SAR-420 Chapter 9

Form 412.09 (Rev. 10)

Idaho National Laboratory

CHAPTER 9 – BUILDINGS AND AUXILIARY SYSTEMS – TREAT

FACILITY FSAR

Identifier: Revision: Effective Date:

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CHAPTER 9

BUILDINGS AND AUXILIARY SYSTEMS

Further dissemination authorized to DOE and DOE contractors only; other requests shall be approved by the originating facility or higher

DOE programmatic authority.

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CONTENTS

ACRONYMS .................................................................................................................................. 9-4

9. BUILDINGS AND AUXILIARY SYSTEMS .......................................................................... 9-5

9.1 Reactor Building ........................................................................................................ 9-5

9.1.1 Design Bases ........................................................................................... 9-5 9.1.2 Description .............................................................................................. 9-7 9.1.3 Safety Evaluation ................................................................................... 9-11

9.2 Control Building ...................................................................................................... 9-11

9.2.1 Design Bases ......................................................................................... 9-11 9.2.2 Description ............................................................................................ 9-12 9.2.3 Safety Evaluation ................................................................................... 9-12

9.3 Reactor Fuel Assembly Storage and Handling ............................................................ 9-13

9.3.1 Reactor Fuel Assembly Storage Facilities ................................................ 9-14 9.3.2 Reactor Fuel Assembly Handling ............................................................ 9-17

9.4 Experiment Loop Storage and Handling .................................................................... 9-19

9.4.1 Experiment Loop Storage ....................................................................... 9-20 9.4.2 Experiment Loop Handling..................................................................... 9-23

9.5 Prevention of Inadvertent Criticality .......................................................................... 9-27

9.5.1 Requirements......................................................................................... 9-27 9.5.2 Criticality Concerns ............................................................................... 9-27 9.5.3 Criticality Controls................................................................................. 9-28 9.5.4 Application of Double Contingency Principle .......................................... 9-29 9.5.5 Criticality Instrumentation ...................................................................... 9-29

9.6 Fire Protection Systems ............................................................................................ 9-29

9.6.1 Reactor Building .................................................................................... 9-30 9.6.2 Control Building .................................................................................... 9-36

9.7 Water Systems ......................................................................................................... 9-37

9.7.1 Domestic Water System ......................................................................... 9-37 9.7.2 Sanitary Sewage System ......................................................................... 9-40 9.7.3 Industrial Waste-Water System ............................................................... 9-40

9.8 Process Gases .......................................................................................................... 9-41

9.8.1 Compressed Air Systems ........................................................................ 9-41 9.8.2 Inert Gas Supply System ........................................................................ 9-42

9.9 Heating, Ventilation, and Air Conditioning ................................................................ 9-43

9.9.1 Design Basis .......................................................................................... 9-43 9.9.2 Description ............................................................................................ 9-43 9.9.3 Safety Evaluation ................................................................................... 9-45 9.9.4 Inspection and Testing............................................................................ 9-46

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9.9.5 Limitations and Setpoints ....................................................................... 9-46

9.10 Communication Systems .......................................................................................... 9-46

9.10.1 Design Basis .......................................................................................... 9-46 9.10.2 Description ............................................................................................ 9-46 9.10.3 Safety Evaluation ................................................................................... 9-46 9.10.4 Inspection and Testing............................................................................ 9-46 9.10.5 Limitations and Setpoints ....................................................................... 9-47

9.11 Lighting Systems ..................................................................................................... 9-47

9.11.1 Design Basis .......................................................................................... 9-47 9.11.2 Description ............................................................................................ 9-47 9.11.3 Safety Evaluation ................................................................................... 9-47 9.11.4 Inspection and Testing............................................................................ 9-47 9.11.5 Limitations and Setpoints ....................................................................... 9-47

9.12 Casks ...................................................................................................................... 9-48

9.12.1 Fuel-Handling Cask ............................................................................... 9-48 9.12.2 TREAT Loop-Handling Cask ................................................................. 9-52

9.13 References ............................................................................................................... 9-54

FIGURES Figure 9-1. Cutaway view of the Reactor Building. ............................................................................9-8

Figure 9-2. Reactor Building floor plan. ............................................................................................9-9

Figure 9-3. Reactor Building basement layout.................................................................................. 9-10

Figure 9-4. Reactor Building mezzanine layout. ............................................................................... 9-10

Figure 9-5. Control Building and adjacent Office Building. .............................................................. 9-12

Figure 9-6. Floor plan for Control Building. .................................................................................... 9-13

Figure 9-7. Cross-sectional view through a TREAT fuel storage pit. ................................................. 9-16

Figure 9-8. Manual fuel-handling tool (fuel gripper) and thermocouple plug tool. .............................. 9-18

Figure 9-9. Fuel-handling cask design. ............................................................................................ 9-49

Figure 9-10. Fuel transfer arrangement. ........................................................................................... 9-50

Figure 9-11. Sequence of fuel assembly removal.............................................................................. 9-51

Figure 9-12. TREAT loop-handling cask. ........................................................................................ 9-53

TABLES Table 9-1. Criticality safety scope for TREAT activities and operations. ........................................... 9-28

Table 9-2. Reactor Building fire hazards and fire protection equipment. ............................................ 9-31

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ACRONYMS

ANL Argonne National Laboratory AR nonsafety-related with augmented requirements ASHRAE American Society of Heating, Refrigeration, and Air Conditioning Engineers ASTM American Society of Testing and Materials AWS American Welding Society

BAR basement auxiliary room BNL Brookhaven National Laboratory

CMAA Crane Manufacturers Association of America CSE Criticality Safety Evaluation Report CSI Criticality Safety Index

DOE U.S. Department of Energy

F/CS filtration/cooling system FHC fuel-handling cask FSAR Final Safety Analysis Report

GDC General Design Criteria

HEPA high-efficiency particulate air HID high-intensity discharge HVAC heating, ventilation, and air conditioning

I&C instrumentation and control INL Idaho National Laboratory LHM loop-handling machine

LRD Laboratory Requirements Document

MFC Materials and Fuels Complex

NFPA National Fire Protection Association NRC Nuclear Regulatory Commission NSR nonsafety-related

TLHC TREAT loop-handling cask TREAT Transient Reactor Test (TREAT) facility

UBC Uniform Building Code

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9. BUILDINGS AND AUXILIARY SYSTEMS

As discussed in Chapter 1, Introduction and General Description of Facility, the light-water reactor (LWR) edition of Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.70 (NRC 1978) was used as the guide to format and content for this chapter. RG 1.70 is designated in 10 CFR 830 (2001) as an acceptable format and content guide for U.S. Department of Energy (DOE) reactor safety analysis reports. Building and auxiliary system design characteristics required by RG 1.70 that directly support the design or accident analyses of the Transient Reactor Test (TREAT) facility are discussed in this chapter.

It should be noted that this section references SAR-400 as the source of specific detailed information, but does not invoke the requirements of SAR-400.

The primary buildings making up the TREAT facility include:

• Reactor Building (MFC-720)

• Control Building (MFC-724)

The auxiliary systems include:

• Reactor fuel assembly storage and handling systems

• Experiment loop storage and handling systems

• Fire protection systems

• Water systems

• Heating, ventilation, and air conditioning (HVAC) system

• Communications

• Lighting systems.

9.1 Reactor Building

9.1.1 Design Bases

The Reactor Building provides for:

• Receiving, handling, storing, and loading/unloading of experiment loops, reactor fuels, calibration assemblies, and handling equipment

• Assembly, preparation, checkout, and partial disassembly of experiment loops

• Storage of loop components and experiment support equipment

• Reactor and experiment operation and maintenance

• Instrumentation and control systems

• Redundant and standby electrical power generation

• Health physics operations and equipment

• Radioactive waste management systems

• Building equipment and building services (ANL 1987).

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According to Chapter 15, Accident Analyses, the Reactor Building is classified as a nonsafety-related (NSR) with augmented requirements (NSR-AR) structure, system, and component to support the cranes during fuel handling. To fulfill these functions for the larger TREAT test loops, transporters, and handling equipment, the original Reactor Building was modified to include an expanded high-bay region and two low-bay additions (see Section 9.1.2).

The buildings and auxiliary systems were evaluated for natural events such as wind, earthquake, snow, flooding, and lightning. Because of the low frequency of a tornado strike (SAR-400, Section 1.5.2), the building design does not provide tornado resistance.

The Reactor Building was designed to support the imposed dead and live loads. The dead load includes the weight of framing, roof, walls, floors, and fixed service equipment. The live load includes the intended use and occupancy loads, and loads imposed by such natural phenomena as wind, snow, earthquakes, and flooding. The principal loads used in the design of the facility are provided in System Design Description for Site and Reactor Building (ANL 1987):

Wind: From 30 to 40 lb/ft2 increasing with height above ground level

Earthquake: Uniform Building Code requirements for Zone 3, importance factor of 1.5

Snow: 30 lb/ft2

Flooding: 100-year rain (2.25 in. of water in 24 hours)

Floor: Ranges from 100 psf to 2,500 psf.

The Reactor Building was designed and constructed in accordance with national codes and standards current at the time of construction and modification. The Code of Record is therefore preserved from the original design/build unless revised through facility modifications and is memorialized in (ANL 1987), including:

• Uniform Building Code (UBC), International Conference of Building Officials, including supplements

• American Institute of Steel Construction: Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, and Manual of Steel Construction

• American Iron and Steel Institute: Specification for the Design of Cold-Formed Steel Structural Members

• American Welding Society: Standards Code for Arc and Gas Welding in Building Construction (AWS D1.1)

• American Concrete Institute: Building Code Requirements for Reinforced Concrete (ACI 318)

• American Society for Testing and Materials specifications as appropriate

• American National Standards Institute standards as appropriate

• National Fire Protection Association International: National Fire Codes

• 29 CFR 1910, “Occupational Safety and Health Standards”

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• U.S. Department of Energy (DOE) Manual General Design Criteria, Appendix 6301

• American Society of Heating, Refrigeration, and Air Conditioning Engineers (ASHRAE) Standard 90-75, “Energy Conservation in New Building Design.”

9.1.2 Description

The Reactor Building is a metal-sided, steel-framed industrial building that features two high-bay sections, two low-bay sections, two belowgrade rooms (the subpile room and basement auxiliary room [BAR]), mezzanines, and an exhaust stack.

The principal feature of the Reactor Building is the south high-bay, which is 74 ft wide × 116 ft long × 80 ft high (see Figures 9-1 and 9-2). The south high-bay houses the reactor, experiment storage holes, experiment assembly tower, a storage stand for the loop-handling machine (LHM), the LHM itself, and a motorized roll-up door that provides access on the west side of the south high-bay. Included in the south high-bay are two mezzanine levels, 54 × 29 ft. The south high-bay is equipped with a 60-ton bridge crane (downrated to 20 tons as discussed in Section 9.4.2.1.1) for handling larger experiments and associated equipment. In the south high-bay, access is provided for experiment transporters and other vehicular traffic by a drive-through passageway under the larger crane; this permits receiving and handling of experiments in heavily shielded containers.

The north high-bay is 38 ft wide × 76 ft long × 35 ft high. The north high-bay contains belowgrade reactor fuel and experiment storage holes, an equipment storage pit, a loop transfer pit, storage areas for the fuel-handling cask (FHC) and TREAT loop-handling cask (TLHC), and a motorized roll-up door at the north end. The north high-bay area also provides space for the assembly, storage, loading, checkout, and disassembly of fuel assemblies and experiment loops, as well as for experiments and reactor operations. A 15-ton crane serves the north high-bay area as well as the reactor in the south high-bay.

The subpile room is a 16 ft wide × 16 ft long × 13.5 ft high belowgrade room, located directly beneath the reactor. The subpile room houses reactor-related equipment (control rod drive mechanisms), an oxygen sensor that monitors the oxygen level in the subpile room, and two 1,500-lb chain pulls/hoists. The BAR is a belowgrade room 18 ft wide × 20 ft long × 18 ft high immediately adjacent to the subpile room (see Figure 9-3). It houses the hydraulic supply systems for the reactor transient rod drive and compensation rod latch systems, a 4-ton overhead crane with a 1-ton hoist, an oil mist detection system, and an oxygen sensor that monitors the oxygen level in the BAR.

Two mezzanine floors are 54 × 29 ft (see Figure 9-4) and provide for equipment and operational storage space. The lower mezzanine is located adjacent to the top of the reactor biological shield; the upper mezzanine is located 15 ft above the lower mezzanine. The roof system consists of structural steel trusses that frame the high-bay columns. Purlins span the trusses and support the metal roof decking. The roof proper is composed of metal deck, two layers of 2-1/2-in. polyisocyanurate (adhered), and 60-mil RubberGard ethylene-propylene-diene-monomer (EPDM) membrane. Flashing, gutters, downspouts, etc., are of steel construction and match the finish of the siding.

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Figure 9-1. Cutaway view of the Reactor Building.

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Figure 9-2. Reactor Building floor plan.

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Figure 9-3. Reactor Building basement layout.

Figure 9-4. Reactor Building mezzanine layout.

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Two low-bay areas, the east and west, comprise the remainder of the Reactor Building (refer back to Figures 9-1 and 9-2). Both low-bay areas have steel-frame, beam-column construction for the walls and roof, with horizontal and vertical bracing systems; both areas have insulated metal siding and roofing. Floors are reinforced concrete slabs on rock or compacted fill. The low-bays provide housing for the mechanical/electrical equipment, experiment equipment, instrument room, health physics, personnel restrooms, and equipment storage.

The east bay includes the diesel generator room, electrical and mechanical equipment rooms, workshops, and filtration/cooling system (F/CS) room. The diesel generator room houses a redundant power diesel generator, a standby power diesel generator, day-use diesel fuel tanks, nickel-cadmium battery banks, and associated instrumentation and controls (see Chapter 7, Instrumentation and Controls). Diesel fuel for the redundant and standby generators is stored in an aboveground storage tank located outside the diesel generator room north of the east low-bay. A motorized roll-up door provides access through the east side of the south high-bay.

The west low-bay includes a shift supervisor/health physics room, an instrumentation and control (I&C) room, an electronics shop, electricians office, the suspect water waste tank room, and restrooms (refer back to Figure 9-2). The health physics room provides space for radiological personnel to support TREAT operations and activities. The room includes a safe for radioactive sources and a decontamination sink that drains to a collection container.

Interior walls are constructed of reinforced concrete block or gypsum board on metal studs. Personnel doors and interior walls are of fire-rated construction where required for personnel and facility safety (see Section 9.6). Exterior walls are constructed of ribbed steel siding, supported by steel girts connected to the exterior columns.

9.1.3 Safety Evaluation

The adequacy of the Reactor Building when subjected to natural phenomena events (i.e., wind, earthquake, snow, and flood) and floor loads is described at Chapter 3, Design of Structures, Components, Equipment, and Systems. Other aspects of building safety, such as fire protection, experiment handling, and fuel storage, are discussed in subsequent sections of this chapter.

9.2 Control Building

9.2.1 Design Bases

The function of the Control Building is to house the control console and cabinets associated with reactor operations, including readouts for reactor parameters, radiation monitoring and the F/CS. All manually controlled operations (e.g., reactor startup, steady-state, and transient initiation) are performed by a reactor operator who is stationed at the control console in the Control Building. Human engineering aspects were addressed in the design and placement of the control panel in the Control Building Reactor Control Room.

The Control Building was designed to be occupied by personnel. It contains an HVAC system, sanitary facilities, and fire-suppression systems. A snow load of 30 lb/ft2, a wind load of 25 lb/ft2, and a floor load of 100 lb/ft2 were used in the design of the building. No seismic criteria were used in designing the building. The reactor seismic trip system will automatically shut the reactor down; therefore, seismic damage to the Control Building is not a factor affecting reactor safe shutdown.

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9.2.2 Description

The Control Building is a one-story structure, 60 × 28 ft, with a floor area of 1,680 ft2 (gross). It is located adjacent to the Office Building (see Figure 9-5). The roof and sides of the building are constructed of enamel-coated prefabricated steel panels and are insulated. Interior partitions are constructed of fire-resistant gypsum board on steel studs, and the ceiling is of fire-resistant acoustical tile. The entire building is supported on reinforced concrete piers extending 5 ft below grade (see Figure 9-6).

The Control Building has an auxiliary evacuation alarm. The radiation monitoring system includes an area monitor for detection of alpha, beta, and gamma radiation with the values displayed in the control room.


There are no safety-related SSCs in the Control Building. The Reactor Control Room is designated NSR-AR. The Control Building is not relied upon to prevent radiological releases that could result in doses in excess of the accident consequence guidelines. The adequacy of the Control Building under the effects of natural events and floor loads is described in Chapter 3, Section 3.3.8.

Figure 9-5. Control Building and adjacent Office Building.

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Figure 9-6. Floor plan for Control Building.

9.3 Reactor Fuel Assembly Storage and Handling

This section discusses the facilities for storing and handling reactor fuel at the TREAT facility. Section 9.4 discusses those for test fuel. Additional information on the specific limitations and procedures for storing and handling fissionable material at the TREAT facility is provided in LST-387. The reactor fuel assembly storage and handling facilities were designed and constructed to meet the requirements of General Design Criteria (GDC) 61, 62, and 63 (see Chapter 3).

The requirement of GDC-61, “Fuel Storage and Handling and Radioactivity Control,” Part 1, “designed with capability to permit appropriate periodic inspection and testing of components important to safety,” as it applies to reactor fuel assembly storage and handling, is met due to the following design features:

Inspection of a belowgrade reactor fuel storage hole (see Section 9.4.1.2) can be achieved by removal of a fuel assembly (see Section 9.4.2) from a storage hole, followed by visual and/or light-assisted inspection. Testing of a reactor fuel storage hole is not required since the holes are passive structures.

• Inspection and maintenance of the FHC (see Section 9.4.2.2) is controlled through the use of the Maintenance Management Program used at the TREAT facility.

• Inspection and testing of the 15-ton crane (see Section 9.4.2.2) complies with the requirements of the INL Safety and Health Manuals (INL Laboratory-wide Manual 14 A, B, C) and the TREAT Maintenance Management Program.

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• Inspection and maintenance of the fuel gripper tool (see Section 9.4.2.2) is controlled through the use of the Maintenance Management Program used at the TREAT facility.

The requirement of GDC-61, Part 2, “designed with suitable shielding for radiation protection,” as it applies to reactor fuel assembly storage and handling, is met due to the following design features:

• Chapter 12, Radiation Protection, demonstrates that the reactor fuel storage holes will provide suitable shielding for the Zircaloy-clad TREAT fuel assemblies

• The FHC provides shielding for the TREAT facility (see Section 9.4.2.1). This will provide more than adequate shielding in the handling of the Zircaloy-clad TREAT fuel assemblies.

The requirement of GDC-61, Part 3, “designed with appropriate containment, confinement, and filtering systems,” as it applies to reactor fuel assembly storage and handling, is met due to the following design features:

• The rigid concrete and steel pipe-lined storage holes provide appropriate containment and confinement

• The impact analysis presented in Chapter 3 describes that the storage holes provide adequate containment and confinement, even under highly improbable impact loadings

• The accident analysis presented in Chapter 15 describes that the consequences of dropping the FHC are well within the accident frequency/consequence guidelines.

The fuel handling and storage systems do not require filtering systems.

The requirement of GDC-62, “Prevention of Criticality in Fuel Storage and Handling,” as it applies to reactor fuel storage and handling, is shown to be met (see Section 9.3.1.3). The results of calculations indicate that the fuel assembly storage areas are safely subcritical under over-batching configurations, as well as single batching with flooding.

The requirement of GDC-63, “Monitoring Fuel and Waste Storage,” Part 1, “to detect excessive radiation levels,” as it applies to reactor fuel storage and handling, is met by the radiation monitoring instrumentation described in Chapter 12. The requirement of GDC-63, Part 2, “to initiate appropriate safety actions,” is met by the audio and visual alarms associated with the area radiation monitors are also described in Chapter 12.

9.3.1 Reactor Fuel Assembly Storage Facilities

9.3.1.1 Design Bases. The reactor fuel storage facilities are capable of storing fueled and unfueled assemblies (see Figure 9-2). The belowgrade storage holes maintain a subcritical condition for the most reactive array (ECAR-1610). Flooding of the fuel assembly storage holes was considered in the criticality analysis.

9.3.1.2 Description. All of the facilities available in the TREAT facility for storing reactor fuel assemblies are located in the north high-bay of the Reactor Building and are serviced by the 15-ton building crane. None of the fuel storage areas contains decay heat removal features beyond the natural convection of air in the holes, and the conduction of heat through the concrete around the holes. The burnup level of TREAT fuel assemblies in the past, as well as that projected for the Zircaloy-clad TREAT fuel assemblies in the future, is insignificant. Therefore, no additional heat removal capability will be required.

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Two areas northwest of the reactor provide below-grade storage of Zircaloy-clad TREAT fuel assemblies. Each storage area consists of 6-in.-OD steel tubes arranged on 10-in. centers (see Figure 9-7). One of the two storage areas (Zone 1) is rectangular and provides space for 120 fuel assemblies for locations of zones). The other storage area (Zones 2 and 3) is trapezoidal and provides space for 147 fuel assemblies. Zone 2 and 3 differs from Zone 1 in that high-density (magnetite-loaded) concrete is used between the tubes. Both storage areas are secured by interlocking, hinged-aluminum doors (1/4-in. thick) and padlocks.

Zone 2 houses 126 holes usually filled with fuel assemblies. Zone 3 houses the remaining 21 holes in two rows along the southeastern diagonal. Zone 3 is reserved for belowgrade storage of capsule experiments and miscellaneous fissionable material. Zone 3 storage holes will be available, however, whenever the reactor core is unloaded.

An additional reactor fuel-storage area (Zone 9) provides for the belowgrade storage of fuel assemblies. This area, 12 × 18 ft, provides space for 198 fuel assemblies in an 11 × 18 array. Steel-lined storage tubes, 10-1/2 ft deep × 6 in. diameter, house the stored fuel assemblies. A removable concrete shield plug, 12 in. thick, can be provided for each storage tube. The storage tubes are arranged on 10-in. centers and are cast in concrete. The concrete mixture contains 145 lb of boron frit per cubic yard of concrete. This storage area provides criticality-safe storage of Zircaloy-clad fuel assemblies or Inconel-clad assemblies as described in ECAR-1610.

9.3.1.3 Safety Evaluation. Two safety concerns for reactor fuel storage are criticality safety and radiation safety. Criticality safety involves the storage of Zircaloy-clad TREAT fuel assemblies. Radiation safety involves shielded storage of the fuel assemblies under normal conditions and during postulated accident conditions.

The Zircaloy-clad TREAT fuel assemblies are stored in the fuel-storage areas. Each assembly contains 38 g of 235U as highly enriched UO2 and is diluted with carbon. The carbon-uranium atom ratio is about 10,000:1. Criticality analyses (ECAR-1610) demonstrate that the Zircaloy-clad fuel assembly or Inconel-clad assembly storage arrays are subcritical under normal and credible abnormal events. Chapter 3, Section 3.3.6, provides additional bases for the safe storage of reactor fuel.

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Figure 9-7. Cross-sectional view through a TREAT fuel storage pit.

Because Zircaloy-clad fuel assemblies experience low burnup, the fuel-storage areas require only minimal radiation shielding. Some shielding is provided by the upper graphite reflector section on each fuel assembly (McVean 1979; Yang and Bhattacharyya 1982; Yang, June 1982). If necessary, the 12-in.-thick concrete closure plugs can provide additional shielding. Access by personnel to the floor area above the fuel-storage areas is unrestricted.

The fuel assembly storage areas, which are below grade and store each assembly in a Schedule 20 steel pipe embedded in concrete, provide considerable protection for the stored fuel assemblies. In the event of an earthquake, the seismic forces might cause some spalling of the concrete and some fracturing at the floor-pit joint. However, the steel pipe would protect the stored fuel assembly from either concrete spallation or fracturing.

9.3.1.4 Testing and Inspections. To meet GDC-61 requirements, inspection of a belowgrade storage holes is performed as described in Section 9.4 to demonstrate the storage hole integrity.

9.3.1.5 Limitations and Setpoints. The reactor fuel assembly storage arrays are limited by size or administrative controls as to their maximum inventory.

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9.3.2 Reactor Fuel Assembly Handling

The reactor fuel-handling equipment consists of a FHC, cranes, and associated handling tools.

9.3.2.1 Design Bases. The reactor fuel-handling equipment at TREAT includes an FHC, a 15-ton crane, a 60/10-ton crane (downrated to 20 tons as discussed in Section 9.4.2.1.1), and manual fuel-handling tools.

9.3.2.1.1 FHC–See Section 9.12.1 for discussion of the FHC Design Bases.

9.3.2.1.2 15-Ton Crane—The 15-ton crane is used to handle the FHC and other equipment in the TREAT facility. It services the area from the reactor to the northernmost fuel storage area. The 15-ton crane was designed in accordance with the following codes and standards:

• U.S. Atomic Energy Commission, Reactor Development and Technology, F8-6T

• Occupational Safety and Health Standards (OSHA), 29 CFR, Chapter XVII, Parts 1900-1999

• Safety Code for Overhead and Gantry Cranes, etc., ASME B30.2

• Crane Manufacturers Association of America: CMAA-70

• American Welding Society: AWS D2.0

• American Institute of Steel Construction Manual

• National Electric Code.

9.3.2.1.3 Manual Fuel-Handling Tools—Manual fuel-handling tools are suitable for Zircaloy-clad fuel assemblies. The tools are generally used for small adjustments and other “fine tuning” fuel-handling operations. Their use is governed by the radiation field above the core and the exposure time, and is monitored by radiological control personnel. When radiation fields are low, the manual fuel-handling tool may be the normal method to move fuel elements. The tool has a positive locking device to minimize the potential for a fuel element drop. Tool storage is provided on top of the reactor shielding.

9.3.2.2 Description. See Section 9.12.1 for description of the FHC.

9.3.2.2.1 15-Ton Crane—The bridge crane located at the TREAT north high-bay has a capacity of 15 tons. The building foundations, columns, and runway beams have been analyzed for the design crane loads. The runway beams are reinforced.

The 15-ton bridge crane spans 33-1/3 ft between crane rails on double girders and serves the north high-bay (reactor area). Its coverage overlaps that of the 60-ton crane by approximately 14 ft. The crane consists of a bridge, trolley, and hoist.

The 15-ton crane is pendant controlled, and the various components have multiple operating speeds. The hook has a total lift of 47 ft, of which more than 26 ft lift is above the main floor. Interference between the two overhead cranes in the common operating area over the reactor is prevented by a system of two-position, positive-action limit switches. Bypass switches for these limits are controlled by key operation.

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9.3.2.2.2 Manual Fuel-Handling Tool—The manual fuel-handling tool (fuel gripper), shown in Figure 9-8, is available for use with fuel assemblies that are fresh or only slightly irradiated. Delatching is effected by a rod attached to the handle. The manual fuel-handling tool requires the 15-ton overhead crane for load lifting and transfer.

Figure 9-8. Manual fuel-handling tool (fuel gripper) and thermocouple plug tool.

The thermocouple plug tool is an aluminum rod (5 ft long) that is rotated within an aluminum tube to engage or disengage an end fitting on the rod with the thermocouple lead wire that plugs into the top of thermocouple fuel assemblies.

9.3.2.3 Safety Evaluation. Radiation safety and materials-handling safety are two important safety concerns for the TREAT reactor fuel-handling equipment, due to the limited capacity, criticality safety is not a concern for the FHC. Radiation safety involves the radiation shielding provided by the FHC. Materials-handling safety involves the design of the crane and in-cask hoisting equipment.

9.3.2.3.1 Fuel-Handling Cask—See Section 9.12.1 for discussion of FHC Safety Evaluation.

9.3.2.3.2 15-Ton Crane—The heaviest load that the 15-ton crane will normally be required to handle in the TREAT facility is the 11-1/2 ton TLHC (see Section 9.4.2.2). Hence, the lifted loads are well within the rated capacity of the crane.

As described in Section 9.3.2.2, interference between the two overhead cranes in the common operating area over the reactor is prevented by a system of limit switches. Tests (Paul 1982) have confirmed that the 60-ton crane cannot be inadvertently moved into the common area. For those occasions when it is necessary to operate both cranes in the common area, a bypass key is available that permits the limit switches to be overridden. This key is under administrative control and is not normally available to the crane operator.

The 15-ton crane was designed and constructed according to the nationally recognized codes and standards listed in Section 9.3.2.2. The use of these codes and standards has historically resulted in safe, functional, and reliable cranes, as evidenced by the good performance record of the 15-ton crane. Reliable operation is ensured by maintenance checkouts.

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The crane and the in-cask hoisting equipment were designed to nuclear and industrial criteria, but neither can be considered single-failure-proof. In addition to preventive maintenance, therefore, other measures are taken to minimize the possibility of failure. These measures include maintenance activities as prescribed in PDD-6000, “INL Nuclear and Non-Nuclear Maintenance Management Program.” Chapter 15 demonstrates that the radiological consequences of postulated accidents involving cask drop or the failure of in-cask hoisting equipment are within the accident frequency/consequence guidelines. Health Physics coverage is provided for all fuel-handling operations.

9.3.2.3.3 Manual Fuel-Handling Tool—The manual fuel-handling tool is adequate to handle the fuel assemblies and is compatible with the fuel assemblies. Because the manual tool is generally used for minor adjustments of fuel assemblies already in the core, the possibility of damage is extremely remote. When radiation fields are low, the manual fuel-handling tool may be the normal method to move fuel elements. The tool has a positive locking device to minimize the potential for a fuel element drop. However, should an assembly be damaged, the radiological consequences would be within the accident frequency/consequence guidelines (see Chapter 15).

9.3.2.4 Testing and Inspections. Section 9.12.1 describes the FHC Testing and Inspections. Based on past operations, the 15-ton crane and cask handling tools have proven to be highly reliable equipment. Preventive maintenance of the crane is conducted at regular intervals. See INL Laboratory-wide Manual 14C, “Worker Safety and Health Program” for additional information on periodic maintenance programs.

9.3.2.5 Limitations and Setpoints. The 15-ton crane has interlocks to prevent interference with the 60-ton crane over the top of the reactor. Section 9.12.1 describes the FHC limitations and set points.

9.4 Experiment Loop Storage and Handling

This section discusses the equipment and systems used for storing and handling experiments at the TREAT facility. LST-387 contains information on the specific limitations and procedures for storing and handling fissionable material at the TREAT facility. The experiment loop storage and handling systems were designed and constructed to meet the requirements of GDC 6, 13, 61, 62, and 63 (see Chapter 3).

The requirement of GDC-6, “Design Bases for Experimental Facilities,” will be met by the ESA process. The ESA shall demonstrate that the design, construction, and placement of each experimental facility in or next to the reactor core shall be analyzed for inherent safety questions associated with the active or passive performance of that facility and with its safety-related interactions with the reactor facility. The design bases for each experimental facility shall reflect appropriate consideration of normal and accident conditions in systems or components next to the facility and normal and accident conditions within the scope of activities expected to be performed in the facility. Design requirements for experiments are discussed in Chapter 10, Experimental Facilities and Utilization, Section 10.2.2.

The requirement of GDC-13, “Instrumentation and Control” will be met by appropriate design and instrumentation provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions to ensure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, and the integrity and effectiveness of the experiment containment. Appropriate instrumentation shall be provided to ensure that TS-420, LCO 3.4.4 is met prior to experiment handling and storage.

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The requirement of GDC-61, “Fuel Storage and Handling and Radioactivity Control,” Part 1, “designed with capability to permit appropriate periodic inspection and testing of components important to safety,” as it applies to experiment loop storage and handling, is met by virtue of the following considerations:

• Inspection of a belowgrade loop storage hole (see Section 9.4.1.2) can be achieved by removing the loop storage hole covers, removing the loop, if necessary, and inspecting the hole visually

• Inspection, maintenance, and testing of the 60-ton crane (see Section 9.4.2.2) and the 15-ton crane (see Section 9.3.2.2) is in compliance with the PDD-6000

• Inspection, maintenance, and testing of the TLHC include cable and winch load testing and are controlled by the TREAT maintenance management system.

The requirement of GDC-61, Part 2, “designed with suitable shielding for radiation protection,” as it applies to experiment loop storage and handling, is met by virtue of the following considerations:

• Chapter 12 describes how the loop storage holes and covers and the TLHC provide suitable shielding.

The requirement of GDC-61, Part 3, “designed with appropriate containment, confinement and filtering systems,” as it applies to experiment loop storage and handling, is met by virtue of the following consideration:

• The rigid concrete and steel pipe lining of the storage holes, together with covers, provide appropriate containment and confinement.

The requirement of GDC-62, “Prevention of Criticality in Fuel Storage and Handling,” as it applies to experiment loop storage and handling, is shown to be met in Section 9.3.1.3. The results of calculations indicate that the experiment loop storage areas in the north end of the Reactor Building are safely subcritical under over-batching configurations, as well as single batching with flooding configurations. The subcriticality of the experiment loop storage holes in the south end of the Reactor Building is ensured when details of specific experiments are known.

The requirement of GDC-63, “Monitoring Fuel and Waste Storage,” Part 1, “to detect excessive radiation levels,” as it applies to the experiment loop storage and handling, is met by the radiation monitoring instrumentation described in Section 12.3.4. The requirements of GDC-63, Part 2, “to initiate appropriate safety actions,” is met by the audio and visual alarms associated with the area radiation monitors, which are described in Chapter 12.

9.4.1 Experiment Loop Storage

The experiment loop storage facility provides processing and storage capability for test loops and its supporting equipment.

9.4.1.1 Design Bases. Operations performed on the experiment loops in the floor pits include partial assembly of components, test train loading, final loop assembly and checkout, and partial disassembly of the posttest loop. Storage capability is provided for both partially assembled and completely assembled loops. The floor pit support flanges and shielding structures limit the number of irradiated components that can be handled simultaneously, because it is not planned to procure equipment for all pits.

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As noted in the foregoing discussion, the experiment storage facilities are able to handle and store a variety of equipment, including test fuel and radioactive and nonradioactive experiment hardware. Because these facilities are used in a variety of ways, a common set of design bases is inappropriate. Requirements for storing an item in a specific facility will be determined by a safety evaluation when the specifics of the subject experiment are known. The purpose of this evaluation is to ensure that criticality and radiation safety criteria are met.

The experiment storage holes in the south end of the building are equipped with steel covers to protect the contents. These covers are capable of supporting a 15,000-lb static wheel load. The area immediately surrounding each pit is capable of supporting the 50-ton LHM when it is positioned atop a pit for loop-handling operations. Ease of decontamination and the need for decay heat removal were addressed in the design.

9.4.1.2 Description. Figure 9-2 identifies the locations of facilities available for storing experiment equipment in the TREAT facility. These facilities are in the north and south high-bays of the Reactor Building. The north high-bay facilities are used for transferring experiment loops by the 15-ton crane and the TLHC. The south high-bay facilities could be used for the large loops, including transfer by the 60-ton crane (downrated to 20 tons as discussed in Section 9.4.2.1.1). The LHM was previously used for handling of experiments involving preirradiated test fuel and for posttransient handling of irradiated test loops. Even though the LHM will not be used with the operation analyzed in this FSAR, its use in this analysis is bounding and conservative.

The largest storage hole in the north end of the Reactor Building is an unlined concrete hole 10 ft wide, 12 ft long, and 10 ft deep. Covering the hole is a two-section concrete slab about 6-in. thick. Usually, the hole is used for storing loop experiment equipment only. Fissionable material allowed in this hole is determined in LST-387.

Near the largest storage hole in the north end of the Reactor Building are nine special experiment storage holes. Two of them, on 30-in. centers, are located north of the largest storage hole. A single hole is located east of the storage hole. A linear array of six pits, on 24-in. centers, is located south of the largest storage hole. These special experiment storage pits are 10 in. diameter and approximately 15 ft deep. Each pit consists of a Schedule 40 steel pipe of stepped construction embedded in concrete. The upper end of each pipe can accommodate a steel-lined, lead-filled cover. The special experiment storage pits are used for the storage of test loop experiments.

Next to one of the rectangular reactor fuel storage areas is an experiment storage hole area 4 ft wide × 8 ft long × 14 ft deep. To provide eight storage holes, the area was modified by filling the top 4 ft of the area with concrete, and embedding steel pipes in the concrete in a 2 × 4 array. The storage pits are 10 in. diameter, and approximately 14 ft deep. This area is used for storage of loop experiments.

The reactor fuel storage facilities are also suitable for storing experiments and fissionable material that are within the geometric envelope of fuel assemblies. The trapezoidal fuel-storage area has been used for experiment storage (see Section 9.3.1 for a description of the fuel storage areas.)

In the south end of the Reactor Building are 27 experiment storage/operations pits (see Figure 9-2). Three of these are 12 in., twenty are 24 in., and four are 40-in. diameter holes. These pits are used for storing large loops and equipment, and for experiment support operations. Each pit is 30 ft deep and consists of a Schedule 20 steel pipe embedded in concrete. The pits were formed by drilling an oversized hole in the lava rock and back-filling it with concrete so that at least 6 in. of concrete surrounds the pipe.

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At the top of each pit, monolithic concrete supports the pit flanges, thereby carrying cask loads directly to solid rock. The pits are equipped with lockable covers and alarms. The 2-in.-thick steel covers are designed to support a 15,000-lb wheel load. The pits are stepped, and radiation shielding plugs can be used if necessary. Some of the pits have provisions for cooling ducts and access for electric cables.

9.4.1.3 Safety Evaluation. The safety concerns for the experiment storage facilities include criticality safety and radiation safety. Criticality safety involves the storage of fissionable material in pits and storage areas. Radiation safety involves requirements for shielded storage of radioactive experiment material.

The criticality safety of the experiment storage facilities is addressed in detail in LST-387. Limits established in LST-387 are ensured by administrative control.

For radiation safety purposes, the combination of shielding and limited exposure is used to keep the cumulative dose within the requirements of DOE O 458.1. The pit shield plug is expected to provide adequate shielding. Under certain circumstances, however, it may be necessary to provide additional temporary shielding, and to restrict access to the area if the exposure criteria cannot be met.

Storage pit shield plugs for the pits located in the north end of the Reactor Building contain 5-3/4 in. of lead in a 3/8-in. steel lining. Experience has shown that these plugs provide adequate shielding for the experiment equipment normally stored in the pits. Typically, the exposure rate above one of the storage pits is about 0.2 mR/h.

Radiation shielding calculations (Yang August 1982) for the experiment storage/operations pits in the south end of the Reactor Building indicate a dose rate of 3.27 × 10-4 mR/h at the top of a storage pit containing a preirradiated test train without pit covers in place. Pit covers are available, and the pits are stepped to accept additional shielding if required.

Each storage pit provides considerable protection for the stored experiment equipment because each pit is below grade and uses a steel pipe embedded in concrete. In the event of an earthquake, seismic forces might cause some concrete spalling and fracturing at the floor-pit joint. However, the steel pipe will protect the stored equipment from the spallation or fracturing of the concrete. The storage pit cover and shield plug will protect the stored experiment equipment against falling objects or tornado missiles.

Contamination of the experiment storage pits is not expected to occur. In the unlikely event that the pits become contaminated, specific decontamination procedures will be developed at that time. The smooth steel sides of the storage pit liners will facilitate decontamination.

The need for forced cooling to remove decay heat from preirradiated test trains was examined. Analysis showed that no forced cooling was required (Rieb 1981). However, certain pits have been so constructed that they can accommodate cooling equipment if experiment requirements should change.

The covers of the storage pits are equipped with lockable covers and with alarms. Radiation monitors provide coverage of the experiment storage areas. Chapter 3, Section 3.3.7 provides additional bases for the safe storage of experiment loops.

9.4.1.4 Testing and Inspections. The pits and storage holes are a passive design feature, and therefore require minimal inspection and maintenance consisting of visual inspection for damage and accumulation of dirt.

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9.4.1.5 Limitations and Setpoints. LST-387 contains the limitations that apply to the experiment storage areas.

9.4.2 Experiment Loop Handling

The experiment-handling equipment at the TREAT facility includes:

• A 60-ton overhead bridge crane

• A 15-ton overhead bridge crane

• Experiment-handling casks.

The 60-ton crane services the reactor and the south end of the Reactor Building. The 15-ton crane services the reactor and the north end of the Reactor Building. The experiment-handling casks are shielded casks used for handling radioactive materials at the TREAT facility.

9.4.2.1 Design Bases.

9.4.2.1.1 60-Ton Crane—A traveling overhead bridge crane with a capacity of 60 tons is provided to service the reactor and experiment loop operations above and south of the reactor. The crane has a 10-ton capacity auxiliary hoist to serve the storage holes and high-bay loop work areas. The 60-ton capacity was dictated by the projected weight of some experiment system components; experiment-handling loads governed the choice of the 10-ton auxiliary hoist. However, the 60-ton crane has been downrated to 20 tons. The downrating of the crane to 20 tons was due to completing only a partial load test (25 tons) following corrective maintenance to the main lift hook (Stevens 1984). A final full load test was planned but never completed. The crane coverage area includes the radiography facility, upper mezzanine, LHM storage stand, experiment assembly tower, BAR, and accessible loop storage holes in the south high-bay.

The crane has both cab and radio control capabilities, with the latter providing remote control by the operator from any point within the high-bay areas. Limit switches are provided for each bridge and trolley motion to prevent over travel in any direction. The crane is the vehicle normally used to transfer fissionable material shipping containers from trucks to storage area locations. The crane is 45 ft above the top of the reactor. The 60-ton crane weighs 63 tons.

The 60-ton crane is designed and constructed to standard industry specifications. The governing specification for the crane is the industry standard, CMAA-70 (2000). The 60-ton overhead crane and support structure meets current Performance Category (PC)-2 design criteria (see Chapter 3, Section 3.3.5.1).

The crane has been assigned Class D (heavy duty) service classification. The higher service classification was selected to ensure greater safety. Welding design and procedures conform to AWS D14.1. The 60-ton hook is constructed of American Society of Testing and Materials (ASTM) A668-Class D high-strength carbon steel. The hoisting cables are designed to meet the requirements of FED SPEC RR-W-410G (2010).

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The radio-control system conforms to the requirements of CMAA-70 (2000), and American Society of Mechanical Engineers (ASME) B30.2, “Overhead and Gantry Cranes, Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist,” specifically for “dead man” switches and spurious signal prevention.

9.4.2.1.2 15-Ton Crane—The 15-ton crane services the areas over the reactor and north of the reactor. Interference with the 60-ton crane is prevented by hard-wired electrical interlocks. Similar to the 60-ton crane, the 15-ton is a standard industrial crane. Section 9.3.2.2 describes this crane and its inherent safety functions.

9.4.2.1.3 Experiment-Handling Casks—The experiment-handling casks consist of the TLHC and the FHC. The design bases for the FHC are described at Section 9.12.1. The other cask is discussed below.

9.4.2.1.4 TREAT Loop-Handling Cask—See Section 9.12.2 for discussion of TLHC design bases.

9.4.2.2 Description.

9.4.2.2.1 60-Ton Crane—The 60-ton overhead crane is a top-riding, double-girder, overhead traveling-type bridge crane with a 67-ft clear span. This heavy-duty crane has an auxiliary hoist with a 10-ton capacity and hook travel that ranges from 20 ft below to 64 ft above the main floor. The 60-ton hook has a travel that ranges from the main floor to 60 ft above the main floor. The crane can be operated either from the operator cab or by remote radio control. The motion control system provides variable-speed operation.

The bridge end trucks are rigidly fastened to the girders with bolts torqued to specified levels. The connections, girders, and end trucks are designed for skewing forces equal to at least 10% of the weight of the fully loaded crane. Rail sweeps are provided.

Trolley frame construction is of welded steel, and trolley rail sweeps are also provided. The main and auxiliary hooks, sheaves, drums, and hoist gearing are steel construction.

The crane uses single hooks for both the main and auxiliary hoists. Each hook was tested before assembly in the load blocks. During the 200% load testing, no elongation was discerned. The hooks have been tested to 125% of the installed crane load. Vendor certification of cables testing showed compliance with the requirements.

All sheaves are made of steel and have guards to keep the wire ropes in the grooves when the hoist is slackened and to prevent the wire ropes from chafing. The drums and all hoist gearing are also of steel. The gearing is totally enclosed.

The hoist-holding brakes are designed for 125% of the maximum possible torque at the point of brake application. Each hoist has one eddy-current electric brake. The 10-ton hoist system has one mechanical load brake. The 60-ton hoist system has two mechanical load brakes.

The crane is also equipped with bridge bumpers, runway stops, trolley bumpers, and trolley stops. The crane technical specification, S3125-1001-AS-01 (ANL 1980), and all approved vendor submittals contain full design details.

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Each hoist is provided with a lower-limit switch, and each load block with two upper-limit switches. Limit switches are provided to prevent bridge and trolley over-travel. Electronic crash limit switches and interlocks prevent the simultaneous operation of the 15-ton and 60-ton cranes over the reactor. The 60-ton crane can be either floor-controlled, using radio signals, or cab-controlled, using a hard-wired system.

9.4.2.2.2 15-Ton Crane—The 15-ton crane services the reactor and the north high-bay area of the Reactor Building (see Section 9.3.2.2).

9.4.2.2.3 Experiment-Handling Casks—Two shielded casks are used for handling experiments at the TREAT facility: the FHC and the TLHC.

9.4.2.2.4 Fuel-Handling Cask—The FHC, described in Section 9.12.1, was specifically designed for loading and unloading TREAT reactor fuel assemblies. This cask is appropriate for experiments with the same general design as a fuel assembly, including the top handling fixture. This cask is not designed to survive high drops, nor is the internal cask hoist designed to be single-failure-proof.

9.4.2.2.5 TREAT Loop-Handling Cask—See Section 9.12.2 for description of TLHC.

9.4.2.3 Safety Evaluation. Safety concerns for the experiment-handling equipment at the TREAT facility are radiation and materials handling. Radiation safety involves the shielding provided by an experiment-handling cask. Materials-handling safety involves the adequacy of the cranes and the in-cask hoisting equipment.

Experimenters are responsible for showing that the experiment-handling cask will provide adequate radiation shielding for an experiment. Information on the shielding characteristics of the casks is made available at the TREAT facility and the assessment of shielding adequacy is presented in the experimenter’s safety analysis report.

9.4.2.3.1 60-Ton Crane—The 60-ton crane design codes and standards are common industry practice and historically have resulted in safe and reliable cranes. The safety and reliability of the 60-ton crane is further enhanced by the incorporation of various safety features (see Section 9.4.2.2).

The 60-ton crane has been in operation since October 1981. Although it is not designed to be single-failure-proof, the crane provides added safety by incorporating selected safety features, some of which are beyond those required on a typical industrial crane.

Tests (Paul 1982) were conducted to confirm that electrical “noise” typical of that present in the Reactor Building could not cause malfunctions in the radio-control system. In addition, TREAT personnel conducted tests to confirm that the radio-control signals do not interfere with reactor control and other equipment. Tests were also conducted to verify the operation of the electrical interlock system prevents the simultaneous operation of both overhead cranes in their common operating area. (ANL 1981)

In summary, all performance and operation checkouts have confirmed the safety and reliability of the 60-ton crane. Its record of satisfactory performance is maintained by a comprehensive preventive maintenance program. In addition, the crane is under daily surveillance by TREAT personnel during day-to-day operations.

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9.4.2.3.2 15-Ton Crane—The radiation and materials-handling safety aspects of using the 15-ton crane in the TREAT facility are discussed in Section 9.3.2.4. Chapter 3, Section 3.3.5 provides additional bases for the safe operation of the TREAT cranes.

9.4.2.3.3 Experiment-Handling Casks—

• Fuel-Handling Cask: Experiments handled by the FHC will be bounded by fuel assemblies (see Section 9.12.1).

• TREAT Loop-Handling Cask: See Section 9.12.2 for discussion of TLHC Safety Evaluation.

9.4.2.4 Testing and Inspections. See Section 9.12.2 for discussion of TLHC Testing and Inspections.

• The 60-ton crane hooks are periodically inspected. The hooks have surface punch marks so located that the distance between them can be measured to confirm any spreading caused by material yielding.

To demonstrate crane conformance to specifications, it is operated in unloaded and loaded condition, through the full range of the bridge, trolley, and hoist coverage. The following crane operations and features are tested:

• Hoisting and lowering

• Trolley travel

• Bridge travel

• Limit switches

• Locking devices

• Safety devices.

Prior to initial service, the crane was load-tested to 125% of its rated capacities of 60 and 10 tons. All load-testing conforms to INL Laboratory-wide Manual 14C, “Worker Safety and Health Program, 10 CFR 851,” and was performed in accordance with established procedure (ANL 1981).

9.4.2.5 Limitations and Setpoints. The building cranes have the following operational performance limitations:

• The main hoist shall have a hook travel of 60 ft above the main floor

• The auxiliary hoist shall travel from 60 ft above to 20 ft below the main floor

• Hoisting speeds and bridge and trolley travel speeds shall be multilevel.

The experiment handling casks have the following operation limitations:

• Shielded cask is moved only slightly (approximately 2 in.) above the floor.

The exceptions are:

• Lifts to the reactor (about 15 ft)

• Lifts to the radiography stand (about 5 ft)

• Cask-to-cask transfers (about 12 ft).

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9.5 Prevention of Inadvertent Criticality

This section summarizes the criticality safety approach and the controls for the prevention of an inadvertent nuclear criticality in the TREAT facility during described operations. The primary reference for TREAT-specific nuclear criticality safety is ECAR-1610, “Criticality Safety Evaluation for the TREAT Reactor Building.” This section describes the controls necessary for criticality accident prevention.

9.5.1 Requirements

The INL Criticality Safety Program is described in SAR-400, Chapter 6, Prevention of Inadvertent Criticality. The unique requirements for TREAT are discussed in the following sections.

The criticality safety program requirements and recommendations for INL facilities, including TREAT, and the basis for deriving operational criticality safety limits, are described in Laboratory Requirements Document (LRD)-18001, “INL Criticality Safety Program Requirements Manual.”

9.5.2 Criticality Concerns

The facility radioactive material inventory includes the fissionable material inventory as allowed by ECAR-1610. As described in Chapter 3, quantities of fissionable materials are associated with TREAT operations and activities that an evaluation of inadvertent criticality is required. The location of the material within TREAT varies depending on the operations or activities at a given time. Generally, fissionable material is either within a designated storage zone or is being actively handled for its intended purpose.

Nonreactor TREAT operations and activities addressed in ECAR-1610 include storage racks and configurations for normal and abnormal conditions. ECAR-1610 provides the technical basis for determining the controlled parameters and applicable limits for fissionable material storage. The Criticality Safety Evaluation (CSE) Report also addresses fissionable material handling, thereby providing the basis for the controlled parameters and applicable limits on handling. The scope of ECAR-1610 includes:

• Belowground storage zones

• Aboveground storage (all storage containers or cages at or above ground level)

• Criticality Safety Index (CSI)-limited package arrays

• Material handling (configuration, minimum critical number, optimum spacing, safe handling limit).

As described in Chapter 3, the radioactive material inventory in TREAT, which includes the fissionable material inventory with respect to this chapter, is expected to vary over time. It is feasible that materials in differing forms and configurations from a variety of programs may be stored or handled in the facility. As necessary, additional criticality safety analysis (e.g., CSEs or other technical studies) may be performed, within the initial boundary provided by ECAR-1610, to determine additional controlling operation parameters or limits. Consequently, the collective criticality safety analysis will continue to evolve over time as necessary to support these changes. ECAR-1610 establishes or defines overall facility scope with respect to criticality safety. These key features are summarized below in Table 9-1. In addition to determining the controlled parameters (e.g., mass, geometry, and form) and limits on those parameters, ECAR-1610 also defines the bounds of the analysis, which should also be considered in control selection.

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As appropriate, important assumptions are protected by controls and nonanalyzed situations are prohibited or otherwise controlled.

Table 9-1. Criticality safety scope for TREAT activities and operations.

Feature Detail

Operational status

• Fissionable material is in storage (storage is only defined as material being located in a storage zone or aboveground storage)

• Fissionable material is out of storage (that is, the material is being handled)

Types of storage equipment

• Belowground storage zones

• Aboveground storage (building)

• CSI-limited package arrays

Types of fissionable material

• Fuel assemblies (Pu, 235U)

• Uranium metal, carbide, hydride, oxide (235U)

• Various forms as approved

Postulated criticality accident scenarios are derived from the CSE, which determines the critical configurations. Although ECAR-1610 addresses specific fissionable material configurations, such as TREAT fuel, a generic set of criticality accident scenarios can be derived for the facility as a whole. These scenarios include:

• Criticality related to fissionable material storage

• Criticality related to fissionable material handling.

Fundamental to the control of both of these scenarios in facility operations is identifying the fissionable material being stored and handled. The specific types of materials (e.g., TREAT fuel elements) can be visually identified by size, shape, and other unique characteristics. However, that is not the case for the fissionable material containers. Over-batching, or loss of mass control, is the fundamental parameter failure necessary for criticality, so knowledge of the process concerning TREAT inventory is essential to criticality safety.

9.5.3 Criticality Controls

This section describes engineered controls and administrative controls necessary to prevent inadvertent criticality.

9.5.3.1 Engineered Controls. If required, engineered (i.e., design) controls provide criticality safety of nonreactor TREAT operations and activities. From a risk perspective, prevention is related to a reduction in the likelihood of the event. As described in the accident analysis portion of Chapter 3, criticality was included in the set of representative, bounding, and unique accidents for TREAT. Further, criticality is the focal theme through ECAR-1610, which concludes no engineered controls are necessary for anticipated TREAT operations and activities related to fuels and materials.

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9.5.3.2 Administrative Controls. The administrative controls provided for TREAT nonreactor operations and activities that affect criticality safety are listed below. The administrative controls for nonreactor criticality accident prevention are derived in LST-387.

9.5.3.2.1 Approved Storage—Each storage position within TREAT is limited to a maximum capacity. For example, a storage position may be limited to a specific quantity of fuel assemblies/elements, maximum uranium equivalent, material loading, or a container array. Refer to ECAR-1610, specifically in Section 7, “Credited Controls,” for a complete description of the derived administrative controls.

9.5.3.2.2 Handling—Similar to the “Approved Storage” administrative control, the “Handling” administrative control places restrictions on the maximum number of fuel assemblies/elements, or maximum uranium equivalent. ECAR-1610 also allows for the development of a list of sources and standards. Refer to ECAR-1610, specifically at Section 7, “Credited Controls,” for a complete description of the derived administrative controls.

9.5.4 Application of Double Contingency Principle

The application of the double contingency principle from the perspective of the INL criticality safety program is described in SAR-400, Chapter 6, Prevention of Inadvertent Criticality, and has been implemented in LRD-18001. In the case of TREAT operations and activities involving the potential for a criticality accident, the engineered and administrative controls necessary for applying the double contingency principle are stated (or are embedded) within the applicable CSEs, in this case ECAR-1610. In turn, these engineered and administrative controls are incorporated into the facility safety basis to ensure a desired margin of safety against the occurrence of a criticality accident. Therefore, incorporation of the engineered and administrative controls as defined in ECAR-1610 is consistent with the guidance contained in DOE-STD-3007-2007.

9.5.5 Criticality Instrumentation

The requirements that determine the need for a criticality alarm system (CAS) are included in the INL Criticality Safety Program Requirements Manual, LRD-18001, which implements the relevant requirements of DOE O 420.1C and its referenced standard, ANSI/ANS-8.3 (1997), “Criticality Accident Alarm System.” From LRD-18001, the need for a CAS is evaluated for those activities with fissionable material inventories that criticality is a concern. In addition, the INL Criticality Safety Program requires that a CAS be provided when there are credible criticality accident scenarios (probability of criticality is greater than 10−6 per year) that result in doses that exceed 12 rads in free air. ECAR-1610, specifically in Section 6, “Evaluation & Results,” states that a CAS is not required in the Reactor Building.

9.6 Fire Protection Systems

Fire protection in the Reactor Building is designed to combat fires of typical combustible materials and fires associated with electrical equipment. Fire protection must also combat diesel-oil fires and liquid-metal (sodium) fires.

The fire protection program at the TREAT facility provides defense-in-depth. Two important aspects of this program are careful storage of combustible materials and caution in the use of hazardous materials such as sodium. Fire detection, another important element of the fire protection program, is accomplished with a modern system that provides comprehensive coverage of the TREAT facility.

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Finally, trained professional fire fighters use the equipment and procedures that are most appropriate for each specific situation.

Compliance with the requirements and recommendations identified in HAD-470 serves to satisfy the objectives of DOE O 420.1C and 10 CFR 851.

The fire protection system of the TREAT facility was designed to the requirements of GDC-3, “Fire Protection” (see Chapter 3,). The requirements of GDC-3 are shown within this FSAR to be met for the reactor, the F/CS, the instrumentation and control system, and the electrical power system in Chapter 4, Reactor, Chapter 5, Reactor Filtration/Cooling System, Chapter 7, Instrumentation and Controls, and Chapter 8, Electric Power Systems, respectively. The requirements of this GDC as it applies to the Reactor Building are met due to the following considerations (see Sections 9.6.1 and 9.6.2):

• Personnel doors and interior walls are of fire-rated construction where required for personnel and facility safety

• Redundant systems are physically separated so that fire-induced common-mode failures are highly unlikely

• By administrative control, the incidental use of combustible materials is strictly limited

• The fire detection and suppression systems include ionization and photoelectric smoke detectors, video-based flame detectors, wet-pipe sprinklers, sprinkler water flow alarms, clean agent suppression, manual alarms, various hand-held fire extinguishers, and a two-main water delivery system

• The facility is provided with the professional fire-fighting services of the INL fire stations located at the MFC, TAN and Central Facilities areas

• The reactor fuel-storage and experiment loop storage areas are equipped with channels that will drain any floor water away from the fuel-storage holes.

9.6.1 Reactor Building

9.6.1.1 Design Bases. The fire protection system at the TREAT facility is governed by DOE Order 420.1C. DOE-STD-1066-2012 provides the criteria and guidance for fire protection programs supporting implementation of this order.

HAD-470 provides a detailed assessment of the risks from fire and fire-related hazards associated with the current operations and processes performed within the TREAT reactor and associated buildings. HAD-470 evaluates fire hazards, building fire construction, and fire protection features. The following sections summarize pertinent details from HAD-470 that are important to the accident analyses in Chapter 15.

9.6.1.2 Description. Table 9-2 summarizes the fire hazards and the fire protection equipment associated with each area within the building from Table 3 in HAD-470. Fires in the following areas are potentially serious because they could affect systems associated with reactor I&C and F/CS:

• Diesel generator and electrical equipment rooms

• High-bay areas

• Instrumentation and control room

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• Basement subpile and auxiliary rooms

• Fan room.

Fires in the high-bay areas are also a concern because they may involve liquid sodium and test fuel. Section 9.6.1.3 presents a safety evaluation of these fires.

The layout of the Reactor Building (refer back to Figure 9-2) provides some separation of the critical fire areas. Fire walls can also isolate fires. A 2-hour fire wall isolates the redundant power diesel generator, the standby diesel-generator room, and the electrical equipment room from the rest of the building.

Similarly, an unrated fire wall isolates the electronics shop and the I&C room from the main area of the Reactor Building. The floor of the Reactor Building isolates the basement subpile and equipment rooms.

As pointed out in Section 9.6, the purpose of the fire protection program at the TREAT facility is to provide defense in depth. The fire protection system for the Reactor Building provides for quick detection and suppression of fires. The two components of the fire protection system consist of (1) fire detection and alarm and (2) fire-suppression. The following subsections discuss each of these components.

9.6.1.2.1 Fire Detection and Alarm—Fire detection equipment includes ionization and photoelectric smoke detectors, flame detectors, sprinkler water flow alarms, and manual fire alarms. Fire alarms are located within the building and the TREAT fire alarm system initiates local building alarm notification and transmits alarm signals to the INL Fire Alarm Center (FAC) located at the Central Facilities Area (CFA) 1611. The main fire alarm control panel integrates the fire detection and alarm systems. This panel has an internal battery backup.

9.6.1.2.2 Fire-Suppression—The fire suppression systems for the Reactor Building include both permanently installed and manual systems. The systems, designed for suppressing all classes of fires, are described in detail in HAD-470.

9.6.1.2.3 Water System—The main water system is described in Section 9.7. The Reactor Building fire water distribution includes fire pumps and fire hydrants, which connect to the TREAT water system. The fire and service water pumps receive their supply from water storage tanks and supply water to the sprinkler distribution system as demanded.

Table 9-2. Reactor Building fire hazards and fire protection equipment. Building Area Fire Hazards Fire Detection Fire Fighting

Diesel Generator Room Diesel Fuela Smoke Detector

Sprinkler Water Flow Alarm Sprinkler; portable fire extinguisher (ABC)

Electrical Equipment Room

Electrical Equipment Smoke Detector; Sprinkler Water Flow Alarm

Sprinkler; portable fire extinguisher (ABC)

Work Shop Miscellaneous Combustibles and Flammable gas


Hodoscope Room Electrical Equipment Smoke Detector; Sprinkler Water Flow Alarm


Fan Room Electrical and Mechanical Equip.


Mechanical Equipment Miscellaneous Combustibles Sprinkler Water Flow Alarm Sprinkler; portable fire

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Building Area Fire Hazards Fire Detection Fire Fighting Room extinguisher (ABC) Storage Areas Miscellaneous Combustibles Smoke Detector; Sprinkler

Water Flow Alarm Sprinkler; portable fire extinguisher (ABC)

High-bay – North Miscellaneous Combustibles Video Detection INL Fire Department High-bay – North (Reactor Area)

Miscellaneous Combustibles Video Detection INL Fire Department; portable extinguisher (Class D)

High-bay – South Miscellaneous Combustibles Video Detection INL Fire Department Work Areas Electrical Equipment; Sodium Smoke Detector INL Fire Department Truck Parkb Gasoline; Diesel Fuel Smoke Detector INL Fire Department Retention Tank Room Electrical Equipment Sprinkler Water Flow Alarm Sprinkler; portable fire

extinguisher (ABC) Experimenters Room Electrical Equipment Smoke Detector; Sprinkler

Water Flow Alarm Sprinkler; portable fire extinguisher (ABC)

Supervisor’s office Electrical equipment Sprinkler water flow alarm

Sprinklers, Portable extinguisher (ABC)

Electronics Shop Electrical Equipment Sprinkler Water Flow Alarm Sprinkler; portable fire extinguisher (ABC)

Instrument & Control Room

Electrical Equipment Smoke detectors at ceiling and under raised floor

Clean agent system; portable fire extinguisher (ABC)

Restrooms Miscellaneous combustibles Sprinkler water flow alarm Sprinklers, Portable extinguisher (ABC)

Health Physics Field Office

Miscellaneous Combustibles Sprinkler Water Flow Alarm Sprinkler; portable fire extinguisher (ABC)

First Mezzanine Electrical Equipment; Hydrogen

Smoke Detector; Sprinkler Water Flow Alarm

Sprinkler; portable fire extinguisher (ABC) INL Fire Department Second Mezzanine Plastic Storage Bags Smoke Detector

Basement Subpile Room

Electrical & Hydraulic Equip. Smoke and oil mist detectors; Sprinkler water flow alarm


Basement Auxiliary Room

Electrical & Hydraulic Equip. Smoke Detector; Sprinkler Water Flow Alarm


a Diesel fuel will be supplied by the day tanks in the rooms, and a 500-gal storage tank outside.

b When truck tractors or forklifts are present in the building, their fuel presents a fire hazard.

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9.6.1.3 Safety Evaluation. Certain areas of the Reactor Building may present fire-related hazards to facility safety. The safety significance of such postulated fires depends on the status of the reactor. The three reactor conditions of relevance are:

• Shutdown

• Operation in the low-power, steady-state mode

• Operation in the transient irradiation mode.

The areas of the Reactor Building that may present fire hazards are:

• Diesel generator and electrical equipment rooms

• High-bay areas

• I&C room

• Subpile and basement auxiliary rooms

• Fan room

• Hodoscope area

• Radiographic facility

• Truck fire in truck park in MFC-720

• Fire in the workshop in MFC-720.

Each of these critical areas is discussed in detail in HAD-470 and summarized below.

9.6.1.3.1 Diesel Generator and Electrical Equipment Rooms—These rooms contain diesel-fueled engines and electrical switchgear. A potential fire would most likely be caused by a malfunction of the engines or the switchgear. HAD-470 postulated fire in the diesel generator room involves the failure of the 10-gal day tank, and the subsequent ignition of the spilled diesel fuel. The pool is assumed to be confined to the bermed area. The effects of a postulated fire in the diesel generator room would be localized because of the fire walls and the small amount of combustible material in the room. The use of these rooms for storage or other activities that might introduce combustibles into the rooms is prohibited by PRD-14401, “INL Fire Protection Program Requirements.” Further, the diesel room and electrical equipment rooms are protected by a smoke detector, sprinkler water flow alarm, the automatic sprinkler system, and portable ABC extinguishers. A leak detector will also shut off the fuel supply.

Loss of electrical power results in reactor shutdown. Electrical power is not needed to maintain the reactor in a safe shutdown condition. Therefore, a fire in the electrical equipment room would not affect reactor safety, nor could it initiate a reactor accident. The redundant power system can supply power to one of the two blowers of the F/CS; the other blower is supplied by the normal power system (see Chapter 5). One F/CS blower can perform the intended function if the other blower is disabled. It is not required to run the 130-kw diesel generator (redundant power) to supply the F/CS during normal operations.

The requirements of the UBC and NFPA (and other codes and standards invoked thereby) have been complied within the design, manufacture, and installation of these rooms and the associated

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equipment. An example is the use of the fire-rated wall between the diesel generator room and the electrical equipment room.

9.6.1.3.2 High-Bay Areas—Loop-support operations take place in the high-bay areas of the Reactor Building. Because experiments at the TREAT facility may contain sodium-cooled test fuel, the potential exists for a sodium fire. However, because of the isolation of experiment loop storage and handling areas, it is unlikely that the fire would spread beyond the high-bay area. The area is fairly open and has a concrete floor. No incidental storage is allowed in the area (loop storage is below ground in the pits), and no other significant combustibles are customarily present in the area. In addition, the area is covered by smoke detectors, sprinkler water flow alarm, video detection, and portable fire extinguishers. During loop-support operations, personnel would be present in the area to provide awareness of potential hazards. The sodium is contained in high-integrity vessels, and all activities involving sodium are conducted in accordance with standard procedures.

The accident with the reactor operating and a sodium fire occurring in the high-bay area was considered. It was qualitatively concluded that there was a low probability that the fire could interfere with the ability to shut the reactor down. As discussed above, such a fire would be confined to the immediate area. The sodium would not be contaminated, and the test fuel cladding would likely survive the effects of a sodium fire without damage (see Chapter 15). Therefore, the radiological impact of a sodium fire that occurred during experiment-filling operations would be insignificant. The first mezzanine, which is at the same elevation as the top of the reactor, has water sprinkler coverage.

9.6.1.3.3 Instrumentation and Control Room—This room contains a variety of electrical and electronic equipment associated with the reactor and with the experiments. The primary source of a potential fire is the equipment itself. A HAD-470 postulated fire scenario assumes that 300 lb of ordinary combustibles occupying a 3 × 6 ft area is ignited by an electrical short. The room is equipped with a smoke-detection system and a clean agent (Novec 1230) system that provides for the prompt detection and suppression of a fire. To ensure that a fire is confined to its origin, the room is constructed of fire-resistant materials throughout (walls, ceiling, and floor) in accordance with UBC requirements. Information and control channels associated with the reactor are redundant and physically separated from each other to ensure against fire-induced common-mode failures.

If a fire occurred in the I&C room with the reactor shut down, it would have no impact on reactor safety. If one occurred with the reactor in low-power, steady-state operation, the reactor could be shut down, either by the operator in the reactor control room, or with one of the manual scram buttons in the Reactor Building. The scram could take place before the control equipment could be seriously damaged.

9.6.1.3.4 Basement Rooms—These rooms house the reactor control-rod-drive mechanisms and associated equipment. The fluid for the compensation/shutdown rod drive latches is Mobil DTE 24 hydraulic oil, which has a flash point of 428°F. The fluid for the transient control rod drives is Mobil DTE 25 hydraulic oil, which has a flash point of 450°F. The hydraulic fluid lines are designed for high-pressure service and the hydraulic reservoir is designed to ASME Boiler and Pressure Vessel Code criteria.

The following features of the subpile and basement auxiliary rooms minimize the chance of a fire and prevent its spread if one should occur:

• No miscellaneous storage that could support combustion exists in either room.

• The rooms are constructed of concrete.

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• Both rooms have water sprinkler systems.

• The hydraulic system has demonstrated reliability, having never experienced an oil mist fire.

• For further defense-in-depth, an oil mist detection system, capable of detecting concentrations below combustible levels, is located in the subpile room and in the adjacent auxiliary room. Upon detection of an oil mist in either of the rooms, the detection system automatically shuts off the pumps (which scrams the reactor) and closes the F/CS bypass valve.

• Smoke detectors are located in the subpile and auxiliary rooms to detect ordinary products of combustion.

A fire in this region has been considered in HAD-470. The transient control rod system, located in the basement, holds a total of 300 gal of hydraulic fluid. The postulated fire involves a leak in the hydraulic system resulting in a spill that subsequently ignites. The pool size is limited by the size of the basement. The control mechanisms would probably be damaged, but would have to be severely warped or melted to allow the control rods to come out of the core. The thickness of the metal and the other heat sinks in this room make this a low (qualitatively) probability event. Should the power supplies to the control/shutdown, compensation/shutdown, or transient rods be lost as a result of a fire, an immediate scram would result.

9.6.1.3.5 Fan Room—The fan room contains portions of the F/CS, consisting of the exhaust fans, fan motors, HEPA filters, and associated instrumentation and controls. The electrical equipment and instrumentation are designed and fabricated according to the requirements of the NFPA. The room contains no incidental storage or miscellaneous combustibles. A potential fire would, therefore, most likely originate in the equipment itself, and its spread would be slow because of the lack of combustibles and the fire-resistant construction of the room (concrete floor and minimum of one-hour fire-rated masonry walls). A sprinkler system protects the entire room.

If a fire occurred in the filtration/cooling room with the reactor shut down and blower operation required, the fire would have no impact on reactor safety. Blower operation under this operating condition is required only as a precautionary measure to support the personnel ALARA program and is not associated with reactor safety (see Chapter 5 for a detailed discussion of the F/CS).

If a fire occurred in the filtration/cooling room during reactor operation, the fire would have no impact on reactor safety, and the reactor could be brought to a safe shutdown condition. Blower operation is required only as a precautionary measure (see Chapter 5).

9.6.1.3.6 Hodoscope Areas—The hodoscope areas are located immediately north and south of the reactor. These areas contain electrical wiring associated with the hodoscope equipment. No other combustible material is located there. Therefore, potential fires would have to originate in the electrical wiring. No automatic fire suppression equipment protects this area of the Reactor Building, although manual ABC extinguishers are located nearby and the area is monitored by smoke detectors.

Damage from a fire in this area would be limited to the immediate vicinity because of the lack of other combustibles in adjacent areas. No significant damage would be sustained by the reactor shielding, and an aluminum plate prevents products of combustion from entering the reactor cavity. Should this plate somehow be breached, some smoke would be drawn through the small openings associated with the hodoscope installation into the reactor core and through the exhaust system. It is unlikely that the F/CS would be operating during this condition.

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The only potential threat to reactor safety relates to the cable conduits containing the reactor control and information cables. These conduits are located on the reactor shielding above the hodoscope on the north side of the reactor. With the reactor shut down, damage to these cables would not present a safety hazard. Damage to the cables with the reactor in either the steady-state or transient mode, with resultant loss of power, would cause a reactor scram (rod-holding magnets would lose power and release the scram rods). Furthermore, the control and information cables are redundant; should a malfunction occur in one channel but not lead to a scram, the other control channel would still be available to scram the reactor.

9.6.1.3.7 Radiography Area—The radiography area is located on the west face of the reactor shielding on the main operating floor. Two small 12-V motors and their associated wiring are located in this area. One of the motors drives the shield door and the other the foil carriage. Approximately 80 gal of mineral oil used as neutron shielding is also located in the area, the bulk of it in a welded steel enclosure. The remainder, approximately two quarts, is contained in a small plastic carboy located on the floor adjacent to the shield, approximately 3 ft from the 12-V electric motors.

A fire in this area would originate from a malfunction of one of the electric motors or the wiring. Such a fire should present no hazard to the mineral oil contained in the steel enclosure, because the “fire” would be small (more like a “smoldering”), and because the steel container of mineral oil would provide a heat sink effect.

The radiography area is monitored by smoke detectors. ABC extinguishers are readily available in the Reactor Building. TREAT personnel, including a supervisor, are always available for preventive and mitigative measures in the Reactor Building during radiographic operations. Hence, an electrical fire or short circuit would probably be observed, or would manifest itself as a malfunction of the door or carriage drives before it could spread. In the unlikely event that a fire should go undetected, the only potential threat to reactor safety relates to the cable conduits located on the reactor shielding above the radiography area. Damage to the control cables contained in the conduits would result in a reactor scram if power were lost to the rod-holding magnets. Furthermore, all control and information cables are redundant; should a malfunction occur in one circuit that did not result in a scram, the other circuit would be available to scram the reactor.

9.6.1.3.8 Truck fire in Truck Park—The fuel from a diesel truck (50 gal) leaks onto the floor in the truck park located on the southeast end of the facility and subsequently the fuel ignites. The pool is assumed to be partially confined with a depth of 0.1875 in.

This postulated fire scenario assumes that 1 gal of flammable liquid is spilled and ignites. The fire is also assumed to ignite nearby combustibles. It is assumed that 150 lb of ordinary combustibles occupy this 3 × 6 ft area.

9.6.2 Control Building

Building 724, the Control Building, is used to house the TREAT control room.

9.6.2.1 Design Bases. Fire protection in the Control Building is designed to combat fires of typical combustible materials and fires associated with electrical equipment. The building is constructed with fire-resistant materials and is protected by both a wet pipe sprinkler system and clean agent system (Novec 1230). It is located more than 20 ft from Office Building (MFC-721), and a minimum 30-ft brush-free area is maintained around the Control Building.

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9.6.2.2 Description. Figure 9-6 shows a plan view of the Control Building, a one-story, metal-sided structure of 1,680 ft2 (gross). Interior partitions are constructed of fire-resistant gypsum board on metal studs, and the same material is used on the interior surface of the exterior walls. The ceilings are of fire-resistant acoustical tile. Cross zoned smoke detectors are used in the control room area. Sprinklers are located in the building except in the reactor control room which is protected by the Novec 1230 Clean Agent gaseous suppression system. In addition, a wall-mounted, hand-held fire extinguisher is available.

9.6.2.3 Safety Evaluation. The fire-resistant construction of the Control Building minimizes the likelihood of a fire. In the unlikely event that a fire did occur, the wet pipe sprinkler or clean agent system would suppress it. Exit doors are located at both the west and east corners of the building and are equipped with panic exit hardware. A fire in the Control Building is highly unlikely, but would have no impact on reactor safety if it started in or spread to the reactor control room while the reactor was not operating. As long as the reactor was shut down, the fire or the fire-fighting measures could, in the extreme, totally destroy the reactor control room without affecting reactor safety.

When the reactor is operating in the low-power, steady-state mode, a reactor operator is at the control console in the control room. In the event of an emergency, the operator could immediately scram the reactor and evacuate the Control Building; no further safety-related actions would be required. The fact that an operator can immediately scram the reactor reduces the chances of a fire developing so rapidly that it would disable the control console or force evacuation of the control room before the scram could take place. If such a situation were to occur, the reactor could still be shut down by using the scram circuits in the Reactor Building, which is occupied during steady-state operations.

When the reactor is operating in the transient-irradiation mode, the operating time is usually a few minutes or less but may be as long as 30 minutes. In this operating mode, the reactor is shut down by using control circuitry, such as the program timer and the control computer, located in the Reactor Building. Hence, a fire in the Control Building would not compromise shutdown of the reactor.

9.6.2.4 Inspection and Testing Requirements. An inspection and testing program ensures that the components of the fire-protection systems perform correctly. This program includes operational tests and checkouts. Inspection, testing, and maintenance of the fire alarm, fire-detection, sprinkler system, fire-sprinkler water flow, clean agent system, supervisory systems, and portable fire extinguishers are performed in accordance with applicable NFPA and DOE guidelines and standards.

9.6.2.5 Limitations and Setpoints. There are no limitations or set points for the Reactor or Control Building fire protection systems.

9.7 Water Systems

Water systems can be particularly important to safety at a nuclear power plant because of: (1) the need to cool critical heat loads, and (2) the presence of large quantities of possibly contaminated water. At the TREAT facility, on the other hand, the water systems do not interface with the reactor proper. However, they are the source of the water required for firefighting.

9.7.1 Domestic Water System

9.7.1.1 Reactor Building Area. The MFC site water supply and distribution system is a combination fire protection, potable, and service water system. For domestic purposes, the system can serve the sanitary needs of TREAT and provide cooling water for equipment. Domestic uses include

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once-through cooling for the building evaporative-cooled HVAC systems and the plant air compressor in the mechanical equipment room. Water for firefighting is received into the facility through two underground piping mains, which are fed from the MFC fire water and potable water system originating at the MFC water pump houses, MFC-707 and MFC-1740. The MFC fire water supply and distribution system is described in HAD-432.

9.7.1.1.1 Design Basis—The design of the domestic water system conforms to the requirements of the National Plumbing Code, the NFPA National Fire Codes, and the American Water Works Association. The water system is designed to the seismic requirements of the UBC for Zone 3, with an occupancy importance factor of 1.5.

9.7.1.1.2 Description—The water supply for the TREAT facility reaches the site through underground piping. It flows from the MFC fire-and-potable-water system through a 14 inch high density polyethylene (HPDE) firewater loop around TREAT which also supplies fire hydrants to provide perimeter coverage. As discussed in HAD-470, water capacity is limited by the normal pumping capacity of the two motor-driven MFC pumps. HAD-432 further describes the MFC firewater system capacities and flowrates.

The fire-water system within the Reactor Building is actuated locally by standard fusible-link sprinkler heads. Flow through the sprinkler system will activate an alarm. An alarm will also be activated if the water pressure available to the system decreases below a predetermined value. Outside the Reactor Building are three fire hydrants for use by the professional fire fighters.

The potable water system provides water to the sanitation facilities, the janitor's sink, and the drinking fountains. The lavatories and floor drains in the restrooms drain into a sump that is part of the suspect water waste retention system. The sink in the health physics laboratory drains into a bottle. The toilets in the restrooms are connected to the sanitary waste system.

The domestic water system also provides once-through cooling water for three building evaporative-cooled HVAC systems, hydraulic oil coolers in the BAR, and scram and instrument air compressors in the mechanical equipment room. The industrial waste water system removes this cooling water. None of the equipment contains radioactive material that could contaminate the domestic water system.

The domestic water system also provides water to the fire protection system with which four distribution manifolds are associated. Each manifold provides a pressure-reducing station, bypass leg, valves, gauges across the pressure reduction boundary, and a cumulative-type water meter. The west manifold is located in Room 105 adjacent to the west entry, the east manifold in the mechanical equipment room, and the guard station manifold above the suspended ceiling of the restroom. The fourth supplies the mezzanine, basement auxiliary room, and the subpile room. The flow alarm for this loop is in parallel with the east manifold. The fire protection system is separate from the suspect water waste retention system. There is no mechanism by which suspect material could contaminate the domestic water system through the fire protection system.

9.7.1.1.3 Safety Evaluation—The safety aspects of the domestic water system relate to interruption of fire protection water when it is needed. Only one kind of active component interfaces with the fire protection system: the pressure-reducing valves that reduce the incoming 100-psi line pressure to 40-psi for use throughout the building. Since the fire line is connected upstream of the pressure reducer, failure of the reducer would not affect the supply of the firefighting water.

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If the pressure reducer were to fail in the closed position (no flow through the reducer), there would be no effect on the fire protection system because its pressure would be unchanged. If a reducer should stick in the full-open position, the main line pressure would exist throughout the downstream piping, and water would still be available for firefighting. Even if the fire water failed, reactor safety would not be compromised.

Failures resulting from an arbitrary “leak” in a fire line are also considered. The worst kind of arbitrary “leak” would be a massive rupture. Should such a failure occur, the low-pressure switch in the fire protection system would alarm and draw attention to the problem. Such a failure is extremely unlikely; its occurrence coincidently with the need for fire-fighting water can be dismissed.

Another type of failure considered is damage to the fire protection system from natural phenomena, i.e., wind, snow, and earthquake. The Reactor Building was designed for wind and snow loads of a minimum recurrence interval of 100 years. Hence, the likelihood of damage to the fire protection piping from wind and snow overloads is remote. Even more remote is the likelihood that such damage would coincide with the need for fire protection. Regarding earthquakes, the building was designed and constructed according to the UBC provisions; and the fire protection system was designed and installed according to the requirements of the National Plumbing Code, the American Water Works Association, and the NFPA National Fire Code. As noted earlier, the ability to shut down the reactor and maintain it shutdown is not compromised by a failure of the water system.

9.7.1.1.4 Inspection and Testing—Routine corrective and preventative maintenance activities involving nonreactor equipment and systems are performed in accordance with facility maintenance instructions when applicable.

9.7.1.1.5 Limitations and Setpoints—There are no limitations or set points for the Reactor Building domestic water system.

9.7.1.2 Control Building. The water supply system can serve the sanitary needs of facility personnel.

9.7.1.2.1 Design Basis—The original Control Building water supply design complied with the UBC in effect at the time of construction.

9.7.1.2.2 Description—The Control Building (Building 724) receives its domestic water supply from the Office Building (Building 721) through an underground 3/4-in. copper pipe. The Control Building has a sprinkler fire protection system.

9.7.1.2.3 Safety Evaluation—The Control Building water supply system does not perform a safety function.


9.7.1.2.5 Limitations and Setpoints—There are no limitations or set points for the Control Building domestic water system.

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9.7.2 Sanitary Sewage System

9.7.2.1 Design Basis—The system was designed and constructed according to the provisions of the National Plumbing Code and the Idaho Department of Health and Welfare, Rules and Regulations for Individual and Subsurface Sewage Disposal Systems.

9.7.2.2 Description—The sanitary sewage system in the Reactor Building area collects waste material from the toilets in the restrooms. The system is designed to dispose of 825 gal of effluent per day through a septic tank and disposal field. The system provides sewage disposal for 55 personnel. Because this system is not interconnected with the suspect water waste retention system, no radioactive contamination of the system is possible. It is not connected to the domestic water supply either.

9.7.2.3 Safety Evaluation—The Reactor Building sanitary sewage system does not perform a safety function.

9.7.2.4 Inspection and Testing—Routine corrective and preventative maintenance activities involving nonreactor equipment and systems are performed in accordance with facility maintenance instructions when applicable.

9.7.2.5 Limitations and Setpoints—There are no limitations or set points for the Reactor Building sanitary sewage system.

9.7.3 Industrial Waste-Water System

9.7.3.1 Design Basis—The system was designed in accordance with the requirements of the ASHRAE and the National Plumbing Code.

9.7.3.2 Description—The Reactor Building industrial waste-water system conveys cooling water from the mechanical equipment room to the site surface drainage system. The maximum design flow rate through this gravity piping system is 300 gal/minute.

The equipment cooled by this system includes building HVAC equipment (air washers and evaporative coolers), compensation hydraulic latch, transient rod drive hydraulic system, and the scram air and instrument air systems. None of this equipment contains radioactive material that could contaminate the industrial waste-water system. Failure of the system to provide a cooling service is not a significant safety concern. Loss-of-water cooling to the compensation hydraulic latch or transient rod-drive hydraulic system would result in a high-oil-temperature alarm and reactor shutdown. The industrial waste water from the Control Building drains into an existing drainage ditch serving the Office Building (Building 721).

9.7.3.3 Safety Evaluation—The industrial waste water system does not perform a safety function.


9.7.3.5 Limitations and Setpoints—There are no limitations or set points for the TREAT industrial waste water system.

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9.8 Process Gases

The process gases at the Reactor Building include a compressed air system and an inert gas supply system.

9.8.1 Compressed Air Systems

Two independent compressed air systems serve the TREAT facility:

• Plant and instrument air system

• Control-rod-drive scram air system.

The compressed air system furnishes plant and instrument air to the Reactor Building through distribution piping systems at 20 to approximately 250 psi. System equipment includes:

• After-coolers

• Receiver tanks

• Pressure relief valves

• Check valves

• Filters and driers

• Pressure reducing valves

• Distribution piping.

The systems have automatically controlled high- and low-pressure switches to actuate motor operation.

9.8.1.1 Plant and Instrument Air System

9.8.1.1.1 Design Basis—The instrument air system conforms to the requirements of ASTM B88, “Standard Specification for Seamless Copper Water Tube.”

9.8.1.1.2 Description—The compressed air system provides plant and instrument air to the Reactor Building through distribution piping systems. This system supplies air at 20 to 120 psi. It serves the following areas and functions:

• F/CS controls

• Building ventilation controls

• Air cleaning stations

• Pneumatic tools.

9.8.1.2 Safety Evaluation. The impact of a loss of instrument air is only important as to its effect on the F/CS. Upon loss of instrument air, the F/CS flow control valve will fail in the open or partial-open position, depending on the instrument air supply pressure. Nevertheless, since the flow control valve would remain at least partially open, its function will be maintained (see Chapter 5).

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9.8.1.2.2 Limitations and Setpoints

• The air pressure is controlled between 90 and 120 psig

• The air receiver is designed for operating pressures to 150 psig

• Air filters are rated at 175 psig

• Tubing is rated well over the maximum pressures that can be generated by the air compressor.

9.8.1.3 Control-Rod-Drive SCRAM Air System.

9.8.1.3.1 Design Basis—The control rod drive scram air system conforms to the requirements of the ASTM B88.

9.8.1.3.2 Description—The control-rod-drive scram air system supplies air to the reactor control/shutdown and compensation/shutdown rods. This system has no other function than to provide pressurized air for scramming the control rods. Hence, the operation of the system cannot be compromised by other equipment or systems. A compressor supplies approximately 250 psi air to eight accumulator tanks, one for each of the control rod drives. Each tank has its own check valves and distribution piping to its control rod drive. Air for scramming comes from the cylinder surrounding the control rod and from the tank; the compressor need not be operating for a scram initiation. The compressed air piping and tanks conform to the appropriate ASTM and ASME standards.

9.8.1.3.3 Safety Evaluation—The pressure in each tank is monitored and each tank is equipped with a low-pressure alarm and with a low-pressure scram circuit. If the tank pressure drops below a preset value, an alarm is sounded; if it drops to the scram threshold, an automatic scram occurs. The independent systems ensure that a single failure can impact only one control rod drive; the remaining rods are capable of bringing the reactor to shut down and maintaining it in the shutdown condition.

9.8.1.3.4 Inspection and Testing—The system receives periodic maintenance, consisting of cleaning the filters and separators and checking the operation of alarms, the scramming circuit, and the compressor. All the equipment and piping is readily visible, and hence, can be easily observed by TREAT personnel.

9.8.1.3.5 Limitations and Setpoints—The maximum control-rod-drive air pressure is set at 250 psig.

9.8.2 Inert Gas Supply System

9.8.2.1 Design Basis—The distribution piping design for the system corresponds to ASTM B88.

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9.8.2.2 Description—Cylinders inside the building supply inert gases to stations in the south high-bay area and to a station adjacent to the reactor. Four cylinders each of argon, helium, and nitrogen can be connected by piping manifolds to distribution piping. The piping extends from the manifolds along the south, north, and east walls of the high-bay to both mezzanines east of the reactor. Each cylinder contains 200 standard cubic feet of gas and the pressure of the supply is reduced to approximately 150 psi at the pressure-reducing station. The inert gas supply system has relief valves vented to the outside atmosphere. A low-pressure alarm is provided for each gas supply. Each tank has a shutoff and check valve, the latter to prevent potentially contaminated gas from contaminating the supply tank.

This system furnishes inert gases for experiment-support operations; it is not used for fire-fighting or for maintaining a cover gas continuously over a liquid-metal surface. The gases are stored in standard industrial high-pressure tanks. The tanks are supported by wall brackets and chains that hold them securely in place.

9.8.2.3 Safety Evaluation—A malfunction of the system could disrupt experiment-support operations, but would not present a reactor safety hazard.


9.8.2.5 Limitations and Setpoints—The maximum flow rate in the distribution piping is 0.057 lb/s.

9.9 Heating, Ventilation, and Air Conditioning

9.9.1 Design Basis

The ASHRAE standard was used at the design/build of the TREAT HVAC systems and components. The Code of Record at the time of original design/build as amended by subsequent modification is preserved in accordance with site-specific configuration management.

9.9.2 Description

The HVAC systems for both the Reactor Building and the Control Building perform no safety functions. Neither system is an engineered safety ventilation or atmospheric cleanup system. The HVAC systems are not designed to service critical heat loads.

9.9.2.1 Reactor Building. The Reactor Building HVAC system provides:

• Electrical heating for the building

• Make-up ventilation air for the reactor cooling system

• Fresh air for the building through ventilating fans and louvers, all regulated by a master control

• Air conditioning for the I&C room, the electronics shop, hodoscope room, the experimenters room, and the health physics room.

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9.9.2.1.1 North High-bay Area—During warm weather, the exhaust air is removed by four exhaust fans. Two are located high in the north wall and two high in the south wall.

During cold weather, most of the building air is recycled through the duct resistance heaters. The recycled air is returned from a high elevation of the high-bay through a sheet-metal return air duct. Only enough of outside air is introduced to keep the space at a slightly positive pressure relative to the outside atmosphere. Make-up air is introduced by automatic actuation of a damper on the fresh-air intake during operation of the reactor exhaust system. Unit heaters are provided in the high-bay above the east and west roll-up doors to compensate for heat loss in those areas.

The air-conditioning system in the north high-bay area consists of two evaporative coolers. One cooler is located on the roof above the diesel generator room; motorized dampers interlock with existing exhaust fans, which are used for warm weather exhaust. The second evaporative cooler is located above the hodoscope room on the east side.

In addition to the electric unit heaters, a make-up air system is located in the mechanical equipment room and connected to existing ductwork serving the north high-bay area. This system, consisting of a filter section, a thermostatically controlled electric resistance duct heater, and a fan cabinet, provides cold weather heating. A wall-mounted unit heater is provided at the north roll-up door to compensate for local area heat losses.

9.9.2.1.2 Instrument and Control Room—The I&C room is cooled by two computer room-type air conditioning units, using two air-cooled condensers located at ground level outside the west wall. Each unit has electric heating and humidification capabilities.

Conditioned air is delivered under the raised floor, from where it is directed through the instrument racks to a plenum region above the suspended ceiling. Pressing through the racks, the air cools the instruments, then is returned to the external units through registers located in the suspended ceiling. Additional floor registers introduce conditioned air directly to the room space for personnel comfort.

A supply fan-and-filter assembly introduces outside air above the suspended ceiling of the I&C room. This air serves as make-up air for exfiltration to the personnel area just south of and adjacent to the I&C room. It mixes in the plenum formed by the suspended ceiling and building roof. Exfiltration from the I&C room to the personnel area corridor is through a register/damper located below the ceiling and above the door. This exfiltration is balanced with the flow of the restroom exhaust fan.

9.9.2.1.3 Electricians Office and Electronics Shop—Both the electrician’s office and the electronics shop are cooled by a computer room-type unit, using an air-cooled condenser located at ground level outside the west wall. The unit has electric heat and humidification capabilities. Conditioned air is supplied to the spaces through ductwork above the suspended ceiling, from where it exits through ceiling diffusers. The air introduced to the rooms, then returns to the outside unit through a return-air grille in the partition between the two rooms. Based on the occupancy of each room, outside air requirement is offset by the infiltration through doorways.

9.9.2.1.4 Shift Supervisor/Health Physics Room—The Shift Supervisor/HP room is serviced by an air-conditioning unit located on the roof of the HP room. Ductwork for this unit is located above the suspended ceiling and provides air through ceiling diffusers to the HP spaces and to the instrument room. During cold weather, the HP office is heated by electric baseboard heaters.

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9.9.2.1.5 Diesel Generator Room—The air requirement for operation of the generators and for general ventilation of the diesel generator room is provided by outside air-intake dampers. This air is balanced by duct connections between the diesel units and outside louvers and by roof exhaust fans. The roof fans are controlled by a thermostat to permit room cooling by outside air flow during warm weather. When the generators are not operating, unit heaters are used to heat the room.

9.9.2.1.6 Electrical Equipment Room—The electrical equipment room is cooled by routing air through air-intake dampers in the side walls to exhaust fans on the roof. The dampers and fans are thermostatically controlled.

9.9.2.1.7 Filtration/Cooling System Room—The exhaust fan and damper is located in the roof above the fan room. A motorized damper is located above the east door. The fan room is equipped with electrical unit heaters.

9.9.2.1.8 Exhaust Duct System—A 6-in. diameter carbon steel pipe line collects potentially contaminated gases from the experiment equipment located along the south and east walls of the high-bay area and along the east face of the reactor. The piping includes a check valve and butterfly valve, and is connected to the reactor-cooling-system piping at the suction side of the HEPA filters. The butterfly valve will be closed and tagged to prevent inadvertent introduction of air into the F/CS. Further, all of the inlets to the 6-in. duct will be closed and tagged, identifying that they should be opened only during venting operations.

9.9.2.1.9 Mechanical Equipment Room—The mechanical equipment room is ventilated in summer through an outside air-intake damper and roof exhaust. The room is heated in winter by electric unit heaters.

9.9.2.1.10 Hodoscope Room, Workshop, and Utility Room—The hodoscope room, workshop, and utility room are heated in the winter by electric unit heaters. The hodoscope room has its own air conditioning system. The workshop has manually operated vent fans. The utility room is not ventilated.

9.9.2.1.11 Restrooms and Janitor’s Closet—The restrooms and janitor’s closet are ventilated by roof exhaust fans. Make-up air infiltrates each room from the adjacent corridor. The restrooms are heated with electric wall heaters.

9.9.2.2 Control Building. The Control Building is a single-story, metal-sided structure that contains the control panels and instrumentation for remote control of the reactor. HVAC are provided by an air conditioner that incorporates supplemental electric heating coils.

9.9.2.3 Contamination Control. A master shutdown switch is located in both the I&C room at the Reactor Building and the Control Building. The switch de-energizes all Reactor Building HVAC equipment, both supply and exhaust. This shutdown feature minimizes the potential spread of contamination if an accident occurs at TREAT. However, shutting down the HVAC equipment will not shut down the F/CS. Hence, any air drawn out of the Reactor Building by the F/CS passes through the HEPA filters.


There are no potential safety impacts resulting from a loss of the HVAC system.

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9.9.4 Inspection and Testing

Routine corrective and preventative maintenance activities involving nonreactor equipment and systems are performed in accordance with facility maintenance instructions when applicable.

9.9.5 Limitations and Setpoints

There are no safety limitations or set points for the TREAT HVAC system.

9.10 Communication Systems

9.10.1 Design Basis

The TREAT communication system design meets the requirements identified in DOE Order 5480.30, “Nuclear Reactor Safety Design Criteria,” as amended.

9.10.2 Description

9.10.2.1 Intra-facility Communications. Intra-facility communication systems include telephone, two-way radio, public address, and multichannel intercom systems. Telephone outlets are provided in each office and work area, including the Control Building. The radio system connects the Control Building and the Reactor Building. Inside the perimeter fence, the public address (PA) system provides complete coverage throughout the facility, including the guard post. PA speakers are provided in every work area in the Reactor Building, except small closet areas covered by speakers in adjacent areas. A multichannel intercom system with channel selector switches, plug-in jacks, and paging control is provided for communications between work locations in the Reactor Building, the Control Building, and the Office Building.

9.10.2.2 Facility-to-offsite Communications. The facility-to-offsite communication systems include a telephone system and a two-way radio. The telephone system provides for external communications from various areas in the Reactor Building and the Control Building. The portable two-way radio, located in the Control Building, can be used to contact the DOE Warning Communications Center.


The principal safety function of the communication systems is to ensure that communication is maintained in the event of an emergency. When normal power is lost, telephone system equipment is powered separately. The two-way radios are battery powered, and hence, independent of site power. The paging system and the multichannel intercom system are connected to the standby power system. The diverse nature of the power sources and the independent communication systems ensure that communications will function when needed.



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There are no safety limitations or set points for the TREAT communication systems.

9.11 Lighting Systems

Both normal lighting and emergency lighting systems are provided at the TREAT facility.

9.11.1 Design Basis

The emergency lighting system meets the requirements of the NFPA 101, “Life Safety Code.” Emergency lighting is tested periodically to ensure its proper operation.

9.11.2 Description

9.11.2.1 Normal Lighting. The lighting system for all interior and exterior areas of the facility meets or exceeds the requirements defined in the Illuminating Engineering Society Handbook. Lighting controls are provided to meet DOE lighting conservation requirements. Wherever practicable, building lighting is supplied by fluorescent lights. For high-bay lighting on the main floor of the Reactor Building, high-intensity discharge (HID) lighting is used in conjunction with fluorescent lighting fixtures. HID sources provide the perimeter fence lighting.

9.11.2.2 Emergency Lighting. In case of loss of normal power, battery-powered emergency lights in the Reactor Building and in the Control Building allow personnel to leave safely. In addition, standby power from the diesel generator supplies selected Reactor Building lighting to supplement the battery-powered emergency lights and to maintain a minimum lighting level throughout the building.


In case of a facility emergency, an emergency lighting system is available. This system is powered by batteries, which are designed to function to allow enough time to restore normal lighting or shut down coupled with facility evacuation.




The emergency lighting system is limited by the capacity of the battery supply.

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9.12 Casks

9.12.1 Fuel-Handling Cask

9.12.1.1 Design Basis. The FHC is designated as a NSR-AR SSC in Chapter 3, Table 3-2. The FHC provides adequate shielding for a Zircaloy-clad fuel assembly removed from the center of the core two hours after shutdown from a 1,000-MJ transient. The radiation field from the cask is less than 1 mR/h at a 3-ft operating distance.

The cask is compatible with Zircaloy-clad fueled and unfueled assemblies. The weight of the cask is accommodated by the 15-ton overhead crane. A storage facility is provided for the cask.

9.12.1.2 Description. The FHC consists of a lead- and steel-shielded cylinder 10 ft long. The cask weighs approximately 8.25 tons (see Figure 9-9). The cask body is double-walled, composed of a 20-in.-diameter outer steel (Schedule 20) tube and an 8-in.-diameter inner steel (Schedule 40) tube. The tubes are welded to a steel base plate 3/4 in. thick. The intervening 6-in. annulus is filled with lead shot of density equal to 75% solid lead. The cask is shielded at the bottom by a lead-filled drawer 4 in. thick.

Within the cask is a cylindrical shroud that houses a square spreading tube and a fuel-assembly gripper. The square spreading tube, driven up or down by a lead screw-and-motor drive, pushes outward on the upper ends of the assemblies adjacent to the fuel assembly being handled. The fuel-assembly gripper, a four-jaw, spring-loaded latching device, grasps the fuel assembly by the upper-end fitting, and is raised or lowered by a steel cable attached to a shaft that is rotated by the cask drive motor. Operation of the cask mechanisms is by means of a control pendant at the end of an 8-ft cable.

Figure 9-9 illustrates the FHC positioned at the reactor for a fuel-handling operation. Because of its size and weight, the cask is attached to the 15-ton overhead crane during the entire sequence of operations. Figure 9-10 illustrates the fuel transfer arrangement, showing the relationship of the worker to the various elements.

At the start of fuel-assembly removal, the base of the cask is positioned above the desired opening in the radial slot of the rotating shield plug (see Chapter 4). An indexed carriage positions the cask, and the operator then uses an external hand-crank mechanism to rotate the cylindrical shroud, aligning the square spreading tube with the desired fuel position in the core. The square tube is driven down into the core to spread the upper ends of the adjacent fuel assemblies, and the gripper is then lowered and latched to the fuel assembly. Next, the gripper is raised, withdrawing the fuel assembly, after which the square tube is drawn into the cask and the bottom drawer closed. The gripper cable drive includes a slip clutch which limits the force applied to an element while it is being raised into the cask. Figure 9-11 shows the sequence of activities performed by the FHC.

The sequence for inserting a fuel assembly into the reactor is the same, except that the square tube is withdrawn before the gripper. At the start of the unlatching action the gripper cable is prevented from rising by a 3-s time-delay relay. During this time the square tube rises to permit small triangular shaped cleats (mounted in each corner of the square tube) to engage with dogs on the gripper jaws and thereby unlatch from the assembly end fitting. The cleats also retain the gripper within the confines of the square tube. Operations at the fuel-storage areas are essentially identical to those at the reactor.

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Figure 9-9. Fuel-handling cask design.

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Figure 9-10. Fuel transfer arrangement.

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Figure 9-11. Sequence of fuel assembly removal.

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9.12.1.3 Safety Evaluation. Cask-drop accidents are considered in evaluating materials-handling safety. Because the available experiment-handling casks are vulnerable to such accidents, a standard practice is to limit the height to which a cask can be raised during movements within the Reactor Building. For safety analysis purposes, it is assumed that if a cask were dropped during one of the high lifts, the cask would be breached and the contained experiment equipment damaged.

The postulated cask-drop accident could result from the effects of severe natural phenomena such as earthquakes or high winds. However, the likelihood of severe natural phenomena occurring during critical cask-handling operations is unlikely. The postulated cask-drop accident could also result from system failure or operator error. This type of cask-drop accident is more likely than an accident resulting from the effects of severe natural phenomena and will be minimized by administrative controls and operator training. The radiological consequences of potential drops are addressed in Chapter 15.

Shielding reduces radiation levels of a loaded cask to safe levels for operators working next to the FHC. The weight (up to 115 lb) of the Zircaloy fuel assembly design is accommodated by the hoisting mechanism in the cask. The FHC is stored upright on the Reactor Building floor. Structural steel angles cantilevered from a support column provide lateral restraint at the top of the cask.

9.12.1.4 Inspection and Testing. Based on past operation, the cask has proven to be reliable equipment. Preventive maintenance of the cask is conducted at regular intervals as required by the maintenance management system.

9.12.1.5 Limitations and Setpoints. To minimize the potential for impacts on the rotating shield plug during cask carries, administratively controlled restrictions limit transfers of any cask to a height of 2 in. over the rotating shield plug, and the lift of the FHC to a maximum of 9 in., when clearing the locating projections of the cask carriage (see Chapter 4, Section 4.2.2.3.2.3).

9.12.2 TREAT Loop-Handling Cask

9.12.2.1 Design Basis. The TLHC is designated as a NSR-AR SSC in Chapter 3, Table 3-2. The TLHC is a shielded cask that is used for handling calibration assemblies at the TREAT facility, and certain test loop experiments. Its hoisting mechanism is designed to lift the 1,200-lb calibration or test assembly, and the cask itself can be handled by either the 15-ton or the 60-ton crane. Storage for the TLHC is provided within the TREAT facility.

As noted, the experiment-handling casks are shielded. However, because they are not designed to Department of Transportation criteria for transport of radioactive material, they are unlikely to survive intact under all severe accident conditions. Also, the in-cask hoists are not designed to be single-failure-proof. Section 9.4.2.3 and Chapter 15 describe the safety implications of using these casks and hoists.

9.12.2.2 Description. The TLHC is used for handling experiments too large to fit in the FHC. The cask is not designed to survive high drops, nor is the winch that moves experiments into and out of the cask designed single-failure-proof.

At the TREAT facility, the TLHC is used for handling calibration assemblies, and experiment test loops. This cylindrical cask is approximately 13 ft long, weighs 11.5 tons, is 10 in. ID, and 24 in. maximum OD. A combination of cast lead and lead shot provides 6-in. shielding over the entire length of the cask; at the bottom is a manually operated drawer.

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Figure 9-12 is a depiction of the TLHC. An electrical hoist, mounted on the side of the cask, is used in loading and unloading loops from the TLHC. The cable runs up the side, through a running line tensiometer, over a pulley arrangement, and into the center of the cask. The hook at the end of the cable engages the lifting fixtures that are manually attached to the calibration assembly. The TLHC may be transferred with either the 15-ton or the 60-ton crane. To lift the cask, the crane uses a lifting yoke that attaches to two lifting lugs, spaced 180 degrees apart, on the cask.

Figure 9-12. TREAT loop-handling cask.

9.12.2.3 Safety Evaluation. See above at Section 9.12.1 for an evaluation of typical cask safety. The TLHC has been equipped with a hoist system for the handling of a 1,200-lb assembly. This hoist system has a digital readout of the weight being lifted. The tensiometer limits the force applied to the item being lifted. The cask design is compatible with both the 15-ton and 60-ton cranes.

No shielding analysis has been performed for the TLHC containing experiment or calibration assemblies. However, the TLHC does provide adequate shielding for handling these assemblies. This conclusion is supported by experience with test loops containing preirradiated fuel, and with calibration assemblies.

9.12.2.4 Inspection and Testing. Routine corrective and preventative maintenance activities involving nonreactor equipment and systems are performed in accordance with facility maintenance instructions when applicable.

9.12.2.5 Limitations and Setpoints. The TLHC has no limitations and set points other than inherent limitations due to design. The tensiometer set point is adjusted based upon the weight of the item being lifted.

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9.13 References

10 CFR 830, 2001, “Safety Basis Requirements,” Subpart B, Code of Federal Regulations, Office of the Federal Register.

10 CFR 851, 2006, “Worker Safety and Health Programs, Final Rule,” U.S. Department of Energy.

29 CFR 1900-99, 2015, “Occupational Safety and Health Standards,” Code of Federal Regulations.

ACI 318, “Building Code Requirements for Reinforced Concrete,” American Concrete Institute.

ANL, 1980, TREAT Upgrade for Construction of 60-Ton Reactor Building Crane, S3125-1001-AS-01, Argonne National Laboratory.

ANL, 1981, Acceptance Test Procedure for 60/10-Ton Crane, S3125-1002-QP-00, Argonne National Laboratory.

ANL, 1987, TREAT Upgrade (TU) and TU-Related Support Facilities System Design Description for Site and Reactor Building, S3100-0001-OJ-03, Argonne National Laboratory.

ANSI/ANS-8.3-1997, “Criticality Accident Alarm System,” American National Standards Institute/American Nuclear Society, revised 2003.

ASHRAE, 1975, “Energy Conservation in New Building Design,” American Society of Heating, Refrigeration, and Air Conditioning Engineers.

ASME B30.2, 2011, “Overhead and Gantry Cranes, Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist,” American Society of Mechanical Engineers, as amended.

ASTM B88, “Standard Specification for Seamless Copper Water Tube,” American Society of Testing and Materials, as applicable.

AWS D1.1, “Structural Welding Code - Steel,” American Welding Society.

AWS D2.0, 1966, “Welded Highway and Railway Bridges” (superseded).

AWS D14.1, “Specification for Welding of Industrial and Mill Cranes and Other Material Handling Equipment,” American Welding Society.

CMAA-70, 2000, “Specifications for Top Running Bridge & Gantry Type Multiple Girder Electric Overhead Traveling Cranes,” Crane Manufacturers Association of America.

DOE O 420.1C, 2012, “Facility Safety,” Change 1, U.S. Department of Energy.

DOE O 458.1, 2013, “Radiation Protection of the Public and the Environment,” Admin Change 3, U.S. Department of Energy.

DOE O 5480.30, 2010, “Nuclear Reactor Safety Design Criteria,” Change 1, U.S. Department of Energy.

DOE-STD-1088, “Fire Protection for Relocatable Structures,” U.S. Department of Energy.

DOE-STD-3007-2007, “Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities,” U.S. Department of Energy.

ECAR-1610, “Criticality Safety Evaluation for the TREAT Reactor Building,” current rev., Idaho National Laboratory.

HAD-470, 2014, “Transient Reactor Test (TREAT) Facility Fire Hazards Analysis - MFC-720 Complex,” current rev., Idaho National Laboratory.

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HAD-432, Materials and Fuels Complex (MFC) Area Fire Hazards Analysis, current revision.

LRD-18001, “INL Criticality Safety Program Requirements Manual,” current rev., Idaho National Laboratory.

LST-387, “Criticality Safety Controls for TREAT,” current rev., Idaho National Laboratory.

Manual 14C, “Worker Safety and Health Program, 10 CFR 851,” current rev., Idaho National Laboratory.

McVean, R. L., ANL, to J. C. Phillips, ANL, August 2, 1979, “Additional Criticality Studies for TU Fuel Storage in TREAT Reactor Building.”

NFPA 101, “Life Safety Code,” current rev., National Fire Protection Association.

NRC, 1978, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants,” Regulatory Guide 1.70, Rev. 3, U.S. Nuclear Regulatory Commission.

Paul, J. M., ANL, to G. K. Rusch, ANL, August 17, 1982, “TREAT 60-ton Radio-Controlled Crane –Safety and Interference Test for the Reactor Facility.”

PDD-6000, “INL Nuclear and Non-Nuclear Maintenance Management Program,” current rev., Idaho National Laboratory.

PRD-14401, “INL Fire Protection Program Requirements,” current rev., Idaho National Laboratory.

Rieb, M. J., ANL, to R. E. Timm, ANL, September 29, 1981, Memo SSE-1-84.

RR-W-410G, 2010, “Federal Specification Wire Rope and Strand.”

SAR-400, 2015, “INL Standardized Safety Analysis Report,” current rev., Idaho National Laboratory.

Stevens, W. W, ANL to L. J. Harrison, ANL, April 16, 1984, Intra-Laboratory Memo.

Yang, S., and S. K. Bhattacharyya, ANL, to D. C. Wade, ANL, January 20, 1982, “Comparative Radiation Doses from Fission Products in TREAT Upgrade and TREAT Fuel Assemblies,” Memo S3412-2-A-424.

Yang, S., ANL, to S. K. Bhattacharyya, ANL, June 15, 1982, “Radiation Doses from Activation Component of TU Fuel Assembly Top Fitting,” Memo S3412-2-A-472.

Yang, S., ANL, to S. K. Bhattacharyya, ANL, August 13, 1982, “TUSHIELD Calculations for 37-pin Test Fuel in Test Train,” Memo S3412-2-A-553.

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INTENTIONALLY BLANK

SAR-420 Chapter 9

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