevent any entrainment of the uranium oxide powder into the ventilation systems, the URSS glovebox is kept at negative pressure by radiological ventilation zone 1 (RVZ1) and equipped with high efficiency particulate air (HEPA) filters to remove particles at both the ventilation supply and exhaust penetrations of the glovebox. The volume of each filter housing is limited as a criticality safety control.

Before the uranium oxide is removed from the furnace, the process is verified to have spent sufficient time at a minimum sustained temperature to ensure metal has been converted to oxide.

The uranium oxide is removed from the furnace and packed into a uranium oxide storage canister by the method described in Subsection 4b.4.2.1.4.

Criticality safety controls in the URSS are further described in Subsection 6b.3.2.7.

4b.4.2.2 Target Solution Preparation System The TSPS produces a LEU, uranyl sulfate solution, which once qualified for use, is referred to as target solution. Solid uranium oxide is dissolved in a sulfuric acid solution to convert the uranium to uranyl sulfate. Hydrogen peroxide may be used as a catalyst to aid the conversion. The solution is adjusted as needed for pH and the batch is verified to be within the specifications of the target solution qualification program. Target solution is released by operators to the RPF for use in the irradiation cycle. Solutions prepared by TSPS may be used either as a fresh target solution batch or as makeup solution for in-use target solution batches in the RPF.

4b.4.2.2.1 Safety Functions The safety functions of TSPS are provided in Section 4b.1. The system interfaces for TSPS are provided in Table 4b.4-6. The process flow diagram for TSPS are provided in Figure 4b.4-3.

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Chapter 4 - Irradiation Unit and Special Nuclear Material Radioisotope Production Facility Description Processing and Storage 4b.4.2.2.2 Dissolution of Uranium Oxide Uranium oxide is converted to a uranyl sulfate solution within the uranyl sulfate dissolution tanks.

Uranium oxide, within a uranium oxide storage canister, is transferred from the uranium oxide storage rack to the [ ]SRI using the transfer cart. The uranium oxide storage canister is imported and opened within the TSPS glovebox. Only one uranium oxide storage canister is imported to the TSPS glovebox at any time. The TSPS glovebox is kept at negative pressure by RVZ1 and equipped with HEPA filters on the supply and exhaust connections. The volume of the filters is limited as a criticality safety control.

Measurement, by mass, of uranium oxide powder is performed in the TSPS glovebox, and the material is transferred to one of the two uranyl sulfate dissolution tanks via a normally closed port in the TSPS glovebox. Two ports are provided, one dedicated to each uranyl sulfate dissolution tank. The ports preclude backflow of liquid from a uranyl sulfate dissolution tank. Unused uranium oxide remains in the uranium oxide storage canister and is returned to the uranium oxide storage rack.

The uranyl sulfate dissolution tanks are designed with favorable dimensions for criticality safety and are spaced from one another to minimize reactivity by interaction. Sulfuric acid used to convert the uranium oxide to uranyl sulfate is added to the tank. Hydrogen peroxide may also be added as a catalyst, and uranyl peroxide is formed as an intermediate. Heat is applied to the uranyl sulfate dissolution tank to aid the conversion to uranyl sulfate. Heat also decomposes excess hydrogen peroxide if it is used as a catalyst. Throughout the conversion process, the tank may be agitated. A reflux condenser on the exhaust ventilation of the uranyl sulfate dissolution tank is used to condense and return evaporated water. On a leak of the reflux condenser into the dissolution tank, a high level in the tank results in an engineered safety features actuation system (ESFAS) dissolution tank isolation, which closes the radioisotope process facility cooling system (RPCS) supply and return cooling water valves. Non-condensable gases are exhausted from the condenser to RVZ1 through a HEPA filter. The reflux condenser size is limited as a criticality safety control as described in Section 6b.3, but operation of the reflux condenser is not required to maintain a safe configuration.

Once operators verify the dissolution process is complete by sampling, the uranyl sulfate is pumped to the target solution preparation tank through a set of filters to remove any potentially undissolved solids. The filters are limited in size as a criticality safety control.

4b.4.2.2.3 Preparation of Target Solution Both qualified target solution batches and uranyl sulfate makeup solutions are prepared for use in the target solution preparation tank. The target solution preparation tank has capacity for an entire batch of target solution and is a favorable geometry for criticality safety. Solutions are pumped into the target solution preparation tank from the uranyl sulfate dissolution tanks and blended to generate a target solution batch. If the solution is to be a qualified target solution batch, reagents, such as water, sulfuric acid, [ ]PROP/ECI are added to the tank to adjust solution properties within the constraints specified by the Target Solution Qualification Program, as described in Section 4a2.2. Makeup solution is adjusted by operators as needed to ensure the batch already in a TSSS tank will meet the Target Solution Qualification Program requirements once the makeup solution is added.

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Chapter 4 - Irradiation Unit and Special Nuclear Material Radioisotope Production Facility Description Processing and Storage Agitation of the target solution preparation tank is provided to ensure the solutions are well-mixed before samples are taken and before they are pumped to the TSSS. RVZ1 provides ventilation for the target solution preparation tank. HEPA filters at the target solution preparation tank minimize contamination of the ventilation system.

4b.4.2.3 Unirradiated SNM Related Equipment The following is a list of process equipment associated with processing unirradiated SNM.

System components meet the criticality safety controls determined in the nuclear criticality safety evaluations described in Section 6b.3. Nominal sizes and specifications are provided below:

a. URSS glovebox

• Quantity: 1

• Design guidance: AGS-G001-2007 (AGS, 2007)

b. Uranium oxidation furnace

• Quantity: 1

• Normal operating temperature: 482°F-1832°F (250°C-1000°C)

c. Uranium storage racks

• Capacity: 80 canisters

d. Uranium metal storage canister

Description:

Storage of uranium metal

• Capacity: 25.8 lbs. (11.7 kg)Content administratively controlled to less than 14.7 lbs.

(6.70 kg)

e. Uranium oxide storage canister

Description:

Storage of uranium oxide

• Capacity:Content administratively controlled to less than 8.8 lbs. (4.0 kg)

f. Transfer cart

• Capacity: 1 canister

g. TSPS glovebox

• Quantity: 1

• Design guidance: AGS-G001-2007 (AGS, 2007)

h. Uranyl sulfate dissolution tank

• Quantity: 2

• Size: 5.3 gal. (20 l)

• Normal operating pressure: atmospheric (vented)

i. Target solution preparation tank

• Quantity: 1

• Size: [ ]PROP/ECI

• Normal operating pressure: atmospheric (vented)

j. Uranyl sulfate dissolution tank filter

• Quantity: 2 SHINE Medical Technologies 4b.4-8 Rev. 1

Chapter 4 - Irradiation Unit and Special Nuclear Material Radioisotope Production Facility Description Processing and Storage Table 4b.4 URSS Interfaces Interfacing System Interface Description Target solution preparation Uranium oxide storage canisters are transferred from the system (TSPS) uranium oxide storage rack to the TSPS glovebox by the transfer cart.

Radiological ventilation zone 1 RVZ1 provides exhaust ventilation to the URSS glovebox.

(RVZ1)

Radiological ventilation zone 2 The URSS glovebox ventilation supply is taken from RVZ2.

(RVZ2)

Solid radioactive waste Spent SNM containers and URSS glovebox filters are packaging (SRWP) processed by SRWP for disposal.

Normal electrical power supply The NPSS is distributed to provide power within the URSS system (NPSS) glovebox, the oxidation furnace, and ancillary equipment.

Process integrated control The URSS provide measurement signals to the PICS. The system (PICS) PICS allows operators to control processes in the URSS.

SHINE Medical Technologies 4b.4-14 Rev. 1

Chapter 4 - Irradiation Unit and Special Nuclear Material Radioisotope Production Facility Description Processing and Storage Table 4b.4 TSPS Interfaces Interfacing System Interface Description Uranium receipt and Uranium oxide storage canisters are transferred from the uranium storage system (URSS) oxide storage rack to the TSPS glovebox by the transfer cart.

Target solution staging Uranyl sulfate solutions are pumped from the target solution system (TSSS) preparation tank to the target solution hold tanks.

Facility chemical reagent The chemical reagent system supplies reagents to the uranyl system (FCRS) sulfate dissolution tanks for the conversion of uranium oxide to uranyl sulfate solution as well as reagents for adjustment of solutions in the TSPS.

Solid radioactive waste Spent TSPS glovebox filters and spent liquid filters are processed packaging (SRWP) by SRWP system.

Radioisotope process The RPCS provides process cooling water to the reflux facility cooling system condensers on the uranyl sulfate dissolution tanks.

(RPCS)

Radiological ventilation RVZ1 provides exhaust ventilation to the TSPS glovebox, the zone 1 (RVZ1) uranyl sulfate dissolution tanks, and the target solution preparation tank.

Radiological ventilation The ventilation supply for the TSPS glovebox, the uranyl sulfate zone 2 (RVZ2) dissolution tanks, and the target solution preparation tank is taken from RVZ2.

Normal electrical power The NPSS is distributed to provide power to the TSPS glovebox, supply system (NPSS) to operate the pumps, heating elements, and ancillary equipment.

Engineered safety features The ESFAS actuates isolation functions on detection of high actuation system (ESFAS) dissolution tank level.

Process integrated control The TSPS provides measurement signals to the PICS. PICS system (PICS) allows operators to control processes in TSPS.

SHINE Medical Technologies 4b.4-15 Rev. 1

onym/Abbreviation Definition ho/cm micromho per centimeter RA as low as reasonably achievable 1 argon-41 ME American Society of Mechanical Engineers British thermal unit hr British thermal units per hour centimeter T dry bulb temperature AS facility compressed air system HS facility chilled water system RS facility chemical reagent system WS facility demineralized water system HS facility nitrogen handling system WS facility potable water system R facility structure 4 facility ventilation zone 4 NE Medical Technologies 5-v Rev. 1

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Chapter 5 - Cooling Systems Primary Closed Loop Cooling System 5a2.2 PRIMARY CLOSED LOOP COOLING SYSTEM 5a2.2.1 DESIGN BASES AND FUNCTIONAL REQUIREMENTS The primary closed loop cooling system (PCLS) provides forced convection water cooling to the target solution vessel (TSV) and neutron multiplier during irradiation of the target solution and immediately prior to transferring target solution from the TSV to the TSV dump tank. The PCLS also provides indirect cooling of the light water pool via natural convection heat transfer to the PCLS components submerged in the pool, as described in Subsection 4a2.7.3. The PCLS rejects heat to the radioisotope process facility cooling system (RPCS). A total of eight independent instances of PCLS are installed in the SHINE facility, one for each irradiation unit (IU). There are no common pressure retaining components between the instances of PCLS. The major PCLS equipment is located in the primary cooling room and the IU cell.

Each instance of PCLS includes two pumps, a heat exchanger, and a cooling water clean-up side stream located in the primary cooling rooms adjacent to the east side of each IU cell. In the IU cell, the PCLS is connected to the subcritical assembly system (SCAS) and includes an air separator, an expansion tank, and a nitrogen-16 (N-16) delay tanks. Figure 5a2.2-1 provides a PCLS flow diagram.

The process functions of the PCLS cooling system are to:

• remove heat from each TSV and neutron multiplier during full-power IU operation;

• cool the light water pool by natural convection heat transfer to PCLS components inside the light water pool;

• maintain water quality to reduce corrosion and scaling;

• limit concentrations of particulate and dissolved contaminants that could be made radioactive by neutron irradiation;

• reduce N-16 radiation exposure within the primary cooling room in support of as low as reasonably achievable (ALARA) goals; and

• remove entrained gases from the cooling water.

PCLS removes heat from the TSV and neutron multiplier during startup and irradiation by circulating water in an upward direction [

]PROP/ECI along the exterior surfaces of the TSV and neutron multiplier walls. The subcritical assembly support structure (SASS) provides the shell side pressure boundary to direct the cooling water flow past the TSV and neutron multiplier. The PCLS is attached to the SASS upper and lower plenums.

PCLS is designed to remove a minimum of 580,000 British thermal units per hour (Btu/hr)

(170 kilowatts [kW]) of heat from each IU during full-power operation and during shutdown conditions when target solution is in the TSV.

PCLS is designed to maintain the pressure of the cooling water in the SASS higher than the internal pressure of the TSV. The TSV is designed and fabricated to prevent target solution from leaking into the PCLS. See Section 4a2.4 for additional information related to the TSV.

The PCLS cleanup side stream maintains system cooling water quality. The PCLS is designed to operate without corrosion inhibiting chemicals in the process fluid. The cleanup side stream can SHINE Medical Technologies 5a2.2-1 Rev. 1

Chapter 5 - Cooling Systems Primary Closed Loop Cooling System Flow instrumentation is provided to monitor the flowrate of the PCLS cooling water. The PCLS is normally operated as a constant flowrate system during irradiation. However, the PCLS may operate with either one or both pumps operating.

If the PCLS temperature or flowrate is outside allowable limits, the TRPS initiates an IU Cell Safety Actuation, resulting in a transfer of the target solution to the TSV dump tank where it is cooled by natural convection to the light water pool.

Expansion tank level instrumentation provides indication of loss of cooling water, such as by evaporation or radiolysis. Addition of makeup cooling water is a manual operation. Expansion tank level instrumentation can also perform a leak detection function as described in Subsection 5a2.2.6.

Conductivity instrumentation is provided to measure the conductivity of the PCLS water and monitor the performance of the PCLS cleanup side stream. Conductivity instrumentation can also perform a leak detection function as described in Subsection 5a2.2.6.

The PCLS pressure, flow, temperature, and expansion tank level indications are available locally and in the control room. Sampling and analysis of cooling water from the PCLS is performed locally. System operational controls are in the control room.

5a2.2.4 RADIATION MONITORS AND SAMPLING The RVZ1e line ventilating the PCLS expansion tank headspace is equipped with radiation monitors as described in Subsection 9a2.1.1. The RVZ1e radiation monitors are intended to detect leakage of target solution or fission product gases from the PSB or neutron multipliers. If radiation exceeding a predetermined setpoint is detected, the TRPS initiates an IU Cell Safety Actuation and the RVZ1e line is isolated.

Sampling and analysis of the water from the PCLS is performed to ensure that the water quality requirements are being maintained and contaminants are not present in the cooling water.

Maintaining water quality ensures functional and safe operation by reducing corrosion damage and scaling. See Table 5a2.2-1 for water quality requirements. Sampling of cooling water for radiological contaminants is performed to detect possible leakage of target solution into the PSB.

5a2.2.5 PCLS INTERFACES The system interfaces of the PCLS are listed in Table 5a2.2-3.

The PCLS cooling water is pumped through the PCLS heat exchanger, where the heat is transferred to the RPCS and subsequently transferred to the process chilled water system (PCHS), where it is dissipated to the environment.

The PCLS cooling water leaves the SCAS and enters the PCLS air separator, which allows entrained radiolytic gas to separate from the cooling water. Besides hydrogen and oxygen, the headspace contains air, water vapor, and small amounts of N-16 and argon-41 (Ar-41). An interface between the RVZ1e and the expansion tank allows radiolytic gases to be purged to RVZ1e, preventing the buildup of hydrogen gas. Ambient air from within the primary confinement boundary is drawn through a flame arrestor and filter for sweeping of the expansion tank headspace.

SHINE Medical Technologies 5a2.2-4 Rev. 1

Chapter 5 - Cooling Systems Primary Closed Loop Cooling System Table 5a2.2 PCLS Components Component Functions Code/Standard PCLS heat exchanger Transfers heat from PCLS cooling ASME BPVC,Section VIII, loop to the RPCS Division 1 (ASME, 2010)

PCLS expansion tank Provides thermal expansion ASME BPVC Section VIII, protection and pump head, and Division 1 (ASME, 2010) facilitates cooling loop level monitoring Piping components PCLS cooling loop piping ASME B31.3 (ASME, 2013)

Nitrogen-16 (N-16) delay Allows for the decay of N-16 that is ASME B31.3 (ASME, 2013) tanks generated in the cooling water by neutron activation of oxygen and for the decay of the gaseous flow path from the PCLS air separator PCLS pumps Circulates PCLS cooling water Note(a) through system components PCLS instrumentation Provides indication of PCLS See Chapter 7 for safety-operating parameters related instrumentation See Note(a) for nonsafety-related instrumentation PCLS air separator Allows entrained radiolytic gas to ASME BPVC Section VIII, leave the cooling water and enter into Division 1 (ASME, 2010) the expansion tank where it is vented to prevent the buildup of hydrogen in the system PCLS flame arrestor with Prevents the ignition of hydrogen in Note(a) filter the PCLS expansion tank if RVZ1e flow through the expansion tank is lost PCLS deionizer bed Removes dissolved ions from the Note(a)

PCLS cooling water a) Commercially available equipment designed to standards satisfying system operation.

SHINE Medical Technologies 5a2.2-7 Rev. 1

Table 5a2.2 PCLS System Interfaces System Interface Description dioisotope process The RPCS interfaces with each of the eight instances of PCLS oling water system inside the radiologically controlled area (RCA). Nonsafety-related PCS) manual isolation valves are located at the interface with PCLS.

cility demineralized The FDWS interfaces with each of the eight PCLS cooling loops ter system (FDWS) inside the RCA. The FDWS interfaces with the PCLS downstream of a FDWS vacuum breaker. Nonsafety-related manual isolation valves are located at the interface with PCLS.

bcritical assembly The SCAS interfaces with the PCLS in each of the eight light water stem (SCAS) pools located in the irradiation facility (IF).

rmal electrical power The NPSS provides power to PCLS process skid, including pumps pply system (NPSS) and instrumentation, located inside the IF.

interruptible electrical The UPSS provides the PCLS safety-related instrumentation with wer supply system electrical power during normal conditions and during and following PSS) design basis events.

V reactivity protection The PCLS provides instrumentation for the TRPS to monitor stem (TRPS) variables important to the safe operation of the PCLS. The TRPS provides controls to the PCLS components to perform safety actuations when monitored variables exceed predetermined limits.

cility compressed The FCANHS provides compressed nitrogen gas to the PCLS loop nitrogen handling pneumatic control mechanisms located inside the IF.

stem (FCANHS) ocess integrated The PICS monitors and controls the PCLS process parameters, ntrol system (PICS) utilizing the instrumentation and controlled components within the IF.

diological ventilation The RVZ1 provides an exhaust path from the headspace of each of ne 1 (RVZ1) the eight PCLS expansion tanks. The PCLS removes radiolytic gas from the cooling water and vents it to prevent combustible gas mixtures from forming.

diological ventilation The RVZ2 provides an indirect source of makeup air into the PCLS ne 2 (RVZ2) expansion tanks via the supply air provided to the IF through the primary confinement.

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Chapter 5 - Cooling Systems Primary Closed Loop Cooling System Figure 5a2.2 Primary Closed Loop Cooling System Flow Diagram SHINE Medical Technologies 5a2.2-9 Rev. 1

Table 5a2.3 RPCS Interfaces (Sheet 1 of 2)

System Interface Description imary closed loop The RPCS interfaces with each of the eight PCLS cooling loops inside oling system (PCLS) the RCA. Nonsafety-related manual isolation valves are located at the interface with PCLS.

V off-gas system Interfaces at the TOGS cooling water supply and return connections OGS) inside the RCA to condense water vapor and remove heat from recombiner condensers and condenser-demisters. Nonsafety-related manual isolation valves are located at the interface with TOGS.

olybdenum extraction Interfaces at the evaporator supply and return connections inside the d purification system RCA to facilitate condensation of water vapor. Nonsafety-related EPS) manual isolation valves are located at the interface with MEPS.

ocess vessel vent Interfaces at the supply and return connections of the PVVS cooler and stem (PVVS) condensers within the RPF section of the RCA to reduce the PVVS process temperature and relative humidity. Nonsafety-related manual isolation valves are located at the interface with PVVS.

ocess chilled water Interfaces at the supply and return connections of the RPCS heat stem (PCHS) exchanger inside the RCA and transfers heat from the RPCS to the PCHS so it can be released to the environment exterior to the RCA boundary. Nonsafety-related manual isolation valves are located at the interface with PCHS.

rget solution Interfaces at the supply and return connections of the TSPS reflux eparation system condensers inside the RCA to mitigate liquid loss during dissolution.

SPS) Nonsafety-related manualSupply and return isolation valves are located at the interface with TSPS.

diological ventilation Interfaces at the supply and return connections of the IU supplemental ne 1 recirculating cooling system fan coil, exterior to the primary confinement boundary, oling subsystem inside the RCA. Nonsafety-related manual isolation valves are located VZ1r) at the interface with RVZ1r.

diological ventilation Interfaces at the supply and return connections of the recirculating unit ne 2 recirculating fan coils inside the RCA. Nonsafety-related manual isolation valves are oling subsystem located at the interface with RVZ2r.

VZ2r) cility demineralized Interfaces upstream of the RPCS pumps inside the RCA to supply ter system (FDWS) makeup water to the RPCS.

utron driver assembly Interfaces with each of the nine NDAS cooling cabinets within the RCA stem (NDAS) to remove heat from the independent NDAS cooling loops.

NE Medical Technologies 5a2.3-6 Rev. 1

Chapter 5 - Cooling Systems Nitrogen-16 Control 5a2.7 NITROGEN-16 CONTROL Nitrogen-16 (N-16) is generated in the PCLS and light water pool by the neutron activation of oxygen. The N-16 control is provided by the primary closed loop cooling system (PCLS) delay tanks. As shown in Figure 5a2.2-1, onea liquid delay tank is located downstream of the air separator, in the PCLS cooling loop flow path, and the other is a gaseous delay tank located downstream of the PCLS expansion tank, inside the primary confinement boundary.

The N-16 delay tanks provides additional holdup time to allow for sufficient decay of N-16 prior to exiting the shielding to meet with as low as reasonably achievable (ALARA) goals and the radiation protection program. In addition, to allowing a portion of the N-16 to decay, a reduction in shielding wall thickness is realized as well as a reduction in the PCLS equipment radiation tolerance requirements.

The PCLS uses an air separator to remove entrained gases from the cooling water flow path.

The PCLS air separators are vented to the headspace of the corresponding IU cell expansion tank inside the primary confinement boundary. The headspace of the PCLS expansion tank accepts separated gases, including N-16, and directs those gases to the gaseous delay tank, via vent lines through the primary confinement boundary, and into the radiological ventilation zone 1 exhaust (RVZ1e). Redundant safety-related isolation valves are provided on the flow path between the gaseous N-16 delay tank and the RVZ1e interface with the PCLS.The gas volumes of the expansion tank headspace and vent lines are sufficient to allow adequate decay of N-16 prior to the gases leaving the IU cell shielding.

Subsection 11.1.1 provides a discussion of airborne and liquid radiation sources at the SHINE facility, including N-16.

SHINE Medical Technologies 5a2.7-1 Rev. 1

Chapter 6 - Engineered Safety Features Summary Description Table 6a2.1 Comparison of Unmitigated and Mitigated Radiological Doses for Select Irradiation Facility DBAs Unmitigated Public Dose (rem) Mitigated Public Dose (rem)

Worker Worker Public Worker Public Worker Limiting Limiting TEDE TEDE TEDE TEDE Representative DBA Organ Organ Mishandling or Malfunction of Target Solution 5.0E+00 4.1E+02 2.4E+03 6.5E-02 1.5E+00 3.0E+00 (Primary Confinement Boundary - IU Cell)

Mishandling or Malfunction of Equipment 4.9E+00 4.0E+02 2.3E+03 2.3E-01 4.8E+00 2.8E+01 (Primary Confinement Boundary - TOGS Cell)

Facility-Specific Events 2.1E+00 2.5E+02 2.4E+02 3.34E-01 4.67.1E-02 4.46.9E-02 (Tritium Confinement Boundary)

SHINE Medical Technologies 6a2.1-4 Rev. 1

Chapter 6 - Engineered Safety Features Detailed Descriptions Tritium in the IF is confined using active and passive features of the TPS. The TPS glovebox is a credited passive confinement barrier that encloses the isotope separation subsystem process equipment. The TPS glovebox is maintained at negative pressure relative to the TPS room and has a nitrogen atmosphere. The TPS glovebox provides confinement in the event of a breach in the TPS process equipment that results in a release of tritium from the isotope separation process equipment.

The ATIS header jacket and ATIS gloveboxes are a credited passive confinement barrier that encloses the ATIS header tritium lines and ATIS subsystem process equipment. The ATIS gloveboxes are maintained at negative pressure relative to the IF and have a nitrogen atmosphere. The ATIS header jacket and ATIS gloveboxes provide confinement in the event of a breach in the ATIS process equipment or ATIS header tritium lines that results in a release of tritium from the ATIS process equipment or ATIS header tritium lines.

The TPS glovebox is equipped with the glovebox stripper system (GBSS) which strips tritium from the nitrogen atmosphere during normal operation and from the process lines during maintenance. The GBSS process equipment exhausts to RVZ1e and is located in an air hood adjacent to the glovebox. The GBSS process equipment is part of the credited passive confinement boundary. The TPS process equipment other than the GBSS is not credited with confinement functions under accident conditions.

The TPS glovebox includes isolation valves on the nitrogen supply for the nitrogen atmosphere.

The TPS process equipment within the TPS glovebox has isolation valves on the process connections to the tritium supply header, the deuterium supply header, and the mixed gas return header. The TPS process equipment within the TPS glovebox also has isolation valves on the process evacuation lines that connect to the GBSS and the instrument nitrogen supply line. The GBSS process equipment has isolation valves on the connection to the ATIS glovebox exhaust and the exhaust to RVZ1e. These valves close automatically upon loss of power or receipt of a confinement isolation signal generated by the ESFAS.

When the isolation valves for a process line or glovebox close, the spread of radioactive material is limited to the glovebox plus the small amount between the glovebox and its isolation valves.

The liquid nitrogen supply and exhaust lines for the TPS equipment are credited to remain intact during a DBA and serve as a passive section of the tritium confinement boundary. Process piping outside the glovebox other than the GBSS piping to the isolation valves are not credited to remain intact during accident conditions to achieve confinement of tritium.

Upon detection of high TPS exhaust to facility stack tritium concentration or high TPS glovebox tritium concentration, the ESFAS automatically initiates a TPS isolation. The active components required to function to maintain the confinement barrier are transitioned to their deenergized (safe) state by the ESFAS. This includes process isolation valves, the GBSS RVZ1e isolation valves, and the RVZ2 dampers that isolate the TPS room from the IF general area. A description of the ESFAS and a complete listing of the active components that transition state with a TPS isolation are provided in Section 7.5.

In the event of a break in the process piping within the TPS glovebox, the release of tritium from the glovebox is uncontrolled for up to ten20 seconds until the isolation valves close. Long-term leakage and permeation of the confinement barrier result in migration of tritium out of the confinement and into the TPS room, IF, and environment. The facility accident analysis considers the effect of this air exchange in its evaluation of radiological consequences. The SHINE Medical Technologies 6a2.2-3 Rev. 1

Chapter 6 - Engineered Safety Features Summary Description Table 6b.1 Comparison of Unmitigated and Mitigated Radiological Doses for Select Radioisotope Production Facility DBAs Unmitigated Public Dose (rem) Mitigated Public Dose (rem)

Worker Worker Public Worker Public Worker Limiting Limiting TEDE TEDE TEDE TEDE Representative DBA Organ Organ Critical Equipment Malfunction 1.16.8E-7.9E-01 5.2E+01 4.6E+02 6.53.2E-01 5.52.9E+00 (Process Confinement Boundary - Supercell) 023 Critical Equipment Malfunction 7.6E-01 5.1E+01 4.4E+02 5.4E-03 4.0E-01 3.9E+00 (Process Confinement Boundary - Below Grade)

SHINE Medical Technologies 6b.1-4 Rev. 1

ESFAS Confinement Isolation Signal Act ive I & C System s Supercell Vent ilation Isolat ion (Su pply and Exhaust )

Supercell - RPF General Ar ea Delay Bed Isolation Valves N2PS Valves Delay Bed Vault Facility-Wide Hot cell Isolat ion Valves Supercell - RPF General Ar ea Guard Bed Isolation Valves PVVS Safet y Exhaust Valves Supercell Facility Mezzanine VTS Safety Actuat ion RVZ1 Isolat ion Dampers Supercell - RPF General Ar ea (RCA Boundary)

Facility Mezzanine Act ive Com ponent s N2PS Piping Product ion Facilit y Biological Facility-wide Supercell Confinement Shield Components Supercell - RPF General Ar ea RPF PVVS Carbon Beds and Piping Supercell and Delay Bed Vaults Passive Com ponents Com busti ble Gas Process Confinem ent PVVS Process Isolat ion M anagem ent Boundary NE Medical Technologies 6b.1-5 Rev. 0

ESFAS Confinement Isolation Signal Active I & C Systems Supercell Ventilation Isolation (Supply and Exhaust)

Supercell - RPF General Area N2PS Valves Facility-Wide Hot cell Isolation Valves Delay Bed Isolation Valves Supercell - RPF General Area Delay Bed Vault PVVS Safety Exhaust Valves Facility Mezzanine VTS Safety Actuation RVZ1 Isolation Dampers Supercell - RPF General Area (RCA Boundary)

Facility Mezzanine Active Components N2PS Piping Production Facility Biological Facility-wide Supercell Confinement Shield Components Supercell - RPF General Area RPF PVVS Carbon Beds and Piping Supercell and Delay Bed Vaults Passive Components Combustible Gas Process Confinement PVVS Process Isolation Management Boundary NE Medical Technologies 6b.1-5 Rev. 1

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Chapter 6 - Engineered Safety Features Detailed Descriptions 6b.2 DETAILED DESCRIPTIONS This section provides the details of the design, initiation, and operation of engineered safety features (ESFs) that are provided to mitigate the design basis accidents (DBAs) in the radioisotope production facility (RPF). The RPF DBAs, the ESFs required to mitigate the DBAs, and the location of the bases for these determinations are listed in Table 6b.1-1.

6b.2.1 CONFINEMENT The confinement systems are designed to limit the release of radioactive material to uncontrolled areas during and after DBAs to mitigate the consequences to workers, the public, and the environment. The principal objective of the confinement systems is to protect on-site personnel, the public, and the environment. The second objective is to minimize the reliance on administrative or active engineering controls to provide a confinement system that is as simple and fail-safe as reasonably possible. Figure 6b.1-1 provides an overview of the structures, systems, and components that provide RPF confinement safety functions.

A listing of the automatic isolation valves included in the confinement boundaries is in Section 7.5.

6b.2.1.1 Supercell Confinement The supercell is a set of hot cells in which isotope extraction, purification, and packaging is performed, and gaseous waste is handled. The supercell provides shielding and confinement to protect the workers, members of the public, and the environment by confining the airborne radioactive materials during normal operation and in the event of a release. The supercell includes features to allow the import of target solution, consumables, and process equipment; transfer between adjacent cells; and export of final products, waste, spent process equipment, and samples for analysis in the laboratory. The export features of the supercell are integrated into the confinement boundary to allow export operations while maintaining confinement. The supercell is described in detail in Section 4b.2.

Figure 6b.2-1 provides a block diagram of the supercell confinement boundary. [

]PROP/ECI The hot cells are fitted with stainless steel boxes for confinement of materials and process equipment. The radiological ventilation zone 1 (RVZ1) draws air through each individual confinement box, drawing air from the general RPF area, to maintain negative pressure inside the confinement, minimizing release of radiological material to the facility. Filters and carbon adsorbers on the ventilation inlets and outlets control release of radioactive material to workers and the public. These are not credited in the accident analysis. RVZ1 is described in Section 9a2.1.

The supercell ventilation exhaust ductwork is fitted with radiation monitoring instrumentation to detect off-normal releases to the confinement boxes. Upon indication of a release exceeding setpoints, isolation dampers or valves on both the inlet and outlet ducts isolate the hot cells from the ventilation system. Additionally, the actuation signal closes isolation valves on the molybdenum extraction and purification system (MEPS) heating loops and conducts a vacuum transfer system (VTS) safety actuation. The active components required to function to maintain SHINE Medical Technologies 6b.2-1 Rev. 1

Chapter 6 - Engineered Safety Features Detailed Descriptions under accident conditions. If sufficient radioactive material reaches the radiation monitors in the RVZ1 exhaust duct, ESFAS will isolate the RVZ building supply and exhaust. The evaluated accident sequence for which the process confinement boundary is necessary is listed in Table 6b.1-1 and discussed further in Section 13b.2.

The requirements needed for process confinement boundary system operability, periodic surveillance, setpoints, and other specific requirements needed to ensure the functionality of the process confinement boundary are located in technical specifications.

6b.2.2 PROCESS VESSEL VENT ISOLATION The process vessel vent system (PVVS) captures or provides holdup for radioactive particulates, iodine, and noble gases generated within the RPF and primary system boundary. The system draws air from the process vessels through a series of processing components which remove the radioactive components by condensation, acid adsorption, mechanical filtration with high-efficiency particulate air (HEPA) filters, and adsorption in carbon beds. Two sets of carbon beds are used; the guard beds located in the supercell, and the delay beds located in the carbon delay bed vault.

Fires may occur in the carbon guard and delay beds which result in the release of radioactive material into the downstream PVVS system, which leads to the facility ventilation system and the environment. The PVVS guard and delay beds are equipped with isolation valves at the inlet and the outlet that isolate the affected guard bed or group of delay beds from the system and extinguish the fire. The isolation valves also serve to prevent the release of radioactive material to the environment. The guard and delay beds are equipped with sensors to detect fires which provide indication to ESFAS. The isolation valves close within 30 seconds of the receipt of the actuation signal. The redundancy in the beds and the ability to isolate individual beds allows the PVVS to continue to operate following an isolation.

The evaluated accident sequence for which the PVVS isolation is necessary is listed in Table 6b.1-1 and discussed further in Section 13b.2.

The requirements to be specified in the technical specifications for system operability, periodic surveillance, setpoints, and other specific requirements needed to ensure the functionality of the PVVS isolations are located in Section 7.5 and Section 9.6, which describes the ESFAS and the PVVS, respectively.

6b.2.3 COMBUSTIBLE GAS MANAGEMENT Hydrogen gas is produced by radiolysis in the target solution during and after irradiation. During normal operation, the PVVS removes radiolytic hydrogen and radioactive gases generated within the RPF and primary system boundary. The PVVS is described in detail in Section 9b.6. If PVVS becomes unavailable, the buildup of hydrogen gas is limited using the combustible gas management system, which uses the nitrogen purge system (N2PS), process system piping, and the PVVS to establish an inert gas flow through the process vessels.

The principle objective of the combustible gas management system is to prevent the conditions required for a hydrogen deflagration in the gas spaces in the RPF process tanks.

SHINE Medical Technologies 6b.2-3 Rev. 1

Nuclear Criticality Safety in the Chapter 6 - Engineered Safety Features Radioisotope Production Facility The lift tanks are capable of drawing solution from the TSSS, RLWS, subcritical assembly system (SCAS), and the RDS for various purposes and supply solution to the TSSS, RLWS, RLWI, SCAS, and MEPS. The tanks are supplied with vacuum through associated vacuum pumps and valves which regulate and maintain vacuum pressure throughout the system.

Vacuum is broken in the lift tanks by venting the tank through a three-way valve which isolates the vacuum header and allows inflow from radiological ventilation zone 2 (RVZ2). Breaking vacuum in a lift tank allows gravity drain of its contents to the desired destination in one of the connected systems. Note that two-way transfers are not possible for the MEPS, RLWI, and RDS.

VTS can only supply to MEPS and RLWI, and it can only remove target solution from the RDS.

Criticality Safety Basis The NCSE for the VTS shows that the entire process will remain subcritical under normal and credible abnormal conditions.

The VTS components which contain target solution are designed with favorable geometry for the most reactive concentration. The components individually have geometry within the evaluated single parameter limits for target solution. In cases where favorable geometry components are in proximity to each other, the interaction between the components is evaluated and controlled.

The VTS components are designed to prevent leaks of solution. Vaults or hot cells containing the VTS tanks or associated piping are equipped with drip trays and adequately sized drains that drain to RDS. The vacuum buffer tank is equipped with a demister that separates potentially entrained liquid in the vapor, which prevents transfer of target solution to downstream components.

The inadvertent transfer of solution to a non-fissile system requires application of the DCP to prevent criticality accidents. The VTS piping design and features prevents transfer of target solution to non-favorable geometry components within the VTS. The vacuum lift tanksheaders are equipped with level controlsliquid detection that stops transfers upon detection of high levelliquid. Additionally, a ball-check valve is located between the vacuum lift tanks and the vacuum headerbuffer tank (VTS knockout pot) to prevent high level transfer of solution to the vacuum buffer tank.

6b.3.2.6 Process Vessel Vent System The PVVS is an off-gas management system for the process equipment which contains radioactive liquids with the potential for excessive hydrogen production in the IXP system, MEPS, RLWI, RLWS, TSSS, and VTS. The PVVS also periodically accepts gas from the target solution vessel (TSV) off-gas system (TOGS). The PVVS supplies ventilation flow and receives radioactive gas from the tanks and other equipment in these systems and processes it through a series of filters, delay beds, and blowers before it is released from the facility stack. The system does not normally contain significant fissile material.

Criticality Safety Basis The NCSE for the PVVS shows that the entire process will remain subcritical under normal and credible abnormal conditions. There are no identified criticality safety controls for the PVVS.

Inadvertent transfer of target solution into the PVVS is prevented in upstream systems.

SHINE Medical Technologies 6b.3-16 Rev. 1

Nuclear Criticality Safety in the Chapter 6 - Engineered Safety Features Radioisotope Production Facility 6b.3.2.7 Uranium Receipt and Storage System The URSS receives and stores enriched uranium oxide and metal and converts uranium metal into oxide for use in the TSPS. A process overview is provided in Figure 6b.3-6.

Activities for the receipt and measurements of uranium and the conversion from metal to oxide occur inside the URSS glovebox. Upon receipt, the convenience cans are removed from the shipping container and imported into the glovebox for measurement and repackaging into metal or oxide storage cans, as appropriate. Once the metal or oxide cans are appropriately loaded, they are moved to the appropriate storage rack via a transfer cart. For conversion activities, a metal can is moved from the storage rack via the transfer cart to the glovebox where it is converted using specified time and temperature constraints to the appropriate uranium oxide.

The oxide is then measured, and an oxide can is loaded with the product which is then transferred to the oxide storage rack via the transfer cart. Oxide may also be transferred to the TSPS via the transfer cart for processing into solution.

Criticality Safety Basis The NCSE for the URSS shows that the entire process will remain subcritical under normal and credible abnormal conditions.

Receipt and handling of shipping containers which contain uranium is in accordance with the approved safety analysis for packaging associated with each container. Areas in which intact shipping containers are stored are controlled by limiting the aggregate criticality safety index for the storage area. Administrative controls are used to ensure the criticality safety index limits are not exceeded.

Under normal process conditions, the mass of uranium metal and oxide areis limited to quantities below evaluated safe subcritical limits for moderated material. Moderators in the room and the glovebox are controlled to establish double contingency protection for the system. For a criticality to occur under normal conditions, a non-credible quantity of metal or oxide would need to be introduced into the system or mass limits would need to be exceeded concurrent with the introduction of a significant quantity of moderator. Moderator exclusions featurescontrols and the glovebox itself prevent the introductionuncontrolled intrusion of moderators into the URSS room and glovebox. The furnace and cooling lines are seismically qualified with isolation valves upstream and downstream of the furnaceareas containing exposed fissile material.

Introduction of high-enrichment uranium requires application of the DCP to prevent criticality accidents. SUpon receipt of uranium, examination of the supplier certification is verified to ensureused to confirm the condition of received material prior to import of material to the glovebox. Confirmation of material form and enrichment is within limits. Additionally, batchby sample analysis verifies the uranium enrichment is withinare used to ensure that appropriate limits are applied.

Accumulation of excess mass requires application of the DCP to prevent criticality accidents. The number of canisters within the glovebox is limited undermass of uranium in- and out-of-storage is administratively controlled. Additionally,Material contained within sealed shipping containers, the glovebox, and the storage racks is considered to be in-storage and is subject to specific limits for each of these areas. Material out-of-storage is administratively limited to a value significantly below the single-parameter subcritical limit. Controls on the use and transport of moderator SHINE Medical Technologies 6b.3-17 Rev. 1

Nuclear Criticality Safety in the Chapter 6 - Engineered Safety Features Radioisotope Production Facility exclusion featuress within the room are used to prevent the introduction of external moderatorsinteraction of material out-of-storage with moderating materials. HEPA filters, which are favorable geometry within single parameter limits, prevent the accumulation of oxide outside of the glovebox or in downstream ventilation.

Upon receipt of uranium, supplier certification is verified to be within moisture limits. Additionally, a sample analysis of received oxide verifies the moisture of the fissile material is within moisture limits. Holdup of fissile material in the process is controlled in the glovebox and furnace by tracking mass and periodic cleanout of the glovebox and furnace based on the throughput of uranium. Cleanout of fissile material holdup is independently verified prior to restarting operations. During maintenance activities, fissile material is removed prior to maintenance and moderators are removed prior to restarting operations and . Confirmation of fissile material and moderator removal is performed under supervisory oversight.

Incomplete oxidation of metal requires application of the DCP to prevent criticality accidents. The furnace oxidation steps are administratively controlled to ensure adequate oxidation.

Additionally, sample analysis following oxidation verifies oxide powder content and moisture content of the oxide. Operators visually confirm that only uranium oxide is added to an oxide canister.

Excess mass added to an oxide or metal canister requires application of the DCP to prevent criticality accidents. Supplier certification is verified that the mass of metal in a canister is within limits. Additionally, operators weigh the mass of metal added to a metal canister. Uranium oxide, either from a received canister or from the furnace, is metered onto a weigh tray to ensure the mass is within limits. Additionally, the uranium oxide transferred from the weigh tray to an oxide canister is weighed to verify the mass in the oxide canister is within limits.

The URSS oxide storage rack and metal storage rack are favorable geometry and maintain the appropriate storage cell size. The maximum number of storage cells is significantly below the allowable number of storage cells based on the mass per storage canister. Transfers from the URSS glovebox to themass in each storage racks is performed one canister at a time using a transfer cart that is administratively controlled. Movement of fissile material out-of-storage is maintainsed at an appropriate separation distance to other fissile material in storage to prevent unfavorable interaction.

6b.3.2.8 Radioactive Drain System The RDS collects overflows and leakage of target solution from systems in the RPF and directs it to two favorable-geometry tanks in below grade vaults. A process overview is provided in Figure 6b.3-7.

The system is comprised of drip pans, piping, and collection tanks. The collection tanks are normally maintained empty and are equipped with instrumentation to alert personnel of an abnormal condition. The system operates by gravity drain, where overflows and leakage flow through installed piping directly to the RDS hold tanks. The hold tank contents can be mixed, sampled, and withdrawn through the VTS to the TSSS or RLWS as appropriate.

Criticality Safety Basis SHINE Medical Technologies 6b.3-18 Rev. 1

NE Medical Technologies 6b.3-26 Rev. 0

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Nuclear Criticality Safety in the Chapter 6 - Engineered Safety Features Radioisotope Production Facility Figure 6b.3 Target Solution Preparation System Overview SHINE Medical Technologies 6b.3-26 Rev. 1

ES-3100 Convenience Shipping Cans Container Metal Oxide Uranium Handling Glovebox Import to Import to Glove Box Glove Box Weigh (Metal)

Tare (Funnel)

Load Stor age Weigh Furnace Tray Can (Oxide) Tare Tare (Funnel)

(Tray)

Furnace Load Stor age Can Tare (Tray)

Weigh Weigh Weigh (Loaded Can) (Oxide) (Loaded Can) Tare (Can)

Tare (Can)

Expor t to Pass to Expor t to Storage Oxide Glove Storage Rack Box Rack Metal Side Oxide Side NE Medical Technologies 6b.3-28 Rev. 0

Nuclear Criticality Safety in the Chapter 6 - Engineered Safety Features Radioisotope Production Facility Figure 6b.3 Uranium Receipt and Storage System Overview Receive Shipping Package Remove Canisters Oxide Metal Uranium Handling Glovebox Import Metal Import Oxide Thermal Oxidation Canister Canister Repackage to Metal Repackage to Oxide Storage Canister Storage Canister Export Metal Export Oxide Storage Canister Storage Canister Storage in Racks Storage in Racks SHINE Medical Technologies 6b.3-28 Rev. 1

1 Instrumentation and Control System Architecture 2 Target Solution Vessel Reactivity Protection System Architecture 3 Engineered Safety Feature Actuation System Architecture 1 HIPS Platform Timing 2 TRPS and ESFAS Programmable Logic Lifecycle Process 1 Process Integrated Control System Interfaces 1 TRPS Logic Diagrams 2 TRPS Mode State Diagram 1 ESFAS Logic Diagrams 2 Extraction Hot Cell 3 Carbon Guard Bed Physical Configuration 43 Vacuum Transfer System 54 Radiologically Controlled Area Isolation 1 Facility Control Room Layout 2 Status Indication Panels 3 Maintenance Workstation 1 Effluent Monitor Locations NE Medical Technologies 7-v Rev. 1

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Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System 7.5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 7.5.1 SYSTEM DESCRIPTION The engineered safety features actuation system (ESFAS) is a three-division safety-related instrumentation and control (I&C) system that performs various control and actuation functions credited by the SHINE safety analysis as required to prevent the occurrence or mitigate the consequences of design basis events within the SHINE facility. The ESFAS provides sense, command, and execute functions necessary to maintain the facility confinement strategy and provides process actuation functions required to shutdown processes and maintain processes in a safe condition. The ESFAS also provides nonsafety-related system status and measured process variable values to the facility process integrated control system (PICS) for viewing, recording, and trending.

The ESFAS monitors variables important to the safety functions for confinement of radiation and tritium within the irradiation facility (IF) and the radioisotope production facility (RPF) and for criticality safety to perform the following functions:

• Radiologically Controlled Area (RCA) Isolation

• Supercell Isolation

• Carbon Guard Bed Isolation

• Guard Bed Misalignment Actuation

• Carbon Delay Bed Isolation

• Vacuum Transfer System (VTS) Safety Actuation

• Tritium Purification System (TPS) Isolation

• Irradiation Unit (IU) Cell Nitrogen Purge

• RPF Nitrogen Purge

• Molybdenum Extraction and Purification System (MEPS) [ ]PROP/ECI Isolation

• Extraction Column Alignment Actuation

• Iodine and Xenon Purification and Packaging (IXP) Alignment Actuation

• Dissolution Tank Isolation The ESFAS monitors the IF and the RPF continually throughout the operation of processes within the main production facility, via the use of radiation monitoring and other instrumentation.

Interlocks and bypass logic necessary for operation are implemented within the ESFAS. If at any point a monitored variable exceeds its predetermined limits, the ESFAS automatically initiates the associated safety function. ESFAS logic diagrams are provided in Figure 7.5-1 and the general architecture of the ESFAS is provided in Figure 7.1-3.

7.5.2 DESIGN CRITERIA The SHINE design criteria are described in Section 3.1. Table 3.1-1 shows the SHINE design criteria applicable to the ESFAS.

7.5.2.1 Access Control ESFAS Criterion 1 - The ESFAS shall require a key or combination authentication input at the control console to prevent unauthorized use of the ESFAS.

SHINE Medical Technologies 7.5-1 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System 7.5.2.3 General Instrumentation and Control Requirements ESFAS Criterion 14 - The ESFAS safety functions shall perform and remain functional during normal operation and during and following a design basis event.

ESFAS Criterion 15 - Manual controls of ESFAS actuation components shall be implemented downstream of the digital I&C portions of the safety system.

7.5.2.4 Single Failure ESFAS Criterion 16 - The ESFAS shall be designed to perform its protective functions after experiencing a single random active failure in nonsafety control systems or in the ESFAS, and such failure shall not prevent the ESFAS and credited redundant passive control components from performing itsthe intended functions or prevent safe shutdown of an IU cell.

ESFAS Criterion 17 - The ESFAS shall be designed such that no single failure can cause the failure of more than one redundant component.

ESFAS Criterion 18 - The ESFAS shall be designed so that no single failure within the instrumentation or power sources concurrent with failures as a result of a design basis event should prevent operators from being presented the information necessary to determine the safety status of the facility following the design basis event.

7.5.2.5 Independence ESFAS Criterion 19 - Interconnections among ESFAS safety divisions shall not adversely affect the functions of the ESFAS.

ESFAS Criterion 20 - A logical or software malfunction of any interfacing nonsafety systems shall not affect the functions of the ESFAS.

ESFAS Criterion 21 - The ESFAS shall be designed with physical, electrical, and communications independence of the ESFAS both between the ESFAS channels and between the ESFAS and nonsafety-related systems to ensure that the safety functions required during and following any design basis event can be accomplished.

ESFAS Criterion 22 - Physical separation and electrical isolation shall be used to maintain the independence of ESFAS circuits and equipment among redundant safety divisions or with nonsafety systems so that the safety functions required during and following any design basis event can be accomplished.

ESFAS Criterion 23 - The ESFAS shall be designed such that no communication - within a single safety channel, between safety channels, and between safety and nonsafety systems -

adversely affects the performance of required safety functions.

ESFAS Criterion 24 - ESFAS data communications protocols shall meet the performance requirements of all supported systems.

ESFAS Criterion 25 - The timing of ESFAS data communications shall be deterministic.

SHINE Medical Technologies 7.5-3 Rev. 1

bon Guard Bed 1 Isolation initiates the following safety functions:

• Energize PVVS carbon guard bed 1 inlet isolation valve

• Energize PVVS carbon guard bed 1 outlet isolation valve

• Energize PVVS carbon guard bed bypass valve

• Deenergize PVVS carbon guard bed 2 inlet isolation valve

• Deenergize PVVS carbon guard bed 2 outlet isolation valve ESFAS initiates a Carbon Guard Bed 1 Isolation based on the following variable:

• High carbon guard bed 1 temperature presentation of the Carbon Guard Bed 1 Isolation is provided in Figure 7.5-3.

3.1.15 Carbon Guard Bed 2 Isolation bon Guard Bed 2 Isolation initiates the following safety functions:

• Energize PVVS carbon guard bed 2 inlet isolation valve

• Energize PVVS carbon guard bed 2 outlet isolation valve

• Energize PVVS carbon guard bed bypass valve

• Deenergize PVVS carbon guard bed 1 inlet isolation valve

• Deenergize PVVS carbon guard bed 1 outlet isolation valve ESFAS initiates a Carbon Guard Bed 2 Isolation based on the following variable:

• High carbon guard bed 2 temperature presentation of the Carbon Guard Bed 2 Isolation is provided in Figure 7.5-3.

3.1.16 Guard Bed Misalignment Actuation rd Bed Misalignment Actuation initiates the following safety functions:

• Energize PVVS carbon guard bed bypass valve

• Deenergize PVVS carbon guard bed 1 inlet isolation valve

• Deenergize PVVS carbon guard bed 1 outlet isolation valve

• Deenergize PVVS carbon guard bed 2 inlet isolation valves

• Deenergize PVVS carbon guard bed 2 outlet isolation valve ESFAS initiates a Guard Bed Misalignment Actuation if an isolation valve is fully closed in h carbon guard beds:

• PVVS carbon guard bed 1 inlet isolation valve OR PVVS carbon guard bed 1 outlet isolation valve fully closed, AND NE Medical Technologies 7.5-11 Rev. 1

presentation of the Guard Bed Misalignment Actuation is provided in Figure 7.5-3.

3.1.17 Carbon Delay Bed Group 1 Isolation bon Delay Bed Group 1 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed group 1 three-way valves

• Energize PVVS carbon delay bed 1 inlet isolation valve

• Energize PVVS carbon delay bed group 1 outlet isolation valves ESFAS initiates a Carbon Delay Bed Group 1 Isolation based on the following variables:

• High carbon delay bed group 1 exhaust carbon monoxide 3.1.18 Carbon Delay Bed Group 2 Isolation bon Delay Bed Group 2 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed group 2 three-way valves

• Energize PVVS carbon delay bed 2 inlet isolation valve

• Energize PVVS carbon delay bed group 2 outlet isolation valves ESFAS initiates a Carbon Delay Bed Group 2 Isolation based on the following variables:

• High carbon delay bed group 2 exhaust carbon monoxide 3.1.19 Carbon Delay Bed Group 3 Isolation bon Delay Bed Group 3 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed group 3 three-way valves

• Energize PVVS carbon delay bed 3 inlet isolation valve

• Energize PVVS carbon delay bed group 3 outlet isolation valves ESFAS initiates a Carbon Delay Bed Group 3 Isolation based on the following variables:

• High carbon delay bed group 3 exhaust carbon monoxide 3.1.20 Carbon Delay Bed 4 Isolation bon Delay Bed 4 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed 4 three-way valve

• Energize PVVS carbon delay bed 4 inlet isolation valve

• Energize PVVS carbon delay bed 4 outlet isolation valves NE Medical Technologies 7.5-12 Rev. 1

• High carbon delay bed 4 exhaust carbon monoxide 3.1.21 Carbon Delay Bed 5 Isolation bon Delay Bed 5 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed 5 three-way valve

• Energize PVVS carbon delay bed 5 inlet isolation valve

• Energize PVVS carbon delay bed 5 outlet isolation valves ESFAS initiates a Carbon Delay Bed 5 Isolation based on the following variables:

• High carbon delay bed 5 exhaust carbon monoxide 3.1.22 Carbon Delay Bed 6 Isolation bon Delay Bed 6 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed 6 three-way valve

• Energize PVVS carbon delay bed 6 inlet isolation valve

• Energize PVVS carbon delay bed 6 outlet isolation valves ESFAS initiates a Carbon Delay Bed 6 Isolation based on the following variables:

• High carbon delay bed 6 exhaust carbon monoxide 3.1.23 Carbon Delay Bed 7 Isolation bon Delay Bed 7 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed 7 three-way valve

• Energize PVVS carbon delay bed 7 inlet isolation valve

• Energize PVVS carbon delay bed 7 outlet isolation valves ESFAS initiates a Carbon Delay Bed 7 Isolation based on the following variables:

• High carbon delay bed 7 exhaust carbon monoxide 3.1.24 Carbon Delay Bed 8 Isolation bon Delay Bed 8 Isolation initiates the following safety functions:

• Energize PVVS carbon delay bed 8 three-way valve

• Energize PVVS carbon delay bed 8 inlet isolation valve

• Energize PVVS carbon delay bed 8 outlet isolation valves NE Medical Technologies 7.5-13 Rev. 1

Proprietary Information - Withheld from public disclosure under 10 CFR 2.390(a)(4)

Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System The ESFAS initiates a Carbon Delay Bed 8 Isolation based on the following variables:

• High carbon delay bed 8 exhaust carbon monoxide 7.5.3.1.25 VTS Safety Actuation VTS Safety Actuation Isolation initiates the following safety functions:

• Deenergize VTS vacuum transfer pump 1 breakers

• Deenergize VTS vacuum transfer pump 2 breakers

• Deenergize VTS vacuum transfer pump 3 breakers

• Deenergize VTS vacuum break valves

• MEPS A extraction column wash supply valve

• MEPS A extraction column eluent valve

• MEPS A [ ]PROP/ECI wash supply valve

• MEPS A [ ]PROP/ECI eluent valve

• MEPS B extraction column wash supply valve

• MEPS B extraction column eluent valve

• MEPS B [ ]PROP/ECI wash supply valve

• MEPS B [ ]PROP/ECI eluent valve

• MEPS C extraction column wash supply valve

• MEPS C extraction column eluent valve

• MEPS C [ ]PROP/ECI wash supply valve

• MEPS C [ ]PROP/ECI eluent valve

• IXP recovery column wash supply valve

• IXP recovery column eluent valve

• IXP [ ]PROP/ECI wash supply valve

• IXP [ ]PROP/ECI eluent valve

• IXP FNHS supply valve

• IXP liquid nitrogen supply valve The ESFAS initiates a VTS Safety Actuation based on the following variables or safety actuations:

• VTS vacuum header liquid detection switch signal

• VTS extraction lower lift tank A liquid detection switch signal

• VTS extraction lower lift tank B liquid detection switch signal

• VTS extraction lower lift tank C liquid detection switch signal

• VTS extraction lower lift tank D liquid detection switch signal

• VTS extraction upper lift tank A1 liquid detection switch signal

• VTS extraction upper lift tank A2 liquid detection switch signal

• VTS extraction upper lift tank B1 liquid detection switch signal

• VTS extraction upper lift tank B2 liquid detection switch signal

• VTS extraction upper lift tank C1 liquid detection switch signal

• VTS extraction upper lift tank C2 liquid detection switch signal

• TSV fill lift tank 1 liquid detection switch signal

• TSV fill lift tank 2 liquid detection switch signal

• TSV fill lift tank 3 liquid detection switch signal

• TSV fill lift tank 4 liquid detection switch signal

• TSV fill lift tank 5 liquid detection switch signal SHINE Medical Technologies 7.5-14 Rev. 1

• TSV fill lift tank 8 liquid detection switch signal

• Target solution storage lift tank liquid detection switch signal

• Uranium liquid waste lift tank liquid detection switch signal

• RDS lift tank liquid detection switch signal

• RDS liquid detection switch signal

• Supercell Area 1 Isolation

• Supercell Area 2 Isolation

• Supercell Area 6 Isolation

• Supercell Area 7 Isolation

• RCA Isolation

• Facility master operating permissive presentation of the VTS Safety Actuation is provided in Figure 7.5-3.

3.1.26 TPS Isolation Isolation initiates the following safety functions:

• Deenergize accelerator tritium interface system (ATIS) header glovebox stripper system (GBSS) isolation valves

• Deenergize ATIS header GBSS bypass isolation valves

• Deenergize GBSS RVZ isolation valves

• Deenergize ATIS header return line isolation valves

• Deenergize TPS process evacuation header isolation valves

• Deenergize ATIS header deuterium supply isolation valves

• Deenergize storage and separation system GBSS raffinate isolation valves

• Deenergize ATIS header tritium supply isolation valves

• Deenergize ATIS glovebox exhaust header isolation valves

• Deenergize TPS process evacuation GBSS isolation valves

• Deenergize TPS glovebox nitrogen supply valves

• Deenergize RVZ TPS ventilation dampers ESFAS initiates a TPS Isolation based on the following variables or safety actuation:

• High TPS exhaust to facility stack tritium concentration

• High TPS glovebox tritium concentration

• RCA Isolation

• Facility master operating permissive 3.1.27 IU Cell Nitrogen Purge Cell Nitrogen Purge transitions the nitrogen purge system (N2PS) IU cell header valves to r deenergized state.

ESFAS also provides the target solution vessel (TSV) reactivity protection system (TRPS) each IU cell with an actuation signal to initiate an IU Cell Nitrogen purge within the TRPS.

NE Medical Technologies 7.5-15 Rev. 1

• UPSS loss of external power

• TRPS IU cell 1 nitrogen purge signal

• TRPS IU cell 2 nitrogen purge signal

• TRPS IU cell 3 nitrogen purge signal

• TRPS IU cell 4 nitrogen purge signal

• TRPS IU cell 5 nitrogen purge signal

• TRPS IU cell 6 nitrogen purge signal

• TRPS IU cell 7 nitrogen purge signal

• TRPS IU cell 8 nitrogen purge signal 3.1.28 RPF Nitrogen Purge F Nitrogen Purge initiates the following safety functions:

• Deenergize PVVS blower bypass valves

• Deenergize radioactive liquid waste immobilization (RLWI) PVVS isolation valve

• Deenergize PVVS carbon guard bed bypass valves

• Deenergize N2PS RPF header valves

• Deenergize N2PS RVZ2 north header valves

• Deenergize N2PS RVZ2 south header valves ESFAS initiates an RPF Nitrogen Purge based on the following variable:

• Low PVVS flow 3.1.29 RCA Isolation A Isolation initiates the following safety functions:

• Deenergize RVZ1 exhaust isolation dampers

• Deenergize RVZ2 exhaust isolation dampers

• Deenergize RVZ2 supply train 1 isolation dampers

• Deenergize RVZ2 supply train 2 isolation dampers

• Deenergize RVZ3 supply isolation dampers shipping/receiving IF

• Deenergize RVZ3 supply isolation dampers shipping/receiving RPF

• Deenergize RVZ3 supply isolation dampers main RCA ingress/egress

• Deenergize RVZ3 supply isolation dampers RPF emergency exit

• Deenergize RVZ3 supply isolation dampers IF emergency exit

• Deenergize RVZ3 exhaust isolation dampers IF emergency exit

• Deenergize RVZ1 exhaust train 1 blower breakers

• Deenergize RVZ1 exhaust train 2 blower breakers

• Deenergize RVZ2 exhaust train 1 blower breakers

• Deenergize RVZ2 exhaust train 2 blower breakers

• Deenergize RVZ2 supply train 1 blower breakers

• Deenergize RVZ2 supply train 2 blower breakers

• Supercell Area 1 Isolation

• Supercell Area 2 Isolation

• Supercell Area 3 Isolation NE Medical Technologies 7.5-16 Rev. 1

• Supercell Area 6 Isolation

• Supercell Area 7 Isolation

• Supercell Area 8 Isolation

• Supercell Area 9 Isolation

• Supercell Area 10 Isolation

• VTS Safety Actuation

• TPS Isolation ESFAS initiates an RCA Isolation based on the following variables:

• High RVZ1 RCA exhaust radiation

• High RVZ2 RCA exhaust radiation presentation of the RCA Isolation is provided in Figure 7.5-4.

3.1.30 Extraction Column A Alignment Actuation action Column A Alignment Actuation initiates the following safety functions:

• Deenergize MEPS area A upper three-way valve

• Deenergize MEPS area A lower three-way valve

• Deenergize MEPS A extraction column eluent valve ESFAS initiates the Extraction Column A Alignment Actuation based on both of the following ts being active:

• MEPS area A upper three-way valve supplying position indication

• MEPS area A lower three-way valve supplying position indication 3.1.31 Extraction Column B Alignment Actuation action Column B Alignment Actuation initiates the following safety functions:

• Deenergize MEPS area B upper three-way valve

• Deenergize MEPS area B lower three-way valve

• Deenergize MEPS B extraction column eluent valve ESFAS initiates the Extraction Column B Alignment Actuation based on both of the following ts being active:

• MEPS area B upper three-way valve supplying position indication

• MEPS area B lower three-way valve supplying position indication 3.1.32 Extraction Column C Alignment Actuation action Column C Alignment Actuation initiates the following safety functions:

• Deenergize MEPS area C upper three-way valve NE Medical Technologies 7.5-17 Rev. 1

ESFAS initiates the Extraction Column C Alignment Actuation based on both of the following ts being active:

• MEPS area C upper three-way valve supplying position indication

• MEPS area C lower three-way valve supplying position indication 3.1.33 IXP Alignment Actuation ne and Xenon Purification and Packaging (IXP) Alignment Actuation initiates the following ty functions:

• Deenergize IXP upper three-way valve

• Deenergize IXP lower three-way valve

• Deenergize IXP recovery column eluent valve ESFAS initiates the IXP Alignment Actuation based on both of the following inputs being ve:

• IXP upper three-way valve supplying position indication

• IXP lower three-way valve supplying position indication 3.1.34 Dissolution Tank Isolation solution Tank Isolation initiates the following safety functions:

• Deenergize target solution preparation system (TSPS) radioisotope process facility cooling system (RPCS) supply cooling valves

• Deenergize TSPS RPCS return cooling valve

• Deenergize RVZ2 TSPS supply damper

• Deenergize RVZ1 TSPS exhaust damper ESFAS initiates the Dissolution Tank Isolation based on the following input being active:

• High TSPS dissolution tank 1 level switch signal

• High TSPS dissolution tank 2 level switch signal 3.2 ESFAS Monitored Variables le 7.5-1 identifies the specific variables that provide input to the ESFAS and includes the rument range for covering normal and accident conditions, the accuracy for each variable, the lytical limit, and the response time of the sensor element.

3.3 Operating Conditions ESFAS control and logic functions operate inside of the facility control room where the ironment is mild and not exposed to the irradiation process. However, the cables for the FAS are routed through the radiologically controlled area to the process areas. The routed NE Medical Technologies 7.5-18 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System The environmental conditions for ESFAS components are outlined in Table 7.2-1 through Table 7.2-3. The facility heating, ventilation and air conditioning (HVAC) systems are relied upon to maintain the temperature and humidity parameters in these areas. The facility HVAC systems are described in Section 9a2.1.

7.5.4 DESIGN ATTRIBUTES 7.5.4.1 Access Control A detailed description of access control is provided in Subsection 7.2.5.

7.5.4.2 Software Requirements Development A detailed description of the development of software requirements is provided in Subsection 7.2.6.

7.5.4.3 General Instrumentation and Control Requirements The ESFAS is powered from the uninterruptible electrical power supply system (UPSS), which provides a reliable source of power to maintain the ESFAS functional during normal operation and during and following a design basis event. The UPSS is designed to provide power to the ESFAS controls for six hours after a loss of off-site power. The UPSS is described in Section 8a2.2.

The actuation and priority logic (APL) portions within an equipment interface module (EIM) support the implementation of different actuation methods. The APL is implemented using discrete components and is not vulnerable to a software common cause failure (CCF). Having the capability for hardwired signals into each EIM supports the capability for additional and diverse actuation means from automated actuation. As an example, a division of APL circuits may receive inputs automatically from the programmable logic portion of the ESFAS, inputs from manual controls in the facility control room, and input signals from a nonsafety control system.

Both the manual controls and nonsafety control system inputs come individually into the APL and are downstream of the programmable logic portion of the ESFAS architecture as shown in Figure 7.1-3.

7.5.4.4 Single Failure The ESFAS consists of three divisions of input processing and trip determination and two divisions of actuation logic (see Figure 7.1-2) arranged so that no single failure can prevent a safety actuation when requiredwithin the ESFAS results in the loss of the protective function.

The only nonsafety inputs into the ESFAS are those from the PICS for controls. The nonsafety control signals from the PICS are implemented through a hardwired parallel interface that requires the PICS to send a binary address associated to the output state of the EIM along with a mirrored complement address. The mirrored complement address prevents any single incorrectly presented bit from addressing the wrong EIM output state. To prevent the PICS from inadvertently presenting a valid address, the ESFAS contains a safety-related enable nonsafety switch that controls when the hardwired parallel interface within the APL is active, thus controlling when the PICS inputs are allowed to pass through the input circuitry and for use in the priority logic within the APL. When the enable nonsafety switch is not active, the nonsafety-related SHINE Medical Technologies 7.5-19 Rev. 1

uires the PICS to send a binary address associated to the output state of the EIM along with a ored complement address. The mirrored complement address prevents any single incorrectly sented bit from addressing the wrong EIM output state. To prevent the PICS from vertently presenting a valid address, the ESFAS contains a safety-related enable nonsafety ch that controls when the hardwired parallel interface within the APL is active, thus controlling n the PICS inputs are allowed to pass through the input circuitry and for use in the priority c within the APL. When the enable nonsafety switch is not active, the nonsafety-related trol signal is ignored. If the enable nonsafety is active, and no automatic or manual actuation mand is present, the nonsafety-related control signal can control the ESFAS output. The dwired module provides isolation for the nonsafety-related signal path.

4.5 Independence escription of the application of independence to the ESFAS is provided in Subsection 7.2.2.

4.6 Prioritization of Functions h division of the ESFAS includes the analog logic circuitry necessary to prioritize the ESFAS ts. Automatic Safety Actuation or Manual Actuation are highest priority and PICS nonsafety trol inputs are lower in priority.

4.7 Fail-Safe fail-safe positions of components upon loss of power to ESFAS are provided in Table 7.5-2.

4.8 Setpoints points in the ESFAS are based on a documented methodology that identifies each of the umptions and accounts for the uncertainties in each instrument channel. The setpoint hodology is described in Subsection 7.2.3.

4.9 Operational Bypass, Permissives and Interlocks ntenance bypasses are described in Subsection 7.1.4.

ther Carbon Guard Bed 1 Isolation or Carbon Guard Bed 2 Isolation is active, then the Guard Misalignment Actuation is bypassed.

ESFAS starts a 180 second timer on loss of external power to the UPSS. If the indication of of external power to the UPSS clears prior to the 180 second timer expiring, the timer resets.

e timer expires, the ESFAS initiates an IU Cell Nitrogen Purge on loss of external power to UPSS.

safety inputs into the ESFAS are transferred from PICS through the hardwired module. The S inputs are bypassed with the enable nonsafety switch permitting the inputs to control FAS outputs when administrative procedures permit the operator to use the switch to enable PICS functionality with the ESFAS.

NE Medical Technologies 7.5-20 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System The manual actuation inputs from the operators in the facility control room are connected directly to the discrete APL. The manual actuation input into the priority logic does not have the ability to be bypassed and will always have equal priority to the automated actuation signals over any other signals that are present.

7.5.4.10 Completion of Protective Actions The ESFAS is designed so that once initiated, protective actions will continue to completion. Only deliberate operator action can be taken to reset the ESFAS following a protective action.

Figure 7.5-1, Sheets 2619 through 231, shows how the ESFAS latches in a protective action and maintains the state of a protective action until operator input is initiated to reset the output of the ESFAS to normal operating conditions.

The output of the ESFAS is designed so that actuation through automatic or manual means of a safety function can only change when a new position is requested. If there is no signal present from the automatic safety actuation or manual actuation, then the output of the EIM remains in its current state. A safety-related enable nonsafety switch allows an operator, after the switch has been brought to enable, to control the output state of the ESFAS with a hardwired binary control signal from the nonsafety-related controls. The enable nonsafety switch is classified as part of the safety system and is used to prevent spurious nonsafety-related control signals from adversely affecting safety-related components. If the enable nonsafety switch is active, and no automatic safety actuation or manual actuation signals are present, the operator is capable of energizing or deenergizing any EIM outputs using the nonsafety-related hardwired control signals. If the enable nonsafety switch is not active, the nonsafety-related hardwired control signals are ignored.

7.5.4.11 Equipment Qualification ESFAS rack mounted equipment is installed in a mild operating environment and is designed to meet the environmental conditions described in Subsection 7.4.3.4. Rack mounted ESFAS equipment is tested to appropriate standards to show that the effects of EMI/RFI and power surges are adequately addressed. Appropriate grounding of the ESFAS is performed in accordance with Section 5.2.1 of Institute of Electrical and Electronics Engineers (IEEE)

Standard 1050-2004, IEEE Guide for Instrumentation and Control Equipment Grounding in Generating Stations (IEEE, 2004b).

7.5.4.12 Surveillance The TRPS supports calibration and testing to ensure operability as described in Subsection 7.2.4.

7.5.4.13 Classification and Identification Each division of the ESFAS is uniquely labeled and identified in accordance with SHINE identification and classification procedures.

7.5.4.14 Human Factors The ESFAS provides manual actuation capabilities for each of the safety functions identified in Subsection 7.5.3. To support the use of manual actuations, the ESFAS includes isolated outputs SHINE Medical Technologies 7.5-21 Rev. 1

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Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System 7.5.5.2 High RVZ1 Supercell Radiation (PVVS Cell)

The high RVZ1 supercell radiation signal protects against hot cell equipment leakage or an accident that could potentially result in excess radiation doses to the workers or to the public. The RVZ1 supercell radiation is measured by an analog interface on three different channels, one for each division of ESFAS. When two-out-of-three or more high RVZ1 supercell radiation channels are active, then a Supercell Isolation for that area and VTS Safety Actuation are initiated.

7.5.5.3 High RVZ1 Supercell Radiation (MEPS Extraction Cells)

The high RVZ1 supercell radiation signal protects against hot cell equipment leakage or an accident that could potentially result in excess radiation doses to the workers or to the public. The RVZ1 supercell radiation is measured by an analog interface on two different channels, one for each Division A and Division B of ESFAS. When one-out-of-two or more high RVZ1 supercell radiation channels are active, then a Supercell Isolation for that area, MEPS [

]PROP/ECI Isolation and VTS Safety Actuation are initiated.

7.5.5.4 High RVZ1 Supercell Radiation (IXP Extraction Cell)

The high RVZ1 supercell radiation signal protects against hot cell equipment leakage or an accident that could potentially result in excess radiation doses to the workers or to the public. The RVZ1 supercell radiation is measured by an analog interface on two different channels, one for each Division A and Division B of ESFAS. When one-out-of-two or more high RVZ1 supercell radiation channels are active, then a Supercell Isolation for that area and VTS Safety Actuation are initiated.

7.5.5.5 High RVZ1 Supercell Radiation (Purification and Packaging Cells)

The high RVZ1 supercell radiation signal protects against hot cell equipment leakage or an accident that could potentially result in excess radiation doses to the workers or to the public. The RVZ1 supercell radiation is measured by an analog interface on two different channels, one for each Division A and Division B of ESFAS. When one-out-of-two or more high RVZ1 supercell radiation channels are active, then a Supercell Isolation for that area is initiated.

7.5.5.6 High MEPS [ ]PROP/ECI Conductivity The high MEPS [ ]PROP/ECI conductivity signal protects against leakage of high radiation solutions into the [ ]PROP/ECI, which is partially located outside the supercell shielding and could potentially result in an excess dose to the workers. The MEPS

[ ]PROP/ECI conductivity is measured by an analog interface on two different channels, one for each Division A and Division B of ESFAS. When one-out-of-two or more high MEPS

[ ]PROP/ECI conductivity channels are active, then a MEPS [ ]PROP/ECI Isolation is initiated.

7.5.5.7 High PVVS Carbon Guard Bed Temperature The high PVVS carbon guard bed temperature signal protects against a fire in the PVVS carbon guard bed impacting the efficiency of the PVVS carbon delay beds. The PVVS carbon guard bed temperature is measured with a temperature interface on two different channels, one for each Division A and Division B of ESFAS. When one-out-of-two or more high PVVS carbon guard bed SHINE Medical Technologies 7.5-23 Rev. 1

5.8 PVVS Carbon Guard Bed Isolation Valves Misaligned VS carbon guard bed isolation valves misaligned protects against inadvertent closure of all VS or N2PS flow paths through the carbon guard beds. PVVS carbon guard bed isolation e indication is measured through a discrete input channel through the respective ESFAS sion the valve is assigned to. When one-out-of-two or more valves show as fully closed in h trains of the carbon guard beds, then a Guard Bed Misalignment Actuation is initiated.

5.9 High PVVS Carbon Delay Bed Exhaust Carbon Monoxide high PVVS carbon delay bed exhaust carbon monoxide signal protects against a fire in the VS delay bed. The PVVS carbon delay bed exhaust carbon monoxide is measured with an log interface on two different channels, one for each Division A and Division B of ESFAS.

en one-out-of-two or more high PVVS carbon delay bed exhaust carbon monoxide channels active, then a Carbon Delay Bed Isolation for the affected bed group is initiated.

5.10 VTS Lift Tank Liquid Detection Switch VTS lift tank liquid detection switch signals protect against an overflow of the vacuum lift

s. The VTS lift tank liquid detection switch signals are measured with a discrete input rface with a single detection signal for each lift tank associated to ESFAS Division A and a undant detection signals common to all lift tanks at the VTS vacuum header associated to FAS Division B. If one-out-of-two or more (Division A and Division B) VTS lift tank liquid ection switch signals are active, then a VTS Safety Actuation is initiated.

5.11 RDS Liquid Detection Switch RDS liquid detection switch signal detects leakage or overflow from other tanks and piping.

RDS liquid detection switch signal is measured with a discrete signal input on two different nnels, one for each Division A and Division B of ESFAS. When one-out-of-two or more RDS id detection switch signal channels are active, then a VTS Safety Actuation is initiated.

5.12 High TPS Exhaust to Facility Stack Tritium high TPS exhaust to facility stack tritium signal protects against a release of tritium from the aust of the TPS glovebox stripper system into the facility ventilation systems. The TPS aust to facility stack tritium is measured with an analog interface on three different channels, for each division of ESFAS. When one-out-of-two or more high TPS exhaust to facility stack m channels are active, then a TPS Isolation is initiated.

5.13 High TPS Glovebox Tritium high TPS glovebox tritium signal protects against a release of tritium from TPS equipment the TPS glovebox. The TPS glovebox tritium is measured with an analog interface on three rent channels, one for each division of ESFAS. When one-out-of-two or more high TPS ebox tritium channels are active, then a TPS Isolation is initiated.

NE Medical Technologies 7.5-24 Rev. 1

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Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 ESFAS Monitored Variables (Sheet 2 of 7)

Variable Analytical Limit Logic Range Accuracy Response Time MEPS [ ]PROP/ECI 0.1 to 50 8.8 micromho/cm 1/2 3 percent 5 seconds conductivity extraction area A micromho/cm MEPS [ ]PROP/ECI 0.1 to 50 8.8 micromho/cm 1/2 3 percent 5 seconds conductivity extraction area B micromho/cm MEPS [ ]PROP/ECI 0.1 to 50 8.8 micromho/cm 1/2 3 percent 5 seconds conductivity extraction area C micromho/cm Carbon guard bed 1 temperature 150°C 1/2 0 to 350°C 2 percent 10 seconds Carbon guard bed 2 temperature 150°C 1/2 0 to 350°C 2 percent 10 seconds PVVS carbon guard bed 1 Discrete Active 1/2 Active/Inactive 1 second inlet isolation valve fully closed input signal PVVS carbon guard bed 1 Discrete Active 1/2 Active/Inactive 1 second outlet isolation valve fully closed input signal PVVS carbon guard bed 2 Discrete Active 1/2 Active/Inactive 1 second inlet isolation valve fully closed input signal PVVS carbon guard bed 2 Discrete Active 1/2 Active/Inactive 1 second outlet isolation valve fully closed input signal Carbon delay bed group 1 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide Carbon delay bed group 2 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide Carbon delay bed group 3 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide SHINE Medical Technologies 7.5-27 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 ESFAS Monitored Variables (Sheet 3 of 7)

Variable Analytical Limit Logic Range Accuracy Response Time Carbon delay bed 4 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide Carbon delay bed 5 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide Carbon delay bed 6 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide Carbon delay bed 7 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide Carbon delay bed 8 20 ppm 1/2 0 to 30 ppm 10 percent 15 seconds exhaust carbon monoxide VTS vacuum header Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction lower lift tank A Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction lower lift tank B Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction lower lift tank C Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction lower lift tank D Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction upper lift tank A1 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction upper lift tank A2 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal SHINE Medical Technologies 7.5-28 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 ESFAS Monitored Variables (Sheet 4 of 7)

Variable Analytical Limit Logic Range Accuracy Response Time VTS extraction upper lift tank B1 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction upper lift tank B2 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction upper lift tank C1 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal VTS extraction upper lift tank C2 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal Target solution storage lift tank Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal Uranium liquid waste lift tank Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal RDS lift tank liquid detection Discrete Active 1/2 Active/Inactive 5.5 seconds switch signal(a) input signal TSV fill lift tank 1 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal TSV fill lift tank 2 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal TSV fill lift tank 3 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal TSV fill lift tank 4 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal SHINE Medical Technologies 7.5-29 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 ESFAS Monitored Variables (Sheet 5 of 7)

Variable Analytical Limit Logic Range Accuracy Response Time TSV fill lift tank 5 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal TSV fill lift tank 6 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal TSV fill lift tank 7 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal TSV fill lift tank 8 Discrete Active 1/2 Active/Inactive 5.5 seconds liquid detection switch signal(a) input signal RDS liquid detection Discrete Active 1/2 Active/Inactive 5.5 seconds switch signal input signal TPS exhaust to 80 µCi/m3 2/3 1 to 100 µCi/m3 10 percent 5 seconds facility stack tritium TPS glovebox tritium 150 Ci/m3 2/3 0.001 to 50,000 Ci/m3 10 percent 5 seconds PVVS flow 5.0 scfm 2/3 1-20 scfm 3 percent 0.5 seconds TSPS dissolution tank 1 Discrete Active 1/2 Active/Inactive 1 second level switch signal input signal TSPS dissolution tank 2 Discrete Active 1/2 Active/Inactive 1 second level switch signal input signal TRPS IU cell 1 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal TRPS IU cell 2 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal SHINE Medical Technologies 7.5-30 Rev. 1

Variable Analytical Limit Logic Range Accuracy Response Time TRPS IU cell 3 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal TRPS IU cell 4 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal TRPS IU cell 5 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal TRPS IU cell 6 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal TRPS IU cell 7 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal TRPS IU cell 8 Discrete Active 1/1 Active/Inactive 500 ms nitrogen purge signal input signal MEPS area A lower Discrete three-way valve supplying Active 1/2 & 1/2 Active/Inactive 1 second input signal position indication(b)

MEPS area A upper Discrete three-way valve supplying Active 1/2 & 1/2 Active/Inactive 1 second input signal position indication(ba)

MEPS area B lower Discrete three-way valve supplying Active 1/2 & 1/2 Active/Inactive 1 second input signal position indication(ba)

MEPS area B upper Discrete three-way valve supplying Active 1/2 & 1/2 Active/Inactive 1 second input signal position indication(ba)

NE Medical Technologies 7.5-31 Rev. 1

Variable Analytical Limit Logic Range Accuracy Response Time MEPS area C lower Discrete three-way valve supplying Active 1/2 & 1/2 Active/Inactive 1 second input signal position indication(ba)

MEPS area C upper Discrete three-way valve supplying Active 1/2 & 1/2 Active/Inactive 1 second input signal position indication(ba)

IXP lower three-way valve Discrete Active 1/2 & 1/2 Active/Inactive 1 second supplying position indication(ba) input signal IXP upper three-way valve Discrete Active 1/2 & 1/2 Active/Inactive 1 second supplying position indication(ba) input signal Discrete UPSS loss of external power Active 1/2 Active/Inactive 1 second input signal Each lift tank is provided one division of liquid detection at the tank; the redundant division of liquid detection is located in the VTS header and serves all lift tanks.

A safety actuation is initiated when both the lower and upper three-way valve supplying position indications show one-out-of-two of the redundant indications are active.

NE Medical Technologies 7.5-32 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 Fail Safe Component Positions on ESFAS Loss of Power (Sheet 1 of 3)

FAIL-SAFE POSITION: CLOSED RVZ1 exhaust isolation dampers RVZ2 supercell area 9 (packaging area 2) inlet isolation dampers RVZ2 exhaust isolation dampers RVZ1 supercell area 9 (packaging area 2) outlet isolation dampers RVZ2 supply train 1 isolation dampers RVZ2 supercell area 10 (IXP area) inlet isolation dampers RVZ2 supply train 2 isolation dampers RVZ1 supercell area 10 (IXP area) outlet isolation dampers RVZ3 supply isolation dampers shipping/receiving IF RVZ TPS ventilation dampers RVZ3 supply isolation dampers shipping/receiving RPF PVVS carbon guard bed bypass valveRLWI PVVS isolation valve RVZ3 supply isolation dampers main RCA ingress/egress MEPS [ ]PROP/ECI A inlet isolation valve RVZ3 supply isolation dampers RPF emergency exit MEPS [ ]PROP/ECI B inlet isolation valve RVZ3 supply isolation dampers IF emergency exit MEPS [ ]PROP/ECI C inlet isolation valve RVZ3 exhaust isolation dampers IF emergency exit MEPS [ ]PROP/ECI A discharge isolation valve RVZ2 TSPS supply damper MEPS [ ]PROP/ECI B discharge isolation valve RVZ1 TSPS exhaust damper MEPS [ ]PROP/ECI C discharge isolation valve RVZ2 supercell area 1 (PVVS area) inlet isolation dampers MEPS A extraction column wash supply valve RVZ1 supercell area 1 (PVVS area) outlet isolation dampers MEPS A extraction column eluent valve RVZ2 supercell area 2 (extraction area A) inlet isolation dampers MEPS A [ ]PROP/ECI wash supply valve RVZ1 supercell area 2 (extraction area A) outlet isolation dampers MEPS A [ ]PROP/ECI eluent valve RVZ2 supercell area 3 (purification area A) inlet isolation dampers MEPS B extraction column wash supply valve RVZ1 supercell area 3 (purification area A) outlet isolation dampers MEPS B extraction column eluent valve RVZ2 supercell area 4 (packaging area 1) inlet isolation dampers MEPS B [ ]PROP/ECI wash supply valve RVZ1 supercell area 4 (packaging area 1) outlet isolation dampers MEPS B [ ]PROP/ECI eluent valve RVZ2 supercell area 5 (purification area B) inlet isolation dampers MEPS C extraction column wash supply valve RVZ1 supercell area 5 (purification area B) outlet isolation dampers MEPS C extraction column eluent valve RVZ2 supercell area 6 (extraction area B) inlet isolation dampers MEPS C [ ]PROP/ECI wash supply valve RVZ1 supercell area 6 (extraction area B) outlet isolation dampers MEPS C [ ]PROP/ECI eluent valve RVZ2 supercell area 7 (extraction area C) inlet isolation dampers IXP recovery column wash supply valve RVZ1 supercell area 7 (extraction area C) outlet isolation dampers IXP recovery column eluent valve RVZ2 supercell area 8 (purification area C) inlet isolation dampers IXP [ ]PROP/ECI wash supply valve RVZ1 supercell area 8 (purification area C) outlet isolation dampers IXP [ ]PROP/ECI eluent valve SHINE Medical Technologies 7.5-33 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 Fail Safe Component Positions on ESFAS Loss of Power (Sheet 2 of 3)

IXP FNHS supply valve Storage and separation system GBSS raffinate isolation valves IXP liquid nitrogen supply valve ATIS header tritium supply isolation valves ATIS header GBSS isolation valves ATIS glovebox exhaust header isolation valves ATIS header GBSS bypass isolation valves TPS process evacuation GBSS isolation valves GBSS RVZ isolation valves TPS glovebox nitrogen supply valves ATIS header return line isolation valves N2PS RVZ2 north header valves TPS process evacuation header isolation valves N2PS RVZ2 south header valves ATIS header deuterium supply isolation valves TSPS RPCS supply cooling valves TSPS RPCS return cooling valve FAIL-SAFE POSITION: OPEN RVZ1 exhaust train 1 blower breakers PVVS carbon delay bed 4 inlet isolation valve RVZ1 exhaust train 2 blower breakers PVVS carbon delay bed 5 inlet isolation valve RVZ2 exhaust train 1 blower breakers PVVS carbon delay bed 6 inlet isolation valve RVZ2 exhaust train 2 blower breakers PVVS carbon delay bed 7 inlet isolation valve RVZ2 supply train 1 blower breakers PVVS carbon delay bed 8 inlet isolation valve RVZ2 supply train 2 blower breakers PVVS carbon delay bed group 1 outlet isolation valves VTS vacuum transfer pump 1 breakers PVVS carbon delay bed group 2 outlet isolation valves VTS vacuum transfer pump 2 breakers PVVS carbon delay bed group 3 outlet isolation valves VTS vacuum transfer pump 3 breakers PVVS carbon delay bed 4 outlet isolation valves VTS vacuum break valves PVVS carbon delay bed 5 outlet isolation valves PVVS blower bypass valves PVVS carbon delay bed 6 outlet isolation valves PVVS carbon guard bed bypass valves PVVS carbon delay bed 7 outlet isolation valves PVVS carbon guard bed 1 inlet isolation valve PVVS carbon delay bed 8 outlet isolation valves PVVS carbon guard bed 1 outlet isolation valve MEPS A extraction feed pump breakers PVVS carbon guard bed 2 inlet isolation valve MEPS B extraction feed pump breakers PVVS carbon guard bed 2 outlet isolation valve MEPS C extraction feed pump breakers PVVS carbon delay bed 1 inlet isolation valve N2PS IU cell header valves PVVS carbon delay bed 2 inlet isolation valve N2PS RPF header valves PVVS carbon delay bed 3 inlet isolation valve SHINE Medical Technologies 7.5-34 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Table 7.5 Fail Safe Component Positions on ESFAS Loss of Power (Sheet 3 of 3)

FAIL-SAFE POSITION: SUPPLYING PVVS carbon delay bed group 1 three-way valves PVVS carbon delay bed 5 three-way valve PVVS carbon delay bed group 2 three-way valves PVVS carbon delay bed 6 three-way valve PVVS carbon delay bed group 3 three-way valves PVVS carbon delay bed 7 three-way valve PVVS carbon delay bed 4 three-way valve PVVS carbon delay bed 8 three-way valve FAIL-SAFE POSITION: DISCHARGING MEPS area A lower three-way valve MEPS area C lower three-way isolation valve MEPS area A upper three-way valve MEPS area C upper three-way isolation valve MEPS area B lower three-way valve IXP upper three-way valve MEPS area B upper three-way valve IXP lower three-way valve SHINE Medical Technologies 7.5-35 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 1 of 32)

Trip Determination SHINE Medical Technologies 7.5-36 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 2 of 32)

Trip Determination SHINE Medical Technologies 7.5-37 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 3 of 32)

Trip Determination SHINE Medical Technologies 7.5-38 Rev. 0

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 4 of 32)

Trip Determination SHINE Medical Technologies 7.5-39 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 5 of 32)

Trip Determination SHINE Medical Technologies 7.5-40 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 6 of 32)

Trip Determination SHINE Medical Technologies 7.5-41 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 7 of 32)

Trip Determination SHINE Medical Technologies 7.5-42 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 8 of 32)

Trip Determination SHINE Medical Technologies 7.5-43 Rev. 0

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 9 of 32)

Trip Determination SHINE Medical Technologies 7.5-44 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 10 of 32)

Trip Determination SHINE Medical Technologies 7.5-45 Rev. 0

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 11 of 32)

Safety Functions SHINE Medical Technologies 7.5-46 Rev. 0

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Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 12 of 32)

Safety Functions SHINE Medical Technologies 7.5-47 Rev. 0

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Chapter 7 - Instrumentation & Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 13 of 32)

Safety Functions SHINE Medical Technologies 7.5-48 Rev. 0

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Chapter 7 - Instrumentation & Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 14 of 32)

Safety Functions SHINE Medical Technologies 7.5-49 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 15 of 32)

Safety Functions SHINE Medical Technologies 7.5-50 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 16 of 32)

Safety Functions SHINE Medical Technologies 7.5-51 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 17 of 32)

Safety Functions SHINE Medical Technologies 7.5-52 Rev. 0

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Chapter 7 - Instrumentation & Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 18 of 32)

Safety Functions SHINE Medical Technologies 7.5-53 Rev. 0

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Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 19 of 32)

Safety Functions SHINE Medical Technologies 7.5-54 Rev. 0

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Chapter 7 - Instrumentation & Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 20 of 32)

Safety Functions SHINE Medical Technologies 7.5-55 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 21 of 32)

Safety Functions SHINE Medical Technologies 7.5-56 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 22 of 32)

Safety Functions SHINE Medical Technologies 7.5-57 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 23 of 32)

Safety Functions SHINE Medical Technologies 7.5-58 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 24 of 32)

Safety Functions SHINE Medical Technologies 7.5-59 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 25 of 32)

Nonsafety Interface Decode SHINE Medical Technologies 7.5-60 Rev. 0

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Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 26 of 32)

Priority Logic SHINE Medical Technologies 7.5-61 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 27 of 32)

Priority Logic SHINE Medical Technologies 7.5-62 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 28 of 32)

Priority Logic SHINE Medical Technologies 7.5-63 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 29 of 32)

Priority Logic SHINE Medical Technologies 7.5-64 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 30 of 32)

Priority Logic SHINE Medical Technologies 7.5-65 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 31 of 32)

Priority Logic SHINE Medical Technologies 7.5-66 Rev. 0

Engineered Safety Features Chapter 7 - Instrumentation & Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 32 of 32)

Legend SHINE Medical Technologies 7.5-67 Rev. 0

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 1 of 24)

SHINE Medical Technologies 7.5-29 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 2 of 24)

Trip Determination SHINE Medical Technologies 7.5-30 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 3 of 24)

Trip Determination SHINE Medical Technologies 7.5-31 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 4 of 24)

Trip Determination SHINE Medical Technologies 7.5-32 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 5 of 24)

SHINE Medical Technologies 7.5-33 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 6 of 24)

Trip Determination SHINE Medical Technologies 7.5-34 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 7 of 24)

Trip Determination SHINE Medical Technologies 7.5-35 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 8 of 24)

SHINE Medical Technologies 7.5-36 Rev. 1

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Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 9 of 24)

SHINE Medical Technologies 7.5-37 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 10 of 24)

SHINE Medical Technologies 7.5-38 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 11 of 24)

SHINE Medical Technologies 7.5-39 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 12 of 24)

Safety Functions SHINE Medical Technologies 7.5-40 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 13 of 24)

SHINE Medical Technologies 7.5-41 Rev. 1

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Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 14 of 24)

SHINE Medical Technologies 7.5-42 Rev. 1

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Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 15 of 24)

SHINE Medical Technologies 7.5-43 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 16 of 24)

Safety Functions SHINE Medical Technologies 7.5-44 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 17 of 24)

Safety Functions SHINE Medical Technologies 7.5-45 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 18 of 24)

Nonsafety Decode SHINE Medical Technologies 7.5-46 Rev. 1

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Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 19 of 24)

SHINE Medical Technologies 7.5-47 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 20 of 24)

Priority Logic SHINE Medical Technologies 7.5-48 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 21 of 24)

Priority Logic SHINE Medical Technologies 7.5-49 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 22 of 24)

Priority Logic SHINE Medical Technologies 7.5-50 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 23 of 24)

Priority Logic SHINE Medical Technologies 7.5-51 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 ESFAS Logic Diagrams (Sheet 24 of 24)

Legend SHINE Medical Technologies 7.5-52 Rev. 1

NE Medical Technologies 7.5-70 Rev. 0

NE Medical Technologies 7.5-71 Rev. 0

NE Medical Technologies 7.5-72 Rev. 0

NE Medical Technologies 7.5-73 Rev. 0

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 Vacuum Transfer System (Sheet 1 of 2)

SHINE Medical Technologies 7.5-72 Rev. 1

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Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 Vacuum Transfer System (Sheet 2 of 2)

SHINE Medical Technologies 7.5-73 Rev. 1

Chapter 7 - Instrumentation and Control Systems Engineered Safety Features Actuation System Figure 7.5 Radiologically Controlled Area Isolation (Sheet 1 of 2)

SHINE Medical Technologies 7.5-74 Rev. 1

Engineered Safety Features Chapter 7 - Instrumentation and Control Systems Actuation System Figure 7.5 Radiologically Controlled Area Isolation (Sheet 2 of 2)

SHINE Medical Technologies 7.5-75 Rev. 1

Chapter 9 - Auxiliary Systems Heating, Ventilation, and Air Conditioning Systems Details of the inspection and testing requirements of safety-related RV systems are provided in Subsection 9a2.1.1.5.

9a2.1.1.2 System Description Radiological Ventilation Zone 1 RVZ1 is divided into two subsystems: RVZ1r and RVZ1e. A flow diagram of RVZ1r is provided in Figure 9a2.1-2. A flow diagram of RVZ1e is provided in Figure 9a2.1-3.

RVZ1r provides cooling for systems within the irradiation unit (IU) cell and the target solution vessel (TSV) off-gas system (TOGS) cell. RVZ1r recirculates, filters, and cools air within the irradiation unit (IU) cell and the TOGS cell. The system includes two fan coil units and associated ductwork and dampers per each set of lU/TOGS cells. Each set of RVZ1r units is located within the confinement boundary for the lU/TOGS cells that it serves. RVZ1r provides sampling, ventilation, and cleanup connections for the primary confinement.

RVZ1e exhausts air from the areas with a high potential for contamination in the facility. The air is filtered and directed out of the SHINE facility through the exhaust stack. The subsystem includes fans, filters, ductwork, dampers, and high efficiency filter banks. It also includes the necessary transfer ductwork to allow makeup from the RCA general area into the exhausted areas.

RVZ1e is designed to maintain ventilation zone 1 areas at a lower pressure than ventilation zone 2 areas. The design inhibits backflow with the use of backflow dampers at the discharge of the RVZ1e and RVZ2e exhaust fans in order to minimize the spread of contamination. RVZ1e ductwork provides sampling locations for radiation detectors, fire detection equipment, stack release monitoring, and an exhaust stack connection point for RVZ2e and the process vessel vent system (PVVS).

The RVZ1 serves the following areas:

• IU cells

• TOGS cells

• Glovebox stripper system (GBSS) process equipment

• Primary closed loop cooling system (PCLS) expansion tank

• Uranium receipt and storage system (URSS) glovebox

• Radioactive liquid waste immobilization (RLWI) shielded enclosure

• Supercell

• Target solution preparation system (TSPS) glovebox

• Target solution dissolution tanks

• Target solution preparation tank

• HVAC enclosures Radiological Ventilation Zone 2 RVZ2 includes three subsystems: RVZ2e, RVZ2s, and RVZ2r. A flow diagram of RVZ2e is provided in Figure 9a2.1-4. A flow diagram of RVZ2s air handling units (AHUs) is provided in Figure 9a2.1-5. A flow diagram of RVZ2s distribution and RVZ2r is provided in Figure 9a2.1-6.

SHINE Medical Technologies 9a2.1-2 Rev. 1

Chapter 9 - Auxiliary Systems Heating, Ventilation, and Air Conditioning Systems

• Preparation room

• Tool crib

• Vestibule

• Primary cooling rooms

• IF general area

• Neutron driver assembly system (NDAS) service cell

• TPS room

• Supercell Radiological Ventilation Zone 3 Under normal operating conditions, RVZ3 transfers air from ventilation zone 4 to ventilation zone 3 then from ventilation zone 3 to ventilation zone 2 via engineered pathways. RVZ3 receives air from RVZ2s in the IF exit labyrinth. A flow diagram of RVZ3 is provided in Figure 9a2.1-7. Under accident conditions, bubble tight dampers close, isolating ventilation zone 2. The design of RVZ3 inhibits backflow within ductwork that could spread contamination.

9a2.1.1.3 System Operation RVZ1e areas draw ambient supply air from adjacent ventilation zone 2 spaces, except for the supercell. During normal operation, areas ventilated by RVZ1e are maintained at negative pressure with respect to their surrounding ventilation zone 2 spaces. The supercell is supplied air directly from RVZ2r. The air supplied to the supercell is exhausted by RVZ1e.

RVZ1e contains redundant fans that are capable of continuous operation. During normal operation, one fan is operating while the other fan is on standby. If the operating fan fails, the standby fan will start automatically.

The exhaust from RVZ1e areas collects in the RVZ1e system duct header and then is drawn through the final filter banks on the mezzanine. These filter banks contain high efficiency particulate air (HEPA) filters and carbon adsorbers upstream of the building isolation dampers.

These filters and adsorbers are equipped with differential pressure monitoring equipment and are periodically monitored by operations personnel. The building isolation dampers are safety-related automatic isolation dampers controlled by ESFAS. These dampers are located at the RCA boundary, upstream of the exhaust fans and exhaust stack.

Negative pressure is maintained in the ductwork to control contamination and maintain pressure gradients. System operation between RVZ1e, RVZ2e, and RVZ2s is coordinated such that the overall airflow and pressure gradients are maintained. The pressure gradients create flow patterns that direct air towards areas of increasing contamination potential. This is maintained by the variable frequency drives (VFDs) on the exhaust fans. Minimum airflow will be maintained during normal system operation.

During upset conditions, affected sections of the RVZ1e, RVZ2s, and RVZ2e ventilation systems are isolated as required for the specific event or indication. Bubble tight dampers close, based on detection of increased radiation. The RVZ1e supply flow path to the supercell includes nonsafety-related HEPA and carbon filters. The RVZ1e exhaust flow path from the supercell includes nonsafety-related HEPA filters and safety-related carbon filters. The remaining RVZ1e flow paths that exhaust confinements for fission products contain non-creditedsafety-related SHINE Medical Technologies 9a2.1-4 Rev. 1

Chapter 9 - Auxiliary Systems Heating, Ventilation, and Air Conditioning Systems HEPA filters and credited,carbon filters. The RVZ1e safety-related, redundant bubble-tight dampers are situated as near to the confinement boundary as practical.

The IU cell exhaust flow path of RVZ1e provides ventilation of the IU cell and TOGS cell via the PCLS expansion tank headspace. This path is equipped with radiation monitoring instrumentation and redundant isolation valves. Between the RVZ1e IU cell radiation instrumentation and RVZ1e IU cell ventilation valves is an isolation lag tank. If radiation measurements exceed predetermined limits, the TSV reactivity protection system (TRPS) initiates an IU Cell Safety Actuation, which closes the RVZ1e IU cell ventilation valves. The isolation lag tank provides an exhaust gas delay time greater than the closing time of the valves.

Upon loss of power, loss of signal, or ESFAS initiation of confinement, dampers seal the affected confinement areas within 30 seconds.

The RVZ1r fan coil units (FCUs) are capable of continuous operations. The RVZ1r recirculates, and cools air within the IU cell and TOGS cell. The IU cell and TOGS cell are established as low leakage boundaries.

RVZ2e fans are capable of continuous operation. RVZ2e exhausts the various normally occupiable rooms within the RCA as well as fume hoods, filters the air via HEPA filter banks and discharges to the facility stack. Exhaust headers are maintained at a negative pressure by the VFD. Negative pressure is maintained in the ductwork to control contamination and maintain pressure gradients. The exhaust from RVZ2 areas collects in the RVZ2 system duct header and then is drawn through final HEPA filters and carbon adsorbers prior to discharge to the exhaust stack.

During normal operation, ventilation zone 2 areas are maintained at negative pressure with respect to RVZ3 airlocks. The speed of the RVZ2e exhaust fans is controlled to maintain a negative pressure setpoint in the RVZ2e exhaust header. Minimum airflow will be maintained during normal system operation.

RVZ2s AHUs are capable of continuous operation. Ventilation zone 2 and portions of ventilation zone 3 areas are directly supplied air via the RVZ2s AHUs. The AHUs supply conditioned, 100 percent outside air. Each AHU contains filters, pre-heat and cooling coils, and supply fans.

The supply system includes redundant AHUs. If a single AHU fails, the standby AHU will start automatically. The AHUs normally supply a constant volume of conditioned air to RVZ2 and RVZ3 areas.

The RVZ2s supply duct contains safety-related automatic isolation dampers controlled by ESFAS. These dampers are located at the RCA boundary.

RVZ2r AHUs are capable of continuous operation. The RVZ2r AHUs further condition the air in the RCA general area to comfort levels.

The RVZ1e and RVZ2e subsystems combine downstream of each subsystems respective filter banks, RCA isolation bubble-tight dampers, and exhaust fans, as shown in Figure 9a2.1-8. The PVVS delay bed discharge is also combined with the RVZ1e and RVZ2e flow downstream of the exhaust fans and upstream of the stack release monitor. The discharge of the stack is approximately 10 feet above the roofline of the facility and will maintain a minimum discharge velocity of 3,000 fpm.

SHINE Medical Technologies 9a2.1-5 Rev. 1

NE Medical Technologies 9a2.1-15 Rev. 0







 

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NE Medical Technologies 9a2.1-15 Rev. 1

Chapter 9 - Auxiliary Systems Fire Protection Systems and Programs 9a2.3 FIRE PROTECTION SYSTEMS AND PROGRAMS 9a2.3.1 FIRE PROTECTION PLAN AND PROGRAM The Fire Protection Plan describes the overall facility Fire Protection Program (FPP). The FPP describes the fire protection organization and responsibilities, design and programmatic approach, and means to limit the probability and consequences of fire at the SHINE facility. The Fire Protection Plan establishes the requirements to be satisfied by the facility fire protection program. This plan establishes a program that represents an integrated effort involving components, procedures, analyses, and personnel used in defining and carrying out activities of fire protection. It includes system and facility design, fire prevention, fire detection, annunciation, confinement, suppression, administrative controls, inspection and maintenance, training, quality assurance, and testing. The established fire protection program elements, systems, structures, and components are subject to the SHINE Quality Assurance Program, as described in the Quality Assurance Program Description. The elements of the FPP work together to satisfy the requirements of applicable regulatory requirements presented in 10 CFR 50.48(a). The FPP is comprised of the following lower tier documents which are developed and maintained as part of the overall FPP.

• Fire Hazards Analysis (FHA);

• Safe Shutdown Analysis;

• Pre-Fire Plans; and

• Administrative controls (e.g., implementing procedures, drawings, calculations, analyses, specifications).

Development of the FPP is informed by the guidance provided in National Fire Protection Association (NFPA) 801, Standard for Fire Protection for Facilities Handling Radioactive Materials (NFPA, 2014). The structure and content of the FPP are based on the precepts of 10 CFR 50.48(a) and NFPA 801. The FPP ensures, through the application of the defense-in-depth concept, that a fire will not prevent the performance of necessary safety-related functions and that radioactive releases to the environment, in the event of fire, will be minimized.

9a2.3.2 DESIGN BASES The concept of defense-in-depth is fundamental to the FPP. Fire protection defense-in-depth for the SHINE facility is defined as follows:

• Prevent fires from starting, including limiting combustible materials;

• Detect, control, and extinguish those fires that do occur to limit consequences; and

• Provide protection for safety-related structures, systems, and components (SSCs) important to safety so that a continuing fire will not prevent the safe shutdown of the irradiation units or cause an uncontrolled release of radioactive material to the environment.

The FPP is developed and implemented to accomplish these criteria.

SHINE Medical Technologies 9a2.3-1 Rev. 1

Chapter 9 - Auxiliary Systems Fire Protection Systems and Programs and impurities allowing for reuse of the solution in the irradiation process. These processes present a possibility of radiological release from processes, with fire presenting an energetic source that can drive release. Radiological release due to fire is typically associated with combustion of radiologically contaminated ordinary combustible materials or fire damage to confinement systems that could allow release of collocated radiological materials.

Uranium oxide and uranium metal are received and stored in the uranium receipt and storage system (URSS) room. Storage of the uranium metal and uranium oxide is in metal storage canisters. Canisters are stored on metal storage racks to ensure a safe configuration of the stored materials. Uranium metal is received in sufficiently massive configurations that it is not pyrophoric.

The TSPS and URSS rooms are protected with automatic fire detection and provided with appropriate portable fire extinguishers for incipient stage fire suppression. Combustible loading in these rooms is maintained low to prevent fire. Fire response using water-based extinguishants is prohibited; elevated floors of the URSS and TSPS fire area are provided to prevent flooding of these rooms.

Irradiation is performed in the irradiation facility (IF). Chemical processing, to extract medical isotopes from the target solution, is performed in the RPF. The irradiation and chemical processing of radiological materials is discussed in detail in Chapter 4.

Once target solution is introduced to the irradiation process, it is contained in pipes and tanks.

These pipes and tanks are located in the IU cells, hot cells, tank vaults, and pipe trenches throughout the IF and RPF. The IU cells, hot cells, tank vaults, and pipe trench structures are constructed of massive steel and concrete barriers to provide radiation shielding. The monolithic construction of these structures provides significant fire separation from the general areas of the IF and RPF. This construction provides protection to the pipes and tanks containing radiological materials. Combustible loading in the spaces within the IU cells, hot cells, tank vaults, and pipe trenches is maintained very low. Combustible materials in these spaces are limited to cable and equipment. Combustible loading in the IF and RPF general areas is maintained low to present a minimal potential for fire. The IF and RPF general areas are equipped with automatic fire detection and provided with portable fire extinguishers to provide incipient fire suppression capability.

Filters contained in the facility ventilation systems that may contain fission products are replaced on a regular basis. Filters are contained in non-combustible ductwork. Areas of the radiologically controlled area (RCA) containing filters are protected with automatic fire detection and portable fire extinguishers. Combustible loading is maintained low in these areas.

The carbon guard beds located in the process vessel vent system (PVVS) are equipped with temperature detection. The guard beds are isolated upon indication of an unacceptable increase in temperature. The carbon delay beds are monitored with in-bed temperature detection and carbon monoxide detectors at eachthe outlet of each carbon delay bed group. The carbon delay beds are equipped with a nitrogen purge line that ismay be used to extinguish hot spots if detected.

Three facility systems are provided to mitigate hydrogen generation due to radiolysis. These systems are the TSV off-gas system (TOGS), PVVS, and nitrogen purge system (N2PS).

SHINE Medical Technologies 9a2.3-4 Rev. 1

Chapter 9 - Auxiliary Systems Handling and Storage of Target Solution and the lift tank is vented to atmosphere. Once pressure is equalized, the solution may be drained, pumped from the lift tank, or transferred to a second stage of lift tanks for additional elevation gain. Liquid transfers using vacuum lift tanks in the RPF are identified in Table 9b.2-1.

• The second method facilitates solution transfers without using a vacuum lift tank. Vacuum from the knockout pot is applied directly to the selected destination tank, and valves in the pathway between the source tank and destination tank are aligned to allow flow. This method is typical for transfers in the RLWS system, the RLWI system, and between laboratory scale processes that are part of the isotope separation process. Direct liquid transfers between tanks facilitated by VTS are identified in Table 9b.2-2.

Two separate vacuum headers are provided based on the liquid being transferred. Tanks designed to contain target solution are provided with vacuum from a separate header than tanks with concentration controls or other tanks or services that are not intended to contain fissile material. The headers are air gapped at the knockout pot as shown in Figure 9b.2-1.

The VTS provides an interface for sampling of solution in the target solution hold tanks, target solution storage tanks, RLWS system tanks, and the radioactive drain system (RDS).

The VTS is the only system used to transport SNM between the RPF and IF. A description of the process used to fill the TSV is provided in Subsection 4a2.6.1. The VTS is one of the systems used to transport solutions of SNM or byproduct material in the RPF. The other systems used to transport solutions containing SNM or byproduct material in the RPF are TSPS, MEPS, IXP, and the RLWI system, which use pumps to provide the motive force to transport the solutions.

Table 9b.2-3 identifies the systems which interface with the VTS.

Figure 9b.2-1 provides a process flow diagram of the VTS.

9b.2.5.3 Instrumentation and Controls Temperature of solution in the source tank is monitored prior to a transfer to ensure that the transfer does not induce the solution to flash in the pipe. Level of each vacuum lift tank is also monitored to allow the process integrated control system (PICS) to control each transfer.

Automatic flow shut-off valves and liquid detection instruments are provided in the VTS to prevent solution from entering the knockout pot. On detection of liquid by these instruments, ESFAS actuates valves that act as vacuum breakers on the knockout pot and trips the breakers on the vacuum pumps, terminating solution transfers. The knockout pot drains to favorable geometry tanks in the RLWS system in the event of high-level alarm. ESFAS also trips the vacuum breaking valves and pump on detection of high radiation in radiological ventilation zone 1 (RVZ1) or on level detection in the RDS. A detailed description of the ESFAS is provided in Section 7.5.

9b.2.5.4 Safety Analysis The VTS is a safety-related system. The VTS structural framework and pipe supports are designed to withstand design basis seismic events. The VTS is classified as Seismic Category I.

SHINE Medical Technologies 9b.2-3 Rev. 1

Chapter 9 - Auxiliary Systems Handling and Storage of Target Solution Figure 9b.2 Vacuum Transfer System Process Flow Diagram (Sheet 6 of 6)

SHINE Medical Technologies 9b.2-15 Rev. 0

Chapter 9 - Auxiliary Systems Handling and Storage of Target Solution Figure 9b.2 Vacuum Transfer System Process Flow Diagram (Sheet 6 of 6)



 

 

SHINE Medical Technologies 9b.2-15 Rev. 1

Chapter 9 - Auxiliary Systems Cover Gas Control in the Radioisotope Production Facility the tank headspace, to the PVVS conditioning and filtration equipment. Gases pass through condensers, cooled with process chilled water, to remove excess heat and reduce absolute humidity of the off-gas.

Condensate is collected in the PVVS condensate tank within the PVVS hot cell, located within the supercell. Condensate may be returned to the target solution staging system (TSSS) tanks as makeup water or to the radioactive liquid waste storage (RLWS) system for waste processing. An in-line heater, the PVVS reheater, downstream of the condenser heats the off-gas back to ambient temperature to reduce the relative humidity. The off-gas then flows through acid adsorber beds, HEPA filters, and the guard beds to neutralize entrained acid droplets or gases, filter particulates, and capture iodine. The gas flows from the hot cell to a below-grade, shielded vault, passing through a series of delay beds packed with carbon to delay the release of fission product noble gases such as xenon and krypton. The eight delay beds are organized into three groups as shown in Figure 9b.6-1. Group 1 includes Delay Beds 1 and 2. Group 2 includes Delay Beds 3, 4, and 5. Group 3 includes Delay Beds 6, 7, and 8. A final set of HEPA filters removes any entrained carbon fines upstream of the blowers, and the treated gases are discharged to the facility stack.

In the event PVVS loses the capability to provide flow, the nitrogen purge system (N2PS) actuates nitrogen flow to the RPF tanks to mitigate hydrogen generation. Upon actuation of the N2PS, the isolation valves at the PVVS intake interface with ventilation zone 2 actuate closed, and the PVVS isolation valve at the radioactive liquid waste immobilization (RLWI) interface actuates closed, to prevent nitrogen backflow. During the nitrogen purge, the PVVS equipment and piping continues to provide the flow path for the off-gas through the RPF. Safety-related bypasses are provided around filtration equipment in the hot cell that could contribute to a blocked pathway and an alternate, safety-related exhaust point to the roof is actuated open.

The branch to the alternate release point is upstream of the PVVS blowers.

Fire protection is provided for the guard beds and delay beds. Temperature instrumentation and carbon monoxide detection are used to monitor for oxidation. The beds may be isolated or purged with nitrogen to smother the reaction. Additionally, operators can attempt to increase the system flow rate to increase convective cooling. The engineered safety features actuation system (ESFAS) automatically isolates an affected delay bed groups on high temperature in the carbon guard beds or high carbon monoxide concentration in the carbon delay bedseffluent gas.

Principal components of the PVVS are identified in Table 9b.6-2.

A process flow diagram of the PVVS is provided in Figure 9b.6-1.

9b.6.1.3 Operational Analysis and Safety Function The PVVS provides confinement of fission products to prevent release of radioactive material.

The PVVS maintains hydrogen concentrations below the lower flammability limit (LFL) to preclude a hydrogen deflagration or detonation, as discussed in Subsection 9b.6.1.3.1. The PVVS passively reduces the concentration of radionuclides in the gaseous effluent to the facility stack, including during postulated transients, as discussed in Subsection 9b.6.1.3.2.

SHINE Medical Technologies 9b.6-2 Rev. 1

Chapter 9 - Auxiliary Systems Cover Gas Control in the Radioisotope Production Facility 9b.6.2.2 System Description N2PS provides back-up sweep gas flow in the form of stored pressurized nitrogen gas.

Downstream pressure is controlled with self-regulating pressure reducing valves with overpressure protection by pressure relief valves. On actuation of the N2PS, nitrogen flows through the irradiation facility (IF) and RPF equipment to ensure the hydrogen concentration is below the LFL. The nitrogen purge flows through the normal PVVS path and filtration equipment, including the delay beds. After exiting the delay beds in PVVS, the nitrogen purge is diverted to a safety-related alternate vent path in case of a downstream blockage. Valves configured to fail open allow the diversion to the alternate vent path. After actuation of the N2PS, the pressurized storage tubes can be refilled by truck deliveries.

A process flow diagram of the N2PS is provided in Figure 9b.6-2.

Purge of an IU Upon loss of normal power as determined by the engineered safety features actuation system (ESFAS) and after a delay or upon loss of normal sweep gas flow in the IU as determined by the TSV reactivity protection system (TRPS), solenoid valves on the nitrogen discharge manifold actuate open, releasing nitrogen into the IU cell supply header. Upon loss of sweep gas flow in any IU cell, nitrogen solenoid isolation valves for the given cell actuate open releasing nitrogen purge gas into the TSV dump tank, and valves in the TOGS actuate open to allow the nitrogen purge gas to flow to the PVVS. The nitrogen purge gas flows through the TSV dump tank, TSV, and TOGS equipment before discharging into PVVS. A flow switch provides indication that nitrogen is flowing to the IU cell. A detailed discussion of the IU Cell Nitrogen Purge is provided in Section 7.4.

Purge of RPF Equipment Upon loss of normal power or loss of normal sweep gas flow through PVVS, as determined by the ESFAS, solenoid valves on the ventilation zone 2 air supply to PVVS fail closed and isolate the sweep gas air flow to the RPF tanks. At the same time, solenoid valves on the nitrogen discharge manifold actuate open, releasing nitrogen into the RPF distribution piping. The nitrogen flows through the RPF equipment in parallel before discharging into PVVS. A flow switch provides indication that nitrogen is flowing to the RPF distribution piping.

Processes that arereceive ventilated byion air from the PVVS during normal conditions are also ventilated by N2PS during deviations from normal operation. In the RPF, the N2PS ventilates tanks in the TSSS, RLWS system, radioactive liquid waste immobilization (RLWI) system, radioactive drain system (RDS), molybdenum extraction and purification system (MEPS),

iodine and xenon purification and packaging (IXP) system, and VTS. A detailed discussion of the RPF Nitrogen Purge is provided in Section 7.5.

9b.6.2.3 Operational Analysis and Safety Function In the event of a loss of normal power, loss of sweep gas flow through PVVS, or loss of sweep gas flow through any TOGS, the N2PS controls the buildup of hydrogen which is released into the primary system boundary and tanks or other volumes which contain fission products to ensure that the system and confinement boundaries are maintained.

SHINE Medical Technologies 9b.6-5 Rev. 1

Table 9b.6 Process Vessel Vent System Interfaces (Sheet 1 of 2)

Interfacing System Interface Description gineered safety features The ESFAS monitors the operation of the process vessel vent system uation system (ESFAS) (PVVS) and. ESFAS actuates the nitrogen purge system (N2PS) on failure ofand opens the PVVS blowersfiltration bypass on low ventilation flow through PVVS, and isolates the delay beds on high carbon monoxide concentration.

ine and xenon purification The PVVS ventilates tanks in the IXP.

packaging (IXP) system lybdenum extraction and The PVVS ventilates the molybdenum eluate hold tank and MEPS ification system (MEPS) condensate tank.

ogen purge system (N2PS) The N2PS provides sweep gas flow through the PVVS piping and filtration equipment on loss of normal power or normal flow in PVVS.

rmal electrical power supply The NPSS is distributed to the PVVS blowers, the PVVS reheater, and tem (NPSS) ancillary equipment.

cess integrated control The PICS controls the PVVS and monitors PVVS instrument signals.

tem (PICS) duction facility biological The PFBS provides shielding to workers from the PVVS. PVVS eld (PFBS) equipment is located in a hot cell and in a below-grade vault.

dioactive drain system The PVVS ventilates the RDS tanks.

S) dioactive liquid waste The PVVS ventilates the RLWS tanks. The PVVS drains condensate rage (RLWS) system water to the RLWS for disposal.

dioactive liquid waste The PVVS ventilates the immobilization feed tank.

mobilization (RLWI) system dioisotope process facility The RPCS provides cooling capacity to the PVVS for the off-gas ling system (RPCS) condensers.

diological ventilation zone 1 The PVVS blowers discharge into a header shared by RVZ1 to the Z1) facility stack. Some PVVS components are located in a hot cell, which is ventilated by RVZ1.

diological ventilation zone 2 The PVVS intake removes air from RVZ2 for use as sweep gas across Z2) the RPF tanks.

ck release monitoring The SRMS monitors the discharge from the PVVS delay beds to the tem (SRMS) stack.

ndby generator system The SGS provides nonsafety-related backup power to PVVS S) components.

get solution staging system The PVVS ventilates the TSSS tanks to mitigate hydrogen generation.

SS) The PVVS may also transfer condensate water to the TSSS for reuse in the irradiation cycle.

NE Medical Technologies 9b.6-8 Rev. 1

Chapter 9 - Auxiliary Systems Cover Gas Control in the Radioisotope Production Facility Figure 9b.6 PVVS Process Flow Diagram SHINE Medical Technologies 9b.6-12 Rev. 0

Chapter 9 - Auxiliary Systems Cover Gas Control in the Radioisotope Production Facility Figure 9b.6 PVVS Process Flow Diagram SHINE Medical Technologies 9b.6-12 Rev. 1

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Chapter 9 - Auxiliary Systems Other Auxiliary Systems

• Reduce radiation exposure during operation in accordance with applicable guideline exposures set forth in 10 CFR 20.

The RLWI system piping is designed and constructed in accordance with ASME B31.3, Process Piping (ASME, 2013). Nonsafety-related components within the RLWI system are designed to standards satisfying system operation.

9b.7.3.2 System Description The RLWI system solidifies blended liquid waste to a form suitable for shipping and disposal. The RLWI system removes selected isotopes, as needed, from the blended liquid waste and then immobilizes the wastes for ultimate disposal. The headspace cover gas in the immobilization feed tank is swept by the PVVS or nitrogen from the nitrogen purge system (N2PS) to maintain radiolytically generated hydrogen gas below the lower flammability limit.

The blended liquid waste sources and radionuclide and uranium concentrations are described in Subsection 9b.7.4.

The immobilization feed tank is filled from the liquid waste blending tanks on a batch basis by vacuum suction applied to the immobilization feed tank from the vacuum transfer system (VTS).

Positive displacement pumps transfer the contents of the immobilization feed tank [

]PROP/ECI, and meter the tank contents to a disposable waste drum.

The waste drums are prefilled with measured amounts of dry, powdered solidification agent in accordance with the process control program (PCP). The prefilled drum is transferred into an enclosure for contamination control. The radiological ventilation zone 1 exhaust subsystem (RVZ1e) equipment processes air from the enclosure through a high efficiency particulate air (HEPA) and carbon filter before discharging to the facility stack. Transfer of the prefilled waste drum inside the enclosure is by remote handling equipment and positioners.

The liquid waste drum is filled with blended liquid waste and mixed. Subsequent to fill and mixing, the fill and vent ports are disengaged. The drum is then remotely transferred to a curing station where the mixed contents of the waste drum hardens prior to removal from the enclosure. The cured drum is remotely transferred into a shielded cask and transported to the material staging building for further radiological decay, as needed, prior to shipment to a licensed disposal facility.

Remote sampling for waste characterization is performed in the RLWS prior to solidification activities. Radiation measurements are performed on the solidified waste drum prior to shipment in the material staging building to verify it meets shipping dose rate requirements.

Table 9b.7-1 identifies the systems which interface with the RLWI system. Figure 9b.7-1 provides a process flow diagram of the RLWI system.

9b.7.3.3 Operational Analysis and Safety Function Liquid waste solidification is performed in accordance with a PCP. The RLWI system is sized to process approximately double the routine liquid waste generation rate from the RPF.

SHINE Medical Technologies 9b.7-6 Rev. 1

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Chapter 9 - Auxiliary Systems Other Auxiliary Systems 9b.7.4.2 System Description The RLWS system collects, stores, blends, conditions, and meters liquid wastes for processing by the RLWI for solidification. Included in the blended liquid wastes is PVVS condensate which can also be recycled through the target solution staging system (TSSS) to minimize waste generation. The RLWS system can also transfer liquid via a normally removed pipe spool to a target solution storage tank for sampling and verification against target solution parameters. The headspace in each RLWS system tank is swept with air by the PVVS or by the nitrogen purge system (N2PS) to remove the potential accumulation of radiolytically generated hydrogen gas.

Laboratory waste is preconditioned and manually processed separately from the RLWS system.

Liquid waste collected, blended, and stored by the RLWS system includes:

• Uranium liquid waste, with uranium concentrations potentially exceeding 25 gU/l. This waste is located in the uranium liquid waste tanks.

• Radioactive liquid waste, with negligible uranium concentration with respect to criticality safety (< 1 gU/l). This waste is stored in the liquid waste collection tanks.

• Blended liquid waste, with low uranium concentrations (< 25 gU/l). Blended waste may originate from uranium liquid waste, radioactive liquid waste, or any combination of the two.

Uranium liquid waste tanks are geometrically favorable annular tanks similar in design to those used in TSSS. These tanks include two redundant overflow lines which drain to the radioactive drain system (RDS) in the event of an overfill. The uranium liquid waste tanks are connected in series to ensure high concentration uranium-bearing waste (greater than 25 gU/l) is not inadvertently transferred to the non-geometrically favorable liquid waste blending tanks.

The uranium liquid waste tanks are configured to operate in series, with the first tank receiving wastes from the following sources:

• Mo-99 extraction column washes

• Iodine-131 (I-131) recovery column washes

• Spent target solution

• Solution in radioactive drain sump tanks

• Solution in VTS knockout pot

• Decontamination liquid waste

• PVVS condensate tank

• Solution from the second uranium liquid waste tank via gravity drain from the uranium liquid waste lift tank The remaining liquid wastes are collected in four liquid waste collection tanks designed and sized to maximize storage capacity. The liquid waste collection tanks are configured to receive wastes from the following sources:

• [ ]PROP/ECI effluent and washes

• MEPS condensate and purification waste

• [ ]PROP/ECI washes SHINE Medical Technologies 9b.7-8 Rev. 1

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Chapter 9 - Auxiliary Systems Other Auxiliary Systems Process liquid wastes determined by sampling to be less than 25 gU/l are consolidated into the eight liquid waste blending tanks. The liquid waste blending tanks are configured to receive wastes from the following sources:

• Second uranium liquid waste tank

• PVVS condensate tank

• Liquid waste collection tanks Uranium liquid waste is combined with the radioactive liquid waste and/or PVVS condensate in the liquid waste blending tanks for homogeneous radionuclide and uranium concentrations in the RLWI system feed.

The tanks are sized to maximize decay time thereby minimizing dose rates from the immobilized waste product. Each uranium liquid waste tank has a minimum nominal capacity of

[ ]PROP/ECI, and each of the liquid waste collection tanks and liquid waste blending tanks has a minimum nominal capacity of 600 gallons.

The RLWS system piping is designed and constructed in accordance with ASME B31.3, Process Piping (ASME, 2013).

Table 11.2-6 provides the chemical composition and radiological properties of liquid waste streams.

Table 9b.7-2 identifies the systems which interface with the RLWS system. Figure 9b.7-2, Figure 9b.7-3, and Figure 9b.7-4 provide process flow diagrams of the RLWS system.

9b.7.4.3 Operational Analysis and Safety Function Solenoid valves isolating radioactive liquid flow paths fail to the normally isolated positions.

Solenoid valves isolating the sweep gas flow path fail to the normally aligned flow from air to the PVVS vent header. Operators align the RLWS tank inlets and outlets based on procedures using information from the position indicators and instrumentation.

The RLWS system tanks, valves, and piping are located in shielded tank vaults, valve pits, and pipe trenches within the RPF. Section 11.1 provides a description of the radiation protection program, and Section 4b.2 provides a detailed description of the PFBS.

Sampling of RLWS system tank contents for pH verifies that waste solution acidity is maintained or is adjusted as necessary. Solution composition can be adjusted via the reagent addition line.

Sampling of waste tanks is performed by vacuum lift to a hot cell, where samples are remotely obtained.

Following a period of decay to reduce dose rates, waste is transferred from the liquid waste blending tanks to the RLWI system immobilization feed tank. Subsection 9b.7.3 provides a detailed discussion of the RLWI system.

RLWS system piping connected to, installed over, or installed adjacent to safety-related equipment is designed to meet seismic requirements because its failure could damage safety-related equipment such that the equipment would be prevented from performing its safety function.

SHINE Medical Technologies 9b.7-9 Rev. 1

Chapter 9 - Auxiliary Systems Other Auxiliary Systems Table 9b.7 Radioactive Liquid Waste Immobilization System Interfaces (Sheet 1 of 2)

Interfacing System Interface Description Nitrogen purge system (N2PS) N2PS supplies sweep gas to RLWI tank headspace cover gas upon a loss of power or loss of normal sweep gas flow to remove potential accumulation of radiolytically generated hydrogen gas.

Process vessel vent system The PVVS supplies sweep gas to the immobilization feed (PVVS) tank headspace to remove potential accumulation of radiolytically generated hydrogen gas.

The immobilization feed tank cover gas and waste drum vent both discharge via a common header to the PVVS vent header.

Radioactive liquid waste storage Immobilization feed tank receives radioactive liquid waste (RLWS) system from the RLWS system.

Vacuum transfer system (VTS) Suction from VTS provides the motive force for waste liquid transfer from the blending tanks to the immobilization feed tank.

Radiological ventilation zone 1 The RLWI shielded enclosure is ventilated by RVZ1.

(RVZ1)

Ventilation of the solidification skid enclosure and glovebox removes residual decay heat in the immobilization feed tank and the heat of hydration from the solidification process in the glovebox.

Radiological ventilation zone 2 The RVZ2 is the source of air supply to the shielded (RVZ2) enclosure through RVZ2 filtration equipment.

The RVZ2 is the source of air for the vacuum break between the VTS suction header and the drum fill head vacuum test tank.

Facility nitrogen handling system The FNHS provides instrument-grade pressurized nitrogen (FNHS) to immobilization feed tank level instrumentation.

Process integrated control The components of the RLWI system are controlled and system (PICS) monitored by the PICS.

Normal electrical power supply The components of the RLWI system are powered by the system (NPSS) NPSS.

SHINE Medical Technologies 9b.7-20 Rev. 1

Chapter 9 - Auxiliary Systems Other Auxiliary Systems Figure 9b.7 RLWI System Process Flow Diagram SHINE Medical Technologies 9b.7-30 Rev. 0

Chapter 9 - Auxiliary Systems Other Auxiliary Systems Figure 9b.7 RLWI System Process Flow Diagram SHINE Medical Technologies 9b.7-29 Rev. 1

Chapter 9 - Auxiliary Systems Other Auxiliary Systems Figure 9b.7 RLWS Liquid Waste Blending Tanks Process Flow Diagram SHINE Medical Technologies 9b.7-33 Rev. 0

Chapter 9 - Auxiliary Systems Other Auxiliary Systems Figure 9b.7 RLWS Liquid Waste Blending Tanks Process Flow Diagram SHINE Medical Technologies 9b.7-32 Rev. 1

Chapter 13 - Accident Analysis Accident-Initiating Events and Scenarios material by the cooling water, minimizes the potential for criticality in the PCLS and dose to workers or the public.

Because of the system characteristics and preventative controls in place, further analysis is not required.

Scenario 5 - Failure in the TOGS Causes High Pressure in the TSV during Fill Mode A failure by the TOGS to control pressure, and a resulting pressure increase during TSV filling operations, may result in a backflow of target solution. Target solution may flow through the fill line into the TSV fill lift tank, into the VTS header, and into the VTS buffer tank. This failure potentially results in radiological exposures to workers or a criticality accident in non-favorable-geometry components in the VTS.

The protection in place for this scenario is the configuration of the TSV fill line to prevent significant volume of target solution from backflowing from the TSV into the VTS lift tank. The TSV fill line connects to the TSV with an air gap. The connection is located at the approximate elevation of the TSV overflow lines. The fill line is sloped to allow it to drain after fill operations have occurred. Therefore, no significant volume of target solution will backflow from the TSV to the VTS lift tank in the event of pressurization of the TSV.

Defense-in-depth measures are also present to mitigate this scenario, which include:

• the VTS vacuum valve to lift tank closes from high liquid level in the lift tank, and

• a drain valve for the buffer tank opens and drains to the RDS sumpRLWS if a high level in the liftbuffer tanks is detected.

Because of the system characteristics and preventative controls in place, further analysis is not required.

Scenario 6 - Target Solution Leakage within a Valve Pit A pipe or valve failure in the valve pit may be caused by overpressurization due to thermal expansion of target solution in an isolated section of piping. This pipe or valve failure results in leakage of target solution from the system into the valve pit, which subsequently could result in:

(1) increased worker or public dose, or (2) a criticality accident in the valve pit. The protections in place to mitigate the consequences of target solution leakage within a valve pit are: (1) drip pans and drains to the radioactive drain system (RDS), which prevent accumulation of solution within the valve pit and prevent criticality, and (2) valve pit shielding and confinement for fission products that could result from leakage, reducing potential dose to workers and the public.

Because this piping is potentially located in either the IF or the RPF, this event and associated dose consequences is further analyzed in Chapter 13b.

13a2.1.4.3 Accident Consequences The release of target solution from the PSB to the light water pool or connected process systems results in potential radiological exposure to workers and the public. The accident consequences associated with the mishandling or malfunction of target solution are evaluated further in Subsection 13a2.2.4.

SHINE Medical Technologies 13a2.1-20 Rev. 1

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences The public dose was generally calculated over a 30-day interval at the site boundary. SThe scenarios resulting in the release of tritium from the TPS glovebox uses a 10-day release interval because it is expected that tritium recovery can be accomplished within this time frame. The /Q values are calculated at the nearest point along the site boundary and at the nearest resident location. The maximum calculated value over all directions of the 50th percentile /Q was used for both receptor locations. A ground release was used as the release point.

The environmental and meteorological conditions used to develop the atmospheric dispersion factors are discussed in Section 2.3.

Conservatism Additional areas of conservatism included in the determination of radiological consequences include:

• Conservative TSV power history and operational cycle: The TSV power history was derived from nearly continuous TSV operation over a [ ]PROP/ECI period at a power level that exceeds the design power level by ten percent. No credit was taken for medical isotope extraction activities that normally occur during the operation of the SHINE facility.

• Conservative statistical bounding of nuclide inventory: Due to inherent uncertainties in MCNP5, multiple unique sets of results were run through ORIGEN-S to determine the nuclide inventories. The nuclide inventories were analyzed such that a 95 percent confident 95th percentile upper bound was determined for each nuclide. These uncertainties on individual nuclides, 0 to 35 percent, were added to the safety basis inventory to account for the uncertainties inherent to the methods used.

• Conservative estimation of nuclide decay (linear interpolation in lieu of exponential decay): Analyses which account for the decay of nuclides between time steps use linear interpolation in lieu of exponential decay, which increases the available radionuclide inventory at the intervening points.

• Condensation was conservatively neglected in the LPF model.

Uncertainties Uncertainty in the radionuclide inventory was evaluated using statistical modeling to account for uncertainties associated with the use of Monte Carlo N-Particle Transport Code (MCNP)

(LANL, 2011) in the SHINE Best Estimate Neutronics Model (BENM). The modeling produced a nuclide-dependent multiplication factor ranging from approximately 0 to 35 percent increase in the nuclide inventory per nuclide. For the radionuclides which were increased, the average increase was approximately 2.5 percent, and the total estimated increase in inventory was approximately 1 percent. The unweighted uncertainty associated with the multiplication factors was approximately 12 percent. Given that the majority of radionuclides either did not receive an increase or received an increase less than 10 percent and that the multiplication factor only increased the inventory this uncertainty is considered to be negligible.

Based on the results of the validation activities for the LPF model, described below, there is no additional uncertainty associated with the LPF model used in the analysis.

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

SHINE Medical Technologies 13a2.2-5 Rev. 1

Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences 13a2.2.3.6 Radiological Consequences Because the postulated reduction in cooling events do not exceed any design limits or cause damage to the PSB, there are no radiological consequences to workers or the public from a reduction in cooling event.

13a2.2.4 MISHANDLING OR MALFUNCTION OF TARGET SOLUTION The bounding scenario analyzed as a design basis accident (DBA) for mishandling or malfunction of target solution is a loss of the PSB integrity which results in a release of target solution into the IU cell. This scenario is described in Subsection 13a2.1.4.2 as Scenario 1b.

13a2.2.4.1 Initial Conditions The TSV is operating at 110 percent of its design power limit at the time of the initiating event.

Additional initial accident conditions are described in Subsection 13a2.1.4.1.

13a2.2.4.2 Initiating Event The accident sequence is initiated by a catastrophic loss of PSB integrity. Potential causes of the initiating event are discussed in Subsection 13a2.1.4.1.

13a2.2.4.3 Sequence of Events It is assumed that the primary confinement boundary is intact and performs a mitigation function with respect to radionuclide transport from the IU cell to the IF. The primary confinement boundary components are designed to maintain their integrity under postulated accident conditions and are maintained in accordance with the facility configuration management and maintenance 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, and through the PCLS expansion tank to RVZ1e.

4. Detection of airborne radiation in RVZ1e actuates the primary confinement boundary isolation valves and an IU trip within 20 seconds of detection. A sufficient time delay is provided by the holdup volume in RVZ1e to prevent radioactive gases from exiting through RVZ1e prior to isolation.

5. The radioactive material is then dispersed throughout the IF and exits the facility to the environment through building penetrations.

6. Detection of airborne radiation in RVZ1e actuates the primary confinement boundary isolation valves and an IU trip within 20 seconds of detection. A sufficient time delay is provided by design to prevent significant radioactive gases from exiting through this path prior to isolation.

7. 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.

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences

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 upon actuation of the radiation alarms.

No operator actions are taken or required to reach a stabilized condition or to mitigate dose consequences.

Following the failure of the PSB, it is assumed that the MAR is instantly well-mixed with the light water pool. Gases immediately evolve out of the pool and into the IU cell gas space. For the purposes of the accident analysis, it is assumed that the N2PS is operating and causes pressurization of the IU cell. Radiation transport is driven by pressure-driven flow between the IU cell and the IF. Reduction in the MAR occurs during the release due to adsorption of iodine onto the IU cell walls and other surfaces until equilibrium conditions are established. The majority of the MAR is transported to the IF through leakage through the primary confinement boundary.

Transport to the environment occurs through leakage around penetrations in the RCA boundary.

Safety Controls The safety controls credited for mitigation of the dose consequences for this accident are:

• Primary confinement boundary

• Ventilation radiation monitors

• Ventilation isolation mechanisms

• Holdup volume in the RVZ1e 13a2.2.4.4 Damage to Equipment Chemical and radiological contamination may occur to systems within the IU cell. The contamination does not affect the safety function of the affected systems.

Following isolation of the primary confinement boundary, leakage between the IU cell and the IF is driven primarily by pressure-driven flow caused by N2PS. The IU cell sealing is a significant contributor to the function of the primary confinement boundary and will maintain its function under accident conditions.

The light water pool is required to act as a passive heat sink to remove decay heat from the irradiated target solution. The light water pool is constructed with a stainless steel liner surrounded by concrete and maintains the light water pool water inventory and will not be affected by the release of target solution.

13a2.2.4.5 Radiation Source Terms The 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 between cycles. The power level used for the analysis is 137.5 kW, which is 110 percent of design operating power. The entire radionuclide inventory in the TSV is instantaneously released to the light water pool and dispersed uniformly throughout the pool.

SHINE Medical Technologies 13a2.2-13 Rev. 1

Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences 13a2.2.6.2 Initiating Event A seismic event is the initiating event for a tritium release into multiple IU cells. All NDAS accelerators experience vacuum boundary component failures and cause a pressurized release of tritium and SF6 gas into the eight IU cells simultaneously as a result of the seismic event.

The initial accident conditions for each IU cell are the same to those accident conditions involving the NDAS of a single IU cell, as described in Subsection 13a2.1.12.1.

13a2.2.6.3 Sequence of Events The 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 and through the PCLS expansion tanks to RVZ1e.

4. Detection of high accelerator TPS interface system (ATIS) mixed gas return line pressure actuates the primary confinement boundary isolation valves and irradiation unit trips within 20 seconds of detection. A sufficient time delay is provided by the holdup volume in RVZ1e to prevent radioactive gases from exiting through RVZ1e prior to isolation.

5. Tritium migrates to the IF through the IU cell plugs and is released to the environment.

6. Detection of high accelerator TPS interface system (ATIS) mixed gas return line pressure actuates the primary confinement boundary isolation valves and irradiation unit trips within 20 seconds of detection. Tritium is assumed to exit by this path until isolation has occurred.

7. 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.

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 alarms.

For the first 20 seconds, the mixed gas flow to RVZ1e equalizes the excess pressure and results in a direct release to the IF. After the primary confinement boundary is isolated, rRadiation transport is driven primarily by barometric breathing between the IU cell and the IF.

The safety-related SSCs in the IU cell do not fail during a seismic event, but the NDAS and its internal components are not safety-related and cannot be relied upon to remain intact following a design basis earthquake.

No operator actions are taken or required to reach a stabilized condition or to mitigate dose consequences.

Safety Controls The safety controls credited for mitigation of the dose consequences for this accident are:

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences

• Primary confinement boundary

• IU Cell Safety Actuation on high ATIS mixed gas return line pressure

• Ventilation isolation mechanisms

• Holdup volume in the RVZ1e It is assumed that the primary confinement is intact and performs a mitigation function with respect to radionuclide transport from the IU cells to the IF. The primary confinement boundary components are designed to maintain their integrity under postulated accident conditions and are maintained in accordance with the facility configuration management and maintenance systems.

13a2.2.6.4 Damage to Equipment Failure of the NDAS vacuum boundary does not cause subsequent damage to equipment. While the NDAS vacuum boundary integrity is not seismically qualified to maintain integrity, the NDAS is designed to maintain structural integrity during and following a design basis earthquake.

After the initial IU cell pressurization has reached equilibrium, leakage between the IU cells and the IF is driven primarily by barometric breathing. The leakage between the cells and the IF is not impacted by the accident sequence.

13a2.2.6.5 Radiation Source Terms The initial MAR for this scenario is a total of [ ]PROP/ECI of tritium from all of the neutron driver assemblies.

The accident source term development is discussed in Section 13a2.2. The LPF model values used in the source term development for the public and worker doses are provided in Table 13a2.2-1 and Table 13a2.2-2, respectively.

13a2.2.6.6 Radiological Consequences The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The results of the determination are provided in Table 13a3-1 and meet the accident dose criteria.

13a2.2.7 MISHANDLING OR MALFUNCTION OF EQUIPMENT The bounding scenario analyzed for mishandling or malfunction of equipment events is a loss of the PSB integrity which results in a release of off-gas into the TOGS cell. This scenario is described in Subsection 13a2.1.7.2 as Scenario 1.

13a2.2.7.1 Initial Conditions Initial accident conditions are described in Subsection 13a2.1.7.1.

13a2.2.7.2 Initiating Event The accident sequence is initiated by a failure of the PSB in the TOGS within the TOGS cell. The cause of the initiating event is discussed in Subsection 13a2.1.7.

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences 13a2.2.7.3 Sequence of Events The accident sequence proceeds as follows:

1. A failure of the PSB in the TOGS causes a release of noble gases and iodine into the TOGS cell.

2. The radioactive material is confined by the primary confinement boundary, which is described in Section 6a2.2.

3. Some radioactive material is transported into the IF through penetrations in the confinement boundary.

4. The radioactive material is then dispersed throughout the IF and exits to the environment through building penetrations.

5. Detection of high radiation in the RVZ1e ventilation from the IU cell actuates ventilation dampers and minimizes the transport of radioactive material to the environment. The assumed response time for RVZ1e ventilation is 20 seconds from detection of high airborne radiation. A sufficient time delay is provided by designthe holdup volume in RVZ1e to prevent significant radioactive gases from exiting through this pathRVZ1e 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.

A portion of the gaseous iodine is adsorbed onto the cell walls, while the majority of the available MAR is transported to the IF through pressure-driven flow caused by the N2PS and leakage through the primary confinement boundary. Transport to the environment occurs through penetrations in the RCA boundary.

Safety Controls The safety controls credited for mitigation of the dose consequences for this accident are:

• Primary confinement boundary

• Ventilation radiation monitors

• Ventilation isolation mechanisms

• Holdup volume in the RVZ1e It is assumed that the primary confinement boundary is intact and performs a mitigation function with respect to radionuclide transport from the TOGS cell to the IF. The primary confinement boundary components are designed to maintain their integrity under postulated accident conditions and are maintained in accordance with the facility configuration management and maintenance systems.

SHINE Medical Technologies 13a2.2-18 Rev. 1

assumed that the primary confinement is intact and performs a mitigation function with pect to radionuclide transport from the IU cell to the IF. The primary confinement is designed aintain its integrity under postulated accident conditions and is maintained in accordance the facility configuration management and maintenance programs.

1. The initiating event is a vacuum boundary component failure in the NDAS, which instantaneously releases tritium and SF6 gas into the IU cell.

2. The IU cell becomes slightly pressurized due to the mass of released SF6 gas.

3. Tritium is transported into the IF through penetrations in the confinement boundary and through the PCLS expansion tank to RVZ1e.

4. Detection of high ATIS mixed gas return line pressure actuates the primary confinement boundary isolation valves and an irradiation unit trip within 20 seconds of detection. A 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 ATIS mixed gas return line pressure actuates the primary confinement boundary isolation valves and an irradiation unit trip within 20 seconds of detection.

Tritium is assumed to exit by this path until isolation has occurred.

7. 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.

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.

the first 20 seconds, a direct release to the environment is modeled. After the primary finement boundary is isolated, rRadiation transport is primarily driven by barometric breathing ween the IU cell and the IF. The accident duration used in this analysis is 10 days, after which assumed that recovery actions will have occurred to stop further release and dispersion of oactive material.

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

• Primary confinement boundary (IU cell plugs and seals)

• IU Cell Safety Actuation on high ATIS mixed gas return line pressure

• IU cell ventilation isolations

• Holdup volume in the RVZ1e 2.2.12.1.4 Damage to Equipment ure of the NDAS vacuum boundary does not cause subsequent damage to equipment.

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences After the initial IU cell pressurization has reached equilibrium, leakage between the IU cells and the IF is driven primarily by barometric breathing. The leakage paths between the cells and the IF are not impacted by the accident sequence.

13a2.2.12.1.5 Radiation Source Terms The initial MAR for this scenario is [ ]PROP/ECI of tritium from the neutron driver assembly in the IU cell.

The accident source term development is discussed in Section 13a2.2. The LPF model values used in the source term development for the public and worker doses are provided in Table 13a2.2-1 and Table 13a2.2-2, respectively.

13a2.2.12.1.6 Radiological Consequences The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The accident duration used in this analysis is 10 days, after which it is assumed that recovery actions will have occurred to stop further release and dispersion of radioactive material.

The radiological consequences of this accident scenario are provided in Table 13a3-1 and meet the accident dose criteria.

13a2.2.12.2 Tritium Release into the Tritium Purification System Glove Box A release of the tritium inventory from the TPS is analyzed as a DBA. This accident is described in Subsection 13a2.1.12.3 as TPS Scenario 1. This analysis establishes bounding radiological conditions for a release of tritium due to a TPS process deflagration, release of tritium to the facility stack, and release of tritium from the tritium storage bed.

13a2.2.12.2.1 Initial Conditions Initial conditions for facility-specific events are described in Subsection 13a2.1.12.1.

13a2.2.12.2.2 Initiating Event An event causes a break in the tritium piping and vessels such that the uncontrolled release of the entire tritium in-process inventory occurs within the tritium confinement boundary. The tritium confinement boundary is described in detail in Section 6a2.2. Potential causes of the initiating event are discussed in Subsection 13a2.1.12.3.

13a2.2.12.2.3 Sequence of Events It is assumed that the tritium confinement boundary is intact and performs a mitigation function with respect to radionuclide transport from the TPS to the IF. The tritium confinement boundary components are designed to maintain their integrity under postulated accident conditions and are maintained in accordance with the facility configuration management and maintenance programs.

SHINE Medical Technologies 13a2.2-28 Rev. 1

2. For the first 120 seconds, tritium escapes from the glovebox to the IF at 10 percent of the maximum GBSS flow rate.

3. The glovebox ventilation shuts down after 120 seconds due to the glovebox tritium monitors.

4. During the 30 seconds after the initiating event, the TPS room vents to the IF at an elevated rate due to the facility RVZ2 ventilation system.

5. The RVZ2 ventilation damper from the TPS room isolates after 30 seconds due to the glovebox tritium monitors.

6. The radioactive material is then dispersed throughout the IF and exits the facility to the environment through building penetrations.

7. Personal dosimeters, local radiation alarms, and alarms in the facility control room notify facility personnel of radiation leakage.

8. Facility personnel evacuate the immediate area within 10 minutes upon actuation of the radiation area monitor alarms.

oughout the accident sequence, the leakage rate between the TPS glovebox and the TPS m is constant. After the TPS room ventilation is isolated, radiation transport is driven by air hange between the TPS glovebox and the IF. Transport to the environment occurs through A boundary leak paths. The accident duration used in this analysis is 10 days, after which it is umed that recovery actions will have occurred to stop further release and dispersion of oactive material.

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

• TPS room ventilation isolations

• GBSS ventilation isolations

• TPS glovebox tritium radiation monitors

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

• TPS glovebox gas space inerted with nitrogen

• TSP glovebox minimum volume prevents deflagration conditions 2.2.12.2.4 Damage to Equipment ure of the TPS piping and vessels does not cause subsequent damage to other equipment.

2.2.12.2.5 Radiation Source Terms initial MAR for this scenario is 236,000 curies of tritium from the TPS equipment in the TPS ebox.

accident source term development is discussed in Section 13a2.2. The LPF model values d in the source term development for the public and worker doses are provided in le 13a2.2-1 and Table 13a2.2-2, respectively.

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences Table 13a2.2 Summary of Radiation Transport Terms (Public)

ARF x LPF Nobles Iodine Non-volatiles Tritium Tritium Accident Category (30-day) (30-day) (30-day) (10-day) (30-day)

Maximum Hypothetical Accident (Subsection 13a2.2.1) 9.98E-01 9.98E-01 0 N/A N/A Mishandling or Malfunction of Target Solution 9.98E-01 1.22E-01 9.058.39E-07 N/A N/A (Subsection 13a2.2.4)

External Events (Subsection 13a2.2.6) N/A N/A N/A 1.75E-01N/A 3.66E-01 Mishandling or Malfunction of Equipment 9.98E-01 5.72E-01 0 N/A N/A (Subsection 13a2.2.7)

Facility-Specific Events (Subsection 13a2.2.12)

• Tritium Release into an IU Cell N/A N/A N/A 1.75E-01N/A 3.66E-01

• Tritium Release into the Tritium Purification N/A N/A N/A 1.537E-01 N/A System Glovebox

• Tritium Release into the Irradiation Facility N/A N/A N/A 9.30E-01N/A 1.00E+00 (Header Release)

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Chapter 13 - Accident Analysis Accident Analysis and Determination of Consequences Table 13a2.2 Summary of Radiation Transport Terms (Worker)

ARF x LPF (10-minute)

Accident Category Nobles Iodine Non-volatiles Tritium Maximum Hypothetical Accident (Subsection 13a2.2.1) 1.19E-02 1.19E-02 0 N/A Mishandling or Malfunction of Target Solution (Subsection 13a2.2.4) 8.24E-03 4.03E-05 9.69E-11 N/A External Events (Subsection 13a2.2.6) N/A N/A N/A 91.47E-021 Mishandling or Malfunction of Equipment (Subsection 13a2.2.7) 1.19E-02 1.17E-02 0 N/A Facility-Specific Events (Subsection 13a2.2.12)

• Tritium Release into an IU Cell N/A N/A N/A 91.47E-021

• Tritium Release into the Tritium Purification System Glovebox N/A N/A N/A 1.852.89E-04

• Tritium Release into the Irradiation Facility (Header Release) N/A N/A N/A 1.00E+00 SHINE Medical Technologies 13a2.2-33 Rev. 1

Chapter 13 - Accident Analysis Summary and Conclusions Table 13a3 Irradiation Facility Accident Dose Consequences Public Worker Dose Dose TEDE TEDE Accident Category (Bounding Scenario) (mrem) (mrem)

Maximum Hypothetical Accident (Subsection 13a2.2.1)

• TOGS failure with complete PVVS blockage 3676 4800 Insertion of Excess Reactivity (Subsection 13a2.2.2) No consequences Reduction in Cooling (Subsection 13a2.2.3) No consequences Mishandling or Malfunction of Target Solution (Subsection 13a2.2.4)

• Primary system boundary leak into an IU cell 65 1480 Loss of Off-Site Power (LOOP) (Subsection 13a2.2.5) No consequences External Events (Subsection 13a2.2.6) 59106 411930 Mishandling or Malfunction of Equipment (Subsection 13a2.2.7) 234 4760 Large Undamped Power Oscillations (Subsection 13a2.2.8) No consequences Detonation and Deflagration affecting the Primary System Boundary No consequences (Subsection 13a2.2.9)

Unintended Exothermic Chemical Reactions other than Detonation No consequences (Subsection 13a2.2.10)

System Interaction Events (Subsection 13a2.2.11) No consequences Facility-Specific Events (Subsection 13a2.2.12)

• Tritium Release into an IU Cell 713 513616

• Tritium Release into the Tritium Purification System Glove 32635 4671 Box

• Tritium Release into the Irradiation Facility (Header Release) 257 3140 SHINE Medical Technologies 13a3-2 Rev. 1

uce the overall risk of the evaluated scenarios to within acceptable limits. For accident uences that are not prevented and have mitigative controls applied, the radiological or mical consequences were quantitatively evaluated to demonstrate the effectiveness of the cted controls. The radiological consequences of accidents that were selected for additional luation are further evaluated in Section 13b.2. The accident analysis for chemical exposures rovided in Section 13b.3.

.1.2.1 Maximum Hypothetical Accident in the RPF MHA in the RPF is a fire in a carbon guard bed with degraded performance of the nstream carbon delay beds.

initiating event for this accident is ignition of transient combustibles or exothermic chemical ction in the bed resulting the formation of a hot spot and eventually a fire. Redundant mperature sensorsindication normally detect the fires an increase in temperature and initiate solation of the affected carbon guard bed is isolated. The carbon guard bed releases its ntory to the downstream carbon delay beds which are normally credited with adsorbing 99 cent of the released iodine. For the MHA, the carbon delay beds are assumed to be operating reduced efficiency of 95 percent. TheAnalysis demonstrates that the carbon guard bed is umed to be isolated on high exit gas temperature after 100 percent of the material-at-risk is ased from the guard bed to prevent damage offire will not propagate to the carbon delay

s. This scenario is described further in Subsection 13b.2.1.

.1.2.2 External Events external initiating events for the RPF that were evaluated include seismic events, tornados igh winds, small aircraft impacts, flooding, fires, and chemical releases. The SHINE main duction facility is designed to withstand credible external events, as described in section 13a2.1.6. External events were considered as potential IEs for a number of accident narios that fall within the other accident categories. The design basis seismic event results in ential chemical consequences, as described below and in Section 13b.3.

esign basis flooding event could result in potential flooding of internal vaults, trenches, and

, as well as the URSS and TSPS rooms. Flooding of the areas that contain fissile material uces the margin to criticality and challenges the double-contingency principle. Water intrusion these areas is minimized by sealed covers for the below-grade locations and by elevated m floors for the URSS and TSPS rooms. The local maximum probable precipitation event ulting in a 100-year flood will not exceed the first-floor entrance elevations, providing itional margin.

ernal event scenarios are further described in Subsection 13b.2.3.

.1.2.3 RPF Critical Equipment Malfunction ical equipment malfunctions in the RPF were evaluated as part of the accident analysis.

tiple scenarios were identified as having potential radiological consequences and were cted for additional evaluation. The identified scenarios are described below. For each nario, the controls that act to reduce the likelihood or consequences of the accident are listed.

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Radioisotope Production Facility Chapter 13 - Accident Analysis Accident Analysis Methodology For scenarios that require mitigative controls, the radiological consequence assessments for limiting exposures are presented in Subsection 13b.2.4.

Scenario 1 - Spill of Target Solution in the Supercell (MEPS Column Misalignment)

A spill of target solution in the supercell has the potential to release radioactive gases, aerosol, and particulates into the hot cell and ventilation system. Potential consequences of spilled target solution in the supercell include radiological dose. To mitigate the impact of spilled target solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (radiological ventilation zone 1 [RVZ1]) is equipped with radiation monitors that provide a signal to the engineered safety features actuation system (ESFAS) to isolate the affected cell and limit the amount of target solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, hot cell inlet (radiological ventilation zone 2

[RVZ2]) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers, and ESFAS-controlled MEPS extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to limit the amount of target solution introduced into the affected hot cell. This scenario is further described in Subsection 13b.2.4.1.

Scenario 2 - Spill of Target Solution in the Supercell (MEPS Overpressurization)

A spill of target solution in the supercell caused by MEPS overpressurization has the potential to release radioactive gases, aerosol, and particulates into the hot cell and ventilation system.

Potential consequences of spilled target solution in the supercell include radiological dose. To prevent deflagrations, which may cause overpressure events, the nitrogen purge system (N2PS) automatically actuates on a failure of PVVS and is relied on to dilute hydrogen concentrations in tanks and vessels in the RPF. Additionally, target solution extraction pumps are provided pressure relief mechanisms. To mitigate the impact of spilled target solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell and limit the amount of target solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers, and ESFAS-controlled.

MEPS extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to limit the amount of target solution introduced into the affected hot cell. This scenario is further described in Subsection 13b.2.4.1.

Scenario 3 - Spill of Molybdenum Eluate Solution in the Supercell (Overfill or Drop of Rotovap Flask)

A spill of the molybdenum solution in the MEPS purification cell may result in the release of radioactive gases, aerosol, and particulates into the hot cell and ventilation system. Potential consequences of spilled eluate solution in a hot cell include radiological dose. To mitigate the impact of spilled eluate solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers. The resulting sequence of events for this scenario is analogous to the MEPS eluate spill described in Subsection 13b.2.4.2.

SHINE Medical Technologies 13b.1-5 Rev. 1

Radioisotope Production Facility Chapter 13 - Accident Analysis Accident Analysis Methodology Scenario 4 - Spill of Target Solution in the Supercell (IXP Column Misalignment)

A spill of target solution in the IXP extraction cell caused by IXP column misalignment has the potential to release radioactive gases, aerosol, and particulates into the hot cell and ventilation system. Potential consequences of spilled target solution in supercell include radiological dose.

To mitigate the impact of spilled target solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell and limit the amount of target solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers, and ESFAS-controlled IXP extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to limit the amount of target solution introduced into the affected hot cell. This scenario is further described in Subsection 13b.2.4.1.

Scenario 5 - Spill of Target Solution in the Supercell (IXP Overpressurization)

A spill of target solution in the IXP extraction cell caused by IXP column overpressurization has the potential to release radioactive gases, aerosol, and particulates into the hot cell and ventilation system. Potential consequences of spilled target solution in the supercell include radiological dose. To prevent hydrogen deflagrations, which may cause overpressure events, the N2PS automatically actuations on a failure of PVVS and is relied on to dilute hydrogen concentrations in tanks and vessels in the RPF. Additionally, target solution extraction pumps are provided pressure relief mechanisms. To mitigate the impact of spilled target solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell and limit the amount of target solution introduced into the cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers, and ESFAS-controlled IXP extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to limit the amount of target solution introduced into the affected hot cell. This scenario is further described in Subsection 13b.2.4.1.

Scenario 6 - Spill of Target Solution in the Supercell (Liquid Nitrogen Leak in IXP Hot Cell)

A liquid nitrogen leak in the IXP hot cell may damage components in the supercell and result in a spill of target solution in the hot cell, with the potential to release radioactive gases, aerosol, and particulates into the supercell and ventilation system. Potential consequences of spilled target solution in the supercell include radiological dose. To mitigate the impact of spilled target solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell and limit the amount of target solution introduced into the cell, hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and ESFAS-controlled IXP extraction pump breakers, VTS vacuum transfer pump breakers, and VTS vacuum break valves are provided to limit the amount of target solution introduced into the affected hot cell. This scenario is further described in Subsection 13b.2.4.1.

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Radioisotope Production Facility Chapter 13 - Accident Analysis Accident Analysis Methodology Scenario 7 - Spill of Iodine Solution in the Supercell (Overfill or Drop of Iodine Solution Bottle)

A spill of iodine eluate solution in the IXP cell results in the release of radioactive gases, aerosols, and particulates into the hot cell and ventilation system. Potential consequences of iodine solution spilling inside the IXP cell include radiological dose. To mitigate the impact of spilled iodine solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers. The resulting sequence of events for this scenario is analogous to the MEPS eluate spill described in Subsection 13b.2.4.2.

Scenario 8 - Spill of Target Solution in the Pipe Trench from a Single Pipe A spill of target solution in the pipe trench results in the release of radioactive gases, aerosols, and particulates into the pipe trench. Potential consequences of spilled target solution inside the pipe trench include radiological dose. To mitigate the impact of spilled target solution, the following controls are applied: the pipe trench is designed as a confinement boundary, RDS drains prevent the accumulation of target solution in the pipe trench, the RDS sump tank liquid detection sensor detects fluid in-leakage and provides a signal to ESFAS to stop any in-process transfers of solution within the facility via opening ESFAS-controlled VTS vacuum transfer pump breakers and VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are equipped with radiation monitors that provide a signal to ESFAS to isolate the building ventilation supply and exhaust dampers on high radiation. This scenario is further described in Subsection 13b.2.4.3.

Scenario 9 - Spill of Target Solution in the Pipe Trench from Multiple Pipes A spill of target solution in the pipe trench results in the release of radioactive gases, aerosols, and particulates into the hot cell and ventilation system. Potential consequences of spilled target solution in the pipe trench include radiological dose. To prevent the failure of multiple target solution-carrying pipes, the pipes are seismically qualified. This scenario is further described in Subsection 13b.2.4.3.

Scenario 10 - Spill of Target Solution in a Tank Vault (Hold Tank Leak or Rupture)

A spill of target solution in a tank vault results in a release of radioactive gases, aerosols, and particulates into the tank vault. Potential consequences of target solution spilling in the tank vault include radiological dose. To mitigate the impact of spilled target solution, the following controls are applied: the tank vault is designed as a confinement boundary, RDS drains prevent the accumulation of target solution in the tank vault, the RDS sump tank liquid detection sensor detects fluid in-leakage and provides a signal to ESFAS to stop any in-process transfers of solution within the facility via opening ESFAS-controlled VTS vacuum transfer pump breakers and VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are equipped with radiation monitors that provide a signal to ESFAS to isolate the building ventilation supply and exhaust dampers on high radiation. This scenario is further described in Subsection 13b.2.4.4.

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Radioisotope Production Facility Chapter 13 - Accident Analysis Accident Analysis Methodology Scenario 11 - Spill of Target Solution in a Tank Vault (Hold Tank Deflagration)

A spill of target solution in a tank vault caused by a hold tank deflagration results a release of radioactive gases, aerosols, and particulates into the tank vault. Potential consequences of target solution spilling in the tank vault include radiological dose. To prevent a deflagration in the hold tank, the N2PS automatically actuates on a failure of PVVS and is relied upon to dilute hydrogen concentrations. To mitigate the impact of spilled target solution, the following controls are applied: the tank vault is designed as a confinement boundary, RDS drains prevent the accumulation of target solution in the tank vault, the RDS sump tank liquid detection sensor detects fluid in-leakage and provides a signal to ESFAS to stop any in-process transfers of solution within the facility via opening ESFAS-controlled VTS vacuum transfer pump breakers and VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are equipped with radiation monitors that provide a signal to ESFAS to isolate the building ventilation supply and exhaust dampers on high radiation. This scenario is further described in Subsection 13b.2.4.4.

Scenario 12 - Spill of Target Solution in a Tank Vault (Seismic Event)

A spill of target solution in a tank vault caused by a seismic event results in a release of radioactive gases, aerosols, and particulates into the tank vault. Potential consequences of target solution spilling in the tank vault include radiological dose. To prevent seismically caused damage, the process tanks and piping are designed to withstand earthquakes. To mitigate the impact of spilled target solution, the following controls are applied: the tank vault is designed as a confinement boundary, RDS drains prevent the accumulation of target solution in the tank vault, the RDS sump tank liquid detection sensor detects fluid in-leakage and provides a signal to ESFAS to stop any in-process transfers of solution within the facility via opening ESFAS-controlled VTS vacuum transfer pump breakers and VTS vacuum break valves, and the RVZ1 and RVZ2 building exhausts are equipped with radiation monitors that provide a signal to ESFAS to isolate the building ventilation supply and exhaust dampers on high radiation. This scenario is further described in Subsection 13b.2.4.4.

Scenario 13 - Spill of Molybdenum Eluate in the Supercell (Deflagration)

Loss of sweep gas flow from PVVS through the eluate tank in the supercell may result in a buildup of hydrogen in the eluate tank and a subsequent deflagration. A spill of molybdenum eluate caused by a deflagration in the eluate tank results in the release radioactive gases, aerosols, and particulates into the hot cell. Potential consequences of spilled eluate solution in a hot cell include radiological dose. To prevent deflagrations in tanks and vessels in the RPF, the N2PS automatically actuates upon a loss of PVVS and is relied upon to dilute hydrogen concentrations. To mitigate the impact of spilled eluate solution, the following controls are applied: the supercell is designed as a confinement boundary, hot cell exhaust ventilation (RVZ1) is equipped with radiation monitors that provide a signal to ESFAS to isolate the affected cell, hot cell outlet (RVZ1) ducts are equipped with carbon filters, and hot cell inlet (RVZ2) and outlet (RVZ1) ventilation ducts are equipped with ESFAS-controlled redundant isolation dampers. This scenario is further described in Subsection 13b.2.4.2.

Scenario 14 - Target Solution Leaking out of the Supercell (MEPS [ ]PROP/ECI Leak)

A leak in the MEPS extraction column [ ]PROP/ECI allows target solution to enter the

[ ]PROP/ECI . Potential consequences of target solution leaking into the [

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nario 19 - Heavy Load Drop onto a Tank Vault or Pipe Trench Cover Block ane failure or operator error resulting in a heavy load drop on a tank vault or pipe trench er block causes a damage to the cover block and internal equipment. Potential sequences of a heavy load drop include radiological dose. To prevent damage to a cover k, the cover blocks have been designed to withstand a heavy load drop. This scenario was luated qualitatively and is not described in Section 13b.2 because the accident sequence is vented.

.1.2.4 RPF Inadvertent Nuclear Criticality lear criticality safety (NCS) in the RPF is accomplished through the use of criticality safety trols to prevent criticality during normal and abnormal conditions. Each process that involves use, handling, or storage of SNM is evaluated by the SHINE nuclear criticality safety staff er the requirements of the NCS program. The results of the criticality safety evaluations are rporated into the ISA. Radiological consequences of criticality accidents are not included in accident analysis because preventative controls are used to ensure criticality events are ly unlikely. Further discussion of the criticality safety bases for RPF processes is included in tion 6b.3.

.1.2.5 RPF Fire RPF was evaluated for internal fire risks based on the fire hazards analysis (FHA). The FHA uments the facility fire areas and each area was individually evaluated for fire risks. Internal lity fires are generally evaluated as an initiating event for the release of radioactive material are included in the scenarios evaluated in Section 13a2.1 and this section. Two unique narios are described below and evaluated in detail in Section 13b.2.

main production facility maintains a facility fire protection plan to reduce the risks of fires, as cribed in Section 9a2.3.

nario 1 - PVVS Carbon Delay Bed Fire upset or malfunction in the PVVS (high moisture or high temperature) results in ignition of the bon media in a delay bed. A fire in the carbon delay bed results in a release of the captured oactive material into the PVVS downstream of the delay bed and to the environment via the lity exhaust stack. A release to the environment results in radiological exposure to the public.

ease of radioactive material in excess of acceptable levels is prevented by the carbon delay carbon monoxide (CO) detectors, which provide a signal to ESFAS to close the PVVS bon delay bed isolation valves for the affected carbon delay bed group and bypass the cted group in the event of high CO concentration indicative of a fire in thea bed. Releases to RPF are further mitigated by the process confinement boundary (carbon delay bed vaults).

scenario is further described in Subsection 13b.2.6.1.

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upset or malfunction in the PVVS (high moisture or high temperature) results in ignition of the bon media in a guard bed. A fire in the guard bed results in a release of the captured oactive material into the PVVS downstream of the guard bed, into the delay beds, and to the ironment via the facility exhaust stack. A release to the environment results in radiological osure to the public. Release of radioactive material in excess of acceptable levels is vented by the carbon guard bed temperature sensors, which provide a signal to ESFAS to e the PVVS carbon guard bed isolation valves for the affected carbon guard bed in the event igh temperature indicative of a fire in the bed. Additionally, the downstream carbon delay s, which reduce or delay radioisotope release. Releases to the RPF are further mitigated by supercell confinement boundary. This scenario is further described in Subsection 13b.2.6.2.

.1.2.6 RPF Chemical Accidents ential chemical exposures in the RPF were evaluated to identify chemical hazards and essary controls. The bounding inventories of chemicals used in the main production facility e identified and evaluated for exposure to workers and the public. Only exposure to uranium e presents a risk that exceeds the applicable evaluation criteria. This scenario is discussed her in Section 13b.3.

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eral design basis accidents described in Section 13b.1 result in a release of radioactive erials into or outside the controlled areas of the facility.

analyses in this section evaluate the applicable radiological consequences of these idents to demonstrate than an individual located in the unrestricted area following the onset of ostulated accidental release of licensed material would not receive a radiation dose in excess 00 mrem total effective dose equivalent (TEDE) for the duration of the accident.

iological consequences to workers are also evaluated and are shown to not exceed 5 rem DE during the accident.

.2.1 MAXIMUM HYPOTHETICAL ACCIDENT IN THE RPF maximum hypothetical accident (MHA) in the radioisotope production facility (RPF) is a fire carbon guard bed with degraded carbon delay bed efficiency. It is postulated that percent of the radionuclide inventory is released from the guard bed and flows downstream the carbon delay beds and is then released to the facility stack. The performance of the bon delay beds is assumed to be degraded to 95 percent. The automatically mitigated release a credible carbon guard bed fire is discussed in Subsection 13b.2.6.2.

al Conditions process vessel vent system (PVVS) is operating normally, with nominal flow through one bon guard bed.

affected carbon guard bed contains radioactive iodine from RPF process streams. The erial-at-risk (MAR) in this scenario is a combination of gases from eight irradiation units (IU),

various modifiers applied to account for decay and operational sequencing.

ating Event upset or malfunction in the PVVS results in high moisture or high temperature flow through carbon guard bed. The high moisture or high temperature results in ignition of the carbon rd bed absorber media. Potential initiating events for this scenario are discussed further in section 13b.1.2.1.

uence of Events

1. Ignition of one of the carbon guard beds occurs, resulting in an exothermic release of 100 percent of the stored radioactive material to the gas piping downstream of the guard bed.

2. The downstream carbon delay beds adsorb 95 percent of the radioactive material.

3. The radioactive material exiting the carbon delay beds is released to the environment through the PVVS and facility stack.

4. Automatic isolation of the guard bed occurs before the gas temperature exiting the bed reaches 180°C to protect the downstream carbon delay beds from damage. The maximum gas temperature entering the carbon delay beds is 130°C, which is insufficient to propagate the fire condition to the carbon delay bed. Temperature instrumentation in NE Medical Technologies 13b.2-1 Rev. 1

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Chapter 13 - Accident Analysis Analyses of Accidents with Radiological Consequences the carbon guard bed detects increased temperature and the carbon guard beds are isolated. However, the isolation of the carbon guard bed is not credited for limiting the release of radioactive materials from the carbon guard bed in this scenario.

Damage to Equipment The occurrence of fire damages the affected carbon guard bed and eliminates its ability to function. No other damage to the PVVS system or its components occurs.

Transport of Radioactive Material The methods used to calculate radioactive material transport are described in Section 13a2.2.

The leak path factor (LPF) model terms used in this accident are provided in Table 13b.2-1. For this accident, the release of material from the guard bed is assumed to be instantaneous and is transported to the environment at an increased rate.

Radiation Source Terms The initial MAR for this scenario is a portion of the iodine gas inventory evolved from target solution during normal operations. Development of the accident source term for this scenario is discussed further in Section 13a2.2.

The iodine gas inventory is produced by fission and decay of fission products and continuously evolved from the target solution and through the target solution vessel (TSV) off-gas system (TOGS) during operations. Partitioning fractions for iodine gas are used to describe the quantities of iodine in solution that move to the RPF. Removal of iodine by the TOGS zeolite beds are credited for all gases that are transported to the RPF. 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. The 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.

Radiological Consequences The radioactive material is contained in the PVVS system and does not result in worker exposure. The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The results of the determination are provided in Table 13b.2-2.

13b.2.2 LOSS OF ELECTRICAL POWER Loss of off-site power (LOOP) was evaluated in the accident analysis as an initiating event for a number of critical equipment malfunction scenarios. A facility-wide LOOP results in automatic actuation of multiple facility engineered safety features, which act to ensure the risk associated with radiological or chemical releases is reduced to within acceptable limits. The facility-wide LOOP does not result in system or component failures within the RPF that result in unacceptable radiological or chemical consequences. The facility-wide LOOP is further discussed in Subsection 13a2.1.5 and Subsection 13a2.2.5.

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event causes a break in the MEPS piping between the extraction pump discharge and the action column. The break downstream of the pump discharge causes spray and osolization of the target solution without any extraction of isotopes by the extraction column.

ential initiating events for this scenario and analogous scenarios for the iodine and xenon fication (IXP) system cell are discussed further in Subsection 13b.1.2.3; Scenarios 1, 2, 4, 5, 6.

uence of Events

1. A break in the MEPS piping between the extraction pump discharge and the extraction column occurs.

2. Aerosolized target solution sprays from the break into the hot cell, releasing radioactive material into the hot cell and causing the cell to become pressurized to the nominal pressure of the cell drain loop seal.

3. Radiation detectors in the hot cell exhaust ventilation detect high airborne radiation and cause the engineered safety features actuation system (ESFAS) to shut down the vacuum transfer system (VTS), shut down the extraction pump, and isolate the hot cell ventilation.

4. Leakage of radioactive material from the hot cell to the RPF and the environment through the ventilation dampers occurs, resulting in radiological consequences to workers and the public.

maximum volume of spilled target solution in this accident scenario is limited by the volume he vacuum lift tanks and installed piping of the MEPS. The ESFAS shutdown of the VTS vents additional target solution from entering the hot cell after high radiation has been ected. The analyzed volume of target solution for this scenario is 30 liters, which is servatively based on the volume of two vacuum lift tanks plus additional pipe volume.

controls credited for mitigation of the dose consequences for this accident are:

• Supercell confinement boundary

• Hot cell radiological ventilation zone 1 (RVZ1) radiation monitors

• Hot cell RVZ1 outlet carbon filters (radioiodine)

• Inlet (radiological ventilation zone 2 [RVZ2]) and outlet (RVZ1) ventilation isolation dampers

• MEPS or IXP extraction pump breakers

• VTS vacuum transfer pump breakers

• VTS vacuum break valves

• ESFAS Supercell Isolation function

• ESFAS VTS Safety Actuation function mage to Equipment leak of target solution in the supercell does not cause subsequent damage to equipment.

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Chapter 13 - Accident Analysis Analyses of Accidents with Radiological Consequences Transport of Radioactive Material The methods used to calculate radioactive material transport are described in Section 13a2.2.

The LPF model terms used in this accident are provided in Table 13b.2-1.

Radiation Source Terms The initial MAR for this scenario is 30 liters of target solution from the IU at [ ]PROP/ECI post-shutdown. The action of the TOGS during this [ ]PROP/ECI period removes more than 67 percent of the iodine present in the solution at shutdown. It is conservatively assumed that 35 percent of the post-shutdown iodine inventory is released to the supercell during the accident. Additionally, partitioning fractions are applied to the noble gases present in target solution. Development of the accident source term for this scenario is discussed further in Section 13a2.2.

Radiological Consequences The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The results of the determination are shown in Table 13b.2-2.

13b.2.4.2 Spill of Eluate Solution in the Supercell Initial Conditions At the time of the initiating event, eluate solution in the MEPS eluate tank is spilled onto the floor of the hot cell, releasing radioactive material into the hot cell atmosphere.

Initiating Event An event causes the failure of the MEPS eluate tank, which results in a spill of eluate solution.

Potential initiating events for this scenario and analogous scenarios for the purification and IXP cells are discussed further in Subsection 13b.1.2.3; Scenarios 3, 7, and 13.

Sequence of Events

1. A break in the MEPS eluate tank occurs.

2. Eluate solution spills from the tank into the hot cell, releasing radioactive material into the hot cell and causing the cell to become pressurized to the nominal pressure of the cell drain loop seal.

3. Radiation detectors in the hot cell exhaust ventilation detect high airborne radiation and cause ESFAS to isolate hot cell ventilation.

4. Leakage of radioactive material from the hot cell to the RPF and the environment through the ventilation dampers occurs, resulting in radiological consequences to workers and the public.

The controls credited for mitigation of the dose consequences for this accident are:

• Supercell confinement boundary

• Hot cell RVZ1 radiation monitors

• Hot cell RVZ1 outlet carbon filters (radioiodine)

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Chapter 13 - Accident Analysis Analyses of Accidents with Radiological Consequences

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.

5. Liquid detection in the RDS sump tank sends a signal to ESFAS, which opens the VTS vacuum pump breakers and the VTS vacuum break valves to stop transfers of solution in the RPF. It is assumed that up to 70 liters of target solution are released in the course of the accident.

The controls credited for mitigation of the dose consequences for this accident are:

• Process confinement boundary (trench or pit cover and cover seal)

• RDS drains to the RDS sump tank

• RDS sump tank liquid detection sensors

• VTS vacuum transfer pump breakers

• VTS vacuum break valves

• ESFAS VTS Safety Actuation function Additional controls described in Subsection 13b.1.2.3 are provided but not credited in the dose analysis.

Damage to Equipment The leak of target solution into the pipe trench does not cause further damage to equipment.

Transport of Radioactive Material The methods used to calculate radioactive material transport are described in Section 13a2.2.

The LPF model terms used in this accident are provided in Table 13b.2-1.

Radiation Source Terms The initial MAR for this scenario is 70 litersa batch of target solution from the IU at [

]PROP/ECI post-shutdown. The action of the TOGS during this [ ]PROP/ECI period removes more than 67 percent of the iodine present in the solution at shutdown. It is conservatively assumed that 35 percent of the post-shutdown iodine inventory is released to the pipe trench during the accident. The volume used in this analysis is based on the available volume of the VTS lift tanks and associated piping. It is assumed that this is maximum volume that can be released from the system before the ESFAS actuation shuts down the VTS. Additionally, partitioning fractions are applied to the noble gases present in target solution. Development of the accident source term for this scenario is discussed further in Section 13a2.2.

Radiological Consequences The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The results of the determination are provided in Table 13b.2-2.

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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.

nsport of Radioactive Material methods used to calculate radioactive material transport are described in Section 13a2.2.

LPF model terms used in this accident are provided in Table 13b.2-1. For this accident, the ase of material for the first 30 seconds is assumed to be instantaneous and is transported to environment at an increased rate. Following isolation valve actuation, the transport occurs at duced rate.

iation Source Terms initial MAR for this scenario is a portion of the noble gas inventory evolved from target tion during normal operations. Development of the accident source term for this scenario is ussed further in Section 13a2.2.

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Chapter 13 - Accident Analysis Analyses of Accidents with Radiological Consequences The noble gas inventory is produced by decay of fission products and continuously evolved from the target solution and through the TOGS during operations. The MAR uses selected time intervals for the most recent purges (i.e., [ ]PROP/ECI) to account for the processing capacity of target solution batches in the supercell for the combined eight IU. The gases accumulate in the carbon delay bed and decay. The MAR assumes the combined noble gas inventory produced by eight IUs over approximately [ ]PROP/ECI of irradiation with the most recent purges of [

]PROP/ECI. Partitioning fractions for noble gases are used to describe the quantities of noble gases in solution that move to the RPF to account for removal during movement of solution.

Radiological Consequences The radioactive material is contained in the PVVS system and does not result in worker exposure. The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The results of the determination are provided in Table 13b.2-2.

13b.2.6.2 PVVS Carbon Guard Bed Fire Initial Conditions The PVVS is operating normally, with nominal flow through a carbon guard bed.

The affected carbon guard bed contains iodine from RPF process streams. The MAR in this scenario is a combination of iodine from eight IUs with various modifiers applied to account for decay and processing capacity of target solution batches in the supercell.

Initiating Event An upset or malfunction in the PVVS results in high moisture or high temperature flow through the carbon guard bed. The high moisture or high temperature results in ignition of the carbon guard bed adsorber material. Potential initiating events are discussed further in Section 13b.1.2.5, Scenario 2.

Sequence of Events

1. Ignition of the carbon guard bed occurs, resulting in an exothermic release of stored radioactive material to the PVVS downstream of the guard bed.

2. Radioactive material is captured by the downstream carbon delay bed and filtered. One percent of the released radioactive material is released through PVVS and the facility stack to the environment.

3. Incipient fire conditions are detected by redundant temperature indication in the guard bed, which send an actuation signal to the ESFAS.

4. ESFAS isolates the carbon guard bed using installed isolation valves. Valve closure is accomplished within 30 seconds of detection of elevated temperature above the actuation setpoint.

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.

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Chapter 13 - Accident Analysis Analyses of Accidents with Radiological Consequences The components credited for mitigation of the dose consequences for this accident areis:

• PVVS carbon guard bed temperature indicators

• PVVS carbon guard bed isolation valves

• ESFAS Carbon Guard Bed Isolation function

• PVVS delay bed filtration

• Supercell confinement boundary Damage to Equipment The occurrence of fire damages the affected carbon guard bed and eliminates its ability to function. No other damage to the PVVS system or its components occurs.

Transport of Radioactive Material The methods used to calculate radioactive material transport are described in Section 13a2.2.

The LPF model terms used in this accident are provided in Table 13b.2-1. For this accident, the release of material for the first 30 secondsguard bed inventory is assumed to be at an elevated rate due to continued PVVS flow. Following isolation damper actuation, theinstantly transport occurs at aed to the delay bed. The delay bed is credited to reduced rate the release of material by 99 percent with no credit taken for carbon guard bed isolation functions.

Radiation Source Terms The initial MAR for this scenario is a portion of the iodine gas inventory evolved from target solution during normal operations. Development of the accident source term for this scenario is discussed further in Section 13a2.2.

The iodine gas inventory is produced by fission and decay of fission products and continuously evolved from the target solution and through the TOGS during operations. Partitioning fractions for iodine gas are used to describe the quantities of iodine in solution that move to the RPF.

Removal of iodine by the TOGS zeolite beds are credited for all gases that are transported to the RPF. 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.

The 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.

Radiological Consequences The radioactive material is contained in the PVVS system and does not result in worker exposure.

The radiological consequences of this accident scenario are determined as described in Section 13a2.2. The results of the determination are provided in Table 13b.2-2.

SHINE Medical Technologies 13b.2-13 Rev. 1

Chapter 13 - Accident Analysis Analyses of Accidents with Radiological Consequences Table 13b.2 Radioisotope Production Facility Accident Dose Consequences Public Dose Worker Dose TEDE TEDE Accident Scenario (mrem) (mrem)

Maximum Hypothetical Accident 4023 No consequences Spill of Target Solution in the Supercell 117 64317 Spill of Eluate Solution in the Supercell 7814 47501 Spill of Target Solution in the RPF Pipe Trench 4 170 Spill of Target Solution from a Tank 5 398 Spill of Waste Solution in RLWI 103 657861 PVVS Carbon Delay Bed Fire 39 No consequences PVVS Carbon Guard Bed Fire 81 No consequences SHINE Medical Technologies 13b.2-15 Rev. 1