ML15147A210

Shine Medical Technologies V. 0 - Chapter 06 - Engineered Safety Features

ML15147A210
Person / Time
Site:	SHINE Medical Technologies
Issue date:	05/14/2015
From:	Bynum R V Shine Medical Technologies
To:	Office of Nuclear Reactor Regulation
Lynch S
References
SHINE, SHINE.SUBMISSION.16, SHN.PSAR.P, SHN.PSAR.P.7
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Chapter 6 - Engineered Safety FeaturesTable of ContentsSHINE Medical Technologies6-iRev. 1CHAPTER 6 ENGINEERED SAFETY FEATURESTable of Contents SectionTitlePage6a1HETEROGENEOUS REACTOR ENGINEERED SAFETY FEATURES............6a1-16a2IRRADIATION FACILITY ENGINEERED SAFETY FEATURES.......................6a2-1 6a2.1SUMMARY DESCRIPTION...............................................................................6a2-26a2.2IRRADIATION FACILITY ENGINEERED SAFETY FEATURES DETAILEDDESCRIPTION...................................................................................................6a2-46a2.2.1CONFINEMENT.................................................................................................6a2-56a2.2.2CONTAINMENT.................................................................................................6a2-86a2.3EMERGENCY COOLING SYSTEM...................................................................6a2-106a2.4IRRADIATION FACILITY ENGINEERED SAFETY FEATURESTECHNICAL SPECIFICATIONS........................................................................6a2-116a

2.5REFERENCES

...................................................................................................6a2-126bRADIOISOTOPE PRODUCTION FACILITY ENGINEERED SAFETYFEATURES........................................................................................................6b-16b.1SUMMARY DESCRIPTION ENGINEERED SAFETY FEATURES...................6b-16b.2RADIOISOTOPE PRODUCTION FACILITY ENGINEEREDSAFETYFEATURES........................................................................................................6b-36b.2.1CONFINEMENT.................................................................................................6b-46b.2.2CONTAINMENT.................................................................................................6b-76b.2.3EMERGENCY COOLING SYSTEM...................................................................6b-86b.2.4RADIOISOTOPE PRODUCTION FACILITY ENGINEERED SAFETYFEATURES TECHNICAL SPECIFICATIONS....................................................6b-96b.3NUCLEAR CRITICALITY CONTROL................................................................6b-11 6b.3.1CRITICALITY-SAFETY CONTROLS.................................................................6b-166b.3.2TECHNICAL SPECIFICATIONS........................................................................6b-206b.4REFERENCES...................................................................................................6b-23 Chapter 6 - Engineered Safety FeaturesList of TablesSHINE Medical Technologies6-iiRev. 0List of Tables NumberTitle6a2.1-1Summary of IF Design Basis Events and ESF Provided for Mitigation6a2.2-1Irradiation Facility Confinement Safety Functions 6b.1-1Summary of RPF Design Basis Eve nts and ESF Provided for Mitigation6b.2-1Radioisotope Production Facility Confinement Safety Functions6b.3-1Tanks Subject to Criticality-Safety Controls6b.3-2SHINE Nuclear Criticality Safety Program Elements Chapter 6 - Engineered Safety FeaturesList of FiguresSHINE Medical Technologies6-iiiRev. 0List of Figures NumberTitleNone Chapter 6 - Engineered Safety FeaturesAcronyms and AbbreviationsSHINE Medical Technologies6-ivRev. 1Acronyms and Abbreviations Acronym/AbbreviationDefinitionCi/m3microcuries per cubic meterACGIHAmerican Conference on Governmental Industrial HygienistsAGSAmerican Glovebox Society AHRIAir-Conditioning Heating and Refrigeration InstituteAIHAAmerican Industrial Hygiene Association ALARAas low as reasonably achievableAMCAAir Movement and Control AssociationANSAmerican Nuclear Society ANSIAmerican National Standards Institute ASHRAEAmerican Society of Heating, Refrigerating and Air-Conditioning Engineers, Inc.ASMEAmerican Society of Mechanical EngineersBSBachelor of ScienceBtubritish thermal unit Btu/hrbritish thermal units per hour CCelsius CAAScriticality accident and alarm systemCAMcontinuous air monitorCAMScontinuous air monitoring system CFRCode of Federal Regulations DACderived air concentrations DBAdesign basis accidentDBEdesign basis eventESFengineered safety feature Chapter 6 - Engineered Safety FeaturesAcronyms and AbbreviationsSHINE Medical Technologies6-vRev. 1 Acronyms and Abbreviations (cont'd)

Acronym/AbbreviationDefinitionESFAengineered safety feature actuation system FFahrenheit FDWSfacility demineralized water system FHAfire hazard analysis FICSfacility integrated control systemFPCfacility process conditionsHCFPShot cell fire detection and suppression system HEPAhigh efficiency particulate air HS&Ehealth safety & environment HVACheating, ventilation, and air conditioningI-125iodine-125I-131iodine-131 IBCInternational Building Code ICBSirradiation cell biological shieldingICCInternational Code CouncilIFirradiation facility IMCInternational Mechanical Code ISGinterim staff guidance IUirradiation unit keffeffective neutron multiplication factorkWhkilowatt-hourLCOlimiting conditions for operations LFLlower flammability limitLOOPloss of off-site powerLSCLife Safety Code Chapter 6 - Engineered Safety FeaturesAcronyms and AbbreviationsSHINE Medical Technologies6-viRev. 0 Acronyms and Abbreviations (cont'd)

Acronym/AbbreviationDefinitionLWPSlight water pool systemLWRlight water reactorMCNPMonte Carlo N-Particle Mo-99molybdenum-99 N-16nitrogen-16 NCSnuclear criticality-safetyNCSEnuclear criticality-safety evaluationNCSPNuclear Criticality Safety Program NEMANational Electrical Manufacturers Association NFDSneutron flux detection systemNFPANational Fire Protection AssociationNGRSnoble gas removal system Np-239neptunium-239 NRCU.S. Nuclear Regulatory Commission PCLSprimary closed loop cooling systemPFBSproduction facility biological shieldPSARPreliminary Safety Analysis Report PSBprimary system boundary Pu-239plutonium-239 PVVSprocess vessel vent systemRAMradiation area monitorRAMSradiation area monitoring systemRCAradiologically controlled area RDSradioactive drain system remroentgen equivalent manRICSradiological integrated control system Chapter 6 - Engineered Safety FeaturesAcronyms and AbbreviationsSHINE Medical Technologies6-viiRev. 0 Acronyms and Abbreviations (cont'd)

Acronym/AbbreviationDefinitionRPFradioisotope production facility RVradiologically controlled area ventilation system RVSAradiologically controlled area supply air RVZ1radiologically controlled area ventilation zone 1 RVZ2radiologically controlled area ventilation zone 2SMACNASheet Metal and Air Conditioning Contractors National AssociationSDDsystem design description SHINESHINE Medical Technologies, Inc.

SNMspecial nuclear material SRMstack release monitoringSSCstructure, system, or componentTOGStarget solution vessel off gas system TPCStarget solution vessel process control system TPStritium purification system TRPStarget solution vessel reactivity protection systemTSVtarget solution vesselU-238uranium-238 UFLupper flammability limit UNuranyl nitrate UNCSuranyl nitrate conversion systemUPSuninterruptible power supplyUPSSuninterruptible electrical power supply system UREXuranium extraction vol%volume percent Xe-133xenon-133 Heterogeneous ReactorChapter 6 - Engineered Safety FeaturesEngineered Safety FeaturesSHINE Medical Technologies6a1-1Rev. 0CHAPTER 6ENGINEERED SAFETY FEATURES6a1HETEROGENEOUS REACTOR ENGINEERED SAFETY FEATURESThe SHINE Medical Technologies, Inc. (SHINE) facility is not a heterogeneous reactor.

Therefore, this section does not apply to the SHINE facility.

Irradiation FacilityChapter 6 - Engineered Safety FeaturesEngineered Safety FeaturesSHINE Medical Technologies6a2-1Rev. 06a2IRRADIATION FACILITY ENGINEERED SAFETY FEATURESThe engineered safety features (ESF) of this portion of the facility are those systems designed to provide safety and defense in depth during start-up, normal operations, and shutdown as well as mitigate the consequences of postulated accidents, in spite of the fact that these postulated accidents are unlikely. These features can be divided into active and passive categories and include the following functions:

ConfinementThese systems provide active and passive protection against the potential release of radioactive material to the environment during normal conditions of operations and following a DBA. The confinement systems provide for active isolation of piping and heating, ventilation, and air conditioning (HVAC) systems penetrating confinement boundaries in certain post-accident conditions.

Chapter 6 - Engineered Safety FeaturesSummary DescriptionSHINE Medical Technologies6a2-2Rev. 06a2.1SUMMARY DESCRIPTIONThe intent of this section is to identify ESFs and provide a cross reference to postulated DBAs and events which the ESF acts to mitigate. The subsequent sections provide a detailed review of the ESF initiation and response to the event.Section 6a2.2 and its related subsections describe the ESFs, their modes of initiation and operation, and describe the structures, systems, and components (SSCs) which provide the features in detail.Table6a2.1-1 provides a reference summary of this information.

Chapter 6 - Engineered Safety FeaturesSummary DescriptionSHINE Medical Technologies6a2-3Rev. 0Table 6a2.1-1 Summary of IF Design Basis Accidents and ESF Provided for MitigationEngineered Safety Feature(ESF)Irradiation Facility Design Basis AccidentsMitigated by ESFSSCs which provide ESF Detailed Description Section or SubsectionConfinement *Mishandling or malfunction of target solution *Mishandling or malfunction of equipment affecting the PSB*Tritium Purification System Design Basis Accident*Irradiation unit cells including penetration seals*RCA ventilation system Zone 1 (RVZ1) ductwork upto bubble-tight isolation dampers*Bubble-tight isolation dampers*Isolation valves on piping systems penetrating the irradiationunit cells*TOGS shielded cells including penetration seals*Engineered safety features actuation system (ESFAS) *Double-walled pipe used for TPS*TPS gloveboxes*TPS confinement system 6a2.2.1 Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Detailed DescriptionSHINE Medical Technologies6a2-4Rev. 06a2.2IRRADIATION FACILITY ENGINEERED SAFETY FEATURES DETAILED DESCRIPTION The ESFs are passive or active features designed to mitigate the consequences of accidents and to keep the radiological and chemical exposures to the public, the facility staff, and the environment within acceptable values. This section provides the details of design, initiation, and operation of ESFs that are provided to mitigate the design basis accidents (DBAs) tabulated in

Section 6a2.1.According to the Final Interim Staff Guidance (ISG) Augmenting NUREG-1537, accident-initiating events (IEs) and scenarios, the following design basis accidents are to be addressed for the irradiation facility (IF):a.Insertion of excess reactivity/inadvertent criticalityb.Reduction in coolingc.Mishandling or malfunction of target solutiond.Loss of off-site powere.External eventsf.Mishandling or malfunction of equipment affecting the PSBg.Large un-damped power oscillations (fuel temperature/void-reactivity feedback)h.Detonation and deflagration in PSB i.Unintended exothermic chemical reactions other than detonationj.Primary system boundary system interaction eventsk.Facility-specific events*Inadvertent exposure to neutrons from neutron driver*Irradiation facility fires*Tritium Purification System Design Basis AccidentThe following design basis accidents (DBAs) do not have radiological consequences that require mitigation by ESFs: a.Insertion of excess reacti vity/inadvertent criticalityb.Reduction in coolingc.Loss of off-site power d.External eventse.Large un-damped power oscillations (fuel temperature/void-reactivity feedback)f.Detonation or deflagration in PSBg.Unintended exothermic chemical reactions other than detonationh.Primary system boundary system interaction eventsi.Facility-specific events*Inadvertent exposure to neutrons from neutron driver*Irradiation facility firesThe three DBAs requiring ESFs to mitigate the consequences are identified in Table 6a2.1-1.

Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Detailed DescriptionSHINE Medical Technologies6a2-5Rev. 06a2.2.1CONFINEMENT6a2.2.1.1Introduction Confinement is a term used to describe the low-leakage boundary that surrounds radioactive materials released during an accident and the associated RCA ventilation system (RV) components. Confinement systems are designed to localize release of radioactive material to controlled areas in normal operational states and mitigate the consequences of DBAs. Radiation protection control features such as adequate shielding and RV minimize hazards normally associated with radioactive materials. The principal design and safety objective of the confinement systems is to protect on-site personnel, the public, and the environment. The second design objective is to minimize the reliance on administrative or complex active engineering controls to provide a confinement system that is as simple and fail-safe as reasonably possible. The target solution vessel (TSV), TSV dump tank, TSV off-gas system (TOGS), and associated components act as the primary pressure boundary and are safety-related SSCs. Together they act as the primary fission product boundary. The confinement boundary of the IU cell and TOGS shielded cell encloses the PSB.The tritium purification system (TPS) double-walled piping and TPS confinement system provides confinement of the tr itium supplied to the IU. The confinement boundary of the TPS gloveboxes encloses the TPS.Confinement of the IU cells is achieved through the RV, ESFAS, and the biological shielding provided by the steel and concrete structures compri sing the walls, roofs, penetrations of the IU cell and TOGS shielded cell. Shielding of the IU cells and TOGS shielded cells is discussed in

detail in Section 4a2.5.

Confinement of the TPS gloveboxes is provided by the RV, ESFAS, and biological shielding comprising the walls, roofs, and penetration seals of the TPS gloveboxes. The biological shielding is described in Section 4a2.5.6a2.2.1.2Confinement System and ComponentsThe ventilation system serving the IU cells, TOGS shielded cells, and TPS gloveboxes includes components whose functions are designated as nonsafety-related and safety-related. The ductwork, the isolation dampers, and the filter trains of RVZ1 are designated as safety-related components. Refer to Table 6a2.2-1 for a description of the system and component safety functions. Active confinement isolation component s are required to operate as described below.The IU cells, TOGS shielded cell, and TPS gloveboxes employ a combination passive-active confinement methodology. During normal operation, passive confinement is achieved through the contiguous boundary between the hazardous materials and the surrounding environment and is credited with confining the hazards generated as a result of DBAs. This boundary includes the biological shield (created by the physical construction of the cell itself) and the extension of that boundary through the RCA Zone1 (RVZ1) ventilation system.

The extent of this passive confinement boundary extends from the upstream side of the intake Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Detailed DescriptionSHINE Medical Technologies6a2-6Rev. 0high efficiency particulate air (HEPA) filter to the final downstream HEPA filter prior to exiting the building. For the TPS the confinement also includes the double-walled piping. In the event of a DBA that results in a release in the IU cell, TOGS shielded cell, or TPS glovebox, radioactive material would be confined by the biological shield and physical walls of the cell itself. Each line that connects directly to the IU cell, TOGS shielded cell, or TPS glovebox atmosphere and penetrates the IU cell, TOGS shielded cell, or TPS glovebox is provided with redundant isolation valves to prevent releases of gaseous or other airborne radioactive material.

Confinement isolation valves on piping penetrating the IU cell, TOGS shielded cell, or TPS glovebox are located as close as practical to the confinement boundary and active isolation valves are designed to take the position that provides greater safety upon loss of actuating power. To mitigate the consequences of an uncont rolled release occurring within an IU cell, TOGS shielded cell, or TPS glovebox, as well as the off

-site consequences of releasing noble fission products through the ventilation system prior to sufficient decay, the confinement barrier utilizes

an active component in the form of bubble-tight isolation dampers (safety-related) on the inlet and outlet ventilation ports of each cell. This ESF effectively reduces the amount of ductwork in the confinement volume that needs to remain intact to achieve IU cell, TOGS shielded cell, or TPS glovebox confinement. These dampers close automatically (fail-closed) upon loss of power or receipt of a confinement isolation signal generated by the ESFAS. Following an initiating event, the ESFAS provides the confinement isolation signal that isolates the IU cells, TOGS shielded cell, or TPS glovebox. Refer to Subsection 7a2.5 for a description of the ESFAS.A failure of the TPS outside the glovebox is mitigated by the TPS confinement system. The TPS confinement system uses isolation valves to stop a tritium leak outside the glovebox when a leak is detected.

Overall performance assurance of the active confinement components is achieved through factory and in-place testing. Duct and housing leak tests are performed in accordance with ASME N511 (ASME, 2008), with minimum acceptance criteria as specified in ASME AG-1 (ASME, 2009). Specific owner's requirements with respect to acceptable leak rates are based on the safety analyses.6a2.2.1.3Functional Requirements Active confinement components are designed to fail into a safe state if conditions such as loss of signal, loss of power, or adverse environments are experienced.

Mechanical, instrumentation, and electrical systems and components required to perform their intended safety function in the event of a single failure are designed to include sufficient redundancy and independence such that a single failure of any active component does not result in a loss of the capability of the system to perform its safety functions.

Mechanical, instrumentation, and electrical systems and components are designed to ensure that a single failure, in conjunction with an initiating event, does not result in the loss of the system's ability to perform its intended safety function. The single failure considered is a random failure and any consequential failures in addition to the initiating event for which the system is required and any failures that are a direct or consequential result of the initiating event.

Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Detailed DescriptionSHINE Medical Technologies6a2-7Rev. 0The design of safety-related systems (including protection systems) is consistent with IEEE Standard 379-2000 (IEEE, 2001)and Regulatory Guide 1.53 in the application of the single-failure criterion.6a2.2.1.4Confinement ComponentsThe following components are associated with the secondary confinement barrier of the IU cells, TOGS shielded cell, or TPS glovebox, as prev iously described. Their specific materials, construction, and installation and operating requirements are evaluated based on the safety analysis.Bubble-tight isolation dampers, designed, cons tructed and tested in accordance with ASME AG-1, Section DA "Dampers and Louvers" (ASME, 2009):*Maintain their functional integrity.*Maintain their rated leak-tightness following a seismic event.

• Maintain their structural integrity under fan shut-off pressure.*Provide bubble-tight isolation upon receipt of a control signal or, in the event of loss of actuator power, by closure of actuator.*Provide bubble-tight isolation when using manual actuator or when locked closed with power actuator removed. *Relay damper full-open and full-closed position for control and indication by the use of limit switches.Dampers are butterfly type, blade and frame fabricated of heavy-gage stainless steel. Total leakage is based on bubble solution test as outlined in ASME AG-1 2009, Section DA-5141

(ASME, 2009).

Ventilation ductwork and ductwork support materials meet the requirements of ASME AG-1, Article SA-3000 "Materials". Supports are designed and fabricated in accordance with the requirements of ASME AG-1 2009, Section SA "Ductwork" (ASME, 2009). Details of the TPS confinement system will be developed in final design and provided in the FSAR.Low leakage seals are provided on each penetration through the IU cell, TOGS shielded cell, and TPS glovebox. For systems open to the IU cell, TOGS shielded cell atmosphere, or TPS glovebox, redundant isolation valves are provided.6a2.2.1.5Engineered Safety Feature Test RequirementsEngineered safety features are periodically tested to ensure that ESF components maintain operability and can provide adequate confidence that the system performs satisfactorily in service during postulated events. To the extent possible, the ESFAS and the confinement ESF whose operation it initiates are designed to permit testing during plant operation. Testing actuation devices and actuated equipment may be done individually or in groups to avoid negative impact to plant operations.

Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Detailed DescriptionSHINE Medical Technologies6a2-8Rev. 06a2.2.1.6Design BasesFor general discussion of the codes and standards used for the SHINE facility, see Chapter 3. For the design basis of the RVZ1, see Subsection 9a2.1.1. For the design basis of the IU cell and TOGS cell, see Section 4a2.5. For the design basis of the ESFAS, see Section 7a2.5. Forthe design basis of the TPS, see Section 9a2.7.1.Potential variables, conditions, or other items that will be probable subjects of a technical specification associated with the IF confinement systems and components are provided in Chapter 14.6a2.2.2CONTAINMENTThe SHINE facility does not employ a containm ent feature. Due to the low temperature and power level of facility operations, the safety analysis demonstrates that confinement features are adequate to mitigate potential accidents. Therefore, Subsection6b.2.1 is provided to describe the use of confinements as an ESF for the RPF.

Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Detailed DescriptionSHINE Medical Technologies6a2-9Rev. 0Table 6a2.2-1 Irradiation Facility Confinement Safety Functions System, Structure, ComponentDescriptionClassification RVZ1 IU cell isolation dampersProvide confinement at IU cell and TOGS shielded cell boundaries Safety-RelatedESFASProvides confinement isolation signalSafety-Related Isolation valves on piping systemsProvide confinement at IU cell and TOGS shielded cell boundaries, and TPS boundary Safety-RelatedIU cell and TOGS shielded cell including penetrationsProvide confinement Safety-Related Chapter 6 - Engineered Safety FeaturesEmergency Cooling SystemSHINE Medical Technologies6a2-10Rev. 06a2.3EMERGENCY COOLING SYSTEM Decay heat removal during accident scenarios is provided by the safety-related light water pool. No emergency core cooling system is required for the SHINE facility to mitigate the consequences of an accident. See Chapter 13.

Irradiation Facility Engineered SafetyChapter 6 - Engineered Safety FeaturesFeatures Technical SpecificationsSHINE Medical Technologies6a2-11Rev. 06a2.4IRRADIATION FACILITY ENGINEERED SAFETY FEATURES TECHNICAL SPECIFICATIONSPotential variables, conditions, or other items that are probable subjects of a technical specification associated with the IF ESFs are provided in Chapter 14.

Chapter 6 - Engineered Safety FeaturesReferencesSHINE Medical Technologies6a2-12Rev. 06a

2.5REFERENCES

ASME, 2008.In-ServiceTestingofNuclearAirTreatment, Heating, Ventilating, and Air-Conditioning Systems, ASMEN511-2007, Ameri can Society of Mechanical Engineers, February 2008.ASME, 2009.CodeonNuclearAirandGasTreatment, ASMEAG-1-2009, American Society of Mechanical Engineers, September 2009.

IEEE, 2001.IEEE StandardApplication of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, IEEEStandard379-2000, Institute of Electrical and Electronics Engineers, Septem ber 2000, Reaffirmed March 2008.

Summary DescriptionChapter 6 - Engineered Safety FeaturesEngineered Safety FeaturesSHINE Medical Technologies6b-1Rev. 16bRADIOISOTOPE PRODUCTION FACILITY ENGINEERED SAFETY FEATURES6b.1SUMMARY DESCRIPTION ENGINEERED SAFETY FEATURESESFs of this portion of the facility are those systems designed to mitigate the consequences of postulated accidents.

ConfinementThese systems provide active and passive protection against the potential release of radioactive materials or chemicals to the environment following a design basis accident. The confinement systems provide for active isolation of pi ping and HVAC systems penetrating confinement boundaries in certain post-accident conditions.The intent of this section is to identify ESFs and provide a cross reference to postulated DBAs and events which the ESF acts to mitigate. The subsequent sections provide a detailed review of the ESF initiation and response to the event.

Section 6b.2 and its subsections describe the ESFs, their modes of initiation and operation, and describe the structures, systems, and components (SSCs) which provide the features in detail.Table6b.1-1 provides a reference summary of this information.

Chapter 6 - Engineered Safety FeaturesSummary Description Engineered Safety FeaturesSHINE Medical Technologies6b-2Rev. 1Table 6b.1-1 Summary of RPF Design Basis Events and ESF Provided for MitigationEngineered Safety Feature(ESF)Radioisotope Production Facility Design Basis EventMitigated by ESFSSCs which provide ESF DetailedDescription Section or SubsectionConfinement*Critical equipment malfunction*Accidents with hazardous chemicals*Hot cells including penetration seals*RCA ventilation system Zone 1 (including ductwork up to filters and

filters) and Zone 2 *Bubble-tight isolation dampers

• Tank vaults

• Radiological integrated controls system(RICS)*Isolation valves on piping systems penetrating hot cells 6b.2.1 Radioisotope Production FacilityChapter 6 - Engineered Safety FeaturesEngineered Safety FeaturesSHINE Medical Technologies6b-3Rev. 06b.2RADIOISOTOPE PRODUCTION FACILITY ENGINEERED SAFETY FEATURESThe ESFs are passive or active features designed to mitigate the consequences of accidents and to keep the radiological and chemical exposures to the public, the facility staff, and the

environment within acceptable values. This section provides the details of design, initiation, and operation of ESFs that are provided to mitigate the DBAs discussed in Section6b.1. This includes chemical storage areas outside the RCA.According to Chapter 13 of the Final ISG Augmenting NUREG-1537, the following DBAs are to be addressed for the RPF:a.Critical equipment malfunction.b.Inadvertent nuclear criticality in the RPF.

c.RPF fire.d.Accidents with hazardous chemicals. e.External events.These DBAs encompass LOOP and operator errors (See Section 13b).The following DBAs do not have consequences that require mitigation by ESFs:a.Inadvertent nuclear criticality in the RPF.b.RPF fire.

c.External events.The two DBAs requiring ESFs to mitigate consequences are identified in Table 6b.1-1.

Chapter 6 - Engineered Safety FeaturesConfinementSHINE Medical Technologies6b-4Rev. 26b.2.1CONFINEMENT6b.2.1.1IntroductionConfinement describes the low-leakage boundary surrounding radioactive or hazardous chemical materials released during an accide nt and parts of RVZ1 and RVZ2. Confinement systems localize releases of radioactive or hazardous materials to controlled areas and mitigate the consequences of DBAs. Personnel protection control features such as adequate shielding and RV minimize hazards normally associated with radioactive or chemical materials. The principal design and safety objective of the confinement system is to protect the on-site personnel, the public, and the environ ment. The second design objective is to minimize the reliance on administrative or complex active engineering controls and provide a confinement system that is as simple and fail-safe as reasonably possible.

This subsection describes the confinement systems for the RPF. The RPF confinement areas include hot cell enclosures for process operations and trench and vault enclosures for process tanks and piping.Confinement is achieved through RV, RICS, and biological shielding provided by the steel and concrete structures comprising the walls, roofs, and penetrations of the hot cells. Shielding of the hot cells is discussed in detail in Subsection 4b.2. 6b.2.1.2Confinement System and ComponentsThe RV serving the RCA, outside of the IF, includes components whose functions are designated as nonsafety-related and safety-related. The ductwork, the isolation dampers, and the filter trains of RVZ1 are designated as safety-related. Refer to Table 6b.2-1 for a description of the system and component safety functions. Active confinement isolation components are required to operate as described below.

The hot cells employ a combination passive-act ive confinement methodology. During normal operation, passive confinement is achieved through the contiguous boundary between the hazardous materials and the surrounding environment and is credited with confining the hazards generated as a result of DBAs.This boundary includes the biological shield (created by the physical construction of the cell itself) and the extension of that boundary through the RVZ1. The intent of the passive boundary is to confine hazardous materials while also pr eventing the introduction of external energy sources that could disturb the hazardous materials from their steady-state condition. The extent of this passive confinement boundary extends from the upstream side of the intake HEPA filter to the final downstream HEPA filter prior to exiting the building.In the event of a DBA that results in a r elease in the hot cells, radioactive material would be confined by the biological shield and physical walls of the cell itself. Each line that connects directly to the hot cell atmosphere and penetrates the hot cell is provided with redundant isolation valves to prevent releases of gaseous or other airborne radioactive material. Confinement isolation valves on piping penetrating the hot cell are located as close as practical to the confinement boundary and active isolation valves are designed to take the position that provides greater safety upon loss of actuating power.

Chapter 6 - Engineered Safety FeaturesConfinementSHINE Medical Technologies6b-5Rev. 2To mitigate the consequences of an uncontrolled release occurring within a hot cell, as well as the off-site consequences of releasing fission pr oducts through the ventilation system, the confinement barrier utilizes an active compone nt in the form of bubble-tight isolation dampers (safety-related) on the inlet and outlet ventilation ports of each hot cell. This ESF effectively reduces the amount of ductwork in the confinement volume that needs to remain intact to achieve hot cell confinement. These dampers close automatically (fail-closed) upon loss of power or receipt of a confinement isolation signal generated by the RICS. Following an initiating event, the RICS isolates the hot cells. Refer to Section 7b for a description of the RICS.

Overall performance assurance of the active confinement components is achieved through factory testing and in-place testing. Duct and housing leak tests are performed in accordance with ASME N511, with minimum acceptance criteria as specified in ASME AG-1 (ASME, 2009). Specific owner's requirements with respect to acceptable leak rates are based on the safety

analyses.6b.2.1.3Functional RequirementsActive confinement components are designed to fail into a safe state if conditions such as loss of signal, loss of power, or adverse environments are experienced.

Mechanical, instrumentation, and electrical systems and components required to perform their intended safety function in the event of a si ngle failure are designed to include sufficient redundancy and independence such that a single failure of any active component does not result in a loss of the capability of the system to perform its safety functions.

Mechanical, instrumentation, and electrical systems and components are designed to ensure that a single failure, in conjunction with an initiating event, does not result in the loss of the system's ability to perform its intended safety function. The single failure considered is a random failure and any consequential failures in addition to the initiating event for which the system is required and any failures that are a direct or consequential result of the initiating event.The design of safety-related systems (including protection systems) is consistent with IEEE Standard 379-2000 and Regulatory Guide 1.53 in the application of the single-failure criterion.6b.2.1.4Confinement ComponentsThe following components are associated with the confinement barriers of the hot cells, tank vaults and pipe trenches, as previously described.

Their specific materials, construction, and installation and operating requirements are evaluated based on the safety analysis.Bubble-tight isolation dampers, designed, c onstructed and tested in accordance with ASME AG-1, Section DA "Dampers and Louvers" (ASME, 2009):*Maintain their functional integrity.*Maintain their rated leak-tightness following a seismic event.*Maintain their structural integrity under fan shut-off pressure.*Provide bubble-tight isolation upon receipt of a control signal or, in the event of loss of actuatorpower, by closure of actuator.*Provide bubble-tight isolation when using manual actuator or when locked closed with power actuator removed.

Chapter 6 - Engineered Safety FeaturesConfinementSHINE Medical Technologies6b-6Rev. 0*Relay damper full-open and full-closed position for control and indication by the use of limit switches.Dampers are butterfly type, blade and frame f abricated of heavy-gage stainless steel. Total leakage based on bubble solution test as outlined in ASME AG-1 2009, Section DA-5141 (ASME, 2009).

Ventilation ductwork and ductwork support materials meet the requirements of ASME AG-1, Article SA-3000 "Materials". Supports are designed and fabricated in accordance with the requirements of ASME AG-1, Section SA "Ductwork" (ASME, 2009).Low leakage seals are provided on each penetration through the hot cells. For systems open to the hot cell atmosphere, redundant isolation valves are provided.6b.2.1.5Engineered Safety Feature Test Requirements Engineered safety features are tested to ensur e that ESF components maintain operability and can provide adequate confidence that the system performs satisfactorily in service during postulated events. To the extent possible, the RICS and the confinement ESF whose operation it initiates are designed to permit testing during plant operation. Testing actuation devices and actuated equipment may be done individually or in groups to avoid negative impact to plant operations.6b.2.1.6Design BasesFor general discussion of the codes and standards used for the SHINE facility, see Chapter 3. For the design basis of the RVZ1 and RVZ2, see Subsection 9a2.1.1. For the design basis of the

hot cells, see Section 4b.2. For the ESF-related design basis of the RICS, see Section 7b.4.1. Potential variables, conditions, or other items that will be probable subjects of a technical specification associated with the RPF confinement systems and components are provided in Chapter 14.

Chapter 6 - Engineered Safety FeaturesContainmentSHINE Medical Technologies6b-7Rev. 06b.2.2CONTAINMENTThe SHINE RPF does not employ a containment feature. Due to the low temperature and power level of facility operations, the safety analysis demonstrates that confinement features are adequate to mitigate potential accidents. Subsection6b.2.1 is provided to describe the use of confinements as an ESF for the RPF.

Chapter 6 - Engineered Safety FeaturesEmergency Cooling SystemSHINE Medical Technologies6b-8Rev. 06b.2.3EMERGENCY COOLING SYSTEM There is no emergency cooling system associat ed with the RPF side of the SHINE facility; therefore, this subsection is not applicable.

Radioisotope Production Facility EngineeredChapter 6 - Engineered Safety FeaturesSafe ty Features Technical SpecificationsSHINE Medical Technologies6b-9Rev. 06b.2.4RADIOISOTOPE PRODUCTION FACILITY ENGINEERED SAFETY FEATURES TECHNICAL SPECIFICATIONS Potential variables, conditions, or other terms that are probable subjects of a technical specification associated with the RPF ESFs are provided in Chapter 14.

Radioisotope Production Facility EngineeredChapter 6 - Engineered Safety FeaturesSafe ty Features Technical SpecificationsSHINE Medical Technologies6b-10Rev. 2Table 6b.2-1 Radioisotope Production Facility Confinement Safety FunctionsSystem, Structure, ComponentDescriptionClassification RVZ1 hot cell isolation dampers, ductwork up to filters and filtersProvide confinement isolation at hot cell boundaries SRRVZ2 isolation dampers, ductwork up to filters and filtersProvide confinement isolation at RCA boundarySRRICS Provides confinement isolation signalSR Isolation valves on piping systemsProvide confinement at hot cell boundaries SRHot cells, tank vaults, and pipe trenchesProvides confinement SR Chapter 6 - Engineered Safety FeaturesNuclear Criticality ControlSHINE Medical Technologies6b-11Rev. 06b.3NUCLEAR CRITICALITY CONTROL The nuclear criticality safety program (NCSP) for t he radioisotope production facility (RPF) is in accordance with applicable America n National Standards Institut e/American Nuclear Society (ANSI/ANS) standards, as endorsed by Regulatory Guide 3.71. Except for irradiation operations in the TSV, the facility is designed to meet the requirements of the following standards and NRC Regulatory Guide:*ANSI/ANS-8.1-1998 (R2007), "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors" (ANSI/ANS

, 2007a) and ANSI/ANS-8.19-2005, "Administrative Practices for Nuclear Criticality Safety" (ANSI/ANS, 2005b) (see Regulatory Guide 3.71), for organization and administration.*ANSI/ANS-8.1-1998 (R2007), "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors" (ANSI/ANS, 2007a) Appendix A will be used to identify nuclear criticality safety (NCS).*ANSI/ANS-8.3-1997 (R2012), "Criticality Accident Alarm System" (ANSI/ANS, 2012a), as modified by Regulatory Guide 3.71, "Nuclear Criticality Safety Standards for Fuels and Materials Facilities."*ANSI/ANS-8.7-1998 (R2007), "Guide for Nuclear Criticality Safety in the Storage of Fissile Materials" (ANSI/ANS, 2007b).*ANSI/ANS-8.10-1983 (R2005), "Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement" (ANSI/ANS, 2005a).*ANSI/ANS-8.20-1991 (R2005), "Nuclear Criticality Safety Training" (ANSI/ANS, 2005c).

• ANSI/ANS-8.21-1995 (R2011), "Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors" (ANSI/ANS, 2011).*ANSI/ANS-8.23-2007 (R2012), "Nuclear Criticality Accident Emergency Planning and Response" (ANSI/ANS, 2012b).*ANSI/ANS-8.24-2007 (R2012), "Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations" (ANSI/ANS, 2012c).*ANSI/ANS-8.26-2007 (R2012), "Criticality Safety Engineer Training and Qualification Program" (ANSI/ANS, 2012d).There is no deviation from any standards or requirements that would require the development of equivalent requirements or standards in the medical isotope production area. These criteria do not apply to the operation of the target solution vessels.Elements comprising the NCSP are presented in Table 6b.3-2. Listed with each program element are the corresponding ANSI/ANS standards and regulations that will serve as the program basis.The RPF will be designed, constructed, and operated such that a nuclear criticality event is prevented. The NCSP will implement the guidance provided in ANSI/ANS-8.19-2005 (ANSI/ANS, 2005b). Nuclear criticality safety at the facility will be ensured by designing the facility and SSCs with safety margins such that safe condit ions will be maintained u nder normal and abnormal operations and design basis accidents. The facility is designed with a preference for NCS controls in the following order:a.Passive engineered.b.Active engineered.

c.Enhanced administrative.

Chapter 6 - Engineered Safety FeaturesNuclear Criticality ControlSHINE Medical Technologies6b-12Rev. 1d.Simple administrative.An accidental criticality is highly unlikely because the SHINE facility is designed with passive engineered design features, including the use of neutron absorbers. The analysis, consequences, and safety controls of such an event have been described in Subsection 13b.2.5. Administrative controls ensure the reliability and availability of the safety controls are adequate to maintain subcriticality.Heterogeneous effects are not considered applicab le because the uranium enrichment is less than20 percent.Management of the Nuclear Criticality Safety ProgramThe NCS criteria are used for managing criticality safety and include adherence to the double contingency principle, as stated in ANSI/ANS-8.1-1998 (R2007) (ANSI/ANS, 2007a). The double contingency principle states "process design should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible." Each process that has accident scenarios that could result in an inadvertent nuclear criticality at the SHINE fa cility meets the double contingency principle. Process design incorporates sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.The NCS program establishes criteria for the administration of those operations in which there exists a potential for nuclear criticality accidents. Responsibilities of management, supervision, and the nuclear criticality safety staff are d efined below. Objectives and characteristics of operating and emergency procedures are also defined. The emergency procedures will include reporting criteria and report content requirements. Reports will be issued based on whether the criticality controls credited were lost (i.e., they were unreliable or unavailable to perform their intended safety functions), irrespective of whether the safety limits of the associated parameters were actually exceeded. Training in criticality safety will be provided to individuals who handle nuclear material at the facility. The training is based upon the training program described in

ANSI/ANS-8.20-1991 (R2005) (ANSI/ANS, 2005c

) and ANSI/ANS-8.26 (R2012) (ANSI/ANS, 2012d).Other aspects of the NCS program include:*Providing distinctive NCS postings in areas, operations, work stations, and storage locations relying on administrative controls for NCS.*Requiring personnel to perform activities in accordance with written, approved procedures when the activity may affect NCS. Unless a specific procedure deals with the situation, personnel shall take no action until the NCS staff has evaluated the situation and provided recovery procedures.*Requiring personnel to report defective NCS conditions to the NCS program management.

Chapter 6 - Engineered Safety FeaturesNuclear Criticality ControlSHINE Medical Technologies6b-13Rev. 0*Describing organizational positions, experience of personnel, qualifications of personnel, and functional responsibilities.*Designating an NCS program director who wi ll be responsible for implementing the NCS program.*Promptly detecting any NCS deficiencies by means of operational inspections, audits, or investigations and reporting those deficiencies. Records of deficiencies will be retained and corrective actions will be documented.*A program for evaluating the criticality significance of NCS events and an apparatus in place for making the required notification to the NRC Operations Center. Qualified individuals make the determination of significance for NCS events. The determination of loss or degradation of double-contingency protection is made against the license.The training program is developed and implemented with input from the criticality safety staff, training staff, and management. The training focuses on the following:*Knowledge of the physics of NCS. *Analysis of jobs and tasks to determine the knowledge a worker must have to perform tasks efficiently.*Design and development of learning objectives based upon the analysis of jobs and tasks that reflect the knowledge, skills, and abilities needed by the worker.*Implementation of revised or temporary operating procedures.*Training all personnel to recognize the CAAS sign al and to evacuate promptly to a safe area.*Providing instruction and training regarding t he policy for NCS organization procedures.Management responsibilities include:*Overall responsibility for safety of operations.*Demonstration of continuing interest in NCS.

• Formulation of NCS policy and communication of the policy to all employees.*Involvement in operations with fissile material.*Assignment of responsibility and delegation of authority to implement established policy.*Responsibility for criticality safety that is assigned in a manner corresponding to that for other safety disciplines.*Provision of a criticality safety staff appropriate for the scope of operations and experienced in the interpretation of criticality data.*Provision of administrative oversight, such as an NCS organization or a Criticality Control Committee.*Provision of administrative and organizational assurance that the SHINE NCS organization is independent of operations to an extent practical.*Establishment of a review system to monitor the NCSP.

• Periodic participation in assessments or audits of the overall effectiveness of the NCSP.*Provision of a committee that will assist m anagement in achieving the objectives of the NCSP.

Chapter 6 - Engineered Safety FeaturesNuclear Criticality ControlSHINE Medical Technologies6b-14Rev. 0 Supervisory responsibilities include:*Safety of operations under their control.*Demonstration of knowledge of aspects of NCS related to operations under their control.*Assurance that personnel under their supervision have an understanding of procedures and safety considerations such that personnel may be expected to perform their functions without undue risk.*Development or participation in the development and maintenance of written procedures that have an impact on NCS.*Verification of compliance with NCS specificat ions prior to the use of new or modified equipment. Verification may make use of inspection reports or other features of the quality control system.*Maintenance of conformance with good safety practices, including unambiguous identification of fissile materials and transient combustibles.

Nuclear criticality safety staff responsibilities include:*Provision of technical guidance for the design of equipment and processes, and for the development of operating procedures.*Maintenance of familiarity with current developments in NCS standards, guides and codes.*Assistance to supervision, on request, for personnel training.*Conduct audits of NCS practices and compliance with procedures, as directed.

• Examination of procedural violations, deficiencies, and mishaps for possible improvement of safety practices and reporting findings to management.*Response to potential NCS non-complia nce and defective NCS conditions.*Reporting of potential NCS non-complianc e and defective NCS conditions to management.The NCS staff member must understand and have experience in the application of NCS programs. The NCS engineer is responsible for implementation of the NCSP. The education requirements of NCS staff members will be addressed in the FSAR.

Process Descriptions Process descriptions of criticality safety for no rmal operations will be developed for the FSAR.

Operating ProceduresThe purpose of operating procedures is to facilitate the efficient and safe conduct of operations. Procedures should be organized and presented for convenient use by operators. The operating

procedures will include:*Controls and limits significant to the NCS of the operation.*Periodic revisions, as necessary.*Procedures will be supplemented by NCS limit s posted on equipment or incorporated in operating check lists or flow sheets.*Periodic review of active procedures by supervision.

Chapter 6 - Engineered Safety FeaturesNuclear Criticality ControlSHINE Medical Technologies6b-15Rev. 1The operating procedures will meet the intent of ANSI/ANS-8.19-1996.Planned Response to Criticality Accidents The criticality accident and alarm system (CAAS) will be used as described in Section 7b.6. The CAAS provides for detection and annunciation of criticality accidents. Emergency procedures shall be prepared and approved by management. On-site and off-site organizations that are expected to respond to emergencies shall be inform ed of conditions that might be encountered. These procedures shall clearly designate evacuation routes. Evacuation should follow the quickest and most direct routes practical. Thes e routes shall be clearly identified. Procedures shall include assessment of exposure to individuals. In addition, personnel assembly stations, outside the areas to be evacuated, shall be designated. Means to account for personnel shall be established. Arrangements shall be made in advance for the care and treatment of injured and exposed persons. The possibility of personnel contamination by radioactive materials shall be considered.Personnel shall be trained in evacuation methods and informed of routes and assembly stations. Drills shall be performed at least annually. Instrumentation and procedures are provided for determining the radiation in the evacuated area following a criticality accident. Information is collected in a central location.Emergency procedures for each area in which special nuclear material is handled, used, or stored will be maintained to ensure that all personnel withdraw to an area of safety upon the sounding of the alarm. The procedures include conducting drills to familiarize personnel with the evacuation plan, designation of responsible individuals for determining the cause of the alarm, and placement of radiation survey instruments in accessible locations for use in such an emergency. The current procedures for each area will be retained as a record for as long as licensed special nuclear material is handled, used, or stored in the area. The superseded portion of the procedures will be retained for three years after the portion is superseded.Fixed and personnel accident dosimeters are provided in areas that require a CAAS. These dosimeters are readily available to personnel resp onding to an emergency, and there is a method for prompt on-site dosimeter readouts.NCS Change ManagementChanges that involve or could affect SNM will be evaluated under 10 CFR 50.59. Such changes include: new design, operation, or modification to existing SSCs; computer programs; processes; operating procedures; or administrative controls.

Prior to implementing the change, it shall be determined that the entire process will be subcritical (with approved margin for safety) under both normal and credible accident scenarios. The above process for NCS change management is in accordance with ANSI/ANS-8.19-1996 (R2005) (ANSI/ANS, 2005b).

Chapter 6 - Engineered Safety FeaturesCriticality-Safety ControlsSHINE Medical Technologies6b-16Rev. 16b.3.1CRITICALITY-SAFETY CONTROLSThe scope of this section is NCS controls for the RPF and it is applicable to SSCs of the facility in which uranium is present in sufficient mass for a nuclear criticality accident to be credible (in certain conditions). Criticality safety controls are listed in Table 13b.2.5-1. The uranium contained within the TSV and TSV dump tank is outside the scope of this section as the TSV and TSV dump tank are within the IF. The CAAS is not a control from perspective of criticality control; however, the CAAS is considered safety-related.The major controlling parameters used in the facility are geometry control and/or limitations on the mass. Nuclear criticality safety evaluations (NCSEs) and analyses are used to identify the significant parameters affected within a particular system. Assumptions relating to process, equipment, material function, and operation, including credible accident scenarios, are justified, documented, and independently reviewed. Where possible, passive engineered controls are used to ensure NCS. The use of geometry as a controlled parameter is acceptable because all dimensions and nuclear properties that use geometry control are verified.

Interaction and neutron absorption may also be us ed as controlling parameters. Interaction as a controlling parameter is acceptable if engineered controls are used to maintain a minimum separation between units. Enhanced administrative controls are used where engineered controls are not feasible.

The use of neutron absorption as a controlled parameter is acceptable by following ANSI/ANS-8.21-1995 (R2011) (ANSI/ANS, 2011),

"Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors." Neutron spectra are considered in the evaluation of absorber effectiveness.

Nuclear Criticality Safety Evaluations Nuclear criticality safety evaluations will be performed to ensure that nuclear processes will remain subcritical during both normal and a ccident scenarios. An approved margin of subcriticality will be included in these evaluations.

Nuclear criticality safety is analyzed for the design features of the plant system or component and for the operating practices that relate to

maintaining NCS. The analysis of individual systems or components and their interaction with other systems or components containing enriched uranium is performed to ensure the NCS criteria are met.

Nuclear criticality safety analyses also meet the following:*NCS analyses are performed using acceptable methodologies.*Methods are validated and used only within demonstrated acceptable ranges.*The analyses adhere to the methodology standards in ANSI/ANS-8.1-1998 (R2007)

(ANSI/ANS, 2007a).*The reference manual and documented reviewed validation report will be kept at the facility.*The reference manual and validation report are incorporated into the Configuration Management Program.

Chapter 6 - Engineered Safety FeaturesCriticality-Safety ControlsSHINE Medical Technologies6b-17Rev. 2*The NCS analyses are performed in accordance with the methods specified and incorporated in the configuration management program.*Nuclear criticality safety controls and controlled parameters ensure that under normal and credible abnormal conditions, all nuclear processes are subcritical, including use of an approved margin of subcriticality for safety that is used.*Process specifications incorporate margins to protect against uncertainties in process variables and against a limit being accidentally exceeded (ANSI/ANS, 2007a).*Use of ANSI/ANS-8.7-1998 (R2007) (ANSI/ANS, 2007b), as it relates to the requirements for subcriticality of operations, the margin of subcriticality for safety, and the selection of controls.*ANSI/ANS-8.10-1983 (R2005) (ANSI/ANS, 2005a), as modified by Regulatory Guide 3.71, as it relates to the determination of consequences of NCS accident sequences, is used.*If administrative keff margins for normal and credible accident scenarios are used, NRC pre-approval of the administrative margins will be sought.*Subcritical limits for keff calculations such that the margin is large compared to the uncertainty in the calculated values, and includes adequate allowance for uncertainty in the methodology, data, and bias to ensure subcriticality are used.*Studies to correlate the change in a value of a controlled parameter and keff value are performed. The studies include changing the value of one controlled parameter and determining its effect on another controlled parameter and keff.*The calculation of keff is based on a set of variables within the method's validated area of applicability.*Trends in the bias support the extension of the methodology to areas outside the area or areas of applicability.The NSCE procedure addresses requirements for:*Normal case operating conditions.*Nuclear criticality hazard identification.*Hazard identification method.*Hazard identification results.*Nuclear criticality hazard evaluation.*Nuclear criticality parameter discussions.*Nuclear criticality safety controls (passive design features, active engineered features and administrative controls).*Nuclear criticality safety peer review requirements.Geometry TanksEach of the tanks within the scope of this section features criticality safety controls that meet the double-contingency principle, i.e., processes incorporate sufficient safety factors so that at least two unlikely, independent, and concurrent changes in process conditions are required before a nuclear criticality accident is possible. The first criticality safety control is that each tank, with the exception of the tanks associated with liquid waste processing, is criticality safe by geometry or by the combination of geometry and a layer of neutron absorbing material integral to the tank construction. The second, independent criticality-safety control is that the most reactive concentration of uranium in any tank results in k eff 0.95, based on MCNP analyses. MCNP is a validated code for calculating reactivity and dosimetry. Small amounts of fissile plutonium-239 (Pu-239), resulting from activation of uranium-238 (U-238) and subsequent decay of Chapter 6 - Engineered Safety FeaturesCriticality-Safety ControlsSHINE Medical Technologies6b-18Rev. 0neptunium-239 (Np-239) during fuel solution irradiation, are accounted for in the MCNP calculations, even if the measured concentration of plutonium is negligible.

PipingPipe runs are single-parameter criticality-safe by geometry, with pipe diameters limited based on MCNP calculations, and with consideration of material and spacing relative to other uranium, reflecting, and moderating materials.In the event of a leak in a double-walled pipe located in the pipe trenches, the target solution in the pipe trench would drain into a criticality-safe by geometry tank.

Tank VaultsConstraints on the facility footprint prevent limiting potential maximum spills in tank vaults to single-parameter criticality safe depths. Theref ore, each tank vault is connected, via a non-valved gravity drain into the criticality safe sump catch tank, which is criticality-safe by geometry.

LimitsSafe geometry limits for each of the following SSCs are established:*Uranium and uranium oxide receipt and storage: storage container's diameter and center-to-center spacing.*Uranium dissolution: process vessels and piping diameter and center-to-center spacing.

• Target Solution Hold Tank (1-TSPS-03T): configuration (e.g., pencil or donut) and dimensions.*TSV Dump Tank (1-SCAS-01T): configuration (e.g., pencil or donut) and dimensions.

• UREX (uranium extraction) recovery system: process vessels and piping diameter and center-to-center spacing.*Denitration system: process vessels and piping diameter and center-to-center spacing.*Criticality-safe sump catch tank (1-RDS-01T): configuration and dimensions.

Mass LimitsThe mass of uranium present in an SSC is determined by analytical measurements and process calculations. Conservative administrative limits for each operation are specified in the operating procedures. Whenever mass control is established for a container, records are maintained for mass transfers into and out of the container. Establishment of mass limits for a container involves consideration of potential moderation, reflection, geometry, spacing, and enrichment. The evaluation considers normal operations and credible abnormal conditions for determination of the operating mass limit for the container and for the definition of subsequent controls necessary to prevent reaching the safety limits.

Mass limits, expressed as total mass of uranium or as uranium concentration, for each of the following SSCs will be established:*Uranium and uranium oxide receipt and storage: total mass.

Chapter 6 - Engineered Safety FeaturesCriticality-Safety ControlsSHINE Medical Technologies6b-19Rev. 0*Uranium dissolution: uranium concentration.*TSV Feed Tank (1-TSPS-03T): uranium concentration.*TSV Dump Tank (1-SCAS-01T): uranium concentration.*Uranium extraction (UREX): uranium concentration.

• Thermal denitration system: uranium concentration.*Waste processing systems: uranium concentration and total mass.

Criticality Accident Alarm SystemThe SHINE CAAS will be designed to meet the following:*The facility CAAS is capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of 2 meters from the reacting material within one minute. Two detectors cover each area needing CAAS coverage.*The CAAS will use gamma and neutron sensitive radiation detectors that energize clearly audible alarm signals if accidental criticality occurs.*The system complies with ANSI/ANS-8.3-1997 (R2012) (ANSI/ANS, 2012a), as modified by Regulatory Guide 3.71.*The CAAS is appropriate for the facility for the type of radiation detected, the intervening shielding, and the magnitude of the mi nimum accident of concern.*The CAAS is designed to remain operational during design basis accidents.*The CAAS is clearly audible in areas that must be evacuated or there are alternative notification methods that are documented to be effective in notifying personnel that evacuation is necessary.*Operations are rendered safe, by shutdown and quarantine, if necessary, in any area where CAAS coverage has been lost and not restored within a specified number of hours. The number of hours will be determined on a process-by-process basis, because shutting down certain processes, even to make them safe, may carry a larger risk than being

without a CAAS for a short time. Compensatory measures (e.g., limiting access, halting SNM movement, or restoring CAAS coverage with an alternate instrument) when the CAAS system is not functional will be determined.*Emergency power will be provided to the CAAS via the UPSS.

Chapter 6 - Engineered Safety FeaturesTechnical SpecificationsSHINE Medical Technologies6b-20Rev. 06b.3.2TECHNICAL SPECIFICATIONSPotential variables, conditions, or other items that are probable subjects of specification associated with the NCSP are provided in Chapter 14.

Chapter 6 - Engineered Safety FeaturesTechnical SpecificationsSHINE Medical Technologies6b-21Rev. 0Table 6b.3-1 Tanks Subject to Criticality-Safety ControlsTag #Description

1. EachCriticality SafeType1-TSPS-01TUranyl Sulfate Preparation Tank2YAnnular1-TSPS-02TUranium Metal Dissolution Tank1YPencil1-TSPS-03TTarget Solution Hold Tank8YAnnular 1-SCAS-01TTSV Dump Tank8YRing 1-MEPS-01TMo Extraction Column3YPencil 1-UNCS-01T Uranyl Nitrate Conversion Tank2YAnnular 1-UNCS-02TUREX Feed Tank2YHarp 1-UNCS-03TTechnetium Removal Column1YCylinder 1-UNCS-04TSolvent Hold Tank1YAnnular 1-UNCS-05TRaffinate Hold Tank2YHarp 1-UNCS-06TRecycle UN (Uranyl Nitrate) Hold Tank1YHarp1-UNCS-07TUN Evaporator Vessel1YProcess1-UNCS-08AThermal Denitrator1Y Tank type will be provided in FSAR 1-UNCS-09T Recycle Target Solution Tank3YAnnular 1-RDS-01TCriticality Safe Sump Catch Tank1YHarp Chapter 6 - Engineered Safety FeaturesTechnical SpecificationsSHINE Medical Technologies6b-22Rev. 1Table 6b.3-2: Nuclear Criticality Safety Program Elements and Requirement Bases Program Element NameRequirement Basis for ElementCriticality Safety PolicyNot applicable Verification & Validation of NCS CodesANSI/ANS 8.1-1998 Nuclear Criticality Safety Evaluations ANSI/ANS 8.1-1998 ANSI/ANS 8.19-1996 Nuclear Criticality Training and Qualifications ANSI/ANS 8.20-1991 R1999 10 CFR 19.12Implementation of Criticality Safety Controls and LimitsANSI/ANS 8.19-1996 Configuration Control / Change Control 10 CFR 50.59Audits and InspectionsANSI/ANS 8.19-1996 Criticality Safety Non-Compliances: Investigating, Reporting, Tracking, and Trending ANSI/ANS 8.19-1996 10 CFR 19.12Criticality Safety Guidelines for Fire FightingNot applicableEmergency Preparedness Plan and Response Procedures Manual ANSI/ANS 8.23-1997 ANSI/ANS 8.19Criticality Detector and Alarms System ANSI/ANS 8.3-1997 10 CFR 19.12 10 CFR 70.24Testing and Calibration of Active-Engineered ControlsNot applicable Criticality Safety Controls ProgramNot applicable Chapter 6 - Engineered Safety FeaturesReferencesSHINE Medical Technologies6b-23Rev. 06b.4REFERENCES ANSI/ANS, 2005a.

Criteria for Nuclear Criticality-Safety Controls in Operations With Shielding and Confinement, ANSI/ANS-8.10-1983 (R2005), ANSI/ANS, 2005.

ANSI/ANS, 2005c. N uclear Criticality-Safety Training, ANSI/ANS-8.20-1991 (R2005),

ANSI/ANS, 2005.

ANSI/ANS, 2007a. N uclear Criticality-Safety in Operations with Fissionable Materials Outside Reactors, ANSI/ANS-8.1-1998 (R2007), ANSI/ANS, 2007.ANSI/ANS, 2007b. G uide for Nuclear Criticality-Safety in the Storage of Fissile Materials, ANSI/ANS-8.7-1998 (R2007), ANSI/ANS, 2007. ANSI/ANS, 2005b. Administrative Practices for Nuclear Criticality-Safety, ANSI/ANS-8.19-2005, ANSI/ANS, 2005.

ANSI/ANS, 2011.

Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors, ANSI/ANS-8.21-1995 (R2011), 2011.

ANSI/ANS, 2012a.

Criticality Accident Alarm System, ANSI/ANS-8.3-1997 (R2012), 2012.

ANSI/ANS, 2012b. Nuclear Criticality Accident Emergency Planning and Response, ANSI/ANS-8.23-2007 (R2012), 2012.

ANSI/ANS, 2012c. Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, ANSI/ANS-8.24-2007 (R2012), 2012.

ANSI/ANS, 2012d. Criticality Safety Engineer Training and Qualification Program, ANSI/ANS-8.26-2007 (R2012), 2012.ASME, 2007.

Testing of Nuclear Air Treatment Systems, ASME N510, American Society of Mechanical Engineers, 2007.ASME, 2009. Code on Nuclear Air and Gas Treatment, ASME AG-1, American Society of Mechanical Engineers, 2009.

IEEE, 2000. Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, IEEE 379, Institute of Electrical and Electronics Engineers, 2000.

IMC, 2012. International Mechanical Code, IMC, 2012.