ML091590734
| ML091590734 | |
| Person / Time | |
|---|---|
| Site: | University of Wisconsin |
| Issue date: | 06/11/2009 |
| From: | Research and Test Reactors Licensing Branch |
| To: | |
| schuster W, NRR/DPR/PRTA 415-3934 | |
| References | |
| TAC MD9592 | |
| Download: ML091590734 (15) | |
Text
TABLE OF CONTENTS 1.0 DEFINITIONS Page 1.1 Reactor Shutdown. . . . . . . . . . . . . . . 1 1.2 Reactor Secured . . . . . . . . . . . . . . . 1 1.3 Reactor Operation . . . . . . . . . . . . . . 1 1.4 Cold Critical . . . . . . . . . . . . . . . . 1 1.5 Steady State Mode . . . . . . . . . . . . . . 1 1.6 Pulse and Square Wave Modes . . . . . . . . . 1 1.7 Shutdown Margin . . . . . . . . . . . . . . . 2 1.8 Abnormal Occurrence . . . . . . . . . . . . . 2 1.9 Experiment. . . . . . . . . . . . . . . . . . 2 1.10 Experimental Facilities . . . . . . . . . . . 3 1.11 Shim-Safety Blade . . . . . . . . . . . . . . 3 1.12 Transient Rod . . . . . . . . . . . . . . . . 3 1.13 Regulating Blade. . . . . . . . . . . . . . . 3 1.14 Fuel Element . . . . . . . . . . . . . . . . 3 1.15 Fuel Bundle . . . . . . . . . . . . . . . . . 3 1.16 Core Lattice Position . . . . . . . . . . . . 3 1.17 Instrumented Element. . . . . . . . . . . . . 4 l 1.18 LEU 30/20 Core. . . . . . . . . . . . . . . . 4 l 1.21 Operational Core. . . . . . . . . . . . . . . 4 1.22 Safety Limits . . . . . . . . . . . . . . . . 4 1.23 Limiting Safety System Settings . . . . . . . 4 1.24 Operable. . . . . . . . . . . . . . . . . . . 5 1.25 Reactor Safety Systems. . . . . . . . . . . . 5 1.26 Experiment Safety Systems . . . . . . . . . . 5 1.27 Measured Value. . . . . . . . . . . . . . . . 5 1.28 Measured Channel. . . . . . . . . . . . . . . 5 1.29 Safety Channel. . . . . . . . . . . . . . . . 5 1.30 Channel Check . . . . . . . . . . . . . . . . 5 1.31 Channel Test. . . . . . . . . . . . . . . . . 5 1.32 Channel Calibration . . . . . . . . . . . . . 6 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits . . . . . . . . . . . . . . . . 7 2.2 Limiting Safety System Setting. . . . . . . . 8 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limitations. . . . . . . . . . . . 10 3.2 Pulse Mode Operation. . . . . . . . . . . . . 10 3.3 Control and Safety System . . . . . . . . . . 11 3.3.1 Scram Time. . . . . . . . . . . . . . . . . . 11 Amendment No. 17
- TS 3 -
1.10 EXPERIMENTAL FACILITIES Experimental facilities shall mean beam ports, including extension tubes with shields, thermal columns with shields, vertical tubes, through tubes, in-core irradiation baskets, irradiation cell, pneumatic transfer systems and in-pool irradiation facilities.
REACTOR COMPONENTS 1.11 SHIM-SAFETY BLADE A shim-safety blade is a control blade having an electric motor drive and scram capabilities. It may have a fueled follower section.
1.12 TRANSIENT ROD The transient rod is a control rod with scram capabilities that can be rapidly ejected from the reactor core to produce a pulse.
It may have a voided follower.
1.13 REGULATING BLADE The regulating blade is a low worth control blade that need not have scram capability and may have a fueled follower. Its position may be varied manually or by the servo-controller.
1.14 FUEL ELEMENT A fuel element is a single TRIGA fuel rod of LEU 30/20 type. l 1.15 FUEL BUNDLE A fuel bundle is a cluster of three or four fuel elements secured in a square array by a top handle and a bottom grid plate adaptor.
1.16 CORE LATTICE POSITION The core lattice position is that region in the core (approximately 3" by 3") over a grid plug hole. It may be occupied by a fuel bundle, an experiment or experimental facility, or a reflector element.
Amendment No. 17
- TS 4 -
1.17 INSTRUMENTED ELEMENT An instrumented element is a special fuel element in which a sheathed chromel-alumel or equivalent thermocouple is embedded in the fuel near the horizontal center plane of the fuel element at a point approximately 0.3 inch from the center of the fuel body.
l 1.18 LEU 30/20 CORE l
l A LEU 30/20 core is an arrangement of TRIGA LEU 30/20 fuel in the l reactor grid plate.
l l 1.21 OPERATIONAL CORE l An operational core is an LEU 30/20 core for which the core parameters of shutdown margin, fuel temperature, power calibration, and maximum allowable reactivity insertion have been determined to satisfy the requirements of the Technical Specifications.
REACTOR INSTRUMENTATION 1.22 SAFETY LIMITS Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
1.23 LIMITING SAFETY SYSTEM SETTINGS Limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions.
l Amendment No. 17
- TS 7 -
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability This specification applies to fuel element temperature and steady-state reactor power level.
Objective The objective is to define the maximum fuel element temperature and reactor power level that can be permitted with confidence that no fuel element cladding failure will result.
Specifications
- a. The temperature in a TRIGA LEU 30/20 fuel element shall not l exceed 1150°C under any conditions of operation.
l
- b. The steady-state reactor power level shall not exceed 1500 l kW under any conditions of operation.
Bases A loss of integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel moderator and the cladding if the fuel temperature exceeds the safety limit.
The pressure is caused by air, fission produce gases, and hydrogen from dissociation of the fuel moderator. The magnitude of this pressure is determined by the fuel moderator temperature and the ratio of hydrogen to zirconium in the alloy.
Amendment No. 17
- TS 8 -
The safety limit for the TRIGA-FLIP fuel element is based on data which indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided the temperature does not exceed 1150°C and the fuel cladding is water cooled (pages 3-1 to 3-23 of GA-9064).
The safety limit for the standard TRIGA fuel is based on data including the large amount of experimental evidence obtained during high performance reactor tests of this fuel. These data indicate that the stress in the cladding (due to hydrogen pressure from the dissociation of zirconium hydride) will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1000°C and the fuel cladding is water cooled (GA-9064, pages 3-1 to 3-23).
It has been shown by experience that operation of TRIGA reactors at a power level of 1500 kW will not result in damage to the fuel. Several reactors of this type have operated successfully l for several years at power levels up to 1500kW. The LEU l Conversion SAR section 4.7.8 shows by analysis that a power level l of 1500 kW corresponds to a peak fuel temperature of 665°C. Thus a Safety Limit on power level of 1500 kW provides an ample margin of safety for operation.
2.2 LIMITING SAFETY SYSTEM SETTING Applicability This specification applies to the scram setting which prevents the safety limit from being reached.
Objective The objective is to prevent the safety limits from being reached.
Specifications l a. The limiting safety system setting for fuel temperature l shall be 400°C as measured in an instrumented fuel element l with a pin power peaking factor between 0.87 and 1.16, or l 500°C as measured in an instrumented fuel element with a pin l power peaking factor of at least 1.16.
l b. The limiting safety system setting for reactor power level shall be 1.25 MW.
Amendment No. 17
- TS 9 -
Bases
- a. The limiting safety system setting is a temperature which, l if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. Analyses l performed in section 4.7.6 of the LEU Conversion Analysis l show that with the IFE in a core location with a pin power l peaking factor of at least 0.87, the maximum fuel l temperature would be no greater than 678°C if the IFE l thermocouple reaches 400°C providing a margin of 472°C to l the safety limit. The same analyses also show that with the l IFE in a core location with a pin power peaking factor of at l least 1.16, the maximum fuel temperature would be no greater l than 678°C if the IFE thermocouple reaches 500°C providing a l margin of 472°C to the safety limit. l l
In the pulse mode of operation, the same limiting safety system setting will apply. However, the temperature channel will have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting of the "tail" of the energy transient in the event the pulse rod remains stuck in the fully withdrawn position.
- b. Analysis in section 4.7 of the Conversion Analysis SAR shows l that at 1.3 MW, the peak fuel temperature in the core will l be approximately 604°C so that the limiting power level l setting provides an ample safety margin to accommodate errors in power level measurement and anticipated operational transients.
Amendment No. 17
- TS 11 -
Specification The reactivity to be inserted for pulse operation shall be determined and mechanically limited such that the reactivity insertion will not exceed 1.4% ) k/k.
Basis The LEU Conversion SAR section 4.7.10 shows by analysis that a l 1.4% ) k/k limitation on pulse reactivity will result in a maximum l fuel temperature of 790°C. This leaves a margin to the 1150°C l Safety Limit of 360°C, and a margin of 40°C to the 830°C l operational limit recommended by General Atomics, Pulsing l Temperature Limit for TRIGA LEU Fuel, GA-C26017 (December, l 2007). l 3.3 CONTROL AND SAFETY SYSTEM 3.3.1 Scram Time Applicability This specification applies to the time required for the scrammable control elements to be fully inserted from the instant that a safety channel variable reaches the Safety System Setting.
Objective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.
Specification The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable control element reaches its fully inserted position shall not exceed 2 seconds.
Basis This specification assures that the reactor will be promptly shut down when a scram signal is initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.
Amendment No. 17
- TS 13 -
TABLE 1 Safety System Or Minimum No. Function & Operating Measuring Channel Operable Mode in Which Required
- a. Fuel Element 1 Scram at 400°C for IFE peaking l Temperature factors 0.87-1.16 or 500°C for l IFE peaking factors >1.16. All l modes.
- b. Reactor Power Level 2 Scram at 125% of full licensed power level; Square Wave &
Steady State Modes.
- c. Manual Pushbutton 1 Scram; All modes.
- d. Reactor Pool-water 1 Scram if water level is less Level than 19 feet above top of core; All modes.
- e. Log N 1 Prevent firing transient rod when drive is not full in and power level is above 1 kW in all modes.
- f. Log Count Rate 1 Prevent control element with-drawal when neutron count rate is less than 2 per second; All modes.
- h. High Voltage Monitor 1 Scram on loss of high voltage supply to neutron and gamma ray power level instrumenta-tion detectors; All modes.
- i. Pulse Mode Control 1 Prevents withdrawal of control Blade Withdrawal blades while in pulse mode.
Interlock.
- j. Reactor Pool-water 1 Scram if water temperature is l Temperature greater than 130°F; All modes. l Amendment No. 17
- TS 14 -
Bases The fuel temperature scram provides the protection to assure that if a condition occurs in which the limiting safety system setting is exceeded, an immediate shutdown will occur to keep the fuel temperature below the safety limit.
The reactor power level scrams are provided in steady state and square wave modes as added protection against abnormally high fuel temperatures and to assure that reactor operation stays within the licensed limits.
The manual scram allows the operator a means of rapid shutdown in the event of unsafe or abnormal conditions.
The reactor pool water level scram assures shutdown of the reactor in the event of a serious leak in the primary system or pool.
The Log N interlock prevents firing of the transient rod at power levels above 1.0 kW if the transient rod drive is not in the full down position. This effectively prevents inadvertent pulses which might cause fuel temperature to exceed the safety limit on fuel temperature.
The Log N interlock does not allow control element withdrawal unless the neutron count rate is high enough to assure proper instrument response during reactor startup.
The preset timer assures reduction of reactor power to a low level after a pulse.
The high voltage monitor prevents operation of the reactor with other systems inoperable due to failure of the detector high voltage supplies.
The pulse mode control blade withdrawal interlock prevents reactivity addition in pulse mode other than by firing the transient rod.
l The thermal-hydraulic analysis in the SAR assumes a pool water l temperature of 130°F. If the temperature exceeds 130°F then the scram l will prevent continued operation in an un-analyzed condition.
Amendment No. 17
- TS 26 -
5.0 DESIGN FEATURES 5.1 REACTOR FUEL Applicability This specification applies to the fuel elements used in the reactor core.
Objective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Specifications l a. TRIGA LEU 30/20 Fuel l The individual unirradiated TRIGA LEU 30/20 fuel elements shall have the following characteristics:
l (1) Uranium content: maximum of 30 Wt-% enriched to maximum of l 19.95 Wt-% with nominal enrichment of 19.75 Wt-% Uranium l 235.
(2) Hydrogen-to-zirconium atom ratio (in the ZrHx ): nominal 1.6 l H atoms to 1.0 Zr atoms with a maximum H to Zr ratio of l 1.65.
(3) Natural erbium content (homogeneously distributed): nominal l 0.9 Wt-%.
(4) Cladding: 304 stainless steel, nominal 0.020 inch thick.
l Amendment No. 17
- TS 27 -
Bases The fuel specification permits a maximum uranium enrichment of 19.95%. l This is about 1% greater than the design value for 19.75% enrichment. l Such an increase in loading would result in an increase in power l density of less than 1%. An increase in local power density of 1% l reduces the safety margin by less than 2% (Texas A&M LEU Conversion l SAR, December 2005). l l
The fuel specification for a single fuel element permits a minimum l erbium content of about 5.6% less than the design value of 0.90 Wt-%. l (However, the quantity of erbium in the full core must not deviate from l the design value by more than -3.3%). This variation for a single fuel l element would result in an increase in fuel element power density of l about 1-2%. Such a small increase in local power density would reduce l the safety margin by less than 2% (Texas A&M LEU Conversion SAR, l December 2005). l The maximum hydrogen-to-zirconium ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than the value resulting from a hydrogen-to-zirconium ratio of 1.60. However, this increase in the clad stress during an accident would not exceed the rupture strength of the clad (M.T. Simnad, The U-ZrHx Alloy: Its Properties and Use in TRIGA Fuel, l General Atomics Report E-117-833, February, 1980). l l
l 5.2 REACTOR CORE Applicability This specification applies to the configuration of fuel and in-core experiments.
Amendment No. 17
- TS 28 -
Objective The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.
Specifications l a. The core shall be an arrangement of TRIGA LEU 30/20 uranium-zirconium hydride fuel-moderator bundles positioned in the reactor grid plate.
l l b. The reactor shall not be operated with a core lattice position vacant except for positions on the periphery of the core assembly.
l c. The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water.
Bases l a. TRIGA cores have been in use for years and their characteristics l are well documented. LEU cores including 30/20 fuel have also l been operated at General Atomics and Texas A&M and their l successful operational characteristics are available. In l addition, the analysis performed at Wisconsin indicates that the l LEU 30/20 core will safely satisfy all operational requirements.
l See chapters 4 and 13 of the LEU Conversion Analysis SAR.
l l b. Vacant core lattice positions will contain experiments or an experimental facility to prevent accidental fuel additions to the reactor core. They will be permitted only on the periphery of the core to prevent power perturbations in regions of high power density.
Amendment No. 17
- TS 29 -
- c. The core will be assembled in the reactor grid plate which is l located in a pool of light water. Water in combination with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation requirements.
5.3 Control Elements Applicability These specifications apply to the control blades and transient control rod.
Objective The objective is to assure that control elements are fabricated to reliably perform their intended control and safety function.
- a. The safety blades shall be constructed of boral plate and shall have scram capability.
- b. The regulating blade shall be constructed of stainless steel.
- c. The transient rod shall contain borated graphite or boron and its compounds in a solid form as a poison in an aluminum or stainless steel clad. The transient control rod shall have scram capability and may incorporate an aluminum or air follower.
Bases The boral safety blades and stainless steel regulating blade used in the reactor have been shown to provide adequate reactivity worth, structural rigidity, and reliability to assure reliable operation and long life under operating conditions. The transient control rod materials and fabrication techniques have been used in many TRIGA reactors and have demonstrated reliable operation and long life.
5.4 Radiation Monitoring Systems Applicability These specifications describe the functional performance and essential components of the radiation monitoring systems.
Objective The objective is to describe those systems which provide information on radiation levels and effluent radioactivity.
Specifications
- a. The area radiation monitoring system shall provide gamma radiation level information at the control console for at least three locations in the Laboratory. It shall cause an alarm at the control console and initiation of an evacuation alarm if high radiation levels occur and prompt remedial action is not taken.
Amendment No. 17
- TS 31 -
5.6 Reactor Building Applicability These specifications apply to the room housing the reactor and the ventilation system controlling that room.
Objective The objective is to provide restrictions on release of airborne radioactive materials to the environs.
Specifications
- a. The reactor shall be housed in a closed room designed to restrict leakage. the minimum free volume shall be 2,000 cubic meters.
- b. All air or other gas exhausted from the reactor room and associated experimental facilities shall be released to the environment a minimum of 30.5 meters above ground level. l Bases Calculations in Chapter 13 of the SAR demonstrate that the occupational l doses in the event of the maximum hypothetical accident do not exceed l limits if the lab volume is at least 2000m3 . Furthermore, calculations l in chapter 13 that assume operation of the ventilation system assume a l stack height of 30.5m. l 5.7 REACTOR POOL WATER SYSTEMS Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.
Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.
Amendment No. 17
- TS 32 -
Specifications
- a. The reactor core shall be cooled by natural convective water flow.
- b. The pool water inlet and outlet pipe to the demineralizer shall not extend more than 15 feet into the top of the reactor pool when fuel is in the core. The outlet pipe from the demineralizer shall be equipped with a check valve to prevent inadvertent draining of the pool.
- c. Diffuser and other auxiliary systems pumps shall be located no more than 15 feet below the top of the reactor pool.
- d. All other piping and pneumatic tube systems entering the pool shall have siphon breakers and valves or blind flanges which will prevent draining more than 15 feet of water from the pool.
- e. A pool level alarm shall indicate loss of coolant if the pool level drops approximately one foot below normal level.
l f. A pool water temperature alarm shall indicate if water temperature l reaches 130°F.
Bases l a. The LEU Conversion SAR section 4.7.8 shows by analysis that the l natural convective cooling of the reactor core is sufficient to l maintain the fuel in a safe condition up to at least a power level l of 1500 kW (the power Safety Limit).
- b. The inlet pipe to the demineralizer is positioned so that a siphon action will drain less than 15 feet of water. The outlet pipe from the demineralizer penetrates the pool below core level and a check valve prevents leakage from the pool by reverse flow from pipe ruptures or improper operation of the demineralizer valve manifold.
- c. In the event of pipe failure and siphoning of pool water, the pool water level will drop no more than 15 feet from the top of the pool.
- d. Other pipes which enter the pool have siphon breakers which prevent pool drainage. Valves are provided for pneumatic tube system lines and primary cooling system pipe. Other piping installed in the pool has blind flanges permanently installed.
- e. Loss of coolant alarm, after one foot of loss, requires corrective action. This alarm is observed in the reactor control room and outside the reactor building.
l f. The thermal-hydraulic analysis in the SAR assumes a pool water l temperature of 130°F. If the temperature exceeds 130°F then the l alarm will prevent continued operation in an un-analyzed l condition.
Amendment No. 17