ML13080A364
| ML13080A364 | |
| Person / Time | |
|---|---|
| Site: | Rhode Island Atomic Energy Commission |
| Issue date: | 03/15/2013 |
| From: | State of RI, Atomic Energy Comm |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML13080A364 (94) | |
Text
Proposed Technical Specifications 130314 1.0 Definitions ................................................................................. 2 2.0 Safety Limits and Limiting Safety System Settings .................................. 10 2.1 Safety Limits 2.2 Limiting Safety System Settings 3.0 Limiting Conditions for Operation .................................................... 16 3.1 Core Parameters 3.2 Reactor Control and Safety System 3.3 Coolant System 3.4 Confinement System 3.5 Ventilation System 3.6 Emergency Power System 3.7 Radiation Monitoring System and Effluents 3.8 Experiments 3.9 Reactor Core Components 4.0 Surveillance Requirem ents ................................................................ 54 4.1 Core Parameter Surveillance 4.2 Reactor Control and Safety System Surveillance 4.3 Coolant System Surveillance 4.4 Confinement System Surveillance 4.5 Ventilation System Surveillance 4.6 Emergency Power System Surveillance 4.7 Radiation Monitoring System and Effluent Surveillance 4.8 Experiment Surveillance 4.9 Facility Specific Surveillance 5.0 Design Features ............................................................................ 76 5.1 Site and Facility Description 5.2 Reactor Coolant System Description 5.3 Reactor Core and Fuel Description 5.4 Fissionable Material Storage Description 6.0 Administrative Controls ................................................................ 80 6.1 Organization 6.2 Review and Audit 6.3 Radiation Safety 6.4 Procedures 6.5 Experiment Review and Approval 6.6 Required Actions 6.7 Reports 6.8 Records 1
-p 1.0 Definitions 1.1 Channel A channel is the combination of sensor, line, amplifier, and output devices that are connected for the purpose of measuring the value of a parameter.
1.2 Channel Calibration A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter that the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a channel test.
1.3 Channel Check A channel check is a qualitative verification of acceptable performance by observation of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same parameter.
1.4 Channel Test A channel test is the introduction of a signal into the channel for verification that it is operable.
1.5 Confinement Confinement is an enclosure of the overall facility that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways.
1.6 Control Rod A control rod is a device fabricated from neutron absorbing material, that is used to establish neutron flux changes and to compensate for routine reactivity losses.
1.7 Core Configuration The core configuration includes the number, type, or arrangement of fuel elements, reflector elements, and control rods occupying the core grid.
2
1.8 Excess Reactivity Excess reactivity is that amount of reactivity that would exist if all of the control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical.
1.9 Experiment An experiment is any operation that is designed to investigate non-routine reactor characteristics, or any material or device not associated with the core configuration or the reactor safety systems that is intended for irradiation within the pool or an experimental facility. Hardware that is rigidly secured to a core or shield structure so as to be part of its design to carry out experiments is not normally considered to be an experiment.
1.10 Experimental Facility An experimental facility is any structure or device which is intended to guide, orient, position, manipulate, or otherwise facilitate a multiplicity of experiments of similar character.
1.11 Explosive Material Explosive material is any solid or liquid which is either:
A. Categorized as a severe, dangerous, or very dangerous explosion hazard in Sax's Dangerous Properties Of Industrial Materials by Richard J. Lewis, Sr., 11 th Ed. (2004), or B. Is given an identification of Reactivity (Stability) Index of 2, 3, or 4 by the National Fire Protection Association (NFPA) in its publication NFPA 704: Standard System for the Identification of the Hazards of Materials for Emergency Response, 2007 Edition.
1.12 Fixed Experiment A fixed experiment is any experiment, experimental apparatus, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than other forces to which the experiment might be subjected that are normal to the operating environment of the experiment, or that can arise as a result of a credible malfunction.
3
1.13 Limiting Conditions for Operation (LCO)
The limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the reactor.
1.14 Limiting Safety System Setting (LSSS)
Limiting Safety System Settings are settings for automatic protective devices related to those variables having significant safety functions, and chosen so that automatic protective action will correct an abnormal situation before a safety limit is exceeded.
1.15 May The word "may" is used to denote permission, neither a requirement nor a recommendation.
1.16 Mode of Operation Mode of operation refers to the type of core cooling that is employed while the reactor is operating. The two modes of operation are forced convection cooling mode, and natural convection cooling mode.
1.17 Moveable Experiment A moveable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
1.18 Operable Operable means that a component or system is capable of performing its intended function.
1.19 Operating Operating means that a component or system is performing its intended function.
1.20 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.
4
a 1.21 Reactivity Worth of an Experiment The reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of:
A. Insertion or removal from the core, B. Intended or anticipated changes in position, or C. Credible malfunctions that alter experiment position or configuration.
1.22 Reactor Operating The reactor is operating whenever it is not secured or shut down.
1.23 Reactor Operator A reactor operator is an individual who is licensed under 10 CFR Part 55 to manipulate the controls of the RINSC reactor.
1.24 Reactor Operator Trainee A reactor operator trainee is an individual who is authorized to manipulate the controls of the RINSC reactor under the direct supervision of a reactor operator.
1.25 Reactor Safety Systems Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
1.26 Reactor Secured The reactor is secured when either:
A. There is insufficient moderator available in the reactor to attain criticality, B. There is insufficient fissile material present in the reactor to attain criticality, or C. The following conditions exist:
- 1. The reactor is shutdown,
- 2. The master switch is in the off position and the key is removed from the lock, 5
- 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 experiments are being moved or serviced that have a reactivity worth of greater than 0.6% dK/K when moved.
1.27 Reactor Shutdown The reactor is shut down if it is subcritical by at least 0.75% dK/K in the reference core condition with the reactivity of all installed experiments included.
1.28 Readily Available on Call Readily available on call shall mean that the individual is aware that they are on call, can be contacted within ten minutes, and is within a 30 minute driving time from the reactor building.
1.29 Reference Core Condition The condition of the core when it is at ambient temperature and the reactivity of xenon is less than 0.2 % dK/K.
1.30 Regulating Rod The regulating rod is a control rod of low reactivity worth used primarily to maintain an intended power level. It is not required to have a scram capability, and may be controlled manually or by servo controller.
1.31 Reportable Occurrence A reportable occurrence is any of the following:
- 1. A violation of a safety limit,
- 2. An uncontrolled or unplanned release of radioactive material which results in concentrations of radioactive materials inside or outside the restricted area in excess of the limits specified in Appendix B of 10CFR20,
- 3. Operation with a safety system setting less conservative than the limiting setting established in the Technical Specifications,
- 4. Operation in violation of a limiting condition for operation established in the Technical Specifications, 6
&
- 5. A reactor safety system component malfunction or other component or system malfunction which could, or threaten to, render the safety system incapable of performing its intended safety functions,
- 6. An uncontrolled or unanticipated change in reactivity in excess of 0.75 %AK/K,
- 7. Abnormal and significant degradation of the fuel cladding, of
- 8. Abnormal and significant degradation of the primary coolant boundary, or
.99. An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the facility.
1.32 Safety Channel A safety channel is a channel in the reactor safety system.
1.33 Safety Limits Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of the principal barriers which guard against the uncontrolled release of radioactivity. The principal barrier is the fuel element cladding.
1.34 Scram Time Scram time is the elapsed time between the initiation of a scram signal and the time when the blades are fully inserted in the core.
1.35 Senior Reactor Operator A senior reactor operator is an individual who is licensed under 10 CFR Part 55 to manipulate the controls of the RINSC reactor and to direct the licensed activities of reactor operators.
1.36 Shall The word "shall" is used to denote a requirement.
7
1.37 Shim Safety Rod A shim safety rod is a control rod of high reactivity worth used primarily to make course adjustments to power level, and to provide a means for very fast reactor shutdown by having a scram capability.
1.38 Should The word "should" is used to denote a recommendation.
1.39 Shutdown Margin Shutdown Margin shall mean the minimum amount of negative reactivity inserted into the core when the most reactive control blade and the regulating rod are fully withdrawn, and the remaining control blades are fully inserted into the core.
1.40 Subcritical The reactor is subcritical if:
A. There is insufficient fissile material or moderator present in the core to attain criticality under optimum available conditions of moderation and reflection, or B. The control rods are providing sufficient negative reactivity in the core to prevent criticality.
1.41 Surveillance Activities Surveillance activities are activities that are performed on a periodic basis for the purpose of verifying the integrity and operability of facility infrastructure and equipment, and ensuring the safe operation of the reactor.
1.42 Surveillance Intervals Surveillance intervals are the periods in which each surveillance activity is to be performed. Each interval is defined in a way that provides operational flexibility. The surveillance intervals are:
A. Daily shall mean during the calendar day on days when the facility is open, B. Weekly shall mean an interval not to exceed 10 calendar days, C. Monthly shall mean an interval not to exceed 6 weeks, 8
D. Quarterly shall mean an interval not to exceed 4 months, E. Semiannually shall mean an interval not to exceed 7 1/2 months, F. Annually shall mean an interval not to exceed 15 months, and G. Biennially shall mean an interval not to exceed 2 1/2 years.
1.43 True Value The true value is the actual value of a parameter.
1.44 Unscheduled Shutdown An unscheduled shutdown is defined as any unplanned shutdown of the reactor that is not associated with testing or check out operations, which is caused by:
A. Actuation of the reactor safety system, B. Operator error, C. Equipment malfunction, or D. Manual shutdown in response to conditions that could adversely affect safe operation.
1.45 Unusual Event An unusual event is defined as any of the following:
A. Permanent changes in the facility organization involving RIAEC membership, the facility Director, or either of the facility Assistant Directors.
B. Significant changes in the transient or accident analyses as described in the Safety Analysis Report.
9
2.0 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits Applicability:
This specification applies to fuel that is loaded in the core.
Objective:
The objective of this specification is to ensure that the integrity of the fuel cladding is not damaged due to overheating.
Specifications:
The true value of the reactor fuel cladding shall be less than or equal to 530 C.
Bases:
NUREG 1313 shows that the integrity of the fuel cladding will not be damaged due to overheating provided that the cladding temperature does not exceed 530 C.
2.2 Limiting Safety System Settings 2.2.1 Limiting Safety System Settings for Natural Convection Mode of Operation Applicability:
These specifications apply to the safety channels that monitor variables that directly impact fuel cladding temperature during natural convection mode operation of the reactor.
Objective:
The objective of these specifications is to ensure that the safety limit for the reactor cannot be exceeded during natural convection mode operation.
10
Specifications:
2.2.1.1 The limiting safety system setting for reactor thermal power shall be 115 kW.
2.2.1.2 The limiting safety system setting for the height of coolant above the top of the fuel meat shall be 23 ft 9.6 inches.
2.2.1.3 The limiting safety system setting for the bulk pool temperature shall be 125 OF.
Bases:
This combination of specifications was set to prevent the cladding temperature from approaching the 530 °C value at which damage to the fuel cladding could occur, even under transient conditions.
The thermal-hydraulic analysis for steady state power operation under natural convection cooling conditions shows that the fuel cladding temperature will remain significantly below the threshold for cladding damage during steady state operation of the reactor if the following combination of limits are in place:
- The steady state power level is less than 200 kW,
- The coolant height above the fuel meat is at least 23 ft 6.5 in, and
- The bulk pool temperature is no greater than 130 OF.
The transient analysis for natural convection cooling was performed for the most conservative case in which all of the safety channels are at their respective limiting trip value when the transient is terminated. The analysis shows that the peak fuel cladding temperature will be approximately 67.5 oC during a transient in which the following combination of limits are in place:
- The initial power level is no greater than 100 kW,
- The coolant height above the fuel meat is at least 23 ft 9.1in,
- The bulk pool temperature is no greater than 128 OF, and
- The transient is terminated by an over power trip at 125 kW.
11
a In both, the steady state and transient analyses, the predicted peak cladding temperature is significantly below the damage threshold temperature of 530 °C. The safety margins are:
- Margin for the transient bounded by the limiting conditions is 530 °C - 67.5 Oc = 462.5 oc.
Measurement uncertainty was based on the nominal operating values of 100 kW and 108 OF for the power and pool temperature respectively, and has been determined to be:
- Power Level +/- 10 kW
- Coolant Height 0.5 in
- Temperature 3 OF Consequently, the bases for these specifications are:
Specification 2.2.1.1 sets the limiting safety system setting for reactor thermal power to be 115 kW.
The analyses show that cladding damage will not occur under any condition if initial power is no greater than 200 kW. Taking into consideration a 10 kW measurement error, if the LSSS is 115 kW, then the Limiting Trip Value could be as high as 125 kW, which still leaves a safety margin of 75 kW.
Specification 2.2.1.2 sets the limiting safety system setting for the height of coolant above the top of the fuel meat to be 23 ft 9.6 in. The analyses show that cladding damage will not occur under any condition if the height is no less than 23 ft 6.5 in. Taking into consideration a 0.5 in measurement error, if the LSSS is 23 ft 9.6 in, then the Limiting Trip Value could be as low as 23 ft 9.1 in, which still leaves a safety margin of 2.6 in.
Specification 2.2.1.3 sets the limiting safety system setting for the bulk pool temperature to be 125 OF.
The analyses show that cladding damage will not occur under any condition if the pool temperature is no greater than 130 OF. Taking into consideration a 12
3 OF in measurement error, if the LSSS is 125 OF, then the Limiting Trip Value could be as high as 128 OF, which still leaves a safety margin of 2 OF.
2.2.2 Limiting Safety System Settings for Forced Convection Mode of Operation Applicability:
These specifications apply to the safety channels that monitor variables that directly impact fuel cladding temperature during forced convection mode operation of the reactor.
Objective:
The objective of these specifications is to ensure that the safety limit for the reactor cannot be exceeded during forced convection mode operation.
Specifications:
2.2.2.1 The limiting safety system setting for reactor thermal power shall be 2.1 MW.
2.2.2.2 The limiting safety system setting for the height of coolant above the top of the fuel meat shall be 23 ft 9.6 inches.
2.2.2.3 The limiting safety system setting for the primary coolant outlet temperature shall be 120 OF.
2.2.2.4 The limiting safety system setting for the primary coolant flow rate shall be 1800 gpm.
Bases:
This combination of specifications was set to prevent the cladding temperature from approaching the 530 oC value at which damage to the fuel cladding could occur, even under transient conditions.
The thermal-hydraulic analysis for steady state power operation under forced convection cooling conditions shows that the fuel cladding temperature will remain significantly below the threshold for cladding damage during operation of the reactor if the following combination of limits are in place:
13
- The steady state power level is less than 2.4 MW,
- The coolant height above the fuel meat is at least 23 ft 6.5 in,
- The primary coolant outlet temperature is no greater than 125 OF, and
- The coolant flow rate through the core is at least 1580 gpm.
The transient analysis for forced convection cooling was performed for the most conservative case in which all of the safety channels are at their respective limiting trip value when the transient is terminated. The analysis shows that the peak fuel cladding temperature will be no greater than 85 oC during a transient in which the following combination of limits are in place:
- The initial power level is no greater than 2.2 MW,
- The coolant height above the fuel meat is at least 23 ft 9.1in,
- The primary coolant inlet temperature is no greater than 123 OF,
- The coolant flow rate through the core is a least 1740 gpm, and
- The transient is terminated by an over power trip at 2.3 MW.
In both, the steady state and transient analyses, the predicted peak cladding temperature is significantly below the damage threshold temperature of 530 °C. The safety margins are:
- Margin for the transient bounded by the limiting conditions is 530 °C - 85 °C = 445 OC.
Measurement uncertainty was based on the nominal operating values of 2 MW, 1950 gpm, and 90 OF to 115 OF for the power, flow and outlet temperature respectively, and has been determined to be:
- Power Level +/- 0.2 MW
- Coolant Height 0.5 in
- Temperature 3 OF
- Flow Rate +/- 60 gpm 14
Consequently, the bases for these specifications are:
Specification 2.2.2.1 sets the limiting safety system setting for reactor thermal power to be 2.1 MW.
The analyses show that cladding damage will not occur under any condition if initial power is no greater than 2.4 MW. Taking into consideration a 0.2 MW measurement error, if the LSSS is 2.1 MW, then the Limiting Trip Value could be as high as 2.3 MW, which still leaves a safety margin of 0.1 MW.
Specification 2.2.2.2 sets the limiting safety system setting for the height of coolant above the top of the fuel meat to be 23 ft 9.6 inches. The analyses show that cladding damage will not occur under any condition if the height is no less than 23 ft 6.5 in.
Taking into consideration a 0.5 in measurement error, if the LSSS is 23 ft 9.6 in, then the Limiting Trip Value could be as low as 23 ft 9.1 in, which still leaves a safety margin of 2.6 in.
Specification 2.2.2.3 sets the limiting safety system setting for the primary coolant outlet temperature to be 120 OF. The analyses show that cladding damage will not occur under any condition if the primary coolant outlet temperature is no greater than 125 OF. Taking into consideration a 3 OF in measurement error, if the LSSS is 120 OF, then the Limiting Trip Value could be as high as 123 OF, which still leaves a safety margin of 2 OF.
Specification 2.2.2.4 sets the limiting safety system setting for the primary coolant flow rate to be 1800 gpm. The analyses show that cladding damage will not occur under any condition if the primary coolant flow rate is at least 1580 gpm. Taking into consideration a 60 gpm in measurement error, if the LSSS is 1800 gpm, then the Limiting Trip Value could be as low as 1740 gpm, which still leaves a safety margin of 160 gpm.
15
3.0 Limiting Conditions for Operation 3.1 Core Parameters 3.1.1 Reactivity Limits Applicability:
This specification applies to all core configurations, including configurations that have experiments installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to make certain that core reactivity parameters will not exceed the limits used in the safety analysis to ensure that a reactor transient will not result in damage to the fuel.
Specification:
3.1.1.1 Core 3.1.1.1.1 The core shutdown margin shall be at least 1.0 % dK/K.
3.1.1.1.2 The core excess reactivity shall not exceed 4.7 % dK/K.
3.1.1.1.3 The temperature coefficient shall be negative.
3.1.1.1.4 The reactor shall be subcritical by at least 3.0 %dKiK during fuel loading changes.
3.1.1.2 Control Rods 3.1.1.2.1 The reactivity worth of the regulating rod shall not exceed 0.6 % dK/K.
16
3.1.1.3 Experiments 3.1.1.3.1 The total absolute reactivity worth of experiments shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies:
Total Moveable and Fixed 0.6 %dK/K Total Moveable 0.08 %dKIK 3.1.1.3.2 The maximum reactivity worth of any individual experiment shall not exceed the following limits, except when the operation of the reactor is for the purpose of measuring experiment reactivity worth by doing criticality studies:
Fixed 0.6 % dK.K Moveable 0.08 % dK/K Basis:
Specification 3.1.1.1.1 provides a limit for the minimum shutdown reactivity margin that must be available for all core configurations. The shutdown margin is necessary to ensure that the reactor can be made subcritical from any operating condition, and to ensure that it will remain subcritical after cool down and xenon decay, even if the most reactive control rod failed in the fully withdrawn position. No credit is taken for the negative reactivity worth of the regulating rod because it would not be available as part of the negative reactivity insertion in the event of a scram. An allowance is made for measuring the reactivity worth of experiments. The reactor can be made critical with experiments of unknown reactivity, so that the criticality data can be used to determine whether or not the reactivity worth of an experiment is within the limits prescribed by this specification.
Specification 3.1.1.1.2 provides a maximum limit for excess reactivity available for all core configurations.
Excess reactivity is necessary to overcome the negative reactivity effects of coolant temperature increase, coolant void increase, fuel temperature increase, and xenon build-17
up that occur during sustained operations. Excess reactivity is also required to be available in order to overcome any negative reactivity effects of experiments that are installed in the core. An allowance is made for measuring the reactivity worth of experiments. The reactor can be made critical with experiments of unknown reactivity, so that the criticality data can be used to determine whether or not the reactivity worth of an experiment is within the limits prescribed by this specification.
Specification 3.1.1.1.3 requires that the temperature coefficient be negative. This requirement ensures that a temperature rise due to a reactor transient will not cause a further increase in reactivity.
Specification 3.1.1.1.4 provides a limit for the minimum core shutdown reactivity during fuel loading changes. This limit takes advantage of the negative reactivity that can be added to the core above and beyond the shutdown margin by the insertion of the highest reactivity worth, and regulating control rods. This limit assures that the core will remain subcritical during these operations.
Specification 3.1.1.2.1 provides a limit for the reactivity worth of the regulating rod. The reactivity limit is set to a value less than the delayed neutron fraction so that a failure of the automatic servo system could not result in a prompt critical condition.
Specification 3.1.1.3.1 provides total reactivity limits for all experiments installed in the reactor, the reactor pool, or inside the reactor experimental facilities. The limit on total experiment worth is set to a value less than the delayed neutron fraction so that an experiment failure could not result in a prompt critical condition. The limit on total moveable experiment worth is set to a value that will not produce a stable period of less than 30 seconds, so that the reactivity insertion can be easily compensated for by the action of the control and safety systems. As part of the Safety Analysis, Argonne National Laboratory modeled a reactivity insertion of + 0.08 % dK/K over a 0.1 second interval, and determined that this reactivity insertion resulted in a stable period of approximately 75 seconds.
18
Specification 3.1.1.3.2 provides total reactivity limits for any individual experiment installed in the reactor, the reactor pool, or inside the reactor experimental facilities.
The reactivity limits for both, fixed and moveable experiments are the same as the limits for total fixed and moveable experiments. Consequently, the safety analysis done for Specification 3.1.1.3.1 applies to this specification as well.
3.1.2 Core Configuration Limits Applicability:
This specification applies to core configurations during operations above 0.1 MW when the reactor is in the forced convection cooling mode.
Objective:
The objective of this specification is to ensure that there is sufficient coolant to remove heat from the fuel elements when the reactor is in operation at power levels greater than 0.1 MW.
Specifications:
3.1.2.1 All core grid positions shall contain fuel elements, baskets, reflector elements, or experimental facilities during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.
3.1.2.2 The pool gate shall be in its storage location during operations at power levels in excess of 0.1 MW in the forced convection cooling mode.
Bases:
Specification 3.1.2.1 requires that all of the core grid spaces be filled when the reactor is operated at higher power levels that require forced convection cooling. This requirement prevents the degradation of coolant flow through the fuel channels due to flow bypassing the actively fueled region of the core through unoccupied grid plate positions.
19
Specification 3.1.2.2 requires that the pool gate that is used for separating the sections of the pool, be in its storage location when the reactor is in operation at higher power levels that require forced convection cooling. This requirement ensures that there will be a sufficient heat sink for high power operations, and ensures that the full volume of the pool water will be available in the event of a loss of coolant accident.
3.2 Reactor Control and Safety System Applicability:
This specification applies to the reactor safety system and safety related instrumentation required for critical operation of the reactor.
Objective:
The objective of this specification is to define the minimum set of safety system and safety related channels that must be operable in order for the reactor to be made critical.
Specification:
3.2.1 The reactor shall not be made critical unless:
3.2.1.1 All shim safety blades are capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated.
3.2.1.2 The reactivity insertion rates of individual shim safety and regulating rods does not exceed 0.02%
dK/K per second.
3.2.1.3 The following reactor safety and safety related instrumentation is operable and capable of performing its intended function:
20
Protection Power Channels Function Set Point Level Required Over Power All 2 Scram by Power Level Less than or 105% of Equal to Licensed Power Low Pool Level All I Scram by Pool Level Less than or 23 ft 9.6 Drop Equal to in Primary Coolant Inlet >100kW 1 Alarm by Inlet Temp Less than or 111 F Temperature Equal to Primary Coolant Outlet >100kW 1 Alarm by Outlet Temp Less than or 117 F Temperature Equal to
>100kW 1 Scram by Outlet Temp Less than or 120 F Equal to Pool Temperature <=100kW 1 Scram by Pool Temp Less than or 125 F Equal to Primary Coolant Flow Primary Flow Less than or 1800 Rate >100kW 1 Scram by Rate Equal to gpm Rate of Change of Less than or 4 Power All 1 Scram by Period Equal to seconds Seismic Disturbance All 1 Scram if Seismic Disturbance Detected Bridge Low Power Position > 100kW 1 Scram if Bridge Not Seated at HP End Bridge Movement All 1 Scram if Bridge Movement Detected Coolant Gates Open >100kW 1 Scram if Inlet Gate Open
>100kW 1 Scram if Outlet Gate Open Detector HV Less than or Detector HV Failure All 1 Scram if Decrease Equal to 50 V Detector HV Less than or All 1 Scram if Decrease Equal to 50 V Detector HV Less than or All I Scram if Decrease Equal to 50 V No Flow Thermal Column > 100kW 1 Scram by No Flow Detected Manual Scram All 1 Scram by Button Depressed All 1 Scram by Button Depressed Servo Control No Automatic Interlock All I Servo if Regulating Blade not Full Out No Automatic 30 All 1 Servo if Period Less than seconds Shim Safety No SS Withdrawal All 1 Withdrawal if Count Rate Less than 3 cps No SS All 1 Withdrawal if Test / Select SW not Off Rod Control Loss of Less than or 10 Communication All 1 Scram if Communication Equal to Jseconds 21
Basis:
Specification 3.2.1.1 requires that all shim safety blades be capable of being fully inserted into the reactor core within 1 second from the time that a scram condition is initiated. As part of the Safety Analysis, Argonne National Laboratory analyzed a variety of power transients in which it was assumed that the time between the initiation of a scram signal, and full insertion of all of the shim safety rods was one second. The analysis showed that if the reactor is operated within the safety limits, this time delay will not cause an over power excursion to damage the fuel.
Specification 3.2.1.2 requires that the reactivity insertion rates of individual shim safety and regulating rods do not exceed 0.02%
dK/K per second. As part of the Safety Analysis, Argonne National Laboratory analyzed ramp insertions of 0.02% dK/K reactivity from a variety of initial power levels. The reactivity insertions are stopped by the over power trip. In all cases, peak fuel and cladding temperatures due to the power overshoot are well below the temperatures required to damage the fuel or cladding.
Consequently, this limit ensures that an over power condition due to a reactivity insertion from raising a control rod will not damage the fuel or cladding.
Specification 3.2.1.3 identifies the safety and safety related instrumentation that is required to be operable when the reactor is operated.
Two independent power level channels are required for both forced and natural convection cooling modes of operation, each of which must be capable of scramming the reactor by 105%
licensed power. The basis section of Specification 2.2.1 shows that this ensures that the power level safety limit of 2.4 MW will not be exceeded. Having two independent power level channels ensures that at least one over power protection will be available in the event of an over power excursion.
One low pool level channel is required for both forced and natural convection cooling modes of operation. This channel ensures that the reactor will not be in operation if the pool level is below the safety limit of 23 ft 6.5 inches above the top of the core.
One primary inlet coolant temperature channel is required for forced convection cooling mode operation. This channel alerts 22
the operator in the event that the inlet temperature reaches 111 F. The steady state thermal hydraulic analysis that was done by Argonne National Laboratory for forced convection flow predicts that the inlet temperature would be 115 F for operation at 2.4 MW, with a primary flow of 1580 gpm and an outlet temperature of 125 F.
One primary outlet temperature channel is required for forced convection cooling mode operation. This channel is capable of scramming the reactor when the temperature reaches 120 F.
The basis section of Specification 2.2.1 shows that this ensures that the coolant outlet temperature safety limit of 125 F will not be exceeded.
One pool temperature channel is required for natural convection cooling mode of operation. This channel is capable of scramming the reactor when the temperature reaches 125 F.
The basis section of Specification 2.2.2 shows that this ensures that the pool temperature safety limit of 130 F will not be exceeded. This channel provides the over temperature protection when the reactor is operated in the natural convection cooling mode.
One primary coolant flow rate channel is required for forced convection cooling mode operation. This channel assures that the reactor will not be operated at power levels above 100 kW with a primary coolant flow rate that is less than the safety limit of 1580 gpm. The basis section of Specification 2.2.1 shows that if this channel is set to scram at a limiting safety system setting of 1800 gpm, the safety limit will not be exceeded.
One rate of change of power channel is required for both cooling modes of operation. The 4 second period limit serves as an auxiliary protection to assure that the reactor fuel would not be damaged in the event that there was a power transient.
As part of the Safety Analysis, Argonne National Laboratory analyzed a power excursion involving a period of less than 1 second, which was stopped by an over power scram when the true power reached the limiting safety system setting of 2.3 MW. The analysis showed that peak fuel temperatures stayed well below the temperature required to damage the fuel. A 4 second period limit provides an additional layer of protection against this type of transient.
23
One seismic disturbance scram is required for both modes of operation. In the event of a seismic disturbance, the shim safety blade magnets would be likely to drop the blades due to the vibration caused by the disturbance. However, this scram ensures that the blades will be dropped in the event of a disturbance.
One bridge low power position scram is required for forced convection cooling mode operation. In order for the forced convection cooling system to work, the reactor must be seated against the high power section pool wall. This scram ensures that the reactor is properly positioned in the pool so that the coolant ducts are properly coupled with the cooling system piping.
One bridge movement scram is required for both modes of operation. This scram assures that the reactor will be shut down in the event that the bridge moves during operation.
One coolant gate open scram on each coolant duct is required during forced convection cooling mode operation. These scrams ensure that coolant flow through the inlet and outlet ducts are not bypassed during forced convection cooling.
One detector HV failure scram is required for each of the power channels, and the period channel. These channels rely on detectors that require high voltage in order to be operable.
These scrams assure that the reactor will not be operated when one of these detectors does not have proper high voltage.
One no flow thermal column scram is required during forced convection cooling mode operation. This scram ensures that there is coolant flow through the thermal column gamma shield during operations above 100 kW.
Two manual scram buttons are required to be operational during both modes of operation. One manual scram button is located in the control room, which provides the operator with a mechanism for manually scramming the reactor. The second scram button is on the reactor bridge, which provides anyone directly over the core with a mechanism for scramming the reactor if there were a reason to do so.
24
One servo control interlock that prevents the regulating blade from being put into automatic servo mode unless the blade is fully withdrawn is required for both modes of operation. As a result of this interlock, when the regulating blade is transferred to automatic servo control, the blade is unable to insert additional reactivity into the core.
One servo control interlock that prevents the regulating blade from being put into automatic mode if the period is less than 30 seconds is required for both modes of operation. This interlock limits the power overshoot that occurs when the regulating blade is put into automatic mode.
One shim safety interlock that prevents shim safety withdrawal if the start up neutron count rate is less than 3 cps is required for both modes of operation. This interlock ensures that the start up channel, which is the most sensitive indication of subcritical multiplication, is operational during reactor start-ups.
One shim safety interlock that prevents shim safety withdrawal if the neutron flux monitor test / select switch is not in the off position is required for both modes of operation. This interlock prevents shim safety withdrawal when this instrument is receiving test signals rather than actual signals from the detector that is part of the neutron flux monitor channel.
One rod control communication scram is required for both modes of operation. The control rod drive system has a communication link between the digital display in the control room, and the stepper motor controllers out at the pool top.
There is a watchdog feature that verifies that this communication link is not broken. In the event that the link is broken, a scram will occur within ten seconds of the break. All of the scram signals are sent independently of this link. The transient analysis performed by Argonne National Laboratory shows that if the control rod drive communication were lost while the reactor were on a period, the over power, and period trips would prevent the power from reaching a level that could damage the fuel cladding.
3.3 Coolant Systems 3.3.1 Primary Coolant System 25
3.3.1.1 Primary Coolant Conductivity Applicability:
This specification applies to the primary coolant.
Objective:
The objective of this specification is to maintain the primary coolant in a condition that minimizes corrosion of the fuel cladding, core structural materials, and primary coolant system components, as well as to minimize activation products produced as a result of impurities in the coolant.
Specifications:
The primary coolant conductivity shall be < 2 jimho
/ cm when averaged over a quarter of a year.
Bases:
Specification 3.3.1.1 is based on empirical data from the facility history. Over the lifetime of the facility, primary coolant conductivity has been maintained within the limit specified, and no corrosion on the fuel cladding, core structural materials, or primary coolant system components have been noted.
3.3.1.2 Primary Coolant Activity Applicability:
This specification applies to the primary coolant.
Objective:
The objective of this specification is to provide a mechanism for detecting a potential fuel cladding leak.
26
Specification:
Cs-137 and 1-131 activity in the primary coolant shall be maintained at levels that are indistinguishable from background.
Basis:
Specification 3.3.1.2 provides a mechanism for detecting a potential fuel cladding leak by requiring that periodic primary coolant analysis be performed to test for the presence of Cs-137 or 1-131. These isotopes are prominent fission products.
Consequently, if either of these isotopes are detected in the primary coolant, it may indicate a fuel cladding leak.
3.3.2 Secondary Coolant System Applicability:
This specification applies to the secondary coolant.
Objective:
The objective of this specification is to provide a mechanism for detecting a potential primary to secondary system leak.
Specifications:
Na-24 activity in the secondary coolant shall be maintained at levels that are indistinguishable from background.
Bases:
Specification 3.3.2.1 provides a mechanism for detecting a potential primary to secondary system leak by requiring that periodic secondary coolant analysis be performed to test for the presence of Na-24. This isotope is produced by the activation of the aluminum structural materials in the primary pool, and a small concentration of it is present in the primary coolant during, and immediately following operation of the reactor. If this isotope is found in the secondary coolant, it may indicate a primary to secondary system leak.
27
3.4 Confinement System 3.4.1 Operations That Require Confinement Applicability:
This specification applies to the operations for which the components of the confinement system must be operable.
Objective:
To assure that operations that have the potential to release airborne radioactive material are performed under conditions in which the release to the environment would be detected, and be limited to levels below 10 CFR 20 limits.
Specification:
- 1. The confinement system shall be operable whenever:
- 1. The reactor is operating.
- 2. Irradiated fuel handling is in progress.
- 3. Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container.
- 4. Any work on the core or control rods that could cause a reactivity change of more than 0.65% dK/K is in progress.
- 5. Any experiment movement that could cause a reactivity change of more than 0.65%
dK/K is in progress.
Bases:
The purpose of the confinement system is to mitigate the consequences of airborne radioactive material release.
During operation of the reactor, the production of radioactive gasses or airborne particulates is possible.
Though unlikely to occur, fuel cladding failure represents the greatest possible source of airborne radioactivity. The potential causes of fuel cladding damage or failure are:
28
- 1. Damage during fuel handling operations.
- 2. Fuel cladding damage due to an unanticipated reactivity excursion.
Additionally, fission products could be released due to damage to a sufficiently fueled experiment that has been irradiated long enough to build up a significant fission fragment inventory. In the event that the experiment is not adequately contained, it is conceivable that it could be damaged during handling operations to the extent that there could be fission fragment release.
These specifications ensure that the confinement system will be operable during conditions for which there is any potential for fuel cladding damage or failure to occur, as well as for experiment failures in which fission products could potentially be released.
3.4.2 Components Required to Achieve Confinement 3.4.2.1 Normal Operating Mode Confinement Applicability:
This specification describes the components of the confinement system that are necessary in order for the system to perform its intended function under normal operating conditions.
Objective:
To assure that the confinement system is capable of detecting a release of airborne radioactive material.
Specification:
- 1. The following confinement system components shall be operable:
- 1. Normal Personnel Access Door
- 2. Roll Up Door
- 3. Roof Hatch 29
Bases:
The personnel access door, roll up door, and roof hatch represent the major potential air access ways through confinement. If these components are operable, the major potential air pathways are capable of being controlled to ensure that any airborne radiological release would be detected either by the confinement radiation monitoring system, or by the stack effluent monitoring system.
Under normal operating conditions, the normal operating mode ventilation system controls the general airflow from outside confinement, through confinement, and back out to the environment through the stack.
3.4.2.2 Emergency Operating Mode Confinement Applicability:
This specification describes the components of the confinement system that are necessary in order for the system to perform its intended function under emergency operating conditions.
Objective:
To assure that the confinement system is capable of mitigating the consequences of a possible release of airborne radioactive material.
Specification:
- 1. The following emergency confinement system components shall be operable:
- 1. Emergency Confinement System Buttons
- 2. Confinement Air Intake Damper
- 3. Confinement Air Exhaust Damper
- 4. Emergency Personnel Access Door 30
Bases:
Under emergency conditions, operability of any of the emergency confinement system buttons allows the path of the airflow from confinement, through the ventilation system to be changed so that it goes through the emergency filter. Operability of the confinement air intake and exhaust dampers allows the confinement building to be isolated from the outside so that no exhaust confinement air escapes through a pathway other than the emergency pathway. Emergency mode operation of the ventilation system ensures that under emergency conditions, confinement air will be drawn through the emergency filter before being exhausted through the stack. Operability of the filter minimizes the environmental consequence of a potential airborne radioactivity release. Emergency mode operation of the ventilation system also ensures that dilution air will be added to the confinement air from the emergency filter. Operability of the emergency personnel access door allows the reactor operator to have a confinement egress route that does not require the individual to go through the main confinement room. When the door is shut, confinement is maintained.
3.4.3 Conditions Required to Achieve Confinement 3.4.3.1 Normal Operating Mode Confinement Applicability:
This specification describes the conditions necessary to assure that normal operating mode confinement is achieved.
Objective:
To assure that the confinement system is functioning sufficiently to prevent airflow from inside confinement to the environment through an uncontrolled pathway.
31
Specification:
The following conditions shall be met in order to ensure that the normal confinement is achieved:
- 1. The Normal Personnel Access Door is closed, except for entry and exit.
- 2. The Roll Up Door is closed.
- 3. The Roof Hatch is closed.
- 4. The Emergency Personnel Access Door is closed, except for entry and exit.
- 5. The Confinement Dampers are Open.
- 6. The negative differential pressure inside confinement with respect to the outside is at least 0.5 inches of water.
Bases:
Normal confinement is maintained by keeping all of the doors and the roof hatch closed, except for entry and exit. A negative differential pressure of 0.5 inches of water makes certain that the confinement system is performing its intended function adequately by ensuring that confinement airflow is directed through a defined pathway that is monitored for radiological release. The differential pressure is achieved by circulating air from outside confinement, through the intake damper, and ultimately back out of confinement through the exhaust damper.
3.4.3.2 Emergency Operating Mode Confinement Applicability:
This specification describes the conditions necessary to assure that emergency operating mode confinement is achieved.
32
Objective:
To assure that the confinement system is functioning sufficiently to prevent airflow from inside confinement to the environment through an uncontrolled pathway, and to assure that the confinement airflow pathway to the environment goes through the emergency filter and is mixed with dilution air prior to being exhausted out of the stack.
Specification:
The following conditions shall be met in order to ensure that the emergency confinement is achieved:
- 1. The Normal Personnel Access Door is closed, except for entry and exit.
- 2. The Roll Up Door is closed.
- 3. The Roof Hatch is closed.
- 4. The Emergency Personnel Access Door is closed, except for entry and exit.
- 5. The Confinement Dampers are Closed.
- 6. The negative differential pressure inside confinement with respect to the outside is at least 0.5 inches of water.
Bases:
Emergency confinement is maintained by closing the confinement intake and exhaust dampers, and by keeping all of the doors and the roof hatch closed, except for entry and exit. This causes all of the make-up confinement air to be drawn in through the spaces around the confinement penetrations, and directed through the Emergency Filter before being exhausted to the stack. A negative differential pressure of 0.5 inches of water makes certain that the confinement system is performing its intended function adequately by ensuring that confinement airflow is directed through the defined pathway that includes the emergency air filter, prior to being released to the environment.
33
3.5 Ventilation System 3.5.1 Ventilation System Components Required for Normal Operating Mode Applicability:
This specification describes the ventilation system components that must be operating in order to assure that the normal operating mode confinement is functioning.
Objective:
To assure that the normal mode confinement system is capable of performing its intended function.
Specification:
- 1. The following normal mode ventilation system components shall be operating:
- 1. Confinement Exhaust Blower
- 2. Confinement Exhaust Filter System, which shall include:
- 1. Roughing Filter
- 2. Absolute Filter
- 3. Confinement Exhaust Stack
- 2. The Confinement Exhaust Filter System Absolute Filter shall be certified by the manufacturer to have a minimum efficiency of 99.97% for removing 0.3 micron diameter particulates.
Bases:
The Confinement Exhaust Blower produces a differential pressure across confinement to ensure that all confinement air pathways are through controlled pathways. The Confinement Exhaust Filter System ensures that the majority of the radioactive particulates that would be likely to be released in the event of a fuel failure would be filtered out prior to being released to the environment, until the emergency operating mode ventilation system is activated.
The Confinement Exhaust Stack ensures that the plume of 34
confinement air that is released to the environment, is released at an elevation of 115 feet above ground level, which provides for an opportunity for the air to disperse prior to the plume reaching ground level.
3.5.2 Ventilation System Components Required for Emergency Operating Mode Applicability:
This specification describes the ventilation system components that must be operating in order to assure that the emergency operating mode confinement is functioning.
Objective:
To assure that the emergency mode confinement system is capable of performing its intended function.
Specification:
- 1. The following emergency mode ventilation system components shall be operating:
- 1. Emergency Blower
- 2. Emergency Filter System, which shall include:
- 1. Emergency Filter Intake System Roughing Filter
- 2. Emergency Filter System Intake Absolute Filter
- 3. Emergency Filter System Charcoal Filter
- 4. Emergency Filter System Exhaust Absolute Filter
- 3. Dilution Blower
- 4. Confinement Exhaust Stack
- 2. The exhaust rate through the emergency filter shall be less than or equal to 1500 cfm.
- 3. The emergency filter is shall be at least 99%
efficient at removing iodine.
35
- 4. The Emergency Filter System Exhaust Absolute Filter shall be certified by the manufacturer to have a minimum efficiency of 99.97% for removing 0.3 micron diameter particulates.
Bases:
Under emergency conditions, the Confinement Exhaust Blower turns off, and differential pressure across confinement is maintained by the Emergency Blower. The Emergency Blower directs confinement air through the Emergency Filter to remove any radioactive iodine that would be expected to be released during a fuel failure. An airflow limit of 1500 cfm though the filter ensures that the flow rate is low enough to allow the filter to adsorb at least 99 % of the iodine that would be expected to be released in the event of a fuel cladding failure. The Emergency Filter System Exhaust Absolute Filter prevents charcoal particulates from the charcoal filter from being released to the building exhaust air stream. The Dilution Blower provides a non-contaminated source of air to mix with the confinement air, so that any airborne radioactivity that is released is diluted prior to release. The Confinement Exhaust Stack ensures that the plume of confinement air that is released to the environment, is released at an elevation of 115 feet above ground level, which provides for an opportunity for the air to disperse prior to the plume reaching ground level.
3.6 Emergency Power System 3.6.1 Required Emergency Power Sources Applicability:
This specification describes the emergency electrical power sources that are neeessai=y required in order to ensure that power is available to confinement system components that are necessary to ensuwe make certain that the confinement system is able to perform its intended function in the event of an electrical power outage.
36
Objective:
To assure that the confinement system is able to perform its intended function even, when normal electrical power is unavailable.
Specification:
- 1. An emergency electrical power source shall be operable whenever the confinement system is required to be operable.
Bases:
Operability of the emergency electrical power source ensures that the blower systems that are necessary in order to maintain emergency operation mode confinement will remain operable, even in the event of a facility electrical power outage.
3.6.2 Components Required to be Supplied with Emergency Power Applicability:
This specification describes the confinement system components that are required to be connected to an emergency electrical power source.
Objective:
To assure that the confinement system is able to perform its
.e..*,, f...nction intended when normal electrical power is unavailable.
Specification:
- 1. The following confinement system components shall be connected to an emergency power source:
- 1. Emergency Blower
- 2. Dilution Blower 37
Bases:
In the event of a power outage, the reactor will scram due to the loss of magnet current to the shim safety blades. The confinement air intake and exhaust dampers are pneumatically operated and will fail closed to isolate the confinement room. The confinement exhaust blower will shut off due to loss of power. As long as the emergency and dilution blowers continue to be operable, the emergency confinement system will continue to perform its intended function. In the event of a power outage, the emergency power source will supply the emergency and dilution blowers with electricity so that they will continue to operate, and the emergency confinement system will continue to be functional.
3.7 Radiation Monitoring Systems and Effluents 3.7.1 Radiation Monitoring Systems 3.7. 1.1 Required Radiation Monitoring Systems Applicability:
This specification applies to the radiation monitoring systems required for critical operation of the reactor, and fuel handling activities.
Objective:
The objective of this specification is to define the minimum set of radiation monitoring systems that must be operable for the reactor to be made critical, or for fuel handling activities.
Specifications:
3.7.1.1. The following air radiation monitoring instrumentation shall be operable when the reacter is in eper-atien, and during fuel handling
.p..a'i-ns whenever:
The reactor is operating, Irradiated fuel handling is in progress, 38
Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container, Any work on the core or control rods that could cause a reactivity change of more than 0.65% dK/K is in progress, or Any experiment movement that could cause a reactivity change of more than 0.65% dK/K is in progress:
3.7.1.1.1 A minimum of one radiation monitor that is capable of warning personnel of high radiation levels in the confinement gaseous or particulate effluent shall be operating.
3.7.1.1.2 If this detector fails during operation, a suitable alternative gaseous or particulate air monitor may be used, or an hourly grab sample analysis may be made in lieu of having a functioning monitor.
3.7.1.2 The following fission product radiation monitoring instrumentation shall be operable when the feaetor is in oper-ation, and during fuel handling ,perati.ns. whenever:
The....reactor AAý ý" ,.. is opera.ting,
- oea~g Irradiated fuel handling is in progress, Experiment handling is in progress for an experiment that has a significant fission product inventory, and for which the experiment is not inside a container, Any work on the core or control rods that could cause a reactivity change of more than 0.65% dK/K is in progress, or 39
Any experiment movement that could cause a reactivity change of more than 0.65%
dK/K is in progress:
3.7.1.2.1 A minimum of one gamma sensitive radiation monitor that is capable of warning personnel of high radiation levels shall be over the pool.
3.7.1.2.2 If this detector fails, a suitable gamma sensitive alternative meter with alarming capability may be placed at the top of the bridge in lieu of the normal detector.
3.7.1.2.3 If failure of the detector occurs during operations than the replacement detector must be put in place within 15 minutes of the recognition of failure.
3.7.1.3 Passive radiation monitors provided by a certified vendor shall be used to provide area radiation monitoring inside confinement when the reactor is in operation, and during fuel handling operations.
3.7.1.4 Passive radiation monitors provided by a certified vendor shall be used to provide environmental radiation monitoring when the reactor is in operation, and during fuel handling operations.
Bases:
A continuing evaluation of the air within confinement will be made in order to ensure that the airborne radioactivity concentration does not exceed 10 CFR 20 limits for personnel working inside confinement, and that the concentration exhausted from confinement does not exceed the limits for the general public.
40
Specification 3.7.1.1.1 identifies the air radiation monitoring instrumentation that is required to be operable when the reactor is operated, and during fuel handling operations.
Specification 3.7.1.1.2 allows for the air monitoring instrumentation to be replaced, or for grab samples to be performed in the event that the normal instrument fails.
Continuous monitoring for fission product release is performed at the pool top. In the event of a release, it is anticipated that the first indication would come from the pool top radiation detector which would detect the noble gasses, particularly Krypton and Xenon.
Specification 3.7.1.2.1 identifies the fission product monitoring instrumentation that is required to be operable when the reactor is operated, and during fuel handling operations.
Specification 3.7.1.2.2 allows for the fission product monitoring instrumentation to be replaced in the event that the normal instrument fails.
Specification 3.7.1.3 requires that passive area radiation monitoring be in place during reactor operation and fuel handling activities. Active monitoring is performed in accordance with the RINTSC Radiation Safety Program, and tailored to meet the specific needs of each experiment as they are performed. The passive monitoring provides a mechanism for reconstructing maximum doses as a backup to the active systems used.
Specification 3.7.1.4 requires that passive environmental radiation monitoring be in place during reactor operation and fuel handling activities. This provides a mechanism for determining doses outside confinement.
41
3.7.1.2 Radiation Monitoring System Alarm Set Points Applicability:
This specification applies to the radiation monitoring systems required for critical operation of the reactor, and fuel handling activities.
Objective:
The objective of this specification is to ensure that personnel are notified in the event of unusually high radiation levels.
Specifications:
3.7.2.1.1 The stack gaseous monitor shall alarm when radiation levels of the stack gas are 2.5 times normal levels, or greater.
3.7.2.1.2 The stack particulate monitor shall alarm when radiation levels of the stack particulates are 2 times normal levels, or greater.
3.7.2.1.3 The area radiation monitors shall alarm when radiation levels are 2 times normal levels, or greater.
3.7.2.1.4 Alarm set points may be adjusted higher with the approval of the Director or Assistant Director.
Bases:
All of the radiation monitors in the confinement room have set points that are in terms of "normal" radiation levels. The purpose of defining set points in terms of "normal" radiation levels is to account for the fact that the radiation levels vary in the confinement room, depending on what kinds of experiments are being performed.
42
3.7.2 Effluents 3.7.2.1 Airborne Effluents Applicability:
This specification applies to the monitoring of airborne effluents from the Rhode Island Nuclear Science Center (RINSC).
Objective:
To assure that the release of airborne radioactive material from the RINSC will not cause the public to receive doses that are greater than the limits established in 10 CFR 20.
Specifications:
Airborne effluents shall be monitored by an air monitor installed, calibrated and maintained in accordance with ANSI 13.1. The annual total effective dose equivalent to the individual member of the public likely to receive the highest dose from air effluents will be calculated using a generally-accepted computer program.
Bases:
10 CFR 20.1101(d) states, in part, "to implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon-222 and its daughters, shall be established by licensees other than those subject to § 50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions."
Since the Rhode Island Nuclear Science Center is located on Narragansett Bay, the wind does not blow in the same direction more than about 10% of the time as shown in the following table taken from historical wind rose data.
43
Wind Blowing From Frequency % Wind Blowing From Frequency %
North 6.20 E-02 6.02 South 5.80 E-02 5.80 North/Northeast 5.80 E-02 5.80 South/Southwest 8.40 E-02 8.40 Northeast 4.40 E-02 4.40 Southwest 1.05 E-01 10.50 East/Northeast 1.30 E-02 1.30 West/Southwest 6.40 E-02 6.40 East 1.20 E-02 1.20 West 6.80 E-02 6.80 East/Southeast 1.30 E-02 1.30 West/Northwest 9.50 E-02 9.50 Southeast 5.80 E-02 6.80 Northwest 1.04 E-01 10.40 South/Southeast 4.90 E-02 4.90 North/Northwest 6.80 E-02 6.80 Thus, during routine operations, no individual would be in the pathway of the plume more than about 10% of the time. Calculations of annual dose equivalent due to the primary airborne effluent, Argon-41, using the COMPLY Code show less than the allowable ALARA limitation given in 10 CFR 20.1101 for the hypothetical maximum exposed individual member of the general public.
3.7.2.2 Liquid Effluents Applicability:
This specification applies to the monitoring of radioactive liquid effluents from the Rhode Island Nuclear Science Center.
Objective:
The objective is to assure that exposure to the public resulting from the release of liquid effluents will be within the regulatory limits and consistent with as low as reasonably achievable requirements.
Specifications:
The liquid waste retention tank discharge shall be batch sampled and the gross activity per unit volume determined before release.
Bases:
10 CFR 20.2003 permits discharges to the sanitary sewer provided that conditions in 10 CFR 20.2003 (a) are met.
44
3.8 Experiments 3.8.1 Experiment Materials Applicability:
This specification describes the limitations on the types of materials that may be irradiated or installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to prevent damage to the reactor, reactor pool, and reactor experimental facilities.
Specification:
- 1. Corrosives Materials
- 1. Corrosive materials shall be doubly contained in corrosion resistant containers.
- 2. Highly Water Reactive Materials
- 1. Highly water reactive materials shall not be placed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
- 3. Explosive Materials
- 1. Explosive materials shall not be placed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
- 4. Fissionable Materials
- 1. The quantity of fissionable materials used in experiments shall not cause the experiment reactivity worth limits to be exceeded.
- 2. Fissionable materials shall be doubly encapsulated.
45
- 3. Containers for experiments that have fissionable material shall be opened inside confinement.
Basis:
ANSI 15.1 recommends that the kinds of materials used in experiments be taken into consideration in order to limit the possibility of damage to the reactor, reactor pool, or reactor experimental facilities. Specifically, ANSI suggests that:
Damage could arise as a result of corrosive materials reacting with core, or experimental facility materials.
Specification 3.8.1.1 reduces the possibility of this by requiring that corrosive materials be doubly contained so that the likelihood of container breach is minimized.
Damage could arise as a result of highly water reactive materials reacting with the pool water. Specification 3.8.1.2 makes this scenario impossible by prohibiting the use of highly water reactive materials in experiments.
Damage could arise as a result of explosive materials reacting inside and experimental facility. Specification 3.8.1.3 makes this scenario impossible by prohibiting the use of explosive materials in experiments.
Failure of experiments that contain fissionable materials have the potential to have an impact on reactor criticality, or on radioactive material release. The consequence of experiment failure on criticality is bounded by limiting the reactivity worths of experiments. The analysis for this is in SAR Chapter 13 as part of the transient analysis. The radioactive material release is bounded by the analysis in SAR Chapter 13 for the Maximum Hypothetical Accident involving a fuel element failure. Double encapsulation of fissionable materials reduces the probability of the release of radioactive material. The requirement that experiments containing fissionable materials be opened inside confinement ensures that in the event of a fission product gas release, the mitigating actions of the confinement system would be available.
46
3.8.2 Experiment Failures or Malfunctions Applicability:
This specification applies to experiments that are installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to ensure that experiments cannot fail in such a way that they contribute to the failure of other experiments, core components, or principle barriers to the release of radioactive material.
Specification:
- 1. Experiment design shall be reviewed to ensure that credible failure of any experiment will not result in releases or exposures in excess of limits established in 10 CFR 20.
- 2. Experiment design shall be reviewed to ensure that no reactor transient can cause the experiment to fail in such a way that it contributes to an accident.
- 3. Experiment design shall be reviewed to ensure that credible failure of any experiment will not contribute to the failure of:
- 1. Other Experiments
- 2. Core Components
- 3. Principle physical barriers to uncontrolled release of radioactivity
- 4. Experiments which could increase reactivity by flooding shall not remain in the core, or adjacent to the core unless the minimum core shutdown margin required would be satisfied with the experiment in the flooded condition.
47
Basis:
ANSI 15.1 recommends that experiment design be taken into consideration in order to limit the possibility that an experiment failure or malfunction could result in other failures, accidents, or significant releases of radioactive material.
Experiments are reviewed by the RINSC Nuclear and Radiation Safety Committee prior to being authorized to be installed in the reactor pool, or inside the reactor experimental facilities. These specifications ensure that experimental design is considered as part of the review, in order to minimize the possibility of these types of problems due to experiment failure or malfunction.
In order to determine the reactivity worth of a new experiment for which there is no data based on similar experiments, the only way to determine the reactivity worth of the experiment is to perform an approach to critical with the experiment loaded in the core. In that case, it is possible that an experiment could be found to have enough positive reactivity that if additional positive reactivity were added due to flooding, the shutdown margin would be less than 1.0 % dK/K. In that event, Technical Specification 3.8.2.4 requires that the experiment be removed immediately.
3.9 Reactor Core Components 3.9.1 Beryllium Reflectors Applicability:
This specification applies to neutron flux damage to the standard and plug type beryllium reflectors.
Objective:
To prevent physical damage to the beryllium reflectors in the core from accumulated neutron flux exposure.
Specification:
The maximum accumulated neutron flux shall be lX 1022 neutrons/cm 2 .
48
Basis:
This limit is based on an analysis that was done by the University of Missouri Research Reactor (MURR). In their analysis, they note that the HFIR Reactor has noticed the presence of small cracks at fast fluences of 1.8 X 10" nvt, and suggest that "a value of 1 X 1022 nvt (>lMeV) could be used as a conservative lower limit for determining when replacement of a beryllium reflector should be considered."
The RINSC limit of I X 1022 nvt is even more conservative than what this analysis considers because it is not limited to fast neutron flux.
3.9.2 Low Enriched Uranium Fuel Applicability:
This specification applies to the physical condition of the fuel elements.
Objective:
To prevent operation with damaged fuel elements.
Specifications:
Fuel elements to shall be inspected for physical defects and reactor core box fit in accordance with manufactured specifications.
Bases:
The RINSC inspects and tests each fuel element for reactor core box fit in accordance with written procedures to assure operation with fuel elements that are not damaged and meet specifications.
3.9.3 Experimental Facilities 3.9.3.1 Experimental Facility Configuration During Reactor Operation 49
Applicability:
These specifications apply to the reactor experimental facilities during reactor operation.
Objective:
These specifications ensure that in the event of a Maximum Credible Accident, the rate at which the pool level would decrease would be low enough to make certain that the fuel cladding would not be damaged due to insufficient cooling.
Specifications:
3.9.3.1.1. Each beam port shall have no more than an area of 1.25 in2 open to confinement during reactor operation.
3.9.3.1.2. When the reactor is in operation, the drain valve to the through port shall be closed.
3.9.3.1.3. When the through port is in use, gate valves shall be installed on the end(s) of the port that will be used for access.
3.9.3.1.4. When the through port is not being monitored for a leak condition, the ends of the port shall be closed.
Bases:
Specification 3.9.3.1.1:
The LOCA analysis shows that as long as the pool level does not drain through an area greater than 1.48 in2 to confinement, then there will be sufficient time for decay power to drop to a point which will not damage the fuel cladding, provided that the pool level does not drop below the elevation of the bottom of the eight inch beam ports. It also shows that if any single port has a catastrophic failure, the un-damaged ports do not become pool drain pathways. Consequently, 50
limiting the areas of each experimental port that is open to confinement to 1.25 in2 is conservative.
Specification 3.9.3.1.2:
Shearing the through port is not considered to be a credible accident. Consequently, a leak in the through port is not anticipated to be catastrophic.
The through port has three potential pool leak pathways. The first is the through port drain. By keeping this drain closed during operation, that potential leak pathway is blocked, and the potential for an unnoticed pool leak though this experimental facility is prevented.
Specification 3.9.3.1.3:
If the end(s) of the through port that will be used for access have gate valves mounted to them, then in the event of a leak, the port can be easily isolated so that the leak is stopped.
Specification 3.9.3.1.4:
The LOCA analysis has shown that the amount of time available for performing mitigating actions in the event of a non-catastrophic pool leak is on the order of hours. Consequently, as long as reactor personnel will become aware of a pool leak though the through port reasonably quickly, and the gate valves are in place, the consequence of the leak can be mitigated quickly by closing the valves.
3.9.3.2 Experimental Facility Configuration Within 4.5 Hours After Shutdown Applicability:
These specifications apply to the reactor experimental facilities for the 4.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period after reactor shutdown.
51
Objective:
These specifications ensure that in the event of a Maximum Credible Accident, the rate at which the pool level would decrease would be low enough to make certain that the fuel cladding would not be damaged due to insufficient cooling.
Specifications:
3.9.3.2.1. If there is no need to open a beam port within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, then the 1.25 in2 area opening to confinement shall be maintained until that time period has passed.
3.9.3.2.2. If there is a need to open a beam port within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, then:
3.9.3.22.1I The reactor shall be moved to the low power section of the pool where it is at the opposite end of the pool from the beam port extensions.
3.9.3.2.2.2. The pool gate shall be positioned so that the high power section of the pool is isolated in such a way that if a beam port extension were sheared off, the pool level in the low power section would not be affected.
Bases:
Specification 3.9.3.2.1 The LOCA analysis shows that if the reactor were operated for an infinite amount of time at 2 MW, the amount of time that it would take for the power fraction to decay after shutdown to a point where the fuel cladding blister temperature could not be reached, even if the pool level were at the elevation of the bottom of the 8 inch beam ports, would be 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The analysis also shows that the 52
maximum area of an opening between a beam port and confinement that limits this drain time to 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is 1.48 in2 . Consequently, maintaining the limit on the area open between confinement and the beam ports to 1.25 in2 for a period of 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown ensures that in the event of a catastrophic beam port failure, the drain time would provide sufficient time for power to decay to a point below which the fuel could not be damaged.
Specification 3.9.3.2.2 In the event that access to a beam port is needed within 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown, a provision is made so that the core can be isolated from the beam port end of the pool. With the core in the low power end of the pool, and the pool gate in place, if a beam port extension were sheared off, and a catastrophic beam port failure were to occur, the coolant level above the core would not be affected.
53
4.0 Surveillance Requirements 4.1 Core Parameter Surveillance 4.1.1 Reactivity Limit Surveillance Applicability:
This specification applies to the surveillance requirements for reactivity limits.
Objective:
The objective of this specification is to ensure that reactivity limits are not exceeded.
Specification:
4.1.1.1 Core Reactivity Limit Surveillance 4.1.1.1.1 The core shutdown margin shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Following control blade changes.
4.1.1.1.2 The core excess reactivity shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Following control blade changes.
4.1.1.1.3 The temperature coefficient shall be shown to be negative at the initial start-up after a fuel type change.
54
4.1.1.2 Control Rod Reactivity Limit Surveillance 4.1.1.2.1 The reactivity worth of the regulating rod shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Whenever maintenance is performed that could have an effect on the reactivity worth of the control rod 4.1.1.2.2 The reactivity worth of the shim safety rods shall be determined:
Annually Whenever the core reflection is changed Whenever the core fuel loading is changed Whenever maintenance is performed that could have an effect on the reactivity worth of the control rod 4.1.1.3 Experiment Reactivity Limit Surveillance 4.1.1.3.1 The reactivity worth of new experiments shall be determined prior to the experiments initial use.
4.1.1.3.2 The reactivity worth of any on going experiments shall be re-determined after the core configuration has been changed to a configuration for which the reactivity worth has not been determined previously.
Basis:
Specification 4.1.1.1.1 requires that the core shutdown margin be determined annually, and whenever there is a change in core loading or core reflection. The annual measurement of the shutdown margin provides a snapshot 55
of how the shutdown margin is increasing due to fuel bum-up. Measurements made whenever the core loading or reflection is changed provide assurance that core reactivity limits are not being exceeded due to changes in core configuration.
Specification 4.1.1.1.2 requires that the core excess reactivity be determined annually, and whenever there is a change in core loading or core reflection. The annual measurement of the excess reactivity provides a snapshot of how it is decreasing due to fuel bum-up. Measurements made whenever the core loading or reflection is changed provide assurance that core reactivity limits are not being exceeded due to changes in core configuration.
Specification 4.1.1.1.3 requires that the temperature coefficient be shown to be negative at the initial start-up after a fuel type change. A negative temperature coefficient makes power increases self limiting by inserting a negative reactivity effect as fuel and coolant temperatures rise. As part of the Safety Analysis, Argonne National Laboratory determined that for the equilibrium core, the temperature and void coefficients are negative over a temperature range of 20 C to 100 C. The fuel temperature coefficient was determined to be negative over a temperature range of 20 C to 600 C.
Specification 4.1.1.2.1 requires that the regulating rod reactivity be determined annually, and whenever there is a change in core loading or core reflection. These determinations provide assurance that the rod worth does not exceed its reactivity limit due to fuel bum-up, changes in core configuration, or control rod degradation.
Specification 4.1.1.2.2 requires that the shim safety rod reactivities be determined annually, and whenever there is a change in core loading or core reflection. These determinations provide assurance that the rod worths do not degrade due to rod changes, or changes in core configuration.
Specification 4.1.1.3.1 requires that the reactivity worth of new experiments be determined prior to initial use. This ensures that reactivity worth limits are not exceeded.
56
Specification 4.1.1.3.2 requires that the reactivity worth of on going experiments be re-determined after the core configuration has been changed to a configuration for which the reactivity worth has not been determined previously. This provides assurance that core configuration changes do not cause experiment reactivity worth limits to be exceeded, without requiring that experiment worths be re-determined every time that a recurring core configuration change, such as equilibrium core re-fuelling, occurs.
4.2 Reactor Control and Safety System Surveillance Applicability:
This specification applies to the safety and safety related instrumentation.
Objective:
The objective of this specification is to ensure that the safety and safety related instrumentation is operable, and calibrated when in use.
Specification:
4.2.1 Shim safety drop times shall be measured:
4.2.1.1 Annually 4.2.1.2 Whenever maintenance is performed which could affect the drop time of the blade 4.2.1.3 When a new core is configured 4.2.1.4 Following control blade changes 4.2.2 All shim safety reactivity insertion rates shall be measured:
4.2.2.1 Annually 4.2.2.2 Whenever maintenance is performed which could affect the reactivity insertion rate of the blade 4.2.2.3 When a new core is configured 4.2.2.4 Following control blade changes 57
4.2.3 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired of-de ene-gie (this is a legacy from tube based technology):
4.2.3.1 Control room manual scram button 4.2.3.2 Power level channels 4.2.3.3 Period channel 4.2.3.4 Red e.ntro.l communication watchd. g seram move to annual - TS 4.2.5.9) 4.2.4 The following reactor safety and safety related instrumentation shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and for which reactor power level will be greater than 100 kW, and after the channel has been repaired or-de ener.giz. d fo*r whi. h r-eaet.r. power- level will be groater- than 100 WA:II 4.2.4.1 All of the reactor safety and safety related instrumentation listed in 424 4.2.3.
4.2.4.2 Primary coolant flow rate scram 4.2.5 The following reactor safety and safety related instrumentation alarms, scrams, and interlocks shall be tested annually:
4.2.5.1 The following detector HV failure scrams:
4.2.5.1.1 Power level channels 4.2.5.1.2 Period channel 4.2.5.2 The following shim safety withdrawal interlocks:
4.2.5.2.1 Start-up count rate 4.2.5.2.2 Test / Select switch position 4.2.5.3 The following servo control interlocks:
4.2.5.3.1 Regulating blade not full out 4.2.5.3.2 Period less than 30 seconds 4.2.5.4 The following coolant system channel temperature alarms and scrams:
58
4.2.5.4.1 Primary inlet temperature alarm 4.2.5.4.2 Primary outlet temperature alarm 4.2.5.4.3 Primary outlet temperature scram 4.2.5.4.4 Pool temperature alarm 4.2.5.4.5 Pool temperature scram 4.2.5.5 The following coolant system channel flow scrams:
4.2.5.5.1 Primary flow scram 4.2.5.5.2 Inlet and outlet coolant gates open scrams 4.2.5.5.3 No flow thermal column scram 4.2.5.6 Low pool level scram 4.2.5.7 The following bridge scrams:
4.2.5.7.1 Bridge manual scram 4.2.5.7.2 Bridge movement scram 4.2.5.7.3 Bridge low power position scram 4.2.5.8 Seismic scram 4.2.5.9 Rod control communication watchdog scram 4.2.6 The following reactor safety and safety related instrumentation shall be calibrated annually:
4.2.6.1 Power level channels 4.2.6.2 Primary flow channel 4.2.6.3 Primary inlet temperature channel 4.2.6.4 Primary outlet temperature channel 4.2.6.5 Pool temperature channel Basis:
Specification 4.2.1 defines the surveillance interval for measuring the shim safety drop times. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1. The requirement that this parameter be measured after maintenance is performed which could affect the drop time of the blade assures that the reactor will not be operated with a shim safety blade that does not meet the LCO requirements due to maintenance activities.
59
Specification 4.2.2 requires that all shim safety reactivity insertion rates shall be measured annually. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.
Specification 4.2.3 indicates the reactor safety and safety related instrumentation that must be verified to be operable prior to the initial reactor start-up of each day. This requirement is consistent with the historical facility requirements.
Specification 4.2.4 provides for the fact that if the reactor is operated at power levels less than or equal to 100 kW, the forced cooling system is not required to be operational. However, for operations above 100 kW, this specification requires that the primary coolant flow rate scram be verified to be operable prior to the initial start-up of the reactor. This requirement is consistent with the historical facility requirements.
Specification 4.2.5 defines the surveillance interval for testing the reactor safety and safety related instrumentation alarms, scrams, and interlocks that are not tested as part of the requirements of Specifications 4.2.3 and 4.2.4. The annual requirement is consistent with the historical facility frequency.
Specification 4.2.6 defines the surveillance interval for calibrating the safety and safety related instrumentation. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.
4.3 Coolant Systems Surveillance 4.3.1 Primary Coolant System 4.3.1 I Primary Coolant Conductivity Surveillance Applicability:
This specification applies to the surveillance of the primary coolant.
Objective:
The objective of this specification is to provide a periodic verification that the primary coolant conductivity is within prescribed limits.
60
Specification:
The conductivity of the primary coolant shall be tested monthly.
Basis:
Specification 4.3.1.1 requires that the conductivity of the primary coolant be tested on a monthly basis.
ANSI 15.1 recommends that this be performed on a weekly to quarterly schedule. Specification 3.1.1.1 sets a limit on the average conductivity when averaged over one quarter of a year. Consequently, a monthly measurement falls within the ANSI recommended schedule, and allows for a running average based on three data points per quarter.
4.3.1.2 Primary Coolant Activity Surveillance Applicability:
This specification applies to the surveillance of the primary coolant.
Objective:
The objective of this specification is to provide a periodic verification that the Cs-137 and 1-131 activity in the primary coolant is not significantly above background.
Specifications:
Cs-137 and .1-131 activity in the primary coolant shall be measured annually.
Basis:
Specification 4.3.1.2 requires that the Cs-137 and I-131 activity in the primary coolant be tested on an annual basis. This schedule is consistent with the schedule recommended by ANSI 15.1. These isotopes are indicators of a fuel failure.
61
4.3.1.3 Primary Coolant Level Inspection Surveillance Applicability:
This specification applies to the surveillance of the primary coolant.
Objective:
The objective of this specification is to ensure that the coolant level is at an adequate height above the core during reactor operation.
Specification:
The primary coolant level shall be verified to be greater than or equal to the Limiting Safety System Setting value prior to the initial start-up each day that the reactor is started up from the shutdown condition.
Basis:
Specification 4.3.1.3 requires that the primary coolant level be inspected prior to the first reactor start-up of each day. A float switch system is used to monitor the pool level 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week. This system is tied into the facility alarm system, which is monitored by an offsite alarm company. In the event that the pool level reaches one inch greater than the LSSS, the automatic pool fill is started. If the pool level drops to the LSSS, then a scram occurs, the operator receives an alarm, and the alarm company receives an alarm. A daily verification of the pool level prior to starting the reactor up provides adequate assurance that the float switch is working to maintain the pool level.
4.3.1.4 Primary Coolant System Inspection Surveillance Applicability:
This specification applies to the surveillance of the primary cooling system components.
62
Objective:
The objective of this specification is to provide a periodic verification that there are no obvious defects in any of the system components.
Specifications:
The components of the primary coolant system shall be inspected annually.
Basis:
Specification 4.3.1.4 requires that the primary coolant system be inspected on an annual basis to ensure that the integrity of the pool and other cooling system components are not degraded.. This schedule is consistent with the historical inspection schedule for the facility.
4.3.2 Secondary Coolant System 4.3.2.1 Secondary Coolant Activity Surveillance Applicability:
This specification applies to the surveillance of the secondary coolant.
Objective:
The objective of this specification is to provide a periodic verification that the Na-24 activity in the secondary coolant is not significantly above background.
Specification:
Na-24 activity in the secondary coolant shall be measured annually.
63
Basis:
Specification 4.3.2.1 requires that the Na-24 activity in the secondary coolant be tested on an annual basis. This schedule is consistent with the schedule recommended by ANSI 15.1.
4.3.2.2 Secondary Coolant System Inspection Surveillance Applicability:
This specification applies to the surveillance of the secondary cooling system components.
Objective:
The objective of this specification is to provide a periodic verification that there are no obvious defects in any of the system components.
Specification:
The components of the secondary coolant system shall be inspected annually.
Basis:
Specification 4.3.2.2 requires that the primary coolant system be inspected on an annual basis.
This schedule is consistent with the historical inspection schedule for the facility.
4.4 Confinement System Surveillance 4.4.1 Normal Operating Mode Confinement System Applicability:
This specification describes the surveillance requirements for the normal operating mode confinement system.
Objective:
The objective of this specification is to verify that the normal operating mode confinement system is functional prior to reactor start-up.
64
Specification:
4.4.1.1. The conditions required to achieve normal operating mode confinement that are specified in section 3.4.3.1 shall be verified to be met prior to the each day of reactor start-up.
Bases:
If the conditions specified in section 3.4.3.1 are met, then the normal operating mode confinement system is functioning. By ensuring that the normal operating mode confinement system is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that any airborne radioactivity release would be directed to the stack, mixed with dilution air, and detected by the stack radiation monitor system.
4.4.2 Emergency Operating Mode Confinement System Applicability:
This specification describes the surveillance requirements for the emergency operating mode confinement system.
Objective:
The objective of this specification is to verify that the emergency operating mode confinement system is functional.
Specification:
4.4.2.1. A functional test of the emergency operating mode confinement system shall be performed:
4.4.2.1.1. Quarterly 4.4.2.1.2. After any maintenance that could affect the operability of the system 4.4.2.2. The functional test of the emergency operating mode confinement system shall verify that the conditions required to achieve emergency operating mode confinement are met when an evacuation button is depressed. The following 65
actions shall occur when an evacuation button is depressed:
4.4.2.2.1. The evacuation horn sounds 4.4.2.2.2. The following dampers close:
4.4.2.2.2.1. Confinement Air Intake Damper 4.4.2.2.2.2. Confinement Air Exhaust Damper 4.4.2.2.3. The negative differential pressure inside confinement with respect to the outside is at least 0.5 inches of water.
4.4.2.2.4. The confinement room HVAC and air conditioners de-energize.
Bases:
A periodic functional test of the emergency confinement system ensures that in the event of an airborne radioactivity release, the emergency confinement system is capable of being activated. The testing periods that are specified conform to ANSI 15.1 recommendations.
4.5 Ventilation System Surveillance 4.5.1 Normal Operating Mode Ventilation System Applicability:
This specification describes the surveillance requirements for the normal operating mode ventilation system.
Objective:
The objective of this specification is to verify that the normal operating mode ventilation system is operable prior to reactor start-up.
Specification:
4.5.1.1. The confinement exhaust blower shall be verified to be in operation prior to each day of reactor start-up:
66
Bases:
By ensuring that the normal operating mode ventilation system is functional prior to each day of reactor start-up, conditions are verified to be in place to make certain that any airborne radioactivity release would be directed to the stack and be detected by the stack radiation monitor system.
4.5.2 Emergency Operating Mode Ventilation System Applicability:
This specification describes the surveillance requirements for the emergency operating mode ventilation system.
Objective:
The objective of this specification is to verify that the emergency operating mode ventilation system is operational and functional.
Specifications:
4.5.2.1. A test of the operability of the emergency operating mode ventilation system shall be performed:
4.5.2.1.1. Quarterly 4.5.2.1.2. After any maintenance that could affect the operability of the system 4.5.2.2. The test of the operability of the emergency operating mode ventilation system shall verify that the following actions occur when an evacuation button is depressed:
4.5.2.2.1. The following blowers are de-energized:
4.5.2.2.1.1. Confinement Exhaust Blower 4.5.2.2.1.2. Rabbit System Blower 4.5.2.2.1.3. Off Gas System Blower 67
4.5.2.2.2. The following blowers are energized:
4.5.2.2.2.1. Emergency Exhaust Blower 4.5.2.2.2.2. Dilution Blower 4.5.2.3. The flow rate at the exhaust of the emergency exhaust blower shall be verified to be less than or equal to 1500 cfm:
4.5.2.3.1. Annually 4.5.2.3.2. After any maintenance that could affect the operability of the system 4.5.2.4. The emergency filter efficiency shall be verified to be at least 99% efficient for removing iodine:
4.5.2.4.1. Biennially 4.5.2.4.2. After any maintenance that could affect the operability of the system Bases:
A periodic test of the operability of the emergency ventilation system ensures that in the event of an airborne radioactivity release, the emergency confinement system is capable of being activated. The verification of the emergency exhaust blower flow rate, and the emergency filter efficiency ensure that the filter will perform its intended function. The testing periods that are specified conform to ANSI 15.1 recommendations.
4.6 Emergency Power System Surveillance Applicability:
This specification describes the surveillance requirements for the emergency power system.
Objective:
The objective of this specification is to verify that the emergency power system is operable and functional.
68
Specification:
4.6.1. An operability test to verify that the emergency power system starts in the event of a facility power outage shall be performed quarterly.
4.6.2. A functional test of the emergency power system under load shall be performed:
4.6.2.1. Biennially 4.6.2.2. Following emergency system load changes Bases:
Periodic tests of the emergency power system ensures that in the event of a facility power outage, the emergency power system would automatically start, and be capable of handling the load required to power the emergency confinement system. The testing periods that are specified conform to ANSI 15.1 recommendations.
4.7 Radiation Monitoring Systems and Effluents Surveillance 4.7.1. Monitoring Systems Applicability:
This specification applies to the radiation monitoring systems that are required to be operable during reactor operation and fuel handling activities.
Objective:
The objective of this specification is to verify the operability of required radiation monitoring instrumentation.
Specifications:
4.7.1.1 The following radiation monitors shall be verified to be operable prior to the initial start-up each day that the reactor is started up from the shutdown condition, and after the channel has been repaired r-de-eenergite:
69
4.7.1.1.1 The experimental level area radiation monitor 4.7.1.1.2 The pool top area radiation monitor 4.7.1.1.3 The gaseous effluent air monitor 4.7.1.1.4 The particulate air monitor 4.7.1.2 The following reactor safety and safety related instrumentation shall be calibrated annually:
4.2.7.1 The experimental level area radiation monitor 4.2.7.2 The pool top area radiation monitor 4.2.7.3 The gaseous effluent air monitor 4.2.7.4 The particulate air monitor Bases:
Specification 4.7.1.1 indicates the radiation monitors that must be verified to be operable prior to the initial reactor start-up of each day. This requirement is consistent with the historical facility requirements.
Specification 4.7.1.2 defines the surveillance interval for calibrating the safety and safety related instrumentation. The annual requirement is consistent with the historical facility frequency, and is within the range recommended by ANSI Standard 15.1.
4.7.2. Effluents Applicability:
These specifications apply to the surveillance of the monitoring equipment used to measure the radioactivity in effluents.
Objective:
The objective of these specifications is to assure that an accurate assessment of the radiological material release from the facility via effluents can be made.
Specifications:
4.7.2.1. The monitoring equipment used to measure the radioactive concentrations in the waste retention tanks shall be calibrated annually.
70
4.7.2.2. The1 ontents of the retention tanks shall be samnpled and cvaluiated for radioactiv-e
.n*entrations and p14 prior. to release. (This is specificed in TS 3.7.2.2)
Bases:
Exper-ienee '%ith the electronic reliability' and calibration stabiliy of the units used by the Rhede island Nuelcar Science Center Reactor- demenstraftes tha the abov.e per-iods are r-easonable surveillance ffeettenci*s An annual calibration period is within the range that is recommended by ANSI Standard 15.1 for radiation monitors.
4.8 Experiments Surveillance Applicability:
This specification applies to experiments that are installed inside the reactor, the reactor pool, or inside the reactor experimental facilities.
Objective:
The objective of this specification is to ensure that experiments have been reviewed to verify that the design is within the limitations of the RINSC Technical Specifications and 10 CFR 50.59.
Specification:
4.8.1 Experiments shall be reviewed to cnsure that the design is within the limitations of the RINSC Technical Specifications and 10 CFR 50.59 prior to the experiments initial use.
Basis:
This specification ensures that all experiments will be reviewed to verify that the experiment designs are within the limitations of the RINSC Technical Specifications and 10 CFR 50.59 prior to its initial use.
71
4.9 Facility Specific Surveillance Re.i.;.e.n..s 4.9.1 Beryllium Reflectors-Elements Applicability:
This specification applies to the surveillance of the standard and plug type beryllium reflectors.
Objective:
To prevent physical damage to the beryllium reflectors in the core from accumulated neutron flux exposure.
Specification:
4.9.1.1 .The maximum neutron fluence of any beryllium reflector shall be:
.9..1.1.1. Less than or equal to I E 22 n.utro..
/-ef.---and (This is specified in TS 3.9.1) 4.9.1.1.2 1. The fluence shall be determined annually.
4.9.1.2. The beryllium reflectors shall be visually inspected and functionally fit into the core grid box on a rotating basis not to exceed five years such that:
4.9.1.2.1. The surveillance each year shall include at least one fifth of the beryllium reflectors, 4.9.1.2.2. If a beryllium reflector is removed from use and the time since its last surveillance exceeds five years, it shall be visually inspected and functionally fit into the core grid box prior to being placed in use, and 4.9.1.2.3. If damage is discovered, then the surveillance shall be expanded to include all of the beryllium reflectors prior to use, and annually thereafter.
72
Bases:
The neutron fluenee limit is based on an analysis tha was done by the Uniiver-sityý of Misseui Researh H4FIR Reateoer has ntieed the preV-sne of small redaeks at fast fluenees of 1.8 X Iro 22ml and they suggest that 14a vnltie of 1.XK 1012 -..- (>!MeX') eould be used as eonservafive lewer- limit for detefmiining whe
... ...
yensider.ed." * .+1.*
The WN 1, .. W J... -,.+
limit of 1 X .Ii. ation . is . even i........ A o fmenre inserasnive than what this analysiseconsiders beaause s the lNSC limit is not lionitid to fast netron Historically, the total lifetime neutron fluence has increased by less than 1% of the maximum limit per year. Consequently, an annual verification of total fluence isreasonable. Additionally, reflector elements are visually inspected and functionally fit into the core grid box in order to verify that there are no observable fuel defects or swelling. The rotating inspection schedule ensures that all of the reflectors in the core will be inspected at least once every five years. Since core element handling represents one of the highest risk opportunities for mechanically damaging the fuel cladding, this schedule is deemed appropriate, given the limited amount of information that is gained from these inspections. The discovery of a damaged reflector triggers an increase in the inspection schedule to an annual period.
4.9.2 LE-U Fuel Elements Applicability:
This specification applies to the surveillance of the LEU fuel elements.
Objective:
To verify the physical condition of the fuel elements in order to prevent operation with damaged fuel elements.
73
Specification:
4.9.2.1. The fuel elements shall be visually inspected and functionally fit into the core grid box on a rotating basis not to exceed five years such that:
4.9.2.1.1. The surveillance each year shall include at least one fifth of the fuel elements, 4.9.2.1.2. The surveillance each year shall include fuel elements that represent a cross section with respect to bum-up, 4.9.2.1.3. If a fuel element is removed from use and the time since its last surveillance exceeds five years, it shall be visually inspected and functionally fit into the core grid box prior to being placed in use, and 4.9.2.1.4. If damage is detected by Technical Specification 4.3.3 or otherwise discovered, then the surveillance shall be expanded to include all of the fuel elements prior to use, and annually thereafter.
Bases:
RINSC Technical Specification 4.3.3 requires periodic pool water analysis to test for the presence of radioactivity that could potentially indicate a fuel cladding failure. Fuel elements are visually inspected and functionally fit into the core grid box in order to verify that there are no observable fuel defects or swelling. The rotating inspection schedule ensures that all of the fuel elements in the core will be inspected at least once every five years. Since fuel handling represents one of the highest risk opportunities for mechanically damaging the fuel cladding, this schedule is deemed appropriate, given the limited amount of information that is gained from these inspections. The pool water analysis is the most sensitive mechanism for detecting fuel cladding failure. A detected fuel failure triggers an increase in the inspection schedule to an annual period. Fuel inspections include a cross section 74
of elements with respect to bum-up history in order to ensure that each inspection includes high bum-up elements that would be most likely to start to fail over time.
75
5.0 Design Features 5.0 Design Features 5.1 Site and Facility Description The Rhode Island Nuclear Science Center (RINSC) is located on a 3 acre section of a 27 acre auxiliary campus of the University of Rhode Island. The 27 acre site was formerly a military reservation prior to becoming the Bay Campus of the university. The parcel of land is located in the town of Narragansett, Rhode Island, on the west shore of the Narragansett Bay, approximately 22 miles south of Providence, and approximately 6 miles north of the entrance of the bay from the Atlantic Ocean.
The facility is one of a number of buildings located on the Bay Campus of the university. The RINSC facility consists of a reactor room and an office wing with one entrance between them. The reactor room acts as the confinement space. The reactor pool is constructed on top of a military gun pad.
A more detailed description of the site and the facility is located in Chapter 1 of the Rhode Island Nuclear Science Center Safety Analysis Report.
5.2 Reactor Coolant System Description The RINSC reactor is located in a concrete pool that is lined with aluminum. For operations up to 100 kW, natural convection cooling is possible. For operations above 100 kW, forced convection cooling must be applied. In both cases, the coolant is light water provided by the local town water supply.
The primary section of the forced convection cooling system takes water from the pool outlet line, and directs it to a delay tank where its progress through the cooling system is held up for approximately 70 seconds in order to reduce the N-16 concentration in the water. From the delay tank, the forced cooling system is divided into two loops. Each primary loop consists of a pump and a heat exchanger.
For each loop, the water from the delay tank goes through a primary pump, through a primary heat exchanger, and back to the pool via the pool inlet line, where the two loops recombine. The piping for the primary cooling system is aluminum. Nominal 76
temperatures and pressures are less than 130 F and less than 100 psig respectively.
The secondary sides of the primary heat exchangers use city water to remove the heat from the primary sides. For each loop, secondary water from the heat exchanger is circulated to a cooling tower, through the secondary pump, and back to the heat exchanger Both of the cooling towers use air cooling to reduce the temperature of the secondary water.
A more detailed description of the reactor coolant system is located in Chapter 5 of the Rhode Island Nuclear Science Center Safety Analysis Report.
5.3 Reactor Core and Fuel Description The RINSC fuel is MTR plate type fuel that has a nominal enrichment of 19.75% U-235. The chemical composition of the fuel is UJ3 Si 2 . Each fuel assembly consists of 22 fuel plates, bound by side plates that hold the plates evenly spaced apart. At each end of the assembly, the side plates are attached to square end boxes, that are capable of being inserted into a core grid box. The cladding, side plates, and end boxes are aluminum. Each fresh fuel assembly is loaded with 275g U-235 nominal.
The core grid box is consists of a 5 15/16 inch thick grid plate that has a 9 X 7 array of square holes, and a box that has four walls that surround the grid plate in such a way that the plate serves as the bottom of the box with the top end open. The grid box is suspended from the top of the pool by four comer posts that occupy the comer grid spaces. The box is oriented so that the open end faces up toward the top of the pool. The reactor core is configured by inserting fuel elements end boxes into grid spaces, so that each fuel assembly is standing up inside the box.
The standard core consists of 14 assemblies in a 3 X 5 array in the center of the grid box, with the central grid space available as an experimental facility. The remaining grid spaces are either filled with graphite or beryllium reflector assemblies, or incore experimental facilities. A non-standard core configuration with 17 fuel elements is also possible. In this configuration, the standard core configuration has been modified so that the three central reflector assemblies on the thermal column edge of the core are substituted with fuel assemblies.
77
Both core configurations include 4 shim safety control blades, and a regulating rod. The shim safety blades are located between the fuel and the reflector assemblies on both of the edges of the fuel array that consist of 5 assemblies. There are two blades on each side of the fuel. The blades are housed in shrouds that are part of the core grid box. The shrouds ensure that the blades have unfettered movement in and out of the core. The regulating rod is positioned one grid space out from the fuel, along the central axis of the fuel on the thermal column side of the core.
A more detailed description of the reactor core, core parameters, and fuel is located in Chapter 4 of the Rhode Island Nuclear Science Center Safety Analysis Report.
5.4 Fissionable Material Storage Description Irradiated fuel is stored in two types of fuel storage racks in the reactor pool:
Fixed racks that are mounted on the pool wall Moveable racks that rest on the pool floor Each fixed rack has 9 spaces for fuel storage arranged in a linear array. Each moveable rack has 18 spaces for fuel storage arranged in a 9 X 2 array.
Non-irradiated fuel is typically stored in the RINSC fuel safe.
Non-fuel fissionable materials are either kept where they are in use, or are stored in the reactor pool or fuel safe depending on size constraints and what is most reasonable from an ALARA standpoint, A more detailed description of the fuel storage racks is located in Chapter 9 of the Rhode Island Nuclear Science Center Safety Analysis Report.
Applicability:
This specification applies to the fissionable material storage facilities used for storing materials while they are not in use, or in an approved shipping container.
78
Objective:
The objective of this specification is to ensure that it is impossible for fissionable material to achieve a critical configuration.
Specification:
5.4.1. Fissionable material that is not in use or not in an approved shipping container shall be in storage.
5.4.2. Fissionable material storage facilities shall have keff <= 0.9, for all conditions of moderation and reflection using light water.
Bases:
These specifications conform to ANSI 15.1. They ensure that fissionable material that is not in use will remain in a configuration that cannot achieve criticality.
79
6.0 Administrative Controls 6.1 Organization 6.1.1 Organization Structure The Rhede island Nuelear Seience Center- (RINSC) Reactor- is li.ensed to the Sae of Rh.de island. The Gove..rn of the state appoints five individuals to the RhCde island Attmic Energy Cemmission (RIAEC) who have the ultiatye rfespoesibility foer the fCoility lioense. The Rhode Island Atomic Energy Commission seleCts a Directorewh rsnble is for the everall operation and administraion of the facility. The Assistant Diretor-s for Raation Operations, for Radiation and Reactor Safey are responsible foe the day to day operation ofethe facilty. The RINSC staff operates and maintains the facility. The Nuclear and Radiation Safety Commitee (NRSC) is an independent reviewcommitteeand audit whose non RNSC mrembers are selected by the PIAEC Commissioeries. Figur-e 1 shows the or-ganizationt chart.
The Rhode Island Nuclear Science Center (RINSC) Reactor is licensed to the State of Rhode Island. The Rhode Island Atomic Energy Commission shall have responsibility for the safe operation of the reactor. The Governor of the state appoints five Commissioners to the Rhode Island Atomic Energy Commission (RIAEC) who have the authority to select a Director, and appoint individuals to the Nuclear and Radiation Safety Committee (NRSC). The Director is the organizational head, and is responsible for the reactor facility license. The Assistant Director for Operations is responsible for the reactor programs and operation of the facility. The Assistant Director for Radiation and Reactor Safety is responsible for the safety programs of the facility. The RINSC staff operates and maintains the facility. The Nucl~ear and R-ad;iatin Safety Committee (NRSC) is an independent review and audit committee. Figure 1 shows the organization chart.
80
Nuclear and Radiation I Safety Committee Figure 1 - Organization Chart 6.1.2 Responsibility 6.1.2.1 Rhode Island Atomic Energy Commission (RIAEC)
The Rhode Island Atomic Energy Commission serves as the liaison between the State of Rhode Island, and the federal regulating authority. RIAEC has the ultimate responsibility for the RINSC Reactor license,-and-sets-th p*liy for- the.r.g....nain. The RIAEC Commissioners set the policy for the organization.
6.1.2.2 Director The Director represents the RIAEC, and is responsible for developing and directing all of the administrative and technical programs that support RIAEC policy. The Director is responsible for ensuring facility compliance with federal and state licenses and regulations, and for all 81
activities in the reactor facility which may affect reactor operations or involve radiation hazards.
6.1.2.3 Assistant Director for Operations The Assistant Director for Operations is responsible for implementing the operations programs and managing the day to day operation of the RINSC facility. The Assistant Director ensures that operation of the reactor is compliant with the provisions of the RINSC License and Technical Specifications.
6.1.2.4 Assistant Director for Reactor and Radiation Safety The Assistant Director for Reactor and Radiation Safety is responsible for implementing and managing the Radiation Safety Program. The Assistant Director ensures that that the public and facility personnel are safeguarded from undue exposure to radiation, and that the facility is compliant with federal and state radiation safety regulation.
6.1.2.5 Senior Reactor Operators The Senior Reactor Operator on duty during reactor operations is responsible for directing the licensed activities of Reactor Operators. The Senior Reactor Operator ensures that the operability of the reactor is compliant with the RINSC License and Technical Specifications during operation, and that any experiments performed during operation have been reviewed and approved by the NRSC, and are installed in accordance with any limitations prescribed by NRSC. The Senior Reactor Operator also ensures that experimenters follow facility procedures.
6.1.2.6 Reactor Operators The Reactor Operator on duty during reactor operations is responsible for manipulating the controls of the reactor.
The Reactor Operator directs the actions of Reactor Operator Trainees, and ensures that the reactor is operated within the limits of the RINSC Technical Specifications.
82
6.1.3 Staffing 6.1.3.1 Minimum Staffing Requirements 6.1.3.1.1 The minimum staffing requirements when the reactor is not secured but all of the shim safety control rods are fully inserted into the core shall be a Reactor Operator in the control room or at the pool top.
6.1.3.1.2 The minimum staffing requirements when all of the shim safety rods are not fully inserted into the shall be two individuals present in the facility:
6.1.3.1.2.1 A Reactor Operator in the control room, and 6.1.3.1.2.2 A second individual present in the facility that is capable of scramming the reactor, initiating a facility evacuation, and notifying RINSC staff members and appropriate response agencies.
6.1.3.1.3 If the Senior Reactor Operator on duty is not serving as the Reactor Operator or the second individual present in the facility, they shall be readily available on call.
6.1.3.2 A Senior Reactor Operator shall be present in the facility during any of the following operations:
6.1.3.2.1 The initial reactor start-up and approach to power for the day, 6.1.3.2.2 Fuel element, reflector element, or control rod core position changes, 6.1.3.2.3 Experiment installation or removal for experiments that have a reactivity worth greater than 0.75 %dK/K, and 6.1.3.2.4 Recovery from an unscheduled significant reduction in power, and 6.1.3.2.5 Recovery from an unscheduled shutdown.
83
6.1.3.3 Staff Contact List 6.1.3.3.1 A staff contact list shall be available in the control room for use by the Reactor Operator.
6.1.4 Selection and Training of Personnel 6.1.4.1 Qualification 6.1.4.1.1 Rhode Island Atomic Energy Commissioners The RIAEC Commissioners shall be aware of the general operational and emergency aspects of the reactor facility.
6.1.4.1.2 Director At the time of the appointment to the position, the Director shall have a minimum of six years of nuclear experience. The individual shall have a Bachelor of Science degree or higher in an engineering or scientific field. The degree may fulfill up to four years of the six years of nuclear experience required.
6.1.4.1.3 Assistant Director for Operations At the time of the appointment to the position, the Assistant Director shall have a minimum of three years of nuclear experience. The individual shall have a Bachelor of Science degree or higher in an engineering or scientific field. The degree may filfill lp to two years of the three years of nuclear experience required.
6.1.4.1.4 Assistant Director for Reactor and Radiation Safety At the time of the appointment to the position, the Assistant Director shall have a minimum of three years of health physics experience. The individual shall have a Bachelor of Science degree or higher in an engineering or scientific field. The degree may fulfill up to two years of the three years of nuclear experience required.
84
6.1.4.1.5 Senior Reactor Operators Senior Reactor Operators shall be licensed pursuant to 10 CFR 55.
6.1.4.1.6 Reactor Operators Reactor Operators shall be licensed pursuant to 10 CFR 55.
6.1.4.2 Initial Training and Licensing Personnel that require a Reactor Operator or Senior Reactor Operator license shall be trained in accordance with the facility Operator Training Program.
6.1.4.3 Re-Qualification and Re-Licensing As a condition of maintaining their operating licenses, Reactor and Senior Reactor Operators shall participate in the facility Operator Re-Licensing Program.
6.1.4.4 Medical Certification Facility senior management shall certify that the health of each Reactor Operator and Senior Reactor Operator is such that they will be able to perform their assigned duties. This certification shall be maintained in accordance with 10 CFR 55.21.
6.2 Review and Audit 6.2.1 Nuclear and Radiation Safety Committee (NRSC) Composition and Qualifications 6.2.1.1 Composition The NRSC shall be comprised of a minimum of five seven individuals:
6.2.1.1.1 The Director 6.2.1.1.4- 2 The Assistant Director for Operations 85
6.2.1.1.2 3 The Assistant Director for Reactor and Radiation Safety 6.2.1.1.3 4 Thfee Four members that are not frem euside-t RIAEC commissioners or and DRIN8G staff 6.2.1.2 Qualification The collective qualification of the NRSC members shall represent a broad spectrum of expertise in science and engineering.
6.2.1.3 Alternates Qualified alternates may serve in the absence of regular members.
6.2.2 Nuclear and Radiation Safety Committee Charter The NRSC shall have a written Charter that specifies:
6.2.2.1 Meeting frequency of not less than once per year; and.
6.2.2.2 Quorum in which the member-s that ere RIAEC er RWiSC staff member-s do net eonstitute the 6.2.2.2 Quorum shall consist of a minimum of seven (7) members, including the Assistant Director for Radiation and Reactor Safety, and the Director or Assistant Director for Operations.
6.2.2.3 NRSC Minutes shall be reviewed and approved at the next committee meeting.
6.2.2.4 If deficiencies that affect reactor safety are found, a written report shall be submitted to the RIAEC Commissioners within three months after the NRSC has completed its audit.
86
6.2.3 Review Function The NRSC shall review the following items:
6.2.3.1 Proposed changes to the Technical Specifications or License, and violations of the Technical Specifications or License, 6.2.3.2 Proposed changes to the NRSC Charter, 6.2.3.3 Proposed changes in reactor safety related instrumentation or systems that have safety significance, and 10 CFR 50.59 evaluations, 6.2.3.4 New procedures, major changes to procedures that have safety significance, and violations of procedures that have safety significance, 6.2.3.5 New experiments, 6.2.3.6 Operating abnormalities that have a safety significance, and 6.2.3.7 Reportable occurrences.
6.2.4 Audit Function The non-RIAEC staff members of the NRSC shall audit the following items:
6.2.4.1 Reactor operations shall be audited at least annually to verify that the facility is operated in a manner consistent with public safety and within the terms of the facility license, 6.2.4.2 The Operator Re-Qualification Program shall be audited at least biennially, 6.2.4.3 The Emergency Plan and Emergency Plan Implementing Procedures shall be audited at least biennially, 6.2.4.4 Actions taken to correct any deficiencies found in the facility equipment, systems, structures, or methods of operation that could affect reactor safety shall be audited at least annually, and 87
6.2.4.5 The Radiation Safety Program shall be audited at least annually.
6.3 Radiation Safety The facility shall have a qualified, designated individual that is responsible for implementing the Radiation Safety Program. The Assistant Director for Reactor and Radiation Safety is the individual in the organization that fulfills this requirement. A qualified alternative may serve in this capacity if the Assistant Director is unavailable for an extended period of time.
6.4 Procedures 6.4.1 Written procedures shall be adequate to assure the safe operation of the reactor, but should not preclude the use of independent judgment and action should the situation require such.
6.4.2 The procedures for the following activities shall be reviewed and approved by the NRSC:
6.4.2.1 Startup, operation, and shutdown of the reactor, 6.4.2.2 Fuel loading, unloading, and movement within the reactor, 6.4.2.3 Maintenance of major components of systems that could have an effect on reactor safety, 6.4.2.4 Surveillance checks, calibrations, and inspections that are required by the Technical Specifications, or have a significant effect on reactor safety, 6.4.2.5 Radiation control, 6.4.2.6 Administrative controls for operations, and maintenance, and experiments that could affect reactor safety or core reactivity, 6.4.2.7 Implementation of Feqttife the Emergency and Security plans, and.
6.4.2.8 Receipt, use, and transfer of byproduct material.
88
6.4.3 Substantive changes to the above procedures shall be made only with the approval of the NRSC. Temporary changes to the procedures that do not change their original intent may be made by a Senior Operator. Temporary changes to procedures shall be documented and subsequently reviewed by the NRSC.
6.5 Experiments Review and Approval 6.5.1 All new experiments shall be reviewed and approved by the NRSC prior to bringing the reactor to power with the experiment loaded.
6.5.2 Substantive changes to previously approved experiments shall be reviewed and approved by the NRSC prior to bringing the reactor to power with the experiment loaded.
6.5.3 Minor changes that do not significantly alter the experiment may be approved by a Senior Reactor Operator or upper management.
6.6 Required Actions 6.6.1 Action to be Taken in the Event of a Safety Limit Violation 6.6.1.1 The reactor shall be shut down and reactor operations shall not be resumed until authorization is obtained from the NRC.
6.6.1.2 Immediate notification shall be made to the Director and to the NRSC members.
6.6.1.3 Notification shall be made to the NRC in accordance with paragraph 6.7.2 of these specifications.
6.6.1.4 A safety limit violation report shall be prepared. The report shall include:
6.6.1.4.1 A complete analysis of the causes of the event, 6.6.1.4.2 The extent of possible damage to facility components, systems, or structures 6.6.1.4.3 Corrective actions taken to prevent or reduce the probability of recurrence.
6.6.1.5 The safety limit violation report shall be submitted to the NRSC for review and appropriate action.
89
6.6.1.6 The safety limit violation report shall be submitted to the NRC in accordance with Paragraph 6.7.2 of these specifications in support of a request for authorization to resume reactor operations.
6.6.2 Action to be Taken in the Event of a Reportable Occurrence Other Than a Safety Limit Violation 6.6.2.1 The Senior Reactor Operator shall be notified promptly and corrective action shall be taken immediately to place the facility in a safe condition until the cause of the reportable occurrence is determined and corrected.
6.6.2.3 2 The occurrence shall be reported to the Director or Assistant Director.
6.6.2.13 If the reactor is shutdown, operations shall not be resumed without authorization from the Senior- Rctc Opelatef Director or Assistant Director for Operations.
6.6.2.4 The occurrence, and corrective action taken shall be reviewed by the NRSC during its next scheduled meeting.
6.6.2.5 Notification shall be made to the NRC in accordance with Paragraph 6.7.2 of these specifications.
6.7 Reports 6.7.1 Annual Report A written report shall be submitted annually to the NRC following the 30th of June of each year, and shall include the following information:
6.7.1.1 A summary of the number of hours that the reactor was critical for the period, the energy produced for the period, and the cumulative total energy output since initial criticality; 6.7.1.2 A summary of the unscheduled shutdowns that occurred during the period, the causes of the shutdowns, and if applicable, corrective action taken to preclude recurrence; 90
6.7.1.3 A summary of any major maintenance performed during the period that has safety significance, and the reasons for any corrective maintenance required; 6.7.1.4 A summary of 10 CFR 50.59 safety evaluations made during the reporting period; 6.7.1.5 A summary of the amount of radioactive effluents, and to the extent possible, an estimate of the individual radionuclides that have been released or discharged to the environs outside the facility as measured at or prior to the point of release.
If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed, a statement to this effect is sufficient for the summary.
6.7.1.6 A summary of the results of environmental surveys performed outside the facility during the reporting period; and 6.7.1.7 A summary of annual radiation exposures in excess of 500 mrem received by facility personnel or visitors.
6.7.2 Special Reports 6.7.2.1 Reporting Requirements for Reportable Occurrences In the event of a reportable occurrence, the following notifications shall be made:
6.7.2.1.1 Within one working day after the occurrence has been discovered, the NRC Headquarters Operation Center shall be notified by telephone at the number listed in 10 CFR 20 Appendix D, and 6.7.2.1.2 Within 14 days after the occurrence has been discovered, a written report that describes the circumstances of the event shall be sent to the NRC Document Control Desk at the address listed in 10 CFR 50.4.
91
b 6 6.7.2.2 Reporting Requirements for Unusual Events Within 30 days following an unusual event, a written report that describes the circumstances of the event shall be sent to the NRC Document Control Desk at the address listed in 10 CFR 50.4.
6.8 Records 6.8.1 Records to be retained for a period of at least five years 6.8.1.1 Reactor operating records, 6.8.1.2 Principal maintenance activities, 6.8.1.3 Surveillance activities required by the Technical Specifications, 6.8.1.4 Facility radiation monitoring surveys, 6.8.1.5 Experiments performed with the reactor, 6.8.1.6 Fuel inventories and transfers, 6.8.1.7 Changes to procedures, and 6.8.1.8 NRSC meeting minutes, including audit findings.
6.8.2 Records to be retained for a period of at least one certification cycle Current Reactor Operator re-qualification records shall be maintained for each individual licensed to operate the reactor until their license is terminated.
6.8.3 Records to be retained for the life of the facility 6.8.3.1 Gaseous and liquid radioactive effluents released to the environs, 6.8.3.2 Off-site environmental monitoring surveys, 6.8.3.3 Personnel radiation exposures, and 6.8.3.4 Drawings of the reactor facility, and 92
6.8.3.5 Reportable occurrences.
93
James E.Matos, PhD Senior Physicist NATIONAL ANuclear LABORATORY Engineering Division HEU Minimization Methods Section Argonne National Laboratory 9700 South Cass Avenue, Bldg. 208 Argonne, IL60439 1-630-252-6758 phone 1-630-252-5161 fax 1-708-262-2055 mobile February 27, 2013 jim.matos@anl.gov Mr. Jeff Davis Acting Director Rhode Island Nuclear Science Center 16 Reactor Road Narragansett, Rhode Island 02882
Subject:
Transmittal of Computer Code Manuals and InpurFiles for Each of the Computer Codes Used in the RAI Analyses for RINSC Operating License Renewal: 2010 - 2013 Enclosed are two DVDs with the manuals for the computer codes and the computer code input files that were used in the RINSC RAI analyses.
We followed the document that listed computer codes and RAI numbers that you sent to me. The only change is that I corrected a typo under PLTEMP from 4.14 to 4.15.
The DVDs contain:
- 1. JDavis Computer Codes Used in Relicensing
- 2. Overview of RERTR Computer Codes Used. More detailed information can be found in each of the manuals.
- 3. RINSC RAI Questions & Answers. This is the file you sent to me, included on the DVD for convenience.
- 4. COMPUTER CODES USED folder with computer code manuals for:
PARET-ANL 7.5 PLTEMP-ANL 4.0 REBUS 1.5 (DIF3D)
VARI3D 1.42 WIMS-ANL 5.08 RELAP5 Mod 3.3 is an NRC code which we are not authorized to distribute. NRC approval is required.
- 5. RINSC INPUT FILES folder containing all of the computer input files for each of the computer codes that were used in the analyses.
J.E. Matos Nuclear Engineering Division GTRI Convert Program HEU Minimization Methods Section Argonne National Laboratory