ML20153D265
| ML20153D265 | |
| Person / Time | |
|---|---|
| Site: | University of California-Davis |
| Issue date: | 08/13/1998 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20153D253 | List: |
| References | |
| R-130, NUDOCS 9809250006 | |
| Download: ML20153D265 (50) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 4
9 . . . . . ,o FACILITY OPERATING LICENSE DOCKET NO. 50-607 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for license, filed by the Department of the Air Force at McClellan Air Force Base, on October 23,1996, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. Construction of the facility was completed in substantial conformity with the provisions of the Act, and the rules and regulations of the Commission; C. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance (i) that the activities authorized by this license can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's regulations; E. The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accordance with the regulatione of the Commission; F. The licensee is a Federal agency and will use the facility for defense programs and research. The licensee, in accordance with 10 CFR Part 140, " Financial Protectior ,uirements and Indemnity Agreements," is not required to furnit r- f of financial protection.
The licensee has executed en indemnt . ,.eement that satisfies the requirements of 10 CFR Part 140 of the Commission's regulations; i
9809250006 980813 PDR ADOCK 05000607 P PDR
2 G. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; H. The issuance of this license is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been l satisfied; and I. The receipt, possession, and use of the byproduct and special nuclear materials as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30 and 70, including Sections 30.33, 70.23, and 70.31.
- 2. Facility Operating License No. R-130 is hereby issued to the Department of l
l the Air Force at McClellan Air Force Base as follows: l A. The license applies to the training reactor and isotopes production, General Atomics (TRIGA) nuclear reactor (the facility) owned by the Department of the Air Force at McClellan Air Force Base (the licensee).
The facility is located on the licensee's site at McClellan Air Force Base and is described in the licensee's application for license of October 23, 1996, as supplemented.
B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the Department of the Air Force at l McC'ellan Air Force Base:
l (1) Pursuant to Section 104c of the Act and 10 CFR Part 50, l
" Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at McClellan Air Force Base, in accordance with the procedures and limitations set forth in this license.
(2) Pursuant to the Act and 10 CFR Part 70, "Do.nestic Licensing of Special Nuclear Material," to receive, possess, and use up to 21.0 kilograms of contained uranium-235 enriched to less than 20 percent in the isctope uranium-235 in the form of reactor fuel; up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of plates; and to possess, but not separate, such special nuclear material as may be produced by the operation of the facility.
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(3) Pursuant to the Act and 10 CFR Part 30, " Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive, possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a 55-millicurie sealed cesium-137 source for instrument calibrations; small instrument calibration and check sources of less than 0.1 millicurie each; and to possess, use, but not separate, except for byproduct material produced in reactor experiments, such byproduct material as may be produced by l the operation of the facility.
C. This license shall be deemed to contain and is subject to the conditions specified in Parts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I; to all applicable provisions of the Act; and to the rules, regulations, and orders of the Commission now or hereafter in effect and to the additional conditions specified below:
(1) Maximum Power Level l The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and in the pulse mode with reactivity insertions not to exceed $1.75 (1.23 %Ak/k).
l (2) Technical Soecifications i The Technical Specifications contained in Appendix A are hereby incorporated in the license. The licensee shall operate j the facility in accordance with the Technical Specifications.
l (3) Physical Security Plan l
l The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security plan, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The approved plan, which is exempt from public disclosure pursuant to the provisions of 10 CFR 2.790, is entitled " Physical Security Plan for the McClellan Nuclear Radiation Center (MNRC) TRIGA l Reactor Facility," Revision 3, dated August 1996.
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D. This license is effective as of the date of issuance and shall expire i twenty (20) years from its date of issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION )
i a or Office of Nuclear Reactor Regulation 1
Enclosure:
Appendix A Technical I Specifications l
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Date of issuance: August 13, 1998 l 1
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TECHNICAL SPECIFICATIONS FOR THE j MCCLELLAN NUCLEAR RADIATION CENTER (MNRC) l REACTOR FACILITY DOCUMENT NUMBER: MNRC-0004-DOC 1
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.t TECHNICAL BPECIFICATION APPROVAL This " Technical Specification for the McClellan Nuclear Radiation Center'(MNRC) Reactor has undergo following coordination. !
s Reviewed by:-
Health Physics Supervisor (Date)
Reviewed by:
Operations Supervisor - (Date)
Approved by:
Chief, Nuclear Licensing and Operations (Date) ,
l Approved by:
Chairman, SM-ALC Nuclear Safety Committee (Date)
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TECHNICAL SPECIFICATIONS 1
TABLE OF CONTENTS 1.0 Definitions 1
2.0 Safety Limits and Limiting Safety System Setting (LSSS) 6 2.1 Safety Limits 6
2.2 Limiting Safety System Setting (LSSS) 6 2.2.1 Fuel Temperature 6
3.0 Limiting Conditions for Operations (LCOS) 7 3.1 Reactor Core Parameters 7 3.1.1 Steady-State Operation 7
3.1.2 Pulse or Square Wave Operation 7 3.1.3 Reactivity Limitations 8
3.2 Reactor Control and Safety Systems 8 3.2.1 Control Rods 8
3.2.2 Reactor instrumentation 9 3.2.3 Reactor Scrams and interlocks 10 3.2.4 Reactor Fuel Elements 12 3.3 Reactor Coolant Systems 13 3.4 Reactor Room Exhaust System 14 3.5 Intentionally Left Blank 15 3.6 Intentionally Left Blank 15 3.7 Reactor Radiation Monitoring Systems 15 3.7.1 Monitoring Systems 15 i
3.7.2 Effluents - Argon-41 Discharge Limit 16 3.8 Experiments 16 3.8.1 Reactivity Limits 16 3.8.2 Materials Limit 17 l
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18 3.8.3 Failure and Malfunctions 19 4.0 Surveillance Requirements 19 4.1 Reactor Core Parameters 19 4.1.1 Steady State Operation 19 4.1.2 Shutdown Margin and Excess Reactivity 20
( 2 Reactor Control and Safety Systems 20 4.2.1 Control Rods 21 4.2.2 Reactor Instrumentation 22 4.2.3 Reactor Scrams and Interlocks 23 4.2.4 Reactor Fuel Elements 24 4.3 Reactor Coolant Systems 25 4.4 Reactor Room Exhaust System 25 4.5 Intentionally Left Blank 25 4.6 Intentionally Left Blank 25 4.7 Reactor Radiation Monitoring Systems 26 4.8 Experiments 27 5.0 Design Features 27 5.1 Site and Facility Description 27 5.1.1 Site 28 5.1.2 Facility Exhaust 28 5.2 Reactor Coolant System 29 5.3 Reactor Core and Fuel 29 5.3.1 Reactor Core 29 5.3.2 Reactor Fuel 30 5.3.3 Control Rods and Control Rod Drives 31 5.4 Fissionable Material Storage
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6.0 Administrative Controls 31 6.1 Organization 31 6.1.1 Structure 31 6.1.2 Responsibilities .
31 6.1.3 Staffing 31 6.1.4 Selection and Training of Personn3l 32 6.2 Review and Inspection 32 6.2.1 NSC Composition and Qualifications 32 l
6.2.2 NSC Charter and Rdes 32 i 6.2.3 Review Function 32 6.2.4 Inspection Function 33 6.3 Radiation Safety 33 6.4 Procedures 33 6.4.1 Reactor Operations Procedures 33 6.4.2 Health Physics Procedures 34 6.5 Experiment Review and Approval 34 6.6 Required Actions 35 6.6.1 Actions to be taken in case of a safety limit 35 violation 6.6.2 Actions to be taken for a reportable occurrence 35 6.7 Reports 35 j 6.7.1 Operating Reports 35 6.7.2 Special Reports 37 g
[ 6.8 Records 38 l Fig 6.1 McClellan AFB Nuclear Operations Organization l.
l Fig 6.2 McClellan Nuclear Radiation Center (MNRC) Organization Fig 6.3 Nuclear Safety and Licensing Organization
TECHNICAL SPECIFICATIONS FOR THE MCCLELLAN NUCLEAR RADIATION CENTER (MNRC)
REACTOR FACILITY General The McClellan Nuclear Radiation Center (MNRC) reactoris operated by the United States Air Force at McClellan Air Force Base, Sacramento CA. The MNRC research reactor is a TRIGA type reactor. The MNRC provides a hig sensitivity inspection capability for detection of early stage corrosion in aluminum aircraft components, thereby reducing aircraft crash risk and reducing repair cost. In addition, the MNRC provides a wide range of irradiation services for both military and non military tasks. The reactoroperates at a nominal steady state power level up to and including 2 MW. The MNRC reactoris also capable of square wave and pulse operationalmodes. The MNRC reactor fuelis less than 20% enriched in uranium-235.
1.0 Definitions 1.1 As low As Reasonably Achievable (ALARA1 The term "as low as reasonably achievable"when applied to radiation exposures and releases of radioactivematerials in effluents means as low as reasonably achievab into account the state of technology, and the economics of improvements in relation to the benefits to the public hea undpublic the safety and other societaland socioeconomicconsiderations,and in relation to the ut;lization of nuclear energ interest.
1.2 Licensed Ooerators. A MNRC reactor operator is an individual licensed by the Nuclear Regulatory Commission (e.g., senior reactoroperator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license.
1.2.1 Senior Reactor Ooerator. An individual who is licensed to direct the activities of reactor operators and to manipulate the controls of the facility.
1.2.2 Reactor r merator An individual who is licensed to manipulate the controls of the facility and perform reactor- related maintenance.
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l 1.3 Channel A channelis the combination of sensor, line, amplifier, processor, and output devices which are connected for the purpose of measuring the value of a parameter.
1.3.1 ChannelTest. A channeltest is the introductionof a signalinto the channel for verification that it is operable.
1.3.2 ChannelCalibratiort A channelcalibrationis an adjustme1t of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip and shall be deemed to include a channel test.
1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification,where possible, shallinclude comparison of the channel with other independent channels or systems measuring the same variable.
1.4 Confinement Confinementmeans isolation of the reactor room air volume such that the move into and out of the reactor room is through a controlled path.
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1.5 Exoeriment Any operation, hardware, or target (excluding devices such as det:ctors, fiss foils, etc), which is designed to investigate specific reactor characteristics or which is in reactor tank, or in a beamport or experiment facility and which is not rigidly secured to a core o as to be a part of their design.
1.5.1 Eperiment. Moveable. A moveable experiment is one where it is intended that th experiment may be moved in or near the reactor core or into and out of the reactor while 1.5.2 Exoeriment. Secured. A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by me restraining force must be substantially greater than those to which the experiment might be s pneumatic, buoyant, or other forces which are normal to the operating environment of the which can arise as a result of credible conditions.
1.5.3 Exoeriment Facilities. Experiment facilities shall mean the pneumatic transfer tube, central thimble, beamtubes, irradiation facilities in the core or in the reactor tank, and radiography bays.
j 1.5.4 Exoeriment Safety System. Experiment safety systems are those systems, including their I associatedinputcircuits, which are designed to initiate a scram for the primary purpose of protecting an ex or to provide information which requires manual protective action to be initiated.
i 1.6 Fuel Element. Standard. A fuel element is a single TRIGA element. The fuelis U-ZrH clad in stainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt %) of ;
uranium can be either 8.5,20 or 30 wt %, with an enrichment of less than 20% U-235. A standard fuel element ma contain a bumable poison. l 1.7 Fuel Element. Instrumented. An instrumented fuel element is a standard fuel eleme thermocouples for temperature measurements. An instrumented element shall have at least one operable thermocouple embedded in the fuel near the axial and radial midpoints.
1.8 Measured Value. The measured value is the value of a parameter as it appears on the output of a channel.
1.9 Mode. Steady-State. Steady-state mode operation shall mean operation of the MNRC reactor with the selector switch in the automatic or manual mode position.
1.10 Mode. Sauare-Wave Square-wavemode operation shall mean operation of the MNRC reactor with the selector switch in the square-wave mode position.
1.11 Mode. Pulse. Pulse mode operation shall mean operation of the MNRC reactor with the selector switch in the pulse mode position.
1.12 Ooerable. Operable means a component or system is capable of performing its intended function.
1.13 Ooeratina. Operating means a component or system is performing its intended function.
1.14 Ooeratina Cvcie. The period of time starting with reactor startup and ending with reactor shutdown.
1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment within the MNRC reactor safety system in response to a variable or condition of the MNRC reactor facility having reached a specified limit.
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l 1.15.1 Channel Level. At the protectiveinstrument channel level, protective action is the generat and transmission of a scram signalindicating that a reactor variable has reached the specified limit.
1.15.2 Subsystem level. At the protective instrument subsystem level, protective action is the generation and transmission of a scram signalindicating that a specified limit has been reached.
NOTE; Protective action at this levelwould lead to the operation of the safety shutdown equipment.
1.15.3 Instrument System level. At the protective instrument system level, protective action is the l
generation and transmission of the command signal for the safety shutdown equipment to operate.
1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.
1.16 Pulse Ooerational Core. A pulse operational core is a reactor state operational core for which the i maximum allowable pulse reactivity insertion has been determined.
1.17 Reactivity. Excess. Excess reactivity is that amount of reactivity that would exist if all control rods (control, regulating, etc.) were moved to the maximum reactive condition from the point where the reactor is at am temperature and the reactor is critical. (ks:1) 1.18 Reactivity Limits. The reactivitylimits are those limits imposed on the reactivity conditions of the reactor Core.
l 1.19 ReactivityWorth of an Exoeriment. The reactivity worth of an experiment is the maximum value of the reactivity change that could occur as a result of changes that alter experiment position or configuration.
1.20 Reactor Controls Reactor controls are apparatus and/or mechanisms the manipulation of which dir affect the reactivity or power level of the reactor.
1.21 Reactor Core. Ooerational. The MNRC reactor operational core is a core for which the parameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worths of control rods and experiments have been determined to satisfy the requirements set forth in these Technical Specifications.
1.22 Reactor Ooeratina. The MNRC reactor is operating whenever it is not secured or not shut down.
1.23 Reactor Safety Systems. Reactor safety systems are those systems, I Jding their associated input channels,which are designed to initiate automatic reactor protection or to provide info.mation for initiation of manual protective action.
1.24 Reactor Secured The MNRC reactoris secured when the console key switch is in the off position and the key is removed from the lock and under the control of a licensed operator, and the conditions of 'a' or 'b' exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required l by technical specifications; and
- a. (2) No work is in progress involving core fuel, core structure, installed control rods, or control rod drives, unless the control rod drives are physically decoupled from the control rods; and 3
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- a. (3) No experimentsin or near the reactor are being moved or serviced that have on movemert, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or 1
- b. The reactor contains insuffident fissile materials in the reactor core, adjacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 Reactor Shutdown The MNRC reactoris shutdown if it is subcritical by at least one dollar ($1.00) both l in the Reference Core Condition and for all allowed ambient conditions with the reactivity worth of all installed l experiments included. l l
1.26 Reference Core Condition The condition of the core when it is at ambient temperature (cold T<28'C), l the reactivity worth of xenon is negligible (< $0.30)(i.e., cold, clean, and critical), and the central irradiation facility I contains the graphite thimble plug and the aluminum thimble plug. f 1
1.27 Research Reactor. A research reactor is defined as a device designed to support a self-sustaining )
l neutron chain reaction for research development, education, and training, or experimental purposes, and which may l have provisions for the production of radioisotopes.
1.28 Rod. Control A control rod is a device fabricated from neutron absorbing material, with or without a fuel or air follower,which is used to establish neutron flux changes and to compensate for routine reactivity losses. The l
follower may be a stainless steel section. A control rod shall be coupled to its drive unit to allow it to perform its control l
function, and its safety function when the coupling is disengaged. This safety function is commonly termed a scram.
1.28.1 Reaulatina Rod. A regulating rod is a control rod used to maintain an intended power level and may be varied manually or by a servo-controller. A regulating rod shall have scram capability. l 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.
i 1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapid reactivity l insertion to produce a pulse or square wave. )
1,29 Safety Channel. A safety channelis a measuring channelin the reactor safety system. I 1.30 Safety Limit. Safety limits are limits on important process variables, which are found to be necessary
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to reasonably protect the integrity of the principal barriers which guard against the uncontrolled release of radioactivity.
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1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the control rods being fully inserted.
1.32 Scram. External. The extemal scrams consist of those shutdown signals that do not originate from the reactor control system.
1 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; the word "may" to denote permission, neither a requirement nor a recommendation.
1.34 Shutdown Marain. Shutdown margin shall mean the minimum shutdown reactivty necessary to provide j confidence that the reactor can be made subcritical by means of the. control and safety system starting from any permissibleoperating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator action.
1.35 Shutdown. Unscheduled An unscheduled shutdown is any unplanned shutdown of the MNRC reactor i
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caused by actuation of the reactor safety syt em, operator error, equipment malfunction, or a manual shutdown in response to conditionswnich could adversely arfect safe operation, not including shutdowns which occur during testing or check-out operations.
1.36 Surveilance Activities. In general, two types of surveillance activities are specified: operability checks and tests, and calibrations. Operability checks and tests are generally specified as daily, weekly or quarterly.
Calibration times are generally specified as quarterly, semi-annually, annually, or biennially.
1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and not to reduce frequency. Established frequencies shall be maintained over the long term. The allowable surveillanceinterval is the interval between a check, test, or calibration, whichever is appropriate to the item being subjected to the surveillance,and is measured from the date of the last surveillance. Allowable surveitance intervals shall not exceed the following:
1.37.1 Annual-interval not to exceed fifteen (15) months.
l 1.37.2 Semiannuel-interval not to exceed seven and a half (7.5) months.
1.37.3 Quarterly - interval not to exceed four (4) months.
1.37.4 Monthly - interval not to exceed six (6) weeks.
1.37.5 Weth!y - interval not to exceed ten (10) days.
1.38 Unrev ewed Safety Questions. A proposed change, test or experiment shall be deemed to involve an unreviewed safety question:
- a. If the probability of occurrenceor the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- b. If a possibility for an accidentor malfunctionof a different type than any evaluated previously in the safety analysis report may be created; or
- c. If the margin of safety as defined in the basis for any technical specification is reduced.
1.39 Value. Measured. The measured value is the value of a parameter as it appears on the output of a channel.
1.40 Value. True. The true value is the actual value of a parameter.
1.41 Watchdoo Circuit. Tie watchdog circuit is a surveillance circuit provided by the Data Acquisition Computer (DAC) and the Control System Computer (CSC) to ensure proper operation of the reactor computerized control system.
1.42 Loss of Coolant Accident (LOCA). The loss of coolant accident (LOCA) assumes the complete loss of coolant to the reactor core. The LOCA assumes (SAB Chapter 13) that the reactor core becomes uncovered instantaneously after extended operation at a power level of two (2) Megawatts.
1.43 Emeraency Core Coolina System (FCCS). The emergency core cooling system (ECCS) provides a domestic source of water to cool the reactor core in the event of a loss of coolant accident (LOCA). The emergency 5
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core cooling system assures that the fuel temperature safety limit will not be exceeded durin 2.0 Safety Limits and Limitina Safety Systen; Settina.
2.1 Safety Limits.
Acolicability - This specification applies to the temperature of the reactor fuel in a standard T element.
Obiective - The objective is to define the maximum temperature that can be permitted with con no damage to the fuel element cladding will result.
Soecification -
- a. The maximum fuel temperature in a standard TRIGA fuel element shall not exceed 930*C du state operation.
- b. The maximum temperature in a standard TRIGA fuel element shall not exceed 1100*C during operation.
BASit -
- a. This fuel safety limit applies for conditions in which the cladding temperature is above 500*C (S Analysis Report (SAR), Section 4.5.4.1.3). The important parameter for a TRIGA reactor is the temperature. This parameter is well suited as it can be measured directly. A loss in the integrity o cladding could arise if the cladding stress exceeds the ultimate strength of the cladding material. T cladding stress is a function of the element's internal pressure while the ultimate strength of the c a function of its temperature. The cladding stress is a result of the internal pressure due to the presence product gases and hydrogen from the disassociation of hydrogen and zirconium in the fuel modera pressure is the most significant. The magnitude of the pressure is determined by the fuel, the cladding fueldue stress moder the ratio of hydrogen to zirconiumin the alloy. At a fuel temperature of 930'C for ZrH,7 to the internal pressure is equal to the ultimate strength of the cladding material at the same temperatu 4.18). This is a conservative limit since the temperature of the cladding material is always lower tha temperature. (See Chapter 4 of the Safety Analysis Report.)
- b. This fuel safety limit applies for conditions in which the cladding temperature is less than 500*C (SA Chapter 4, Section 4.5.4.1.3).This analysis shows that a maximum temperaturefor the clad during a pu a peak adiabatic fuel temperature of 1000'C is conservatively estimated to be 470*C. SAR, figure 4.1 the ultimate strength of the clad at a temperature of 470*C is 59,000 psi. Therefore, if the stress produced b hydrogen overpressure inside the fuel element is less than 59,000 psi, the fuel element will not und cladding integrity. Referring to Figure 4.18 and consideringf UZrH,7 uel with a peak temperature of 1000'C, t on the clad is found to be 24,000 psi. Therefore, a fuel temperaturelimit of 1100*C is a limit where no loss of claddi integritywilloccur. Operationaldata shows that GA fuel with a hydrogen to zircenium ratio of at least 1.65 pulsed to temperatures of about 1150'C without damage to the clad. (SAR, ref 4.21.)
2.2 Limitina Safety System Settina.
2.2.1 Fuel Temoerature.
t Aeolicability - This specification applies to the protective action for the reactor fuel element temperature.
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Obiective-The objectiveis to prevent the fuel element temperature safety limit from being reached.
1 Soecification- The limiting safety system setting shall be 750'C (operationally this may be set more conservatively) as measured in an instrumented fuel element. One instrumented element shall be located in the analyzed peak power location of the reactor operational core.
B.as.is - For steady-state operation of the reactor, the limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 750'C provides a safety margin at the point of measurement of at least 137'C for standard TRIGA fuel elementsin any condition of operation. A part of the safety margin is used to account for the difference between the true and measured temperaturesresulting from the actuallocation of the thermocouple. If the thermocouple element is located in the hottest position in the core, the differer.ce between the true and measured temperatures will be on a few degrees since the thermocouple junction is near the center and the mid-plane of the fuel element.
l For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperaturechannel
! will have no effect on limiting the peak power 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 limit the energy release after the pulse if the transient rod should not reinsert and the fuel temperature continues to increase.
i 3.0 Limitina Conditions For Ooeration i 3.1 Reactor Core Parameters 3.1.1 Steadv-State Ooeration Acolicabilitv- This specification applies to the maximum reactor power attained during steady-state ;
operation. '
Objective - The objective is to assure that the reactor safety limit (fuel temperature) is not exceeded, and to provide for a setpoint for the high flu,v limiting safety systems, so that automatic protective action will prevent the safety limit from being reached during steady-state operations.
Soecification- The nominai reactor steady-state power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. For the purpose of testing the reactor steady-state power level scram, the power shall not exceed 2.3 MW.
Basis - Operationalexperience and thermal-hydraulic calculations demonstrate that MNRC TRIGA fuel elements may be safely operated at power levels up to 2.3 MW with natural convection cooling. (Ref SAR Section 4.6.2.)
l 3.1.2 Pulse or Sauare Wave Ooeration Aeolicability - This specification applies to the peak temperature generated in the fuel as the result of a step insertion of reactivity.
Obiective - The objective is to assure that the fuel temperature safety limit will not be exceeded.
Soecification- (a) For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23%
Ak/k ($1.75);(b) For the square wave mode of operation the maximum step insertion of reactivity shall be 0.63% Ak/k
($0.90).
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Basisi - Standard TRIGA Fuelis fabricated with a nominal hydrogen to zirconium ratio of 1.6 to 1.7.
This yields delta phase zirconium hydride which has a high creep strength and undergoes no phase changes at temperaturesin excess of 100*C. However, after extensim steady state operation at two (2) MW the hydrogen wi redistributedue to migration from the central high temperature regions of the fuel to the cooler outer regions. When the fuelis pulsed, the instantaneoustemperature distribution is such that the highest values occur at the radial edge of the fuel. The higher temperaturesin the outer regions occur in fuelwith a hydrogen to zirconium ratio that has now increased above the nominalvalue. This produces hydrogen gas pressures considerably in excess of that expected.
If the pulse insertionis such that the temperatureof the fuel exceeds about 875"C, then the pressure may be sufficie to cause expansion of microscopic holes in the fuel that grow with each pulse. The analysis in SAR, section 13.2.2.21, shows that the limiting pulse, for the worst case conditions, is 1.34% Ak/k ($1.92). Therefore, the 1.23%
Ak/k ($1.75) limit is below the worst case reactivityinsertion accidentlimit. The $0.90 square wave step insertion limit is also well below the worst case reactivity insertion accident limit.
3.1.3 Reactivity Limitations Acolicability - These specifications apply to the reactivity conditions of the reactor core and the reactivity worths of the control rods and apply to all modes of reactor operation.
Objective - The objectNe is to assure that the reactor can be placed in a shutdown condition at all times and to assure that the safety limit shall not be exceeded.
Soecifications -
- a. Shutdown Marain - The reactor shall not be operated unless the shutdown margin provided by the control rods is greater than 0.35% Ak/k ($0.50) with:
(1) The reactor in any core condition, (2) The most reactive control rod assumed fully withdrawn, (3) Absolute value of all movable experiments analyzed in their most reactive condition or $1.00 whichever is greater.
- b. Excess Reactivity- The maximum available excess reactivity (reference core condition) shall not exceed 6.65% Ak/k ($9.50).
Baill-
- a. This specification assures that the reactor can be placed in a shutdown condition from any operating condition and remain shutdown, even if the maximum worth control rod should stick in the fully withdrawn position (SAR Section 4.5.5).
- b. This specification sets an overall reactivity limit which provides adequate excess reactivity to override the xenon buildup, to overcome the temperature change in going from zero power to 2 MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments and account for fuel bum up over time. An adequate shutdown margin exists with an excess of $9.50 for the two analyzed cores: (SAR Section 4.5.5).
3.2 Reactor Control and Safety Systems 3.2.1 Control Rods 8
I 1
Aoolicability - This specification applies to the function of the control rods.
Obiective - The objective is to determine that the control rods are operable.
Soecifications - The reactor shall not be operated unless the control rods are operable and,
- a. Control rods shall not be considered operableif damage is apparent to the rod or drive assemblies
- b. The scram time measured from the instant a signal reaches the value of a limiting safety system setting to the instant that the slowest control rod reaches its fully inserted position shall not exceed one (1) second Basis -
- a. The apparent condition of the control rod assemblies shall provide assurance that the rods shall continue to perform reliably as designed.
- b. This assures that the reactor shall shutdown promptly when a scram signalis initiated (see SAR Chapter 13).
3.2.2 Reactor Instrumentation Acolicability - This specification applies to the information which shall be available to the reactor operator during reactor operations.
Obiective - The objectiveis to require that sufficientinformationis available to the operator to assure safe operation of the reactor.
i Soecification(s)- The reactor shall not be operated unless the channels described in Table 3.2.2 are j' operable and the information is displayed on the reactor console.
l Table 3.2.2 Reauired Reactor Instrumentation (Minimum Number Operable)
Measuring Steady Square Channel Surveillance Channel State Pulse Wave Function Reauired*
- b. Linear Power 1 0 1 Automatic D1,M,A1 Channel Power Control l c. Log Power 1 0 1 Startup D1,M A1 Channel Control
- d. Fuel Temperature 2 2 2 Fuel D1,M,A1 Channel Temperature
, 9 e
l 0 1 0 Measures M.A1
- e. Pulse Channel Pulse NV f
& NVT l
D1 - Channel check during each day's operation
(*) Where:
M1 - Channel test' monthly A1 - Channel calibration annually )
Basis -
Soecification (a) Table 3.2.2. The two reactor power level safety channels assure that the reactor power level is properly monitored and indicated in the reactor control room (SAR Sections 7.1.2 & 7.1.2.2).
l Soecification(b. c. & e) Table 3.2.2. The linear channel, log channel, and pulse channel assures that the reactor power level and energy are adequately monitored (SAR Sections 7.1.2 & 7.1.2.2).
Soecification (d) Table 3.2.2. The fuel temperature channels assure that the fuel temperature is properly monitored and indicated in the reactor control room (SAR Section 4.5.4.1). l l
t 3.2.3 Reactor Scrams and interlocks Acolicabihtv - This specification applies to the scrams and interlocks.
Objective - The objective is to assure that the reactor is placed in the shutdown condition promptly and that the scrams and interlocks are operable for safe operation of the reactor.
Soecifications-The reactor shall not be operated unless the scrams and interlocks described in Table 3.2.3 are operable:
Table 3.2.3 Reauired Scrams and interlocks Steady Square Channel Surveillance Function Reauirements' Scram State Pulse Wave
- a. Console 1 1 1 Manual Scram M Manual and Automatic Scram Scram Alarm Manual Scram M
- b. Reactor Room 1 1 1 Manual Scram and Automatic Scram Alarm
- c. Radiography 4 4 4 Manual Scram M Bay Manual and Automatic Scrams Scram Alarm
n q l
l
'l I
y e. High Voltage 2 :1 2. Automatic Scram
~
M Power Supplies Alarm & Scram on L Scram Loss of High Voltage to the Reactor Power Level Safety Channels
- f. Fuel 2 2 2 . Automatic Scram M' Temperature . Alarms & Scrams
- Scram on indicated fuel -
temperature of .
750*C orless 1
- h. Extemal 2 2 2 Automatic Scram M Scram ' and Alarm if i experimental or radiography scram interlocks are activated I. One Kilowatt 0 1 1 Prevents initiation of a M
' Pulse & - a step reactivity insertion Square Wave above a reactor power - -;
interlock . levelof 1 KW ;
'j. Low Source 1 1 1 Prevents withdrawalof M
- Level Rod any control rod if the log Withdrawal channel reads less than Prohibit 1.5 times the indicated Interlock log channel current level with the neutron source removed from the core
- k. Control Rod 1 1 1 Prevents simultaneous M Withdrawal withdrawalof twoormore Interlock rods in manual mode
- 1. Magnet 1 1 1 Deenergizes the M Power Key control rod magnets Switch Scram
(*) Where: M - channel test monthly Bann -
Ssecir,ceGon(a) Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the control room (SAR Section 7.1.2.5).
11 n t o = m ,e ' , - - , - -
Soecification(b) Table 3.2.3. The reactor room manual scram allows rapid shutdown of the reactor from the reactor room.
Soecification (c) Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor from any of the radiography bays (SAR Section 9.6.3).
Soecification (d) Table 3.2.3. The automatic power level safety scram assures the reactorwill be shutdown if the power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Section 4.7.2).
Soecification fe) Table 3.2.3. The loss-of-high voltage scram assures that the reactor power level safety channels operate within their intended range as required for proper functioning of the power level scrams (SAR Sections 7.1.2.1 & 7.1.2.2).
Soecification (f) Table 3.2 3. The fuel temperature scrams assure that the reactor will be shutdown if the fuel temperature exceeds 750'C, therefore er'suring the safety limit will not be exceeded (SAR Sections 4.5.4.1 & 4.7.2).
Soecification (a) Table 3.2.3. The watchdog circuits assure that the control system computer I
and the data acquisition computer are functioning properly (SAR Section 7.2).
l l
Soecification(h) Table 3.2.3. The extemal scrams assure that % reactor will be shutdown
! if the radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry into the bays (SAR Section 9.6). Experimentalscrams, a subset of the extemal scrams, also assures the integrity of the reactor system, the experiment, the facility, and the safety of the facility personnel and the public.
Soecification(i) Table 3.2.3. The interlock preventing the initiation of a step reactivity insertion l
l at a level above one (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature to exceed the safety limit (SAR Section 7.1.2.5).
Soecifcation (i) Table 3 2.3. The low source level rod withdrawal prohibit interlock ensures an adequate source of neutrons is present for safe startup of the reactor (SAR Section 7.1.2.5).
Soecification f k) Table 3.2.3. The control-rod withdrawalinterlock prevents the simultaneous withdrawalof two or more control rods thus limiting the reactivity-insertion rate from the control rods in manual mode (SAR Section 7.1.2.5).
Soecification (I) Table 3.2.3. The magnet current key switch prevents the control rods from being energized without inserting the key. Turning off the magnet current key switch de-energizes the control rod magnets and results in a scram (SAR Section 7.1.2.5).
3.2.4 Reactor Fuel Elements Acolicability-This specificationapplies to the physicaldimensionsof the fuel elements as measured on the last surveillance test.
Obiective - The objective is to verify the integrity of the fuel-element cladding.
Soecification- The reactor shall not be used for normal operation with damaged fuel. All fuel elements shall be inspected visualy for damage or deterioration as per Technical Specifications Section 4.2.4. A fuel element shall be considered damaged and must be removed from the core if:
12
1 1
- a. In measuring the transverse bend, the bend exceeds l
0.125 inch (3.175 mm) over the full length 23 inches (584 mm) of the cladding, or, 1
- b. In measuring the elongation, its length exceeds its initial length by 0.125 inch (3.175 mm), or,
- c. A cladding failure exists as indicated by measurable release of fission products.
- d. Visual inspection identifies bulges, grcss pitting, or corrosion.
i Basis - The most severe stresses induced in the fuel elements result from pulse operation of the reactor,during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to i strains that are considerably lower than the strain expected to cause rupturing of a fuel element. Limited operation ;
in the steady state or pulsed mode may be necessary to identify a leaking fuel element especially if the leak is small. '
3.3 Reactor Coolant Svstems Acolicability - These specifications apply to the operation of the reactor water measuring systems.
Obiective- The objective is to assure that adequate cooling is provided to maintain fuel temperatures below the safety limit, and that the water quality remains high to prevent damage to the reactor fuel.
Soecifications- The reactor shall not be operated unless the systems and instrumentationchannels described in Table 3.3 are operable, and the information is displayed locally or in the control room.
l Table 3.3 Reauired Water Systems and Instrumentation Minimum '
Measuring Number Surveillance Channel / System Ooerable Function: Channel / System He.quirements'
- a. Tank Core 1 For operation of the reactor D1 l Inlet Temperature at 1.5 MW or higher, alarms Monitor on high heat exchanger outlet temperature of i 35'C (95'F)
- b. Tank Low Water 1 Alarms if water level drops M l Level Monitor below a depth of 23 feet
- c. Purification" 1 Alarms if the water D1,M.S Inlet Conduc- conductivity is greater
! tivity Monitor than 5 micrombos/cm l
l d. Emergency Core 1 Provides domestic supply D1,S Cooling System of water to cool fuelin the event of a LOCA for a minimum of 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> at 20 gpm from an appropriate nozzle 13 I
t l
4 m._
(*) Where: D1 - channel check during each day's operation A- channel calibration annually O .nnel test quarterly S - channel calibration semiannually M - channel test monthly
(") The purificationinlet conductivity monitor can be out-of-service for no more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> before the reactor shall be shutdown.
Basis -
Soecification (a) Table 3.3. The core inlet temperature alarm assures that large power fluctuations will not occur (SAR Section 4.7.2).
Soecification (b) Table 3.3. The minimum height of 23 feet of water above the tank bottom guarantees that there is sufficientwater for effective cooling of the fuel and that the radiation levels at the top of the reactor are within acceptable limits. The tank water level monitor alarms if the water level drops below 23 feet. (7.01 meter) (SAR Section 11.1.5.1).
Soecification(c) Table 3.3. The water conductivity remaining below 5 micrombos/cm averaged over a week, resultsin a minimizationof activation of waterimpuritiesand contaminationof reactor fuel and minimization of reactor structure corrosion.
Soecification (d) Table 31 This system will mitigate the LOCA event analyzed in the SAR Chapter 13, Section 13.2.
3.4 Reactor Room Exhaust System Aeolicability - These specifications apply to the operation of the reactor room exhaust system.
Obiective - The objectives of this specification are as follows:
- a. Reduce concentrations of airborne radioactive material in the reactor room.
- b. Maintain the reactor room pressure negative with respect to surrounding areas.
- c. As.sure continuous air flow through the reactor room in the event of a LOCA.
Soecification -
- a. The reactor shall not be operated unless the reactor room exhaust system is in operation and the pressure in the reactor room is negative relative to surrounding areas.
Basis - peration of the reactor room exhaust system assures that;
- a. Concentrationsof airbome radioactive material in the reactor room and in air leaving the reactor room will be reduced due to mixing with exhaust system air (SAR Section 9.5.1).
- b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fuel 14
l temperature below the safety limit in the event of a LOC /
l
- c. Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created l by the reactor room exhaust system (SAR Section 9.5.1). l l
3.5 This section intentionally left blank.
3.6 This section intentionally left blank.
3.7 Reactor Radiation Monitorina Systems 3.7.1 Mgnitorina systems Acolicabilitv - This specification applies to the information which shall be available to the reactor operator during reactor operation.
Ob_iective - The objective is to require that sufficient information regarding radiation levels and radioactive effluents is available to the reactor operator to assure safe operation of the reactor.
Soecifications - The reactor shall not be operated unless the channels described in Table 3.7.1 are l
operable, the readings are below the alarm setpoints, and the informationis displayed in the control room. The stack and reactor room CAMS shall not be shutdown at the same time during reactor operation.
Table 3.7.1 Reauired Radiation Monitorina Instrumentation i
Minimum Measuring Number Channel Surveillance Eouioment Ooerable Function Recuirements*
- a. Facility 1 Monitors Argon-41 and D1,W,S Stack Monitor radioactive particu-lates and alarms
- b. Reactor Room 1 Monitors the radiation level D 1,W,S Radiation in the reactor room and alarms Monitor
- c. Purification 1 Monitors radiation level at the D1,W,S System Radia- demineralizer station and tion Monitor alarms
- d. Reactor Room 1 Monitors air from the reactor D1,W,S Continuous room for particulate and Air Monitor gaseous radioactivity and alarms l
l
(*) Where: D1 - channel check during each days operation S - channel calibration semiannually W- channel test 15 l
- monitors may be placed out-of-service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for calibration and maintenance. During this n service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)
Rati.s -
Soecification (a) Table 3.7.1. The facility stack monitor provides information to operating personnel regarding the release of radioactivematerialto the environment (SAR Section 11.1.1.1.4). The alarm setpoin facility stack monitor is set to limit Argon-41 concentrationsto less than 10 CFR Part 20, Appendix B, Table 2, 1 values (averaged over one year) for unrestricted locations outside the operations area.
Soecification (b) Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor room during reactor operation (SAR Section 11.1.5.1).
Soecification (c) Table 3.7.1. The radiation monitor located next to the purification system resin canistersprovidesinformation regarding radioactivity in the primary system cooling water (SAR Section 11.1.5.4.2 Soecification (d) Table 3.7.1. The reactor room continuous air monitor provides information regarding airbome radioactivity from the reactor room (SAR Sections 11.1.1.1.2 & 11.1.1.1.5).
3.7.2 Effluents - Araon-41 Discharoe Limit Aeolicability This specificationappliesto the concentrationof Argon-41 that may be discharged from the MNRC reactor facility.
Obiective. The objective is to ensure that the health and safety of the public is not endangered by the discharge of Argon-41 from the MNRC reactor facility.
Soecification. The annual average unrestricted area concentration of Argon-41 due to releases of this radionuclide from the MNRC, and the corresponding annual radiation dose from Argon-41 in the unrestrcted area (at the operations boundary), shall not exceed the applicable levels in 10 CFR Part 20.
Hal.si The annual average concentrationlimit for Argon-41 in air in the unrestricted area is specified in Appendix B, Table 2, Column 1 of 10 CFR Part 20.10 CFR 20.1301 specifies dose limitations in the unrestricted area (at the operations boundary.) 10 CFR 20.1101 specifies a constraint on air emissions of radioactive materials to the environment. The Safety Analysis Report, Section 11.1.1.1.4 estimates that the routine Argon-41 releases and the corresponding doses in the unrestricted area will be below these limits.
3.8 Exoeriments 3.8.1 Reactivity Limt Acolicabilitv. This specificationapplies to the reactivity limits on experiments installed in the reactor and in-tank experiment facilities.
Obiective. The objective is to assure control of the reactor during the irradiation or handling of experiments adjacent to or in the reactor core.
Soecification Thereactorshallnotbeoperatedunlessthefollowingconditionsgoverningexperimenb exist:
16
Ak/k).
a The absolute reactivity worth of any moveable experiment shall be less than one (1) dollar (0
- b. The absolute reactivity worth of any single secured experiment shall be less than the maximum allowea pulse ($1.75)(1.23% Ak/k).
- c. The absolute total reactivity worth of in-tank experiments shall not exceed an absolute value of one dollar and ninety-two cents ($1.92) (1.34% Ak/k), including the potential reactivity which might result fr ficoding, voiding, or removal and insertion of the experiment.
BD$il-
- a. A limitation ofless than $1.00 on the reactivity worth of a single moveable experiment will assure that the pulse limit of $1.75 is not exceeded (SAR, Chapter,13). In addition, limiting the worth of each mov experiment to less than $1.00 will assure that the additional increase in transient power ark' Mmperature will b enough so that the fuel temperature scram will be effective (SAR, Chapter 13).
- b. The absolute worst event which may be considered !n conjunction with a single secured expe is it.s sudden accidentator unplarned removalwhile the reactoris operating. This would result in a r less than a pulse of $1.92, analyzed in Chapter 13, 9iven time shall not exceed the maximum reactivityinsertionlimit. Chap that an insertion of $1.92 worth of reactivity would be needed to reach the fuel temperature safety limit.
3.8.2 Materials Limit -
facilities. Anolicabilitv. This specification applies to experiments installed in the reactor and its experiment I
by limiting material quantity and radioactive material inventory of the experime
, materials exist - Soscincisen The reactor shall not be operated unless the following conditions governing experim
- a. Experiments containing materials corrosive to reactor components, compounds highly reactive wi water, potentially explosive materials, and liquid fissionable materials shall be appropriately encapsulated.
- b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 5 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is n millicuries.
- c. Explosive materials in quantities greater than 25 milligrams shall not be h, wated in the reactor tank. Explosive materials in quantities of 25 milligrams or less may be irradiated provided the pressu
'pressure detonation of the explosive has been calculated and/or experimentally demonstrated to be less th of the container,
- d. Explosive materials in quantities of three (3) pounds TNT equivalent or less may be irradiated in any radiography bay. The irradiationof explosives in any bay is limited to those assemblies where a sa has been performed that shows that there is no damage to the reactor safety systems upon detonati 17
13.2.6.2).
Basis .-
- a. Appropriate encapsulation is required to lessen the experimental hazards of some types of materials.
- b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueled experimentlead'ig r to total release of the iodine, occupational doses and doses to members of the general the unrestricted areas shall be within the limits in 10 CFR 20 (SAR Section 13.2.6.2).
- c. This specification is intended to prevent damage to vital equiptrent by restricting the quantity of explosive materials within the reactor tank (SAR Section 13.2.6.2).
- d. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in any radiography bay extemal to the reactor tank will not result in damage to the reactor controls or the reactor tank. Sa Analyses have been performed (SAR Section 13.2.6.2) which show that up to six (6) Ibs of TNT equivalent can safely irradiated in any radiography bay. Therefore, the three (3) Ib limit gives a safety margin of two (2). ,
3.8.3 Failure and Malfunctions Acolicability. This specification applies to experiments installed in the reactor and its experimental facilities.
Obiective. The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.
Soecifications
- a. All experiment materials which could off-gas, sublime, vo'atilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, or (3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the urirestricted area, the quantity and type of material to be irradiated shall be limited such that the ,
airbome concentration of radioactivity shall not exceed the applicable limits of 10 CFR Part 20 (at the operations boundary), assuming 100% of the gases or aerosols escape.
- b. In calculations pursuant to (a) above, the following assumptions shall be used:
(1) If the effluent from an experiment facility exhausts through a stack which is closed on high radiation levels, at least 10% of the gaseous activity or aerosols produced will escape.
(2) If the effluent from an experiment facility exhausts through a filter installation designed for ;
greater than 99% efficiency for 0.3 micron and larger particles, at least 10% of these will escape.
(3) For materialswhose boiling point is above 130'F and where vapors formed by boiling this materialcan escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, an evaluation shall be made to determine the need for corrective action. Inspection and any corrective action taken shall be reviewed by the Facility Director or his designated alternate and determined to
- 8
be satisfactory before operation of the reactor is resumed.
Bahia- >
- a. This specification is intended to reduce the likelihood that airbome radioactivity in excess of the limits of 10 CFR Part 20 shall be released into the reactor building or to the unrestricted area (SA
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment failure (SAR Section 13.2.6.2).
- c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited to avoid release of fission products. Potentialdamage to reactor fuel or structure must be brought to the attentio Facility Directoror his designated attemate for review to assure safe operation of the reactor (SAR S
' 4.0 Surveillance Reauirements General The surveillance frequencies denoted herein are based on continuing operation of the reactor.
Surveillance activities scheduled to occur during an operating cycle which can not be performed with the react operating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable t reactor system or measuring channel surveillance requirement may be waived during the associated time perio to reactor system or measuring channel operation, the surveillance shall be performed for each reactor s measuring channel for which surveillancewas waived. A reactor system or measuring channel shall not be considere operable until it is successfully tested.
4.1 Reactor Core Parameters 4.1.1 Steadv State Ooeration channels. Anslic.abi;;tv. This specification applies to the surveillance requirement for the power level monito Obiectiva The objectiveis to verify that the maximum power level of the reactor does not exceed the authorized limit.
Soecification. An annual channel calibration shall be made of the power level monitoring channel.
- If a channel is removed, replaced, or unscheduled maintenance is performed, or a significant change in core configurationoccurs, a channelcalibration shall be required. Discoveryof noncompliancewith specification 4.1.1 shal limit reactor operations to that required to perform the surveillance.
Baals. The annual power level channel calibration will assure that the indicated reactor power level is correct.
4.1.2 Shutdown Marain and Excess Randivity
{
Ass;ic.abilitv. These specificationsapply to the surveillance requirements for reactivity control of the
- reactor core.
Obiectiva The objectke is to measure and verify the reactivity worth, performance, and operability of those systems affecting the reactivity of the reactor.
)
Soecifications.
19
- a. The total reactivity worth of each control rod and the shutdown margin shall be determined anr.ualy or following any significant change in core or control rod configuration. The shutdown margin shall be verite meeting the requirements of Section 3.1.3(a).
- b. The core excess reactivity shall be verified: ,
(1) Prior to each startup operation and, (2) Following any change in core loading or configuration.
Discovery of noncompliancewith Specifications 4.1.2(a-b)shalllimit reactor operationsto that required to p surveillance.
M Basis -
- a. The reactivityworth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the excess reactivity of the core. Past experience w similar reactors gives asserance that measurementsof the control rod reactivity worth on an annual basis is adequ to assure no significantchanges in the shutdown margin, provided no core loading or configuration changes have made.
- b. Determining the core excess reactivity prior to each reactor startup shall assure that specification 3.1.3(b) shall be met, and that the critical rod positions do not change unexpectedly.
4.2 Reactor Control and Safety Systems 4.2.1 Control Rods Acolicability. This specification applies to the surveillance of the control rods.
Obiective. The objectiveis to inspectthe physicalconditionof the reactor control rods and establish the operable condition of the rods.
I Soecification(sl Control rod worths shall be determined annually or after physical removal or any significant change in core or control rod configuration.
l
- a. Each control rod shall be inspected at annualintervals by visualobservationof the fueled sections and absorber sections plus examination of the linkages and drives.
- b. The scram time of each control rod shall be measured semiannually.
Discovery of noncompliancewith specificatica4.2.1 (a-b) shalliimit reactor operations to that required to perform the surveillance.
Sal si (Specification 4.2.1.a-b). Annual determination of control rod worths or measurements after any physical removal or significantchange in core loading or control rod configuration provides information about changes in reactor total reactivity and individual rod worths. The frequency of inspection for the control rods shall provide periodicverification of the condition of the control rod assemblies. The specification intervals for scram time assure operable performance of the control rods.
20
.. , .. - _ - = - - - - ~ _ _ - - -_ ...
4.2.2 Reactor Instrumentation Acolicabilitv. These specifications apply to the surveillance requirements for measurements, tests, calibration and acceptability of the reactor instrumentation.
Obiective The objective is to ensure that the power level and fuel temperature instrumentation are operable.
Soecifications.
- a. The reactor power level safety channels shall have the following:
(1) A channel test monthly or after any maintenance which could affect their operation.
(2) A channel check during each day's operation.
l (3) A channel calibration annually.
- b. The Linear Power Channel shall have the following:
(1) A channel test monthly or after any maintenance which could affect its operation.
{
(2) A channel check during each day's operation.
(3) A channel calibration annually. !
- c. The Log Power Channel shall have the following:
(1) A channel test monthly or after any maintenance which could affect operations.
(2) A channel check during each day's operation.
(3) A channel calibration annually.
- d. The fuel temperature measuring channels shall have the following:
(1) A channel test monthly or after any maintenance which could affect operations.
(2) A channel check during each day's opc. ration.
(3) A channel calibration annually.
- e. The Pulse Energy Integrating Channel shall have the following:
(1) A channel test monthly.
(2) A channel ca3bration annually.
Discovery of noncompliancewith specificatons 4.2.2(a-e) shall limit reactor operation to that required to perform the
. surveillance.
Eilll-i 21 i
7
- a. A daily channel check and monthly test, plus the annual calibration, will assure that ths reactor power level safety channels operate properly,
- b. A channeltest monthly of the reactor powerlevel multi-range channel will assure that the channel is operable and responds correctly. The channel check will assure that the reactor power level multi-range l channelis operable on a daily basis. The channel calibration annually of the multi-range linear channel will assur that the reactor power will be accurately measured so the authorized power levels are not exceeded.
- c. A channel test monthly will assure that the reactor power level wide range log channel is operable and responds correctly. A channel check of the reactor power level wide range fog channel will assure that the channel is operable on a daily basis. A channel ca!Lration will assure that the channel will indicate properly at the corresponding power levels.
- d. A channel test monthly and check during each day's operation, plus the annual calibration, will assure that the fuel temperature measuring channels operate properfy.
- e. A channel test monthly plus the annual channel calibrationwill assure the pulse energy integrating channel operates properfy.
4.2.3 Reactor Scrams and Interlocks Acolicability. These specifications apply to the surveillance requirements for measurements, test, calibration, and acceptability of the reactor scrams and interlocks.
Obiective. The objective is to ensure that the reactor scrams and interlocks are operable.
Soecification.
- a. Console Manual Scram. A channel test shall be performed monthly.
- b. Reactor Room Manual Scram. A channel test shall be performed monthly.
- c. Radiography Bay Manual Scrams. A channel test shall bo performed monthly.
- d. Reactor Power Level Safety Scram. A channel test sha'! be performed monthly.
- e. High-Voltage-Power Supply Scrams. A channel test sha!Lhe performed monthly,
- f. Fuel Temperature Scram. A channel test shall be performed monthly.
- g. Watchdog Circuits Scrams. A channel test shall be performed monthly.
I
- h. Extemal Scrams. A channel test shall be performed monthly.
- 1. The One Kilowatt Pulse Interlock. A channel test of the one kilowatt interlock shall be performed monthly.
J. Low Source Level Rod Withdrawal Prohibit interlock. A channel test shall be performed monthly. 1
- k. Control Rod Withdrawal interlocks. A channel test shall be performed monthly.
22 1
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- 1. Magnet Power Key Switch Scram. A channel test shall be performed monthly.
Discovery of noncompliancewith Specifications 4.2.3(a-l) shall limit reactor operation to that required to perform the surveillance.
Baall- i l
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- a. A channel test monthly of the Console Manual Scram will assure that the scram is operable.
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- b. A channel test monthly of the Reactor Room Manual Scram will assure that the scram is operable.
- c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scrams are
- d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scrams are I operable,
- e. A channeltest monthly of the Loss-of-High-Voltap Scram will assure that the high voltage power supplies are operable and respond correctly.
- g. A channel test monthly of the Watchdog Circuits Scrams weekly will assure that the scram circuits .
pre operable.
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- h. A channel test monthly of the Extemal Scrams will assure that the scrams are operable and respond correctly.-
- 1. A chanr'el test monthly will assure that the One Kilowatt Pulse Interlock works properly.
- j. A channel test monthly (4 the Low Source Level Rod Withdrawal Prohibit Interlock will assure that the interlock is operable.
l k. A channel test monthly of the Control Rod Withdrawal interlock will assure that the interlock is
- l. operable.
1-
- 1. A channel test monthly of the Magnet Current Key Switch will assure that the scram is operable. 1 4.2.4 Reactor Fuel Elements Anolicabilitv. This specification applies to the surveillance requirements for the fuel elements.
Oblective The objective is to verify the continuing integrity of the fuel-element cladding.
Specification To assure the measurementlimitationsin Sect!- 3.2.4 are met, the following te di be done:
- a. The lead elements (i.e., all elements adjacent to the transient rod, with the exception of the instrumented fuel elements), and all elements adjacent to the central irradiation facility shall be inspected annually.
23 i
- b. Tha instrumented futi ci;ments shall be inspected if the elimsnts adjtcznt to it fail to pass tha visual and/or physical measurement requirements of Section 3.2.4 Discovery of noncompliancewith specification 4.2.4 shaltlimit operations to that required to perfo Smis fSoecification4.2.4 a-b1 The above specification assures that the lead fuel elements inspected regularlyand the integrity of the lead fuel elements shall be maintained. These are the the highest power density as analyzed in the SAR. The instrumented fuel element is exclud damage to the thermocouples.
' Anolicability. This specificationapplies to the surveillanm requirements for the reactor water measuring an the emergency core cooling systems.
Objective The objective is to assure that the reactor tank water temperature monitoring system water level alarm, the water conductivity cells and the ECCS are operable.
I Soecification.
a The reactor tank core inlet temperature monitor shall have the following:
(1) A channel check during each day's operation.
(2) A channel test quarterly.
(3) A channel calibration annually.
- b. The reactor tank low water level monitoring system shall have the following:
(1) A channeltest monthly.
- c. The purification inlet conductivity monitors shall have the following:
(1) A channel check during each day's operation.
(2) A channel test monthly.
(3) A channel calibration semiannually.
- d. The Emergency Core Cooling System (ECCS) shall have the following:
(1) A channel check prior to operation.
(2) A channel calibration semiannually.
Discovery of noncompliance with specifications 4.3(a-d) shall limit operations to that required to perform the surveillance.
Basis -
- a. A channeltest quarterly assures the core inlet temperature monitor responds correctly to an input signal.
24
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A channelcheck during each day's operation assures the channelis operable. A channel calibration annually assures the monitoring system reads properly.
- b. A channel test monthly assures that the low water level monitoring system responds correctly to an input
- c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to an input l
signal. A channel check during each day's operation assures that the channel is operable. A channel calibration l semiannually assures the conductivity monitoring system reads properfy.
l d. A channel check prior to operation assures the ECCS is operable. A channel calibration semiannually assures the ECCS performs as required.
4.4 Reactor Room Exhaust System Aoolicability. This specification applies to the surveillance requirements for the reactor room exhaust system.
Obiective. The objective is to as,:ure that the reactor room exhaust system is operating properly.
Soecification. The reactor room exhaust system shall have a channel check during each day's operation.
Discovery of noncompliance with this specification shall lim;t operations to that required to perform the surveillance.
flasjs - A channelcheck during each day's operation of the reactor room exhaust system shall verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surro"nd ig facility areas.
4.5 This section intentionally left blank.
4.6 This section intentionally left blank.
4.7 Reactor Radiation Monitorina Systems Aoolicability This specification applies to the surveillance requirements for the reactor radiation monitoring systems.
Objet.u; The objective is to assure that the radiation monitoring equipment !s operating properly.
Soecification.
- a. The facility stack rrionitor shall have the following:
(1) A channel check during each day's operation.
(2) A channel test weekly.
(3) A channel calibration semiannually.
- b. The reactor room radiation monitor shall have the following:
(1) A channel check during each day's operation.
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. . . ~ . ~ . . . - - . . - - . - . - . - . ~ . - .- -. . . . ~ , . - - - - - _ -
(2) Achannelt:stweekly (3) A channel calibration semiannually. ;
- c. The purification system radiation monitor shall have the following:
(1) A channel check during each day's operation. ,
t (2) A channel test weekly. ;
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, (3) A channel calibration semiannually,
. d. The reactor room Continuous Air Monitor (CAM) shall have the following:
1 s
(1) A channel check during each day's operation.
(2) A channel test weekly.
2 (3) A channel calibration semiannually.
Discovery of noncompliance with specifications 4.7(a-d) shall limit operations to that required to perform the surveillance. ;
Basis
- o A channelcheck of the facility stack monitor system during each day's operation will assure the monitor is opersb e. : A channel test weekly will assure that the system responds correctly to a known source. A channel- '
calibiation semiannually will assure that the monitor reads correctly.
- b. A channel check of the reactor room radiation monitor during each day's operation will assure that the . t monitor is operable.L. A channel test weekly will ensure that the system responds to a known source. A channel calibration of the monitor semiannually will assure that the monitor reads correctly. .
- c. A channelcheck of the purification system radiation monitor during each day's operation assures that the ,
- monitor is operable. A channel test weekly will ensure that the system responds to a known source. A channel calibration of the ny>nitor semiannually will assure that the monitor reads correctly. 1
- d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day's operation will assure '
that the CAM is operable. A channel test weekly will assure that the CAM responds correctly to a known source. A channel calibration semiannually will assure that the CAM reads correctly.
4.8 Expenments l
4 Anolicability. This specification applies to the surveillance requirements for experiments installed in the reactor ,
' or experiment facilities.
.. Qhg;tly.s The objective is to prevent the conduct of experiments or irradiationswhich may damage the reactor or release excessive amounts of radioactive materials as a result of experimental failure.
Specificatons 26 l l
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- a. A new experimentshall not be installed in any MNRC reactor irradiation facility until a safety anal been performed and reviewed for compliance with the Limitations on Experiments, (Technical Specifications 3.8) and 10 CFR 50.59 by the Facility Director or his designee.
- b. A!! experiments performed at the MNRC shall meet the conditions of an approved Facility Use Authoriz Facility use authorizations and experiments carried out under these authorizations shall be reviewed an in accordance with the Utilization of Technology and Industrial Support Nuclear Facilities Document (MNR DOC). An experimentclassifiedas an approvedexperimentshailnot be placed in any MNRC reactorirr untilit has been reviewed for compliance with the approved experimentand facility use authorization by the O Supervisor and the Health Physics Supervisor, or their designated alternates.
- c. The reactivity worth of any in-tank experiment shall be estimated or measured, as appropriate, before reactor cperation with said experiment. Whenever a measurement is done it shall be done at ambient conditions.
l Experiments shall be identified and a log or other record maintained while the experiment is in the reactor or in on l of the in-tank experiment facilities.
Basis a & b. Experienceat most TRIGA reactor facilities verifies the importance of reactor ataff arid safety com reviews of proposed experiments.
- c. Measurementof the reactivityworth of an experimentor estimationof the reactivityworth based on previo or similar measurements shall verify that the experiment is within authorized reactivity limits.
5.0 Desian Features 5.1 Site and Facility Des;riotion 5.1.1 Site Aeolicabilitv. This specification applies to the McClellan Nuclear Radiation Center site location and specific facility design features.
Obiective. The objective is to specify those features related to the Safety Analysis evaluation.
Soecification.
- a. The site location is sit approximately 8 miles (13 km) north-by-noituast of downtown Sacramento, Califomia. The site of the MNku facility is about 3000 ft (0.6 mi or 0.9 km) west of Watt Avenue, and 4500 ft (0.9 mi or 1.4 km) south of E Street.
- b. The restricted area is that area inside the fence surrounding the reactor buildng. The unrestricted area is that area outside the fence surrounding the reactor building.
- c. The TRIGA reactoris located in Building 258, Room 201 of the McClellan Nuclear Radiation Center (MNRC). This building has been designed with special safety features.
- d. The core is below ground levelin a water filled tank and surrounded by a concrete shield.
- e. Restricted access areas include the reactorcontrol room, reactor room, equipment room, and the radiography bays during normal duty hours.
27
flaliS-
- a. Information on the surrounding population, the hydrology, seismology, and climatography of the site have been presented in Chapter 2 of the Safety Analysis Report.
- b. The room enclosing the reactor has been designed with systems related to the safe operation of the facility,
- c. The below grado core design is to negate the consequences of an aircraft hitting the reactor building. This accident was analyzed in Chapter 13 and found to be beyond a credible accident scenario.
- d. The restricted access to specific facility areas assure that proper controls are established for the l safety of the public and for the security of the special nuclear materials.
5.1.2 Facility Exhaust Acolicability. This specification applies to the facility which houses the reactor.
Obiective The objectiveis to assure that provisions are made to restrict the amount of radioactivity released into the environment, or during a LOCA, the system is to assure proper removal of heat from the reactor room.
Soecifications.
- a. The MNRC reactor facility shall be equipped with a system designed to filter and exhaust air from the MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2m (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor contrcl room.
Basis. The MNRC facility exhaust system is designed such that the reactor room and radiography bays are maintained at a negative pressure with respect to the surrounding areas. The free air volume within the MNRC facility is confined to the facility when there is a shutdown of the exhaust system. Controls for startup, filtering, l
and normal operation of the exhaust system are located in the reactor control room. Proper ha: idling of airbome radioactive materials (in emergency situations) can be conducted from the reactor control room with a minimum of exposure to operating personnel.
5.2 Reactor Coolant System Anoticability. This specification applies to the reactor coolant system.
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Obiective. The objectiveis to assure that adequate water is available for cooling and shielding during normal reactor operation or during a LOCA.
Soecification.
- a. During normal reactor operation the reactor core shall be cooled by a natural convection flow of water.
- b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below a depth !
of 23 ft.
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- c. During a LOCA the reactor core shall be cooled for a minimum of 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> at 20 gpm by a source of water from the ECCS.
Basis
- a. SAR, Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930*C will not be exceeded under natural convection flow conditions.
- b. A reactor tank water low level alarm sounds when the water level drops significantly, This alarm annunciatesin the reactor control room and at the command post during off duty hours so that appropriate corrective action can be taken to restore water for cooling and shielding. .
- c. SAR, Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel and shows that the fuel safety limit is not exceeded under LOCA conditions during this water cooling.
5.3 Reactor Core and Fuel 5.3.1 Reactor Core AssncaNutv. This specification applies to the configuration of the fuel.
Obtective The objective is to assure that provisions are made to restrict the arrangement of fuel elements so as to provide assurance that excessive power densities will not be produced.
SoecificaGon For operation at 0.5 MW or greater, the core shall be an arrangement of 100 or more .
fuel elements. Below 0.5 MW there is no minimum required number of fuel elements, in a mixed 20/20 and 8.5/20 fuel j loading: '
Mixed J Core a.. No fuel shall be loaded into Hex Rings A or B.
- b. A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.
20E Core
- a. No fuel shall be loaded into Hex Rings A or B.
- b. Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.
- c. Variations to 20E having 20 wt% fuelin Hex Ring C requires the 20 wt% fuel to be loaded into comer positions gak and graphite dummy elements in the flat positions. The performance of fuel temperature measurements shall apply to variations to the as-analyzed 20E configurations.
Balls in order to meet the power density requirementsdisce;sedin Chapter 4 of the Safety Analysis Report, no less than 100 fuel elements and the above load;ng restrictionswill be allowed in an operational 0.5.MW or greater core. Specification 20E(3) allows for variations of the as-analyzed core with the condition that temperature limits are being maintained.
5.3.2 Reactor Fuel i
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I Aeolicability. Th:ss specifications apply to the fu:I el:ments us*d in the r: actor cora.
Obiective. The objectiveis to assure that the fuel elements are of such design and fabricated in such I' a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Soecification TheindividualunirradiatedTRIGAfuelelementsshallhavethefollowingcharacteristics a Uranium content: 8.5,20 or 30 wt%, uranium, enriched to a nominal less than 20% U-235.
- c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.
HaSiG -
- a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limiting operation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between the maximum temperature generated in the fuel and the safety limit for fuel temperature. The fuel temperatures are not epected f
to exceed 630*C during any cond. tion of normal operation.
- b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, is equal to the clad yield strength when both fuel and cladding temperature are at the safety limit 930*C. Since the fuel temperatures are not expected to exceed 630*C during any condition of normal operation, there is a margin between the fuel element clad stress and its ultimate strength.
- c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances of purchased materials.
5.3.3 Control Rods and Control-Rod Drives Acolicability This specification applies to the control rads and control-rod drives used in the reactor core.
l Obiective The objectiveis to assure the controlrods and control-rod drives are of such a design as to permit their use with a high degree of reliability with respect to their physical, nuclear, and mechanical i
characteristics. l Soecification. f l
- a. All control rods sh'4ll have scram capability and contain stainlese steel, borated graphite, B 4C powder, or boron and its compoundsin alid form as a neutron poison. The shim and regulating rods shall have fuel )
followers sealed in stainless steel. The tra 'sient rod shall have an air filled follower and is sealed in an aluminum tube.
- b. The control-rod drives shall De the standard GA rack and pinion type with an electromagnet and armature attached.
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- a. The neutron poison requirements for the control rods are satisfied by using neutron absorbing borated graphite B C powder, or boron and its compounds. These materials shall be contained in a suitable clad 30 l l
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I material such as stainless steel or aluminum to assure mechanical stability during movement and to isolate the neutrii poison from the tank-water environmentc Scram capabilities are provided for rapid insertion of the control rods. I
- b. The standard GA TRIGA control-rod drive meets the requirements for driving the control rods at the proper speeds, and the electromagnetand armature provide the requirements for rapid insertion capai drives have been tested and proven in many TRIGA reactors.
5.4 Fissionable Material Storace Anchcabehty This specificationapplies to the storage of reactor fuel at a time when it is not in the reactorcore.
Obpective The objective is to assure that the fuel which is being stored will not becon, :ritical and will not reach an unsafe temperature.
. Soecification
- a. All fuel elements not in the reactorcore shall be stored (wet or dry) in a geometrical array where the k,,,l less than 0.9 for all conditions of moderation. l
- b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection coolin I by water or air such that the fuel-element temperature shall not exceed the safety limit.
Baals. a&b. The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.
6.0 Adirdiiististive Controls 6.1 Oroanization. The MNRC facility shall be under the direct control of the MNRC Director or a licensed senior reactor operator (SRO) designated by him to be in direct control. The MNRC Director shall be accountable to l
'i the Responsible Commander for the safe operation and maintenance of the reactor and its associated equipment.
y 6.1.1 Structura The managementforoperationof the MNRC facility shallconsistof the organizational structure as shown in Figures 6.1 through 6.3.
6.1.2 Resoonsibilities. The MNRC Director shall be accountable to the Responsible Commander for
' . the safe operation and maintenance of the reactor and its associated equipment. The MNRC Director shall review and approve all experiments and experimental procedures prior to their use in the reactor. Individuals in the
)
management organization (i.e., reactor operations supervisor, health physics supervisor, etc.) shall be responsible for implementing the policies and operation of the facility, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating l'cense and technical specifications.
6.1.3 Staffing.
6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:
- a. A reactor operator in the control room.
- b. A second person in the facility area that can perform prescribed instructions.
- c. A senior reactor operator shall be present whenever a reactor startup is performed, fuel is being moved, or experiments are being placed in the reactor tank.
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- d. A senior reactor operator readily available. The available senior reactor operator should be within thirty (30) minutes of the facility and reachable by telephone.
6.1.3.2 Alistof reactor facility personnel by name and telephone number shall be available to the reactor operator in the control room. The list shall include:
- a. Management personnel.
- b. Health Physics personnel,
- c. Reactor Operations personnel.
6.1.4 Selection and Trainina of Personnel The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Se'ection and Train Personnelfor Research Reactors (ANS 15.4). Qualificationand requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.
6.2 Review and insoection General Poliev. Nuclear facilities shall be designed, constructed, operated, and maintainedin such a manner that facility perscnnel, the general public, and both govemment and nongovemment property are not exposed to undue risk. These activities shall be conducted in accordance with applicable governmental regulatory requirements.
The Responsible Commander shallinstitute this policy as the facility license holder. The Nuclear Safety Committee (NSC) has been chartered to assist in meeting this responsibility by providing objective and independent reviews, evaluations, advice and recommendations on matters affecting nuclear safety. The following describes the composition and conduct of the NSC.
6.2.1 NSC Comoosition and Qualifications. The ResponsibleCommandershall appointthe chairman of the NSC. The chairman shall appoint a Nuclear Safety Committee of at least five (5) members knowledgeable in fields which relate to nuclear safety. The NSC shall evaluate and review safety standards associated with the operation and use of the nuclear facilities.
6.2.2 NSC Charterand Rules. The NSC shall conduct its review and audit functions in accordance with a written charter. This charter shall include provisions for
- a. Meeting frequency (full committee shall meet at least semiannually).
- b. Voting rules.
- c. Quorums.
- d. Method of submission and content of presentations to the committee.
- e. Use of subcommittees.
- f. Review, approval and dissemination of meeting minutes.
6.2.3 Review Function. The responsibilities of the NSC, or designated subcommittee thereof, shall include but are not limited to the following:
32
- a. Review of approved experiments utilizing the reactor facilities.
- b. Review of all proposed changes to the facility Technical Specifications or SAR.
- c. Determination of whethera proposed change, test, or experiment would constitute an unreviewed safety question, in accordance with 10 CFR 50.59, or require a change to the Technical Specifications. This j
determination may be in the form of verifying a decision already made by the Facility Director,
! d. Review of the operation and operational records for both reactor operations and health physics
'e. Review of abnormal performance of facility equipment and operating anomalies.
- f. Review of all reportable events.
6.2.4 insnection Functiort The NSC, or a subcommittee thereof, shall inspect reactor operations and l
health physics annually, The annualinspection shallinclude, but not be limited to the following:
l
- a. Inspection of the reactor operating and health physics records.
- b. Inspection of the reactor facility.
- c. Examination of reportable events.
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- d. Determination of the adequacy of standard operating procedures.
- e. Verification of the effectiveness of the training program.
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- f. Verification of conformance of operations with the operating license and Technical Specifications 4 and applicable regulations.
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6.3 Radiation Safety. The Health Physics Supervisor shall be responsible for implementation of the MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI /ANS 15.11). The Health Physics Supervisor shall report to the Director.
6.4 Procedures. Written procedures shall be prepared and spproved prior to initiating any of the activities listed in this section. The procedures shall be approved by the MNRC Director. A periodic review of procedures sha be performed and documented in a timely manner by the MNRC staff to assure they are current. Procedures shall be adequate to assure the safe operation of the reactor but shall not preclude the use of independent judgemen action should the situation require. Procedures shall be in effect for the following items:
6.4,1 Reactor Ooerations PrrrMures.
- a. Startup, operation, and shutdown of the reactor,
- b. Fuel loading, unloading, and movement within the reactor, i
- c. Control rod removal or replacement.
- d. Routine maintenance of the control rod drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety.
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, -,.% , ,w-', ------r -- . ,. ., c- . , - - .- . n- 4 w-rye.,- r
...___ --- _. _ ._. _ ._._._.. _ ._.__. - _ _ _ _ __.s
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
- f. Administrative controls for operations, maintenance, and conduct of irradiations and experiments ,
that could affect reactor safety or core reactivity.
- g. Implementation of required plans such as emergency or security plans.
- h. Actions to be taken to correct specific and foreseen potential malfunctions of systems, including responses to alarms and abnormal reactivity changes.
6.4.2 Health Physics Procedures.
- a. Testing and calibration of area radiation monitors, facility air monitors, laboratory radiation detection systems, and portable radiation monitoring instrumentation.
- b. Working in laboratories and other areas where radioactive materials are used.
- c. Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solid and liquid waste, and gaseous effluents released from the facility. The program shallinclude a managementcommitment to maintain exposures and releases as low as reasonably achievable (ALARA).
- d. Monitoring radioactivity in the environment surrounding the facility,
- e. Administrativeguidelines for the facility health physics program to include personnel orientation and training..
- f. Receiptof radioactive materialsat the facility,and unrestrk.ted release of materials and items from the facility which may contain induced radioactivity or radioactive contamination.
- g. Leak testing of sealed sources containing radioactive materials.
- h. Special nuclear material acco' 9 ability. -
- 1. Transportation of radioactive materials.
Changes to the above procedures shall require approval of the MNRC Director. All substantive changes shall be documented.
6.5 Exoeriment Review and Acoroval All new classes of experiments (designated Facility Use Authorization) shall be approved by the Nuclear Safety Committee (NSC) and the MNRC Director. All specific experiments to be performed under the provision of an approved Facility Use Authorization shall be approved by the MNRC Director,
- a. Approved experiments shall be carried out in accordance with established approval procedures.
i
- b. Substantive change to previously approved experiments shall require the same review and approval as a new experiment. +
- c. Minor changes to an experimentthat do not significantlyalter the experiment may be approved by a senior reactor operator.
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l 6.6 Recuired Actions 6.6.1 Actions to be taken in case of a safety limit violation.
In the event of a safety limit violation (fuel temperature), the following actions shall be taken:
- a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the Nuclear Regulatory Commission.
- b. The safety limit violation shall be promptly reported to the MNRC Director.
- c. The safety limit violation shall be reported to the chairman of the NSC and to the Nuclear Regulatory Commission by the MNRC Director.
- d. A safety limit violation report shall be prepared. The report shall describe the following:
(1) Applicablecircumstancesleading to the violation including, when known, the cause and contributing factors.
(2) Effect of violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public.
(3) Corrective action to be taken to prevent recurrence.
- e. The safety limit violation report shall be reviewed by the NSC and then be submitted to the Nuclear Regulatory Commission when authorization is sought to resume operation of the reactor.
6.6.2 Actions to be taken for a reoortable occurrence.
In the event of a reportable occurrence, the following actions shall be taken: '
- a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it is necessary to shutdown the reactorto correct the occurrence, operations shall not be resumed unless authorized by the MNRC Director or his designated alternate.
- b. The occurrence shall be reported to the MNRC Director or his designated alternate. The MNRC Director shall report the occurrence to the Nuclear Regulatory Commission as required by these Technical Specifications or any applicable regulations.
- c. All occurrences shall be reported to the NSC at the same time the NRC is notified. All occurrence reports should be reviewed by the NSC before being sent to the NRC.
6.7 Reoorts. All written reports shall be sent within the prescribed interval to the Nuclear Regulatory Commission, Washington DC 20555, Attn: Document Control Desk.
6.7.1 Ooeratina Reoorts. An arsnual report covering the activities of the reactor facility during the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, performance characteristicsand operating procedures related to reactor safety occurring during the reporting period, and results of surveillance tests and inspections.
35
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- b. A tabulation showing the energy generated by the reactor (in meg w:tt hours), hours tha reactor was critical, and the total cumulative energy output since initial criticality.
- c. The number of emergency shutdowns and inadvertent scrams, including reasons therefor.
- d. Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required.
- e. A brief description, including a summary of the safety evaluations, of changes in the facility or in procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50.
- f. A summary of the nature and amount of radioactive effluents released or discharged to the environment beyond the effective control of the licensee as measured at or prior to the point of such release or discharge, including the following:
(1) Liould Effluents (summarized on a monthly basis)
(a) Liquid radioactivity discharged during the reporting period tabulated as follows:
(i) The total estimated quantity of radioactivity released (in curies).
(ii) An estimation of the specific activity for each detectable radionuclide present if the specific activity of the released material after dilution is greater than 1 x 10 microcuries/ml.
(iii) A summary of the total releasein curies of each radionuclidedetermined in (ii) above for the reporting period based on representative isotopic analysis.
(iv) An estimated average concentration of the released radioactivematerial at the point of release for each month in which a release occurs, in terms of microcuries/ml and the fraction of the applicable limit in 10 CFR 20.
(b) The total volume (in gallons) of effluent water (including diluent) released during each period of liquid effluent release.
(2). Airborne Effluents (summarized on a monthly basis)
(a) Airborne radioactivity discharged during the reporting period (in curies) tabulated as follows:
(i) The total estimated quantity of radioactivity released (in curies) determined by an appropriate sampling and counting method.
(ii)The total estimated quantity (in curies) of Argon-41 released during the reporting period based on data from an appropriate monitoring system.
(iii) The estimated maximum annual average concentration of Argon-41 in the unrestricted area (in microcuries/mi), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of the applicable 10 CFR 20 limits for these values.
(iv) The total estimated quantity of radioadivity in particulate form with half-lives greater than eight days (in curies) released during the reporting period as determined by an appropriate 36 l
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particulate monitoring system.
l (v) The average concentration of radbactive particulates with half-tives greater than eight days released (in microcuries/ml) during the reporting period.
(5) Solid Waste (summarized on an annual basis) l (a) The total amount of solid waste packaged (in cubic feet).
(b) The total activity in solid waste (in curies).
(c) The dates of shipment and disposition (if shipped offsite).
- g. An annualsummary of the radiation exposurereceived by facility operations personnel, by facility users, and by visitors in terms of the average radiation exposure per individualand the greatest exposure per individual in each group.
- h. An annual summary of the radiation levels and levels of contamination observed during routine surveys performed at the facility in terms of average and highest levels.
- 1. An annual summary of any environmental surveys performed outside the facility.
6.7.2 Soecial Reoorts Specialreports are used to report unplanned events as weli as planned major facility and administrativechanges. The following classificationsshall be used to determine the appropriate reportin schedule.
- a. A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone or similar conveyance to the NRC Operations Center of:
(1) Any accidentairelease of radioactivityinto unrestrictedareas above applicable unrestricted area concentration limits, whether or not the release resulted in property damage, personalinjury or exposure; (2) Any violation of a safety limit; (3) Operation with a limited safety system setting less conservative than specified in Section 2.0, Limiting Safety System Settings; (4) Operation in violation of a Limiting Condition for Operation; (5) Failure of a required reactor or experiment safety system component which could render the system incapable of performing its intended safety function unless the failure is discovered during maintenance tests or a period of reactor shutdown; (6) Any unanticipated or uncontrolled change in reactivity greater than $1.00; (7) An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of a condition which could have resulted in operation of the reactor outside the specified safety limits, and; (8) A measurable release of fission products from a fuel element j b. A report within 14 days in writing to the NRC, Document Control Desk, Washington DC 20555.
37
(1) Those events reported as required by Sections 6.7.2.a.(1) through 6.7.2.a.(8).
(2) The written report (and, to the extent possible, the preliminary telephone report or re by similar conveyance) shall describe, analyze, and evaluate safety implications,and outline the taken or planned to prevent recurrence of the event.
i
- c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC 20 (1) Any significantvariation of measured values from a corresponding predicted or previ measured value of safety-connected operating characteristics occurring during operation of the reactor; (2) Any significant change in the transient or accident analysis as described in the Safety Analysis Report (SAR);
(3) Any changes in facility organization or personnel; (4) Any observed inadequaciesin the implementation of administrativeor proceduralcontr such that the inadequacy causes or could have caused an existence or development of an unsafe condition wi!
regard to reactor operations.
6.8 Baceda Records may be in the form of logs, data sheets, or other suitable forms .The required information may be contained in single or multiple records, or a combination thereof. Records and logs shall b -
prepared for the followingitems and retained for a period of at least five years for items a. through f., a foritems g. through k. (Note: Annual reports, if they contain all of the required information, may be used as for items g. through k.)
- a. Normal reactor operation.
- b. Principal maintenance activities.
- c. Those events reported as required by Sections 6.7.1 and 6.7.2.
- d. Equipment and component survelilance activities required by the Technical Specifications.
- e. Experiments performed with the reactor.
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- f. Airbome and liquid radioactive effluents released to the environments and solid radioactive waste shippe off-site.
- g. Offsite environmental mor,itoring surveys.
- h. Fuelinventories and transfers.
4
- 1. Facility radiation and contamination surveys.
J. Radiation exposures for all personnel. .
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- k. Updated, corrected, and as-built drawings of the facility.
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