ML20003H131
| ML20003H131 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos Nuclear Center |
| Issue date: | 04/30/1981 |
| From: | Leighty C GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20003H129 | List: |
| References | |
| NEDO-12725, NUDOCS 8105050173 | |
| Download: ML20003H131 (67) | |
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NED0-12725 80NED026 Class I April 1981 t
PROPOSED TECHNICAL SPECIFICATIONS FOR THE i GENERAL ELECTRIC NUCLEAR TEST REACTOR 4
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i FACILITY LICENSE R-33 DOCKET No. SD-73 4
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PROPOSED TECHNICAL SPECIFICATIONS FOR THE GENERAL ELECTRIC NUCLEAR TEST REACTOR These Technical Specifications apply to the General Electric Nuclear Test Reactor (Facility License R-33, Docket No. 50-73) and represent those parameters which define the boundaries of licensed activities. While not a part of these specifications, additional bases for these specifications are included in the Nuclear Test Reactor Safety Analysis Report,NEDO-12727.
The dimensions, measurements, and other numerical values given in these specifications may differ slightly from actual values owing to normal construction and manufacturing tolerances, or normal accuracy of instrumentation.
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NEDO-12725 CONTENTS P!!ge, 1.0 DEFINITIONS 1-1
) 1.1 Reactor Safety 1-1 1.2 Reactor Operations 1-2 1.3 Reactor Experiments 1-5 1.4 Maintenance and Testing 1-8 1.5 General 1-10 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS 2-1 2.1 Safety Limits 2-1 2.2 Limiting Safety System Settings 2-2 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.1 Reactivity Limits 3-1 3.2 Safety Rod System 3-3 3.3 Reactor Safety System 3-4 3.4 Reactor Ventilation System 3-9 3.5 Reactor Coolant System 3-11 3.6 Criticality Limits 3-12 3.7 Limitations of Experiments 3-15 4.0 SURVEILLANCE REQUIREMENTS 4-1 4.1 Reactivity Limits 4-1 4.2 Safety Rod System 4-3 4.3 L4a.t . isr> System 4-4 4.4 Reactor Ventilation System 4-8 4.5 Reactor Coolant System 4-9 5.0 DESIGN FEATURES 5-1 5.1 Site 5-1 5.2 Reactor 5-1 5.3 Fuel 5-2 6.0 ADMINISTRATIVE CONTROLS ,
6-1 i 6.1 Organization and Staffing 6-1 6.2 Independent Review and Audit 6-6 6.3 Procedures 6-7
! 6.4 Records. 6-9 6.5 Required Actions 6-10 6.6 Reports 6-11 1
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NEDO-12725 TABLES l Table Title Pg
- 3-1 Scram Systems 3-5 3-2 Safety Related Systems 3-6 3-3 Stack Release Rate Limits 3-9 4-1 Surveillance Requirements of Scram Systems 4-6 4-2 Surveillance Requirements of Safety-Related Systems 4-7 I
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4 ILLUSTRATIONS Figure Title Pm 6-1 Facility Organization 6-2 l
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NEDO-12725 1.0 DEFINITIONS 1.1 REACTOR SAFETY 1.l.1 Safety Limits (SL)
Limits upon important process variables which are found to be necessary to reasonably protect the reactor fuel from melting 1.1.2 Process Variables As generally used, these are the measurable operational variables of reactor power, inlet temperature, and primary flow. For the.NTR, the only important process variable is the reactor power.
1.1.3 Limiting Safety System Settings (LSSS)
Settings for automatic protective devices related to those variables having significant reactor safety functions.
1.1.4 Limiting conditions of Operation (LCO)
The lowest functional capability of performance levels of equipment required for safe operation of the facility.
l 1.1.5 Measuring Channel j A measuring channel is the combination of sensor, lines, amplifiers, and l output devices which are connected for the purpose of measuring the value f of a process variable.
1.1.6 Measured Value l
l The measured value of a parameter is the value as it appears at the output of a measuring channel.
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NEDO-12725 1.1.7 True Value The true value for a parameter is its exact value at any instant.
1.1.8 Indicated Value The true value of a variable offset by the particular instrument uncertainty; e g.,
ca.libration, circuit bias, and least count.
1.1.9 Inlet Temperature The core inlet temperature.
1.1.10 Reactor Power The total reactor thermal power, as determined by a primary coolant system heat balance.
1.1.11 Scram Systems That combination of measuring channels and associated circuitry which forms the protective system of the reactor as listed in Table 3-1.
1.1.12 Nuclear Safety Channels The linear and Log N nuclear channels identified in Table 3-1 that are used to monitor reactor power.
l 1.1.13 Safety Related Systems That combination of measurink channels and associated circuitry which provides information which requires utnual protective action to be initiated as listed in Table 3-2.
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NEDO-12725 1.1.14 Reactor Safety System The reactor safety system consists of the scram systems listed in Table 3-1 and the safety related systems listed in Table 3-2.
1.2 REACTOR OPERATIONS 1.2.1 Reactor Operation The reactor is considered to be operating (started up) when all of the follow-ing conditions are satisfied:
- a. Console key is in the console key switch and the console is energized.
- b. More than one of the installed safety and control rods are withdrawn.
- c. Porcntial excess reactivity is greater than, or equal to, zerv.
1.2.2 Reactor Shutdown That suberitical condition of the reactor where the negative reactivity of the Xenon-free core would be equal to or greater than the shutdown margin with experiments removed or when there is insufficient fuel to go critical with all manual poiso- sheets, safety rods, and control rods removed.
1.2.3 Reactor Secured That overall condition where all of the following conditions are satisfied:
- a. Reactor is shut down.
- b. Console is de-energized and the console key switch is in proper custody.
- c. No work is in progress involving in-core components, installed rod drives, or experiments.
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1.2.4 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 operation excluding shutdowns which occur during planned equipment testing or check-out operations.
1.2.5 Operable A system or component is operable when it is capable of performing its intended function in a normal manner.
1.2.6 Abnormal Occurrence An abnormal occurrence is any of the following:
- a. Any actual reactor safety system setting less conservative than specified in 2.2, Limiting Safety System Settings.
b* Operation in violation of a Limitihg Condition for Operation in Section 3.
- c. Incidents or conditions which prevented, or could have prevented, the reactor safety system from performing its intended safety function.
- d. An uncontrolled or unplanned release of radioactivity in excess of the limits specified in the Nuclear Regulatory Cammission requirements:
Title 10, Code of Federal Regulations, Part 2D (10 CFR 20).
- e. An uncontrolled or unanticipated change in reactivity greater than 0.50$ while the reactor is critical. l 1
- f. An observed inadequacy in the implementation of either administra-tive or precedural controls, so that the inadequacy has caused the l
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i NEDO-12725 existence or development of an unsafe condition in connection with the i
j operation of the reactor.
1.2.7 Potential Excess Reactivity That excess reactivity which can be added by the remote control pf poison rods plus the maximum credible reactivity addition from primary coolant tempera--
tute change plus the potential reactivity worth of all installed experiments.
1.2.8 Shutdown Margin The minimum 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, and that the reactor will remain subcritical without further operator action.
1.3 REACTOR EXPERIMENTS 1.3.1 Experiment An experiment is any of the following:
- a. An activity utilizing the reactors experimental facilities or its components or the neutrons or radiation generated therein.
- b. An evaluation or test of a reactor system operational, surveillance, or maintenance technique.
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- c. The material content of any of the preceding, including structural components, encapsulation or confining boundaries, and contained fluids or solids.
1.3.2 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.
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1.3.3 Secured Experiment Any experiment, experimental facility, or component of an experiment is deemed to be secured, or in a secdred position, if it is held in a 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 1.3.4 Unsecured Experiment Any experiment, experimental facility, or component of an experiment is deemed to be unsecured if it is not secured as defined in 1.3.3 above. Moving parts of experiments are deemed to be unsecured when they are in motion.
1.3.5 Movable Experiment A movable experiment is one which may be inserted, removed, or manipulated in an experimental facility while the reactor is critical.
1.3.6 Static Reactivity Worth The static reactivity worth of an experiment is the absolute value of the re-activity change which may be measurable by calibrated control or regulating rod
- comparison methods between two defined terminal positions or configurations of the experiment.
1.3.7 Potential Reactivity Worth The potential reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or antici-pated changes or credible malfunctions that alter experiment position or configuration.
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NEDO-12725 The evaluation must consider possible trajectories of the experiment in motion relative to the reactor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or filling of-void spaces or motion of mechanical components. For removable experiments, the potential reactivity worth is equal to or greater than the static reactivity worth.
1.3.8 Explosive Material Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N. I. Sax, Third Ed. (1968), or is given an Identifi-cation of Reactivity (Stability) Index of 2, 3, or 4 by the National Fire Protection Association in its publication 704.M, 1966, " Identification System for Fire Hazards of Materials," also enumerated in the " Handbook for Laboratory Safety", 2nd edition (1971) published by The Chemical Rubber Co.
1.3.9 Accumulation Accumulation is any single item or group of fissile materials. One can of UO 2
powder, one can of UO pellets, or one fuel rod are examples of an accumulation 2
of fissile material. Examples of fissile materials are U-235, U-233, and PU-239.
1.3.10 Assembly Accumulations of fissile material, which are separated from each other by less than 12 inches.
l 1.3.11 Array A collection of two or more accumulations and/or assemblics wherein each accumulation and/or asserisly in the array is separated from the other by more than 12 inches, but is not nuclear isolated from each other.
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NEDO-12725 1.3.12 Nuclear Isolation Two accumulations (assemblies) or arrays or accumulations (assemblies) may be considered as being nuclearly isolated from each other only if an edge-to-edge separation exists which is not less than one of the following or its nuclear equivalent: 8 inches of water; the larger of 12 feet or the greatest distance across an orthographic projection of either accumulation (assembly) or array on a plane perpendicular to a line jeining their centers.
1.4 MAINTENANCE AND TESTING 1.4.1 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.
1.4.2 Channel Test The introduction or interruption of a signal into the channel to verify that it is operable (not to include the alarm or trip test).
1.4.3 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 possible, includ-ing equipment actuation, alarm, or trip test and shall include the Channel Test.
1.4.4 Instrument Check A qualitative verification of acceptable performance by observation of the instrument's behavior. This verification where possible shall include com-parison of the instrument with other independent instruments measuring the same variable.
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NEDO-12725 1.4.5 Inctrument Test The introduction of interruption of a signal (i.e., built-in internal signals) into an instrument to verify that it is operable (not to include verification of the alarm or trip test).
1.4.6 Instrument Calibration
' A comparison and/or an adjustment of an instrument so that its output corresponds with acceptable accuracy to known values of the parameter which the instrument measures. Calibration shall include instrument actuation, alarm, or trip test, and shall include the Instrument Test.
1.4.7 Component Check A quantitative verification of acceptable performance by observation of the components behavior. This verification, where possible, shall include comparison of the component with other independent components or systems measuring the same variables.
1.4.A Component Test The introducti)n or interruption of a signal into the component to verify it l is operable (not to include the alarm or trip test).
1.4.9 Component Calibration l
A comparison and/or an adjustment of the component so that its output corre-sponds with acceptable accuracy to known values of the parameter which the i
l component measures. Calibration shall encompass the entire component if l
possible, including equipment actuation, alaru, or trip test, and shall include I the component test.
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NEDO-12725 1.4.10 Alarm or Trip Test The introduction or interruption of a signal into a channel, instrument, component, etc., as required to verify that the alarm or trip circuitry is functioning as intended.
1.4.11 Daily Prior to the first startup of each day of operation following a period when the reactor was secured.
1.4.12 Monthly Approximately at 1 month intervals (usually 'the first week of each month).
1.4.13 Quarterly Approximately at 3 month intervals (usually the first month of each fiscal quarter).
1.4.14 Yearly Approximately at 1 year intervsis.
1.5 GENERAL 1.5.1 Site The area (41600 acres) within the confines of the Vallecitos Nuclear Center (VNC) owned and operated by General Electric.
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l j That portion of the building occupied by the reactor, reactor control room, and associated support areas.
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NEDO-12725 1.5.3 Licensed Operator A person who is licensed as a reactor operator (RO) or senior reactor operator (SRO) pursuant to 10CFR55 to operate the controls of the Nuclear Test Reactor (NTR)..
1.5.4 Operator Trainee An individual who manipulates the controls of the. reactor as part of a training program to qualify for an operator's license. This individual will be under
! the direction, and in the presence of, a licensed operator while manipulating the controls.
1.5.5 Readily Available on Call (Senior Reactor Operator) ,
A senior reactor operator who satisfies all of the following:
- a. Is within a reasonable driving time (1/2 hour) from the reactor facility;
- b. Can be promptly contacted by telephone; and
- c. Has informed the reactor operator on duty where he may be contacted.
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l- 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l
r 2.1 SAFETY LIMITS (SL) 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 prevent fuel element damage from
, occurring as a resu'.t of normal operation or anticipated operational occurrences.
2.1.3 Speciff;ation The true value of the reactor thermal power shall not exceed 190 kW during
- steady-state or quasi steady-state operating conditions.
2.1*. 4 Basis l The criterion for the safety limit is that departure from the nucleate boiling (DNB) will not occur anywhere on the fuel disk surfaces. This criterion would permit operation at a departure from nucleate boiling ratio (DNBR) approaching unity (DNBR <1).
To allow a generous safety margin in the calculation of the DNB surface heat flux the safety limit is specified to correspond to a mini-mum allowcble value of DNBR = 1.5. The DNBR is defined by the equation:
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I DNBR i Operating surface heat flux 1
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[ The safety-limit analysis used a thermal-hydraulic computer model to examine l the reactor for steady-state and quasi steady-state conditions at power levels high enough to cause steam generation. The computer model includ'es a multi-( channel core model and a circulation loop consisting of the core, a heat 2-1
NEDO-12725 exchanger, and a primary coolant pump. Situations with the secondary coolant flow to the heat exchanger on or off and the primary coolant pump on or off were analyzed using the program. When the primary pump is off, the core is cooled by natural circulation.
The analysis shows 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 need be considered. The steady-state safety limit for reactor power ensures that the DNBR is never less than 1.5.
The most severe anticipated steady-state or off-normal quasi steady-state event is one in which the reactor power.is at the safety limit value (150 kW) and under natural circulation with the primary flow and inlet temperature simultane-ously at their least favorable values (reactor pressure assumed rormal). For tFis highly unlikely set of conditions the DNBR = 1.8. The safety limit must ce assumed to be 150 kW because this value was used as the trip point for postulated accidents.
Analyses of certain postulated accidents indicate the reactor power may exceed the specified safety limit without causing damage to the reactor fuel. The amount by which the safety limit may be exceeded is a time-dependent variable for each case. Application of the limiting safety system settings (LSSS) specified in Technical Specification 2.2 for reactor power ensures that damage to the fuel will not occur for any postulated accident.
The specified safety limit for reactor power is adequate for all operating conditions including normal operation, anticipated abnormal occurrences, and postulated accidents.
! 2.2 LIMITING SAFETY SYSTEM SETTING (LSSS) l 2.2.1 Applicability l
This specification applies to the scram set point for the nuclear safety channels which monitor reactor power level.
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2.2.2 Objective The objective of this specification is to prevent fuel damage resulting from
, excessive reactor thermal power during normal operation, anticipated operational
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occurrences, and postulated accidents.
2.2.3 Specification The LSSS for reactor thermal power level is 138 kW.
2.2.4 Bases The LSSS has been chosen ,to ensure that reactor scram is initiated in time to prevent fuel damage resulting from excessive reactor thermal power during normal operation, anticipated operational occurrences, and postulated accidents. - The safety margin (the difference between the safety limit and the LSSS) includes systematic and random types of instrument uncertainties and, for. transient events, also includes system delay times. The value of 138 kW is derived from the trip point of 150 kW used in the analysis of the postulated accidents and is used in place of the 190 kW steady-state safety limit because it is more restrictive.
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l l NED0-12725-3.0 f.IMITING CONDITIONS FOR OPERATION l
i 3.1 REACTIVITY LIMITS 311.1 Applicability This specification applies to the reactivity condition of the reactor and to the reactfeity worths of control rods, safety rods, and experiments.
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 Core reactivity shall be limited so that:
- a. The reactor configuration shall be controlled to ensure that the potential excess reactivity shall <0.76$. If it is determined that the potential excess reactivity is, or will be >0.76$, the reactor shall be shutdown immediately. Corrective action shall be taken as required to ensure the potential excess reactivity is <0.76$.
i b. The reactor shall be suberitical whenever the four safety rods are l
withdrawn from the core and the three control rods are fully inserted.
3.1.3.2 No more than one safety rod shall be allowed to be moved in an outward
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a 3.1.3.3 The rate of withdrawal of each safety rod during operation of the l
l reactor shall be less than 1-1/4 inches per second. ,
3.1.3.4 The rate of withdrawal of each control rod during operation of the reactor shall be less than 1/6 inch per second.
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NEDO-12725 3.1.3.5 Each manual poison sheet used to satisfy the requirements of Specification 3.1.3.la shall be restrained in its respective graphite reflector slot in a manner which will prevent movement by more than 1/2 inch relative to the reactor core.
3.1.3.6 The reactivity worth of experiments shall be limited so that:
- a. The sum of the potential worths of all experiments performed at any one time shall be limited to comply with 3.1.3.la.
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- b. The negative potential reactivity worth of any components of an experiment which could be ejected from the reactor by a chemical reaction shall be limited to comply with 3.1.3.la.
3.1.3.7 The temperature coefficient of reactivity of the reactor primary coolant shall be negative above 124*F.
3.1.4 Bases Operation in compliance with Specification 3.1.3.la ensures that there would not be any mechanism for addition of reactivity greater than 0.76$. Detailed analyses have been made of reactivity insertions in NEDO-12727, the NTR Safety Analyses Report (SAR). The analyses show that the reactor will take a reactivity step addition of 0.76$ without fuel damage.
Operation in accordance with Specification 3.1.3.lb ensures that criticality will not be achieved during safety rod withdrawal. Adherence to the 0.76$
i limit also ensures that the reactor will not go critical during safety rod withdrawal.
Operation in accordance with specification 3.1.3.2 and specification 3.1.3.3 limits the rate of reactivity addition during safety rod withdrawal to that from one safety rod. This value is easily controlled by the operator.
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Operation in accordance with specification 3.1.3.4 limits the rate of reactivity addition during control rod withdrawal._ Experience has shown that this is a value which is easily manually controlled by the operator. The rate is also less than the value analyzed in the rod withdrawal accident in the SAR.
, Operation in accordance with specification 3.1.3.5 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 3.1.3.6a and specification 3.1.3.6b ensures that there would not be any mechanism for addition of reactivity greater than 0.76$, including experiments. See the bases for Specificar'v 3.1.3.la.
Operation in accordance with specification 3.1.3.7 provides negative-reactivity feedback to prevent the consequences of an accident exceeding that analyzed in the SAR.
3.2 , SAFETY ROD SYSTEM 3.2.1 Applicability This specification applies to the reactor safety rod system.
l 3.2.2 obj ective The objective of this specification is to specify the lowest acceptable level of performance to reasonably ensure proper operation of the reactor safety rod l system.
3.2.3 Specifications 3.2.3.1 The reactor shall not be made critical unless all safety' rods are operable. The reactor shall be shutdown immediately if it is known that a safety rod is not operable.
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NEDO-12725 3.2.3.2 The total time for the safety rod insertion from the withdrawn position to the shock absorber shall be < 300 ms after a nuclear safety high-power scram signal is initiated (to include the electronic delay time and the insertion of the safety rod).
3.2.4 Bases 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 ensures that-the time for the safety rods to move to the edge of the core including the electronic delay time for a nuclear safety high-power scram shall not exceed 200 ms as used in Section 11 of the SAR.
3.3 REACTOR SAFETY SYSTEM 3.3.1 Applicability This specification applies to the Reactor Safety System.
3.3.2 Obj ec tive The objective of this specification is to specify the lowest acceptable level of performance, instrument alarm, or trip points, and the minimum number of operable components for the Reactor Safety Syscem.
3.3.3 Specifications l 3.3.3.1 The reactor shall not be made critical unless the reactor safety system is operable in accordance with Tables 3-1 snd 3-2, including the maximum or minimum alarm or trip points.
3.3.3.2 The reactor shall be shutdown immediately if any portion of the I reactor safety system malfunctions except as provided for in Tables 3-1 and 3-2.
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Table 3-1 SCRAM SYSTEMS Item No. System ' Condition Trip Point Function 1 Linear High reactor power No higher than Scram (2 out of 3 138 kW or 1 out of 2)
Loss of positive high No less than 90% of Scram (2 out of 3 voltage to ion chambers operating voltage or 1 out of 2)
(if ion chambers are used) 2 Log N High reactor power No higher than 138 kW Scram z
Short reactor period No less than +5 see Scram @
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Y Amplifier Mode switch N/A Scram kl not in operate position U u
Loss of positive high No less than 90% of Scram voltage to ion chambers operating voltage.
(if ion chambers nre used) 3 Primary Coolant High core outlet No greater than 222*F Scram Temperature temperature 4 Primary Coolant Low flow No less than 15 gpm when Scram Flow reactor power is > 0.1 kW s
5 Manual Console betton depressed N/A Scram 6 Electrical Power Reactor console key in N/A Scram off position Loss of ac power to N/A Scram console
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Table 3-2 SAFETY RELATED SYSTEMS Item Function No. System Condition Set Point At a value to ensure Visible and audible alarm; audible 1 Reactor Cell Low differential alarm may be bypassed after Pressure pressure compliance with Specification 3.4.3.1. recognition.
At a value to ensure Visib]r and audible alarm; audible 2 Fuel Loading Tank Low level alarm may be bypassed after Water Level compliance with Specificaticq 3.5.3.1. recognition.
Primary Coolant High core outlet At a value to warn of a Visible and audible alarm; audible 3
high core coolant alarm may be bypassed after Temperature temperature outlet temperature recognition 4 Primary Coolant Core delta N/A Provide information for the heat Core Temperature t emperat ure balance determination.
Differential Radiation Moni- High level At a value approved in Visible and audible alarm; audible S
accordance with alarm may be bypassed after recog-tors (Reactor nition. May be temporarily out of Cell, South cell, Specification 6.3.1 service if portable instruments are and North Room) used during personnel entry and occupancy.
At a value to ensure Visible and audible alarm; audible 6 Stack High level alarm may be reset after recognition.
Radioactivity compliance with Specifications 3.4.3.3 and 3.4.3.5.
Low power indication At a value to ensure Safety or control rods cannot be 7 Linear Power that the indication is withdrawn (1 out of 3 or 1 out of 2) on scale.
N/A Safety rod magnets cannot be reener-8 Control Rod Rods not in gized; may be bypassed to allow with-drawal of one control rod, or one safety rod, or one safety rod drive for purposes of inspection, mainte-nance, and testing.
N/A Control Rods cannot be withdrawn; 9 Safety Rod Rods not in safety rods must be withdrawn in sequence; may be bypassed to allow withdrawal of one control rod, or one safety rod, or one safety rod drive for purposes of inspec-tion, maintenance, and testing.
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I 3.3.4 Bases Operation in accordance with Specifications 3.3.3.1 and 3.3.3.2 ensures that the reactor is equipped with sufficient scram channels and safety-related channels to control operation of the facility, measure operating parameters, warn of abnormal conditions, and scram the reactor automatically if required.
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 to provide protective action to prevent exceeding the safety limit during normal operation and anticipated operational occurrences. Scram action as a result of loss of positive high voltage to ion chambers for the linear channels provides assurance that the ion chambers are capable of detecting neutrons.
The Log N high reactor power scram will be set no higher than the LSSS to pro-vide redundant protective action to prevent exceeding the safety limit during normal operation and anticipated operational occurrences. The period scram limits the rate of rise of the reactor power to periods which are manually con-trollable. The Log N amplifier mode switch scram ensures that the Log N ampli-fier 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 ion chamber is capable of detecting neutrons.
The primary coolant high core outlet temperature scram provides assurance, when the reactor is at power levels which require forced cooling, that the reactor will be shut down if the primary coolant outlet temperature is high.
l 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.
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l The loss of electrical power from either the reactor console key in the off l position, or a 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 gives adequate assurance.
that operation of the reactor will be in compliance with specification 3.4.3.1.
The fuel loading tank low water level alarm gives adequate assurance that operation of the reactor will be in compliance with specification 3.5.3.1.
The primary coolant high core outlet temperature alarm gives adequate assurance that warning vill be given to the operator of high primary coolant core outlet temperature.
The use of an area radiation monitor system will ensure that the area (s) of the facility in which a potential high radiation area exists are monitored to ensure
' protection of personnel.
The stack radioactivity high level alarm gives adequate assurance that operation of the reactor will be in compliance with specification 3.4.3.2.
The low power indication .a the linear channel ensures that the operator has a nuclear safety scram channel operating and indicating neutron flux levels during rod withdrawal.
The control rods not in condition ensures that the reactor will be started up by withdrawing the four safety rods prior to withdrawing the control rods.
l The safety rods not in out condition ensures that the normal method of reactivity control with no reduction in safety margin is used during reactor operation.
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NEDO-12725 3.4 REACTOR VENTILATION SYSTEM 3.4.1 Applicability This specification applies to the reactor ventilation system.-
3.4.2 Obj ective The objective of this specification is to ensure the release of airborne radio-active materials via the ventilation system is below authorized limits.
3.4.3 Specifications 3.4.3.1 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 co < 0.1 kW immediately and corrective action shall be taken as required.
3.4.3.2 The limits for radioactive material discharged through the reactor ventilation system shall be as specified in Table 3-3.
Table 3-3 STACK RELEASE RATE LIMITS Isotope Group Annual Average (pC1/sec) Short Term (uci/sec) l Halogen, > 8d T (2.0 x 10 ) x (MPC,) (1.4 x 10') x (MPC,)
l/2 l Particulate, > 8d T g
Beta-Gamma (2.0 x 10 ) x (MPC,) (1.4 x 10') x (MPC,)
j Alpha (2.0 x 10 ) x (MPC,) (1.4 x 10') x (MPC,)
All Other (including (1.4 x 10 0) x (MPC*) (1.0 x 10 ) x (MPC* )
Noble Gas) where MPC *
= the concentration in pCi/ml shown in Table II, Appendix B,10 CFR, Part 20.
3-9 i
i
NEDO-12725 3.4.3.3 Alarm points for particulate and noble gas continuous monitors shall not exceed a value corresponding to the annual average release rate limit for the most restrictive isotope in the category (except as specified in 3.4.3.4):
Monitor Release Rate Beta-Gamma particulate < Annual Average of Sr-90 limit Noble Gas < Annual Average of Kr-87 limit 3.4.3.4 The alarm points may be raised to a value correspending to the short term release rate, not to exceed the short term release limit, provided grab samples are taken and analyzed for isotopic content. The short term release rate shall be based on the rate of release of the most restrictive isotope.
The releases shall be monitored to ensure that the annual release limit is not exceeded.
3.4.3.5 During operation of the reactor above 0.1 kW or the performance of activities that could release redioactivity, the stack particulate activity monitor and the gaseous activity monitor shall be operating.
NOTE If either unit is not operable, the reactor shall be shut down, or the activity involving releases shall be terminated, or the unit shall be promptly repaired or replaced with one of cocparable monitoring capaoility. During this period, j any indication of abnormal reactor operation shall be cause to shut down the reactor immediately.
3.4.4 Bases operation in accordance with Specification 3.4.3.1 ensures that potentially contaminated reactor cell air due to reactor operation, is released through the ventilation system.
3-10
l l NEDO-12725 l
! The ventilation system release limits in Specification 3.4.3.2 are based on the following.
The annual average dilution factor from the NTR stack to the site boundary based on 1976 and 1977 meteorological conditions and a stack flow rate of 3000 cu ft/ min equals approximately 20,000. That is, the concentration at the site boundary of any release from the NTR stack will be < 1/20,000 of the concentration at the stack when averaged over 1 year.
The above listed annual average limit contains a reduction factor of 2 to account for discharges from other VNC stacks. Additionally, the release limits for all but the "all other" catetory contain a reduction factor of 700 to account for reconcentration in the environs.
The alarm points in Specification 3.4.3.3 are set for the release rate of the most restrictive isotope in the category which is equivalent to the annual average release limit.
The alarm points in Specification 3.4.3.4 will be set for the release rate of the most restrictive isotope in the category which is equivalent to the short term release limit. In no case will the annual release limit be exceeded.
3.5 REACTOR COOLANT SYSTEM l 3.5.1 Applicability
(
This specification applies to the reactor primary coolant system.
, 3.5.2 Objective l
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.
3-11
NEDO-12725 3.5.3 Specifications j l
3.5.3.1 The reactor shall not be started up unless the core tank is filled with water. If during operation of the reactor it is determined or suspected that the core tank is not filled with water, the reactor shall be shut down immediately and corrective action will be taaen as required.
3.5.4 Bases Operation in accordance with Specification 3.5.3.1 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.
3.6 CRITICALITY LIMITS 3.6.1 Applicability This section applies to all Special Nuclear Material (SNM) at the NTR Facility except for the SNM located in the reactor core.
3.6.2 Objective The objective of this specification is to ensure that all Special Nuclear Material as defined in 10CFR70 which is outside the NTR reactor core is handled, used, and stored, so that inadvertent criticality will not occur.
3.6.3 Specifications SNM, excluding the material located within the NTR core, shall be handled, used, and stored in accordance with the following requirements:
3.6.3.1 For individual accumulations and/or assemblies of fissile material, the k-effective shall be < 0.9 under conditions of optimum water moderation with full water reflection.
3-12
NEDO-12725 3.6.3.2 The k-effective shall be 1 0.9 for optimumly moderated (water) individual accumulations of fissile material in a fully reflected (water) array.
3.6.3.3 For arrays of fissile material, the solid angle criterion omega 1 9-10 k-effective shall'be satisfied where k-effective is for the unreflected accumulation of fissile material in the array under conditions of optimum water moderation, if the following conditions are met:
- a. The unreflected k-effective of any unit shall not exceed 0.8.
- b. Each accumulation shall have a k-effective 1 0.9 when completely reflected by water.
- c. Each accumulation shall be evaluated as optimum 1y moderated.(e.g.,
aqueous solutions) at the point of maximum k-effective,
- d. The minimum separation between each accumulation shall be at least 12 inches.
- e. The allowed solid angle shall not exceed 6 steradians.
- f. The effectiveness of the reflector around the array shall be no greater than that of a thick (12 inches) water reflector spaced at distances from the accumulations comparable to the spacing between the accumulations.
- g. For full reflection of an array by concrete thicker than 4.8 inches, the allowable solid angle shall be reduced by 40 percent.
3.6.3.4 The k-effective value for an individual accumulation may be determined from hand calculations using nuclear properties given in the following:
" Criticality Handbook", by Carter, R.D. g al, ARH-600 Volumes 1-3, Atlantic Richfield Hanford, 9/25/70 or " Critical and Safe Masses and Dimensions of Lattices of U and U02 Rods in Water" by Clark, H. K., DP-1014, Savannah River Laboratory, February, 1966.
3-13
NED0-12725 3.6.3.5 The k-effective v..iue of an individual accumulation or assembly of fissile material or an array of fissile material may be determined by l
using the KENO-IV Criticality Code in " KENO-IV, An Improved Monte Carlo Criticality Program" by Petrie, L. M. , and Cross, N. F. , ORNL-4938, November, 1975, provided the upper limit k-effective of at least two sigma is used in these criteria.
3.6.3.6 Different computational tools may be used to determine the k-effective values or to demonstrate the criticality safety of arrays provided they demon-strate conservatism as good or better than those shown in Figure 7-6 and Tables 7-2 and 7-3 of the Safety Analysis Report (NED0-12727).
3.6.3.7 In evaluating the suberiticality of individual accumulations or arrays of fissile material, the 2-contingency criterion shall be applied (two independent unrelated accidents must occur before criticality is credible).
3.6.3.3 If special nuclear materials are to be stored, handled, or used in the presence of heavy water, beryllium, graphite, or diphenyl, possible moder-ating and reflecting effects of these materials shall be considered.
3.6.4 Bases Operation in accordance with Specification 3.6.3.1 ensures the safe handling and storage of fissile material by limiting k-effective to a conservative value.
For example, for pure U-233, U-235, or Pu-239, the mass of the fissile material would have to be increased by about 1.7 along with the achievement of optimum water moderation and full water reflection for a sphere to achieve criticality compared with the mass required to achieve a k-effective of 0.9. For UO 2 enriched to 5 w/o U-235 in uranium, the mass of fissile material would have to be increased by more than 1.9 times that required for a k-effective of 0.9, along with the achievement of optimum water moderation and full water reflection for a sphere to achieve criticality.
Operation in accordance with Specification 3.6.3.2 and 3.6.3.3 ensures safe handling and storage materials in arrays. For example, results presented in 3-14 4
l NEDO-12725 the SAR show that observance of the Solid Angle Criterion to be conservative relative to the critical arrays presented.
Operation in accordance with Specifications 3.6.3.4, 3.6.3.5, and 3.6.3.6 ensures that the computational tools used to evaluate the criticality safety of arrays and accumulations of fissile materials give conservative results when compared to critical experiment data.
Operation in accordance with specification 3.6.3.7 ensures that failure of only one criticality contrcl will not result in a criticality accident. For example, the application of a safe batch limit (0.45 x the minimum critical mass) is an application of this criterion when double batching is credible due to permissi-ble operating procedures. For fissile materials permitted at the NTR, a safe batch for an accumulation of fissile material would have a k-effective of less than 0.9.
Operation in accordance with Specification 3.6.3.8 ensures that proper evalua-tion of criticality safety at the NTR is made if super-normal moderators are present. For example, if super-normal moderators such as heavy water, graphite, bery}lium, or diphenyls are present in or around fissile material external to the NTR core, minimum critical rasses and minimum fissile material masses required to achieve an accumulation whose k-effective is 0.9 may be smaller than those required for water moderated and reflected fissile materials.
3.7 LIMITATIONS OF EXPERIMENTS 3.7.1 Applicability
[
This specification applies to reactor expt.iments.
l 3.7.2 Obj ective ,
The objective of this specification is to prevent an experiment from resulting in a hazard to the operating personnel or the general public and ' damage to the reactor.
3-15
NEDO-12725 3.7.3 Specifications 3.7.3.] The maximum amount of explosives permitted in the NTR facilities are:
- a. South Cell, W 5,(D/2) with W < 9 lbs and D > 3 ft.
- b. North Room (without Modular Stone Monument), W < D with W < 16 lbs and D > 1 ft.
- c. North Room (with Modular Stone Monument), W 5, 2.
- d. Setup Room, W 5,25 lbs.
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 l wall.
3.7.3.2 Experimental objects shall not be allowed inside the core tank when the reactor is at a power greater than 0.1 kW.
3.7.3.3 Experimental objects located in the fuel loading chute shall be secured during reactor shutdown to prevent their entry into the core region during reactor operation.
3.7.3.4 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 i
be at least 5 ft from any explosive device. Radioactive materials, other
~
than those produced by the neutron radiography of the explosive devices and imaging systems, are not permitted in the Setup Room if explosive devices are present.
i 3-16 1
NEDO-12725 3.7.3.5 Unshielded high frequency generating equipment shall not be I operated within 50 ft of any explosive devices.
t l
l 3.7.3.6 The cumulative radiation exposure for any explosive device shall not 12 I
- exceed 3 x 10 2
neutrons /cm from thermal neutrons and 1 x 104 roentgens from gamma.
( 3.7.3.7 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 its contents and/or leakage of corrosive or flammable materials will not damage the reactor.
3.7.3.8 Experimental fuel elements containing plutonium to be utilized in the exr'rimental facilities shall be clad and other experimental devices conta .g plutonium shall be encapsulated.
i 3.7.3.9 The maximum possible chemical energy release from the combustion of flammable substances contained in any experimental 2cility shall not exceed 1000 kW/sec. The total possible energy release from chemical combination or decomposition of substances contained in any experimental capsule shall l be limited to 5 kW/sec, rate of the reaction in the capsule could l
l exceed 1 W. Experimental iacilities containing flammable materials shall be vented external to the reactor graphite pack.
3.7.3.10 A written description and analysis of the possible hazards involved for each type of experiment shall be evaluated and approved by the facility manager, or his designated alternate, before the experiment may be conducted.
Records of such evaluation and approval shall be maintained.
l 3.7.3.11 No irradiation shall be performed which could credibly interfere with the scram action of the safety rods at any time during reactor operation.
3.7.3.12 The radioactive material content, including fission products, of any singly encapsulated experiment to be utilized in the experimental 3-17 t
l _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _
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% cf the equivalent annual doses stated in 10CFR, 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.
3.7.3.13 The radioactive m2terial content, including fission products, of any doubly encapsulated or vented experiment to be utilized in the exp2rimental facilities shall be limited so that the complete release of all gaseous, par-ticulate, or volatile components from the encapsulation or confining boundary of the experiment could not result in a dose to any peraon 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 timo 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 la excess of 5 rem to the whole body or 30 rem to the thyroid.
f Bases Specifications 3.7.3.1 through 3.7.3.11 are intended to reduce the likelihood of damage to 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 3.7.3.12 and 3.7.3.13 ensure the radiological effects of experiment failures do not pose a hazard to the general public or to operating personnel.
3-18
- 4. SURVEILLANCE REQUIREMENTS 4.1 REACTIVITY LIMITS 4.1.1 Applicability This specification applies to the surveillance requirements for reactivity limits.
l 4.1.2 Objective To ensure that the reactivity limits of Specification 3.1 are not exceeded.
4.1.3 Specification l
4.1.3.1
- a. 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$.
i b. Neutron multiplication will be observed throughout each startup.
Safety rod withdrawal shall be stopped if it appears critical will l be reached before all safety rods are out.
- c. The following restrictions shall be verified at least yearly except during periods when the reactor is in a nonfueled condition. Prior to refueling these restrictions shall be performed and the ncrmal surveillance schedule shall be resumed.
, 1. No more than one safety rod can be moved out at a time.
l l
! 2. The rate of withdrawal of each safety rod shall be less than 1-1/4 in./sec.
l 4-1
-l l
r
- 3. The rate of withdrawal of each control rod shall be less than 1/6 in./sec.
4.1.3.2 Each manual poison sheet in the core region of the reactor used to satisfy specification 3.1.3.la shall be verified to be properly restrained upon insertion per specification 3.1.3.5.
4.1.3.3
- a. The potential reactivity worth of experiments shall be assessed before irradiation. After establishing critical rod position, the potential excess reactivity will be determined which verifies the acceptability of the experiment worth,
- b. Types of experiments with the capability for increasing reactivity by a chemical reaction will be designed and reviewed in accordance with Specification 6.2.
4.1.3.4 The temperature coefficient of reactivity of the reactor primary coolant shall be verified to be negative above 124*F whenever new fuel is used in the reactor core.
4.1.4 Bases Operation in accordance with Specification 4.1.3.la will ensure that the reactor is not operated with a potential excess reactivity of > .76$.
Operation in accordance with Specification 4.1.3.lb will ensure that the reactor will be suberitical when the safety rods are in the full-out position.
Operation in accordance with Specification 4.1.3.lc will ensure that the safety rod and control rod withdrawal ratas are not exceeded and that the safety rod withdrawal interlock is functioning.
4-2
NED0-12725 Verification that the manual poison sheets are properly restrained as specified in Specification 4.1.3.2 ensures that they cannot be ejected during a seismic disturbance.
The reactivity worth of each installed experiment as required in Specification 4.1.3.3a is evaluated before the experiment irradiation to ensure the limits of these specifications are not exceeded. If it cannot reasonably be determined that the reactivity worth is less than the limits in these specifications, then the reactivity worth is measured at low power subsequent to insertion but before full-power operation.
Compliance with Specification 4.1.3.3b ensures that negative reactivity added during reactor operation by a chemical reaction from an experiment in an experi-mental facility will be limited such that there will be no damage to the reactor components and no radioactive releases in excess of that in specifications 3.7.3.13 and 3.7.3.14.
Compliance with specification 4.1.3.4 ensures that the temperature coefficient is negative above 124*F. It is not affected by reactor configuration and fuel burnu'p and is therefore not expected to vary significantly with core life (but could be affected by fuel design changes).
4.2 SAFETY ROD SYSTEM i
i 4.2.1 Applicability l
This specification applies to the surveillance requirements for the safety rod system, l
4.2.2 Obj ective t
l The objective of this specification is to specify the minimum surveillance require-ments to reasonably ensure proper performance of the safety rod system.
l l
l t
3
1 4.2.3 Specifications 4.2.3.1 Safety rod insertion times shall be acasured whenever maintenance that could affect the insertion times is performed on the safety rods, whenever there is reason to suspect the insertion times have changed, and at least yearly except during periods when the reactor is in a nonfueled condition. Prior to refuel-ing th reactor, the insertion times shall be measured and the normal safety rod surveillance schedule shall be resumed.
4.2.4 Bases
] 4.2.4.1 The safety rod insertion time is a direct measurement of the tot:1 time I (4.2.3.1) for the safety rod insertion from the withdrawn position to the shock absorber after a nuclear safety power scram signal has been introduced (includes i instrument delay time). This measurement is to confirm that the rods will insert in less than 300 ms. This measurement is performed as part of the preventive main-tenance program, whenever any repair or maintenance is performed on the safety rods thet could conceivably affect the insertion times or 'for any other reasor.
that is determined to conceivably affect the insertion times. The measurement is to confirm that the rods will insert within the time limits as analyzed in the SAR.
4.3 REACTOR SAFETY SYSTEM L
4.3.1 Applicability This specification applies to the surveillance of the reactor safety system.
4.3.2 Objective The objective of this specification is to ensure that the reactor safety system is operable as required by Specification 3.3.
l 4-4 l i
1
. - __-- - - =- - - _ - - .
NEDO-1d735 4.3.3 Specifications 4.3.3.1 Checks, tests, calibrations, etc., shall be performed as specified in Tables 4-1 and 4-2.
~
4.3.3.2 A channel check of the Nuclear Safety' Channels utilized during reactor operation comparing the channel outputs with a heat balance, shall be performed shortly efter reaching equilibrium primary water core differential temperature on each reactor run the power level is maintained above 15 kW.
4.3.3.3 The Nuclear Safety Channels utilized during reactor operation shall be observed during startup. If it is determined that a channel does not have suf-ficient operating range and trip capability as required, the reactor shall be
] shutdown immediately. Corrective action shall be taken as appropriate.
i 4.3.4 Bases i
l Operation in accordance with Specification 4.3.3.1 ensures the operability of the Scram systems and the safety-related systems by requiring daily prestartup and post-service tests and' checks or others, as required, based on-operating experi-l ence and equipment reliability.
l The linear system is verified to be responding to neutrons. If the Log N system is not sensitive to the source check, the Log N detector operability will be tested by varying the high voltage and observing the detector capacitance effect l
on the period meter.
The calibration schedule presented in Tables 4-1 and 4-2 is considered to be ade-quate. It will detect any long-term drift that is not detected by normal. inter-comparison of channels or instruments.
i p
1 4-5
__ _ - . - __ .._._ - - _ _ __ ____ _ - _. , _ . . _ _ _ _ . -_~
NEDO-12725 Table 4-1 SURVEILLANCE REQUIREMENTS OF SCRAM SYSTEMS Item No. Item Surveillance Frequency"
- 1. Linear System Channel Check Daily Channel Test Daily Instrument Alarm or Trip Test Daily Instrument Calibration Quarterly Channel Calibration (compari- Quarterly sion against a heat balance)
- 2. Log N System Instrument Test Daily Instrument Alarm or Trip Test Daily b
Instrument Calibration Quarterly Channel Calibration (compari- Quarterly sion against a heat balance)
- 3. Primary Coolant Alarm or Trip Test of the Relay Daily emperatur Yearly Detector Calibration
- 4. Primary Coolant Flow Channel Check Daily nastrument Alarm or Trip Test Daily Instrument Calibration Yearly
- 5. Manual Alarm or Trip Test Daily
- 6. Electrical Power Alarm or Trip Test Daily i
" Excluding periods when the reactor is in a nonfueled condition. The normal sur-veillance schedule as appropriate shall be resumed prior to restdrt.
b Prior to placing into service an instrument which has been repaired, the instru-ment check, test, calibration etc., as appropriate shall be performed.
4-6
Table 4-2 SURVEILLANCE REQUIREMENTS OF SAFETY RELATED SYSTEMS Item No. Item Surveillance Frequency"
- 1. Reactor Cell Pressure Alarm or Trip Test Monthly Instrument Calibration Quarterly
- 2. Fcel Loading Tank Alarm or Trip Test Monthly Water Level
- 3. Primary Coolant Alarm or Trip Test Monthly Temperature Instrument Calibration Yearly
- 4. Primary Coolant Core Channel Check Quarterly D e al nstmment Calibration Yearly
- 5. Radiation Monitors Instrument Check Daily Alarm or Trip Test Monthly Channel Calibration Quarterly
- 6. Stack Radioactivity Instrument Cher.k Daily Alarm or Trip Test Monthly Instrument Calibration Quarterly
- 7. Linear Power Channel Test Monthly
- 8. Control Rod Channel Test Quarterly l
- 9. Safety Rod Channel Test Quarterly I
l
( " Excluding periods when the reactor is in a nonfueled condition. The normal sur-veillance schedule as appropriate shall be resumed prim to restart. .
b Prior to placing into service an instrument which has been repaired, the instru-ment check, test, calibration, etc., as appropriate, shall be performed.
4 .7
NEDO-12725 The channel check of the nuclear safety channels in Specif! cation 4.3.3.2 ensures that the detectors are properly adjusted to accurately monitor the parameter they are measuring. Equilibrium primary water core differential temperature is normally recched after approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of stable reactor power operation.
Operation in accordance with Specification 4.3.3.3 ensures that the nuclear safety scram channels provide adequate neutron flux-level manitoring and trip protection.
4.4 REACTOR VENTIIATION SYSTEM 4.4.1 Applicability This specification applies to the surveillance requirements for the reactor venti-lation system.
4.4.2 Obj ec tive The objective of this specification is to ensure that the reactor ventilation system is in satisfactory condition to provide adequate confinement and to con-trol the release of radioactivity to the environment.
4.3.3 Specification 4.4.3.1 The reactor cell negative pressure, with respect to the control room as stated in Specification 3.4.3.1, shall be verified prior to the first reactor startup of each day, r..
4.4.3.2 Surveillance requirements of the instrumentation and equipment required
- to comply with Specification 3.4.3.2 shall be as listed in Table 4-2.
4-8
NEDO-12725' i
f 4.4.3.3 Procedures delineating administrative controls as required to comply with Specification 3.4.3.4 shall be in accordance with Specification 6.3.
4.4.4 Bases Operation in accordance with Specification 4.4.3.1 ensures that contaminated reactot celi air is exhausted through the stack ventilation system. This minimizes the possibility of airborne contamination release to surrounding areas.
Operation in accordance with Specification 4.4.3.2 ensures that all required channels are operational.
Operation in accordance with Specification 4.4.3.3 ensures that appropriate sur-veillance of particulate and gaseous activity releases will.be maintained.
4.5 REACTOR COOLANT SYSTEM 4.5.1 Applicability This* specification applies to the surveillance of the reactor primary coolant system.
L 4.5.2 Objective l
l To ensure that the limits of Specification 3.5 are not exceeded.
l l
4.5.3 Specification l Surveillance requirements of the instrumentation and equipment required to comply t
l with Specification 3.5.3 shall be as listed in Table 4-2.
9 4.5.4 Bases l Having a low-level alarm on the fuel loading tank ensures that the core is filled with water when required by these specifications.
l l
l 4-9/4-10
NED0-12725 l
l
- 5. DESIGN TEATURES 5.1 SITE 5.1.1 The Nuclear Test Reactor (NTR) facility shall be located on the site of the Vallecitos Nuclear Center (VNC) which is owned and controlled by the Cer.eral Electric Company.
5.1.2 The minimum distance from the reactor to the exclusion area boundary as defined in 10CFR100 of the Commission's regulations shall be approximpcely 1600 feet. The restricted area, as defined in 10CFR20 of the Commission's regula-tions, shall be the Vallecitos Nuclear Center.
5.2 REACTOR 5.2.1 A core reel assembly shall support the fuel assemblies a fixed distance from the core center 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.
5.2.2 The control system shall consist of four scrammable, spring-actuated safety rods, three nonscrammable control rods, and a number of manual poison sheets. When tha poison rods and sheets are inserted, they shall be located in the graphite reflector at the outer periphery cf the core tank. The safety and control rods shall be boron carbide clad in stainless steel. The manual poison sheets shall contain metallic cadmium.
5.2.3 Each safety rod shall be connected to a rod drive through an electro-magnet that is energized by a current from a magnet power supply passing in series through the contacts of the manual scram switch, logic unit controlled power switch, and scram relays.
5.2.4 Each safety rod drive assembly shall be equipped with a drive-out limit switch, a drive-in limit switch, a safety-rod, in position switch and a magnet 5-1 l
l
l separation switch. The drive-out- limit switch, in conjunction with the magnet separation switch, gives indication of safety rod out. These switches actuate lights on the control console and interrupt motor power to the rod drives as required.
5.2.5 Each control rod shall be equipped with rod out and rod in limit switches which shall actuate lights on the console.
5.2.6. Position indication for the control rods shall have a tolerance of 1.0 inch.
5.2.7 The graphite reflector shall contain slots to accommodate no more than six manual poison sheets. Each installed manual poison sheet shall be locked in place before the safety rods may be withdrawn.
5.2.8 The core tank shall be protected from overpressure by a vent line to the atmosphere of the cell.
5.3 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 follcwing:
- b. Enrichment High enriched (approximately 93%)
- c. Cladding Aluminum, 0.022 in, thickness
- d. Fuel disk active diameter 2.685 in,
- e. Fuel disk spacing on shaft 0.35 to 0.45 in., cent'er-to-center 5-2
- 6. ADMINISTRATIVE CONTROLS l
l 6.1 ORCAhIZATION AND STAFFING 6.1.1 The NTR shall be owned and operated by the General Electric Company with management and operations organization as shown in Figure 6-1 or equivalent.
6.1.2 The section-level manager shall be responsible for the NTR facility license.
6.1.3 The subsection-level manager (Operations) is designated the facility manager and shall be responsible for the overall safe operation and maintenance of the facility.
6.1.4 The unit-level manager is responsible for the routine safe operation and maintenance of the facility including the training and retraining of operating personnel.
6.1.5 The Reactor Supervisor as required will be the individual responsible for the daily operations.
I 6.1.6 Responsibilities of one level may be assumed by alternates of a lower level when designated in writing.
6.1.7 Functions performed by one level may be performed by a higher level, provided the minimum qualifications are met (i.e., Reactor Operator's license, etc.)
( 6.1.8 The minimum staffing when the reactor is not secured shall be composed of:
1
( a. A licensed operator in the control room with access to the nuclear controls.
( b. A second person present at the site familiar with NTR Emergency Procedures.
- c. A licensed senior operator shall be present at the NTR Facility or readily available on call at all times the reactor is not secured.
l l
6-1 l
NEDO-12725 GENERAL ELECTRIC COMPANY UPPER MANAGEMENT LEVEL SECTION LEVEL (MANAGEGl r- - - -- - - - q --_
'" ' " SUBSECTION LEVEL p -- - --VALL ECITOS l NuCU'/R!A E ND l l g (F ACILf. Y M AN AGER) TECHNOLOGICAL g
l QUALITY ASSURANCE SAFETY COUNCIL l t_ _ _ _ _ _. J L. _. _. _ _ _ _ J r- - - - - -- -
l INTERN AL REVIEW 7 r q
ENGINEERING ,g AND AND e l AUDIT FUNCTION l , MAINTENANCE l L _ _ _ _ _ _ .J L______J UNIT LEVEL 7 .
NUCLEAR SAFETY !
M N GE.
FUNCTION l
__.._._._.__.__J NUCLEAR SAFETY r - - --
l REACTOR r- 7 i
FUNCTION l AD SUPERVISOR MAINTENANCE !
l l L. _ _ _ _ _ .- L _.,_ - J NTR OPERATORS NOTE: DASHED LINES REPRESENT ALTERNATE / OPTIONAL ,
l ORGANIZATION FUNCTIONS
- Figure 6.1. Facility Organization l
l 6-2 l
NED0-12725
- d. A licensed senior operator shall be present at the FIR Facility during the following events:
- 1. During the first startup each day (except when startup is within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of a reactor shutdown).
- 2. During the recovery from an unplanned or unscheduled shutdown or significant reduction in power.
- 3. During reactor fuel loading or reactor fuel relocation.
- 4. During safety and control rod removal, or safety and control rod or rod drive maintenance.
6.1.9 The minimum facility staff qualifications shall be as follows:
- a. Facility Manager At the time of appointment to the active position, the Facility Manager shall have at least 6 years of nuclear experience. Addition-ally, he shall have a baccalaureate or higher degree in an engineering or scientific field and have previous managerial experience. Equiva-lent education or experience may he substituted for a degree. The degree may fulfill 4 years of the 6 years of nuclear experience required on a one-for-one basis.
- b. Unit Manager At the time of appointment to the active position, the Unit Manager shall have at least 4 years of nuclear related experience. Additionally, he shall have a baccalaureate or higher degree in an engineering or scientific field and shall have, or immediately pursue, the obtaining of an NTR Senior Reactor Operator license. Equivalent education and experi-ence may be substituted for a degree. A maximum of 3 years equivalent full-time academic training may be substituted for 3 years of the 4 years of nuclear-related experience required.
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- c. Internal Review and Audit Function l At the time of appointment to the active position, the individual assigned the Internal Review and Audit Function shall have at least 3 years of nuclear related experience. Additionally, he shall have a baccalaureate or higher degree in an engineerink or scientific field.
Equivalent education and experience may be substituted for eLdegree.
A maximum of 2 years equivalent full-time academic training may be substituted for 2 years of the 3 years of nuclear related' experience required.
- d. Reactor Supervisor At the time of appointment to the active position, the Reactor Super-visor shall have at least 3 years of nuclear related experience. He shall have an NTR Senior Reactor Operator license or shall have received sufficient training at the facility, or elsewhere, to satisfy the requirements of a Senior Reactor Operators license and shall immedi-ately pursue the obtaining of an NTR Senior Reactor Operator license.
A maximum of 2 years equivalent full-time academic training may be substituted for 2 years of the 3 years of nuclear-related experience required.
- e. NTR Operators
- 1. Senior Reactor Operator A Senior Reactor Operator, as a minimum, shall have the following qualifications as determined by the Unit Manager:
(a) A suf ficient level of experience in NTR reactor operat ions, experiment setup and operation, and a high level of leadership.
(b) An NTR Senior Operator's license. -
6-4
l NED0-12725 l (c) Mature judgement and a capability for handling diverse 1
problems under rapidly changing conditions.
(d) Reactor Operators' qualific.ations.
- 2. Reactor Operator A Reactor Operator, as a minimum, shall meet the following quali-fications as determined by the Unit Manager:
(a) A high school diploma or equivalent, with a high degree of mechanical dexterity.
(b) A NTR Operators license.
(c) Sufficient training or experience in related nuclear fields.
- 3. Trainee A trainee, as a minimum, shall meet the following qualifications as determined by the Unit Manager:
(a) A high school diploma or equivalent, with a high degree of mechanical dexterity.
(b) Sufficient applicable training or experience.
Senior Reactor Operator and Reactor Operator candidates are required to obtain 1
! licenses issued by the U.S. Nuclear Regulatory Commission in accordance with the provisions of 10CFR55. Licensed personnel participate in a comprehensive requalification program. In addition, medical examinations are required at least every 2 years to determine the individual's capability to perform assigned
' duties without undue risk of operating errors or impairment of the ability to cope with emergencies. .
6.1.10 A designated licensed senior operator shall'be responsible for Operator and Senior Operator license candidate training and license requalification.
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l 6.2 INDEPENDENT REVIEW Ah'D AUDIT l 6.2.1 Independent review and cudit is performed by a technically qualified function, responsible through optional intervening management, to the section manager.
6.2.2 The function as required shall perform under a written charter contain-ing the following information as a minimum:
- a. Subjects reviewed and/or audited
- b. Responsibilities
- c. Authorities
- d. Records, including provisions for dissemination, review, and approval of audit findings in a timely manner
- 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, including safety evaluations, to be conducted without prior NRC approval, pursuant to 10CFR50.59, to verify the proposed activities, do not constitute a change in the Technical Specifications or an unreviewed safety question.
- b. Proposed changes to the procedures or the facility, as described in the Safety Analysis Report, including safety evaluations, to be com-pleted without prior NRC approval, pursuant to 10CFR50.59 to verify the activity, does not constitute a change in the Technical Specifica-tions or an unreviewed safety question.
- c. All new procedures and revisions thereto having safety significance
~ required by the specifications in Section 6.3.
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- d. Proposed changes to the Technical Specifications or the facility operating license.
- e. Violations of the Federal Regulations, Technical Specifications, and facility license requirements having safety significance.
- f. Unusual or abnormal occurrences which are reportable to the NRC under provisions of the Federal Regulations or the Specifications in Section 6.6.
- g. Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect, or could affect, nuclear safety.
6.2.4 The independent audit function shall include selective examinations of facility records and operations. Written reports shall be provided to the nuclear safety function management and the facility manager. Activities requiring periodic independcat audit shall include the following:
- a. The conformance of facility operation to the Federal Regulations, Technical Specifications, and facility license requirements.
- b. The results of all actions taken to correct deficiencies or increase effectiveness in facility equipment, structures, systems, or methods of operation that affect nuclear safety.
- c. The facility emergency proceduras, security plan, requalification program, and their implementing procedures.
6.3 PROCEDURES 6.3.1 Written procedures shall be prepared for the following activities as required:
- a. Normal startup, operation, and shutdown of the reactor
- b. Systems and components as specified by the facility manager or designated alternate involving nuclear safety of the facility 6-7
NZDO-12725 l
'c. Defueling, refueling, and fuel transfer operations when required l
- d. Preventive or corrective maintenance which could have an effect on the safety of the reactor .
- e. Of f-normal conditions relative to reactor safety for which an alarm is received
- f. Response to abnormal reactivity changes C. Surveillance, testing, and calibrations required by the Technical Specifications
- h. Emergency conditions involving potential or actual release of radioactive materials
- i. Radiation protection consistent with 10CFR20 requirements -
- j. Review and approval of changes to all required procedures
- k. Security plan, the operator requalification program, and emergency procedure =
6.3.2 Written procedures shall be prepared which describe the review, approval, and implementing requirements for all types of experiements utilizing the reactor experiment facilities. These procedures shall require the following as a minimum:
- a. All new types of experiments which could be postulated to affect reactivity, to result in unusual radiation exposure to personnel or an unusual release of radioactive materials, shall be reviewed for compliance with the facility license and the Technical Specifications.
Each type of experiment shall be approved by the facility manager or his designated alternate before the experiment is performed.
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- b. Changes to approved experiments shall receive appropriate review and approval.
- c. Approved experiments are implemented in accordance with written procedures.
6.3.3 The facility manager, or his designated alternate, shall approve all procedures (including changes thereto) required by Specifications 6.3.1 and 6.3.2 before implementation.
6.4 RECORDS i
6.4.1 In addition to those records required by applicable government regula-tions, the following records shall be maintained for the periods specified below.
The records may be in the form of logs, data sheets, or cther suitable forms.
6.4.2 Records to be retained for a period of at least 5 years or for the life of the component, whichever is smaller, are as follows:
- a. Normal reactor operation records,
- b. Records of unplanned or unscheduled shutdowns and scrams, including the reasons therefor.
- c. Principal maintenance activities including substitution or replace-ment of reactor equipment or components,
- d. Occurrences reported to the NRC as required by the Technical Specifications.
- e. Surveillance," testing, and calibrations required by the Technical Specifications.
- f. Records of experiments performed, including unusual events involved in their handling and performance.
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- g. Reviews performed for changes made to . procedures or equipment, and new tests and experiments not submitted for NRC approval pursuant to 10CFR50.59.
- h. Records of reviews and audit reports of the independent. review and audit function.
- 1. Records of off-site radioactive shipments and receipts.
6.4.3 Records to be retained for 2 years from the time of the events are as follows:
Retraining and requalification of licensed operators.. Records of the most recent complete cycle shall be maintained at all times the individual is employed at the Facility.
5.4.4 Records to be retained for the lifetime of the reactor facility are as
- ollows:
- a. Gaseous radioactive effluents released or discharged to the environ-ment beyond the effective control of the licensee, as monitored at or before the point of release or discharge.
- b. Off-site environmental monitoring surveys.
- c. Radiation exposure for all facility personnel monitored.
- d. Updated drawings of the facility and records of facility changes.
6.5 REQUIRED ACTIONS 6.5.1 in the event of an abnormal occurrence, action shall be taken to ensure the safety of the plant and personnel and to take appropriate corrective measures.
If required, the reactor shall be shut down. If the reactor is shut down because 6-10
NED0-12725 of an abnormal occurrence, the reactor operation shall not be resumed unless authorized by unit level or higher management. The NRC shall be notified as required in accordance with Section 6.6.
6.5.2 In the event a safety limit is known to have been exceeded, the reactor shall be shut down and reactor operation shall not be resumed without authori-zation by the NRC. The NRC shall be notified in accordance with Section 6.6.
6.6 P.EPORTS 6.6.1 Notification shall be made for the itema listed below not later than the following working day by telephone and confirmed by telegraph or similar convey-ance to the Director of the appropriate Regional U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement listed in Appendix D of 10CFR20. A written report shall follow within 14 days to the appropricte regional USNRC Office of Inspection and Enforcement with copies to the Director, Of fice of Inspection and Enforcement, USNRC, Washington, D.C. 20555, and to the Director, Of fice of Management Planning and Analysis, USNRC, Washington, D.C.
20555.
- a. Violation of safety limits (Subsection 6.5.2)
- b. Release of radioactivity from the site above allowable limits.
- c. Operation of the reactor with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in the Technical Specifications.
- d. Operation in violation of Limiting Conditions for Operation estab-lished in the Technical Specifications unless prompt remedial action, which is permi,tted by the Technical Specifications, is taken.
- e. A reactor safety system condition or component malfunction which renders or could render the reactor safety system incapable of per-forming its intended 7afety function unless the malfunction or 6-11
NEDO-12725 condition is discovered during maintenance tests or periods of reactor shutdown and is caused by the maintenance activity or shutdown.
NOTE Where components or systems are provided in addition to those required by the Technical Specifications, the failure of the extra components or systems is not con-sidered reportable, provided that the minimum number of components or systems specified or required perform their intended reactor safety functions.
To the extent possible, the prelimincry telephone notification shall:
- a. Describe, analyze, and evaluate safety implications;
- b. Outline the measures taken, or to be taken, to ensure the cause of the conditions is determinti; and
- c. Indicate the corrective action taken, or to be taken, to prevent repetition of the occurrence.
6.6.2 A written report shall be forwarded within 30 days to the Director of the Regional USNRC Office of Inspection and Enforcement listed in Appendix D of 10CFR20 in the event of:
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- a. Discovery of any substantial errors in the safety analyses or in the methods used for such analyses, as described in the Safety l
l Analysis Report or in the bases for the Technical Specifications; l b. Discovery of any substantial variance disclosed by operation of the reactor from performance specifications contained in the Safety Analysis Report; and
- c. A permanent change in the facility management personnel (unit manager and higher).
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NEDO-12725 DISTRIBUTION Name M/C-W. A..Erwin V13 D. L. Gilliland V13 D.~ R. Smith V14 P. W. Swartz V14 E. J. Strain -V18 W. H. King .V18
. W. R. Lloyd V18 l B. M. Murray 'V18 G. E. Cunningham V18 D. L. Brown V10 W. B. Johnson (2) V09 4 S. R. Thompson V09 j D. J. Morena V09 l C. E. Leighty (7) V09 P. F. Kachel V13 G. L. Stinsnell VIO VNC Library. V01 P. S. Webb V06 4
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NUCLEAR ENERGY DlVillONS e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GEN ER AL $ ELECTRIC TECHNICAL INFORMATION EXCHANGE TITLE PAGE
[ AUTHOR SUBJECT TigNgg C. E. Leighty Nuclear Science and Technology DA N il 1981 TITLE GE CLASS Technical Specifications for the I General Electric Nuclear Test GOVERNMENT CLAM Reactor NUMBER OF PAGES REPRODUCIBLE COPY FILED AT TECHNICAL SUPPORT SERVICES.R&UO. SAN JOSE, CALIFCRNIA 95125 (Mail Code 211) 67
SUMMARY
These Technical Specifications apply to the General Electric Nuclear Test Reactor sfacility license R-33, Docket No. 50-73) and represent those parameters which define the boundaries of licensed activities.
By cutting out this rectangle and folding in half, the above information can be fitted into a standard card file.
NEDO-12725 DO'UMENT NUMBER Irradiation Processing Operation INFORM ATION PREPARE D FOR Reactor Irradiations (Nuclear Test Reactor)
BUILDING AND ROOM NUMBER M All CODE NEDC1416D7) l