ML22112A240
| ML22112A240 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos Nuclear Center |
| Issue date: | 04/22/2022 |
| From: | GE-Hitachi Nuclear Energy Americas |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22112A236 | List: |
| References | |
| M220027 NEDO 32765, Rev 4 | |
| Download: ML22112A240 (44) | |
Text
TECHNICAL SPECIFICATIONS FOR
THE
NUCLEAR TEST REACTOR FACILITY
LICENSE R-33
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TABLE OF CONTENTS 1 INTRODUCTION................................................................................................................ 1 -1 1.1 PURPOSE................................................................................................................... 1-1 1.2 DEFINITIONS.............................................................................................................. 1-1 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS...................................... 2-1 2.1 SAFETY LIMITS.......................................................................................................... 2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS (LSSS)....................................................... 2-2 3 LIMITING CONDITIONS FOR OPERATION (LCO)........................................................... 3-1 3.1 REACTOR CORE PARAMETERS.............................................................................. 3-1 3.2 REACTOR CONTROL AND SAFETY SYSTEM......................................................... 3-2 3.3 REACTOR COOLANT SYSTEM................................................................................. 3-8 3.4 CONFINEMENT.......................................................................................................... 3-9 3.5 REACTOR CELL AND VENTILATION SYSTEM........................................................ 3-9 3.6 EMERGENCY POWER............................................................................................. 3-11 3.7 RADIATION MONITORING SYSTEM....................................................................... 3-11 3.8 EXPERIMENTS......................................................................................................... 3-12 4 SURVEILLANCE REQUIREMENTS.................................................................................. 4-1 4.1 REACTIVITY LIMITS................................................................................................... 4-1 4.2 REACTOR CONTROL AND SAFETY SYSTEM......................................................... 4-2 4.3 REACTOR COOLANT SYSTEM................................................................................. 4-6 4.4 CONFINEMENT.......................................................................................................... 4-6 4.5 REACTOR CELL AND VENTILATION SYSTEM........................................................ 4-6 4.6 EMERGENCY POWER............................................................................................... 4-7 4.7 RADIATION MONITORING SYSTEM......................................................................... 4-7 4.8 EXPERIMENTS........................................................................................................... 4-7 4.9 FREQUENCY OF TESTING....................................................................................... 4-7 5 DESIGN FEAT URES.......................................................................................................... 5-1 5.1 SITE AND FACILITY DESCRIPTION.......................................................................... 5-1 5.2 REACTOR PRIMARY COOLANT SYSTEM............................................................... 5-1 5.3 REACTOR CORE AND FUEL..................................................................................... 5-1 5.4 FISSIONABLE MATERI AL STORAGE....................................................................... 5-2 6 ADMINISTRATIVE CONTROLS......................................................................................... 6-3 6.1 ORGANIZATION AND STAFFING.............................................................................. 6-3 6.2 INDEPENDENT REVIEWS......................................................................................... 6-5 6.3 RADIATION SAFETY.................................................................................................. 6-6 6.4 PROCEDURES........................................................................................................... 6-7
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6.5 EXPERIMENTS REVIEW AND APPROVAL............................................................... 6-8 6.6 REQUIRED ACTIONS................................................................................................. 6-8 6.7 REPORTS................................................................................................................... 6-9 6.8 RECORDS................................................................................................................. 6-10
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1 INTRODUCTION
1.1 PURPOSE
These Technical Specifications provide limits within which operation of the reactor will assure the health and safety of the public, the environment and on-site personnel.
Areas addressed are Definitions, Safety Limits (SL), Limiting Safety System Settings (LSSS), Limiting Conditions for Operation (LCO), Surveillance Requirements, Design Features and Administrative Controls.
1.2 DEFINITIONS
1.2.1 Channel
The combination of sensors, lines, amplifiers and output devices which are connected for the purpose of measuring the value of a parameter.
1.2.2 Channel Calibration
A comparison and/or an adjustment of the channel so that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel if reasonable, including equipment actuation, alarm, or trip test and shall include the Channel Test.
1.2.3 Channel Check
A qualitative verification of acceptable performance by observation of channel behavior.
This verification where possible shall include comparison of the channel with other independent channels or systems measuring the same variable.
1.2.4 Channel Test
The introduction or interruption of a signal into the channel to verify that it is operable.
1.2.5 Experiment
Any operation, hardware or target (excluding devices such as detectors, foils, etc.),
which is designed to investigate non-routine reactor characteristics, or which is intended for irradiation in an experiment facility and which is not rigidly secured to a core or shield structure so as to be a part of their design.
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1.2.6 Experimental Facility
Any location for experiments which is on or against the external surfaces of the reactor main graphite pack, thermal column, or within any penetration thereof.
1.2.7 Explosive Material
Any chemical compound or mixture, the primary or common purpose of which is to function by an essentially instantaneous release of gas and heat. Explosive Material in the NTR includes:
Detonating, DOT Type I Deflagrating, DOT Type II - IV
1.2.8 Facility
That portion of the building and adjacent outside areas occupied by the reactor, reactor control room, and associated support areas.
1.2.9 Flammable
A flammable liquid is any liquid having a flash point under 100°F. A flammable solid is any solid material, other than one classified as an explosive, which is liable to cause fires through friction or which can be ignited easily and when ignited burns so vigorously and persistently as to create a serious hazard. Flammable solids include spontaneously combustible and water-reactive materials.
1.2.10 Licensed Operator
A person who is licensed as a Reactor Operator (RO) or Senior Reactor Operator (SRO) pursuant to 10 CFR Part 55 to operate the controls of the Nuclear Test Reactor.
1.2.11 Limiting Conditions of Operation (LCO)
Administratively established constraints on equipment and operational characteristics that shall be adhered to during operation of the facility. The LCOs are the lowest functional capability or performance level required for safe operation of the facility.
Deviations from LCOs may be allowed under specified conditions, such as action statements.
1.2.12 Limiting Safety Sy stems Settings (LSSS)
Limiting values for settings of the safety channels by which point protective action must
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be initiated.
1.2.13 Measured Value
The measured value of a parameter is the value as it appears at the output of a channel.
1.2.14 Operable
A system or component is operable when it is capable of performing its intended function.
1.2.15 Potential Excess Reactivity
That excess reactivity which can be added by the remote manipulation of control rods plus the maximum credible reactivity addition from primary coolant temperature change plus the reactivity worth of all installed experiments.
1.2.16 Protective Action
The initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.
1.2.17 Reactivity Worth (Experiment)
The reactivity worth of an experiment is the maximum value of the reactivity change that would occur as a result of planned changes or credible malfunctions that alter experiment position or configuration.
1.2.18 Reactor Operati ng (Reactor Operation)
The reactor is operating when it is not secured or shut down (see 1.2.21 and 1.2.22).
1.2.19 Reactor Thermal Power
The reactor thermal power, as determined by a primary coolant system heat balance.
1.2.20 Reactor Safety Systems
Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
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1.2.21 Reactor Secured
The reactor is considered secured under either of the following two conditions:
- 1. The core contains insufficient fissile material to attain criticality under optimum conditions of moderation and reflection.
- 2. That overall condition where all of the following conditions are satisfied.
- a. Reactor is shut down.
- b. The console keylock switch is OFF and the key is removed from the lock.
- c. No work is in progress on core components that can directly affect core reactivity, including core fuel, the core structure, installed control or safety rods, or control rod drives unless they are physically decoupled from the control rods.
- d. No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment, or one dollar, whichever is smaller.
1.2.22 Reactor Shutdown
That subcritical condition of the reactor where the negative reactivity of the Xenon-free core would be equal to or greater than the minimum shutdown margin and the reactivity worth of all experiments is limited in accordance with Specification 3.8.3.1.
1.2.23 Readily Available Senior Reactor Operator
A Senior Reactor Operator is readily available on call when the SRO:
- a. has been specifically designated and the designation is known to the Reactor Operator on duty, and
- b. can be rapidly contacted by phone by the RO on duty, and
- c. once contacted, is capable of arriving at the NTR within a reasonable time (approximately 1/2 hour / 30-mile radius) under normal conditions.
1.2.24 Safety Limit (SL)
Limits upon important process variables which are found to be necessary to reasonably protect the reactor fuel.
1.2.25 Secured Experiment
Any experiment, experimental facility, or component of an experiment that is held in a
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stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible natural phenomena or malfunctions.
1.2.26 Shutdown Margin
Shutdown margin shall mean the shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operating condition, although the most reactive rod is stuck in its most reactive position, and that the reactor will remain subcritical without further operator action.
1.2.27 Site
The area within the confines of the Vallecitos Nuclear Center (VNC) controlled by the licensee (see Safety Analysis Report, Figure 2-3).
1.2.28 True Value
The true value for a parameter is its actual value at any instant.
1.2.29 Unsafe Condition
An unsafe condition can exist related to either nuclear safety or radiological safety. An unsafe condition relative to nuclear safety exis ts if the ability to shut down the reactor is compromised or the ability to maintain the reactor subcritical is compromised as verified in Chapter 13 analysis. An unsafe condition relative to radiological safety can only exist if any combination of failures in equipment or administrative radiological work controls results in an individual being assigned an unplanned dose greater-than-or-equal-to 100 mRem. Determination of an unsafe condition should consider the single failure of an active component or a single administrative barrier when assessing radiological safety.
1.2.30 Unscheduled Shutdown
Any unplanned shutdown of the reactor caused by actuation of the scram channels, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operat ion excluding shutdowns which occur during planned equipment testing or check-out operations.
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2 SAFETY LIMITS AND LIMITI NG SAFETY SYSTEM SETTINGS
2.1 SAFETY LIMITS
2.1.1 Applicability
This specification applies to reactor thermal power level during either forced convection or natural circulation operation.
2.1.2 Objective
The objective of this specification is to specify a reactor power safety limit which provides the basis for the LSSS.
2.1.3 Specifications
The true value of the reactor thermal power shall not exceed 190 kW under any operating condition.
Basis
Safety Limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of the NTR fuel. The only accidents which could possibly cause fuel damage and a release of fission products from the NTR fuel are those resulting from large reactivity insertions. With the 0.76$ potential excess reactivity limit, a large reactivity insertion is not possible. Therefore, there is no mechanistic way of damaging the fuel and Safety Limits should not be required (refer to SAR, Chapter 13).
The Code of Federal Regulations, however, requires a reactor to have Safety Limits.
Therefore, a Safety Limit was chosen to restrict the ratio of the actual heat flux to the Departure from Nucleate Boiling (DNB) surface heat flux in the hottest fuel element coolant passage below 1.5 to preclude any subsequent fuel damage due to a rise in surface temperature. Thermal-hydraulic analyses show that the DNB heat flux for the NTR is not significantly affected by the core flow rate or the core inlet temperature.
Reactor power is the only significant process variable that needs to be considered (refer to SAR, Chapter 13).
The safety limit for the reactor operating under steady-state or quasi steady-state conditions is 190 kW. A DNB ratio equal to 1.5 was selected as a conservatively safe operating condition for the steady-and quasi steady-state. The reactor thermal power 2-1
level when the DNBR=1.5 is 190 kW (refer to SAR, Chapter 13).
Another Safety Limit under Reactor transient conditions is not required. Conservative transient analyses show that with the potential excess reactivity limit of 0.76$, fuel damage does not occur even if all scrams fail to insert the safety rods. Although the power level may safely attain 4000 kW during this transient event (refer to SAR, Chapter 13), the Safety Limit of 190 kW was conservatively selected to apply to the transient condition.
2.2 LIMITING SAFETY SYSTEM SETTINGS (LSSS)
2.2.1 Applicability
This specification applies to the scram set point for the linear neutron channels which monitor reactor power level.
2.2.2 Objective
The objective of this specification is to ensure that automatic action will prevent the most severe postulated or anticipated transient from causing fuel damage.
2.2.3 Specification
The linear neutron power monitor channel set point shall not exceed the measured value of 125 kW.
Basis
Transient analyses presented in Chapter 13 of the SAR were performed assuming greater than 0.76$ maximum potential reactivity and an overpower scram set point at 150 kW. None of the anticipated abnormal occurrences or postulated accidents resulted in fuel damage using these values. The LSSS of 125 kW is extremely conservative for the NTR.
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3 LIMITING CONDITIONS FOR OPERATION (LCO)
3.1 REACTOR CORE PARAMETERS
3.1.1 Applicability
This specification applies to the reactivity condition of the reactor and to the reactivity worths of control rods, safety rods, manual poison sheets, experiments and the coolant temperature coefficient of reactivity.
3.1.2 Objective
The objective of this specification is to ensure the reactor can be safely controlled at all times and shut down when required.
3.1.3 Specifications
3.1.3.1.
The reactor configuration shall be controlled to ensure that the potential excess reactivity shall be 0.76$. If it is determined that the potential excess reactivity is
>0.76$, the reactor shall be shut down immediately. Corrective action shall be taken as required to ensure the potential excess reactivity is 0.76$.
3.1.3.2.
The reactor shall be subcritical whenever the four safety rods are withdrawn from the core and the three control rods are fully inserted.
3.1.3.3.
The minimum shutdown margin with the maximum worth safety rod stuck out shall be 1$.
3.1.3.4.
Each installed manual poison sheet used to satisfy the requirements of Specification 3.1.3.1 shall be restrained in its respective graphite reflector slot in a manner which will prevent movement by more than 1/2 inch relative to the reactor core.
3.1.3.5.
The temperature coefficient of reactivity of the reactor primary coolant shall be negative above a primary coolant temperature measured value of 124°F.
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Basis
Operation in compliance with Specification 3.1.3.1 ensures that there would not be any mechanism for addition of reactivity greater than 0.76$. Detailed analyses have been made of reactivity insertions in the NTR Safety Analyses Report {SAR) Chapter 13. The analyses show that a reactivity step addition of 0.76$ will not cause significant fuel degradation.
Operation in accordance with Specification 3.1.3.2 ensures that criticality will not be achieved during safety rod withdrawal. Adherence to the 0.76$ limit also ensures that the reactor will not go critical during safety rod withdrawal.
Operation in accordance with Specification 3.1.3.3 ensures that the reactor can be brought and maintained subcritical without further operator action under any permissible operating condition even with the most reactive safety rod stuck in its most reactive position.
Operation in accordance with Specification 3.1.3.4 ensures that the manual poison sheets will not be removed from the reactor core during the maximum postulated seismic event.
Operation in accordance with Specification 3.1.3.5 ensures there is no significant positive reactivity feedback from coolant temperature change during reactor power transients.
3.2 REACTOR CONTROL AND SAFETY SYSTEM
3.2.1 Applicability
This specification applies to the reactor safety rods, control rods and reactor safety systems.
3.2.2 Objective
The objective of this specification is to specify the lowest acceptable level of performance to reasonably ensure proper operation of the reactor safety rod, control rod and reactor safety systems.
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3.2.3 Specifications
3.2.3.1.
Reactor operation shall be permitted only when all safety and control rods are operable.
The reactor shall be shut down immediately if it is known that a safety or control rod is not operable.
3.2.3.2.
No more than one safety rod at a time shall be allowed to be moved in an outward direction.
3.2.3.3.
The rate of withdrawal of each safety rod during reactor operation shall be less than 1-1/4 inches per second.
3.2.3.4.
The rate of withdrawal of each control rod during reactor operation shall be less than 1/6 inch per second.
3.2.3.5.
The average scram time (inflight time) of the four safety rods shall not exceed 300 msec.
3.2.3.6.
Reactor operation shall be permitted only when the reactor safety system is operable in accordance with Tables 3-1 and 3-2.
Table 3-1 specifies automatic trip set points and scram system components necessary to ensure protective actions can be taken to shut down the reactor.
The Trip Points in Table 3-1 reflect the minimum values necessary to avoid entering the LCOs in Sections 3.1 and 3.2.
Table 3-2 specifies alarm set points and fuel rod interlock features that prompt operator actions that ensure the facility is not placed in an unsafe condition.
The reactor shall be shut down immediately if any portion of the reactor safety system malfunctions, except as provided for in Tables 3-1 and 3-2.
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Basis
Operation in accordance with Specification 3.2.3.1 ensures that during normal operation adequate shutdown margin is provided.
Operation in accordance with Specification 3.2.3.2 and specification 3.2.3.3 limits the rate of reactivity addition during safety rod withdrawal to that from one safety rod. This value is easily controlled by the operator.
Operation in accordance with Specification 3.2.3.4 limits the rate of reactivity addition during control rod withdrawal. Experience has shown that this is a value which is easily controlled manually by the operator. This rate is also less than the value analyzed in the rod withdrawal accident in the SAR.
Operation in accordance with Specification 3.2.3.5 ensures that the safety rod system performs satisfactorily. The specified time is approximately the inflight time originally established for this type reactor when higher potential excess reactivities were permitted. With the current limit on potential excess reactivity (see Technical Specification 3.1.3.1), a scram is not required during postulated events to prevent significant fuel degradation (see SAR, Chapter 13).
Maintaining the safety rod system, then, is conservative.
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Table 3-1 REACTOR SAFETY SYSTEM - SCRAM
Item Point* Function No. System Condition Trip
High reactor power 125 kW Scram (2-out-of-3 or 1-out-of-2)
- 1. Linear No less than Loss of positive high voltage to 90% of Scram (2-out-of-3 ion chambers (if used) operating or 1-out-of-2) voltage
Fast reactor period No less than +5 sec Scram
- 2. Log N Amplifier Mode switch not in operate N/A Scram
No less than Loss of positive high voltage to 90% of Scram ion chambers (if used) operating voltage Primary Coolant
- 3. Temperature High core outlet temperature 222 °F Scram (Fenwall)
No less than
- 4. Primary Coolant 15 gpm when Scram Flow Low Flow reactor power
> 0.1 kW
- 5. Manual Console button depressed N/A Scram
Reactor console key in off
- Trip points are the nominal measured values and need not consider the uncertainty in the channel.
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Table 3-2 REACTOR SAFETY-RELATED COMPONENTS
Item Point* Function No. System Condition Set
Low Differential > 0.5 in. Visible and audible alarm; audible
- 1. Reactor Cell Pressure pressure water P alarm may be bypassed after recognition.
Fuel Loading Visible and audible alarm; audible 2 Tank Water Low Flow < 3 ft. below the overflow alarm may be bypassed after Level recognition.
Primary High core outlet Visible and audible alarm; audible
- 3. Coolant temperature <200°F alarm may be bypassed after Temperature recognition.
Primary Core Delta
- 4. Coolant temperature N/A Provide information for the heat balance determination Temperature At a level to ensure Visible and audible alarm; audible
- 5. Stack compliance alarm may be bypassed after Radioactivity High Level with TS recognition.
3.5.3.3 and 3.5.3.4
- 6. Linear Power Low Power 2% on any Safety or control rods cannot be indication scale withdrawn 2-out-of-3 or 2-out-of-2).
- 7. Control or Safety Rod Rods not in N/A Safety rod magnets cannot be reenergized
Control rods cannot be withdrawn; safety rods must be withdrawn in sequence; may be bypassed to
- 8. Safety Rod Rods not out N/A allow withdrawal of one control rod, or one safety rod (drive) out of sequence for purposes of inspection, maintenance, and testing
- Set Point values are the nominal measured values and need not consider the uncertainty of the channel.
Operation in accordance with Specification 3.2.3.6 ensures that the reactor safety system is adequate to control operation of the facility, measure operating parameters, warn of abnormal conditions, and scram the reactor automatically if required.
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The bases for items listed in Table 3-1 are as follows:
The linear high reactor power scram will be set no higher than the LSSS. Scram action as a result of a predetermined decrease of positive high voltage to ion chambers for the linear channels provides assurance that the high voltage power supply is functioning, and the ion chambers are operating in the ionization region of the gas amplification curve.
The fast period scram limits the rate of rise of the reactor power to periods which are manually controllable. The Log N amplifier mode switch scram ensures that the Log N amplifier is in the Operate Mode. Scram action as a result of loss of positive high voltage to the ion chamber for the Log N channel provides assurance that the high voltage power supply is functioning, and the ion chamber is operating in the ionization region of the gas amplification curve.
The primary coolant high core outlet temperature scram provides assurance that the reactor will be shut down if the primary coolant outlet temperature is high.
The primary coolant low-flow scram provides diversification in the safety system to ensure, when the reactor is at power levels which require forced cooling, that the reactor will be shut down if sufficient primary coolant flow is not maintained.
The manual console scram button provides a method for the reactor operator to manually shut down the reactor if an unsafe or abnormal condition should occur and the automatic reactor protection action as appropriate does not function. The loss of electrical power with the reactor console key in the off position (loss of ac power to the console) means that the reactor cannot be operated because ac power is no longer provided to the reactor safety system.
The bases for items listed in Table 3-2 are as follows:
The reactor cell low differential pressure alarm alerts the operator that reactor power must be lowered below 0.1 kW according to specification 3.5.3.1. Remedial action may be taken to correct the condition prior to shutting down the reactor.
The fuel loading tank low water level alarm alerts the operator to verify that the core tank is filled prior to exceeding 0.1 kW according to specification 3.3.3.1, or to shut down the reactor according to specification 3.3.3.2 if operating at > 0.1 kW.
The primary coolant high core outlet temperature alarm alerts the operator prior to
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reaching a high core outlet temperature trip and has no associated LCO.
Core delta temperature is the difference between the core outlet (TC-2) and core inlet (TC-5) thermocouples.
The stack radioactivity high level alarm gives adequate assurance that operation of the reactor will comply with specification 3.5.3.2. The alarm alerts the operator that action is necessary to ensure discharges stay within the limits specified in Table 3-3.
The control rods "not in" interlock ensures that the reactor will be started up by withdrawing the four safety rods prior to withdrawing the control rods and must be functional prior to startup.
The safety rods "not-out" interlock ensures that the method of reactivity control is with the control rods during reactor operation and must be functional prior to startup.
3.3.1 Applicability
This specification applies to the reactor primary coolant system.
3.3.2 Objective
The objective of this specification is to minimize the adverse effects on reactor components and to ensure the proper conditions of the coolant system for reactor operation.
3.3.3 Specifications
3.3.3.1.
Above 0.1 kW the reactor shall be cooled by light water forced coolant. At or below 0.1 kW forced coolant flow is not required.
3.3.3.2.
Reactor operation shall not be permitted unless the core tank is filled with water. If during operation of the reactor it is determined that the core tank is not filled with water, the reactor will be shut down immediately and corrective action will be taken as required.
3.3.3.3.
The specific conductivity of the primary coolant water shall be maintained less than 10 3-8
µS/cm except for time periods not exceeding 7 consecutive days when the specific conductivity may exceed 10 µS/cm but shall remain less than 20 µS/cm. If the specific conductivity exceeds 10 µS/cm, steps shall be taken to assure the specific conductivity is reduced to less than 10 µS/cm.
Basis
During a complete loss of primary coolant flow without a reactor scram, fuel damage does not occur (SAR,Section I 3.4.5). Natural convection cooling is sufficient. Requiring forced coolant flow above 0.1 kW, then, is extremely conservative.
Operation in accordance with Specification 3.3.3.2 ensures that there will be no reactivity insertions due to the removal of voids or the sudden addition of water into the core tank during reactor operation.
The minimum corrosion rate for aluminum in water (< 50°C) occurs at a pH of 6.5.
Maintaining water purity below 10 µS/cm will maintain the pH between 4.8 and 8.7.
These values are acceptable for NTR operation. High specific conductivity can be tolerated for shorter durations during unusual circumstances. Operation in accordance with Specification 3.3.3.3 ensures aluminum corrosion is within acceptable levels and that activation of impurities in the primary water remain below hazardous levels.
3.4 CONFINEMENT
3.4.1 Applicability
The scenarios analyzed in Chapter 13 of the NTR Safety Analysis Report conservatively demonstrate that building confinement is not necessary to ensure radiological doses will not exceed 10 CFR 20 allowable limits. The reactor cell works in conjunction with the ventilation system to ef fectively limit small radioactive releases.
3.5 REACTOR CELL AND VENTILATION SYSTEM
3.5.1 Applicability
This specification applies to the reactor cell and ventilation system.
3.5.2 Objective
The objective of this specification is to limit the release of airborne radioactive materials to the environment.
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3.5.3 Specifications
3.5.3.1
Reactor power shall not be increased above 0.1 kW unless the reactor cell is maintained at a negative pressure of not less than 0.5 in. of water with respect to the control room. If during operation of the reactor above 0.1 kW, the negative pressure with respect to the control room is not maintained, the reactor power shall be lowered to 0.1 kW immediately and corrective action shall be taken as required.
3.5.3.2
The limits for radioactive material discharged through the reactor ventilation system to the atmosphere shall be as specified in Table 3-3.
Table 3-3
STACK RELEASE RATE LIMITS
Isotope Group Annual Average
Halogen, > 8-day Half-life 180 mCi/wk
Particulate > 8-day Half-life:
Beta-Gamma 870 µCi/wk
Alpha 8.7 µCi/wk
All Other (including Noble Gas) 9 Ci/wk
3.5.3.3
Alarm set points for particulate and noble gas continuous monitors shall not exceed a value corresponding to the annual average release rate limit shown in Table 3-3.
3.5.3.4
The stack particulate activity monitor and the gaseous activity monitor should be operating when the reactor is operated above 0.1 kW. If either monitor is not functional:
- a. All evolutions that could precipitate airborne releases should be discontinued within the facility.
- b. The failed monitor should be restored to functionality within a reasonable amount of
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time (i.e., by the end of the run or at the discretion of management).
If these actions cannot be completed, the reactor shall be shut down and shall not be returned to operation above 0.1 kW until both monitors are functional.
Basis
Operation in accordance with Specification 3.5.3.1 ensures that potentially contaminated reactor cell air due to reactor operation is released and monitored through the ventilation system.
The ventilation system release limits in Specification 3.5.3.2 are based on the annual average dilution factor from the NTR stack to the site boundary and was developed from meteorological date collected on site over a two-year period. A stack flow rate of 1,800 cu ft/min yields an annual dilution factor of about 33,000. That is, the concentration at the site boundary from a continuous uniform release from the NTR stack will be 1/33,000 of the concentration at the stack when averaged over 1 year. These annual average limits (Table 3-3) include a reduction factor of 2 to account for discharges from other VNC stacks.
The alarm set points in Specification 3.5.3.3 are set for the annual average release rate limit of the most restrictive isotope in all categories except for the category of noble gases, which uses the most probable isotope, Ar-41.
3.6 EMERGENCY POWER
3.6.1 Applicability
Then NTR has no emergency power systems.
3.7 RADIATION MO NITORING SYSTEM
3.7.1 Specifications
3.7.1.1
The radiation monitors ensure protection of personnel by contributing to maintaining exposures ALARA. Remote Area Monitors are located at the:
North Room (adjacent to the CHRIS)
South Cell Reactor Cell
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Control Room North Room (MSM)
3.7.1.2
A functioning radiation monitor should be in service in the subject room during an experiment irradiated sample unloading evolution. In lieu of a functioning radiation monitor, portable instruments, surveys, or analyses may be substituted for any of the normal installed monitors in section 3.7.1.1 for periods of one week or for the duration of a reactor run in cases where the reactor is continuously operated. Irradiation experiments may continue in that room once a functioning radiation monitor is in service or if reentry dose rates are calculated and decay times established prior to each irradiation experiment and portable instruments are used upon reentry.
3.8 EXPERIMENTS
3.8.1 Applicability
This specification applies to reactor experiments.
3.8.2 Objective
The objective of this specification is to prevent an experiment from resulting in a hazard to the operating personnel or the general public or damage to the reactor.
3.8.3 Specifications
3.8.3.1
The reactivity worth of all experiments shall be limited so that the sum of the reactivity worths of all experiments performed at any one time shall be limited to comply with Specification 3.1.3.1.
3.8.3.2
- a. The amounts of explosives (detonating and deflagrating, DOT Hazard Class/Divisions 1.1, 1.2, 1.3 and 1.4) permitted in the NTR facilities are as follows:
- i. South Cell, W (D/2)2 with W 9 lbs and D 3 ft.
ii. North room (without Modular Stone Monument), W D2 with W 16 lbs and D 1ft.
iii. Setup Room, W 25 lbs.
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- b. The amounts of explosives allowed in the North room MSM (inclusive in the limit of a.ii. above) are as follows:
- i. for DOT Hazard Class Divisions 1.1, 1.2, and 1.3 (detonating): W 2 pounds ii. for DOT Hazard Class Division 1.4 (deflagrating): W 4 pounds
where: W = Total weight of explosives in pounds of equivalent TNT.
D = Distance in feet from the South Cell blast shield or the North Room wall.
3.8.3.3
Experimental objects shall not be allowed inside the core tank when the reactor is at a power greater than 0.1 kW.
3.8.3.4
Experimental objects located in the fuel loading chute shall be secured to prevent their entry into the core region during reactor operation.
3.8.3.5
A maximum of 10 Ci of radioactive material and up to 50 g of uranium may be in storage in a neutron radiography area where explosive devices are present (i.e., in the South Cell or North Room). The storage locations must be at least 1.5 m (5 ft) from any explosive device.
Radioactive materials, other than byproduct irradiated explosive devices and imaging systems, are not permitted in the Setup Room if explosive material is present.
Exception. Devices containing not more than 10 grams TNT equivalent of explosives with up to 200 mCi of tritium in the form of tritiated metal (hydride) are permitted.
However, no more than one device may be in a neutron radiography area or the setup room at any one time, and no other explosive material may be in the same area at that time.
3.8.3.6
No explosive device shall be placed in a radiation field greater than 1 x 104 roentgens or consisting of greater than 3 x 1012 n/cm2 thermal neutrons.
3.8.3.7
With the exception of communication equipment utilizing low-energy electromagnetic 3-13
waves in radiofrequencies, such as mobile phones and two-way hand-held radios, unshielded high-frequency generating equipment shall not be operated within 50 feet of any explosive device.
3.8.3.8
Experimental capsules to be utilized in the experimental facilities shall be designed or tested to ensure that any pressure transient produced by chemical reaction of their contents and/or leakage of corrosion or flammable materials will not damage the reactor.
3.8.3.9
Experimental fuel elements containing uranium to be utilized in the experimental facilities shall be clad.
3.8.3.10
The maximum possible chemical energy release from the combustion of flammable substances contained in any experimental facility shall not exceed 1000 kW-sec. The total possible energy release from chemical combination or decomposition of substances contained in any experimental capsule shall be limited to 5 kW-sec, if the rate of the reaction in the capsule could exceed 1 W. Experimental facilities containing flammable materials shall be vented external to the reactor graphite pack.
3.8.3.11
A written description and analysis of the possible hazards involved for each type of experiment shall be evaluated and approved by the area manager, or his designated alternate, before the experiment may be conducted.
3.8.3.12
No irradiation shall be performed which could credibly interfere with the scram action of the safety rods at any time during reactor operation.
3.8.3.13
The radioactive material content, including fission products, of any singly encapsulated experiment to be utilized in the experimental facilities shall be limited, so that the complete release of all gaseous, particulate, or volatile components from the encapsulation could not result in doses in excess of 10% of the equivalent annual doses
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stated in 10 CFR Part 20. This dose limit applies to persons occupying unrestricted areas continuously for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> starting at time of release or restricted areas during the length of time required to evacuate the restricted area.
3.8.3.14
The radioactive material content, including fission products, of any doubly encapsulated or vented experiment to be utilized in the experimental facilities shall be limited so that the materials at risk from the encapsulation or confining boundary of the experiment could not result in a dose to any person occupying an unrestricted area continuously for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> starting at the time of release in excess of 0.5 rem to the whole body or 1.5 rem to the thyroid or a dose to any person occupying a restricted area during the length of time required to evacuate the restricted area in excess of 5 rem to the whole body or 30 rem to the thyroid.
Basis
Operation in accordance with Specification 3.8.3.1 ensures that there would not be any mechanism for addition of reactivity greater than 0.76$, including experiments. See the bases for Specification 3.1.3.1.
Specifications 3.8.3.1 through 3.8.3.11 are intended to reduce the likelihood of damage to the reactor components and/or radioactivity releases resulting from experiment failure and serve as a guide for the review and approval of new and untried experiments by the facility personnel.
Specifications 3.8.3.5 assures that any radiological effects in storage areas will not pose hazards to the public.
Specifications 3.8.3.12 and 3.8.3.13 ensure the radiological effects of experiment failures do not pose a hazard to the general public or to operating personnel.
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4 SURVEILLANCE REQUIREMENTS 4.1 REACTIVITY LIMITS
4.1.1 Applicability
This specification applies to the surveilla nce requirements for reactivity limits.
4.1.2 Objective
To ensure that the reactivity limits of Specification 3.1 are not exceeded.
4.1.3 Specification
4.1.3.1
Potential excess reactivity will be calculated before each startup. Actual critical rod position shall then be used to verify that the measured value is 0.76$.
4.1.3.2
Neutron multiplication will be observed throughout each startup. Safety rod withdrawal shall be stopped if it appears criticality will be reached before all safety rods are out.
4.1.3.3
The minimum shutdown margin shall be determined by calculation or measurement whenever a decrease in the reactivity worth of a safety rod is suspected.
4.1.3.4
Each manual poison sheet in the core region of the reactor shall be verified to be properly restrained upon insertion.
4.1.3.5
The temperature coefficient of reactivity of the reactor primary coolant shall be verified to be negative above 124 °F whenever changes made to the reactor could affect the temperature coefficient.
Basis
Operation in accordance with Specification 4.1.3.1 will ensure that the reactor is not operated with a potential excess reactivity of >0.76$.
Operation in accordance with Specification 4.1.3.2 will ensure that the reactor will be
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subcritical when all the safety rods are in the full-out position with control rods inserted.
Minimum shutdown margin is assured when the potential excess reactivity is limited to 76¢ and safety rod reactivity worths are unchanged. The shutdown margin, then, should be determined as specified in Specification 4.1.3.3 when changes to the reactor are made which could decrease the reactivity worth of a safety rod.
Verification that the manual poison sheets are properly restrained as specified in Specification 4.1.3.4 ensures that they cannot be ejected during any postulated natural phenomena or operational occurrence.
Compliance with Specification 4.1.3.5 ensures that the temperature coefficient is negative above 124°F. It is not affected by reactor configuration and fuel burnup and is therefore not expected to vary significantly with core life (but could be affected by fuel, core or moderator design changes).
4.2 REACTOR CONTROL AND SAFETY SYSTEM
4.2.1 Applicability
This specification applies to the surveilla nce requirements for the reactor control and reactor safety systems.
4.2.2 Objective
The objective of this specification is to specify the minimum surveillance requirements to reasonably ensure proper performance of the safety rod, control rod and safety systems.
4.2.3 Specifications
4.2.3.1
Each safety rod and control rod drive shall be tested for operability annually.
4.2.3.2
The interlock which restricts safety rod withdrawal to one rod at a time shall be tested annually.
4.2.3.3
The rate of withdrawal of each safety rod shall be measured annually.
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4.2.3.4
The rate of withdrawal of each control rod shall be measured annually.
4.2.3.5
The safety rod scram time (inflight time) shall be measured semi-annually. The scram time (inflight time) shall additionally be measured after any work is performed which could affect the scram time or rod travel time.
4.2.3.6
Checks, tests and calibrations of the reactor safety system shall be performed as specified in Tables 4-1 and 4-2.
4.2.3.7
A thermal power verification shall be performed monthly when the reactor is operating above 50 kW.
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Table 4-1
SURVEILLANCE REQUIREMENTS OF REACTOR
SAFETY SYSTEM SCRAM INSTRUMENTS
Item No. Item Surveillance Frequency*
Channel Check (neutron source check) Daily
Channel Test (high level trip test) Daily
- 1. Linear System Channel Check (comparison against a heat
balance Semi-annually
Channel Calibration Annual
Channel Test Daily
- 2. Log N System Channel Check Monthly
Channel Calibration Annually
Primary Coolant Channel Test Daily
- 3. Temperature (Fenwall) Channel Calibration Annually
Channel Check Daily
- 4. Primary Coolant Flow Channel Test Daily
Channel Calibration Annually
- 5. Manual Channel Test Daily
- 6. Electrical Power Channel Test Daily
- Prior to placing into service an instrument which has been repaired, the instrument check, or test or calibration, as appropriate will be performed.
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Table 4-2
SURVEILLANCE REQUIREMENTS OF REACTOR SAFETY-RELATED ITEMS (INFORMATION INSTRUMENTS)
Item Frequency*
No. Item Surveillance
- 1. Reactor Cell Pressure Channel Test Quarterly
- 2. Fuel Loading Tank Water Level Channel Test Quarterly
Channel Test Quarterly
- 3. Primary Coolant Temperature (TC-7)
Channel Calibration Annually
Channel Check Quarterly
- 4. Primary Coolant Conductivity Channel Calibration Biennially
Channel Check Monthly
- 5. Primary Coolant Core Temperature (TC2 & TC5)
Channel Calibration Annually
Channel Check Daily
- 6. Stack Radioactivity (Gas and particulate channels) Channel Test Monthly
Channel Calibration Annually
- 7. Linear Power Channel Test Monthly
- Prior to placing into service an instrument which has been repaired, the instrument check, or test or calibration, as appropriate will be performed.
Basis
Specification 4.2.3.1 ensures that each safety and control rod is maintained operable.
Specification 4.2.3.2 ensures that the safety rod interlock preventing the simultaneous withdrawal of more than one safety rod functions properly.
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Specifications 4.2.3.3 and 4.2.3.4 ensure that the control and safety rod withdrawal rates are within limits.
Specification 4.2.3.5 provides for the periodic measurement of safety rod insertion times to ensure they are within limits.
Specification 4.2.3.6 ensures that the safety system is periodically tested and checked to maintain all instruments operable.
Specifications regarding surveillance requirements of the reactor coolant system are included in the reactor safety system, Specification 4.2, Tables 4-1 and 4-2.
4.4 CONFINEMENT
Surveillance requirements related to confinement are limited to the ventilation system and are included in Section 4.5.
4.5 REACTOR CELL AND VENTILATION SYSTEM
4.5.1 Applicability
This specification applies to the surveilla nce requirements for the reactor cell and ventilation system.
4.5.2 Objective
The objective of this specification is to ensure that the reactor ventilation system is in satisfactory condition to provide adequate confinement and to control the release of radioactivity to the environment.
4.5.3 Specification
4.5.3.1
The reactor cell negative pressure, with respect to the control room, shall be verified prior to the first reactor startup of each day.
4.5.3.2
Surveillance requirements for the instrumentation and equipment required to comply with Specifications 3.5.3.2, 3.5.3.3, and 3.5.3.4 shall be as listed in Specification 4.2, Table 4-2.
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Basis
Operation in accordance with Specification 4.5.3.1 ensures that contaminated reactor cell air is exhausted through the ventilation system. This minimizes the possibility of airborne contamination release to surrounding areas.
Operation in accordance with Specification 4.5.3.2 ensures that all required channels are operable, and that proper notification and surveillance will occur.
4.6 EMERGENCY POWER
Then NTR has no emergency power systems.
4.7 RADIATION MO NITORING SYSTEM
Surveillances for the area Radiation Monitors are performed according to procedural guidance at the prescribed frequencies.
4.8 EXPERIMENTS
Specific surveillance activities shall be established during the review and approval process as specified in Section 6.2.3 "Review Function" and are not part of the Technical Specifications.
4.9 FREQUENCY OF TESTING
4.9.1 Applicability
This specification applies to all surveillance requirements in Section 4 of these Technical Specifications.
4.9.2 Objective
The objective of this specification is to establish maximum time intervals for surveillance periods. It is intended that this specification provides operational flexibility and not reduce surveillance frequency.
4.9.3 Specifications
4.9.3.1
Time intervals used elsewhere in these specifications shall be defined as follows:
- a. Biennially - Interval not to exceed 30 months.
- b. Annually - Interval not to exceed 15 months.
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- c. Semi-annual - Interval not to exceed 32 weeks.
- d. Quarterly - Interval not to exceed 18 weeks.
- e. Monthly - Interval not to exceed 6 weeks.
- f. Weekly - Interval not to exceed 10 days.
- g. Daily - Must be done prior to the first startup of the calendar day following a shutdown greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.9.3.2
Surveillance tests (except those required for safety while the reactor is shut down) may be deferred during a reactor shutdown. Deferred surveillance tests must be completed prior to reactor startup.
4.9.3.3
Surveillance tests scheduled to occur during reactor operation, which cannot be performed with the reactor operating, may be deferred until the subsequent scheduled reactor shutdown.
Basis
Specification 4.9.3.1 establishes maximum time intervals for surveillance requirements which define the terms and makes them objectively quantifiable.
Specification 4.9.3.2 permits deferring tests which are not required if the reactor will not be operating.
Specification 4.9.3.3 permits deferring tests which might require a reactor shutdown for the sole purpose of performing the test.
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5 DESIGN FEATURES
5.1 SITE AND FACILITY DESCRIPTION
5.1.1
The Nuclear Test Reactor (NTR) facility shall be located on the site of the Vallecitos Nuclear Center (VNC).
5.1.2
The controlled area, as defined in 10 CFR Part 20 of the Commissions regulations, shall be the area within the VNC site boundary. The restricted area, as defined in 10 CFR Part 20 of the Commissions Regulations, is the NTR facility.
5.1.3
The fuel assemblies shall be positioned in a reel assembly inside the core tank. The core reel assembly shall be rotated only when the reactor is shut down and by manual operation of a crank inside the NTR cell.
5.1.4
The control system shall consist of four scrammable, spring-actuated safety rods, three nonscrammable control rods, and a number of manual poison sheets. When the poison rods and sheets are inserted, they shall be located in the graphite reflector at the outer periphery of the core tank. The safety and control rods shall be boron carbide clad in stainless steel. The manual poison sheets shall contain metallic cadmium.
5.1.5
The discharge of the gaseous effluent stack shall be approximately 14 meters (45 feet) above grade level of Building 105.
5.2 REACTOR PRIMARY COOLANT SYSTEM
The reactor coolant system shall be protected from overpressure by a vent line to the atmosphere of the cell.
5.3 REACTOR CORE AND FUEL
The core shall consist of 16 fuel element assemblies. Each fuel element assembly shall consist of 40 disks spaced on an aluminum support shaft. Other nominal specifications of the assemblies shall include the following:
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- a. Fuel 23.5% (by weight uranium) / 76.5%
- b. Enrichment Approximately 93% U-235 (unburned)
- c. Cladding Aluminum, 0.022-inch thickness
- d. Fuel disk active diameter 2.685 inch
- e. Fuel disk spacing on shaft 0.35 to 0.45-inch, center-to-center
5.4 FISSIONABLE MATERIAL STORAGE
Fuel including fueled experiments and fuel devices not in the reactor shall be stored in a geometrical array where k eff is no greater than 0.9 for all conditions of moderation and reflection using light water.
BASIS
The bases for the items in sections 5.1 to 5.4 are as follows:
The fixed NTR core configuration and control element arrangement ensures a temperature coefficient turnover from positive to negative above the operational coolant temperature of 124°F and yields a negative void coefficient above that temperature.
Ensuring the reactor primary coolant system is vented to atmosphere protects the system from overpressure damage.
Loss of coolant will not result in damage to the fuel system comprised of this proven clad, metallic fuel plates. Neglecting natural convection air cooling of the fuel plates, the loss-of coolant inventory from the reactor results in a worst-case fuel temperature peak at about 800°F about 20 minutes after coolant loss at which time it begins to decline. At this peak, the commensurate power is 1.5 kW, which can be tolerated indefinitely without increasing graphite temperatures to over 150°F. Limiting excess reactivity to 0.76$ (TS 3.1.3.1) ensures a step reactivity insertion also will not cause fuel damage; even with a failure to scram, operation of the reactor will not pose a threat to the health and safety of the public.
The effluent stack is of sufficient height to disperse the exhaust upward.
Limits imposed in 5.4 on the fueled experiments and fuel devices not in the reactor are
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conservative and ensures safe storage.
6 ADMINISTRATIVE CONTROLS
6.1 ORGANIZATION AND STAFFING
6.1.1 Structure
The NTR shall be owned and operated by the licensee with management and operations organization as shown in Figure 6-1 or equivalent.
6.1.2 Responsibilities
6.1.2.1
The Level 1 manager shall be responsible for the NTR facility License.
6.1.2.2
The Level 2 manager (Operations) is designated the area manager and shall be responsible for the overall safe oper ation and maintenance of the facility.
6.1.2.3
The Level 3 Reactor Supervisor (if utilized) is the individual responsible for supervising daily operations. In the absence of this position, the Level 2 area manager is responsible for supervising daily operations.
6.1.2.4
Responsibilities of one level may be assumed by alternates when designated in writing.
6.1.2.5
Functions performed by one level may be performed by a higher level, provided the minimum qualifications are met (e.g., Senior Reactor Operator license).
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Figure 6-1. Facility Organization
6.1.3 Staffing
6.1.3.1
The minimum staffing when the reactor is not secured shall be composed of:
- a. A licensed operator in the control room.
- b. A second person present at the site who is familiar with the VNC Radiological Emergency Plan and Emergency Procedures relevant to the NTR and is capable of carrying out facility written procedures.
- c. A licensed Senior Reactor Operator shall be present at the VNC Emergency NTR Facility or readily available.
6.1.3.2
A licensed Senior Reactor Operator shall be present at the NTR Facility during the following events:
- a. During the recovery from an unscheduled shutdown.
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- b. During reactor fuel loading or reactor fuel movement.
- c. During any experiment or facility changes with a reactivity worth greater than one dollar.
6.1.4 Selection and Training of Personnel
The selection, training and requalification of operations personnel shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Reactors, ANSI/ANS 15.4-2016, Sections 4 through 6; Title 10 of the Code of Federal Regulations, Part 55; and the latest revision of the Facility Operator Requalification Program.
6.2 INDEPENDENT REVIEWS
6.2.1
Independent reviews are performed by the cognizant Nuclear Safety Review Groups responsible to the Level 1 manager.
6.2.2
The independent review function shall be performed under a written charter or directive containing the following information as a minimum:
- a. Subjects reviewed
- b. Responsibilities
- c. Authorities
- d. Records
- e. Other matters as may be appropriate
6.2.3
Activities requiring independent review shall include the following:
- a. Proposed types of tests and experiments (or substantive changes thereto) including safety evaluations that could affect core reactivity or result in an uncontrolled release of radioactivity that do not require prior NRC approval pursuant to 10 CFR 50.59, to ensure the proposed activity does not require a change in the Technical Specifications.
- b. Proposed changes to the procedures or the facility as described in the Safety Analysis Report including safety evaluations that do not require prior NRC
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approval, pursuant to 10 CFR 50.59, to verify the activity does not constitute a change in the Technical Specifications.
- c. Proposed changes to the Fire Protection program as described in the Safety Analysis Report that do not require prior NRC approval, to verify the change would not adversely affect the ability to achieve and maintain safe shutdown of the NTR in the event of a fire.
- d. Except for minor or administrative changes, all new procedures and revisions of existing procedures having safety significance that are required by the specifications in Section 6.4.
- e. Proposed changes to the Technical Specifications or the facility operating license.
- f. Violations of the Federal Regulations, Technical Specifications, and facility license requirements.
- g. Unusual or abnormal occurrences which are reportable to the NRC under provisions of the Federal Regulations or the Specifications in Section 6.6.
- h. Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect, or could affect, nuclear safety.
6.2.4
Independent periodic examination and verification shall be performed of facility operations, maintenance and administration. These periodic examinations and verifications shall be performed by staff that do not have direct responsibility for the safe operation of the reactor.
6.3 RADIATION SAFETY
6.3.1
The radiation safety program must achieve the requirements of 10 CFR Part 20.
6.3.2
Safety is foremost at the facility. Regulatory compliance personnel have the authority to intercede and suspend activities which could involve or result in radiologically hazardous situations.
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6.3.3
The ALARA program shall be applied to all facility staff, facility. users, visitors, the public and the environment.
6.4 PROCEDURES
6.4.1
Written procedures shall be prepared for the following activities as required:
- a. Startup, operation, and shutdown of the reactor.
- b. Defueling, refueling, and fuel transfer operations, when required.
- c. Preventive or corrective maintenance which could have an effect on the safety of the reactor.
- d. Off-normal conditions relative to reactor safety for which an alarm is received.
- e. Response to abnormal reactivity changes.
- f. Surveillance testing, and calibrations required by the Technical Specifications.
- g. Emergency conditions involving potential or actual release of radioactive materials.
- h. Radiation protection consistent with 10 CFR Part 20 requirements.
- i. Review and approval of changes to all required procedures.
- j. Security plan, the operator requalification program, and emergency procedures.
- k. Operation and maintenance of experiments that could affect reactor safety or core reactivity.
6.4.2
The Level 2 area manager shall approve all procedures (including revisions) required by Specification 6.4.1 before implementation.
6.4.3
Minor changes to the original procedures which do not change their original intent may be made by the Level 3 Reactor Supervisor or Level 2 manager. Changes made by approval of the Level 3 Reactor Supervisor must be subsequently approved by the
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Level 2 area manager.
6.4.4
Temporary deviations from established procedures may be made by a Licensed Senior Reactor Operator in order to deal with special or unusual circumstances. These deviations shall be documented and reported to the area manager.
6.5 EXPERIMENTS REVIEW AND APPROVAL
6.5.1 New Experiments
All new experiments or class of experiments shall undergo independent review according to 6.2.3 and approved in writing by the Level 2 area manager or designee.
6.5.2 Previously Approved Experiments
- 1. Substantive changes made to previously approved experiments shall undergo independent review according to 6.2.3 and be approved in writing by the Level 2 manager or designee.
- 2. Changes made to previously approved experiments that do not significantly alter the experiment do not require independent review and may be approved by Level 3 manager or higher.
6.6 REQUIRED ACTIONS
6.6.1 Action to be Taken in Case of Safety Limit Violation
- 1. The reactor shall be shut down, and reactor operations shall not be resumed until authorized by Level 1 management.
- 2. The safety limit violation shall be promptly reported to the Level 2 area manager.
- 3. The safety limit violation shall be reported to the NRC.
- 4. A safety limit violation report shall be prepared. The report shall describe the following:
- a. Applicable circumstances leading to the violation including, when known, the cause and contributing factors.
- b. Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public.
- c. Corrective action to be taken to prevent recurrence.
- 5. The report shall be reviewed by Regulatory Compliance and any follow-up report
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shall be submitted to the NRC when authorization is sought to resume operation of the reactor.
6.6.2 Action to be taken in the event of an occurrence of the type Identified in Section 6.6.2.a.1 or 6.6.2.a.2.
- 1. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Level 2 area manager or the Level 1 manager.
- 2. Occurrence shall be reported to the area manager and to the NRC addressed in accordance with 10 CFR 50.4.
- 3. Occurrence shall be reviewed by Regulatory Compliance.
6.7 REPORTS
6.7.1 Operating Reports
Annual operating report(s) shall be submitted to the NRC Document Control Desk. The report(s) shall include the following:
- a. A narrative summary of reactor operating experience including the hours the reactor was critical and total energy produced.
- b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence.
- c. Tabulation of major preventive and corrective maintenance operations having safety significance.
- d. A summary report in accordance with 10 CFR 50.59.
- e. A summary of the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge.
- f. Summarized results of environmental su rveys performed outside the facility.
- g. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
6.7.2 Special Reports
Special reports are used to report unplanned events as well as planned major facility and administrative changes. The following special reports shall be forwarded to the NRC addressed in accordance with 10 CFR 50.4:
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- a. There shall be a report not later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance, to be followed by a written report within 14 days, that describes the circumstances of any of the following events:
- 1. Release of radioactivity from the site above allowed limits.
- 2. Any of the following:
Operation with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in the technical specifications.
Operation in violation of limiting conditions for operation established in the technical specifications unless prompt remedial action is taken.
A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or c ondition is discovered during maintenance tests or periods of reactor shutdowns. (Note: Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems or not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)
An unanticipated or uncontrolled change in reactivity greater than 0.50$.
Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary, which could result in exceeding prescribed radiation limits for personnel or the environment.
An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.
- b. A written report within 30 days to the NRC for the
- 1. Permanent changes in the facility organization involving Level 1 or Level 2 management.
- 2. Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
6.8 RECORDS
Records may be in the form of logs, data sheets, or other suitable forms. The required
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information may be contained in single, or multiple records, or a combination thereof.
6.8.1
Records to be retained for a period of at least five years or for the life of the Component, whichever is less:
- a. Normal reactor facility operation (supporting documents such as checklists, log sheets, etc., shall be maintained for a period of at least one year).
- b. Principal maintenance operations.
- c. Reportable occurrences.
- d. Surveillance activities required by the Technical Specifications.
- e. Reactor facility radiation and contamination surveys where required by applicable regulations.
- f. Experiments performed with the reactor.
- g. Fuel inventories, receipts, and shipments.
- h. Approved changes in operating procedures.
- i. Records of meeting reports of the review groups.
6.8.2
Records of the requalification programs shall be maintained in accordance with 10 CFR 55.59(c)(5).
6.8.3
Records to be Retained for the Lifetime of the Reactor Facility.
(Note: Applicable annual reports, if they contain all the required information, may be used as records in this section.)
- a. Gaseous and liquid radioactive effluents released to the environs.
- b. Radiation exposure for all personnel monitored.
- c. Drawings of the reactor facility.
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