R406_Student_Manual.pdf

USNRC RESEARCH AND TEST REACTOR

HANDS-ON OPERATIONS TRAINING COURSE

03/2015

STUDENT MANUAL

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USRNC REACTOR BASED TRAINING PROGRAM

TABLE OF CONTENTS

EXERCISES

EXERCISE 1; NETL Reactor and Facility Design, Walkthrough, Pulse E1 Attachment 1.1, NETL Reactor and Facility Design Review E1.1 Attachment 1.2, UT TRIGA Reactivity Summary E1.2 Attachment 1.3, Exercise 1 Review E1.3 PROCEDURE OPER-1, Startup-Shutdown Checks OPER-1 PROCEDURE OPER-2, Reactor Startup and Shutdown OPER-2 PROCEDURE OPER-3, Reactor Operation Modes OPER-3 EXERCISE 2: Reactor Startup E2 Attachment 2.1, Relevant Technical Specifications E2.1 Attachment 2.2: Exercise 2 Review E2.2 EXERCISE 3: Subcritical Multiplication and 1/M Approach to Critical E3 EXERCISE 4: Control Rod Worth and Calibration E4 PROCEDURE SURV-3, Excess Reactivity and shutdown Margin SURV-3 PROCEDURE SURV-6, Control Rod Calibration SURV-6 EXERCISE 5: Fuel Element Worth versus Position E5 PROCEDURE FUEL-1, Movement of Fuel FUEL-1 EXERCISE 6: Power Level Instrument Calibration E6 PROCEDURE SURV-2, Reactor Power Calibration SURV-2 EXERCISE 7: Reactor Pulsing and Square Wave Pulsing E7 Attachment 7.1: Relevant Technical Specifications E7.1 EXERCISE 8: Experiment Operations E8 Attachment 8.1: Relevant Technical Specifications E8.1 Attachment 8.2: Experiment Request E8.2 PROCEDURE ADMN-6, Authorization of Experiments ADMN-6 PROCEDURE FUEL-2, Procedure for Movement of Experiments FUEL-2

USNRC HANDS-ON OPERATIONS TRAINING COURSE SCHEDULE & DESCRIPTION

The schedule for exercises planned in the USNRC Hands-on Operations Training Course over for March 2 through March 6, 2015 is provided. A brief description of the exercises follows the schedule. Exercises are designed to require approximately ½ day periods; adjustments may be possible if exercises are completed outside of the anticipated schedule.

SCHEDULE

(Mon) Mar 2, (Tue) Mar 3, 2015 (Wed) Mar 4, 2015 (Thu) Mar 5, (FRI) Mar 6, Morning INTRODUCTION:

Arrival, course overview, safety & security briefing

EXERCISE 2:

Reactor Startup

EXERCISE 4:

Control Rod Worth and Calibration

EXERCISE 6:

Power Level Instrument Calibration

EXERCISE 8:

Experiment Operations

Break Afternoon EXERCISE 1:

NETL Reactor and Facility Design, Pulse Observation

EXERCISE 3:

Subcritical Multiplication & 1/M Approach to Critical

EXERCISE 5:

Fuel Element Worth versus Position

EXERCISE 7:

Reactor Pulsing and Square Wave Pulsing

CLOSEOUT:

Summary, exit survey

DESCRIPTION

INTRODUCTION: Arrival, course overview, safety & security briefing

The UT Austin TRIGA Security Plan establishes a system of color coded badges to assist identification of authorized building occupants. In addition to a badge, entry into radiological areas requires personal monitoring.

EXERCISE 1: NETL Reactor and Facility Design, Walkthrough, Pulse

Based on significant RTR operating experience, significant facility systems and equipment have been identified; familiarization with material in this exercise will ensure walkthroughs are more effective in communicating concepts and issue in the context of the UT TRIGA reactor facilities. During the facility walkthrough, participants will inspect systems and equipment identified in the facility design review.

Technical Specifications requirements will be emphasized. The exercise will conclude with observation of a reactor pulse.

EXERCISE 2: Reactor Startup

This training program provides experience to aid an intuitive grasp of reactor behavior. This exercise will introduce the participants to reactor controls as a basis for focusing on reactor response.

EXERCISE 3: Subcritical Multiplication and 1/M Approach to Critical

Achieving a critical condition following major changes in the reactor configuration is generally performed using a 1/M process. This exercise will introduce the participants to the use of subcritical multiplication in guiding an approach to critical.

EXERCISE 4: Control Rod Worth and Calibration

Technical Specifications requires knowledge and control of reactor reactivity parameters. Evaluating changes in reactivity parameters requires baseline and calibration of the control rods. This exercise will demonstrate two commonly used methods of control rod reactivity worth determination.

EXERCISE 5: Fuel Element Worth versus Position

Spatial distribution of the neutron flux across the reactor core determines a material’s reactivity value. This exercise will demonstrate spatial dependence of the reactivity on fuel.

EXERCISE 6: Power Level Instrument Calibration

Technical Specifications requires knowledge and control of reactor thermal power. Neutron flux is the leading parameter in the process that results in thermal power from fissions. Power level instruments measure neutron flux, and this exercise will demonstrate calibration of neutron detectors as reactor power level instrumentation.

EXERCISE 7: Reactor Pulsing and Square Wave Pulsing

Temperature effects in TRIGA hydride fuel have a significant impact on reactor response. This exercise will demonstrate changes in power and temperature as a function of prompt reactivity.

EXERCISE 8: Experiment Operations

The conduct of research reactor experiments integrates Technical Specification requirements for administrative procedures and review and approval processes in planning and design with operating procedures and radiological controls. This exercise will demonstrate experiment proposal, review, insertion, and removal.

USNRC TRAINING PROGRAM – 01/2011

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USNRC TRAINING PROGRAM – 03/2015

EXERCISE 1: NETL Reactor and Facility Design, Walkthrough, Pulse

1. PURPOSE AND DISCUSSION

The UT Austin TRIGA Security Plan establishes a system of color coded badges to assist identification of authorized building occupants. In addition to a badge, entry into radiological areas requires personal monitoring.

Certain facility systems and equipment have been identified as significant for training purposes; familiarization with material in this exercise will ensure walkthroughs are more effective in communicating concepts and issue in the context of the UT TRIGA reactor facilities.

During the facility walkthrough, participants will inspect systems and equipment identified in the facility design review. The exercise will conclude with observation of a reactor pulse.

2. EXERCISE OBJECTIVES

Following completion of this exercise, participants should be able to

2.1 Locate normal entrances, egresses and common areas

2.2 Respond appropriately to an emergency

2.3 Employ proper radiological precautions for entering radiological areas and/or handling radioactive materials

2.4 Comply with security requirements related to badging

2.5 Comply with radiological monitoring requirements

2.6 Explain for the following systems:

• Reactor pool and cooling systems

• Control rod system

• Confinement system

• Facility radiation monitoring systems

• Power level instrument systems

• Process instrument systems & controls

• Experiment facilities

• Facility sumps & drains

A. Purpose of the systems

B. Safety functions associated with the system

C. Technical Specifications related to the systems

Facility Introduction, Badge Issue, Safety Briefing E1-2

D. Approximate dimensions and flow paths of key system components

E. The purpose of system control actions, setpoints and alarms

2.7 Observe and read process instruments, and correlate instrument readings with proper operation of systems and equipment

2.8 Explain the properties of ZrH that provide inherent, passive safety

3. PREREQUISITES

3.1 Completion of safety briefing prior to entry of restricted areas

4. BACKGROUND/REFERENCES

4.1 University of Texas Safety Analysis Report

4.2 NUREG 1135, Safety Evaluation report related to the Construction Permit and

Operating License for the Research reactor at the University of Texas (Docket No. 50-602)

4.3 NETL Procedures (ADMN, FUEL, HP, and OPER series)

4.4 NETL Security Plan

4.3 Attachment 1.1, NETL Reactor and Facility Design

4.4 Attachment 1.2, UT TRIGA Reactivity Summary

4.5 Attachment 1.3, Exercise 1 Review

5. INSTRUCTIONS

5.1 CHECKOUT security badge

5.2 CHECKOUT dosimetry

5.3 PARTICIPATE in safety briefing

5.4 REVIEW NETL Reactor and Facility Design (ATTACHED)

5.5 PARTICPATE in a tour

A. NETL laboratory area

Facility Introduction, Badge Issue, Safety Briefing E1-3

B. Equipment and machine rooms

C. Reactor bay

D. Pool cooling room

E. Auxiliary room

F. Control room

G. HVAC rooms

5.6 OBSERVE a reactor pulse (pool side)

6. REINFORCEMENT & REVIEW

Attachment 1.3, Exercise review

6.1 Using the letter labels for components on the pool cooling and the pool cleanup schematics, label equipment in the photographs or draw lines connecting the labeled components to the equipment in the photographs

6.2 Draw lines connecting the labels to the core components in the photograph, and

6.3 Indicate which beam port supports the experiment described experiment facilities

6.4 Write the Technical Specifications value for reactivity limit described in the table

USNRC TRAINING PROGRAM – 11/2014

Attachment 1.1: NETL Reactor and Facility Design Review

Facility Introduction, Badge Issue, Safety Briefing E1.1-1

1.0 Background

Research and test reactor significant operating experience was analyzed (Fig 1.1). The relative contribution of systems and equipment to the experiences include (ordered by contribution to occurrences):

• Reactor pool

• Control elements

• Shipping (of radioactive materials)

• Miscellaneous

• Confinement (or containment)

• Radiation Instruments & Controls

• Power Level Monitoring Instruments & Controls

• Process Systems Instruments & Controls

Figure 1: Contribution of Equipment or System Failures to Operating Experience

Facility Introduction, Badge Issue, Safety Briefing E1.1-2

USNRC Key Topics (10/20/2010) indicate current issues relevant to the NETL facility including:

• Contaminated water releases

• Facility security

• Reactor experiments

• Radiation protection

• Safety culture

Based on this review, the following facility characteristics will be addressed and (where possible) observed in a facility walkthrough:

1.1.1 Reactor Pool and cooling

1.1.2 Control Rod System

1.1.3 Confinement

1.1.4 Facility Radiation Monitoring

1.1.5 Power level instruments

1.1.6 Process System Instruments and controls

1.1.7 Experiment Facilities

1.1.8 Facility Sumps & Drains

1.1 Outline

1.1.1 Reactor Pool and Cooling

Pool cooling function:

a. Reactor heat removal

b. Vertical shielding of radiation from the reactor and allow access to the reactor core

Facility Introduction, Badge Issue, Safety Briefing E1.1-3

Purification:

a Reactor component (particularly fuel elements) corrosion control

Facility Introduction, Badge Issue, Safety Briefing E1.1-4

b. Radioactive contamination control

c. Water (optical) clarity

d. Makeup water interface

SIGNIFICANT SHIELDING & POOL LEVEL VLAUES

Parameter of Interest Level (meters) Notes

CONCRETE PAD -2.191 SAR FIG 7-1: Concrete below pool; SAR 7.2.1 indicates 1.22 meter pad; Drawing RS-6, 5 ft min

FLOOR 0 Pool floor ½” aluminum; pool wall ¼” aluminum

TOP OF SAFETY PLATE 0.197 SAR FIG 4-25, 4.4.4: ½” th. 16” <grid plate, 0.775 m below CL (7.75” above floor)

TOP OF LOWER GRID PLATE 0.642 SAR FIG 4-25, SAR 4.4.3: 1 ¼” thick, 0.332 m below

CL

BOTTOM OF FUEL 0.781 SAR FIG 4-25: ½ fuel length, 7.5 in./0.191 below CL

BEAM PORT CL 0.902 SAR FIGS 4-25/7-1, 8.1.4.3: 6 in. diameter (0.191 m) below fuel CL

CORE CL 0.972 SAR FIGS 4-25/7-1

TOP OF FUEL 1.162 SAR FIG 4-25: ½ fuel length , 7.5 in. (0.191 m) above fuel CL

GRID PLATE 1.295 SAR 4.4.3: ⅝ in. thick (0.324 m) over CL

Facility Introduction, Badge Issue, Safety Briefing E1.1-5

SIGNIFICANT SHIELDING & POOL LEVEL VLAUES

Parameter of Interest Level (meters) Notes

MAIN LOWER SHIELDING 4.267 SAR FIG 7-1: 7.97 feet thick around pool TRANSITIONAL CONCRETE STEP 4.877 SAR FIG 7-1: 3 feet thick around pool

SHIFT TO HIGH DENSITY

CONCRETE 5.944 SAR FIG 7-1: 2.3 g/cc to 2.9 g/cc

MIN POOL LEVEL (TS) 6.5 TS 3.3, A.3.3.b: 5.25 m over the core VACUUM BREAKERS 6.7 OPER-4 II.A.2.b

LOW POOL LEVEL SCARM 7.8 SCRAM MAIN-1 ATT 1

LOW POOL LEVEL 8.05 Minimum, fill required OPER-4 II.A.2.a LOW POOL LEVEL ALARM 8.07 Alarm MAIN-1 ATT 1 NORMAL POOL LEVEL 8.1 NOMINAL OPER-4 II.A.2.a HIGH POOL LEVEL 8.15 Maximum OPER-4 II.A.2.a HIGH POOOL LEVEL ALARM 8.17 Alarm MAIN-1 ATT 1

TOP OF TOP DECK (NOT LEDGE) 8.534 SAR FIG 7-1

1.1.2 Reactor Core & Control Rods

Facility Introduction, Badge Issue, Safety Briefing E1.1-6

Facility Introduction, Badge Issue, Safety Briefing E1.1-7

CONTROL ROD SPECIFICATIONS

Transient Shim & Regulating

Cladding Material Al Type 304 SS OD 1.25 in. (3.18 cm) 1.35 in. (3.34 cm) Length 36.75 in. (93.35 cm) 43.13 in. (109.5 cm) Wall Thickness 0.028 in. (0.071 cm) 0.020 in. (0.051 cm)

Absorber Material Boron Carbide (solid form) Boron Carbide (solid form) OD 1.19 in. (3.02 cm) 1.31 in. (3.32 cm) Length 15 in. (38.1 cm) 14.25 in. (36.20 cm)

Follower Material Air Fuel OD 1.25 in. (3.18 cm) 1.31 in. (3.34 cm) Length 20.88 in. (53.02 cm) 15 in. (38.1 cm)

Facility Introduction, Badge Issue, Safety Briefing E1.1-8

1.1.3 Confinement System

SAR 7.2.2, The ventilation system is designed to maintain a negative Pressure within the reactor bay with respect to the building exterior and other building areas. Confinement and isolation is achieved by air control dampers and leakage prevention material at doors and other room penetration points. A separate system is designed to exhaust air from several locations within the reactor that could contain airborne radionuclides such as argon-41. Manual operation of start/stop controls of both main and purge air systems will be available in the reactor control room.

Facility Introduction, Badge Issue, Safety Briefing E1.1-9

Facility Introduction, Badge Issue, Safety Briefing E1.1-10

Velocity 3900 fpm Argon purge flow velocity 1800 fpm HVAC flow velocity

Flow Rate 7200 cfm HVAC flow 1100 cfm Argon purge flow

1.1.4 Facility Radiation Monitoring

Safety Analysis Report, 9.2. RADIATION MONITORING:

Radiation monitoring shall consist of fixed, portable, or sampling type systems.

Monitoring systems will be applied to measurement of radiation areas and high radiation areas around the reactor facility, significant contamination within and adjacent to the reactor facility, and radioactive materials and their concentrations in effluents.

Continuous monitoring or control of radiation fields in the restricted area around the reactor shall occur whenever levels greater than 100 mrem/hr are produced in accessible areas.

Airborne radioactive monitoring shall consist of continuous sampling of air particulate activity in the reactor area. Warning levels and action levels will be determined relative to allowable maximum permissible concentrations. Measurements should be sensitive to one maximum permissible concentration change in one hour.

Facility Introduction, Badge Issue, Safety Briefing E1.1-11

Effluent monitoring shall be provided for the discharge of the radioactive noble gas argon-41. Monitoring will consist of either the use of integrating dosimeters at a location of interest or sampling of a point in the release path. Measurements shall determine that the dose at a location of interest is either less than ten mrem per year above natural background or two percent of the allowable maximum permissible concentration for the year.

Fixed area gamma monitors shall have remote readouts with audible and visual alarms at the reactor control console. Local readouts should be provided in areas with significant radiation levels and routine personnel access. A multiple channel area monitoring system with GM type detector probes will be installed. The system should have at least four channels functional. Measurement should include the dose range of 1 millirem/hour to 1 rem/hour.

A continuous air particulate monitor with audible and visual alarms shall be functional in the reactor vicinity during reactor operations. A fixed filter beta particulate monitor with 70 1pm flow rate capacity or equivalent system will provide air particulate monitoring.

Detectors such as thin window GM detectors will monitor activity and provide alert and alarm conditions with visual and audible annunciators. Count rate of the instrument should include the range of 50 to 50,000 counts/minute.

A gas monitor system for the noble gas effluent, argon-41, shall also be operable during operation, or sufficient data available to demonstrate a calculated release quantity.

Design goal for the argon-41 monitor is a sensitivity of 50% of 1 mpc (maximum permissible concentration) for unrestricted areas. Test measurements indicate sensitivity for a ten minute count of 2 x 10-8 μCi/cc.

1.1.5 Power Level Instrumentation

Safety Analysis Report, Chapter 6: The Instrumentation and Control System for the TRIGA reactor is a computer-based design incorporating the use of one multifunction, NM-1000 microprocessor neutron flux monitoring channel and two companion NP-1000 current mode neutron monitoring safety channels. The combination of these two systems provides an independent operating channel and the redundant safety function of percent power with scram.

The NM-1000 provides wide range log power and multi-range linear power from source level to full power. The control system logic is contained in a separate control system computer (CSC) with graphics and text displays which are the interface between the operator and the reactor. Another system for data acquisition and control (DAC) functions as the interface point for interface circuitry, process signals and communications.

Facility Introduction, Badge Issue, Safety Briefing E1.1-12

1.1.6 Process System Instruments and controls

Facility Introduction, Badge Issue, Safety Briefing E1.1-13

Facility Introduction, Badge Issue, Safety Briefing E1.1-14

1.1.7 Experiment Facilities & Operations

Facility Introduction, Badge Issue, Safety Briefing E1.1-15

1.1.8 Facility Sumps & Drains

Pedestal foundation drain; automatically discharges to sanitary sewerage

Facility Introduction, Badge Issue, Safety Briefing E1.1-16

Liquid radiological waste holding tank; services radiological labs, safety shower, decontamination shower, with batch drain from waste holding tanks

Reactor building foundation drains sump in the truck access ramp; automatically discharges to concrete diversion structure between NETL and Burnet Rd

1.2 References

1.2.1 Reactor Pool and Cooling

Safety Analysis report The University of Texas TRIGA, Chapter 5 (Reactor Coolant System)

Pool and Cooling

Heat dissipation is satisfied by natural convective flow of pool water through the reactor core and forced circulation of the pool water through an external heat exchanger. The pool coolant volume is composed of approximately 41.0 cubic meters in a two by three meter oval pool with a vertical depth of 8.1 meters. A vertical shield is provided by about 6.8 meters of water above the reactor core.

Facility Introduction, Badge Issue, Safety Briefing E1.1-17

Supplemental cooling of the reactor pool is required for continuous operation at the rated power level. A heat rate of 20.7°C/hour is expected with the reactor operated at 1000 kW. Heat removal from the pool is provided by heat exchange with a chilled water supply. The chilled water supply is operated by the University for cooling of Research Center buildings and equipment. Chilling capacity is provided by multiple 1200 ton (4220 kW) units. At reactor rated power the heat removal capacity required is represented by about 25% of the chilling system capacity of one unit. A tube and shell heat exchanger is installed for heat removal from the reactor pool to the available chilled water system.

Maximum design temperature of the coolant system, coolant inlet temperature, is 120°F (48.9°C). The maximum allowable peak heat flux at this temperature is 325 kBtu/hr-ft2 (103 watts/cm2) corresponding to a power level of 1900 kW for an 85 element core. Since the maximum licensed power level is 1100 kW, the resulting maximum heat flux will be 188 kBtu/hr-ft2 (59.4 watts/cm2) which is well below the value at which clad integrity may be questioned.

Suction of water from the pool is provided by an inlet which extends no more than 2 meters below the top of the reactor tank. The coolant water is drawn through the coolant pump and forced through the heat exchanger. Return of cooled water to the reactor pool is provided by single or multiple discharge outlets above the reactor core or an outlet near the tank bottom. A diffused water jet is created at the outlets above the reactor core by a nozzle. Delay and diffusion of the reactor core convective coolant column is enhanced by the action of the coolant discharge nozzle.

Accidental siphoning of reactor pool water is prevented by the presence of suction breaks on both suction and discharge lines of the coolant system. Siphon breaks are created by holes located in the lines approximately half a meter below the normal water level.

Indication of the reactor pool status is determined by two sensors located in the pool. Pool level and bulk pool temperature sensors in the pool are monitored in the control room. An annunciator alarm indication is generated in the control room by abnormal pool levels and by high pool temperatures.

Heat removal capacity and thus pool heat rate is specified by analysis of a tube and shell heat exchanger. At a flow rate of 400 gal/min (25.2 liters/sec) of chilled water at 48°F (8.89°C) a heat removal rate of 1140 kW is expected.

The heat exchanger and pump, the major components of the cooling system, are located in room 1.104b at about the same vertical level as the reactor core. Valves are provided in the coolant loop for control and isolation of the cooling system function.

Specifications of cooling system components are listed in Table 5-1. A positive pressure difference of 1 psi (7 kilopascals) between the shell side outlet and tube side

Facility Introduction, Badge Issue, Safety Briefing E1.1-18 inlet of the heat exchanger is designed to prevent leakage of primary pool coolant into the secondary chilled water system. The pressure difference is maintained under varying flow conditions by a differential pressure controller which regulates the position of a throttle valve in the heat exchanger shell side outlet pipe. Coolant water supply temperature is regulated by a temperature controller coupled to a mixing valve in the chilled water supply line.

The cooling system parameters normally available in the control room include coolant temperatures, flowrates, and differential pressure status. Two temperature probes, one in the pool suction line and one in the pool discharge line, allow monitoring of heat exchanger cooling function. Typical temperature probes used are resistance temperature detectors (RTD's). Two flow meters, one in the chilled water line and one in the pool water line provide information on system flow rates. A differential pressure monitor provides an alarm if the pressure at the high pressure point on the heat exchanger tube side is not less than the low pressure point on the shell side. The differential pressure is designed for a difference substantially greater than 7 kilopascals (1 lb/sq. in.).

Suction of water from the pool is provided by two inlets, neither of which extend more than 2 meters below the top of the reactor tank. Valves at the pool surface allow suction from either a subsurface inlet or from a surface skimmer designed to collect and remove floating debris. Accidental siphoning of reactor pool water is prevented by siphon breaks similar to those on the coolant piping.

Purification System

The purification skid is located in room 1.104b at about the same level as the reactor core. The skid consists of a pump, flowmeter, filter, resin bed, and instrumentation.

Normally the purification system is operated continuously to provide removal of suspended particles and soluble ions in the coolant water. The system flow rate is about 10 gpm (0.6 1ps).

Purification functions of the loop are generated by two components, a filter for removal of suspended materials and a resin bed for removal of soluble elements. Typical filtration is provided with 25 micron filters. Typical ion exchange is provided by .085 cubic meters of mixed cation and anion resin. Water purity is measured by conductivity cells at the inlet and outlet of the resin beds.

Return flow to the pool is thru a subsurface discharge pipe. Valves are provided for isolation of the suction or return lines, and for isolation of system components for maintenance or resin replacement.

Water quality is measured by two conductivity cells in the purification loop. The cells are located on inlet and outlet lines of the demineralizer that readout locally in the control room. Typical conductivity cells are composed of two parts, titanium electrodes shielded by ryton for conductivity measurement, and a thermister for temperature

Facility Introduction, Badge Issue, Safety Briefing E1.1-19 compensation. A Wheatstone bridge circuit on the purification skid is connected to the cells. A switch allows selection of either inlet or outlet conductivity.

A connection from the purification system to the domestic water system provides makeup water to replenish pool evaporation losses. The makeup water connection includes features that isolate the two water systems. Two of these features are valves for flow control and a quick release connection. The valves include a check valve for limiting flow direction and one or more block valves for stopping water flow. Use of the quick release connection allows physical separation of the two systems except during periods in which the makeup process is operating.

An indication of radioactive release as alternatives to CAM and Ar41 monitor is provided if a water activity monitor is installed or by a GM detector area monitor.

1.2.2 Reactor Core & Control Rods

The TRIGA reactor system has three major areas which are used to define the reactor design bases:

a. Fuel temperature,

b. Prompt negative temperature coefficient,

c. Reactor power.

These areas are interrelated, and fuel temperature is selected as the safety parameter most directly applicable to fuel cladding integrity. A ratio of 1:1.6 was selected to ensure stable structure and volume.

Zr H Phase Diagram

Facility Introduction, Badge Issue, Safety Briefing E1.1-20

Thermal hydraulic analysis was performed using radial power peaking factors, TRIGA fuel geometry, and cooling channel geometry defined by the core grid plate and fuel elements.

Facility Introduction, Badge Issue, Safety Briefing E1.1-21

Fuel temperatures were found to be acceptable up to departure form nucleate boiling. The analysis demonstrates that TRIGA fuel cooled by natural convection is capable of about

1.9 MW power without experiencing departure from nucleate boiling.

Facility Introduction, Badge Issue, Safety Briefing E1.1-22

The UT TRIGA maximum power level limit was established at 1.1 MW to provide a wide margin to departure from nucleate boiling under all conditions. TRIGA fuel has an extremely large negative feedback coefficient associated with the fundamental physics of ZrH.

Relative magnitude of Contributing Components to the Prompt Negative Temperature Coefficient of

TRIGA Reactors (ZrH1.6 – SS cladding)

Contributing Component (%) Zr Hydride Effect 55 Inhomogeneities (control rod locations) 5 Doppler broadening of 238U resonances 20 Leakage from Core 20

Therefore enough uranium has to be added to compensate for (1) operation at elevated temperatures, (2) xenon generation, and (3) fuel burnup.

Control rod design is required to assure the reactor remains subcritical with the most reactive rod failed in the fully withdrawn position.

Facility Introduction, Badge Issue, Safety Briefing E1.1-23

1.2.3 Confinement System

Safety Analysis Report, Chapter 6: Ventilation of the reactor bay is provided by two modes of system operation. One mode is for standard operation with recirculation of air.

The other mode is an exhaust operation with high volume flow that has no air recirculation. Design during exhaust mode operation is a rate of air exchange in excess of two per hour. Total volume for the room exceeds 4120 cubic meters. Normal operation of the ventilation system uses a roof stack for the exhaust of air from the reactor bay.

Isolation of the reactor bay is provided by ventilation dampers. These dampers will shut in response to either manual or automatic signal actuation. An automatic signal will initiate shutdown of the ventilation system by closure of the dampers if a set point for airborne particulate radioactivity exceeds a set point, Protective switches within the ventilation system will cause the air fans to respond to the position change of the dampers. Damper design is for fail-safe operation so that loss of control power will isolate the reactor bay. The separate air purge system is designed to exhaust air that may contain radionuclide products by a low volume system.

The exhaust stack will extend 14 feet (4.24 meters). Ground elevation in the area is 794 feet. Roof elevation at the stack is 843 feet. Therefore, the effective release point is at least 50 feet above the maximum ground elevation at the building.

Relative to ambient atmospheric pressure, the design goals for the reactor bay, adjacent zones and academic wing of the building is a negative pressure difference. The differential pressures are 0.06: 0.04: 0.03 inches (0.15: 0.10: 0.80 cm.) of water. Air balance values relate typical or standard values except in the case of the reactor bay.

Argon-4l is produced in the reactor pool as a result of the (n,γ) reaction with argon-40 dissolved in the pool water. Most of the argon-4l remains in solution but some of it is transferred to the reactor room air at the pool surface. Calculations of the argon released from the pool surface estimate a concentration in the room of 1.6 x 10-6 pCi/cc with the reactor operating at 1000 kW.

1.2.4 Facility Radiation Monitoring

SAR 9.5. EVALUATION OF MONITORING SYSTEMS

The radiation monitors provide information to operating personnel of impending or existing hazards from radiation so that there will be sufficient time to take the necessary steps to control the exposure of personnel and the release of radioactivity or to evacuate the facility.

Three types of radiation monitors are used: a continuous air particulate monitor for determining radiation levels due to particulate radioisotopes suspended in the reactor room air, a continuous air gaseous monitor for determining radiation levels due to argon-

Facility Introduction, Badge Issue, Safety Briefing E1.1-24

41 in the room air, and area radiation monitors for determining the gamma field at several locations in the facility.

Each type of radiation monitor has a specific radiological purpose. The particulate air monitor is used to detect radioisotopes released due to fuel element failure (a design basis accident). The gaseous air monitor is used to determine the effluent radiation release of argon-41. Argon, a component of air (.04% by volume) may be activated to produce argon-41 in potentially significant quantities. Finally, the area radiation monitors are used to minimize personnel radiation exposures. The facility radiation monitors discussed in SAR sections 9.5.1., 9.5.2, and 9.5.3 are typical instruments at the time of original installation. Replacements may have slightly different characteristics.

Set points for the particulate continuous air monitor warn of the presence of particulate fission product nuclides. Since gaseous and volatile elements such as krypton, xenon, bromine, and iodine have particulate decay products, the presence of some of their radioisotopes should also be detected. An alarm set point at 2000 picoCuries/milliliter detects particulate activity concentrations at the occupational values of 10 CFR 20 for 70% of the relevant isotopes in the ranges 84-105 and 129-149. These ranges of isotopes represent the one percent yield for fission products of uranium-235.

The air monitor in use is a Ludlum Model 333-2 beta air monitor, configured for continuous sampling of airborne beta-emitters. It uses two standard pancake G-M tubes, each having a 1.75 inch effective diameter. The two detectors are arranged in line so that gamma background subtraction is performed. This increases the accuracy of the beta count.

The 333-2 will accept air flow rates ranging from 10-100 liters per minute. Particle accumulation on a fixed filter continues at a constant rate. Activity on the filter, however, is a function of the air flow rate, filter collection efficiency, and the decay rates for nuclides that are present on the filter.

A high radiation area, defined in 10 CFR 20 as having a radiation level >100 mrem/hr, may exist above the pool access area during some operations. The area radiation monitor located above the pool access area will have an alarm set point of 100 memr/hr. Although the radiation levels within the pool protection railings may exceed doses of 100 mrem/hr, the dose exists only in the immediate area of the pool surface. At other locations of the pool shield platform level, the dose rates are significantly less than 100 mrem/hr, but may exceed the 2 to 5 mrem/hr range. While the reactor is operating, one area radiation monitor will operate above the pool access area, as well as at least two additional area radiation monitors located at other positions around the reactor shield and at the beam port facilities. The radiation monitor system consists of six units with GM tubes that detect dose rates from 0.1 mR/hr to 10 mR/hr.

1.2.5 Power Level Instrumentation

Safety Analysis Report, 5.1.1. NM-1000 Neutron Channel The NM-1000 nuclear channel has multifunction capability to provide neutron monitoring over a wide power range from a single detector. The selectable functions are any or all of the following:

Facility Introduction, Badge Issue, Safety Briefing E1.1-25

a. Percent power.

b. Wide-range log power

c. Power rate of change.

d. Multi - range linear power.

For the TRIGA ICS, one NM-1000 system is designated to provide the wide-range log power function and the multirange linear power function. The wide-range log power function is a digital version of the patented GA 10-decade log power system to cover the reactor power range from below source level to l50% power and provide a period signal.

For the log power function, the chamber signal from startup (pulse counting) range through the Campbelling (root mean square, RMS, signal processing) range covers in excess of 10-decades of power level. The self-contained microprocessor combines these signals and derives the power rate of change (period) through the full range of power.

The microprocessor automatically tests the system to ensure that the upper decades are operable while the reactor is operating in the lower decades and vice versa when the reactor is at high power.

For the multirange function, the NM-1000 uses the same signal source as for the 1og function. However, instead of the microprocessor converting the signal into a log function, it converts it into 10 linear power ranges. This feature provides for a more precise reading of linear power level over the entire range of reactor power. The same self-checking features are included for the log function. A linear power level signal is available for the percent power safety function.

The NM-1000 system is contained in two National Electrical Manufacturers Association (NEMA) enclosures, one for the amplifier and one for the processor assemblies. The amplifier assembly contains modular plug-in subassemblies for pulse preamplifier electronics, bandpass filter and RMS electronics, signal conditioning circuits, low voltage power supplies, detector high-voltage power supply, and digital diagnostics and communication electronics. The processor assembly is made up of modular plug-in subassemblies for communication electronics (between amplifier and processor), the microprocessor, a control/display module, low-voltage power supplies, isolated 4 to 20 mA outputs, and isolated alarm output. Outputs are Class lE as specified by IEEE.

Communication between the amplifier and processor assemblies is via two twisted-pair shielded cables.

The system is automatically calibrated and checked (including the testing of trip levels) prior to operation. The checkout data is recorded for future use, and operation cannot proceed without a satisfactorily completed checkout. The neutron detector uses the standard 0.2 counts per nv fission chamber that has provided reliable service in the past.

It has, however, been improved by additional shielding to provide a greater signal-to-noise ratio. The low noise construction of the chamber assembly allows the system to respond to a low reactor shutdown level which is subject to being masked by noise.

The NP-1000 Power Safety Channel is a complete linear percent power monitoring system mounted within one compact enclosure which contains current to voltage

Facility Introduction, Badge Issue, Safety Briefing E1.1-26 conversion signal conditioning, power supplies, trip circuits, isolation devices, and computer interface circuitry. The power 1evel trip circuit is normally hardwired into the scram system and the isolated analog outputs are monitored by the CSC as well as being hardwired to a bar graph indicator. A special version of the safety channel, the NPP- 1000, provides measurement functions for peak pulse power, total pulse energy, automatic gain change and related trip points. The control system automatically selects proper gain setting for steady-state or pulse mode when the operator determines the reactor operating mode. Peak pulse power and total pulse energy are also set by the pulse operation mode. Both safety channels, the NP-1000 and the NPP-1000, are identical except for the peak and energy circuits. The detector for each safety channel is either an ionization chamber or self-powered in-core detector.

1.2.6 Process System Instruments and controls

Several monitoring sensors are installed to allow remote readout of water system parameters in the reactor control room. Other system parameters are indicated by local monitoring devices.

Indication of the reactor pool status is determined by two sensors located in the pool.

Pool level and bulk pool temperature sensors in the pool are monitored in the control room. An annunciator alarm indication is generated in the control room by abnormal pool levels and by high pool temperatures. The cooling system parameters normally available in the control room include coolant temperatures, flowrates, and differential pressure status. Two temperature probes, one in the pool suction line and one in the pool discharge line, allow monitoring of heat exchanger cooling function. Typical temperature probes used are resistance temperature detectors (RTD's). Two flow meters, one in the chilled water line and one in the pool water line provide information on system flow rates. A differential pressure monitor provides an alarm if the pressure at the high pressure point on the heat exchanger tube side is not less than the low pressure point on the shell side.

The differential pressure is designed for a difference substantially greater than 7 kilopascals (l lb/sq. in.).

Numerous water system parameters are measured by local pressure or temperature sensors in the system lines. Both temperature and pressure probe are located on the inlet and outlet lines of the pool water side and chilled water side of the heat exchanger. A local indication of flow in chilled water side of the coolant loop is provided by the pressure drop across a venturi in the flow path. Purification loop flow is measured by an in line flow meter. Water pressure before and after the filter in the purification loop is measured for indication of filter condition. Water quality is measured by two conductivity cells in the purification loop. The cells are located on inlet and outlet lines of the demineralizer that readout locally in the control room. Typical conductivity cells are composed of two parts, titanium electrodes shielded by ryton for conductivity measurement, and a thermister for temperature compensation. A Wheatstone bridge circuit on the purification skid is connected to the cells. A switch allows selection of either inlet or outlet conductivity.

Facility Introduction, Badge Issue, Safety Briefing E1.1-27

A connection from the purification system to the domestic water system provides makeup water to replenish pool evaporation losses. The makeup water connection includes features that purify the domestic water before introduction into the pool water, and valves that isolate the two water systems. Two of these features are valves for flow control and a quick release connection. The valves include a check valve for limiting flow direction and one or more block valves for stopping water flow. Use of the quick release connection allows physical separation of the two systems except during periods in which the makeup process is operating.

1.2.7 Experiment Facilities & Operations

Central thimble: aluminum tube that fits through the center hole of the top and bottom grid plates 1.5 in. o.d. (3.81 cm.) and 1.33 in. i.d. (3.38 cm.). Holes in the tube above the lower grid plate assure that it is normally filled with water in the entire core region.

Rotary Specimen Rack: multiple-position (40) specimen rack located in a well in the top of the graphite reflector provides for the large scale production of radioisotopes and for the activation and irradiation of multiple samples. All positions in this rack are exposed to neutron fluxes of comparable intensity. Specimen positions at 1.23 in. (3.18 cm.) in diameter by 10.80 in. (27.4 cm.) in depth. Samples are loaded from the top of the reactor through a water-tight tube. The facility rotates either manually or via a motor which shares the power supply for the pool lights

GA pneumatic specimen tube: sample capsule (rabbit) is conveyed to a receiver-sender station via 1.25 in. o.d. (3.18 cm.) aluminum tubing. Effective space in the specimen transfer capsules is 0.68 in. (1.7 cm.) diameter by 4.5 in. (f1.4 cm.) height. An optional transfer box may be employed to permit the sample to be sent and received from up to three different receiver-sender stations.

UT-design pneumatic tube system: the sample capsule is conveyed through a transit section of polyethylene tubing ¾” i.d. to the in-pool/core terminus slightly less than 5/8” i.d. Two in-pool/core assemblies provide cadmium covered section or bare.

Facility Introduction, Badge Issue, Safety Briefing E1.1-28

There are five beam ports divided into two categories as follows:

Tangential beam ports: Two beam ports are oriented tangential to the reactor core, penetrate the graphite reflector, the coolant water, and the concrete shield. A hole is drilled in the graphite tangential to the outer edge of the core. One beam port terminates at the tangential point to the core. The other beam tubes extend both directions from the reflector and out opposite sides of the reactor shield.

Radial beam ports: There are two radial beam ports, each which penetrate the concrete shield structure and the coolant water. One radial port terminates at the inner edge of the reflector. The second radial port also terminates at the outer edge of the reflector, however a hole drilled in the graphite reflector extends the effective source of radiation to the core fuel region.

Facility Introduction, Badge Issue, Safety Briefing E1.1-29

Three element feature: the 3-El is a 2.062 in. (5.24 cm) diameter hole. Location of the hole center coincides with the center of a three element sub array. Total area of the hole is

3.34 in.2 (21.5 cm2), but also contains additional area for each of the three elements of the original sub array. The area available under the cutout for experiments is limited by clearance for adjacent elements to less than the area of the central and three adjacent fuel element positions. Three single element holes have a total area of 5.34 in.2 (34.4 cm2).

One three element sub array cutout in the sub array is one D ring and two E ring locations. The other three element sub array is one E ring and two D ring locations.

Six element feature: A 6-El removable grid space is located at the center of the core grid structure controls the position of the center hole and adjacent B ring elements. Removal of the A and B ring element array will create a 4.005 in. (10.17 cm) diameter hole. This hole has an area of 12.6 in.2 (81.3 cm2), centered at reactor core center. Additional space is available in the partial element holes that remain in the B ring. This additional space is the result of the experiment hole diameter, which exceeds the diameter of the center hole, but is less than a diameter that would include all the B ring holes. A limit shall exist on the use of this space such that the total area available to the experiment remains less than

15.8 in.2 (30 cm2).

Seven Element Feature: during reflector change out, a facility similar to the 6-El was fabricated, but displaced into the outer rings.

Experiment evaluation requirements:

Double encapsulation is required for experiments containing materials:

• Corrosive to reactor components, compounds highly reactive which water, potentially explosive materials, and liquid fissionable materials shall be doubly encapsulated.

• NFPA reactivity ratings greater than two or fire ratings equal to four should be doubly encapsulated. The encapsulation is to protect against the energy release and corrosion properties.

• Materials having a reactivity rating of 1 or a flammability rating greater than 1 should be evaluated individually to determine if double encapsulation is warranted.

• More than 100 ppm of thorium, uranium, or plutonium.

No more than 25 mg TNT equivalent explosive material is permitted

In fueled experiments, total radioactive iodine inventory (131-135) shall be no greater than 750 mCi, with maximum strontium inventory no greater than 2.5 mCi

Experiment materials, except fuel materials which could off -gas, sublime, volatize, or produce aerosols under one of the following conditions shall be limited to occupational

Facility Introduction, Badge Issue, Safety Briefing E1.1-30 levels for airborne radioactivity concentration, as specified in 10CFR20, where averaging over a year for:

• Normal operating conditions of the experiment or reactor

• Credible accident conditions in the reactor

• Possible accident conditions in the experiment.

1.2.8 Facility Sumps & Drains

1.3 Technical Specifications

1.3.1 Reactor Pool and Cooling

SECT DESCRIPTION SPECIFICATION SR FREQ Procedure/record

3.3.1.a Pool cooling 48°C 4.3.1.a Annual 4.3.1.a Monthly 4.3.1.a During operation

3.3.1.b Pool cooling 6.5 m 4.3.1.b Annual 4.3.1.b Monthly 4.3.1.b During operation

3.3.1.c Pool cooling 5.0 µmho <mo>

4.3.1.c Annual

4.3.1.c Monthly SURV-4 Reactor Water Systems

Surveillance 4.3.1.c During operation

3.3.1.d Pool cooling <1 psig HX dp 4.3.1.d Daily & retest 4.3.1.d During operation

3.3.1.e Pool cooling Periodic measurement ph 4.3.1.e Quarterly

SURV-4 Reactor Water Systems

Surveillance

3.3.1.e Pool cooling Periodic measurement α-β 4.3.1.e Quarterly

SURV-4 Reactor Water Systems

Surveillance

3.3.1.e Pool cooling Periodic measurement γ spec

4.3.1.e Annually

5.2.1 Reactor coolant

system Natural convection

5.2.2 Reactor coolant

system Siphon breaks

5.5.a Reactor pool irradiator ≤ 10KCi Co-60

5.5.b Reactor pool irradiator ≥4.5 m under water

5.5.b Reactor pool irradiator ≥ 0.5 m form wall

5.5.c Reactor pool irradiator

Pool water 2.5X10x test gamma irradiator

Facility Introduction, Badge Issue, Safety Briefing E1.1-31

1.3.2 Reactor Core & Control Rods

2.1 Safety limit Fuel

Temp Clad <500°C, <1150°C

2.1 Safety limit Fuel

Temp Clad >500°C, <950°C

2.2.1 Fuel temp LSSS < 550°C

2.2.1 Fuel temp LSSS IFE in B or C ring

SURV-1 Fuel temperature calibration

2.2.2 Power level 1100 kW

2.2.3 Reactivity insertion 2.2% Δk/k (pulse)

3.1.1 Excess reactivity 4.9% Δk/k 4.1.1

Annually or post core/control rod changes

SURV-3 Excess Reactivity and

Shutdown Margin

3.1.2 Shutdown margin

0.2% Δk/k (ref core, 1 rod out, moveable experiments most reactive)

4.1.2 Annually or post core/control rod

changes

SURV-3 Excess Reactivity and

Shutdown Margin

3.1.3 Transient rod 2.8% Δk/k worth

SURV-6 Control Rod Calibration

4.1.3 Annual

SURV-7 Pulse Characteristic Comparison

3.1.3 Transient rod < 15 s withdrawal time 4.1.3

Annually or post core/control rod changes

SURV-6 Control Rod Calibration

3.1.4.a Fuel elements < 2.54 mm elongation 4.1.4 3.1.4.b Fuel elements < 1.5875 mm bend 4.1.4

3.1.4.c Fuel elements No clad defect (fission product release) 4.1.4 Biennial

3.2.1.a Control rods No apparent damage 4.2.1.a Biennial

3.2.1.b Control rods < 1 s scram 4.2.1.b Annually or retest SURV-6 Control Rod Calibration

3.2.1.c Control rods < 0.2% Δk/k s-1 4.2.1.c Annually or retest SURV-6 Control Rod Calibration

3.2.2.b Control rods multiple simultaneous manual rod withdrawal prevented

4.2.2 Semiannually or

retest

3.2.2..c Transient Rod Interlock prevents pulsing unless down 4.2.2 Semiannually or retest

3.2.2.d Standard Control rods Withdrawal in pulse mode prevented 4.2.2 Semiannually or retest

3.2.2..e Transient Rod Pulse prevented if > 1 kW 4.2.2 Semiannually or retest

3.2.3.a Safety System 550°C (2)

4.2.3 Annually

SURV-1 Fuel temperature calibration

4.2.3 Daily channel

check

3.2.3.b Safety System <1.1 MW SS 4.2.3 Annually or retest SURV-2 reactor power calibration

Facility Introduction, Badge Issue, Safety Briefing E1.1-32

4.2.3 Daily

3.2.3.b Safety System <2000 MW P 4.2.3 Annually or retest

3.2.3.e Manual SCRAM Operable 4.2.3 Semiannually or retest

1.3.3 Reactor Pool & Pool Cooling

3.3.1.a Pool cooling 48°C 4.3.1.a Annual 4.3.1.a Monthly 4.3.1.a During operation

3.3.1.b Pool cooling 6.5 m 4.3.1.b Annual 4.3.1.b Monthly 4.3.1.b During operation

3.3.1.c Pool cooling 5.0 µmho <mo>

4.3.1.c Annual

4.3.1.c Monthly SURV-4 Reactor Water Systems

Surveillance 4.3.1.c During operation

3.3.1.d Pool cooling >1 psig HX dp 4.3.1.d Daily & retest 4.3.1.d During operation

3.3.1.e Pool cooling Periodic measurement ph 4.3.1.e Quarterly

SURV-4 Reactor Water Systems

Surveillance

3.3.1.e Pool cooling Periodic measurement α-β 4.3.1.e Quarterly

SURV-…