ML20196C705
| ML20196C705 | |
| Person / Time | |
|---|---|
| Site: | 05000294 |
| Issue date: | 08/30/1984 |
| From: | MICHIGAN STATE UNIV., EAST LANSING, MI |
| To: | |
| Shared Package | |
| ML20196A128 | List: |
| References | |
| NUDOCS 8812080067 | |
| Download: ML20196C705 (77) | |
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.3 JUSTIFICATION FOR REVISION TO:
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TECHNICAL SPECIFICATIONS i
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MICHIGAN STATE UNIVERSITY j
j TRIGA REACTOR l
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REVISION 2 AUGUST 1984 f
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LICENSE NO. R-114 DOCKET NO. 50-294 i
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8812080067 881201 j
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FORMARD The following is a reproduction of the current Technical Specifications for the HSU TRIGA Reactor Operating License (R-114).
Each section of this document has been evaluated to determine its applicability to the "Possession Only" Status versus the current "Operational i
Status".
The results of this evaluation have been categorized into (3) specific actions:
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Delete the section in its entirety; (2)
Retain the section (unchanged); or (3)
Revise the section as necessary for the "Possession Only" Status A description of the proposed action and the justification for such action has O
been inserted in bold face type at the ceginning o each section.
Those sections that have been retained and/or revised have been assembled to produce Revision 3 to the technical specifications.
The completed document (Technical Specifications, Revision 3 - attached) should become Appendix A to the license as amended to "Possession Only" Status.
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SUMMARY
OF AFFECTED TECHNICAL SPECIFICATIONS MICHIGAN STATE UNIVERSITY TRIGA REACTOR SECTION STATUS
- 1.0 Definitions R
2.0 Safety Limits & Limiting Safety System Settings 2.1 Safety Limit Fuel Element Temperature D
2.2 Limiting Safety System Setting D
3.0 Limiting Conditions for Operation 3.1 Non-Pulsing Operation D
3.2 Reactivity Limitations D
4 3.3 Pulse Mode Operation D
3.4 Control and Safety System 0
3.5 Radiation Monitoring System 0
3.6 Argon 41 Discharge Limit D
3.7 Engineered Safety Feature-Ventilation System D
3.8 Limitations on Experiments D
3.9 Irradiations D
I 4.0 Surveillance Requirenents 4.1 General R
4.2 Safety Limit - Fuel Element Temperature D
4.3 Limiting Conditions for Operation D
4.3.1 Reactivity Requiremerts D
4.3.2 Control and Safety System D
4.3.3 Radiation Monitoring System R
4.3.4 Ventilation System R
4.3.5 Experimentation and D
Irradiation Limits 4.4 Reactor Fuel Elements D
5.0 Design Features 5.1 Reactor Fuel D
l 5.2 Reactor Core (Fuel Configuration)
D 5.3 Control Rods D
5.4 Radiation Monitoring System R
5.5 Fuel Storage D
5.6 Reactor Building & Ventilation System N
5.7 Reactor Pocl Water System R
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SECTION STATUS
- 6.0 Administrative Controls 6.1 Organization R
6.2 Review and Audit R
6.3 Action to be Taken in the Event a Safety D
Limit is Exceeded 6.4 Action to be Taken in the Event of a D
Reportable Occurrence 6.5 Operating Procedures D
6.6 Facility Operating Records R
6.7 Reporting Requirements R
- N - There is no change being made to this section.
- R - These specifications are being revised to reflect the possession of the residual byproduct material only.
- D - Reactor operation is no longer possible since the fuel has been shipped off-site; therefore, limits and/or operations pertaining to reactor safety are no longer applicable.
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TABLE OF CONTENTS t
(s Page Introduction.......................................................
1-1 1.0 Definitions.................................................
1-1 2.0 Safety Lir.its and Limiting Safety System Settings...........
2-1 2.1 Safety Limi t Fuel El ement Temperature................
2-1 2.2 Limiting Safety System Setting.......................
2-2 3.0 Limiting Conditions for Operation...........................
3-1 3.1 Nonpulsing...........................................
3-1 3.2 Reactivity Limitations...............................
3-2 3.3 Pulse Mode Operations................................
3-3 3.4 Control and Safety Systems...........................
3-4 3.J Radiation Monitoring Systems.........................
3-10 3.6 Argon-41 Discharge Limit.............................
3-11
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3.7 Engineered Safety Feature-Ventilation System.........
3-12 3.8 Limitations on Experiments...........................
3-14 3.9 Irradiations.........................................
3-18 4.0 Surveillance Requirements..,...............................
4-1 4.1 Genera 1..............................................
4-2 4.2 Safety Limi t Fuel Temperature........................
4-3 4.3 Limiting Conditions for Operation...................
4-3 4.4 Reac tor fuel El ement s................................
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J 5.0 Design Features.............................................
5-1 5.1 Reactor Fue1.........................................
5-1 5.2 Reactor Core.........................................
5-2 5.3 Control Rods.........................................
5-4 5.4 Radiation Monitoring System..........................
5-5 5.5 Fuel Storage.........................................
5-7 5.6 Reactor Building and Ventilation System...............
5-8 5.7 Reactor Pool Hater Systems...........................
5-9 6.0 Administrative Controls.....................................
6-1 6.1 Organization.........................................
6-1 6.2 Review and Audit.....................................
6-1 6.3 Action Taken in the Event a Safety Limit is Exceeded.................................
6-4 6.4 Action to be Taken in the Event of a Reportable Occurrence.............................
6-5 6.5 Operating Procedure..................................
6-6 6.6 Facility Operating Records...........................
6-7 6.7 Reporting Requirements...............................
6-8
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i Included in this document are the Technical Specifications and the "Base;" for the Technical Specifications.
These bases, which provide the technical support for the individual technical specifications, are included for information purposes only.
They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.
The dimensions, measurements and other numerical values given in these specifications may differ from measured values owing to normal construction and manufacturing tolerances, or normal accuracy of instrumentation.
1.0 DEFINITIONS Definitions that pertain only to reactor operations have been deleted since the fuel has been shipped off-site (in preparation for decommissioning) and reactor operations are no longer possible. All other definitions have been either retained "as is" or revised as necessary to relate directly to the "Possession Only" conditions.
REACTOR OPERATING CONDITIONS - Revised 1.1 Reactor Shutdown - Deleted The reactor is shut down when the reactor is suberitical by at least 0.7% AK/K or $1.00.
1.2 RP R ior Secured - Deleted The reactor is secured when all' the following conditions are satisfied:
a.
The reactor is shut down; O
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b.
The console key switch is in the "off" position and the key is removed from the console and under the control of a licensed operator or stored in a locked storage area; and c.
No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms, or insertion or withdrawal of in-core experiments.
1.3 Reactor Ooeration - Deleted Reactor operation is any condition wherein the reactor is not secured.
1.4 Cold critical - Deleted The reactor is in the cold critical condition when it is critical with 0
the fuel and bulk water temperatures both below 50 C.
I 1.5 Nonouisino Mode - Deleted Nonpulsing mode operation shall mean operation of the reactor with the mode selector switch in the manual position.
1.6 Pulse Mode - Deleted l
Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.
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1.7 Shutdown Margin - Deleted Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting from any permissible operating conditions and that the reactor will remain suberitical without further operator action.
1.8 Abnormal Occurrence - Retained An "Abnormal Occurrence" is defined for the purposes of the reporting requirements of Section 208 of the Energy Reorganization Act of 1974 (P.L.93-438) as an unscheduled incident or event which the Nuclear Regulatory Commission determines is significant from the standpoint of public health or safety.
1.9 Reoortable Occurrence - Deleted A reportable occurrence is any of the following which occurs during reactor operation:
a.
Operation with any safety system setting less conservative that specified in Section 2.2 Limiting Safety System Settings; l
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Operation in violation of a Limiting Condition for Operation; c.
Failure of a required reactor or experiment safety system component which could render the system incapable of performing its intended safety function; d.
Any unanticipated or uncontrolled change in reactivity greater than one dollar; O
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e.
An observed inadequacy in the implementation of either d
administrative or procedural controls, such that the inadequacy could have caused the existence or develop-ment of a condition which could result in operation of the reactor outside the specified safety limits; and f.
Release of fission products from a fuel element.
REACTOR EXPERIMENTS - This entire sectior. should be deleted.
1.10 Exoeriment - Deleted Any operation, hardware, or target (excluding devices such as detectors, foils, etc.), which is designated to investigate non-routiner reactor characteri: tics or which is intended for irradiation within the pool, on or in a beamport or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design.
The HSU Reactor irradiations includa exposure of samples AV to neutrons and/or gamma radiation in either the rotary specimen rack, central thimble or other experimental assembly.
1.11 Secured Exoeriment - Deleted A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means.
The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experimel.t or by forces which can arise as a result of credible malfunctions.
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1.12 Moveable Exoeriment - Deleted A moveable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
1.13 Exoerimental Facilities - Deleted Experimental facilities shall mean in-core positions including the central thimble, the rotary sample rack, nd in-pool irradiation facilities.
REACTOR COMPONENTS - Revised (See proposed Technical Specifications, Revision 3, "Definitions")
1.14 Shim Rod - Retained A shim rod is a control rod having an electric motor drive and scram capability.
1.15 Safetv-Transient Rod - Retained The safoty-transient rod is a control rod with scram capability that can be rapidly ejected from the reactor core to produce a pulse.
1.16 Reaulatino Rod - Retained l
The regulating rod is a low worth control rod having an electric motor l
drive and scram capability.
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1.17 En1_flement - Retained l
A fuel element is a single TRIGA fuel rod of standard type.
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1.18 Instrumented Element - Retained O.
An instrumented element is a special fuel element in which a sheathed chromel-alumel or equivalent thermocouple is embedded in the fuel at the vertical center plane of the fuel element. More than one thermocouple may be located in each element.
1.19 Standard Core - Deleted A standard core is an arrangement of standard and/or instrumented TRIGA fuel in the reactor grid plate.
(Refer to Sec. 5.1) 1.20 Ooerational Core - Deleted An operational core is a standard core for which the core parameters of shutdown margin, fuel temperature, power calibration, and maximum allowable reactivity insertion have been determined to satisfy the requirements of the Technical Specifications.
REACTOR INSTRUMENTATION - Revised - (See proposed Revision 3. "Definitions")
1.21 Safety Limit - Retained Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
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1.22 Limitina Safetv System Settina - Retained Limiting safety systems setting is the setting for automatic protective devices related to those variables having significant safety functions.
1.23 Ooerable - Retained A tystem, device, or component shall be considered operable when it is capable of performing its intended functions in a normal manner.
1.24 Reactor Safety Systems - Deleted Reactor safety systems are those systems, including their associated input circuits, which are designed to initiate a reactor scram for the primary purpose of protecting the reactor or to provide information which rcquires manual protective action to be initiated.
1.25 Exoeriment Safety Systems - Deleted O
Experiment safety systems are those systems, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated.
1.26 Heasured Value - Retained The measured value is the magnitude of that variable as it appears on the output of a measuring channel.
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1.27 Measurina Channel - Retained A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.
1.28 Safety Channel - Retained A safety channel is a measuring channel in the reactor safety system.
1.29 Channel Check - Retained A channel check is a qualitative verification of acceptable performance by observation of channel behavior.
1.30 Channel Test - Reta)ned A channel test is the introduction of a signal into the channel to verify that it is operable.
1.31 Channel Calibration - Retained A channel calibration consists of comparing a measured value from the measuring channel with a corresponding known value of the parameter so that the measuring channel output can be adjusted to respond with acceptable accuracy to known values of the measured variable.
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,q 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS b
SECTION 2.1, "SAFETY LIMIT-FUEL ELEMENT TEMPERATURE" - Deleted This limit is no longer applicable since all reactor fuel has been removed from the site in preparation for decommissioning.
2.1 Safety Limit-Fuel Element Temoerature Aeolicability This specification applies to the temperature of the reactor fuel.
Objective The objective is to define the maximum fuel element temperature that can be permitted with confidente that no damage to the fuel element cladding will result.
Soecifications The temperature in a standard TRIGA fuel element (Refer to Sec. 5.1) 0 shall not exceed 1000 C under any conditio v, of operation.
Halli The important parameter for a TRIGA reactor is the fuel element temperature.
This parameter is well suited as a single specification especially since it can be measured. A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit.
The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel moderator.
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The magnitude of this pressure is determined by the fuel-moderator C) temperature and the ratio of hydrogen to zirconium in the alloy.
The safety limit for the standard TRIGA fuel is bastd on data, including the large mass of experimental evidence, obtained during high performance reactor tests on this fuel.
These data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided 0
0 that the temperature of the fuel does not exceed 1830 F (1000 C)
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and the fuel cladding is water cooled.
SECTION 2.2, "LIMI!!NG SAFETY SYSTEM SETTINGS" (SCRANSETTINGS)-Delete $
These settings pertain to Reactor Operation which is no longer possible since the fuel has been shipped off-site.
2.2 LIMITING SAFETY SYSTEM SETTINGS ALo11cability O
This specification applies to the scram settings which prevent the safety limit from being reached.
Obiective The objective is to prevent the safety limits from being reached.
Soecifications 0
The limiting safety system setting shall be 450 C as measured in an instrumented fuel element relative to the ambient temperature.
Instrument element shall be located in the B or C ring of the core configuration.
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D oU The limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit 0
from being exceedd. A setting of 450 C provides a safety margin of 0
500 C for standard TRIGA fuel elements. A part of the safety margin is used to account for the difference between the true and measured temperatures resulting from the actual location of the thermocouple.
If the thermocouple element is located in the hottest position in the core, the difference between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the fuel and close to the anticipated hot spot.
If the thermocouple element is located in a region of lower temperature, such as on the periphery of the core, the measured temperature will differ by a greater amount from that actually occurring at the core hot spot.
Calculations indicate that, for this case, the true temperature at the 0
hottest location in the core would be no greater than 900 C providing 0
a margin to the safety limit of at least 100 C for standard fuel elements.
This margin is ample to account for the remaining uncertainty in the accuracy of the fuel temperature measurement channel and any overshoot in reactor power resulting from a rector transient during nonpulsing mode operation.In the pulse mode of operation, the same limiting safety sysum setting will apply. However, the temperature channe! will nave no effect on limiting the peak power generated beccuse of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).
In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting of the "tail" of the energy transient in the event the pulse rod remains stuck in the fully withdrawn position.
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3.0 LIMITING CONDITIONS FOR OPERATION SECTION 3.1, "NONPULSING OPERATION" - Deleted It is impossible for the Reactor to go critical since the fuel is no longer on site.
3.1 Nonouisina Ooeration Aeolicability This specification applies to the energy generated in the reactor during nonpulsing operation.
Obiective This objective is to assure that the reactor does not go critical.
$seci fications The reactor core is to be maintained with a Keff less than.996 at all times.
4 Bases This specification is designed to permit control of the reactor in the fulfillment of other Technical Specification requirements e.g., section 4.3.2.a without exceeding the limitation of Section 3.2.
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SECTION 3.2, "(EACTOR LIMITATIONS" - Deleted The Reactor has been shutdown permanently in preparation for decosals-sioning; therefore, limitations pertaining to reactivity worth or fuel temperature are no longer applicable.
3.2 Reactivity Limitations Acolicability These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.
They apply for all modes of operations.
Obiective The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit will not be exceeded.
Soecifications a.
The reactor shall not be operated unless the shutdown margin provided by control rods shall be greater than 0.47. AX/K with:
(1) the highest worth non-secured experiment in its most reactive state.
(2) the highest worth control rod fully withdrawn; and (3) the reactor in the cold critical condition without Xenon.
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The excess reactivity above cold critical, without Xenon, shall not exceed 2.25% AK/K with experiments in place.
c.
The maximum rate of reactivity insertion associated with movement of a standard control rod shall be no greater than 0.2%
AK/K/sec.
AA111 a.
The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth
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control rod should remain in the fully withdrawn position.
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The value for maximum excess reactivity provides an adequate margin for experiment insertion while minimizing the possibility l
of exceeding the safety limits.
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The limit on maximum rate of reactivity insertion assures that f
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achieving super-criticality is dependent upon prompt ar.d delayed i
neutrons rather than prompt neutrons alone.
l SECTION 3.3, "PUl.SE MODE OPERATION" - Deleted i
The Reactor cannot be operated since the fuel is no lo y er on-site.
AI) specifications pertaining to Reactor Operation should be deleted.
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result of a pulse insertion of reactivity.
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Obiective The objective is to assure that the reactor is not pulsed.
b soecification The reactor will not be operated in a pulsed mode.
Bases
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This section would ordinarily be removed since operation of the reactor in any mode is not being planned.
It has been revised to maintain this section available and not to change other section nur.bers i. this f
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3.4 control and safety Systim SECTION 3.4.1, "SCRAM TIME" - Deleted t
i Specifications pertaining to the time requirements for prompt shutdown (scram) of the Reactor are no longer appilcable since the Reactor fuel has been shipped off-site.
3.4.1 itrAm Time
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Aeolicability i
This specification applies to th6 time required for the
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scramable control rods to be fully insertad from the instant I
that a safety channel variable reaches the Safety System Setting.
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I 4 a Obiective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.
Soecification The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the control rod reaches its fully inserted pcsition shall not exceed 2 j
seconds for the pulse (transient) rod and I second for the regular and shim rods.
Bases This specification assures that the reactor will be promptly i
shut down when a scram signal is initiated.
Experience and
'v analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified cram time is adequate to assure the safety of the reactor.
l SECTION 3.4.2, "REACTOR CONTROL SYS1EM" - Deleted t
This section applies to Reactor Operation which is impossible since the Reactor fuel is no longer or, site.
3.4.2 Reactor Control Syltas Anolicab111tv This specification appIles to the channel; monitoring the reactor core, which must provide ir. formation to the reactor j
operator during reactor operation.
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I Obiective The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.
I Soecifications The reactor shall not be operated unless the measuring channels listed in the following table are operable.
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Effective Mode l
i littigf na Channel Ooerable N.P.
Plusino Fuel Elenent Temperature 1
X X
Linear Power Level I
X
% Power Level 1
X Integrated Pulse Power 1
X W
Fuel temperature displayed at the control console gives continuous information on this parameter which has a specified safety limit. The power level monitors assure that the reactor f
power level is adequately monitored for both nonpulsing and pulsing modes of operation.
The specifications on reactor poor level indication are included in this section since the power level is related to the fuel temperature.
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O SECTION 3.4.3, "REACTOR SAFETY SYSTEM" - Deleted These specifications pertain to the various Reactor safety system channels which were necessary for reactor operations. Since the fuel has been shipped off-site for decommissioning, the reactor can no Ionger be operated.
3.4.3 Reactor Safety System Aonlicability This specification applies to the reactor safety system channels.
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Minimum Reactor Safety Channels Number Effective Mode Safety Channel Onorable Function N.P.
Pulse Fuel Element Temperature 1 SCRAM f LSSS X
X Linear (Power Level) 1 SCRAM f 110%
X of scale
% of Pcwer 1
SCRAM f 110%
X of full power Console Scram Bar 1
SCRAM X
X Detector Power SCRAM on loss of Supply (High Voltage)
I supply voltage X
Preset Timer 1
O 15 seconds or less after pulse Shim and Regulating Rod Position 1
Prevent withdrawal X
Start-up Channel 1
Prevent shim or X
regulating rod withdrawal with less than 2 neutron induced counts per second Shim and Regulating 1
Prevent simultaneous X
Rod Controls I
withdrawal NV/NVT SCRAM 1
Prevent excessive power X
during a pulse Pulse Rod Interlock 1
Prevent withdrawal X
of pulse rod when shim and/or regulating rod are off the bottom O
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O Obiective The objective is to specify the minimum number of reactor safety system channels that must be operable for safe operation.
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Saltifications The reactor shall not be operated unless the safety channels described J
in Table 1 are operable.
Balti The fuel temperature and power level scrams provide protection to assure that the reactor can be shut down before the safety limit on the 1
fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system if an unsafe or abnormal condition j
occurs.
In the event of failure of the power supply for the safety chambers, operation of the reactor without adequate instrumentation is prevented.
The preset timer assures the the reactor power level will reduce to a low level after pulsing, j
The interlock to prevent startup of the reactor at neutron count rates l
1ess than 2 cps, which corresponds to approximately 2.5 x 10-4 watts, assures that sufficient neutrons are available for proper startup.
The interlock to prevent withdrawal of the shim or regulating rod in j
the pulse mode is to prevent the reactor from being pulsed while on a positive period.
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i 0O SECTION 3.5, "RADIATION MONITORING SYSTEM - Deleted It is no longer possible to operate the Reactor since the Reactor fuel has been shipped off-site. Specifications pertaining to the Radiation Monitoring System for "Possession Only" status have been included in Section 4.0 "Surveillance Requirements" of proposed Revision 3 to the Technical Specifications.
3.5 Radiation Monitorina System i
Aeolicability This specification applies to the radi:. tion monitoring information which must be available to the reactor operator during reactor operation.
Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.
I Soecifications The reactor shall not be operated unless the radiation monitoring l
channels listed in the following table are operable.
I Radiation Monitoring Channels Function No.
Area Radiation Monitor Monitor radiation levels I
within the reactor room l
Continuous Air Radiation Monitor radiation levels 1
l Monitor within the reactor room and in exhaust stream 2600J l
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O The radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take necessary steps to prevent the spread of radioactivity to the surroundings.
I For periods of time for maintenance to the radiation monitoring channels, the intent of this specification will be satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be kept under visual observation.
SECTION 3.6, "ARGON-41 DISCHARGE LIMIT" - Deleted l
At the Reactor Facility, Argon-41 is produced only by Reactor Operations (Halflife =1.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br />). Since the fuel has been shipped off-site and Reactor Operation is no longer rassible, there will be no discharge (s) of Argon-41 free the reactor facility.
3.6
&tgnD-41 Discharge Lirqtt Aeolicability This specification applies to the concentration of Argon-41 that may be discharged from the TRIGA reactor facility, j
Qbjective i
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To ensure that the health and safety of the public is not endangered by the discharge of Argon-41 from the TRIGA reactor facility.
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O Snecifications The concentration of Argon-41 in the effluent gas from the facility as diluted by atmospheric air in the lee of the facility due to the turbulent wake effect shall not exceed 4.8 x 10-8 pCi/ml averaged over one year.
The maximum allowable concentration of Argon-41 in air in unrestricted areas as specified in Appendix B. Table II of 10CFR 20 is 4.8 x 10-8 pCi/ml.
l SECT.10N 3.7 "ENGINEERED SAFETY FEATURE - VENTILATION SYSTEM" - Deleted The Reactor will (can) not be operated since the fuel has been shipped off-site. Therefore, this specification is no longer applicable.
Specifications for the Ventilation System during "Possession Only" I
status will be contained in Section 4.0 "Surveillance Requirements" of l
proposed Revision 3 to these Technical Specifications.
3.7 Enaineered Safety Feature - h ntilation System ADolicability r
I This specification applies to the operation of the facility ventilation i
system.
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O Objective i
The objective is to assure that the ventilatior, system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation.
Soecifications The reactor shall not be operated unless the facility ventilation system is operable except for periods of time necessary to permit repair of the system.
In the event of a substantial release of airborne radioactivity, the ventilation system will be secured automatically by a signal from an exhaust air radiation monitor.
BAIM During normal operation of the ventilation system, the concentration of Ar-41 in unrestricted areas is below HPC.
In the event of a clad rupture resulting in a substantial release of airborne particulate radioactivity, the ventilation system will be diverted through an 3
l absolute filter. Moreover, radiation monitors within the laboratory independent of those in the ventilation system will give warning of j
high levels of radiation that might occur during operation with the l
ventilation system secured, i
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G SECTION 3.8, "LIMITATIONS ON EXPERIMENTS" - Deleted The Reactor can no longer be operated since the fuel has been shipped off-site.
Experiments in the Reactor and its experimental facilities will no longer be conducted.
i 3.8 Limitations On Exoeriments 8policability 4
This specification applies to experiments installed in the reactor and its experimental facilities.
Obiective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
Soecifications The reactor shall not be operated unless the following conditions j
governing experiments exist.
a.
Non-secured experiments shall have reactivity worth less than 0.7% AK/K.
b.
The reactivity worth of any single experiment shall be less than 1.4% AK/K.
The total reactivity worth of in-core experiments shall not exceed 2.1% AK/K.
]
2600J 3-14
O c.
Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the limit of Appendix B of 10 CFR Part 20.
d.
In calculations pursuant to (c) above, the following assumptions shall be used:
(1) If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiction level, at least 10% of the gaseous activity or l
aerosols produced will escape.
(2) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99%
I efficiency for 0.3 micron particles, at least 10% of those vapors can escape.
0 l31 For materials whose boiling point is above 130 F and where vapors formed by boiling this material can escape only through an undisturbed column of water above this core, at l
1 east 10% of these vapors can escape.
I 1
O 2-3-15
l l
4 i
!O
)
e.
Each fueled experiment shall be controlled such that the total l
inventory of iodine isotopes 131 through 135 in the experiment f
is no greater than 1.5 millicuries.
f i
f.
If a capsule fails and releases material which could damage the l
reactor fuel or structure by corrosion or other means, removal l
and physical inspection of appropriate core components shall be i
performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Reactor Safety Committee and the Reactor Supervisor or his designated alternate and determined to l
be satisfactory before operation of the reactor is resumed.
I g.
Experiments containing materials corrosive to reactor components, compounds highly reactive with water, or liquid fissionable materials shall be doubly encapsulated.
O h.
Explosive materials such as (but not limited to) dynamite. TNT, nitroglycerine or PETN shall not be irradiated in the reactor or
{
experimental facilities.
[
Bun I
i a.
This specification is intended to provide assurance that the
[
worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were to be suddenly inserted, f
i i
I O
2600J 3-16
O b.
The maximum worth of a single experiment is limited so that its removal from the cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.
[
i c.
This specification is intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of 10 i
CFR Part 20 will be released to the atmosphere outside the l
facility boundary.
i d.
The 1.5 millicurie limitation on todine 131 through 135 assures that in the event of failure of a fueled experiment leading to I
total release of the iodine, the exposure dose at the exhaust vent will be less than that allowed by 10 CFR Part 20 for an J
unrestricted area, e.
Operation of the reactor with the reactor fuel or structure damaged is prohibited to avoid release of fission products.
f.
Double encapsulation minimizes the chance of contaminating 3
irradiation facilities or causing structural damage to the irradiation facilities.
1 g.
Explosive material will not be irradiated in order to prevent the possibility of an explosion which might damage the core 1
components.
l l
O 2600J 3-17
._----_r_.-.
__.y
,,_--,__.,.._7
t i
O SECTION 3.9, "IRRADIATIONS" - Deleted Irradiations can no longer be performed since the Reactor can no longer be gerated due to the fuel being shipped off-site for decommissioning.
3.)
Irradiations
[
i Aeolicability J
i This specification applies to irradiations performed in the irradiation 4
facilities contained in the reactor pool as defined in Secuon 1.10.
Irradiations are a subclass of experiments that fall within the specifications hereinafter stated in this section.
The surveillance L
requirements for irradiations are given in Section 4.3.5.b.
1 Objective l
The objective is to prevent damage to the reactor, excessive release of radioactive materials or excessive personnel radiation exposure during the performance of an irradiation.
Snecifications I
I A device or material shall not be irradiated in an irradiation faellity i
i j
under the classification of an irradiation unless the following l
I conditions exist
- r I
\\
l a.
The irradiation meets all the specifications of Section 3.8 for an experiment, i
l b.
The expected radiation field produced by the device or sample upon removal from the reactor is not more than 10 res/hr at one i
foot, otherwise it shall be classed as an experiment;
,O l
2600J i
3-18 l
P I
c.
The device or material is encapsulated in a suitable container.
l d.
The reactivity worth of the device or material is 0.175% AK/K or less, otherwise it shall be classed as an experiment; and e.
The device or material does not remain in the reactor for a period of over 15 days, otherwise it shall be classed as an experiment.
Bases This specification is intended to provide assurance that the special class of experiments called irradiations will be performed in a manner that will not permit any safety limit to be exceeded.
(
O i
t i
i r
O u>
3-19 l
t I
4.0 SURVEILLANCE REOUIREMENTS SECTION 4.1, "GENERAL" has been revised. It has been included under Section 4.0, "Surveillance Requirements" in proposed Revision 3 to these Technical Specifications.
It will be appropriate to maintain some surveillance requirements under the "Possession Only" status. The objective will be related to the surveillance of the radioactive materials / sources as authorized by the R-114 License (Revised).
4.1 General Aeolicability This specification applies to the surveillance requirements of any system related to reactor safety.
Obiective The objective is to verify the proper operation of any system related to reactor safety.
Soecifications Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the l
specifications to which the systems were originally designed and f
fabricated or to specifications approved by the Reactor Safety Committee. A system shall not be considered operable until after it is l
successfully tested. A licensed reactor operator shall be present j
during maintenance of the reactor control and safety system.
l O
2601J 4-1
~
O This specification relates to changes in reactor systems which could directly affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presently accepted operating criteria.
SECTION 4.2, "SAFETY LIMIT, Furl ELEMENT TEMPERATURE" - Deleted l
This specification is no longer applicable since the Reactor fuel has been shipped off-site.
4.2 Safety li;ait - Fuei Element TemeerA19T.t Aeolicabi111y This specification applies to the surveillance requirements of the fuel element temperature measu-ing channel, Objectiva The objective is to assure that the fuel element temperatures are I
properly monitored.
SapcificatioD1 4.
Whenever a reactor scram caused by high fuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature safety limit was exceeded.
O 2601J 4-2
b.
A Channel Check of the fuel element temperature measuring channel shall be made quarterly whenever tne reactor is operated by recording a measured value of a meaningful temperature indication.
Raiti Operational experience with the TRIGA system gives assurance that the thermocouple measurements of fuel element temperatures have been sufficiently reliable to assure accurate indication of this parameter.
4.3 Limitina Conditions For Ocaration SECTION 4.3.1, "REACTIVITY REQUIREMENTS" - Deleted This specification does not apply, since the fuel has been shipped off-site and thus requirements for reactivity control are no longer necessary.
O 4.3.1 Reactivity Reguirements Anolicability These specifications apply to the surveillance requirements for reactivity control of experiments and systems.
Objective The objective is to measure and verify the worth, performance, and operability of those systems affecting the reactivity of the reactor.
O 2601J 4-3
Snecifications a.
The reactivity worth of an experiment shall be estimated or measur-ed, as appropriate, before reactor operation with said experiment.
b.
The control rods shall be visually inspected for deterioration at intervals not to exceed 2 years.
c.
The transient rod drive cylinder and associated air supply system shall be inspected, cleaned and lubricated as necessary semi-annually at intervals not to exceed 8 months, d.
(deleted) 111111 The visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation in the reactor.
Transient control rod checks and semi-annual maintenance ensure proper operation of this control rod.
SECTION 4.3.2, "CONTROL AND SAFETY SYSTEM" - Deleted These specifications apply to control and safety systens related to Reactor Operation which is no longer possible since the fuel has been shipped off-site.
O 2601J 4-4
4.3.2 control And Safety System I
. O.
Anolicability These specifications apply to the surveillance requirements for measurements, tests, ar.d calibrations of the control and safety systems.
Objective l-The objective is to verify the performance and operability of l
those systems and components which are directly related to l
reactor safety.
i I
)
Snecifications a.
The scram time shall be measured annually but at intervals l
l not to exceed 14 months.
f b.
A Channel Test of each of the reactor safety system channels for the intended mode of operation shall be performed prior to each day's operation to prior to each operation extending more than one day.
i c.
(deleted) l i
Bases l
Measurement of the scram time on an annual basis is a check not I
only of the scram system electronics, but also is an indication f
of the capability of the control rods to perform properly.
The I
channel tests will assure that the safety system channels are operatable on a daily basis or prior to an extended run.
l I
O 2601J 4-5
SECTION 4.3.3, "RADIATION MONITORING SYSTEM" - Revised This section is applicable to the conditions under the "possession Only" status.
It has been included under Section 4.2, "Limiting Conditions for Possession Only" in proposed Revision 3 to the Technical Specifications.
4.3.3 Radiation Monitorina System Apolicability This specification applies to the surveillance requirements for the area radiation monitoring equipment and the continuous air monitoring system.
Obiective The objective is to assure that the radiation monitoring equip-ment is operating and to verify the appropriate alarm settings.
Seetifications The area radiation monitoring system and the continuous air monitoring system shall be calibrated annually but at intervals not to exceed 14 months and shall be verified to be operable at weekly intervals.
Balti Experience has shown that weekly verification of area radiation and air monitoring system set points in conjunction with annual calibration is adequate to correct for any variation the the system due to a change of operating chracteristics over a long time span.
O 2601J 46 n
SECTION 4.3.4, "VENTILATION SYSTEM" - Revised This section is applicable to the conditions under the "possession Only" status.
It has been included under Section 4.2, "Limiting Conditions for Possession Only" in proposed Revision 3 to the Technical Specifications.
4.3.4 Ventilation System Apolicability This specification applies to the building confinement ventilation system.
Objective The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled environment.
Scea.i fi cation s It shall be verified weekly that the ventilation system is operable in both normal and emergency conditions.
Hun Experience accumulated over several years of operation has demonstrated that the tests of the ventilation system on a weekly basis are sufficient to assure the proper operation of the system and control of the release of radioacthe mterial.
O 2601J 4-7
SECTION 4.3.5, "EXPERIMENT AND IRRADIATION LINITS" - Deleted These specifications apply to experiments to be instclied in and performed during Rea(. tor Operations. Since the fuel has been removed from the site, this section is no longer applicable.
4.3.5 Exoeriment and Irradiation limits l
Aeolicability This specification applies to the surveillance requirements for l
experiments installed in the reactor and its experimental facilities and for irradiations performed in the irradiation facilities.
J Objective i
The objective is to prevent the conduct of experiments or mC irradiations which may damage the reactor or release excessive amounts of radiative materials as a result of failure, i
igeci fications
]
a.
A new experiment shall not be installed in the reactor or its experimental facilities untti a hazards analysis has been performed and reviewed for compliance with the limitations on Experiments, Section 3.9, by the Reactor l
Safety Committee. Minor modifications to a reviewed and approved experiment may be made at the alscretion of the senior reactor operator responsible for the operation provided that the hazards associated with the modifications have been reviewed and a determinatica made and documented that the m)difications do not create a significantly different, a new, or a greater than the original approved experiment.
2601J 4-8 i
I b.
An irradiation of a new type of device or material shall not I
be performed until an analysis of the irradiation has been f
performed and reviewed for compliance with the Limitations on Irradiations Section 3.9, by a licensed senior operator qualified in health physics, or a licensed senior operator and a person qualified in health physics.
t It has been demonstrated over a number of years of experience that experiments and irradiations reviewed by the Reactor Staff i
and the Reactor Safety Comittee as appropriate can be conducted i
without endangering the safety of the reactor or exceeding the limits in the Technical specifications.
SECTION 4.4, "REACTOR FUEL ELEMENTS" - Deleted I
These specifications are no longer applicable since the Reactor fuel has been shipped off-site for deccanissioning.
4.4 Reactor Fuel Elements, l
Aeolicability t
This specification applies to the surveillance requirements for the fuel elements, f
Objective l
I The objective is to verify the continuing integrity of the fuel element l
cladding.
i l
O 2601J 49 l
l
l l
[
Snecifications All fuel elements in the reactor core (except instrumented) shall be measured for length and bend at intervals not to exceed the sum of 25%
AK/K in pulse reactivity or 3 years, whichever comes first. The l
reactor shall not be operated with damaged fuel. A fuel element shall l
be considered damaged and must be removed from the core if*
a.
In measuring the transverse bend, the bend exceeds 0.125 inch over the length of the cladding; b.
In measuring the elongation, its length exceeds its original length by 0.25 inch; or c.
A clad defect exists as indicated by release of fission products.
Raiti I
The frequency of inspection and measurement schedule is based on the parameters most likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose characteristics are well known. The limit of transverse bend has been l
shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that there l
will be no hot spots resulting in damage to the fuel caused by this touching.
Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious
[
effects. The elongation limit has been specified to assure that the I
cladding material will not be subject to stresses that could cause a loss of integrity in the fuel containnent and to assure adequate
[
coolant flow, t
l I
O 2601J t
4-10 l
5.0 DESIGN FEATURES SECTION 5.1, "REACTOR FUEL" - Deleted This section pertains to the fuel elements used in the Reactor core.
Since all of the fuel has been shipped off-site (for decommissioning),
this specification no longer applies.
i i
5.1 Etactor Fuel Apolicability This specification applies to the fuel elements used in the reactor core.
l
]
Obiettin i
The objective is to assure that the fuel elements are of such a design j
and fabricated in such a manner as to permit their use with a high j
degree of reliability with respect to their physical and nuclear characteristics.
l Soecifications i
Standard TRIGA fuel l
The individual untrradiated standard TRIGA fuel elements shall have the following characteristics:
a.
Uranium content: maximum of 12.0 wt1 enriched to a nominal 19.9% Uranium 235.
I b.
Hydrogen-to-ztrconium atos ratio (in the ZrH,): maximum 1.7 H atoms.
l c.
Cladding:
304 stainless steel, nominal 0.020 inch thick, 2602J 5-1
r Bases A maximum uranium cet it of 12 HT1 in a standard TRIGA element is about 41% greater than the design value of 8.5 Nt%. Such an increase in loading results in an increase in local power density of approximately 411. An increase in local power density of 41% reduces the safety margin by at most 15%.
The maximum hydrogen-to-ztrconium ratio of 1.7 will produce a maximum pressure within the clad during an accident well below the rupture strength of the clad.
i SECTION 5.2, "REACTOR CORE" - Deleted i
I I
This section applies to the ccnfiguration of Reactor fuel (and i
experiments). All Reactor fuel has been shipped off-site fur
[
]
decommissioning; therefore, this section no longer applies.
i 5.2 Reacter Core i
Amtl.' eabili tv t
i This specification applies to the configuration of fuel and in-core expo *1ments.
I Objective i
The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.
Speci fica tions 1
i i
a.
The core 6 hall be an arrangement of TRIGA uranium-zirconium f
i hydride fuel-moderator alements positioned in the reactor grid plate.
i i
!O t
2602J i
5-2 r
1 b.
Ths reactor shall not be operated with a core lattice position vacant except for (1) replacement of single individual elements with in-core irradiation facilities of control rods; (2)
Two separated experiment positions in the D through E i
rings, each occupying a maximum of three fuel element positions; or (3) positios:s on the periphery of the core assembly, c.
The reflector, excluding experiments and experimental facilities, sha15 be a combination of graphite and water.
Bases a.
Standard TRIGA cores have been in use for years and their characteristics are well documented.
b.
Vacant core lattice positions will contain experiments or an experimental facility to prevent accidental fuel additions to the reactor core.
They will be permitted only on the periphery of the core to prevent power perturbation; in regions of high power density.
c.
The core will be assembled in the reactor grid plate which is located in a pool of light water. Hater in combination with graphite reflectors cea be used for neutron economy and the enhancement of experimental facility radiation requirements.
t O
2602J 5-3
j SECTION S.3, "CONTROL ROOS" - Deleted V
These design specifications pertaining to the control rods are no longer necessary since reactor critically cannot be achieved without the Reactor fuel. The Reactor fuel has been shipped off-site.
5.3 Control Rods Aeolicability This specification applies to the control rods used in the reactor core.
Obiective The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Soecifications a.
The shim and regulating control rods shall have scram capability and contain borated graphite, B4 powder or boron and its compoun d in solid form as a poison in aluminum or stainless steel cladJ!ng.
i j
b.
The safety-transient rod Jhall have scram capability and contain borated graphite or boron and its compound in a solid form as a
[
poison in aluminum or stainless steel clad. The I
safety-transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions.
l l
llO 2602]
5-4 l
l
Blin (3
%)
The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, 84 powder or boron and its compounds.
These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment.
Scram capabilities are provided for rapid insertion of the control rods which is the primary safety feature of the reactor.
The safety-transient rod is designed for a reactor pulse.
SECTION 5.4, "RADIATION MONITORING SYSTEM" - Revised The objective of these design specifications applies to Reactor Operation.
However, it will be necessary to retain a "Radiation Monitoring System" (or equipment) under the "Possession Only" status. Therefore, this section has been included under Section 5.0 in proposed Revision 3 to these Technical Specifications.
O 5.4 Radiation Monitorina System Aeolicability This specification describes the functions and essential components of the area rad'ation monitoring equipment and the system for continuously monitoring airborne radioactivity.
Obiective The objective is to describe the radiation monitoring equipment that is available to the operator to assure safe operation of the reactor.
L1 cations The radiatica monitoring equipment listed in the following table will be
(
available for Reactor Operation i
5-5
Radiation Monitorina Channel And Function O
Area Radiation Monitor (gamma sensitive instruments)
Function-Honitor radiation fields in key locations, alarm and readout at control console.
Continuous Air Radiation Monitor (Beta, Gamma sensitive detector with air collection capability)
Function - Monitor concentration of radioactive particulate activity in building, alarm and rea6ut at control console.
Gas (Ar4l) Radiation Monitor (gamma sensitivo detector with air collection capability)
Function - Monitor concentration of radioactive gases in building exhaust, readout at control console.
O am The radiation monitoring system is intended to provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity.to the surroundings.
6 3
!O 2602J l
5-6 l
l l
. - _ _ _ _. _ - _ ~
SECTION 5.5, "FUEL STORAGE" - Deleted h
(G All Reactor fuel has been shipped off-site for decommissioning. This section no longer applies.
5.5 Fuel Storaae Aeolicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.
Obiective The objective is to assure that fuel which is being stored will not become critical and will not reach an unsafe temperature.
Sae.ci fications O
a.
All fuel elements shall be stored in a geometrical array where the k-effective is less than 0.8 for all conditions of moderation, b.
Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.
111191 The limits imposed by Specifications 5.5.a and 5.5.b are conservative and assure safe storage.
O 2602J S-7
SECTION 5.6, "REACTOR BUILDING AND VENTILATION SYSTEM" - Retained This section is applicable to the "Possession Only" status.
It has been ticluded under Section 5.0, "Design Features" in proposed Revision 3 to these Technical Specifications.
5.6 Reactor Buildina and Ventilation System Aeolicability This specification applies to the building which houses the reactor.
Obiective The objective is to assure that provistor.s are made to restrict the amount of release of radioactivity into the environment.
Soecifications O
a.
The reactor shall be housed in a facility designed to restrict 8
leakage.
The minimum free volume in the facility shall be 2 x 10 J
cubic centimeters.
b.
The reactor laboratory shall be equipped with a ventilation system designed to filter and exhaust air or other gases from the reactor laboratory and release them from a stack at a minimum of 13.7 meters from ground level.
The filter shall be used during emergency situation specified by the continuous air monitor or by the operator.
c.
Emergency filtering controls for the ventilation system shall be located in the control room and the system shall be designed to filter in the event of a substantial release of fission products.
O 2602]
5-8 i
l OO T'
'ity is designed such that the ventilation system will normally
, aegative pressure with respect to the atmosphere so that there uncontrolled leakage to the environment.
The free air volume e'
as reactor laboratory is confined when emergency filtering is being
.ne d. Controls for emergency filtering and normal operation of the ventilation system are located in the control room.
Proper handling of airborne radioactive materials (in emergency situation) can be conducted from the control room with a minimum of exposure to operating personnel.
SECTION 5.7, "REACTOR POOL MATER SYSTEM" - Revised It has been included under Section 5.0, "Design Features", in proposed Revision 3 to these Technical Specifications. The design features of the Reactor Pool Mater System are applicable to the conditions during the "Possession Only" status.
5.7 Reactor Pool Water System Aeolicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.
Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor-core and adequate radiation shielding.
Soecificationi a.
The reactor core shall be cooled by natural convective water flow.
2602J 5-9
-,,_r
b.
The pool water inlet pipe to the demineralizer and heat g/
(,
exchanger shall cot extend more than 4.5 meters below the top of the reactor pool when fuel is in the core.
c.
Pool water inlet to the demineralizer and heat exchanger shall have vacuum breaker holes machined into the pipe no more than one meter below the top of the reactor pool, in case of pool water loss due to external pipe system failure.
d.
The reactor shall not be operated if the pool water level is less than 5.48 meters above the top grid plate of the core, e.
The bulk pool temperature shall be monitored while the reactor is in operation and the reactor shall be shut down if the 0
temperature exceeds 50 C, f.
The pool water shall be sampled for conductivity at least weekly. Conductivity averaged over a month shall not exceed 5 micromhos per centimeter.
H1111 a.
This specification is based on thermal and hydraulic calculations which show that the TRIGA core can operate in a safe manner at power levels up to 2,700 kw with natural convection flow of the coolant water.
b.
In the event of accidental siphoning of pool water through inlet and outlet pipes of the demineralizer and heat exchanger system, the pool water level will drop no more than 4.5 meters from the top of the pool.
O 2602J 5-10
c.
In the event of external pipe system failure, the vacuum breaker holes machined into the pipe will cause the cessation of water pumping after the loss of not more than one meter of water.
d.
This specification assures that adequate shielding is provided by the pool water while the reactor is 4
- operated, e.
The water conductivity is an indicator of the water purity and can be used to monitor for the leakage of ground water into the tank. Maintaining low conductivity readings should allow early detection of leaks of this type. Another reason to maintain low conductivity is to insure low ton or mineral concentration in the water.
Thus there is only a small likelihood of inducing activity in the mineral ions which are in the solution.
The result is to limit the remediation levels experienced in the reactor room.
I i
O 2602J 5-11 I
[
t rw c,,
_,,,,,,-_-v n..
c
m SECTION 6.0, "ADMINISTRATIVE CONTROLS" - Revised
_~
[G' This Section is applicable to the Possession Only status and has been included (with the necessary modifications) in proposed Revision 3.
6.0 ADMINISTRATIVE CONTROLS 6.1 Oraanization - Revised
- a. The facility shall be under the direct control of the Reactor Supervisor or a licensed senior operator designated by him to be in direct control.
The Supervisor shall be responsible to the Dean of the College of Engineering and the Associate Dean for Graduate Studies and Research for safe operation and maintenance of the reactor and its associated equipment. The Supervisor or his appointee shall re$iew and approve all experiments and experimental procedures prior to their use in the reactor. He shall enforce rules for the protection of personnel against radiation.
- b. The safety of operation of the HSV TRIGA Nuclear Reactor shall be related to the University Administration as shown in the following chart.
6.2 Review and Audit - Revised
- a. A Reactor Safety Committee (RSC) of at least five (5) members knowledgeable in fields which relate to Nuclear Safety shall review, evaluate, and approve safety standards associated with the operation and use of the facility.
The University Radiation Safety Officer and the Reactor Supervisor shall be members of the Reactor Safety Committee.
The jurisdiction of the RSC shall include all nuclear operations in the facility and general safety standards.
lO 2603J 6-1 l
- b. The operations of the Reactor Safety Comittee shall be in accordance with a written charter, including provisions for:
(1) Meeting frequency, (2) Voting rules, (3) Quorums, I
(4) Method of submission and content of presentation to the Comnittee, and (5) Use o? subcomittees.
- c. The RSC or a Subcomittee thereof shall audit reactor operations at least quarterly, but at intervals not to exceed four months.
O t
t i
I e
t 2603J 6-2 l
i i
r3 REVISED G
Presidsnt l
l l
l Provost l_______________
l l
l l
l Dean, Coll. of l_____1_________
Engineering l
1 I
I I
I l
l Assoc. Dean for l
Faciity Graduate Studies Reactor Safety l
Advisor and Research Comittee l
l l
l l
l 1
l l__________
Reactor Radioisotope Supervisor Comittce l
J l__
Office of Radiation, Chemical and Reactor Biological Staff Safety Line Responsibility
Advisory Responsibility i
O 2603J 6-3
l
- d. The responsibilities of the committee or designated Sub-committee thereof include, but are not limited to, the following:
(1) Review and approval of experiments utilizing the reactor facilities; (2) Review and approval of all proposed changes to the facility, procedures, and Technical Specifications; (3) Review of the operation and operational records of the facility; (4) Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50; (5) Determination of whether a proposed change, test, or experiment would constitute an unreviewed safety question or a change in the Technical Specifications; and (6) Review of abnormal performance of facility equipment and operating anomalies.
SECTION 6.3, "ACTION TO BE TAKEN IN THE EVENT A SAFETY LINIT IS EXCEEDED" - Deleted Since the fuel has been shipped off-site and Reactor Operations no longer possible, these actions are no longer applicable.
6.3 Action To Be Taken In The Event A Safety Limit Is Exceeded In the event a safety limit is exceeded:
- a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC; O
2603J 6-4
(V3
- b. An imediate report of the occurrence shall be made to the Chairman, Reactor Safety Comittee, and reports shall be made to the NRC in accordance with Section 6.7 of these specifications; and
- c. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recomendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safety Comittee for review and then submitted to the NRC when authorization is sought to resume operation to the reactor.
SECTION 6.4, "ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE F4URRENCE" - Deleted By definition (see Section 1.0) a reportable occurrence is associated with Reactor Operations only. Therefore, the requirements contained in this Section do not pertain to th Possession Only" status.
6.4 Action To Be Taken In The Event Of A Reoortable Occurrence In the event of a reportable occurrence, the following action shall be taken:
- a. The Supervisor or his designated alternate shall be notified and corrective action taken with respect to the operations involved;
- b. The Supervisor or his designated alternate shall notify the Chairman of the Reactor Safety Comittee; l
2603J 6-5 l
- c. A report shall be made to the Reactor Safety Committee which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence; and
- d. A report shall be made to the NRC in accordance with Section 6.7 of these specifications.
SECTION 6.5, "0PERATING PROCEDURES" - Deleted This Section pertains to Reactor Operations and the maintenance requirements associated with such operations. Since the Reactor will not (cannot) be operated under the Possession Only status, this Section no longer applies.
6.5 Ooeratina Procedures Written operating procedures shall be adequate to assure the safety of operation of the reactor, but shall not preclude the use of independent judgment and action should the situation require such. Operating procedures shall be in effect for the following items;
- a. Testing and calibration of reactor operating instrumentation and controls, control rod drives, area radiation monitors, and air parti wiato monitors;
- b. Reactor startup, operation, and shutdown;
- c. Emergency and abnormal conditions, including provisions for evacuation, reentry, and medical support;
- d. Fuel element loading or unloading; lO 26033 6-6 1
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- e. Control rod removal or replacement;
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- f. Routine maintenance of the control rod drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety;
- g. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms and abnormal reactivity changes, and
- h. Civil disturbances on or near the facility site.
Substantive changes to the above procedures shall be made only with the approval of the Reactor Safety Committee.
Temporary changes to the procedures that do not change their original intent may be made by the Supervisor or his designated alternate. All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Committee.
6.6 Facility Ooeratina Records - Revised In addition to the requirements of applicable regulations, and in no way substituting therefor, records and logs shall be prepared of at least the following items and retained for a prior of at least five years for items a through f i:nd indefinitely for items g through k.
r
- a. Normal reactor operation,
- b. Principal maintenance activities,
- c. Reportable occurrences, l
- d. Equipment and component survelliance activities required by the Technical Specifications, 2603J 6-7
- e. Experiments performed with the reactor,
- f. Gaseous and liquid radioactive effluents released to the environs,
- g. Offsite environmental monitoring surveys,
- h. Fuel inventories and transfers,
- i. Facility radiation and contamination surveys,
- j. Radiation exposures for all personnel, and
- k. Updated, corrected, and as-built drawings of the facility.
6.7 Reoortina Reauirements - Revised In addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC Region III, Office of Inspection and Enforcement as follows:
- a. A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone or telegraph.
(1) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the the release resulted in property damage, personal injury, or exposure.
- b. A report within 10 days in writing of:
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(1) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal it: jury or exposure.
The written report (and, to the extent possible, the preliminary telephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent reoccurrence of the event; (2) Any violation of a safety limit; and (3) Any reportable occurrence as defined in Section 1.9 of these specifications,
- c. A report within 30 days in writing of:
(1) Any significant variation of measured values from a corresponding predicted or previously measured value of safety-connected operating characteristics occurring during operation of the reactor; (2) Any significant change in the transient or accident analysis as described in the Safety Analysis Report:
(3) Any changes in facility organization; and (4) Any observed inadequacies in the implementation of administrative or procedural controls.
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2603J 6-9
SECTION 6.7.1 - Deleted
,y
- U 6.7.1 A Report within 90 days after completion of startup testing of the reactor upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing tha measured values of the operating conditions of characteristics of the reactor under the new conditions including
a.
An evaluation of facility performance to date in comparison with design predictions and specifications, and b.
A reassesscent of the safety analysis submitted with the license application in light of measured operating charteteristics when such measurements indicate that there may be substantial variance from prior analysis.
SECTION 6.7.2 - Deleted Note that an annual report will be prepared for calendar year 1988.
In addition, the decommissioning plan which is to be submitted to the NRC in December 1988 will contain a summary of all operations that will occur, and lead to ultimate release of the Reactor Facility from License R-114.
6.7.2 An annual report covering the operation of the unit during the previous calendar year submitted prior to March 31 of each year providing the following information:
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O 26033 6-10
a.
A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections; b.
Tabulation of the energy output (in megawatt days) of the reactor and the hours the reactor was critical; c.
The number of emergency shutdowns and inadvertent screams, includirg reasons therefore; i
d.
Discussion of the major maintenance operations performed i
during the period, including the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required; O
e.
A brief description, including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50; q
f.
A sutenary of the nature and amount of radioactive effluents l
released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.
Llpuid Haste _(summarized on a monthly basis)
(1) Radioactivity discharged during the reporting period O
26033 6-11
fm (a) Total radioactivity released (in curies).
C (b) The HPC used and the isotopic composition if greater than 1 x 10~7 microcuries/cc for fission and activation products.
(c) Total radioactivity (in curies), released by nuclide, I
during the reporting period based on representative isotopic analysis.
(d) Average concentration at point of release (in microcuries/cc) during the reporting period. (2)
Total volume (in gallons) of effluent water (including diluent) released during ei.ch period of release Gaseous Haste (summarized on an annual basis)
(1)
Radioactivity discharged during the reporting period (in curies)
(a) Total estimated quantity of radioactivity released (in curies) determined by an appropriate sampling and counting method.
I (b) Total estimated quantity of Argon-41 released (in curies) during the reporting perivd based on data fram an appropriate monitoring system.
(c) Estimated average atmospheric diluted concentration of l
Argon-41 released during the reporting period in terms of microcuries/cc and fraction of the applicable HPC values.
i (d) Total estimated quantity of radioactivity in particulate form with f..lf lives greater than eight days (in curies) released during the reporting period as determined by an appropriate particulate monitoring system.
!O 26033 6-12 l
p (e) Average concentration of radioactive particulates with half k,)
lives greater than eight days released in microcuries/cc during the reporting period.
(i) An estimate of the average concentration of other significant radionuclides present in the gaseous waste discharge in terms of microcuries/cc and fraction of the applicable NPC value for the reporting period if the estimated release is greater than 20% of the appilcable HPC.
(g) An annual summary of the radiation exposure received by facility personnel and visitors in terms of the average radiation exposure per individual and greatest exposure per j
individual in the two groups.
Cach significant exposure in excess of the limits of 10 CFR 20 should be reported including the time and date of the exposure as well as the f
name of the individual and the circumstances leading up to the exposure.
(h) An annual summary of the radiation levels and levels of l
contamination observed during routine surveys performed at the facility in terms of the average and highest levels.
(1) A description of any environmental surveys performed I
outside the facility.
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