SAR-420, Chapter 1

Form 412.09 (Rev. 10)

Idaho National Laboratory

CHAPTER 1 – INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY

– TREAT FACILITY FSAR

Identifier: Revision: Effective Date:

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CHAPTER 1

INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY

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

1. INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY.......................................... 1-5

1.1 Introduction ........................................................................................................................... 1-5

1.2 General Plant Description ..................................................................................................... 1-5

1.2.1 Purpose of the TREAT Facility ...................................................................... 1-5

1.2.2 Summary Design Description of the TREAT Facility .................................... 1-6

1.2.3 Experiments Performed in the TREAT Facility ........................................... 1-21

1.3 Summary of Safety Analysis for the TREAT Facility ........................................................ 1-22

1.4 Detailed Information and Drawings .................................................................................... 1-30

1.5 References ........................................................................................................................... 1-30

FIGURES

Figure 1-1. Schematic of the TREAT reactor. ......................................................................................... 1-11

Figure 1-2. Diagram of principal reactor core interfaces and paths of information flow. ....................... 1-12

Figure 1-3. TREAT reactor cross section. ............................................................................................... 1-13

Figure 1-4. Standard Zircaloy-clad TREAT fuel assembly. .................................................................... 1-14

Figure 1-5. TREAT reactor core. ............................................................................................................. 1-17

TABLES

Table 1-1. Typical operating parameters of the TREAT reactor. .............................................................. 1-7

Table 1-2. TREAT accident radiological consequence summary. ........................................................... 1-23

Table 1-3. Summary list of key parameter and analysis assumption SSCs. ............................................ 1-26

Table 1-4. Summary list of key parameter and analysis assumption TS controls. .................................. 1-26

Table 1-5. Summary list of SSCs and TS controls derived from the accident analysis. .......................... 1-27

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ACRONYMS/ABBREVIATIONS ANL Argonne National Laboratory AR nonsafety-related with augmented requirements ARCS automatic reactor control system

BNL Brookhaven National Laboratory

CF Clad Failure (accident) CFR Code of Federal Regulations

DBA design-basis accident DOE Department of Energy

EAB exclusion area boundary EH Experiment-Handling (accident)

F/CS filtration/cooling system FH Reactor-Fuel-Assembly Handling (accident) FSAR Final Safety Analysis Report

HEPA high-efficiency particulate air

INL Idaho National Laboratory

LMFBR liquid-metal fast breeder reactor LCO(s) limiting condition(s) for operation LCS limiting control setting LWR light-water reactor

MFC Materials and Fuels Complex MHA maximum hypothetical accident

NRC Nuclear Regulatory Commission NSR nonsafety-related

RG Regulatory Guide RIA reactivity insertion accident RTS reactor trip system

SI System Impact (accident) SL safety limit SR safety-related SSC structure, system, and component

TED total effective dose TF TREAT Facility Fire (accident) TREAT Transient Reactor Test (TREAT) facility TS technical specifications

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INTENTIONALLY BLANK

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1. INTRODUCTION AND GENERAL DESCRIPTION OF FACILITY

1.1 Introduction

This final safety analysis report (FSAR) supersedes S3942-0001-YT, “TREAT Final Safety Analysis Report (FSAR)” (INL 2014). Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants,” (NRC 1978) was used as a guide for the Transient Reactor Test (TREAT) facility FSAR format and content. RG 1.70 is designated in 10 CFR 830 as an acceptable format and content guide for U.S. Department of Energy (DOE) reactor safety analysis reports. In addition to RG 1.70, NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” (NRC 1987) was used as a guide for TREAT facility FSAR content.

Because of the significant differences between the TREAT facility and NRC-licensed commercial power reactors, the design and operating requirements for the TREAT facility are not necessarily the same as those that apply to commercial reactors. The approach to demonstrating that the TREAT facility can be operated safely is based on the guidelines presented in Brookhaven National Laboratory (BNL) 50831-V, Design Guide for Category V Reactors, Transient Reactors, (BNL 1979a) and BNL 50831-VI, Design Guide for Category VI Reactors, Air Cooled Graphite Reactors (BNL 1979b). The BNL design guides provide standards, guides, and codes for DOE-owned reactors such as the TREAT facility, which are comparable to those applied to similar reactors licensed by the NRC.

In addition, the following guides were consulted on an FSAR chapter-by-chapter basis to tailor the TREAT facility FSAR content commensurate with the design, systems, operating requirements, and safety analyses typical for a nonpower transient test/research reactor such as TREAT:

• NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors,” (NRC 1996)

• ANSI/ANS-15.21, “Format and Content for Safety Analysis Reports for Research Reactors,” (ANSI/ANS 1996).

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

1.2 General Plant Description

1.2.1 Purpose of the TREAT Facility

The TREAT facility is designed to perform transient testing of nuclear fuels and materials with the flexibility of the facility considered in developing the facility safety envelope (MacFarlane, Freund, and Boland 1958). In addition to the power response of the TREAT reactor, the flexibility in the core configuration allows the reactor to accommodate a variety of experiments.

As discussed in Chapter 9, Buildings and Auxiliary Systems, the primary buildings making up the TREAT facility include the Reactor Building (MFC-720) and the Control Building (MFC-724). The TREAT facility is able to subject experimental specimens to various transient nuclear conditions of interest.

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Transient testing involves placing fuel or material, either previously irradiated or unirradiated, contained in a test assembly (see Chapter 10, Experimental Facilities and Utilization, Section 10.1.2) into the core and subjecting it to severe conditions. These conditions include transient overpower and loss-of-cooling that might result from malfunctions in either the reactivity control systems or core cooling systems of liquid-metal fast breeder reactors (LMFBRs) and light-water reactors (LWRs). During testing, the test assembly is monitored using specialized instruments. After the transient experiment is completed, the fuel or material is examined to determine the effects of the radiation. Fuel meltdowns, metal-water reactions, thermal interaction between overheated fuel and coolant, and the transient behavior of reactor fuel for a variety of systems can be studied.

Since the TREAT facility provides thermal neutrons over relatively large sample volumes, it can be used in other studies including neutron radiography of experimental capsules and elements that will be or that have been run in the TREAT facility or in other reactors. Neutron radiography provides an image of the final placement of the test item before irradiation. This capability can also provide an image of the posttransient configuration of the test items.

In addition, the TREAT gamma scan facility is used to remotely locate radioactive isotopes within an experiment. Gamma scanning of reactor fuel elements and other experiments both before and after their irradiation in reactors can provide information on the movement of radioisotopes as a result of the irradiation and also detect the presence of irradiated fissionable material.

1.2.2 Summary Design Description of the TREAT Facility

Table 1-1 summarizes the principal performance characteristics and parameters of the TREAT reactor. The principal features of the TREAT reactor are shown in Figure 1-1. Figure 1-2 shows the principal reactor core interfaces and paths of information flow.

The TREAT reactor, a graphite-moderated thermal reactor, is designed primarily for operation in the transient or pulsed mode for destructive testing of prototypic fast reactor fuel pins. The TREAT reactor is designed conservatively to produce a pulse with a thermal neutron fluence of at least 3.5 × 1015 neutrons/cm2 averaged over the core.

As discussed in Chapter 4, Reactor, Section 4.2.1.1, the TREAT core peak temperature limit is 600°C, corresponding to approximately 2,100 MJ in a typical reactor core. The 600°C core temperature limit permits the use of a variety of core configurations, and a variety of experiments that can be performed. The factor that limits transient operations is the peak core temperature, due to the increasingly rapid oxidation rate of Zircaloy clad core fuel at higher temperatures in air.

The TREAT core consists of an array of individually removable fuel assemblies. Surrounding the assemblies is a permanent graphite reflector. Enclosing the sides of the reflector is a high-density concrete shield. Removable concrete shield blocks and a rotating steel shield plug provide access for fuel handling and for inserting experiments. The rotating shield plug also provides for core position indexing and work space above the core during fuel handling.

Control rods provide reactivity control. They extend through the lower concrete shield and attach to the control rod drives located in the subpile room directly beneath the reactor. Control rods are driven vertically upward to add positive reactivity and are driven downward to insert negative reactivity.

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Viewing slots may be installed in the reactor core to provide visual or instrument access to the central experiment location from three of the four faces of the reactor (the north, south, and west faces). Access is obtained by moving adjustable shield blocks in the radial shield, lifting reflector blocks in the permanent reflector, and installing slotted fuel assemblies in the core itself. Test fuel motion diagnostic systems (fast neutron hodoscopes) can be installed in the two access slots leading to the center of the core from the north and south faces. Photographic or video recording systems can be installed from the south face or north face if the hodoscope is removed. In addition, a neutron radiography facility is located in the west face of the reactor (see Figure 1-1).

Table 1-1. Typical operating parameters of the TREAT reactor.

Item Parameter

General

Overall dimensions of fuel assemblies 4 in. nominal square, 8.94 ft long, 5/8-in. chamfered corners

Active fuel height 4 ft

Top and bottom reflectors, each Graphite, 2 ft long

Average cross section of core 5 ft, 8-in.-dia pseudo-circle with 210 fuel assemblies

Maximum cross section of core (radial) 6 ft, 4 in. square if 361 fuel and reflector assemblies are inserted

Peripheral permanent reflector Graphite, min. of 2 ft thickness 235U loading in typical core 7.45 to 11.63 kg 235U enrichment in fuel 93.24%

Carbon to 235U atomic ratio 10,000:1

Chemical form of U UO2

UO2 content in fuel 0.248 mass %

Fuel section components Six fuel blocks, 3.8 in. square, 8 in. long

Fuel cladding 0.025-in.-thick Zircaloy-3

Cladding in the axial reflector region 0.050-in.-thick type 1100 aluminum

Cross section of top and bottom reflectors 3.8 in. nominal square with chamfered corners

Fuel-reflector cans joint Riveted to ends of fuel section

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Item Parameter

Assemblies in the core

Shapes Squares with chamfered corners in most of the core, and rectangles with chamfered corners in the central zone to accommodate test loops

Cross section nominal dimensions of assemblies

Standard/Full 4 in. square

Control/Shutdown 4 in. square

Compensation/Shutdown 4 in. square

Transient 4 in. square

Instrumented 4 in. square

Slotted/Access hole 4 in. square

Aluminum clad reflector (dummy) 4 in. square

Zircaloy-clad reflector (dummy) 4 in. square

Half dummy 4 × 2-in. rectangle

Access-hole half dummy 4 × 2-in. rectangle

Compensation/shutdown rod and drive

Total number of rods 4

Number of drives 4

Number of rods per drive 1

Poison material B4C

Geometry:

Poison section 1.75-in.-dia, 60-in.-long, 1/8-in.-thick carbon steel tube

Top follower 1.75-in.-dia, 58.5-in.-long, 1/8-in.-thick Zircaloy-2 tube filled with graphite

Bottom 2 followers 1.75-in.-dia, 75.8-in.-long, 1/8-in.-thick carbon steel tube filled with graphite

Type of latch Hydraulic

Scram power Pneumatic and gravity

Drive power Electric

Typical rod worth, total of 4 6.9% nominally (Value for Core Load 1469. Measured for each core configuration)

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Item Parameter

Control/shutdown rod and drive


Number of drives 4


Poison material B4C

Geometry:


Top follower 1.75-in.-dia, 60-in.-long, 1/8-in.-thick Zircaloy-2 tube filled with graphite

Bottom 2 followers 1.75-in.-dia, 84-in.-long, 1/8-in.-thick carbon steel tube filled with graphite

Type of latch Magnetic

Scram power Pneumatic and gravity

Drive power Electric

Typical rod worth, total of 8 8.8% nominally (Value for Core Load 1469. Measured for each core configuration)

Transient rod and drive


Number of drives 4


Poison material B4C

Geometry:


Top follower 1.75- in.–dia, 60-in.-long, 1/8-in.-thick Zircaloy-2 tube filled with graphite

Bottom 2 followers 1.75-in.-dia, 78.37-in.-long, 1/8-in.-thick, carbon steel tube filled with graphite

Type of latch Not applicable

Scram power Hydraulic

Drive power Hydraulic

Calculated rod worth, total 8.5% nominally (Value for Core Load 1469. Measured for each core configuration)

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Item Parameter

Operating parameters

Shaped transient

Energy <2,900 MJ (core dependent)

Peak power <10,000 MW (core dependent)

Peak fuel temperature ≤600°C

Peak clad temperature ≤600°C

Minimum reactor period 0.1 sec

Cooling time to 37°C approximately 5 hours (core dependent)

Natural burst transient (self-limited transient mode)

Energy <2,100 MJ (core dependent)

Peak power <18,000 MW (core dependent)

Peak fuel temperature ≤600°C

Peak clad temperature ≤600°C

Minimum reactor period 0.023 sec

Cooling time to 37°C approximately 5 hours (core dependent)

Steady-state

Peak power <120 kW

Peak fuel temperature approximately 270°C (with 180 fuel assemblies)

Minimum reactor period 5 sec

Coolant Air

Physics parameters

Prompt neutron generation time approximately 9.0 × 10-4 sec

Effective delayed neutron fraction approximately 0.0072

Total temperature coefficient of reactivity approximately -1.7 × 10-4 (delta k/k2)/°C

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Figure 1-1. Schematic of the TREAT reactor.

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Figure 1-2. Diagram of principal reactor core interfaces and paths of information flow.

1.2.2.1 Reactor Core and Fuel Assemblies. Figure 1-3 shows the TREAT reactor cross section. It consists of a 19 × 19 square array of fuel and reflector assemblies. Surrounding the array is a permanent reflector 2 ft thick. Experiments customarily replace one or two of the fuel assemblies at the center of the core. However, the TREAT core can be reconfigured to accommodate a variety of other experiment geometries.

Chapter 4, Section 4.2.1 provides a detailed description of the TREAT fuel assemblies. Figure 1-4 shows the standard TREAT fuel assembly. The assembly consists of upper and lower graphite reflector sections and a central section of uranium oxide-bearing graphite fuel. The fuel section is 4 ft long and contains six fuel blocks, each 8 in. long. The reactor fuel blocks consist of small particles (mean size, 10 microns) of fully enriched 235U dispersed in a graphite matrix. The carbon-to-uranium atom ratio is nominally 10,000:1.

To prevent oxidation of the graphite at operating temperatures and to contain fission products, the graphite-urania blocks are sealed within evacuated Zircaloy-3 cans. Urania is a term used in the original TREAT safety analysis and design documentation and retained in this FSAR, and is defined as any of several impure mixtures of uranium oxides. The upper and lower graphite reflector sections are each 2 ft long and are clad in cans fabricated of 0.05-in.-thick Type 1100 aluminum, which are riveted to the ends of the fuel section. The complete fuel assembly is slightly less than 9 ft long, has a nominal cross section of 3.96 × 3.96 in., and weighs 95 lb. The 1-in.-diameter aluminum alignment and support pins go into matching openings in the grid plate.

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Figure 1-3. TREAT reactor cross section.

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Figure 1-4. Standard Zircaloy-clad TREAT fuel assembly.

(Note: Section A-A illustrates cross-section of two adjacent fuel assemblies to show distances between two fuel cans and between fuel blocks and can)

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The chamfers at the corners of adjacent assemblies combine to form coolant passages, 5/8 in. square, through the assembled core. Air flows downward through these coolant passages to carry away the heat generated in operation. As discussed in Chapter 4, Section 4.2.1.1, the programmatic fuel temperature limit for these assemblies is 600°C.

In addition to this basic fuel assembly, there are a number of special TREAT fuel assemblies, including:

• Control rod fuel assemblies that each contain a Zircaloy tube and a Graphitar (U.S. Graphite Inc.’s trade name for a family of carbon-graphite (mechanical carbon) products) bearing to guide the control rods

• Slotted reflector and fuel assemblies that can be placed in line to provide a viewing slot into the center of the core

• Instrumented fuel assemblies that contain thermocouples

• Reflector (also called dummy assemblies in Chapter 4) assemblies that consist of graphite blocks in unsealed cans with either aluminum or Zircaloy-clad over the central 4 ft of the graphite

• Instrumented reflector (dummy) assemblies.

These assemblies are color-coded with anodized aluminum round gripping fixtures. Color coding enables easy identification of the fuel assemblies during fuel-handling operations. Each assembly also has a unique serial number engraved into the gripping fixture.

1.2.2.2 Reactor Shield. The sides of the TREAT reactor are enclosed by a high-density concrete shield 5 ft thick (see Figure 1-3). The bottom concrete shielding forms the ceiling of the basement subpile room. The TREAT reactor has 32 control rod penetrations in the lower shield and the core support grid plate to provide for flexibility in control rod locations. This location layout, shown in Figure 1-3, permits a total of 16 control rod drives (one or two rods per drive) to be located in the basement subpile room. Shielding above the reactor consists of removable concrete blocks 3 ft thick and a steel rotating shield plug 1 ft thick. This plug is supported on a large ball-bearing assembly anchored in the concrete shield. A radial slot in the plug, which is customarily filled by movable cast boron-steel blocks during reactor operation, provides access to any core location for fuel or possible experiment handling.

As indicated in Figure 1-3, the east face of the reactor shield houses a graphite thermal column 5 ft square and 5 ft thick, which is shielded by a movable concrete door. A design criterion for the reactor is that any experiment at the center of the core should be visible, preferably along at least two lines divergent by 90 degrees. Accordingly, horizontal access slots are provided in the other three faces of the reactor shield. Two of the three openings in the concrete shield can be converted from a shielding configuration to a viewing configuration (Figure 1-1). The convertible openings are designed to provide an offset opening 4 in. wide × 24 in. high, for taking photographs with equipment external to the shield, or to provide an opening in the shielding 2 ft square, stepping up to 3 ft square, for installation of instrumentation. Converting each of these slots to the viewing configuration requires the use of a screw jack arrangement to move a pair of concrete shielding blocks. The third opening, which is not convertible, provides an offset access slot 4 in. wide × 24 in. high.

Provisions for installing loop thimbles include a removable shield plug, made of depleted uranium, in the concrete shield beneath the core center line. With that plug removed, the reactor can accommodate experiments that extend as much as 41 in. below the fuel-element-support grid plate and up to 5 in. in

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diameter. The design of the concrete reactor shielding allows personnel access around the reactor during steady-state (<120-kW) operations.

See Chapter 4, Section 4.2.2 for a detailed description of the TREAT reactor structure.

1.2.2.3 Experiment Support Facilities in the TREAT Facility. The TREAT reactor can be equipped with two fast-neutron hodoscopes and a radiographic facility. The hodoscope for the south access slot has a viewing area 2.24 in. wide × 20.08 in. high; the north access slot hodoscope has a viewing area 2.60 in. wide × 47.24 in. high. The neutron radiography facility is located in the shield access hole on the west face of the reactor.

Gamma scanning provides a useful tool for remotely determining the location of radioactive isotopes within an experiment. Gamma scanning of reactor fuel elements and other experiments, both before and after their irradiation in reactors can provide information on the movement of radioisotopes as a result of the irradiation. Gamma scanning can also detect the presence of irradiated fissionable material.

Two lead-shielded handling casks and the overhead bridge crane are used for the experiment-handling operations at the top of the reactor. Gaining access to the core for these operations involves removing the large top shield blocks and working from the rotating shield plug. Additional experiment support facilities, e.g., power cables, instrument lead wires, gas and/or liquid lines, are installed as needed by individual experiment requirements.

See Chapter 10, Section 10.1.2 for a description of the TREAT facility experiment support equipment and facilities.

1.2.2.4 Reactivity Control System. Three sets of control rods are used:

• Compensation/shutdown rods, located as (C) as shown in Figure 1-5, compensate for the reactivity worth of an experiment, and provide a fast-acting shutdown capability. The drives for these rods have fast-acting hydraulic latches and a compressible-fluid shock absorber for scram-impact deceleration force reduction

• Control/shutdown rods, located as (S) as shown in Figure 1-5, are the primary means of reactivity control during reactor startup and steady-state operation. The drives for these rods have fast-acting magnetic latches and a compressible-fluid shock absorber for scram-impact deceleration force reduction

• Transient rods, located as (T) as shown in Figure 1-5, are used for high-speed transient control. The rods are mechanically fastened to the drives. The drives are manually controlled in low-speed during manual reactor operation and computer controlled in high-speed during transient reactor operation, and hydraulically driven in both directions including downward during reactor scrams.

See Chapter 4, Section 4.2.3 and Chapter 7, Instrumentation and Control, Section 7.3 for a detailed description of the TREAT reactivity control systems. As shown in Figure 1-5, the control rods in the TREAT core are located in two concentric rings. Four compensation/shutdown rods are located in the inner ring. Eight transient rods and eight control/shutdown rods are located in the outer ring. The eight transient rods are yoked in pairs to four transient rod drives. Similarly, the eight control/shutdown rods are yoked in pairs to four control/shutdown drives (indicated by the ovals in Figure 1-5). Only one compensation/shutdown rod is driven by each compensation/shutdown drive.

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Figure 1-5. TREAT reactor core.

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The rod drives are located below the core in the subpile room (Figure 1-1) and can drive the control rod poison sections down into or up out of the core. The control rods move in a vertical direction through guide tubes. These tubes are located below the grid plate and within special fuel assemblies and have Graphitar bearings for guiding the rods. The rods are driven upward to remove their poison elements from the core; downward motion causes poison insertion.

Compensation/Shutdown Rods and Drives: Each of the four compensation/shutdown rod drives in the inner ring moves a single control rod. The total reactivity control worth of these four rods varies somewhat with core loading. As shown in Table 1-1, the nominal rod worth is 6.9% (Core load 1469) and is core dependent.1 The drives are lead-screw positioned. A pneumatically driven piston is incorporated to scram the rods rapidly downward, and a special high-speed hydraulic latch-release mechanism is used for fast response. The drives have a 58-in. stroke. The compensation/shutdown rods are used as one of three rod shutdown systems, and are also used to compensate for the reactivity worth of the test vehicle.

Control/Shutdown Rod Drives: Each of the four control/shutdown drives located in the outer ring is attached to a pair of control rods. The rod drives are positioned by means of lead-screw drives. They have a pneumatic cylinder-driven scram assist in the downward direction. The drives have a 58-in. stroke.

At the beginning of transient operation, the control/shutdown rods are normally banked at partially withdrawn positions such that the reactivity available in the transient rods is that required for the particular transient being run. From this position, the rods can be scrammed at high speed. The operator customarily uses the control/shutdown rods for control during steady-state reactor operation. The total reactivity control worth of the eight control/shutdown rods varies somewhat with core loading. As shown in Table 1-1, the nominal worth is 8.8% (Core load 1469) and is core dependent (see footnote 1).

Transient Rods and Drives: Each of the four transient rod drives in the outer ring is attached to two rods. These drives are hydraulically actuated and have a 40-in. stroke.

Separate servovalves control each of the transient drive hydraulic cylinders. Each servovalve receives input from a dedicated controller that, in turn, receives commands from the reactor control system. In the steady-state mode, the hydraulic fluid is supplied by a pump at such a rate that the rod drives can move only slowly, and control is under manual switch command from the control room. When the transient mode is enabled by the key switch, the rods can be moved (reactivity addition and removal) at high speed. In this mode with hydraulic fluid supplied from a precharged accumulator, the control signals are derived from the digital computer-based automatic reactor control system (ARCS). As shown in Table 1-1, the nominal rod group worth is 8.5% (Core load 1469) and is core dependent (see footnote 1).

1.2.2.5 Reactor Control System. The following computer and control systems are supplied (see Chapter 7, Section 7.6 for detailed descriptions):

• A manual reactor control system

• A transient control computer and backup monitor computer

• Ancillary computers to provide system interfaces to the Reactivity Control System, Reactor Trip System, and Experiment Support system.

1 The nominal control rod worths are core dependent to a limited degree due to the core configuration (e.g., size, slotted assemblies, and by the size, shape, location, and constituents of the experiment).

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Manual Reactor Control System: The control/shutdown rods, the compensation/shutdown rods, and the transient rods are controlled manually during reactor startup and steady-state operation, and during the rod-positioning phase before a transient is initiated. In the manual mode, the control computers and the capability for rapid reactivity insertion by the transient rod drives are locked out.

The capability for rapid negative reactivity insertion (i.e., a reactor scram) is fully functional during the manual reactor operation mode. See Chapter 7, Section 7.6.1.1 for a detailed description of the manual reactor control system.

Automatic Reactor Control System: The ARCS is computer-based and is the second system identified above. A control computer operating on a fixed sampling interval is used to implement a prescribed transient profile. A monitor computer compares the transient profile implemented by the control computer to a set of self-contained trip boundaries and intercedes with a scram command if the transient proceeds beyond the prescribed limits.

The ARCS controls the transient control rods in one of two modes to produce the desired transient profile:

• Position demand, in which the transient rod drives move only according to a preprogrammed control rod position demand

• Closed-loop, in which the transient rod drives move in response to feedback signals derived from position demand as well as reactor period or reactor power, to execute a prescribed reactor power-versus-time profile.

Operation in either mode requires activating a key switch. The transient rods always move in unison during a transient to maintain maximum core radial power distribution flatness.

The ARCS does not control movement of either the compensation/shutdown or control/shutdown rod systems. When the reactor is under automatic control, the compensation/shutdown rods are banked fully out, and the control/shutdown rods are banked to establish initially a critical reactor configuration.

The TREAT reactor is operated in a self-limited mode. In self-limited operation, the maximum amount of reactivity available in the transient rods is limited such that, if the available reactivity were rapidly added (as a step), the resulting natural burst would be terminated by the strong negative temperature coefficient and would yield a temperature below that of the fuel-assembly damage threshold. See Chapter 7, Section 7.6.1.2 for a detailed description of the ARCS.

1.2.2.6 Reactor Trip System. The reactor trip system (RTS) is a three-channel, hard-wired analog system arranged for one-out-of-two trip circuit logic for steady-state operations and one-out-of-three trip circuit logic for transients. It operates independently of and is electrically isolated from the reactor control system. These systems are described in detail in Chapter 7, Section 7.2. The RTS components include:

• A three-channel, hard-wired RTS that includes energy, power and period trips

• A variable-trip-setting RTS concept for reactor power level and integrated power level trip circuits. Two groups of trip settings are used: one independent of the type of transient being run, the other tailored to the specific transient

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• A two-channel seismic disturbance trip system. This trip system acts directly on the rod latch current power and does not use the RTS logic circuitry; therefore, the RTS logic circuitry need not be seismically qualified

• A two-channel RTS for steady-state without an energy trip.

The design of the RTS accommodates four modes of operation as follows: (1) simulated full transient mode for the purpose of verifying proper ARCS control functions with the simulator computer substituting for the reactor, and partial simulated mode with the control/shutdown and compensation/shutdown rods incapable of withdrawal to guarantee reactor shutdown, (2) test mode for functional testing and calibration of the RTS subsystems: this mode shall also guarantee reactor shutdown as for the simulated transient mode, (3) steady-state mode for reactor startup, low-power steady-state irradiations, and pretransient staging operations up through 120 kW, and (4) transient enable mode for the generation of prescribed reactor transients, at which time the steady-state nuclear channel trips are bypassed and transient-only trips enabled.

No specific requirements for operability or reliability of RTS instrumentation are identified in the accident analysis. However, to provide additional levels of protection for the analyzed scenarios, RTS trip settings are selected to limit core temperatures to prevent exceeding the fuel temperature limit of 820°C for even the most unfavorable postulated time sequence of reactivity insertion. Reactor period, power level, temperature, and energy release are monitored in the transient mode. All of these parameters, except energy release, are also monitored in the steady-state mode. Two sets of trip circuits are used with each of the three trip channels. The settings of one set, called “transient-independent settings,” envelop all transients and do not change for different experiments. The other more restrictive trips are tailored to accommodate the needs of each experiment; they are called “transient-dependent settings” and will be either more restrictive or the same as the transient independent trip settings. Requirements to ensure RTS operability are defined in Technical Specifications (TS)-420.

1.2.2.7 Filtration/Cooling System. The filtration/cooling system (F/CS) components include:

• Two blowers

• A 140-ft coolant exhaust stack

• A high-efficiency particulate air (HEPA) filter assembly with provision for testing and controlled bagout.

The F/CS system is described in detail in Chapter 5, Filtration/Cooling System. The TREAT facility is a test reactor designed primarily to produce short-duration, high-power bursts. Typically, high-power operation persists for only a few tens of seconds at most, and as discussed in Chapter 4, Section 4.2.1.2, the high heat-absorbing capability of the carbon and graphite moderator provides an effective heat sink for transient-generated heat, without dependence on forced air cooling during or after the transient.

The cooling system is also used to remove the generated reactor heat during the various types of steady-state operations. Because of the large heat capacity of the reactor fuel, cooling is not usually needed in the steady-state power mode. As discussed in Chapter 5, Section 5.6.1, a steady-state power level of 120 kW could be maintained for several hours with no cooling before the cladding temperature limit of 600°C is reached.

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Therefore, the main reason for providing the cooling function is to reduce the amount of time needed to cool the core after a power burst. The selection of system cool-down time takes into account effects of fuel cool-down time, oxidation damage, and temperature-related stress, as well as experiment operations efficiency.

The F/CS is also designed to entrain and remove radioactive material by providing subatmospheric pressure in the reactor cavity. Two blowers operating in parallel, located downstream of the reactor, pull coolant air through the system. The flow path starts at inlet prefilters on the top of the reactor shield (see Figure 1-1), and ends at an exhaust stack. Air flows down through the reactor core and out the ducts from a plenum beneath the core. It then flows to a HEPA filter system. Finally, the air passes through the blowers and out through the exhaust stack.

To ensure the reliability of this system, the blowers can be powered from independent power sources. One power source is the normal site electric power; the other is a local diesel generator. See Chapter 5, Section 5.5 for a detailed discussion of the operation requirements of the blowers.

1.2.3 Experiments Performed in the TREAT Facility

Since the design of the facility is such that a variety of experiment hardware configurations and fuel types can be accommodated, a rather broad spectrum of experiments in addition to those cited above can be performed in the TREAT facility.

Historic LMFBR and other experiment designs are summarized in Woolstenhulme and Wiest (2013). Future transient testing capability will be provided for all nuclear fuel types, including fuels for LWRs, high-temperature gas reactors, and fast reactors (Woolstenhulme and Wiest 2013).

The analyses in TEV-1832 (2013) suggest that a seven-pin MOx fast reactor fuel experiment will provide a bounding inventory for accident analysis. Therefore, analyses using a seven-pin Mark-III loop with sodium are used in this FSAR to develop an enveloping set of design and safety analysis requirements in Chapter 10, Section 10.2 that will allow maximum flexibility for future transient testing needs.

See Chapter 10, Section 10.1.2 for a description of the TREAT facility experiment support equipment and facilities.

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1.3 Summary of Safety Analysis for the TREAT Facility

Accidents analyzed in Chapter 15, Accident Analyses, were selected for analysis based on an evaluation of the various operations to be performed at the TREAT facility, including the experiment operations described in Chapter 10, Section 10.1. The limiting events selected for further analysis include the following:

• Reactivity Insertion Accidents (TREAT Design Basis Accident)

• Experiment-Handling Accidents (EH)

• Reactor-Fuel-Assembly Handling Accidents (FH)

• Inadvertent Nuclear Criticality

• System Impact Accidents (SI)

• Reactor-Fuel-Assembly Clad Failure Accidents (CF)

• Loss of Cooling

• Experiment Malfunctions (Experiment Design Basis Accident)

• TREAT Facility Fires (TF)

• Natural Phenomenon Events

• Maximum Hypothetical Accident (TREAT MHA).

Table 1-2 summarizes the results for those accidents from the above list that were analyzed for radiological dose consequences, specifically the EH, FH, SI, CF, and TF accidents, and the TREAT MHA. The consequences are in terms of total effective dose (TED) in rem and in percentage of the consequence guidelines from Chapter 15, Table 15-2. All accident consequences are well within the applicable limits.

The limiting credible accident scenario for determining consequences to the public at the exclusion area boundary (EAB) is the Plant Condition 3 transportation fire accident TF-3; however, the consequences at the EAB are <1% of the dose limits for a Plant Condition 3 event for the public of 5 rem.

The limiting credible accident scenario for determining consequences to the collocated worker at 100 m and 300 m is the Plant Condition 2 clad failure accident CF-1; however, the consequences are less than the dose limits for a Plant Condition 2 event for the onsite worker of 5 rem.

The limiting credible accident scenario for determining consequences to the collocated worker at the Materials and Fuels Complex (MFC) is also the Plant Condition 2 clad failure accident CF-1; however, the consequences at the receptor at MFC are also <1% of the dose limits for a Plant Condition 2 event for the onsite worker of 5 rem.

The limiting credible accident scenario for determining consequences to the worker in the TREAT control room during a transient test (Reactor Building unoccupied) is also a Plant Condition 2 clad failure accident CF-1. The consequences at the receptor at TREAT control room, however, are <1% of the Plant Condition 2 dose limits for the onsite worker of 5 rem.

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Table 1-2. TREAT accident radiological consequence summary.

Accidenta Plant

Condition

Collocated Worker

(100 m)b, c

Collocated Worker

(300 m)b, c

TREAT Control Room

(770 m)b, c MFC

(1,000 m)b, c EAB

(6,000 m)b, c

LPZ (Mud Lake) (32 km)b, c

Idaho Falls (48 km)b, c

EH-1 4 6.77 × 10-1 (<1%)

1.33 × 10-1 (<1%)

3.04 × 10-2 (<1%)

2.01 × 10-2 (<1%)

1.95 × 10-2 (<1%)

2.88 × 10-3 (<1%)

1.83 × 10-3 (<1%)

EH-2 4 6.62 × 10-1 (<1%)

1.30 × 10-1 (<1%)

2.97 × 10-2 (<1%)

1.96 × 10-2 (<1%)

1.76 × 10-2 (<1%)

2.58 × 10-3 (<1%)

1.63 × 10-3 (<1%)

EH-3d 3 6.77 × 10-1 (2.7%)

1.33 × 10-1 (<1%)

3.04 × 10-2 (<1%)

2.01 × 10-2 (<1%)

1.95 × 10-2 (<1%)

2.88 × 10-3 (<1%)

1.83 × 10-3 (<1%)

EH-4d 3 6.62 × 10-1 (2.7%)

1.30 × 10-1 (<1%)

2.97 × 10-2 (<1%)

1.96 × 10-2 (<1%)

1.76 × 10-2 (<1%)

2.58 × 10-3 (<1%)

1.63 × 10-3 (<1%)

FH-1 4 5.44 × 10-1 (<1%)

1.07 × 10-1 (<1%)

2.44 × 10-2 (<1%)

1.61 × 10-2 (<1%)

1.45 × 10-2 (<1%)

2.13 × 10-3 (<1%)

1.34 × 10-3 (<1%)

FH-2e 3 5.44 × 10-1 (2.2%)

1.07 × 10-1 (<1%)

2.44 × 10-2 (<1%)

1.61 × 10-2 (<1%)

1.45 × 10-2 (<1%)

2.13 × 10-3 (<1%)

1.34 × 10-3 (<1%)

SI-8 4 6.29 × 10-3 (2.2%)

1.11 × 10-3 (<1%)

2.02 × 10-4 (<1%)

1.21 × 10-4 (<1%)

4.87 × 10-5 (<1%)

6.26 × 10-6 (<1%)

3.88 × 10-6 (<1%)

CF-1 2 9.0 × 10-1 (18.0%)

1.57 × 10-1 (3.14%)

2.83 × 10-2 (<1%)

1.68 × 10-2 (<1%)

2.93 × 10-3 (<1%)

2.91 × 10-4 (<1%)

1.74 × 10-4 (<1%)

TF-3 3 1.67 × 10-1 (<1%)

3.28 × 10-2 (<1%)

7.46 × 10-3 (<1%)

4.92 × 10-3 (<1%)

3.69 × 10-2 (<1%)

5.89 × 10-3 (<1%)

3.90 × 10-3 (<1%)

TREAT MHA

N/A N/Af N/Af 43.8 (44%)g

27.7 (28%)g

21.4 (86%)g

3.0 (12%)g

2.0 (8%)g

a. The specifics for each accident are given in Chapter 15. b. All doses are TED in rem c. Percentage of applicable limits in parentheses d. Accident consequences same as EH-1 and EH-2, but compared to Plant Condition 3 dose limits. e. Accident consequences same as FH-1, but compared to Plant Condition 3 dose limits. f. Accident occurs during transient operations when Reactor Building is unoccupied and facility workers are relocated to

TREAT control room (MFC-724), therefore, facility and collocated worker consequences are N/A. g. Although not required, hypothesized accidents are compared to the Plant Condition 4 limits.

As discussed in detail in Chapter 3, Design of Structures, Components, Equipment, and Systems, Section 3.2.1, the methodology from ANSI/ANS-58.14-2001 is used to classify TREAT structures, systems, and components (SSCs) as safety-related (SR), nonsafety-related (NSR), or NSR with augmented requirements (NSR-AR).

As demonstrated from the result of the accident analyses summarized in Table 1-2, engineered safety features are not required for the mitigation of the consequences for any analyzed event. Active SR-SSCs are not required for the protection of the public or collocated onsite worker.

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The principal restrictions on TREAT reactor operations are derived from the maximum credible reactivity insertion accident (RIA) (TREAT DBA) in Chapter 15, Section 15.2. The safe shutdown of the TREAT reactor is ensured by the reactor’s physically inherent strong negative temperature coefficient of reactivity, and not on scram action from the RTS. Therefore, due to their importance in providing this function, the fuel assemblies are designated as SR-SSCs.

A three-level administrative approach to reactivity control is established in Chapter 15, Section 15.2.2.1. The first level is to administratively limit core excess reactivity to 8.0%. This restriction on excess reactivity prevents exceeding the TS-420 safety limit (SL) limit of 820°C if all the reactivity administratively allowed in the pretransient rod position in the transient rods is inserted as a step and the core temperature approaches the TS-SL limit of 820°C, followed by a reactor operator error which inserts all the remaining reactivity in the shutdown rods. In addition, the limit on excess reactivity prevents the TS-SL limit of 820°C from being exceeded if this reactivity were added slowly (i.e., the reactor mode selector switch is in steady state, even though the reactor power may be increasing) (see Chapter 16, Derivation of Technical Specifications, Section 16.5.2.1).

The second level of the three-level administrative control approach to reactivity control is based on extrapolating peak core temperature data from temperature-limited transients performed with the core of interest as discussed in Chapter 4, Section 4.3.3.4.5. This extrapolation determines the maximum reactivity permitted from the motion of the transient rods. The reactivity available in the transient rods is limited such that the maximum reactor-fuel temperature would not exceed 820°C if all of the reactivity available in transient rods, plus reactivity feedback effects that could be caused by experiment response to the power transient, were added as a step and the power transient was terminated by the negative temperature coefficient only and assuming no scram action. The reactivity available in the transient rods is manually set by rod axial position and verified before transient operation (see Chapter 16, Section 16.5.2.2).

The third level consists of an administrative control on the transient prescription, such that the TS-420 limiting control setting (LCS) fuel temperature of 600°C will not be exceeded. The transient rod reactivity is limited in the transient prescription to meet the needs of the experimenters and result in a planned temperature less than or equal to 600°C. Ensuring this transient prescription limit is met also ensures that the TS-SL is not exceeded during a transient (see Chapter 16, Section 16.5.2.3).

The analyses in Chapter 15, Section 15.2 show these steps to be effective with no scram action from the RTS. No specific requirements for operability or reliability of RTS instrumentation are identified in the accident analysis, and as such, they are not designated as SR-SSCs. RTS instrumentation is designated as a NSR-AR SSC. No credit is assumed in the accident analysis for the reactor interlock set points, or scram set points on reactor temperature, period, energy, and power. As such, the associated SSCs are not required to be designated as SR-SSCs as a result of the accident analyses.

However, to align TREAT to industry precedent with other test/research reactors licensed by the NRC or operated by DOE, the RTS manual scram and seismic trip subsystems; compensation/shutdown rod and control/shutdown rod systems; and core support, alignment, and concrete structures are designated in Chapter 15 as SR-SSCs to meet the Chapter 3, Section 3.2.1 SR-SSC Criterion 1 to ensure the reactor is shut down and maintained in a safe shutdown condition for the applicable accident scenarios. TREAT SR-SSCs and NSR-AR-SSCs designated in Chapters 15 and 16 (see Table 16-3) are listed in Table 1-5.

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TS-SLs and TS-LCSs are also not required as a result of the accident analyses. As discussed in Chapter 16, Section 16.4, although not required, to align the TREAT facility to industry precedent with other test/research reactors licensed by NRC or operated by DOE, a TS-SL and a TS-LCS on fuel temperature are established. TS limiting conditions for operation (LCOs) on the three reactivity control administrative limits are derived in Chapter 16 to ensure that the TS-SL and TS-LCS on fuel temperature are not exceeded. Therefore, there are no radiological consequences from the TREAT DBA.

The radiological consequences from the TREAT Experiment DBA are not analyzed in this document. Instead, specifications are placed upon the operation of the experimental apparatus such that if these are met, then the operation of the experiment will be safe. The mechanical design criteria in Chapter 10, Section 10.2 require that the experiment containment retain its integrity during all normal operation and accident conditions, including the maximum unplanned reactivity addition; therefore, no radioactive materials are released to the environment. The experiment containment is designated as a SR-SSC to ensure that the experiment containment retain its integrity during all normal operation and accident conditions. Maintaining experiment containment integrity will ensure that the failure of an experiment in the reactor will not result in consequences that exceed the consequence guidelines in Chapter 15, Table 15-2.

Chapter 10, Section 10.2 requires that a safety analysis be performed for each experiment or group of similar experiments. The primary purpose of the safety analysis for each experiment is to show that the proposed experiment can be conducted in TREAT within the approved envelope and that all safety concerns are adequately accounted for. Therefore, there are no radiological consequences from this event.

As discussed in Chapter 15, Section 15.2, the frequency of the TREAT DBA is less than 10-6/year as a result of the controls identified to prevent the event from occurring. Therefore, the frequency of the TREAT DBA causing an excessive reactivity addition resulting in an experiment malfunction (TREAT Experiment DBA) is also considered to be much less than 10-6/year, and is not required to be analyzed. However, an MHA involving a noncredible, nonmechanistic RIA scenario resulting in total reactor core fuel failure and total failure of the experimental apparatus is analyzed quantitatively in Section 15.12 to assess the residual risk of TREAT transient experiment operations.

An analysis is performed in ECAR-2800 to demonstrate the magnitude of the radiological consequences associated with the noncredible, nonmechanistic TREAT MHA release scenario. Although not required for such a noncredible, hypothesized event, the consequences to the collocated worker at the TREAT control room and the public at the EAB were shown to be well within the consequence guidelines for a more conservative Plant Condition 4 event.

Dose consequences to the facility worker are assumed in the accident analysis to be unacceptable. SR and NSR-AR SSCs and TS controls are identified for each specific accident to ensure that protection is provided for the facility worker, as well as the public, collocated worker, and environment. Safety management programs (SMPs) that provide additional facility worker protection are also identified for each specific accident as programmatic administrative controls (ACs) in the accident analyses. Specific attributes for the SMPs to be effective are specified in Chapter 16 and TS-420.

Table 1-3 summarizes the SSCs and Table 1-4 summarizes the TS controls on key parameters and analysis assumptions derived from Chapter 15, Sections 15.1.4, 15.1.5, and 15.1.6.

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Table 1-3. Summary list of key parameter and analysis assumption SSCs.

Chapter 15 Section SSC Description

15.1.6 TREAT Fuel Assembly (SR-SSC)

15.1.5.3.4 Area Radiation Monitors and Stack Monitoring (NSR-AR-SSC)

15.1.5.4 Reactor Filtration/Cooling System (NSR-AR-SSC) Table 1-4. Summary list of key parameter and analysis assumption TS controls.

Chapter 15 Section Title TS Controls (Chapter 16 Section)

15.1.4.6.1 Total Energy A total energy of 6,990,000 MJ integrated over time from the initial startup of TREAT is the limit that can be deposited in the core during normal operation before additional analysis of fission product production is required (16.5.2.4).

15.1.4.6.1 Steady-State Reactor Power

The reactor power shall be less than 120 kW (16.6.3.2).

15.1.6.1 Fuel Temperature The maximum temperature of the fuel in any assembly shall not exceed 820°C (TS-SL) during the RIA DBA (16.4.1.1).

15.1.6.1 Fuel Temperature The maximum temperature of the fuel in any assembly shall not exceed 600°C (TS-LCS) during a planned transient (16.4.2.1).

As discussed above, the principal restrictions on TREAT reactor operations are derived from the

reactivity insertion DBA in Chapter 15, Section 15.2. Table 1-5 summarizes the TS controls and SR-SSCs or NSR-AR-SSCs derived from all of the accidents analyzed in Chapter 15 as listed above.

Certain NSR-AR-SSCs designated in Table 1-5 (e.g., the RTS, Transient Rod System, and ARCS), have LCO operability requirements defined in TS-420. For all other active NSR-AR-SSCs designated in Table 1-5, operability requirements are defined in TREAT operating instructions.

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Table 1-5. Summary list of SSCs and TS controls derived from the accident analysis.

Chapter 15 Section/Accident

SSC Descriptiona TS Controls (Chapter 16 Section)

15.2 Reactivity Insertion Accidents (TREAT DBA)

TREAT Fuel Assembly (SR-SSC) Compensation/Shutdown Rod Systemb (SR-SSC) Control/Shutdown Rod Systemb (SR-SSC) Manual Scram Subsystemb (SR-SSC) Transient Rod System (NSR-AR-SSC)* Reactor Trip System (NSR-AR-SSC)* Transient Rod Electrical Interlock Subsystem (NSR-AR-SSC) Automatic Reactor Control System (NSR-AR-SSC)* Dedicated Information System (NSR-AR-SSC) Manual Reactor Control System (NSR-AR-SSC) Reactor Control Room (NSR-AR-SSC) Standby Power Subsystem (NSR-AR-SSC) Uninterruptible Power Supply (UPS) Subsystem (NSR-AR-SSC)

RTS Instrumentation Operability (16.5.3.1) Shutdown Rods Operability (16.5.3.2) ARCS Operability (16.5.3.3) Fuel Temperature Limits (16.4.1.1, 16.4.2.1) Limits on Reactor Reactivity (16.5.2.1, 16.5.2.2) TREAT Building Occupancy Restriction (16.6.3.3) Staff Qualification and Training (16.6.2.8) Procedures (16.6.2.4)

15.3 Experiment-Handling Accidents

TREAT Fuel Assembly (SR-SSC) Experiment Containment Subsystem (SR-SSC) TREAT Cranes and Chain Hoists/Falls (NSR-AR-SSC) Fuel Handling and Lifting Apparatus (NSR-AR-SSC) Reactor Building Structure (NSR-AR-SSC) TREAT Loop Handling Cask (NSR-AR-SSC)

Fuel Assembly and Experiment Decay Time (16.5.4.3) Experiment Molten Sodium Control (16.5.4.4) Staff Qualification and Training (16.6.2.8) Procedures (16.6.2.4) Radiation Protection (16.6.2.6.8) Hoisting and Rigging (16.6.2.6.10)

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15.4 Reactor-Fuel-Assembly Handling Accidents

TREAT Fuel Assembly (SR-SSC) TREAT Cranes and Chain Hoists/Falls (NSR-AR-SSC) Fuel Handling and Lifting Apparatus (NSR-AR-SSC) Reactor Building Structure (NSR-AR-SSC) Fuel Handling Cask (NSR-AR-SSC)

Source-Range Instruments Operability (16.5.4.1) Fuel Assembly and Experiment Decay Time (16.5.4.3) Staff Qualification and Training (16.6.2.8) Procedures (16.6.2.4) Radiation Protection (16.6.2.6.8) Hoisting and Rigging (16.6.2.6.10)

15.5 Criticality Events Compensation/Shutdown Rod Systemb (SR-SSC) Control/Shutdown Rod Systemb (SR-SSC) Experiment Storage Subsystem (NSR-AR-SSC) Fuel Storage (NSR-AR-SSC)

Source-Range Instruments Operability (16.5.4.1) Prevention of Inadvertent Criticality (16.6.3.8) Shutdown Margin (16.5.2.3) Nuclear Criticality Safety (16.6.2.6.6)

15.6 System Impact Accidents

Core Support, Alignment, and Concrete Structure (SR-SSC) TREAT Cranes and Chain Hoists/Falls (NSR-AR-SSC) Rotating Shield Plug, Bearing, and Drive Motor Subsystem (NSR-AR-SSC) Fuel Handling and Lifting Apparatus (NSR-AR-SSC) Reactor Building Structure (NSR-AR-SSC)

Crane Lift Height Restriction (16.5.4.2) Staff Qualification and Training (16.6.2.8) Procedures (16.6.2.4) Radiation Protection (16.6.2.6.8) Hoisting and Rigging (16.6.2.6.10) Lifting and Handling Restrictions (16.6.3.7)

15.7 Reactor Fuel Assembly Clad Failure Accidents

TREAT Fuel Assembly (SR-SSC) TREAT Building Occupancy Restriction (16.6.3.3) Staff Qualification and Training (16.6.2.8) Procedures (16.6.2.4) Radiation Protection (16.6.2.6.8)

15.8 Loss of Cooling None None

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15.9 Experiment Malfunction (TREAT Experiment DBA)

Experiment Containment Subsystem (SR-SSC) Radiography Facility (NSR-AR-SSC)

Experiment Safety Analysis (16.5.4.5) ARCS Checkout (16.6.3.1) Core Loading Verification (16.6.3.4) Compliance with Design and Safety Analysis Criteria (16.6.3.6.1) Safety and Operations Review Committee (16.6.3.6.2) Quality Assurance (16.6.3.6.3)

15.10 TREAT Facility Fires

Fire Alarm System (NSR-AR-SSC) Firewater System (NSR-AR-SSC)

Staff Qualification and Training (16.6.2.8) Procedures (16.6.2.4) Radiation Protection (16.6.2.6.8) Fire Protection (16.6.2.6.3)

15.11 Natural Phenomena Hazard Events

TREAT Fuel Assembly (SR-SSC) Seismic Trip Subsystemb (SR-SSC) Compensation/Shutdown Rod Systemb (SR-SSC) Control/Shutdown Rod Systemb (SR-SSC) Manual Scram Subsystemb (SR-SSC) Core Support, Alignment, and Concrete Structure (SR-SSC) Transient Rod System (NSR-AR-SSC)* Rotating Shield Plug, Bearing, and Drive Motor Subsystem (NSR-AR-SSC) Reactor Building Structure (NSR-AR-SSC)

RTS Instrumentation Operability (16.5.3.1) Shutdown Rods Operability (16.5.3.2)

Notes: a. NSR-AR-SSCs, or portions of NSR-AR-SSCs, identified with an * have operability requirements defined in TS-420. For all other active NSR-AR-

SSCs, operability requirements are defined in TREAT operating instructions. b. SR-SSC boundaries are defined in the applicable SAR Chapter 4 or Chapter 7 SSC sections.

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1.4 Detailed Information and Drawings

Detailed information about the TREAT facility is included in the TREAT Baseline Description Document (ANL 1992) and the System and Component Design Descriptions as well as subsequent chapters of this document. Drawing lists are included in the referenced documents.

1.5 References

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

ANL, 1992, TREAT Baseline Description Document, Z0003-0004-OJ, Rev. 12, Argonne National Laboratory.

ANSI/ANS-15.21, 1996, “Format and Content for Safety Analysis Reports for Research Reactors,” American National Standards Institute/American Nuclear Society.

ANSI/ANS-58.14-2011, Safety and Pressure Integrity Classification Criteria for Light Water Reactors, American National Standards Institute/American Nuclear Society, April 22, 2011.

BNL, 1979a, Design Guide for Category V Reactors, Transient Reactors, BNL 50831-V UC-80, TID 4500, Brookhaven National Laboratory.

BNL, 1979b, Design Guide for Category VI Reactors, Air Cooled Graphite Reactors, BNL 50831-VI UC-80, TID 4500, Brookhaven National Laboratory.

INL, 2014, “Transient Reactor Test Facility (TREAT) Final Safety Analysis Report,” S3942-0001-YT, Rev. 8, Idaho National Laboratory.

MacFarlane, D. R., G. A. Freund, and J. F. Boland, 1958, Hazards Summary Report on the Transient Reactor Test Facility (TREAT), ANL-5923, Argonne National Laboratory.

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.

NRC, 1987, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” NUREG-0800, U.S. Nuclear Regulatory Commission.

NRC, 1996, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors,” NUREG-1537, U.S. Nuclear Regulatory Commission.

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

TEV-1832, 2013, “Overview of Anticipated Transient Test Experiments,” Rev. 0, Idaho National Laboratory.

TS-420, “TREAT Technical Specifications,” current rev., Idaho National Laboratory.

Woolstenhulme, N. E., and J. D. Wiest, 2013, ATF Transient Testing Pre-Conceptual Design and Engineering Considerations Summary, INL/EXT-13-29898, Idaho National Laboratory.

Documents

SAR-420, Chapter 1