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O UNION CARBIDE RESEARCH REACTOR TECHNICAL SPECIFICATIONS License No. R-81 Docket No. 50-54 Dated: May 1, 1980 8006290363 | |||
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, i TABLE OF CONTENi$ | |||
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1.0 DEFINITIONS 4 | |||
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4 | |||
2.1 - Safety Limits of Reactor Operation 7 | |||
2.2 - Limiting Safety System Settings 9 | |||
3.0 LIMITING CONDITIONS FOR OPERATION 9 | |||
3.1 - Reactivity Limitations 3.2 - Control and Safety Systems 10 13 3.3 - Radiation Monitoring Systems 14 3.4 - Engineered Safety Features 3.5 - Limitations on Experiments 16 20 3.6 - Fuel 3.7 - Pool Water Quality 22 23 3.8 - Radioactive Releases (Airborne) 3.9 - Radiological Environmental Monitoring 27 28 3.10 - Land Use Census 3.11 - Bases for Environmental Specifications 28 33 4.0 SURVEILLANCE REQUIREMENTS 33 4.1 - General 4.2 - Safety Channel Calibration 33 4.3 - Reactivity Surveillance 33 4.4 - Control and Safety System Surveillance 34 4.5 - Radiation Monitoring System 34 | |||
, | |||
35 4.6 - Engineered Safety Features 35 4.7 - Reactor Fuel ' | |||
36 4.8 - Sealed Sources 36 4.9 - Pool Water | |||
^4.10 - Core Spray 36 37 4.11 - Flux Distribution | |||
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37 5.0 DESIGN FEATURES 37 5.1 - Reactor Fuel 5.2 - Control and Safety Systems 38 5.3 - Rod Control System 40 41 5.4 - Cooling System 5.5 - Containment System 42 5.6 - Fuel Storage 44 45 6.0 ADMINISTRATIVE CONTROLS 45 6.1 - Organization 51 6.2 - Procedures 6.3 - Experiment Review and App. oval 52 6.4 - Required Actions 53 6.5 - Reports 54 57 6.6 - Records 59 | |||
==7.0 REFERENCES== | |||
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1.0 DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Condition of Operation (LC0) are defined in | |||
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50.36 of 10 CFR Part 50. | |||
1.1 Safety Channel - A Safety Channel is a measuring or protective channel in the reactor safety system. | |||
J 1.2 Reactor Safety System - The Reactor Safety System is a combina-tion of safety channels and associated circuitry which forms the automatic protectise system for the reactor, or provides informa-tion which requires the initiation of manual protective action. | |||
1.3 Operable - Operable means a component or system is capable of performing its intended function in its required manner. | |||
1.4 Operating - Operating means a component or system is performing its intended function in its normal manner. | |||
4 1.5 Channel Check - A Channel Check is a qualitative verification of acceptable performance by observation of channel behavior. | |||
. | |||
1.6 Channel Test -A Channel Test is the introduction of a cali-bration or test signal into the channel to verify that it responds in the specific manner. | |||
1.7 Channel Calibration - A Channel Calibration is an adjustment of the channel components such that its output responds, within specified range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including readouts, alarm, or trip. | |||
1.8 Unscheduled Shutdown - An Unscheduled Shutdown is any unplanned | |||
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shutdown of the reactor, after startup nas been initiated. | |||
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1.9 Reactor Shutdown - The reactor is shut down when the negative reactivity of the cold, clean core including the reactivity worths of all experiments is equal to or gre'"- than the shut-down margin. | |||
, 1.10 Reactor Operating - The reactor is considered to be operating whenever it is not shut down. | |||
1.11 Reactor Secured - The reactor is secured when: | |||
: a. The core contains insuf cient fuel to attain criticality under optimum conditions i moderation and reflection, or | |||
: b. The moderator has been removed, or | |||
: c. (1) Minimum number of control rods fully inserted as required by Technical Specifications, and (2) The console key switch is in the off position and the key is removed from the lock, and 4 | |||
(3) No work is in progress involving core fuel, core structure, installed control rods or control rod drives unless they are physically decoupled from the control rods and (4) No in-core experiments are being moved or serviced with a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, whichever is smaller. | |||
1.12 T.ue Value - The True Value of a parameter is its actual value at any instant. | |||
1.13 Measured Value - The Measured Value of a parameter is as it appears on the output of a measuring channel. | |||
1.14 Measuring Channel -A Measuring Channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a parameter. | |||
1.15 Reportable Occurrence - A Reportable Occurrence is any of those conditions described in Section 6.5.3 of this specification. | |||
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An Experiment is an apparatus, device or 1.16 An Experiment - | |||
material, placed in the reactor core, in an experiment f acility, or in line with a beam of radiation emanating from the reactor, | |||
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excluding devices designed to measure reactor characteristics such as detectors and foils, | |||
: a. Secured Experiment - Any experiment, experiment facility, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor core. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, or other forces which are normal to the operating environment of the experi-ment (or by forces which can arise as a result of credible malfunctions). | |||
: b. Movable Experiment - A movable experiment is one which may be removed, or manipulated while the reactor is inserted, critical. | |||
: c. Untried Experiment - is a single experiment or class of experiments that has not been previously evaluated and ap-proved by the Nuclear Safeguards Committee. | |||
1.17 Experiment Facilities - An Experiment Facility is any structure, orient, device or pipe system which is intended to guide, manipulate, control the environment or otherwise position, facilitate a multiplicity of experiments of similar character. | |||
1.18 Control Rod - A control rod is a rod fabricated from neutron absorbing material which is used to compensate for fuel burnup, temperature, and poison effects. A control rod is magnetically coupled to its drive unit allowing it to perform the safety function when the magnet is de-energized. | |||
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1.19 Readily Available on Call - Readily available on -call means an individual who (1) has been specifically designated and the designation known to the operator on duty, (2) keeps the operator on duty informed of where he may be rapidly contacted (e.g. by i phone, etc.) (3) is capable of getting to the reactor f acility within a reasonable time under normal conditions (e.g., I hr. or within a 30 mile radius). | |||
d 1.20 Scram Time - is the elapsed time between the instant a limiting safety system set-point is reached and the instant that the slowest control rod is fully inserted. | |||
1.21 Safety Limits - are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against release of radio- | |||
. activity. The principal physical barrier is the fuel cladding. | |||
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits of Reactor Operation 2.1.1 Limits in Forced Cooling Mode | |||
: a. Applicability - This specification applies to the variables that affect thermal and hydraulic performance of the core during forced cooling. They are: | |||
(1) Power in MW. | |||
(2) Flow in GPM. | |||
(3) Height of water above the core. | |||
, | |||
: b. Objective - To assure fuel cladding integrity. | |||
: c. Specifications . | |||
(1) The maximum steady power level under various flow con-ditions shall be as shown in Figure 1. | |||
! (2) The pool water level shall not be less than 20 feet above the core. | |||
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! d. Bases - The analysis given in Ref. 1, Sec. Al, forms the basis for this specification. The superposition method of Gambill is used to derive the burnout heat flux as a function of primary flow rate. A safety f actor of 1.25 is applied to i allow for uncertainties in the correlation. Pool temperature | |||
, | |||
(or core inlet temperature) is not included in the specifi-cation as this variable changes very slowly and has only a minor effect, e.g., a 10 F change results in only a 5% | |||
variation in burnout flux. The latter, however, is evaluated conservatively near the high end of the pool temperature range that is exoected in practice. A de-rating f actor can 0 | |||
: be applied for pool temperatures in excess of 120 F. The relationship between total power and peak heat flux is derived for the core situation with the greatest peaking factors, viz. a new fuel element adjacent to a central in-core flux trap. Reactor power, primary flow rate, and water level will be maintained well within safety limit specifications through limiting safety system scram settings. | |||
(see2.2.1). | |||
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't 2000 2100 Ic00 1900 Fie. 1 | |||
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t PRIMARY FL0h'-RATE, 0 (GPm .- . . | |||
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2.1.2 Limits in Free Convection Mode | |||
: a. Applicability - This specification applies to the thermal and hydraulic variables affecting the core during natural convection cooling. Tney are: | |||
, | |||
(!) Power in MW. | |||
(2) Height of water above the core, | |||
: b. Objective - To assure fuel cladding integrity. | |||
: c. Specifications - | |||
(1) The maximum reactor power level shall be 6.7 MW. | |||
q a | |||
(2) The pool water level shall not be less than 20 ft. above i | |||
; | |||
the top of the core. | |||
l d. Bases - The analysis given in Ref. 1, Sect. A2, forms the basis of this specification. The homogeneous method of Gambill aid Bundy, used in this analysis, has been employed succe sfully to predict natural convection burn-out in ORR and HFIR fuel. The former fuel is close in design to UCRR fuel. A safety factor of 1.24 is applied to account for 1 random variations and uncertainties. A pool temperature near U | |||
the high end of the operating range (120 F) is assumed. | |||
The safety system settings on power and pool level (2.2.2) assure adherence to these specifications. | |||
; 2.2 Limiting Safety System Settings 2.2.1 Safety Channel Set-Points in Forced Cooling Mode | |||
: a. Applicability - This specification applies to the setpoints of the safety channels, | |||
: b. Objective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded. | |||
: c. Specification - For operation in the forced cooling mode the limiting safety system settings are: | |||
(1) Power level at any flow rate shall not exceed 7.5 MW. | |||
(2) Power level settings for various conditions of flow and of pool temperature shall be in accordance with Fig. 1. | |||
1 | |||
(3) Coolant flow shall not be less than 1800 gpm for powers above 250 KW. | |||
(4) Pool level shall not be less than 20 f t. above the top of the core. | |||
: d. Bases - Safety limits have been shown previously (Sec. 2.1 and Ref.1) to lie at a low flow-to-power ratio. To provide adequate assurance that these limits are not approached too closely, the LSSS are chosen conservatively so as to minimize the chance of boiling in the core. This results in a much larger flow / power ratio. In Ref. 1, Sect. A3, power levels derived using conservative correlations for incipient boiling are tabulated for various values of pool temperatures and flow rates to illustrate the resulting temperature margins. | |||
Through a comparison with experiments at ORNL (ORR) this method is shown to be conservative. To preserve the desired temperature margins for all comoinations of variables, the LSS settings are a combination of two fixed set-points, viz. | |||
scrams at 7.5 MW and 1800 gpm, plus an adjustable one that provides automatic power reduction at a setting that depends on the pool temperature and flow rate. Rates of change of U | |||
poci temperature are very slow - a few F/hr. at most - and thus allow adequate lead time for adjustment. | |||
For a reactivity transient the case considered is the step insertion of 0.25% a K positive reactivity with the reactor operating at a steady power of 7.5 MW. The analysis given in Ref.1, Sect. B3, s%ws that the power at the end of .75 sec. | |||
(the scram time, Sect. 3.2.1 below) will be no more than 11 MW. This is well below the safety limit for this mode of operation. No automatic scram is associated with pool temperature as this parameter varies very slowly allowing ample time for appropriate operator action. | |||
-g-2.2.2 Safety Channel Set-Points in Natural Convection Mode | |||
: a. Applicability - This specification applied to the set-points of the safety channels, | |||
: b. Objective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded. | |||
: c. Specification - For operation in the natural convection mode, the limiting safety system settings are: | |||
(1) Power Level 6 250 KW. | |||
(2) Pool level 3 20 ft. above the core, | |||
: d. Bases - The set-points are chosen to avoid boiling in the core during routine operation with natural convection cooling. The analysis given in Ref. 1, Sec. A4, shows that a power of 0.35 MW is needed for incipient boiling to occur. | |||
1 To allow for uncertainties a safety f actor of 1.3 is applied to this, resulting in a safety system set-point of 0.25 MW. | |||
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' | |||
The latter is well below the safety limit of 6.7 MW given above (Sect. 2.1.2) . In the case of reactivity transient, a step insertion of 0.25% AK positive' reactivity at an initial power level of 0.25 MW will, following the analysis rf Ref. 1 (Sect. 83), result in a transient power of 0.38 Mei after 1 second. The latter is well below the safety limit of 6.7 MW for the natural convection mode (2.1.2). | |||
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3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limitations 3.1.1 Shutdown Margin The minimum shutdown margin provided by control rods in the cold, xenon-free condition with the highest-worth rod fully withdrawn and with the highest-worth non-secured experiment in its most positive reactive state shall not be less than 0.5% 6K. | |||
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This specification ensures that the reactor can be shut dcwn from any operating condition and remain shut down after cool-down and xenon decay even if the highest-worth control rod is stuck in its fully withdrawn condition. | |||
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3.1.2 Excess Reactivity The core shall not be loaded with an excess reactivity of greater than 10.2% 6 K when located in the stall position and 8.2% aK when the core is located in the open pool position. | |||
1 | |||
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3.1.3 Experiments i Reactivity limits on experiments are specified in 3.5 below. | |||
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3.l.4 Regulating Rod The integral worth of the regulating rod shall not exceed 0.6% a K. This ensures that a malfunction of the control system cannot make the reactor prompt critical. | |||
I 3.2 Control and Safety Systems 3.2.1 Scram Time The scram time shall not exceed .75 second and the control rod magnet release time shall not exceed .05 second. In the transient analysis (Ref. 1, Sect. 83), these values were assumed, s 3.2.2 Measuring Channels The minimum number and type of measuring channels operable and providing information to the control room operator required for i | |||
reactor operation are given as follows: | |||
Channel No. Operable Operating Mode in Which Required Power Level (normal) 2 All-Power Level (intermediate) 1 All Period Channel 1 All | |||
.. | |||
._ | |||
. | |||
Channel No. Operable Operating Mode in Which Required a | |||
Count Rate l Startup Coolant Flow 1 Forced Cooling Core ST 1 Forced Cooling Rod Position 1/ rod All Pool Temperature 1 All Pool Level 1 All i | |||
Note: a. Operable below 50 W. | |||
Bases - The normal power level instruments (" Level Safeties") | |||
provide redundant information on reactor power in the range 25%-150% of the normal operating power level of 5 MW. | |||
The intermediate power instrument (" Log N") provides usable | |||
, | |||
reactor power information in the logarithmic range 10-4%- | |||
, | |||
300% of the normal power of 5 MW. | |||
The count rate channel covers the neutron flux range from the 4 | |||
source level ( :::1 cps) to 10 cps on a logarithmic scale. | |||
It enables the operator to start the reactor safely from a shutdown condition, and to bring the power to a level that | |||
) can be measured by the Log N instrument. | |||
Coolant flow rate and A T instruments allow the operator to calculate reactor power and calibrate the neutron flux chan-nels in terms of power. | |||
Rod position indicators show the operator the relative positions of control rods, and enable rod reactivity cali-brations to be made. | |||
Pool temperature information allows the operator to adjust the cooling system to keep pool temperature within a pre-ferred range, and to adjust the overpower reverse set-point. | |||
(see 3.2.3). | |||
_ _ | |||
. | |||
1 3.2.3 Safety Channels The minimum number and type of channels providing automatic action that are required for reactor operation are as follows: | |||
No. Operating | |||
; Channel Operable Function Mode Power Level (normal) 2 Scram 0 7.5 MW All Power Level (intermediate) 1 Scram @ 3 sec. period All Reverse @ 10 sec. period All Inhibit @ 30 sec. period All c | |||
Reverse All i Count Rate la Inhibit 0 2 cps Startup Inhibit @ 30 sec. period Startup Pool Watc. Level 1 Scram @ 22 ft. All | |||
) Pool Temperature 1 Alarm @ 120 F All b | |||
Coolant Flow l Scram @ 1800 gpm Forced Circ. | |||
Manual Button 1 Scram All Bridge Lock 1 Scram All Guide Tube Lift 1/ rod Scram All Flapper Valve 1 Scram (above 250 KW w/ valve open) All Keyswitch 1 Scram All Notes: a. Operable below 50 W. | |||
: b. Operable above 250 KW. | |||
: c. Overpower reverse set-points shall be set so that the relationship of pool temperature, flow and power levels shown. | |||
, in Figure 1 are never exceeded. | |||
, | |||
Bases - The power level scram provides redundant automatic protective action to prevent exceeding the safety limit on reactor power. | |||
, | |||
, ~ - | |||
. | |||
, | |||
The period scram, assisted by the intermediate level period reverse and rod inhibit, limits the rate of increase in reactor power to values that are controllable without reaching excessive power levels or temperature. These functions are not limiting safety system settings. | |||
The two inhibits on the count rate channel prevent inad-vertent criticality during cold startup that could arise from lack of neutron information or from too rapid reactivity insertion by control rods. | |||
The scram on pool level provides an adequate head of water above the core and guards against loss of coolant and loss of building containment. | |||
The overpower reverse on the intermediate power channel provides automatic action to reduce power and minimize the chance of incipient boiling in the core. | |||
The coolant flow and flapper valve scrams ensure adequate coolant ficw to prevent boiling in the core. | |||
l l The scrams on bridge lock and guide tubes prevent unplanned reactivity ,anges that could occur through core and control element movements respectively. | |||
The keyswitch scram prevents unauthorized operation of the reactor. | |||
Bypass is permitted on those parameters that can be monitored by alternate means if the initiating circuit malfunctions. | |||
3.3 Radiation Monitoring Systems The minimum acceptable monitoring instrumentation required for reactor operation is as follows: | |||
_ __ -__ | |||
- _ . | |||
No, Max. Alarm Type Operable Setpoint Function Excursion Monitor i SR/hr Detect high radiation: | |||
Alarm and isolate at > 5R/hr. | |||
* Detect Particulate, gas and Ex5aust Duct Monitor 1 iodine activities; alarm in | |||
(" Stack Monitor") | |||
Control room. | |||
Building CAM 1 | |||
** Detect particulate activity in reactor building; alarm. | |||
50mr/hr Detect radiation (Y) in key Fixed Area Monitron 3 locations; alarm in Control room. | |||
Evacuation Switch 1 -- Alarm and initiate evacuation sequence. (manual) | |||
Note: For maintenance or repair, required radiation monitors (except for excursion monitor) may be replaced by portable or substitute instruments for periods up to 24 hours provided the function will | |||
. | |||
still be accomplished. Interruption for brief periods to permit checking or calibration is permissible. | |||
3.4 Engineered Safety Features These specifications apply to required equipment for the confine-ment of activity through controlled release of reactor building air to the atmosphere. | |||
* The alarm set-point for the stack gas monitor shall not be set above a value that would result in an exposure greater than 2 mrem / hour assuming a dilution factor of 2000 and the isotope mixture determined by the most recent analysis. The alarm I set-point for the stack I-131 and stack particulate monitor shall not be set above a value corresponding to that listed in Appendix B, Table II, Column I of 10 CFR Part 20 assuming a dilution factor of 2000 and averaging over one week. | |||
** 25% of the maximum permissible concentration at restricted areas according to Appendix B of 10 CFR 20. | |||
! | |||
__ ._ | |||
1 | |||
. | |||
. 3.4.1 Excursion Monitor | |||
: a. Specification: see 3.3. | |||
: b. Basis - | |||
This monitor senses excessive radiation at the reactor bridge and automatically initiates the " evacuation sequence", which consists of a distinctiva alarm, closure of damper valves in the building ventilation system and hold-up tank vent, and starting of the emergency exhaust fan (see 5.5.2). | |||
3.4.2 Emergency Electric Generator | |||
: a. Specification Equipment No. Operable Function Electric Generator i Upon loss of utility power, start automatically and supply emergency power to the exhaust fan and ventilation system controls. A six day supply of fuel shall be maintained. | |||
O | |||
: b. Basis - Upon loss of utility power the reactor scrams auto-matically. Controlled release confinement requires the ability to run the emergency exhaust fan and to close building damper valves. The latter are pneumatically-operated but are electronically-controlled. | |||
3.4.3 Containment | |||
: a. Specification (1) The emergency exhaust fan shall be capable of sustaining a negative pressure within the reactor building of at least .01-in w.g. at an exhaust flow rate of not greater than 200 cfm. | |||
(2) Filters in the emergency exhaust shall be HEPA and charcoal with tested efficiencies of 99.5% for particle removal and 95% for iodine removal respectively. | |||
. | |||
(3) Depth of water in the canal shall be at least 10 f t. | |||
This is equivalent to a water height above the core of 22 ft. | |||
(4) At least one door of the double airlock doors and the truck doors shall be closed while the reactor is operating. | |||
: b. Bases - To effect controlled release under accident con- | |||
, | |||
ditions of gaseous activity present in the building atmos-phere, a negative pressure is required so that any building leakage will be inward. Reference 1 (Sec. C, 2) contains an analysis of a hypothetical accident resulting in release of airborne activity to unrestricted areas. The assumed exhaust rate is 200 cfm and the filter efficiency for elemental iodine is 95%. In the design of the containment building (5.5) the water seal in the canal is effected when the water depth is > 10 ft. | |||
, 3.5 Limitations on Experiments | |||
. | |||
3.5.1 Experiments l a. Applicability - This specification applies to those experi-ments installed in the reactor and its experiment facilities. | |||
: b. Objective - The objective is to prevent damage to the reactor or excessive release of radioactive material in the event of an experiment failure and also to prevent the safety limits from being exceeded. | |||
: c. Specificatir - Experiments installed in the reactor shall meet the following conditions: | |||
(1) The combined worth of all experiments which can add positive reactivity to the core due to a common-mode failure shall not exceed 2% a K. | |||
; (2) The combined worth of all non-secured experiments which can add positive reactivity to the core due to a common-mode failure shall not exceed 1.7% A K. | |||
l | |||
-- , | |||
-- . _ - _. . _ _ . | |||
.. , | |||
' | |||
! | |||
, | |||
. | |||
. | |||
' | |||
, | |||
(3) The reactivity of any single experiment shall not exceed 0.5% A K. | |||
, (4) An experiment worth less than 0.25% 6 K may be moved | |||
' | |||
when the reactor is critical. | |||
(5) An experiment worth more than 0.25% A K but less than 0.5% A K may be moved with the reactor subcritical by at | |||
' | |||
least 0.75% AK. | |||
(6) All material to be irradiated in the reactor shall either be corrosion resistant or encapsulated within corrosion- resistant containers. | |||
(7) Where failure of the pressure-containing walls of an | |||
; experiment container can cause a hazard to personnel or 1 | |||
to the reactor, the container shall be designed and tested in accordance with the applicable pressure vessel codes. | |||
(8) In-core experiments exposed to reactor water shall be designed to prevent surface boiling. | |||
} (9) Experimental apparatus, material, or equipment to be j | |||
' | |||
inserted in the reactor shall not interfere with the j safe operation of the reactor. | |||
l (10) The total primary coolant ficw utilized by all in-core experiments shall be limited to the same as that in six standard fuel elements. | |||
(11) Experiments on the grid-plate extension are limited to a total reactivity of 0.2% AK and a total load of 100 lbs. | |||
(12) Each class of experiment irradiation in the reactor must have been previously reviewed and approved by the Nuclear Safeguards Committee (6.8). | |||
: d. Bases , | |||
(1) See Ref. 2, Sect. G 5.c. | |||
(2) It is shown in Ref. 3 that the reactor can safely self-limit a step reactivity insertion of- $2.14. This corresponds to an insertion of 2.14 x .81 = 1.73% A K. | |||
, | |||
a | |||
- , - | |||
,w- , ,_ | |||
_ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
- | |||
( | |||
(3) The method of Ref. 1, Sect. 83, shows that a step insertion of 0.5% a K with the reactor critical at 5 MW (or 0.25 MW, in natural convection mode) will result at the end of .75 sec. in a power of not more than 14 MW and .4 MW, for natural convection. Each of these power levels does not exceed the corresponding safety limit. | |||
(4) Similarly it is shown that a step increase of 0.25% a K will produce a power level at the end of the scram time that is much less than the safety limit in either mode of operation. In addition, .25% a K is well within the automatic control capability of the reactor control system. | |||
(5) This specification ensures that, even with a 45% error in estimation of the reactivity of an experiment, the reactor will not be made critical. Even if the reactor were critical, the resulting period (~ 3 secs.) will automatically initiate corrective control action. | |||
(6) This requirement guards against release of activation products in the pr,imary coolant or chemical interaction with core components. | |||
(7) This specification ensures that there will be no mechanical damage to the reactor core nor hazards to personnel due to f ailure of experiment containers where pressure exists or builds up during irradiation. In the case of fueled experiments, it further ensures against hazardous and uncontrolled release of fission products into the reactor building or the environment from the same cause. | |||
(8-9) Ensures that no physical or nuclear interference with the safe operation of the reactor will occur. | |||
(10) This condition is assumed in the analysis given in Ref. | |||
- | |||
1, Sect. A. | |||
(11) These limits ensure that movement of these experiments will not result in reactivity changes in excess of that in Sect. c(4) above. | |||
' .. - - | |||
_ _ _ _ _ _ _ _ _ _ _ | |||
._ | |||
. | |||
- | |||
(12) Ensures that all experiments are evaluated by an inde-pendent group knowledgeable in the appropriate fields. | |||
3.5.2 Fueled Experiments | |||
: a. Applicability - These specifications apply to experiments containing nuclear fuel that are installed in the reactor or its experiment facilities. | |||
: b. Oojectives - The objective is to prevent damage to the reactor, prevent excessive release of fission products in the event of an experiment failure, and also to ensure that safety limits are not exceeded. | |||
: c. Specifications - Fuel-bearing experiments in the reactor shall meet the following conditions: | |||
(1) All fueled experiments are to conform to the specifi-cations listed above in Sect. 3.5.1. | |||
(2) The inventory of solid fuel bearing material being irradiated in the reactor core at any time shall be limited to 200 g of source and/or 750 g of special nuclear material. | |||
(3) The inventory of solid fuel-bearing materials in a single irradiation capsule shall be limited to 200 g of source and/or 50 g of special nuclear material. | |||
(4) The fission power of an irradiation capsule containing special nuclear material shall be limited to 13 KW. | |||
(5) The iodine inventory of a single capsule shall be limited to 500 curies 131 I dose oquivalent for a doubly-encapsulated capsule and 70 curies 131 I dose equivalent for a singly-encapsulated capsule. | |||
: d. Bases - These specifications place - limits on the f,ission product inventory in a fueled capsule such that capsule failure and the hypothetical release of all contained fission products to the reactor coolant will not result in excessive exposure to personnel on and off site. | |||
. | |||
The detailed analyses that form the bases of this specifi-cation are given in Ref. 1, Sect. C3. The total amount of special nuclear material permitted in the core at any time has been increased to 750 g. This increase does not affect the consequences of the release from a single capsule as analyzed in Ref. 1 for it has been established (see License Amendmeret No. 10) that failure of a single capsule will not initiate failure in other neighboring capsules. The core limit of 750 g is based on approximately 15 irradiation positions, each holding 50 g of SNM. The limit of 15 positions is dictated by availability of primary cooling capacity. | |||
3.6 Fuel | |||
: a. Applicability - These specifications apply to the number and condition of the fuel elements present in the core. | |||
: b. Objective - To ensure that power is distributed in the t ore among a sufficient number of fuel elements to avoid excessive peak / average ratio, and to avoid excessive release of fission products. | |||
: c. Specifications | |||
, | |||
: l. The minimum number of fuel elements in the core shall be | |||
: 30. Each control element shall count as 1/2 fuel element for this purpose. | |||
: 2. Control rods shall be kept within + 10% of their mean position whenever the reactor power exceeds 500 KW. | |||
l | |||
; | |||
i | |||
- - - - - - y | |||
. _ . - . .. . - _ _ _ _ _ | |||
4 | |||
: 3. Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive identification, be removed from the core. An increase in the normal gaseous fission product release (due to system contamination) by a factor of 100 shall constitute initial evidence of cladding rupture and require identification of the cause. | |||
: 4. Fuel element loading and distribution in the core shall be such that peak / average thermal flux will not exceed 3.3. | |||
: 5. The fuel plates are composed of enriched uranium-aluminum sandwiched between high purity aluminum clad. Fuel plates may be fabricated by alloying the uranium-aluminum or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in the fine powder form. Burnup of the fuel assemblies 2I fission /cm3 . Fuel shall be limited to .94x10 plates may also be fabricated from uranium oxide-aluminum (U3g0 -Al) using the powder metallurgy 21 process and the burnup shall be limited to 1.5x10 3 | |||
fission /cm , | |||
: d. Bases | |||
: 1. A minimum of 30 elements is assumed a the analysis given in Ref. 1, Sects. A-1, A-2. | |||
: 2. This specification minimizes flux tilts that could cause concentrations or shifts in power distribution across the core. Such shifts are only significant in power operation, and thus this limitation is restricted to power levels above 10% of the normal 5 MW. | |||
: 3. Release of fission products from the compact fuel plates used in this reactor (Sect. 5.1), due to a localized cladding defect, is confined to the defect locality. A relatively small defect, thus, cannot release large quantities of fission products. There is a normal small and variable background of fission product release due | |||
; | |||
to uranium contamination in the coolant and on fuel | |||
; plates. It is, thus, safe to specify a recognizable and substantial increase in this background as a possible indication of cladding rupture. If the rupture were | |||
, | |||
extensive, there would be no doubt at all of this condition. | |||
: 4. This peak / average value is used in the Ref. 1 analysis. | |||
3.7 Pool Water Quality | |||
: a. Applicablity - This specification applies to primary cooling system water in contact with fuel elements. . | |||
: b. Objective - To minimize corrosion of the aluminum cladding of fuel plates and activation of dissolved materials. | |||
: c. Specifications | |||
: 1. Pool water temperature will not exceed 130 F. | |||
: 2. Pool water specific resistance is to be not less than 200,000 ohm-cm, except that for periods not greater than 14 days it may be 70,000 ohm-cm. | |||
: 3. The pH of the pool water shall normally be maintained between 5.0 and 7.5. | |||
. . __ | |||
__ | |||
l l | |||
. | |||
I 3.8 Radioactive Releases 3.8.1 Airborne Stack Release Limit Maximum yearly release rates for noble gases, radioiodines and particulates of half-life greater than eight days shall be | |||
) limited by the following expression: | |||
1 Qj (7/Q)/MPCj < 1/6 where: | |||
Qj = The average release rate for any 12-consecutive months of radionuclide, i, in gaseous effluent from tne stack in Ci/sec. | |||
MPCj= Activity concentration of radionuclide, i, as given in Table II, Column 1 of Appendix B to 10 CFR 20, in uCi/cc. | |||
7./Q = Shall be calculated monthly from measured values of iodine concentration sampled at or above the tree line 380 meters NE of the exhaust stack. | |||
ACTION: Shoulci the _ limit of this Section be exceeded, the license shall notify the Commission within 24 hours, and take action to reduce the release to within the limits immediately. | |||
3.8.2 Dose in Unrestricted Areas | |||
: a. Total body dose due to noble gases releases and dose from radiciodines in gaseous effluents fnr the critical | |||
, | |||
individuals in unrestricted areas should be calculated at least once per calendar quarter and reported in the annual report (TS 6.5.2g). | |||
_ _ _ ___. _ _ _ _ _ _ _ _ _ _. | |||
: b. The total body dose to any individual in unrestricted areas due to noble gases released in gaseous effluents from the site shall be limited to the following expressions: | |||
a | |||
: 1. During any calendar quarter: | |||
3.17x10-8 Mj (7/Q) Qj d 2.5 mrem | |||
: 2. During any calendar year: | |||
3.17x10-8 Mj (%/Q) Qj 6 5 mrem where: | |||
Qj = The release of noble gas radionuclide, i, (measured concentration x flow rate) in uCi. | |||
Releases shall be cumulative over the calendar quarter or year as appropriate. | |||
M4= | |||
The total body dose factors due to gamma emissions for each identified noble gas 3 | |||
radionuclide, mrad / year per uCi/m from Table B-1 of Rev. 1, Reg. Guide 1.109. | |||
T/Q = Shall be calculated from measured values of iodine concentration sampled at the environmental l | |||
monitoring station in Laurel Ridge. This | |||
' | |||
measured value shall be increased by a factor of 2 when calculating the body dose limits. | |||
! i l | |||
i l | |||
i | |||
) | |||
. | |||
., | |||
ACTION: With the calculated air dose from radioactive noble gases in gaseous effluent exceeding any of the above | |||
. | |||
limits, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, a special report which identifies the cause(s) for exceeding the limit (s) and defines corrective actions to be taken to reduce the releases. | |||
: c. The dose to an individual from radiciodines in gaseous effluents released to unrestricted areas shall be limited to the following expression: | |||
i | |||
: 1. During any calendar quarter: | |||
3.17x 10-8 (Rj WQj):$ 7.5 mrem, and | |||
' | |||
: 2. During any calendar year: | |||
3.17x10-8 (Rj WQj) 615 mrem where: | |||
The release of radiciodines in gaseous effluents, Q5= | |||
i, in uCi. Release viall be cumulative over the | |||
, | |||
calendar quarter or year, as appropriate. | |||
The dose factor for each radiciodine, i, in R$= 3 mrem / year per uCi/m (from Reg. Guide 1.109) | |||
> except for I-125 which is determined to be as follows: | |||
Adult Thyroid (inhalation): 1.1x10-3 mrem /pCi | |||
! | |||
- _. - | |||
_. -_. | |||
. | |||
Infant Thyroid (inhalation): 6.8x10-3 mrem /pci (ingestion): 8.9x10-3 mrem /pCi W = The average dispersion parameter for estimating the dose to an individual in the controlling location from radiciodines in gaseous effluents released to unrestricted areas. | |||
- 3 | |||
, | |||
W = ( X/Q) for the inhalation pathway, in sec/m (as determined in 3.8.2a). | |||
W=(h)forthefoodpathways,inm-2 the calculated dose from the release of ACTION: With radioiodines exceeding any of the above limits, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases. | |||
NOTE: The present controlling dose pathway ir via infant inhalation at the Laurel Ridge Residential site. If the | |||
' | |||
Land Use Census (Section 3.10 of this specification) identifies a location or pathway which yields a calculated dose or dose commitment greater than via the presently calculated dose pathway, the dispersion parameter (%/Q or D/Q) and dose f actor (R j ) for this more restrictive pathway shall be used in this specification. | |||
. | |||
3.8.3 Liquid Effluent Releases | |||
: a. Liquid ' waste from all radioactive operations shall be collected in hold tanks. | |||
: b. Before release from the hold tanks, the liquid waste shall be sampled and the activity level measured. | |||
: c. Liquid waste shall not be released from the site unless its activity concentration, including dilution with non-radioactive waste water, is below that specified in 10 CFR, Part 20, Appendix B, Table II, Column 2. This activity concentration shall be determined at least once per month by an analysis of a composite sample of all tanks released j during that period. | |||
: d. Total radioactivity released in liquid effluents shall not exceed 0.01 Ci (Sr-90 equivalent) in any 12-consecutive month | |||
, | |||
period. If the above limit is exceeded, make a special report to the NRC within 30 days explaining the cause of | |||
' | |||
exceeding the limit and the corrective action to reduce the release to within the limit. | |||
: e. Records of and reports on liquid radioactive effluent | |||
- | |||
releases shall be as specified in Section 6 of these Technical Specifications. | |||
3.9 Radiological Environmental Monitoring The radiological environmental monitoring program shall be conducted as specified in Table 3.9.1. The results of analyses performed on the radiological environmental monitoring samples shall be summarized in an Annual Radiological Environmental Report. | |||
l | |||
! | |||
I | |||
_ _ . | |||
3.10 Land Use Census A land use census shall be conducted at least once per 12 months between June 1st and Oct. 1st, and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 square feet producing fresh, leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. | |||
ACTION: With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than at a location for which dose is currently being calculated in Specification 3.8.2b and from which samples are currently being obtained in accordance with Specificaton 3.9, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted. | |||
3.11 Bases for Environmental Specifications | |||
: a. Specification 3.8.1 is provided to ensure that the -dose at the exclusion area boundary from gaseous effluents from the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the l | |||
; | |||
! | |||
i i | |||
. | |||
restricted area boundary to 6 500 mrem / year, to the total body. These release rate limits also restrict, at all times, the corresponding thyroid dos *e rate above background via the inhalation pathway to 61500 mrem / year. | |||
: b. Specification 3.8.2 is provided to demonstrate compliance with 10 CFR 20.l(c) which requires releases of radioactive materials released to unrestricted areas to as low as reasonably achievable. The action statements provide the operating flexibility and at the same time implement the design objective of minimizing the release to unrestricted areas to as low as reasonably achievable. The specifications for noble gas releases are based on limiting the total body dose at the limiting populated area to less than 5 mrem / year. The specification for radiciodine is based on the assumption that 'the limiting dose pathway for these radioisotopes is via infant inhalation at the Laurel Ridge , | |||
Residential site, and limits the inf ant thyroid dose to less than 15 mrem / year. | |||
: c. The radiological monitoring program required by specification 3.9 provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides, which lead to the nighest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected cn the basis of the effluent measurements and modeling of the environmental exposure pathways. This monitoring program may change based on operational experience and results of the land use census. | |||
i | |||
. | |||
_. ._ .. . . . | |||
! | |||
: j. d. Specification 3.10 is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the technical specification limit of dose, 4 | |||
' | |||
via the most restrictive dose ' pathway and the monitoring l program, can be made if required by the results of this Census, i | |||
.i 4 | |||
1 | |||
; | |||
3 f | |||
i i | |||
. | |||
1 | |||
! | |||
! | |||
i I | |||
. | |||
! | |||
i | |||
< | |||
-n. . , ...~,. . . _ - .. | |||
n...g, . , . , , . ,ane. - . . , , - ....,,,,.,,.,,--..,-e.. | |||
- - ~, ,-...a-- | |||
._ - .-. ._ | |||
TABLE 3.9.1 (cont'd) | |||
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations Collection Frequency of Analysis | |||
: 3. INGESTION Food Products Location to be At time of harvest. *I-131 analysis. | |||
determined from One sample of broad Land Use Census. leaf vegetation. | |||
Water Indian Kill inlet. Monthly. Gross beta - Monthly. | |||
Indian Kill Monthly. Gross beta - Monthly. | |||
outlet. | |||
Warwick Brook. Monthly. Gross beta - Monthly. | |||
Sterling Lake Monthly. Gross beta - Monthly. | |||
outlet Ramapo River Monthly. Gross beta - Monthly. | |||
O The maximum values for the lower limit for I-131 are 7x10-2 pC1/m3 airborne concentration and 60 pCi/kg, wet weight leafy vegetables. . | |||
> | |||
-= _. _ - ~_ - - -. . - . - _ _ _ . - _ . - - . - -. - .. | |||
TABLE 3.9.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM | |||
, Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations Collection Frequency of Analysis | |||
: 1. AIRBORNE | |||
: a. Radioiodine 1 sample from 380 Continuous operation Radioiodine canister. | |||
. and Particu- meters NE of of sampler with | |||
* Analyze at least | |||
' | |||
lates stack. sample collection as once per 7 days for required by dust I-131. | |||
I sample from loading but at least Laurel Ridge once per 7 days. Particulate sampler. | |||
Area. Analyze for gross beta radioactivity 24 hours following filter change. | |||
Perform gamma isotopic | |||
' analysis on each sample when gross beta activity is 10 times the mean of control samples for any | |||
. | |||
medium. Perform gamma isotopic analysis on composite (by location) sample at least once per 92 days. | |||
: 2. DIRECT RADIATION Same as #1 above. At least once per Gamma dose. At least once 31 days. per 31 days. | |||
At least once per Ganna dose. At least once 92 days. (Read- per 92 days. | |||
out frequencies are determined by type of dosimeters selected. | |||
. | |||
. | |||
4.0 SURVEILLANCE REQUIREMENTS | |||
* 4.1 General The requirements listed below generally prescribe tests or inspections to verify periodically that the performance of required syst ms is in accordance with specifications given above in Sections 2 and 3. In all instances where the specified frequency is annual, the interval between tests is not to exceed 14 months; when semiannual, the interval should not exceed 7 months; when monthly, the interval shall not exceed 6 weeks; when weekly, the interval shall not exceed 10 days; and when daily, the interval shall not exceed 3 days. | |||
4.2 Safety Channel Calibration A channel calibration of each safety channel shall be performed annually (see Sect. 3.2.3). | |||
i t 4.3 Reactivity Surveillance | |||
: 1. The reactivity worth of each control rod (inclucing the regulating rod) and the shut-down margin shall be determined whenever operation requires a reevaluation of core physics parameters, or annually, whichever occurs first. The rod worth will be determined using the reactivity-period or rod-drop methods. | |||
: 2. The reactivity worth of an experiment shall be estimated, or measured at low power, before conducting the experiment. | |||
. | |||
_ - _ _ _ ._ . _ _ _ _ | |||
- | |||
: 3. ' Boron / Carbide rods shall be gauged quarterly and any dimensional changes reported promptly to the Commission. | |||
Silver / Indium / Cadmium control rods shall be gauged annually, or, in the case of newly installed rods, at the end of the first six months. If any Ag/In/Cd rod should be found not be meet the acceptance criteria it shall be l removed from service. In addition, all other rods l manufactured of the same batch shall be inspected. | |||
4.4 Control and Safety System Surveillance | |||
: 1. The scram time shall be measured annually. If a control rod is removed from the core temporarily, or if a new rod is installed, its scram time shall be measured before reactor operation. If the bridge is moved, the scram time will be measured before subsequent reactor operation. | |||
: 2. A channel test of each measuring channel in the reactor safety system shall be performed monthly or prior to each reactor operating period whichever occurs first unless the preceding shut-down period is 8 hours or less. A channel test before startup is, however, required on any channel receiving maintenance during the shut-down period. | |||
: 3. A channel check of each measuring channel (except for the pool level) in the reactor safety system shall be performed daily when the reactor is in operation. | |||
4.5 Radiation Monitoring System | |||
: 1. The excursion, stack, and area monitors shall be calibrated l | |||
annually. | |||
l | |||
: 2. The excursion, stack, and area monitors shall receive a channel test monthly. | |||
l | |||
, | |||
4 | |||
. | |||
: 3. The excursion, stack, and area monitors shall receive a channel test monthly. | |||
4.6 Engineered Safety Features | |||
: | |||
4.6.1 Excursion Monitor: see above 4.5. | |||
4.6.2 Emergency Generator | |||
: 1. The ability of the emergency generator to start, to run normally, and to generate 440 VAC shall be checked weekly. | |||
: 2. The generator shall be tested for its ability to accept, via the automatic transfer switch, the reactor electrical load once every six months. A commercial power outage and subsequent pickup of load by the emergency generator will a count as a successful load test. | |||
4.6.3 Containment | |||
: 1. The efficiency of the charcoal filters and of the absolute filters in the emergency exhaust system shall be measured annually and the flow rate verified. | |||
: 2. The operability of the evacuation alarm and containment isolation system shall be tested, and negative pressure verified, semiannually. A utility power outage may be used | |||
, to initiate such tests. | |||
4.7 Reactor Fuel i 1. Upon receipt from the fuel vendor, all fuel elements shall be visually inspected and the accompanying quality control documents checked for compliance with specifications. | |||
l l | |||
_ . | |||
__ | |||
: 2. Each new fuel element will be inspected for damage and flow obstructions prior to insertion into the core. | |||
4.8 Sealed Sources The antimony-beryllium sealed source shall be leak tested in accordance with the procedures described in the application for license amendment dated March 21, 1963, except that the frequency of leak testing will be in accordance with 10 CFR Part 34.25(b). The strontium-90 sealed source shall be tested for leakage and/or contamination semiannually. | |||
4.9 Pool Water i | |||
: 1. The pH and specific resistance of the pool water shall be determined each week. | |||
: 2. An analysis of the pool water for radioactive material shall be done at monthly intervals. This analysis is to include Sb-124 as an indicator of Sb-Be neutron source integrity. | |||
I | |||
: 3. Activity of the pool water will be measured weekly. | |||
4.10 Core Spray The core spray in the reactor operating position shall be tested for operability semiannually. ! | |||
I | |||
, | |||
, | |||
y -- | |||
.*4 | |||
1 l | |||
4.11 Flux Distribution In order to verify that power gradients among fuel elements do not cause peaking factors to exceed those used in the bases of Sect. 2.1, the radial neutron flux distribution will be determined whenever a significant core configuration change is made. | |||
5.0 DESIGN FEATURES Those design features relevant to operation safety and to limits that have been previously specified are described below. These features shall not be changed without appropriate review. | |||
5.1 Reactor Fuel | |||
, | |||
Fuel elements shall be of the general MTR/0RR type with thin plates containing uranium fuel enriched to about 93% U-235 and clad with aluminum. The fuel matrix may be fabricated by alloying high purity aluminum-uranium or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in fine powder form. Fuel matrix may also be fabricated from uranium oxide-aluminum (U 0 -Al) 38 using the powder metallurgy process. Elements shall conform to these nominal specifications: | |||
Overall Size: 3 in., x 3 in. x 34 in. | |||
Clad Thickness: 0.015 in. | |||
Plate Thickness: 0.050 in. | |||
Water Channel Width: 0.12 in. | |||
No. of Plates: standard element - 16 fueled plates (min.) | |||
window element - 16 fueled plates (min.) | |||
control element - 9 fueled plates (min.) | |||
partial element - 9 fueled plates (min.) | |||
Plate | |||
==Attachment:== | |||
swagged or pinned. | |||
Fuel ',antent (Total): 200 g U-235 nominal. | |||
Fuei Burnup: The fuel burnup shall not exceed .94x1021 l fission /cm3 except for U38 0 -Al which l shall not exceed 1.5x1021 fission /cm3, | |||
5.2 Control and Safety Systems Design features of the components of this system (3.2.2, 3.2.3) that are important to safety are given below. | |||
5.2.1 Power Level (Normal Channels) | |||
For this function three independent measuring channels are provided, two of which are required to be operable as a minimum. Each channel covers reliably the range from about 25% | |||
to 150% (of 5 MW). Each channel comprises an uncompensated | |||
- | |||
baron-coated ion chamber feeding an amplifier that controls electronic switches in the DC current that flows through each control rod electromagnet. Each channel controls and scrams all control rods. Each channel is fail-safe. The "f ast" scram (~ 5 ms) from each channel also produces, and is backed up by a | |||
" slow" scram (~20 ms) through interruption of the AC supply to the rod electromagnet DC power supply. Each channel indicates power level on a panel meter allowing channel checks to be done during reactor operation. Each chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power. | |||
5.2.2 Power Level (Intermediate) Channel For this function, a single channel is provided, covering reliably the range 10-3% to 300% (of 5 MW) with a logarithmic output incication on both a panel meter and a chart recorder. | |||
To cover the range under all core conditions, a gamma-compensa-ted boron-ion chamber is used to supply a logarithmic amplifier. | |||
The chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power. Rate of change of power information is also derived, in the form of a period, that can produce a fast | |||
scram (and backup slow scram) in the same way as in Sect. | |||
5.2.1. From this channel is also derived control and inhibit actions, viz. bypassing of count rate channel functions, bypassing of flow and flapper scrams, reversal and inhibit of centrol rods. To negate the effect of overcompensation in the ion chamber, which can occur under certain conditions even in an initially undercompensated chamber, provision is made to supply an adjustable small current to the channel amplifier (up to 1.5x10-10 A) so as to facilitate startup. | |||
5.2.3 Count Rate Channel A fission chamber is used to supply pulses to a linear amplifier and logarithmic count rate circuitry. Pulse height discrimination selects pulse amplitudes that correspond to fission events and rejects those from alpha particles. Count rate on a logarithmic scale is displayed on a panel meter and a chart recorder. In addition, count rate period information is derived and similarly displayed. The channel covers a range of 4 | |||
1-10 cps, corresponding roughly to 0.25 mw - 2.5 W, but the upper limit can be increased many decades by repositioning the chamber. The motor-operated chamber drive is operated from the control room, the drive position being indicated on a meter. To prevent control-rod withdrawal when the neutron count rate information may not be reliably indicated, inhibits are provided on count rate and period, and when the fission chamber is being repositioned. All except the latter are bypassed at a power of | |||
>50 W. A scaler is also provided for obtaining accurate values at low count rates if needed (e.g., approach to critical with new fuel or new core configuration). | |||
.____ | |||
5.2.4 Neutron Source For obtaining the reliable neutron information necessary for startup from a cold shutdown condition, an antimony-beryllium neutron source is provided for insertion into the core as needed. This source, nominally 50 curies of Sb-124, is renewed by neutron activation in the core. Its presence in the core is not essential except after extended shutdowns. Integrity of the source is checked by periodic sampling of pool water (4.9). | |||
5.3 Rod Control System 5.3.1 Control Rods Up to five control rods are provided for the control of core reactivity. These rods may be either boron-carbide or silver-indium-cadmium (see 4.3.1) . Individual integral worths vary from about 1-4% AK, depending on position and core configuration. The rods are coupled to drive shafts through electromagnets that allow release of the rods within 50 ms after receiving a scram signal. Position indicators on the control console show the extent of withdrawal for each rod and a digital readout can be switched to any one rod. To limit the rate of reactivity increase upon startup, the rod drive speeds are limited to 5 in/ min. and no more than two rods can be withdrawn simultaneously. Switches on the guide tubes attached to the control fuel elements are arranged to produce a scram if any guide tube is lifted. This guards against lifting of the attached fuel element. | |||
5.3.2 Regulating Rod One regulating rod is provided to aid in fine control and | |||
maintenance of constant reactor power for long periods. The rod is non-fueled, is limited to a total worth of 0.6% oK for safety reasons (3.1.4), and can be either manually or servo-controlled. The drive speed is 24 in./ min. Coarse and fine position readouts are provided. | |||
5.4 Cooling System 5.4.1 Primary Cooling System Core cooling is effected by gravity flow of demineralized water from the reactor pool to an underground . holdup tank that provides an approximate 10-minute delay to allow N-16 activity to decay. The water is then pumped back to the pool through the | |||
' | |||
primary side of a heat exchanger where heat is transferred to a secondary cooling system. The holdup tank is vented to the building exhaust duct. The driving force for the coolant is the fixed head between the pool overflow gutter and the water level in the boldup tank, the latter being fixed by the total volume of water in the system. Flow is adjusted to a desired amount with a valve in the core exit line. Core cooling is not immediately affected by pump f ailure as a flow will continue until the water levels equalize; neither will the pool be drained. To prevent leakage of water through the pool walls, a continuous steel shell is located within the concrete pour of the pools. All embedments penetrating the pools are welded to this shell. To eliminate corrosion of inaccessible piping, the embedded portion of the reactor primary cooling piping under the pools is stainless steel. To change over automatically to natural convection cooling at low flow rates, a weighted flapper valve seals tne core exit plenum. This valve, held closed by the core pressure drop, opens by gravity when the flow drops to j approximately 700 gpm. Leakage at the flapper valve seat, or in l the plenum, is monitored by a plenum leak detector that senses l | |||
l l | |||
1 | |||
] | |||
. | |||
4 | |||
_42 plenum pressure increase and alarms in the control room. | |||
Primary flow is measured by taking the pressure drop across an orifice plate in the core exit line, indication being both in the pump room and on a recorder in the control room. | |||
, Temperature sensors in the pool (above the core) and in the core exit line allow the core T to be measured. These are resistance thermometers having alternative recorder and digital readout in the control room. Float switches are provided to monitor pool level. Normal pool level (at the overflow gutters) provides 24 ft. of water over the top of the fuel. | |||
5.4.2 Secondary Cooling System Reactor power transferred through the heat exchanger is dissipated to the atmosphere via a cooling tower. To minimize corrosion,the exchanger has stainless steel shell and tuces. To prevent water from entering the secondary system should a tube leak occur, the static pressure in the secondary is made higher than that of the primary through the relative elevations of the two systems. | |||
5.4.3 Core Spray For backup in the event a hot core is exposed, two spray nozzles are located at the two alternative operating positions of the reactor. They are controlled from a manually-operated valve located outside the reactor building. | |||
, | |||
l l 5.5 Containment System | |||
. | |||
l l | |||
! | |||
. | |||
5.5.1 Physical Features | |||
* The containment structure consists of the reactor building, with a free air volume of about 7700 m . This building houses the 3 | |||
reactor, the primary cooling system including holdup tank, and the heat exchanger. Personnel access is via double airlock doors or sliding doors with inflatable seals. A 12-foot deep water-filled canal penetrates the building with containment provided by a 25-in. deep water-seal weir. A water-tight gate with inflatable seals can be used to shut the canal off from the reactor pools when needed. Ventilation access to the building is through pneumatically operated damper valves that can be used These dampers are fail-shut upon to seal the building. | |||
reduction of air pressure. | |||
5.5.2 Emergency Sequence While negative pressure within the containment building is not a requisite for reactor operation, it is required in the event of a release for the controlled-release containment of airborne radioactive material. The emergency sequence is initiated either automatically by the excursion monitor (see 3.3) or manually by the console operator. The sequence is that all air supply ducts and the pool sweep dampers are closed immediately, followed later by the exhaust duct damper as soon as negative pressure (-l in, w.g.) is attained in the building (but not more than 7 sec. later). Closure of the pool sweep damper prevents activity released above the core from reaching the exhaust duct before it closes. Also closed immediately are the isolation valves in the vent line and the air purge to the holdup tank. | |||
This prevents activity in this tank from reaching the exhaust duct. Upon closure of the exhaust duct damper, the emergency exhaust f an starts and maintains a nominal negative pressure in the containment building. This f an exhausts building air at a | |||
_ | |||
low rate (f 200 cfm) through absolute and charcoal filters before connecting into the normal exhaust duct. The latter discharges to the atmosphere through a stack at a high elevation. The entire evacuation sequence is fail-safe upon loss of utility electric power. It will operate with either utility or emergency generator power. | |||
5.5.3 Exhaust Duct Monitor (" Stack Monitor") | |||
Air in the exhaust duct is continuously sampled for particulate iodine, and gaseous activities, each being read by separate detectors. The relative proportions of each type of activity i | |||
can thus be determined. The results are indicated on chart records, with repeaters in the control room. Detection or indication of a release is not dependent on all three detectors being operational, for any release will have associated with it all three types of activity or will affect each detector to some extent. Alarms, when setpoints are exceeded, are given at the monitor and repeated in the control room. | |||
5.6 Fuel Storage 5.6.1 New Fuel Unirradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with the Security Plan. Elements are stored upright in metal racks in | |||
, | |||
which the separation between elements is a minimum of 2 inches. | |||
' | |||
With such an arrangement, subcriticality is assured (Ref. 4). | |||
5.6.2 Irradiated Fuel | |||
< | |||
Irradiated fuel is stored upright under water in the storage pool I | |||
within the reactor building in criticality-safe racks. Each rack accommodates 16 elements in wells with center-to-center spacing | |||
of 6 inches. Ref. 5 states that an infinite number of elements so stored would be subcritical. Each well has a bottom hole to allow circulation of water for cooling. | |||
6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The organization for the management and operation of the reactor facility shall be as a minimum the structure shown in Fig. 2. | |||
Job titles shown are for illustration and may vary. Four levels of authority are provided, as follows: | |||
' | |||
Level 1: Individual responsible for the facility license and site administration. | |||
Level 2: Individual responsible for the reactor facility operation and management. | |||
Level 3: Individual responsible for daily reactor operations. | |||
Level 4: Reactor operating staff. | |||
4 The Nuclear Safeguards Committee shall report to Level 1. | |||
Radiation safety personnel shall report to Level 2 or higher. | |||
, | |||
J 6.l.2 Responsibility Responsibility for the safe operation of the reactor facility shall be within the chain of command shown in Figure 2. | |||
I Management levels, in addition to having responsibility for the policies and operation of the reactor facility shall be responsible - for safeguarding the public and f acility personnel from undue radiation exposures and for adhering to all requirements of the operating license and technical ! | |||
specifications. In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications. | |||
l | |||
_ | |||
FIGURE 2 I | |||
l LEVEL 1 | |||
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r - - GENERAL MANAGER i | |||
1 l | |||
1 | |||
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1 RADIATION LEVEL 2 REVIEW SAFETY __ _ _ _ | |||
OPERATIONS AND STAFF MANAGER AUDIT | |||
' | |||
COMMITTEE | |||
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f LEVEL 3 REACTOR SUPERVISOR | |||
: | |||
1 LEVEL 4 OPERATING STAFF (LICENSEDOPERATORS) | |||
. | |||
6.1.3 Staffing | |||
: a. The minimum staffir.3 when the reactor is not secured shall be: | |||
: 1. A licensed Reactor Operator in the control room. | |||
: 2. A second licensed reactor operator present at the reactor facility. Unexp~cted absence for two hours is acceptable provided immediate action is taken to obtain a replacement. | |||
: 3. A licensed Senior Reactor Operator shall be readily available on call. | |||
: 4. A member of the operating shall shall be designated by Level 2 management as knowledgeable in radiation control. | |||
: b. Events requiring tne presence of a Senior Operator: | |||
: 1. All fuel-element or control-rod alterations within the reactor core region. | |||
: 2. Relocations of any experiments witn reactivity worth greater tnan or equal to one dollar. | |||
: 3. Recovery from unplanned or unscheduled shutdowns unless they are of. a type excluded by the Level 2 authority. | |||
Such exclusions shall be posted in the control room or placed in the appropriate procedures. Furthermore, the presence of a senior operator at the facility shall not be required during recovery from unplanned or unscheduled shutdown or significant reduction in power in instances which result from: | |||
: 1. Electrical power interruptions from internal or external f ailures exclusive of power supply failures of the reactor instrumentation, control and safety systems; | |||
- _ _ _ = - - | |||
J 2. False signals which, in the opinion of the Senior Operator, were properly verified to be false and to have resulted from monitoring, experimental, or control equipment, or from personnel inadvertence; and | |||
: 3. Intentional shutdowns made by the Reactor Operator which are not related to the safety of the reactor; provided that prior to the initiation of' such recovery, the Senior Operator shall be notified of the shutdown of power reduction, and shall determine that the shutdown was caused by one of the enumerated occurrences, and shall determine that his presence at j the facility during recovery is not required. | |||
6.1.4 Selection and Training of Personnel The selection, training, and requalification of personnel | |||
' | |||
shall meet or exceed the requirements of ANS-15.4/N 380 and Appendix A of 10 CFR Part 55 and be in accordance with the requalification plan approved by the Commission. | |||
6.1.5 Review and Audit The independent review and audit of reactor facility operations shall be performed by the Nuclear Safeguards Committee. | |||
6.1.5.1 Composition and Qualifications The Nuclear Safeguards Committee shall be composed of a minimum of 5 members. The members shall collectively provide I a broad spectrum of expertise in the appropriate reactor technology. Members and alternates shall be appointed by and report to the Level 1 authority. They may include t | |||
- . - - | |||
.. -. _ - | |||
individuals from within and/or outside the operating organization. Qualified and approved alternates may serve in the absence of regular members. | |||
6.1.5.2 Charter and Rules i | |||
The committee shall function under the following operating rules: | |||
: a. Meetings shall be held not less than semi-annually or more frequently as circumstances warrant, consistent with effective monitoring of facility activities. | |||
: b. A quorum shall consist of not less than one-half the membership, where the operating staff does not constitute | |||
. | |||
a majority. | |||
: c. Subgroups may be appointed to review specific items. | |||
4 | |||
: d. Minutes shall oe kept, and shall be disseminated to members and to the Level I authority within one month j after the meeting. | |||
: e. The Committee shall appoint one or more qualified individuals to perform the Audit Function. | |||
l 6.1.5.3 Review Function The following items shall be reviewed by the review group or a subgroup thereof: | |||
: a. Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question. | |||
: b. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility | |||
, | |||
equipment, or systems having safety significance. | |||
: c. Tests-and experiments in accordance with section 6.3. I i | |||
i | |||
* -% | |||
.. | |||
. | |||
2 | |||
: d. Proposed changes in technical specifications, license, or charter. | |||
' | |||
: e. Violations of technical specifications, license, or | |||
; | |||
charter. Violations of internal procedures or instructions having safety significance. | |||
: f. Operating abnormalities having safety significance, and audit reports. | |||
: g. Reportable occurrences listed in section 6.5.3. | |||
6.1.5.4 Audit Function | |||
. | |||
The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Where necessary, discussions with responsible personnel shall take place. In no case shall the individual or individuals conducting the audit be immediately | |||
; responsible for the area being audited. Tne following items shall be audited: | |||
: a. The conformance of facility operations to the technical specifications and applicable license or charter conditions, at least once per calendar year (interval not to exceed 18 months). | |||
: b. The retraining and requalification for the operating - | |||
staff, at least once every other calendar year (interval | |||
' | |||
not to exceed 30 months). | |||
: c. The results of actions taken to correct deficiencies occurring in reactor facility equipment, systems, structures, or methods of operations that affect reactor safety, at least once per calendar year (interval not to exceed 18 months), | |||
: d. The reactor facility Security Plan and implementing procedures at least once every other calendar year (interval not to exceed 30 months). | |||
l | |||
~ _ _ _ _ _ . . . - _ . | |||
Deficiencies uncovered that affect reactor safety shall immediately be reported to the Level 2 authority. A written report of the findings of the audit shall be submitted to the Level 1 authority and the Nuclear Safeguards Committee members within 90 days after the audit has been completed. | |||
6.2 Procedures There shall be written procedures for, and prior to, initiating any of the activities listed in this section. The procedures shall be reviewed by the Nuclear Safeguards Committee and approved by Level 2 or designated alternates, and such reviews and approvals shall be documented. Several of the following activities may be included in a single manual or set of procedures or divided among various manuals or procedures. | |||
: a. Startup, operation, and shutdown of the reactor. | |||
: b. Fuel loading, unloading, and movement within the reactor, | |||
: c. Routine maintenance of major components of systems that could have an effect on reactor safety, | |||
: d. Surveillance tests and calibrations required by the technical specifications or those that may have an effect on reactor safety. | |||
: e. Personnel radiation protection, consistent with applicable regulations. | |||
: f. Administrative controls for operat ions ar.d maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity. | |||
: g. Implementation of the Security Plan and the Emergency Plan. l Substantive changes to the above procedures shall be made only after documented review by the Nuclear Safeguards | |||
! | |||
f | |||
, | |||
Committee and approval by Level 2 or designated alternates. | |||
Minor modifications to the original procedures which do not change their original int'ent may be made by the Level 3 authority (Reactor Supervisor) and must be approved by Level 2 or designated alternates within 14 days. Temporary changes to the procedures that do not affect reactor safety may be made by a Senior Reactor Operator and are valid for a period of one month. Such temporary changes shall be documented and reported to Level 2 or designated alternate. | |||
6.3 Experiment Review and Approval | |||
: a. All new experiments or classes of experiments that could affect reactivity or result in release of radioactive materials shall be reviewed by the Nuclear Safeguards Committee. This review shall assure that compliance with the requirements of the license, technical soecifications, and applicable regulations has been satisfied, and shall be documented. | |||
Prior to review, an experiment plan or proposal shall be prepared describing the experiment including any safety | |||
' | |||
considerations. | |||
Review comments of the Nuclear Safeguards Committee setting forth any conditions and/or limitations shall be documented in Committee minutes and submitted to Level 2. | |||
: d. All new experiments or classes of experiments shall be approved in writing by Level 2 or designated alternates prior to their initiation. | |||
: e. Substantive changes to approved experiments shall be mde only af ter review by the Nuclear Safeguards Committee and written approval by Level 2 or designated alternates. | |||
Minor changes that do not significantly alter the experiment may be approved by the Level 3 authority (Reactor Supervisor). | |||
: f. Approved experiments shall be carried out in accordance with established approved procedures. | |||
6.4 Required Actions - | |||
6.4.1 Action to be Taken in Case of Safety Limit Violation | |||
: a. The reactor shall be shutdown, and reactor operations shall not be resumed until authorized by the. Commission. | |||
: b. The safety limit violation shall promptly be reported to | |||
' | |||
the Level 1 authority or designated alternates. | |||
: c. The safety limit violation shall be reported to the Commission in accordance with section 6.5.3. | |||
: d. A safety limit violation report shall be prepared. The report shall describe the following: | |||
: 1. Applicable circumstances leading to the violation. | |||
: 2. Effect of the violation upon reactor facility l | |||
' | |||
components, systems, or structures. | |||
: 3. Corrective action to be taken to prevent recurrence. | |||
The report shall be reviewed by the Nuclear Safeguards Committee. A follow-up report describing extant activities shall be submitted to the Commission when authorization is sought to resume operation of the reactor. | |||
6.4.2 Action to be taken in the event of an occurrence as defined in section 6.5.3, a-1, 3: | |||
l | |||
, | |||
: a. Corrective action shall be taken to return conditions to normal; otherwise, the reactor shall be shut down and reactor operation shall not be resumed unless authorized 1 by the Level 2 authority or designated alternates. | |||
: b. All such occurrences shall be promptly reported to the Level 2 authority or designated alternates. l | |||
< | |||
. . - - - - - | |||
: c. All such occurrences where applicable shall be reported to the Comission in accordance with section 6.5.3. | |||
: d. All such occurrences, including action taken to prevent or reduce the probability of a recurrence, shall be reviewed by the Nuclear Safeguards Comittee. | |||
6.5 Reports i | |||
In addition to the requirements of applicable regulations, reports shall be made to the Commission as follows: | |||
6.5.1 Startup Reports Three months after completion of requisite startup and power-escalation testing of the reactor, or nine months af ter criticality, a written report shall be submitted to the Comission. The report shall include a sumary of the following: | |||
: a. Description of measured values of operating conditions or characteristics obtained and comparison of these values with design predictions or specifications. | |||
: b. Descriptions of major corrective actions taken to obtain satisfactory operation. | |||
: c. Re-evaluation of safety analyses where measured values indicate substantial variance from those values used in the Safety Analysis Report. | |||
6.5.2 Operating Reports Routine annual reports, covering the activities of the reactor f acility during the previous calendar year, shall be submitted to the appropriate NRC Regional Office with a copy to the Director of Inspection & Enforcement within 3 months following the end of each prescribed year. Each annual operating report shall include the following information: | |||
! | |||
: a. A narrative summary of reactor operating experience, including the energy produced by the reactor. | |||
: b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence, but excluding those of the types listed in Section 6.1.3. b(3) l above. | |||
: c. Tabulation of major preventive and corrective maintenance operations having safety significance. | |||
: d. Tabulation of major changes in the reactor facility procedures, and new tests and/or experiments significantly different from those performed previously and which are not described in the Safety Analysis Report, including conclusions that no unreviewed safety questions were involved. | |||
: e. A summary of the nature and amount of radioactive effluents from the reactor facility released or discharged to the environs. The summary shall include, where practicable, an estimate of individual radionuclides present in the effluent if the estimated average release after dilution or diffusion is greater than 25% of the ccncentration allowed or recommended, | |||
: f. A summary of exposures received by f acility personnel and visitors where such exposures are greater than 25% of that allowed or recommended. | |||
: g. A summary of the calculated doses to a critical individual in the unrestricted area due to the airborne releases of | |||
' | |||
noble gases and radioiodines. | |||
, | |||
1 | |||
. -- ,- ._ 4 _ _ | |||
_ | |||
6.5.3 Special Reports (Reportable Occurrences) 4 | |||
: a. There shall be a report not later than the following working day by telephone and confirmed by telegraph or similar conveyance to the Commission to be followed by a written report within 14 days of any of the followong: | |||
: 1. Release of radioactivity from the reactor above allowed limits, as provided by section 3.8.1 of this specification. | |||
: 2. Violation of Safety Limits | |||
: 3. Any of the following: | |||
: a. Operation with actual safety-system settings less conservative than the limiting safety-system settings specified in the Technical Specifications. | |||
: b. Operation in violation of Limiting Conditions for Operation established in the Technical Specifications. | |||
: c. 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 condition is discovered during 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 is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.) | |||
: d. An unanticipated or uncontrolled change in reactivity greater than or equal to 1% AK/K. | |||
. | |||
. | |||
. | |||
. | |||
: e. Abnormal and significant degradation in reactor fuel, and/or cladding, coolant boundary, or containment boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel | |||
; | |||
and/or environment. | |||
: f. An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused an unsafe condition with regard to reactor operations. | |||
: b. A written report within 30 days to the Commission of: | |||
: 1. Permanent changes in the facility organization structure. | |||
: 2. Significant changes in the transient or accident analysis as described in the Safety Analysis Report. | |||
: 3. Exceeding the liquid effluent limit as specified in section 3.8.3d above. | |||
: c. A report within 30 days to the Commission and to New York State as specified in the ACTION items of sections 3.8.2b, 3.8.2c, and 3.10 above. | |||
6.6 Records Records of the following activities shall be maintained and retained for the periods specified below. The records may be in the form of logs, data sheets, or other suitable forms. | |||
, The required information may be contained in single, or | |||
! multiple records, or a combination thereof. Recorder charts showing operating parameters of the reactor (i.e. power level, flow, temperature, etc.) for unscheduled shutdown and significant unplanned transients shall be maintained for a minimum period of two years. | |||
l 6.6.1 Records to be retained for a period of at least five years or | |||
, | |||
for the life of the component involved, whichever is smaller. | |||
, | |||
,,. - . , - , ~ . - | |||
- | |||
. | |||
: a. Normal reactor facility operations (including scheduled and unscheduled shutdowns). Note: Supporting docume.its such as checklists, log sheets, etc. shall be maintained for a period of at least two years. | |||
: b. Principal maintenance operations. | |||
: c. Reportable occurrences. | |||
: d. Surveillance activities required by the Technical Specifications. | |||
: e. Reactor facility radiation and contamination surveys where i required by applicable regulations. | |||
: f. Experiments performed with the reactor. | |||
: g. Special Nuclear Materials (SNM) inventories, receipts, and shipments. | |||
: h. Approved changes in operating procedures. | |||
: i. Records of meeting and audit reports of the Nuclear Safeguards Committee. | |||
: j. Sealed Source leak test results. | |||
6.6.2 Records to be Retained for at least One Requalification Cycle or for the Length of Employment of the Individual whichever is Smaller: | |||
: a. Retraining and requalification of licensed operations personnel. However, records of the most recent complete cycle shall be maintained at all times the individual is employed. | |||
6.6.3 Records to be Retained for the Lifetime of the Reactor Facility: (Note: Annual reports may be used where applicable as records in this section.) | |||
: a. Gaseous and liquid radioactive effluents released to the environs, | |||
: b. Off-site environmental-monitoring surveys required by the Technical Specifications. l | |||
: c. Radiation exposure of all personnel monitored. | |||
: d. Updated drawings of the reactor facility. | |||
, . | |||
+ | |||
==7.0 REFERENCES== | |||
: | |||
: 1. Safety Analysis Report, Union Carbide Research Reactor (May 1980), Appendix 2 | |||
: 2. Final Hazards Summary Report, UCNC Research Reactor (May 1960) | |||
: 3. Reactcr Power Excursion Tests in the SPERT IV Facility. | |||
100-17000 (August 1964) | |||
: 4. Critical Experiments with SPERT-D Fuel Elements, ORNL-TM-1207 (July 14, 1965), by E.B. Johnson and P.K. Reedy, Jr. | |||
l 1 | |||
l l}} | |||
Revision as of 03:11, 14 December 2019
| ML19323F712 | |
| Person / Time | |
|---|---|
| Site: | 05000054 |
| Issue date: | 05/01/1980 |
| From: | UNION CARBIDE CORP. |
| To: | |
| Shared Package | |
| ML19323F693 | List: |
| References | |
| NUDOCS 8005290363 | |
| Download: ML19323F712 (62) | |
Text
{{#Wiki_filter:__ _ _ _ _ _ _ _ _ _ _ . . . O UNION CARBIDE RESEARCH REACTOR TECHNICAL SPECIFICATIONS License No. R-81 Docket No. 50-54 Dated: May 1, 1980 8006290363
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, i TABLE OF CONTENi$
l ! 1.0 DEFINITIONS 4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4 2.1 - Safety Limits of Reactor Operation 7 2.2 - Limiting Safety System Settings 9 3.0 LIMITING CONDITIONS FOR OPERATION 9 3.1 - Reactivity Limitations 3.2 - Control and Safety Systems 10 13 3.3 - Radiation Monitoring Systems 14 3.4 - Engineered Safety Features 3.5 - Limitations on Experiments 16 20 3.6 - Fuel 3.7 - Pool Water Quality 22 23 3.8 - Radioactive Releases (Airborne) 3.9 - Radiological Environmental Monitoring 27 28 3.10 - Land Use Census 3.11 - Bases for Environmental Specifications 28 33 4.0 SURVEILLANCE REQUIREMENTS 33 4.1 - General 4.2 - Safety Channel Calibration 33 4.3 - Reactivity Surveillance 33 4.4 - Control and Safety System Surveillance 34 4.5 - Radiation Monitoring System 34
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35 4.6 - Engineered Safety Features 35 4.7 - Reactor Fuel ' 36 4.8 - Sealed Sources 36 4.9 - Pool Water
^4.10 - Core Spray 36 37 4.11 - Flux Distribution
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37 5.0 DESIGN FEATURES 37 5.1 - Reactor Fuel 5.2 - Control and Safety Systems 38 5.3 - Rod Control System 40 41 5.4 - Cooling System 5.5 - Containment System 42 5.6 - Fuel Storage 44 45 6.0 ADMINISTRATIVE CONTROLS 45 6.1 - Organization 51 6.2 - Procedures 6.3 - Experiment Review and App. oval 52 6.4 - Required Actions 53 6.5 - Reports 54 57 6.6 - Records 59
7.0 REFERENCES
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1.0 DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Condition of Operation (LC0) are defined in , 50.36 of 10 CFR Part 50. 1.1 Safety Channel - A Safety Channel is a measuring or protective channel in the reactor safety system. J 1.2 Reactor Safety System - The Reactor Safety System is a combina-tion of safety channels and associated circuitry which forms the automatic protectise system for the reactor, or provides informa-tion which requires the initiation of manual protective action. 1.3 Operable - Operable means a component or system is capable of performing its intended function in its required manner. 1.4 Operating - Operating means a component or system is performing its intended function in its normal manner. 4 1.5 Channel Check - A Channel Check is a qualitative verification of acceptable performance by observation of channel behavior.
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1.6 Channel Test -A Channel Test is the introduction of a cali-bration or test signal into the channel to verify that it responds in the specific manner. 1.7 Channel Calibration - A Channel Calibration is an adjustment of the channel components such that its output responds, within specified range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including readouts, alarm, or trip. 1.8 Unscheduled Shutdown - An Unscheduled Shutdown is any unplanned
shutdown of the reactor, after startup nas been initiated. 1
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1.9 Reactor Shutdown - The reactor is shut down when the negative reactivity of the cold, clean core including the reactivity worths of all experiments is equal to or gre'"- than the shut-down margin. , 1.10 Reactor Operating - The reactor is considered to be operating whenever it is not shut down. 1.11 Reactor Secured - The reactor is secured when:
- a. The core contains insuf cient fuel to attain criticality under optimum conditions i moderation and reflection, or
- b. The moderator has been removed, or
- c. (1) Minimum number of control rods fully inserted as required by Technical Specifications, and (2) The console key switch is in the off position and the key is removed from the lock, and 4
(3) No work is in progress involving core fuel, core structure, installed control rods or control rod drives unless they are physically decoupled from the control rods and (4) No in-core experiments are being moved or serviced with a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, whichever is smaller. 1.12 T.ue Value - The True Value of a parameter is its actual value at any instant. 1.13 Measured Value - The Measured Value of a parameter is as it appears on the output of a measuring channel. 1.14 Measuring Channel -A Measuring Channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a parameter. 1.15 Reportable Occurrence - A Reportable Occurrence is any of those conditions described in Section 6.5.3 of this specification.
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An Experiment is an apparatus, device or 1.16 An Experiment - material, placed in the reactor core, in an experiment f acility, or in line with a beam of radiation emanating from the reactor, , excluding devices designed to measure reactor characteristics such as detectors and foils,
- a. Secured Experiment - Any experiment, experiment facility, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor core. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, or other forces which are normal to the operating environment of the experi-ment (or by forces which can arise as a result of credible malfunctions).
- b. Movable Experiment - A movable experiment is one which may be removed, or manipulated while the reactor is inserted, critical.
- c. Untried Experiment - is a single experiment or class of experiments that has not been previously evaluated and ap-proved by the Nuclear Safeguards Committee.
1.17 Experiment Facilities - An Experiment Facility is any structure, orient, device or pipe system which is intended to guide, manipulate, control the environment or otherwise position, facilitate a multiplicity of experiments of similar character. 1.18 Control Rod - A control rod is a rod fabricated from neutron absorbing material which is used to compensate for fuel burnup, temperature, and poison effects. A control rod is magnetically coupled to its drive unit allowing it to perform the safety function when the magnet is de-energized.
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1.19 Readily Available on Call - Readily available on -call means an individual who (1) has been specifically designated and the designation known to the operator on duty, (2) keeps the operator on duty informed of where he may be rapidly contacted (e.g. by i phone, etc.) (3) is capable of getting to the reactor f acility within a reasonable time under normal conditions (e.g., I hr. or within a 30 mile radius). d 1.20 Scram Time - is the elapsed time between the instant a limiting safety system set-point is reached and the instant that the slowest control rod is fully inserted. 1.21 Safety Limits - are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against release of radio- . activity. The principal physical barrier is the fuel cladding. 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits of Reactor Operation 2.1.1 Limits in Forced Cooling Mode
- a. Applicability - This specification applies to the variables that affect thermal and hydraulic performance of the core during forced cooling. They are:
(1) Power in MW. (2) Flow in GPM. (3) Height of water above the core.
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- b. Objective - To assure fuel cladding integrity.
- c. Specifications .
(1) The maximum steady power level under various flow con-ditions shall be as shown in Figure 1. ! (2) The pool water level shall not be less than 20 feet above the core.
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_ _ __ . .
! d. Bases - The analysis given in Ref. 1, Sec. Al, forms the basis for this specification. The superposition method of Gambill is used to derive the burnout heat flux as a function of primary flow rate. A safety f actor of 1.25 is applied to i allow for uncertainties in the correlation. Pool temperature
, (or core inlet temperature) is not included in the specifi-cation as this variable changes very slowly and has only a minor effect, e.g., a 10 F change results in only a 5% variation in burnout flux. The latter, however, is evaluated conservatively near the high end of the pool temperature range that is exoected in practice. A de-rating f actor can 0
- be applied for pool temperatures in excess of 120 F. The relationship between total power and peak heat flux is derived for the core situation with the greatest peaking factors, viz. a new fuel element adjacent to a central in-core flux trap. Reactor power, primary flow rate, and water level will be maintained well within safety limit specifications through limiting safety system scram settings.
(see2.2.1). i . I
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2.1.2 Limits in Free Convection Mode
- a. Applicability - This specification applies to the thermal and hydraulic variables affecting the core during natural convection cooling. Tney are:
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(!) Power in MW. (2) Height of water above the core,
- b. Objective - To assure fuel cladding integrity.
- c. Specifications -
(1) The maximum reactor power level shall be 6.7 MW. q a (2) The pool water level shall not be less than 20 ft. above i
the top of the core. l d. Bases - The analysis given in Ref. 1, Sect. A2, forms the basis of this specification. The homogeneous method of Gambill aid Bundy, used in this analysis, has been employed succe sfully to predict natural convection burn-out in ORR and HFIR fuel. The former fuel is close in design to UCRR fuel. A safety factor of 1.24 is applied to account for 1 random variations and uncertainties. A pool temperature near U the high end of the operating range (120 F) is assumed. The safety system settings on power and pool level (2.2.2) assure adherence to these specifications.
- 2.2 Limiting Safety System Settings 2.2.1 Safety Channel Set-Points in Forced Cooling Mode
- a. Applicability - This specification applies to the setpoints of the safety channels,
- b. Objective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded.
- c. Specification - For operation in the forced cooling mode the limiting safety system settings are:
(1) Power level at any flow rate shall not exceed 7.5 MW. (2) Power level settings for various conditions of flow and of pool temperature shall be in accordance with Fig. 1. 1
(3) Coolant flow shall not be less than 1800 gpm for powers above 250 KW. (4) Pool level shall not be less than 20 f t. above the top of the core.
- d. Bases - Safety limits have been shown previously (Sec. 2.1 and Ref.1) to lie at a low flow-to-power ratio. To provide adequate assurance that these limits are not approached too closely, the LSSS are chosen conservatively so as to minimize the chance of boiling in the core. This results in a much larger flow / power ratio. In Ref. 1, Sect. A3, power levels derived using conservative correlations for incipient boiling are tabulated for various values of pool temperatures and flow rates to illustrate the resulting temperature margins.
Through a comparison with experiments at ORNL (ORR) this method is shown to be conservative. To preserve the desired temperature margins for all comoinations of variables, the LSS settings are a combination of two fixed set-points, viz. scrams at 7.5 MW and 1800 gpm, plus an adjustable one that provides automatic power reduction at a setting that depends on the pool temperature and flow rate. Rates of change of U poci temperature are very slow - a few F/hr. at most - and thus allow adequate lead time for adjustment. For a reactivity transient the case considered is the step insertion of 0.25% a K positive reactivity with the reactor operating at a steady power of 7.5 MW. The analysis given in Ref.1, Sect. B3, s%ws that the power at the end of .75 sec. (the scram time, Sect. 3.2.1 below) will be no more than 11 MW. This is well below the safety limit for this mode of operation. No automatic scram is associated with pool temperature as this parameter varies very slowly allowing ample time for appropriate operator action.
-g-2.2.2 Safety Channel Set-Points in Natural Convection Mode
- a. Applicability - This specification applied to the set-points of the safety channels,
- b. Objective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded.
- c. Specification - For operation in the natural convection mode, the limiting safety system settings are:
(1) Power Level 6 250 KW. (2) Pool level 3 20 ft. above the core,
- d. Bases - The set-points are chosen to avoid boiling in the core during routine operation with natural convection cooling. The analysis given in Ref. 1, Sec. A4, shows that a power of 0.35 MW is needed for incipient boiling to occur.
1 To allow for uncertainties a safety f actor of 1.3 is applied to this, resulting in a safety system set-point of 0.25 MW.
' The latter is well below the safety limit of 6.7 MW given above (Sect. 2.1.2) . In the case of reactivity transient, a step insertion of 0.25% AK positive' reactivity at an initial power level of 0.25 MW will, following the analysis rf Ref. 1 (Sect. 83), result in a transient power of 0.38 Mei after 1 second. The latter is well below the safety limit of 6.7 MW for the natural convection mode (2.1.2). . 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limitations 3.1.1 Shutdown Margin The minimum shutdown margin provided by control rods in the cold, xenon-free condition with the highest-worth rod fully withdrawn and with the highest-worth non-secured experiment in its most positive reactive state shall not be less than 0.5% 6K. 4
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This specification ensures that the reactor can be shut dcwn from any operating condition and remain shut down after cool-down and xenon decay even if the highest-worth control rod is stuck in its fully withdrawn condition.
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3.1.2 Excess Reactivity The core shall not be loaded with an excess reactivity of greater than 10.2% 6 K when located in the stall position and 8.2% aK when the core is located in the open pool position. 1
3.1.3 Experiments i Reactivity limits on experiments are specified in 3.5 below. ' 3.l.4 Regulating Rod The integral worth of the regulating rod shall not exceed 0.6% a K. This ensures that a malfunction of the control system cannot make the reactor prompt critical. I 3.2 Control and Safety Systems 3.2.1 Scram Time The scram time shall not exceed .75 second and the control rod magnet release time shall not exceed .05 second. In the transient analysis (Ref. 1, Sect. 83), these values were assumed, s 3.2.2 Measuring Channels The minimum number and type of measuring channels operable and providing information to the control room operator required for i reactor operation are given as follows: Channel No. Operable Operating Mode in Which Required Power Level (normal) 2 All-Power Level (intermediate) 1 All Period Channel 1 All
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Channel No. Operable Operating Mode in Which Required a Count Rate l Startup Coolant Flow 1 Forced Cooling Core ST 1 Forced Cooling Rod Position 1/ rod All Pool Temperature 1 All Pool Level 1 All i Note: a. Operable below 50 W. Bases - The normal power level instruments (" Level Safeties") provide redundant information on reactor power in the range 25%-150% of the normal operating power level of 5 MW. The intermediate power instrument (" Log N") provides usable , reactor power information in the logarithmic range 10-4%- , 300% of the normal power of 5 MW. The count rate channel covers the neutron flux range from the 4 source level ( :::1 cps) to 10 cps on a logarithmic scale. It enables the operator to start the reactor safely from a shutdown condition, and to bring the power to a level that ) can be measured by the Log N instrument. Coolant flow rate and A T instruments allow the operator to calculate reactor power and calibrate the neutron flux chan-nels in terms of power. Rod position indicators show the operator the relative positions of control rods, and enable rod reactivity cali-brations to be made. Pool temperature information allows the operator to adjust the cooling system to keep pool temperature within a pre-ferred range, and to adjust the overpower reverse set-point. (see 3.2.3).
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1 3.2.3 Safety Channels The minimum number and type of channels providing automatic action that are required for reactor operation are as follows: No. Operating
; Channel Operable Function Mode Power Level (normal) 2 Scram 0 7.5 MW All Power Level (intermediate) 1 Scram @ 3 sec. period All Reverse @ 10 sec. period All Inhibit @ 30 sec. period All c
Reverse All i Count Rate la Inhibit 0 2 cps Startup Inhibit @ 30 sec. period Startup Pool Watc. Level 1 Scram @ 22 ft. All ) Pool Temperature 1 Alarm @ 120 F All b Coolant Flow l Scram @ 1800 gpm Forced Circ. Manual Button 1 Scram All Bridge Lock 1 Scram All Guide Tube Lift 1/ rod Scram All Flapper Valve 1 Scram (above 250 KW w/ valve open) All Keyswitch 1 Scram All Notes: a. Operable below 50 W.
- b. Operable above 250 KW.
- c. Overpower reverse set-points shall be set so that the relationship of pool temperature, flow and power levels shown.
, in Figure 1 are never exceeded. , Bases - The power level scram provides redundant automatic protective action to prevent exceeding the safety limit on reactor power.
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The period scram, assisted by the intermediate level period reverse and rod inhibit, limits the rate of increase in reactor power to values that are controllable without reaching excessive power levels or temperature. These functions are not limiting safety system settings. The two inhibits on the count rate channel prevent inad-vertent criticality during cold startup that could arise from lack of neutron information or from too rapid reactivity insertion by control rods. The scram on pool level provides an adequate head of water above the core and guards against loss of coolant and loss of building containment. The overpower reverse on the intermediate power channel provides automatic action to reduce power and minimize the chance of incipient boiling in the core. The coolant flow and flapper valve scrams ensure adequate coolant ficw to prevent boiling in the core. l l The scrams on bridge lock and guide tubes prevent unplanned reactivity ,anges that could occur through core and control element movements respectively. The keyswitch scram prevents unauthorized operation of the reactor. Bypass is permitted on those parameters that can be monitored by alternate means if the initiating circuit malfunctions. 3.3 Radiation Monitoring Systems The minimum acceptable monitoring instrumentation required for reactor operation is as follows: _ __ -__
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No, Max. Alarm Type Operable Setpoint Function Excursion Monitor i SR/hr Detect high radiation: Alarm and isolate at > 5R/hr.
- Detect Particulate, gas and Ex5aust Duct Monitor 1 iodine activities; alarm in
(" Stack Monitor") Control room. Building CAM 1
** Detect particulate activity in reactor building; alarm.
50mr/hr Detect radiation (Y) in key Fixed Area Monitron 3 locations; alarm in Control room. Evacuation Switch 1 -- Alarm and initiate evacuation sequence. (manual) Note: For maintenance or repair, required radiation monitors (except for excursion monitor) may be replaced by portable or substitute instruments for periods up to 24 hours provided the function will
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still be accomplished. Interruption for brief periods to permit checking or calibration is permissible. 3.4 Engineered Safety Features These specifications apply to required equipment for the confine-ment of activity through controlled release of reactor building air to the atmosphere.
- The alarm set-point for the stack gas monitor shall not be set above a value that would result in an exposure greater than 2 mrem / hour assuming a dilution factor of 2000 and the isotope mixture determined by the most recent analysis. The alarm I set-point for the stack I-131 and stack particulate monitor shall not be set above a value corresponding to that listed in Appendix B, Table II, Column I of 10 CFR Part 20 assuming a dilution factor of 2000 and averaging over one week.
** 25% of the maximum permissible concentration at restricted areas according to Appendix B of 10 CFR 20.
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. 3.4.1 Excursion Monitor
- a. Specification: see 3.3.
- b. Basis -
This monitor senses excessive radiation at the reactor bridge and automatically initiates the " evacuation sequence", which consists of a distinctiva alarm, closure of damper valves in the building ventilation system and hold-up tank vent, and starting of the emergency exhaust fan (see 5.5.2). 3.4.2 Emergency Electric Generator
- a. Specification Equipment No. Operable Function Electric Generator i Upon loss of utility power, start automatically and supply emergency power to the exhaust fan and ventilation system controls. A six day supply of fuel shall be maintained.
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- b. Basis - Upon loss of utility power the reactor scrams auto-matically. Controlled release confinement requires the ability to run the emergency exhaust fan and to close building damper valves. The latter are pneumatically-operated but are electronically-controlled.
3.4.3 Containment
- a. Specification (1) The emergency exhaust fan shall be capable of sustaining a negative pressure within the reactor building of at least .01-in w.g. at an exhaust flow rate of not greater than 200 cfm.
(2) Filters in the emergency exhaust shall be HEPA and charcoal with tested efficiencies of 99.5% for particle removal and 95% for iodine removal respectively.
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(3) Depth of water in the canal shall be at least 10 f t. This is equivalent to a water height above the core of 22 ft. (4) At least one door of the double airlock doors and the truck doors shall be closed while the reactor is operating.
- b. Bases - To effect controlled release under accident con-
, ditions of gaseous activity present in the building atmos-phere, a negative pressure is required so that any building leakage will be inward. Reference 1 (Sec. C, 2) contains an analysis of a hypothetical accident resulting in release of airborne activity to unrestricted areas. The assumed exhaust rate is 200 cfm and the filter efficiency for elemental iodine is 95%. In the design of the containment building (5.5) the water seal in the canal is effected when the water depth is > 10 ft. , 3.5 Limitations on Experiments
.
3.5.1 Experiments l a. Applicability - This specification applies to those experi-ments installed in the reactor and its experiment facilities.
- b. Objective - The objective is to prevent damage to the reactor or excessive release of radioactive material in the event of an experiment failure and also to prevent the safety limits from being exceeded.
- c. Specificatir - Experiments installed in the reactor shall meet the following conditions:
(1) The combined worth of all experiments which can add positive reactivity to the core due to a common-mode failure shall not exceed 2% a K.
- (2) The combined worth of all non-secured experiments which can add positive reactivity to the core due to a common-mode failure shall not exceed 1.7% A K.
l
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-- . _ - _. . _ _ .
.. ,
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. '
, (3) The reactivity of any single experiment shall not exceed 0.5% A K. , (4) An experiment worth less than 0.25% 6 K may be moved ' when the reactor is critical. (5) An experiment worth more than 0.25% A K but less than 0.5% A K may be moved with the reactor subcritical by at ' least 0.75% AK. (6) All material to be irradiated in the reactor shall either be corrosion resistant or encapsulated within corrosion- resistant containers. (7) Where failure of the pressure-containing walls of an
- experiment container can cause a hazard to personnel or 1
to the reactor, the container shall be designed and tested in accordance with the applicable pressure vessel codes. (8) In-core experiments exposed to reactor water shall be designed to prevent surface boiling. } (9) Experimental apparatus, material, or equipment to be j
'
inserted in the reactor shall not interfere with the j safe operation of the reactor. l (10) The total primary coolant ficw utilized by all in-core experiments shall be limited to the same as that in six standard fuel elements. (11) Experiments on the grid-plate extension are limited to a total reactivity of 0.2% AK and a total load of 100 lbs. (12) Each class of experiment irradiation in the reactor must have been previously reviewed and approved by the Nuclear Safeguards Committee (6.8).
- d. Bases ,
(1) See Ref. 2, Sect. G 5.c. (2) It is shown in Ref. 3 that the reactor can safely self-limit a step reactivity insertion of- $2.14. This corresponds to an insertion of 2.14 x .81 = 1.73% A K. , a
- , -
,w- , ,_
_ _ _ _ _ _ _ _ _ _ _ _ _ _
-
( (3) The method of Ref. 1, Sect. 83, shows that a step insertion of 0.5% a K with the reactor critical at 5 MW (or 0.25 MW, in natural convection mode) will result at the end of .75 sec. in a power of not more than 14 MW and .4 MW, for natural convection. Each of these power levels does not exceed the corresponding safety limit. (4) Similarly it is shown that a step increase of 0.25% a K will produce a power level at the end of the scram time that is much less than the safety limit in either mode of operation. In addition, .25% a K is well within the automatic control capability of the reactor control system. (5) This specification ensures that, even with a 45% error in estimation of the reactivity of an experiment, the reactor will not be made critical. Even if the reactor were critical, the resulting period (~ 3 secs.) will automatically initiate corrective control action. (6) This requirement guards against release of activation products in the pr,imary coolant or chemical interaction with core components. (7) This specification ensures that there will be no mechanical damage to the reactor core nor hazards to personnel due to f ailure of experiment containers where pressure exists or builds up during irradiation. In the case of fueled experiments, it further ensures against hazardous and uncontrolled release of fission products into the reactor building or the environment from the same cause. (8-9) Ensures that no physical or nuclear interference with the safe operation of the reactor will occur. (10) This condition is assumed in the analysis given in Ref.
-
1, Sect. A. (11) These limits ensure that movement of these experiments will not result in reactivity changes in excess of that in Sect. c(4) above.
' .. - -
_ _ _ _ _ _ _ _ _ _ _
._
.
-
(12) Ensures that all experiments are evaluated by an inde-pendent group knowledgeable in the appropriate fields. 3.5.2 Fueled Experiments
- a. Applicability - These specifications apply to experiments containing nuclear fuel that are installed in the reactor or its experiment facilities.
- b. Oojectives - The objective is to prevent damage to the reactor, prevent excessive release of fission products in the event of an experiment failure, and also to ensure that safety limits are not exceeded.
- c. Specifications - Fuel-bearing experiments in the reactor shall meet the following conditions:
(1) All fueled experiments are to conform to the specifi-cations listed above in Sect. 3.5.1. (2) The inventory of solid fuel bearing material being irradiated in the reactor core at any time shall be limited to 200 g of source and/or 750 g of special nuclear material. (3) The inventory of solid fuel-bearing materials in a single irradiation capsule shall be limited to 200 g of source and/or 50 g of special nuclear material. (4) The fission power of an irradiation capsule containing special nuclear material shall be limited to 13 KW. (5) The iodine inventory of a single capsule shall be limited to 500 curies 131 I dose oquivalent for a doubly-encapsulated capsule and 70 curies 131 I dose equivalent for a singly-encapsulated capsule.
- d. Bases - These specifications place - limits on the f,ission product inventory in a fueled capsule such that capsule failure and the hypothetical release of all contained fission products to the reactor coolant will not result in excessive exposure to personnel on and off site.
.
The detailed analyses that form the bases of this specifi-cation are given in Ref. 1, Sect. C3. The total amount of special nuclear material permitted in the core at any time has been increased to 750 g. This increase does not affect the consequences of the release from a single capsule as analyzed in Ref. 1 for it has been established (see License Amendmeret No. 10) that failure of a single capsule will not initiate failure in other neighboring capsules. The core limit of 750 g is based on approximately 15 irradiation positions, each holding 50 g of SNM. The limit of 15 positions is dictated by availability of primary cooling capacity. 3.6 Fuel
- a. Applicability - These specifications apply to the number and condition of the fuel elements present in the core.
- b. Objective - To ensure that power is distributed in the t ore among a sufficient number of fuel elements to avoid excessive peak / average ratio, and to avoid excessive release of fission products.
- c. Specifications
,
- l. The minimum number of fuel elements in the core shall be
- 30. Each control element shall count as 1/2 fuel element for this purpose.
- 2. Control rods shall be kept within + 10% of their mean position whenever the reactor power exceeds 500 KW.
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. _ . - . .. . - _ _ _ _ _
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- 3. Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive identification, be removed from the core. An increase in the normal gaseous fission product release (due to system contamination) by a factor of 100 shall constitute initial evidence of cladding rupture and require identification of the cause.
- 4. Fuel element loading and distribution in the core shall be such that peak / average thermal flux will not exceed 3.3.
- 5. The fuel plates are composed of enriched uranium-aluminum sandwiched between high purity aluminum clad. Fuel plates may be fabricated by alloying the uranium-aluminum or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in the fine powder form. Burnup of the fuel assemblies 2I fission /cm3 . Fuel shall be limited to .94x10 plates may also be fabricated from uranium oxide-aluminum (U3g0 -Al) using the powder metallurgy 21 process and the burnup shall be limited to 1.5x10 3
fission /cm ,
- d. Bases
- 1. A minimum of 30 elements is assumed a the analysis given in Ref. 1, Sects. A-1, A-2.
- 2. This specification minimizes flux tilts that could cause concentrations or shifts in power distribution across the core. Such shifts are only significant in power operation, and thus this limitation is restricted to power levels above 10% of the normal 5 MW.
- 3. Release of fission products from the compact fuel plates used in this reactor (Sect. 5.1), due to a localized cladding defect, is confined to the defect locality. A relatively small defect, thus, cannot release large quantities of fission products. There is a normal small and variable background of fission product release due
to uranium contamination in the coolant and on fuel
- plates. It is, thus, safe to specify a recognizable and substantial increase in this background as a possible indication of cladding rupture. If the rupture were
, extensive, there would be no doubt at all of this condition.
- 4. This peak / average value is used in the Ref. 1 analysis.
3.7 Pool Water Quality
- a. Applicablity - This specification applies to primary cooling system water in contact with fuel elements. .
- b. Objective - To minimize corrosion of the aluminum cladding of fuel plates and activation of dissolved materials.
- c. Specifications
- 1. Pool water temperature will not exceed 130 F.
- 2. Pool water specific resistance is to be not less than 200,000 ohm-cm, except that for periods not greater than 14 days it may be 70,000 ohm-cm.
- 3. The pH of the pool water shall normally be maintained between 5.0 and 7.5.
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I 3.8 Radioactive Releases 3.8.1 Airborne Stack Release Limit Maximum yearly release rates for noble gases, radioiodines and particulates of half-life greater than eight days shall be ) limited by the following expression: 1 Qj (7/Q)/MPCj < 1/6 where: Qj = The average release rate for any 12-consecutive months of radionuclide, i, in gaseous effluent from tne stack in Ci/sec. MPCj= Activity concentration of radionuclide, i, as given in Table II, Column 1 of Appendix B to 10 CFR 20, in uCi/cc. 7./Q = Shall be calculated monthly from measured values of iodine concentration sampled at or above the tree line 380 meters NE of the exhaust stack. ACTION: Shoulci the _ limit of this Section be exceeded, the license shall notify the Commission within 24 hours, and take action to reduce the release to within the limits immediately. 3.8.2 Dose in Unrestricted Areas
- a. Total body dose due to noble gases releases and dose from radiciodines in gaseous effluents fnr the critical
, individuals in unrestricted areas should be calculated at least once per calendar quarter and reported in the annual report (TS 6.5.2g).
_ _ _ ___. _ _ _ _ _ _ _ _ _ _.
- b. The total body dose to any individual in unrestricted areas due to noble gases released in gaseous effluents from the site shall be limited to the following expressions:
a
- 1. During any calendar quarter:
3.17x10-8 Mj (7/Q) Qj d 2.5 mrem
- 2. During any calendar year:
3.17x10-8 Mj (%/Q) Qj 6 5 mrem where: Qj = The release of noble gas radionuclide, i, (measured concentration x flow rate) in uCi. Releases shall be cumulative over the calendar quarter or year as appropriate. M4= The total body dose factors due to gamma emissions for each identified noble gas 3 radionuclide, mrad / year per uCi/m from Table B-1 of Rev. 1, Reg. Guide 1.109. T/Q = Shall be calculated from measured values of iodine concentration sampled at the environmental l monitoring station in Laurel Ridge. This ' measured value shall be increased by a factor of 2 when calculating the body dose limits. ! i l i l i
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ACTION: With the calculated air dose from radioactive noble gases in gaseous effluent exceeding any of the above
.
limits, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, a special report which identifies the cause(s) for exceeding the limit (s) and defines corrective actions to be taken to reduce the releases.
- c. The dose to an individual from radiciodines in gaseous effluents released to unrestricted areas shall be limited to the following expression:
i
- 1. During any calendar quarter:
3.17x 10-8 (Rj WQj):$ 7.5 mrem, and '
- 2. During any calendar year:
3.17x10-8 (Rj WQj) 615 mrem where: The release of radiciodines in gaseous effluents, Q5= i, in uCi. Release viall be cumulative over the
,
calendar quarter or year, as appropriate. The dose factor for each radiciodine, i, in R$= 3 mrem / year per uCi/m (from Reg. Guide 1.109) > except for I-125 which is determined to be as follows: Adult Thyroid (inhalation): 1.1x10-3 mrem /pCi !
- _. -
_. -_. . Infant Thyroid (inhalation): 6.8x10-3 mrem /pci (ingestion): 8.9x10-3 mrem /pCi W = The average dispersion parameter for estimating the dose to an individual in the controlling location from radiciodines in gaseous effluents released to unrestricted areas.
- 3
, W = ( X/Q) for the inhalation pathway, in sec/m (as determined in 3.8.2a). W=(h)forthefoodpathways,inm-2 the calculated dose from the release of ACTION: With radioiodines exceeding any of the above limits, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases. NOTE: The present controlling dose pathway ir via infant inhalation at the Laurel Ridge Residential site. If the
'
Land Use Census (Section 3.10 of this specification) identifies a location or pathway which yields a calculated dose or dose commitment greater than via the presently calculated dose pathway, the dispersion parameter (%/Q or D/Q) and dose f actor (R j ) for this more restrictive pathway shall be used in this specification.
.
3.8.3 Liquid Effluent Releases
- a. Liquid ' waste from all radioactive operations shall be collected in hold tanks.
- b. Before release from the hold tanks, the liquid waste shall be sampled and the activity level measured.
- c. Liquid waste shall not be released from the site unless its activity concentration, including dilution with non-radioactive waste water, is below that specified in 10 CFR, Part 20, Appendix B, Table II, Column 2. This activity concentration shall be determined at least once per month by an analysis of a composite sample of all tanks released j during that period.
- d. Total radioactivity released in liquid effluents shall not exceed 0.01 Ci (Sr-90 equivalent) in any 12-consecutive month
,
period. If the above limit is exceeded, make a special report to the NRC within 30 days explaining the cause of ' exceeding the limit and the corrective action to reduce the release to within the limit.
- e. Records of and reports on liquid radioactive effluent
-
releases shall be as specified in Section 6 of these Technical Specifications. 3.9 Radiological Environmental Monitoring The radiological environmental monitoring program shall be conducted as specified in Table 3.9.1. The results of analyses performed on the radiological environmental monitoring samples shall be summarized in an Annual Radiological Environmental Report. l ! I _ _ .
3.10 Land Use Census A land use census shall be conducted at least once per 12 months between June 1st and Oct. 1st, and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 square feet producing fresh, leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. ACTION: With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than at a location for which dose is currently being calculated in Specification 3.8.2b and from which samples are currently being obtained in accordance with Specificaton 3.9, prepare and submit to the Commission and New York State Department of Environmental Conservation, within 30 days, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted. 3.11 Bases for Environmental Specifications
- a. Specification 3.8.1 is provided to ensure that the -dose at the exclusion area boundary from gaseous effluents from the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the l
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restricted area boundary to 6 500 mrem / year, to the total body. These release rate limits also restrict, at all times, the corresponding thyroid dos *e rate above background via the inhalation pathway to 61500 mrem / year.
- b. Specification 3.8.2 is provided to demonstrate compliance with 10 CFR 20.l(c) which requires releases of radioactive materials released to unrestricted areas to as low as reasonably achievable. The action statements provide the operating flexibility and at the same time implement the design objective of minimizing the release to unrestricted areas to as low as reasonably achievable. The specifications for noble gas releases are based on limiting the total body dose at the limiting populated area to less than 5 mrem / year. The specification for radiciodine is based on the assumption that 'the limiting dose pathway for these radioisotopes is via infant inhalation at the Laurel Ridge ,
Residential site, and limits the inf ant thyroid dose to less than 15 mrem / year.
- c. The radiological monitoring program required by specification 3.9 provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides, which lead to the nighest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected cn the basis of the effluent measurements and modeling of the environmental exposure pathways. This monitoring program may change based on operational experience and results of the land use census.
i .
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- j. d. Specification 3.10 is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the technical specification limit of dose, 4
'
via the most restrictive dose ' pathway and the monitoring l program, can be made if required by the results of this Census, i .i 4 1
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TABLE 3.9.1 (cont'd) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations Collection Frequency of Analysis
- 3. INGESTION Food Products Location to be At time of harvest. *I-131 analysis.
determined from One sample of broad Land Use Census. leaf vegetation. Water Indian Kill inlet. Monthly. Gross beta - Monthly. Indian Kill Monthly. Gross beta - Monthly. outlet. Warwick Brook. Monthly. Gross beta - Monthly. Sterling Lake Monthly. Gross beta - Monthly. outlet Ramapo River Monthly. Gross beta - Monthly. O The maximum values for the lower limit for I-131 are 7x10-2 pC1/m3 airborne concentration and 60 pCi/kg, wet weight leafy vegetables. . >
-= _. _ - ~_ - - -. . - . - _ _ _ . - _ . - - . - -. - ..
TABLE 3.9.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM , Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations Collection Frequency of Analysis
- 1. AIRBORNE
- a. Radioiodine 1 sample from 380 Continuous operation Radioiodine canister.
. and Particu- meters NE of of sampler with
- Analyze at least
' lates stack. sample collection as once per 7 days for required by dust I-131. I sample from loading but at least Laurel Ridge once per 7 days. Particulate sampler. Area. Analyze for gross beta radioactivity 24 hours following filter change. Perform gamma isotopic ' analysis on each sample when gross beta activity is 10 times the mean of control samples for any
.
medium. Perform gamma isotopic analysis on composite (by location) sample at least once per 92 days.
- 2. DIRECT RADIATION Same as #1 above. At least once per Gamma dose. At least once 31 days. per 31 days.
At least once per Ganna dose. At least once 92 days. (Read- per 92 days. out frequencies are determined by type of dosimeters selected. .
.
4.0 SURVEILLANCE REQUIREMENTS
- 4.1 General The requirements listed below generally prescribe tests or inspections to verify periodically that the performance of required syst ms is in accordance with specifications given above in Sections 2 and 3. In all instances where the specified frequency is annual, the interval between tests is not to exceed 14 months; when semiannual, the interval should not exceed 7 months; when monthly, the interval shall not exceed 6 weeks; when weekly, the interval shall not exceed 10 days; and when daily, the interval shall not exceed 3 days.
4.2 Safety Channel Calibration A channel calibration of each safety channel shall be performed annually (see Sect. 3.2.3). i t 4.3 Reactivity Surveillance
- 1. The reactivity worth of each control rod (inclucing the regulating rod) and the shut-down margin shall be determined whenever operation requires a reevaluation of core physics parameters, or annually, whichever occurs first. The rod worth will be determined using the reactivity-period or rod-drop methods.
- 2. The reactivity worth of an experiment shall be estimated, or measured at low power, before conducting the experiment.
.
_ - _ _ _ ._ . _ _ _ _
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- 3. ' Boron / Carbide rods shall be gauged quarterly and any dimensional changes reported promptly to the Commission.
Silver / Indium / Cadmium control rods shall be gauged annually, or, in the case of newly installed rods, at the end of the first six months. If any Ag/In/Cd rod should be found not be meet the acceptance criteria it shall be l removed from service. In addition, all other rods l manufactured of the same batch shall be inspected. 4.4 Control and Safety System Surveillance
- 1. The scram time shall be measured annually. If a control rod is removed from the core temporarily, or if a new rod is installed, its scram time shall be measured before reactor operation. If the bridge is moved, the scram time will be measured before subsequent reactor operation.
- 2. A channel test of each measuring channel in the reactor safety system shall be performed monthly or prior to each reactor operating period whichever occurs first unless the preceding shut-down period is 8 hours or less. A channel test before startup is, however, required on any channel receiving maintenance during the shut-down period.
- 3. A channel check of each measuring channel (except for the pool level) in the reactor safety system shall be performed daily when the reactor is in operation.
4.5 Radiation Monitoring System
- 1. The excursion, stack, and area monitors shall be calibrated l
annually. l
- 2. The excursion, stack, and area monitors shall receive a channel test monthly.
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- 3. The excursion, stack, and area monitors shall receive a channel test monthly.
4.6 Engineered Safety Features
4.6.1 Excursion Monitor: see above 4.5. 4.6.2 Emergency Generator
- 1. The ability of the emergency generator to start, to run normally, and to generate 440 VAC shall be checked weekly.
- 2. The generator shall be tested for its ability to accept, via the automatic transfer switch, the reactor electrical load once every six months. A commercial power outage and subsequent pickup of load by the emergency generator will a count as a successful load test.
4.6.3 Containment
- 1. The efficiency of the charcoal filters and of the absolute filters in the emergency exhaust system shall be measured annually and the flow rate verified.
- 2. The operability of the evacuation alarm and containment isolation system shall be tested, and negative pressure verified, semiannually. A utility power outage may be used
, to initiate such tests. 4.7 Reactor Fuel i 1. Upon receipt from the fuel vendor, all fuel elements shall be visually inspected and the accompanying quality control documents checked for compliance with specifications. l l _ .
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- 2. Each new fuel element will be inspected for damage and flow obstructions prior to insertion into the core.
4.8 Sealed Sources The antimony-beryllium sealed source shall be leak tested in accordance with the procedures described in the application for license amendment dated March 21, 1963, except that the frequency of leak testing will be in accordance with 10 CFR Part 34.25(b). The strontium-90 sealed source shall be tested for leakage and/or contamination semiannually. 4.9 Pool Water i
- 1. The pH and specific resistance of the pool water shall be determined each week.
- 2. An analysis of the pool water for radioactive material shall be done at monthly intervals. This analysis is to include Sb-124 as an indicator of Sb-Be neutron source integrity.
I
- 3. Activity of the pool water will be measured weekly.
4.10 Core Spray The core spray in the reactor operating position shall be tested for operability semiannually. ! I
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1 l 4.11 Flux Distribution In order to verify that power gradients among fuel elements do not cause peaking factors to exceed those used in the bases of Sect. 2.1, the radial neutron flux distribution will be determined whenever a significant core configuration change is made. 5.0 DESIGN FEATURES Those design features relevant to operation safety and to limits that have been previously specified are described below. These features shall not be changed without appropriate review. 5.1 Reactor Fuel , Fuel elements shall be of the general MTR/0RR type with thin plates containing uranium fuel enriched to about 93% U-235 and clad with aluminum. The fuel matrix may be fabricated by alloying high purity aluminum-uranium or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in fine powder form. Fuel matrix may also be fabricated from uranium oxide-aluminum (U 0 -Al) 38 using the powder metallurgy process. Elements shall conform to these nominal specifications: Overall Size: 3 in., x 3 in. x 34 in. Clad Thickness: 0.015 in. Plate Thickness: 0.050 in. Water Channel Width: 0.12 in. No. of Plates: standard element - 16 fueled plates (min.) window element - 16 fueled plates (min.) control element - 9 fueled plates (min.) partial element - 9 fueled plates (min.) Plate
Attachment:
swagged or pinned. Fuel ',antent (Total): 200 g U-235 nominal. Fuei Burnup: The fuel burnup shall not exceed .94x1021 l fission /cm3 except for U38 0 -Al which l shall not exceed 1.5x1021 fission /cm3,
5.2 Control and Safety Systems Design features of the components of this system (3.2.2, 3.2.3) that are important to safety are given below. 5.2.1 Power Level (Normal Channels) For this function three independent measuring channels are provided, two of which are required to be operable as a minimum. Each channel covers reliably the range from about 25% to 150% (of 5 MW). Each channel comprises an uncompensated - baron-coated ion chamber feeding an amplifier that controls electronic switches in the DC current that flows through each control rod electromagnet. Each channel controls and scrams all control rods. Each channel is fail-safe. The "f ast" scram (~ 5 ms) from each channel also produces, and is backed up by a
" slow" scram (~20 ms) through interruption of the AC supply to the rod electromagnet DC power supply. Each channel indicates power level on a panel meter allowing channel checks to be done during reactor operation. Each chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power.
5.2.2 Power Level (Intermediate) Channel For this function, a single channel is provided, covering reliably the range 10-3% to 300% (of 5 MW) with a logarithmic output incication on both a panel meter and a chart recorder. To cover the range under all core conditions, a gamma-compensa-ted boron-ion chamber is used to supply a logarithmic amplifier. The chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power. Rate of change of power information is also derived, in the form of a period, that can produce a fast
scram (and backup slow scram) in the same way as in Sect. 5.2.1. From this channel is also derived control and inhibit actions, viz. bypassing of count rate channel functions, bypassing of flow and flapper scrams, reversal and inhibit of centrol rods. To negate the effect of overcompensation in the ion chamber, which can occur under certain conditions even in an initially undercompensated chamber, provision is made to supply an adjustable small current to the channel amplifier (up to 1.5x10-10 A) so as to facilitate startup. 5.2.3 Count Rate Channel A fission chamber is used to supply pulses to a linear amplifier and logarithmic count rate circuitry. Pulse height discrimination selects pulse amplitudes that correspond to fission events and rejects those from alpha particles. Count rate on a logarithmic scale is displayed on a panel meter and a chart recorder. In addition, count rate period information is derived and similarly displayed. The channel covers a range of 4 1-10 cps, corresponding roughly to 0.25 mw - 2.5 W, but the upper limit can be increased many decades by repositioning the chamber. The motor-operated chamber drive is operated from the control room, the drive position being indicated on a meter. To prevent control-rod withdrawal when the neutron count rate information may not be reliably indicated, inhibits are provided on count rate and period, and when the fission chamber is being repositioned. All except the latter are bypassed at a power of
>50 W. A scaler is also provided for obtaining accurate values at low count rates if needed (e.g., approach to critical with new fuel or new core configuration).
.____
5.2.4 Neutron Source For obtaining the reliable neutron information necessary for startup from a cold shutdown condition, an antimony-beryllium neutron source is provided for insertion into the core as needed. This source, nominally 50 curies of Sb-124, is renewed by neutron activation in the core. Its presence in the core is not essential except after extended shutdowns. Integrity of the source is checked by periodic sampling of pool water (4.9). 5.3 Rod Control System 5.3.1 Control Rods Up to five control rods are provided for the control of core reactivity. These rods may be either boron-carbide or silver-indium-cadmium (see 4.3.1) . Individual integral worths vary from about 1-4% AK, depending on position and core configuration. The rods are coupled to drive shafts through electromagnets that allow release of the rods within 50 ms after receiving a scram signal. Position indicators on the control console show the extent of withdrawal for each rod and a digital readout can be switched to any one rod. To limit the rate of reactivity increase upon startup, the rod drive speeds are limited to 5 in/ min. and no more than two rods can be withdrawn simultaneously. Switches on the guide tubes attached to the control fuel elements are arranged to produce a scram if any guide tube is lifted. This guards against lifting of the attached fuel element. 5.3.2 Regulating Rod One regulating rod is provided to aid in fine control and
maintenance of constant reactor power for long periods. The rod is non-fueled, is limited to a total worth of 0.6% oK for safety reasons (3.1.4), and can be either manually or servo-controlled. The drive speed is 24 in./ min. Coarse and fine position readouts are provided. 5.4 Cooling System 5.4.1 Primary Cooling System Core cooling is effected by gravity flow of demineralized water from the reactor pool to an underground . holdup tank that provides an approximate 10-minute delay to allow N-16 activity to decay. The water is then pumped back to the pool through the ' primary side of a heat exchanger where heat is transferred to a secondary cooling system. The holdup tank is vented to the building exhaust duct. The driving force for the coolant is the fixed head between the pool overflow gutter and the water level in the boldup tank, the latter being fixed by the total volume of water in the system. Flow is adjusted to a desired amount with a valve in the core exit line. Core cooling is not immediately affected by pump f ailure as a flow will continue until the water levels equalize; neither will the pool be drained. To prevent leakage of water through the pool walls, a continuous steel shell is located within the concrete pour of the pools. All embedments penetrating the pools are welded to this shell. To eliminate corrosion of inaccessible piping, the embedded portion of the reactor primary cooling piping under the pools is stainless steel. To change over automatically to natural convection cooling at low flow rates, a weighted flapper valve seals tne core exit plenum. This valve, held closed by the core pressure drop, opens by gravity when the flow drops to j approximately 700 gpm. Leakage at the flapper valve seat, or in l the plenum, is monitored by a plenum leak detector that senses l l l 1
]
.
4 _42 plenum pressure increase and alarms in the control room. Primary flow is measured by taking the pressure drop across an orifice plate in the core exit line, indication being both in the pump room and on a recorder in the control room.
, Temperature sensors in the pool (above the core) and in the core exit line allow the core T to be measured. These are resistance thermometers having alternative recorder and digital readout in the control room. Float switches are provided to monitor pool level. Normal pool level (at the overflow gutters) provides 24 ft. of water over the top of the fuel.
5.4.2 Secondary Cooling System Reactor power transferred through the heat exchanger is dissipated to the atmosphere via a cooling tower. To minimize corrosion,the exchanger has stainless steel shell and tuces. To prevent water from entering the secondary system should a tube leak occur, the static pressure in the secondary is made higher than that of the primary through the relative elevations of the two systems. 5.4.3 Core Spray For backup in the event a hot core is exposed, two spray nozzles are located at the two alternative operating positions of the reactor. They are controlled from a manually-operated valve located outside the reactor building. , l l 5.5 Containment System
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5.5.1 Physical Features
- The containment structure consists of the reactor building, with a free air volume of about 7700 m . This building houses the 3
reactor, the primary cooling system including holdup tank, and the heat exchanger. Personnel access is via double airlock doors or sliding doors with inflatable seals. A 12-foot deep water-filled canal penetrates the building with containment provided by a 25-in. deep water-seal weir. A water-tight gate with inflatable seals can be used to shut the canal off from the reactor pools when needed. Ventilation access to the building is through pneumatically operated damper valves that can be used These dampers are fail-shut upon to seal the building. reduction of air pressure. 5.5.2 Emergency Sequence While negative pressure within the containment building is not a requisite for reactor operation, it is required in the event of a release for the controlled-release containment of airborne radioactive material. The emergency sequence is initiated either automatically by the excursion monitor (see 3.3) or manually by the console operator. The sequence is that all air supply ducts and the pool sweep dampers are closed immediately, followed later by the exhaust duct damper as soon as negative pressure (-l in, w.g.) is attained in the building (but not more than 7 sec. later). Closure of the pool sweep damper prevents activity released above the core from reaching the exhaust duct before it closes. Also closed immediately are the isolation valves in the vent line and the air purge to the holdup tank. This prevents activity in this tank from reaching the exhaust duct. Upon closure of the exhaust duct damper, the emergency exhaust f an starts and maintains a nominal negative pressure in the containment building. This f an exhausts building air at a
_ low rate (f 200 cfm) through absolute and charcoal filters before connecting into the normal exhaust duct. The latter discharges to the atmosphere through a stack at a high elevation. The entire evacuation sequence is fail-safe upon loss of utility electric power. It will operate with either utility or emergency generator power. 5.5.3 Exhaust Duct Monitor (" Stack Monitor") Air in the exhaust duct is continuously sampled for particulate iodine, and gaseous activities, each being read by separate detectors. The relative proportions of each type of activity i can thus be determined. The results are indicated on chart records, with repeaters in the control room. Detection or indication of a release is not dependent on all three detectors being operational, for any release will have associated with it all three types of activity or will affect each detector to some extent. Alarms, when setpoints are exceeded, are given at the monitor and repeated in the control room. 5.6 Fuel Storage 5.6.1 New Fuel Unirradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with the Security Plan. Elements are stored upright in metal racks in , which the separation between elements is a minimum of 2 inches. ' With such an arrangement, subcriticality is assured (Ref. 4). 5.6.2 Irradiated Fuel < Irradiated fuel is stored upright under water in the storage pool I within the reactor building in criticality-safe racks. Each rack accommodates 16 elements in wells with center-to-center spacing
of 6 inches. Ref. 5 states that an infinite number of elements so stored would be subcritical. Each well has a bottom hole to allow circulation of water for cooling. 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The organization for the management and operation of the reactor facility shall be as a minimum the structure shown in Fig. 2. Job titles shown are for illustration and may vary. Four levels of authority are provided, as follows: ' Level 1: Individual responsible for the facility license and site administration. Level 2: Individual responsible for the reactor facility operation and management. Level 3: Individual responsible for daily reactor operations. Level 4: Reactor operating staff. 4 The Nuclear Safeguards Committee shall report to Level 1. Radiation safety personnel shall report to Level 2 or higher. , J 6.l.2 Responsibility Responsibility for the safe operation of the reactor facility shall be within the chain of command shown in Figure 2. I Management levels, in addition to having responsibility for the policies and operation of the reactor facility shall be responsible - for safeguarding the public and f acility personnel from undue radiation exposures and for adhering to all requirements of the operating license and technical ! specifications. In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications. l
_ FIGURE 2 I l LEVEL 1
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r - - GENERAL MANAGER i 1 l 1 , I ! I i 1 RADIATION LEVEL 2 REVIEW SAFETY __ _ _ _ OPERATIONS AND STAFF MANAGER AUDIT ' COMMITTEE
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f LEVEL 3 REACTOR SUPERVISOR
1 LEVEL 4 OPERATING STAFF (LICENSEDOPERATORS)
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6.1.3 Staffing
- a. The minimum staffir.3 when the reactor is not secured shall be:
- 1. A licensed Reactor Operator in the control room.
- 2. A second licensed reactor operator present at the reactor facility. Unexp~cted absence for two hours is acceptable provided immediate action is taken to obtain a replacement.
- 3. A licensed Senior Reactor Operator shall be readily available on call.
- 4. A member of the operating shall shall be designated by Level 2 management as knowledgeable in radiation control.
- b. Events requiring tne presence of a Senior Operator:
- 1. All fuel-element or control-rod alterations within the reactor core region.
- 2. Relocations of any experiments witn reactivity worth greater tnan or equal to one dollar.
- 3. Recovery from unplanned or unscheduled shutdowns unless they are of. a type excluded by the Level 2 authority.
Such exclusions shall be posted in the control room or placed in the appropriate procedures. Furthermore, the presence of a senior operator at the facility shall not be required during recovery from unplanned or unscheduled shutdown or significant reduction in power in instances which result from:
- 1. Electrical power interruptions from internal or external f ailures exclusive of power supply failures of the reactor instrumentation, control and safety systems;
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J 2. False signals which, in the opinion of the Senior Operator, were properly verified to be false and to have resulted from monitoring, experimental, or control equipment, or from personnel inadvertence; and
- 3. Intentional shutdowns made by the Reactor Operator which are not related to the safety of the reactor; provided that prior to the initiation of' such recovery, the Senior Operator shall be notified of the shutdown of power reduction, and shall determine that the shutdown was caused by one of the enumerated occurrences, and shall determine that his presence at j the facility during recovery is not required.
6.1.4 Selection and Training of Personnel The selection, training, and requalification of personnel ' shall meet or exceed the requirements of ANS-15.4/N 380 and Appendix A of 10 CFR Part 55 and be in accordance with the requalification plan approved by the Commission. 6.1.5 Review and Audit The independent review and audit of reactor facility operations shall be performed by the Nuclear Safeguards Committee. 6.1.5.1 Composition and Qualifications The Nuclear Safeguards Committee shall be composed of a minimum of 5 members. The members shall collectively provide I a broad spectrum of expertise in the appropriate reactor technology. Members and alternates shall be appointed by and report to the Level 1 authority. They may include t
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individuals from within and/or outside the operating organization. Qualified and approved alternates may serve in the absence of regular members. 6.1.5.2 Charter and Rules i The committee shall function under the following operating rules:
- a. Meetings shall be held not less than semi-annually or more frequently as circumstances warrant, consistent with effective monitoring of facility activities.
- b. A quorum shall consist of not less than one-half the membership, where the operating staff does not constitute
. a majority.
- c. Subgroups may be appointed to review specific items.
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- d. Minutes shall oe kept, and shall be disseminated to members and to the Level I authority within one month j after the meeting.
- e. The Committee shall appoint one or more qualified individuals to perform the Audit Function.
l 6.1.5.3 Review Function The following items shall be reviewed by the review group or a subgroup thereof:
- a. Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question.
- b. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility
, equipment, or systems having safety significance.
- c. Tests-and experiments in accordance with section 6.3. I i
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- d. Proposed changes in technical specifications, license, or charter.
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- e. Violations of technical specifications, license, or
charter. Violations of internal procedures or instructions having safety significance.
- f. Operating abnormalities having safety significance, and audit reports.
- g. Reportable occurrences listed in section 6.5.3.
6.1.5.4 Audit Function . The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Where necessary, discussions with responsible personnel shall take place. In no case shall the individual or individuals conducting the audit be immediately
; responsible for the area being audited. Tne following items shall be audited:
- a. The conformance of facility operations to the technical specifications and applicable license or charter conditions, at least once per calendar year (interval not to exceed 18 months).
- b. The retraining and requalification for the operating -
staff, at least once every other calendar year (interval
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not to exceed 30 months).
- c. The results of actions taken to correct deficiencies occurring in reactor facility equipment, systems, structures, or methods of operations that affect reactor safety, at least once per calendar year (interval not to exceed 18 months),
- d. The reactor facility Security Plan and implementing procedures at least once every other calendar year (interval not to exceed 30 months).
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Deficiencies uncovered that affect reactor safety shall immediately be reported to the Level 2 authority. A written report of the findings of the audit shall be submitted to the Level 1 authority and the Nuclear Safeguards Committee members within 90 days after the audit has been completed. 6.2 Procedures There shall be written procedures for, and prior to, initiating any of the activities listed in this section. The procedures shall be reviewed by the Nuclear Safeguards Committee and approved by Level 2 or designated alternates, and such reviews and approvals shall be documented. Several of the following activities may be included in a single manual or set of procedures or divided among various manuals or procedures.
- a. Startup, operation, and shutdown of the reactor.
- b. Fuel loading, unloading, and movement within the reactor,
- c. Routine maintenance of major components of systems that could have an effect on reactor safety,
- d. Surveillance tests and calibrations required by the technical specifications or those that may have an effect on reactor safety.
- e. Personnel radiation protection, consistent with applicable regulations.
- f. Administrative controls for operat ions ar.d maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity.
- g. Implementation of the Security Plan and the Emergency Plan. l Substantive changes to the above procedures shall be made only after documented review by the Nuclear Safeguards
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, Committee and approval by Level 2 or designated alternates. Minor modifications to the original procedures which do not change their original int'ent may be made by the Level 3 authority (Reactor Supervisor) and must be approved by Level 2 or designated alternates within 14 days. Temporary changes to the procedures that do not affect reactor safety may be made by a Senior Reactor Operator and are valid for a period of one month. Such temporary changes shall be documented and reported to Level 2 or designated alternate. 6.3 Experiment Review and Approval
- a. All new experiments or classes of experiments that could affect reactivity or result in release of radioactive materials shall be reviewed by the Nuclear Safeguards Committee. This review shall assure that compliance with the requirements of the license, technical soecifications, and applicable regulations has been satisfied, and shall be documented.
Prior to review, an experiment plan or proposal shall be prepared describing the experiment including any safety ' considerations. Review comments of the Nuclear Safeguards Committee setting forth any conditions and/or limitations shall be documented in Committee minutes and submitted to Level 2.
- d. All new experiments or classes of experiments shall be approved in writing by Level 2 or designated alternates prior to their initiation.
- e. Substantive changes to approved experiments shall be mde only af ter review by the Nuclear Safeguards Committee and written approval by Level 2 or designated alternates.
Minor changes that do not significantly alter the experiment may be approved by the Level 3 authority (Reactor Supervisor).
- f. Approved experiments shall be carried out in accordance with established approved procedures.
6.4 Required Actions - 6.4.1 Action to be Taken in Case of Safety Limit Violation
- a. The reactor shall be shutdown, and reactor operations shall not be resumed until authorized by the. Commission.
- b. The safety limit violation shall promptly be reported to
' the Level 1 authority or designated alternates.
- c. The safety limit violation shall be reported to the Commission in accordance with section 6.5.3.
- d. A safety limit violation report shall be prepared. The report shall describe the following:
- 1. Applicable circumstances leading to the violation.
- 2. Effect of the violation upon reactor facility l
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components, systems, or structures.
- 3. Corrective action to be taken to prevent recurrence.
The report shall be reviewed by the Nuclear Safeguards Committee. A follow-up report describing extant activities shall be submitted to the Commission when authorization is sought to resume operation of the reactor. 6.4.2 Action to be taken in the event of an occurrence as defined in section 6.5.3, a-1, 3: l
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- a. Corrective action shall be taken to return conditions to normal; otherwise, the reactor shall be shut down and reactor operation shall not be resumed unless authorized 1 by the Level 2 authority or designated alternates.
- b. All such occurrences shall be promptly reported to the Level 2 authority or designated alternates. l
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- c. All such occurrences where applicable shall be reported to the Comission in accordance with section 6.5.3.
- d. All such occurrences, including action taken to prevent or reduce the probability of a recurrence, shall be reviewed by the Nuclear Safeguards Comittee.
6.5 Reports i In addition to the requirements of applicable regulations, reports shall be made to the Commission as follows: 6.5.1 Startup Reports Three months after completion of requisite startup and power-escalation testing of the reactor, or nine months af ter criticality, a written report shall be submitted to the Comission. The report shall include a sumary of the following:
- a. Description of measured values of operating conditions or characteristics obtained and comparison of these values with design predictions or specifications.
- b. Descriptions of major corrective actions taken to obtain satisfactory operation.
- c. Re-evaluation of safety analyses where measured values indicate substantial variance from those values used in the Safety Analysis Report.
6.5.2 Operating Reports Routine annual reports, covering the activities of the reactor f acility during the previous calendar year, shall be submitted to the appropriate NRC Regional Office with a copy to the Director of Inspection & Enforcement within 3 months following the end of each prescribed year. Each annual operating report shall include the following information:
!
- a. A narrative summary of reactor operating experience, including the energy produced by the reactor.
- b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence, but excluding those of the types listed in Section 6.1.3. b(3) l above.
- c. Tabulation of major preventive and corrective maintenance operations having safety significance.
- d. Tabulation of major changes in the reactor facility procedures, and new tests and/or experiments significantly different from those performed previously and which are not described in the Safety Analysis Report, including conclusions that no unreviewed safety questions were involved.
- e. A summary of the nature and amount of radioactive effluents from the reactor facility released or discharged to the environs. The summary shall include, where practicable, an estimate of individual radionuclides present in the effluent if the estimated average release after dilution or diffusion is greater than 25% of the ccncentration allowed or recommended,
- f. A summary of exposures received by f acility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
- g. A summary of the calculated doses to a critical individual in the unrestricted area due to the airborne releases of
' noble gases and radioiodines. , 1
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6.5.3 Special Reports (Reportable Occurrences) 4
- a. There shall be a report not later than the following working day by telephone and confirmed by telegraph or similar conveyance to the Commission to be followed by a written report within 14 days of any of the followong:
- 1. Release of radioactivity from the reactor above allowed limits, as provided by section 3.8.1 of this specification.
- 2. Violation of Safety Limits
- 3. Any of the following:
- a. Operation with actual safety-system settings less conservative than the limiting safety-system settings specified in the Technical Specifications.
- b. Operation in violation of Limiting Conditions for Operation established in the Technical Specifications.
- c. 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 condition is discovered during 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 is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)
- d. An unanticipated or uncontrolled change in reactivity greater than or equal to 1% AK/K.
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- e. Abnormal and significant degradation in reactor fuel, and/or cladding, coolant boundary, or containment boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel
;
and/or environment.
- f. An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused an unsafe condition with regard to reactor operations.
- b. A written report within 30 days to the Commission of:
- 1. Permanent changes in the facility organization structure.
- 2. Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
- 3. Exceeding the liquid effluent limit as specified in section 3.8.3d above.
- c. A report within 30 days to the Commission and to New York State as specified in the ACTION items of sections 3.8.2b, 3.8.2c, and 3.10 above.
6.6 Records Records of the following activities shall be maintained and retained for the periods specified below. The records may be in the form of logs, data sheets, or other suitable forms.
, The required information may be contained in single, or
! multiple records, or a combination thereof. Recorder charts showing operating parameters of the reactor (i.e. power level, flow, temperature, etc.) for unscheduled shutdown and significant unplanned transients shall be maintained for a minimum period of two years. l 6.6.1 Records to be retained for a period of at least five years or , for the life of the component involved, whichever is smaller. ,
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- a. Normal reactor facility operations (including scheduled and unscheduled shutdowns). Note: Supporting docume.its such as checklists, log sheets, etc. shall be maintained for a period of at least two years.
- b. Principal maintenance operations.
- c. Reportable occurrences.
- d. Surveillance activities required by the Technical Specifications.
- e. Reactor facility radiation and contamination surveys where i required by applicable regulations.
- f. Experiments performed with the reactor.
- g. Special Nuclear Materials (SNM) inventories, receipts, and shipments.
- h. Approved changes in operating procedures.
- i. Records of meeting and audit reports of the Nuclear Safeguards Committee.
- j. Sealed Source leak test results.
6.6.2 Records to be Retained for at least One Requalification Cycle or for the Length of Employment of the Individual whichever is Smaller:
- a. Retraining and requalification of licensed operations personnel. However, records of the most recent complete cycle shall be maintained at all times the individual is employed.
6.6.3 Records to be Retained for the Lifetime of the Reactor Facility: (Note: Annual reports may be used where applicable as records in this section.)
- a. Gaseous and liquid radioactive effluents released to the environs,
- b. Off-site environmental-monitoring surveys required by the Technical Specifications. l
- c. Radiation exposure of all personnel monitored.
- d. Updated drawings of the reactor facility.
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7.0 REFERENCES
- 1. Safety Analysis Report, Union Carbide Research Reactor (May 1980), Appendix 2
- 2. Final Hazards Summary Report, UCNC Research Reactor (May 1960)
- 3. Reactcr Power Excursion Tests in the SPERT IV Facility.
100-17000 (August 1964)
- 4. Critical Experiments with SPERT-D Fuel Elements, ORNL-TM-1207 (July 14, 1965), by E.B. Johnson and P.K. Reedy, Jr.
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