ML20087E423
| ML20087E423 | |
| Person / Time | |
|---|---|
| Site: | Dow Chemical Company |
| Issue date: | 08/07/1995 |
| From: | DOW CHEMICAL CO. |
| To: | |
| Shared Package | |
| ML20087E405 | List: |
| References | |
| NUDOCS 9508140142 | |
| Download: ML20087E423 (47) | |
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TECHNICAL SPECIFICATIONS FOR THE DOW TRIOA RESEARCH REACTOR FACILITY LICENSE R-108 PROPOSED AMENDMENT 7 AUGUST 1995 This document includes the Technical Specifications and the
- bases for the Technical Specifications. The bases provide the technical support for the individual Technical Specifications and 4 are included for information purposes only The bases are not part of the Technical Specifications and they do not constitute' limitations or requirements to which the licensee must adhere.
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't ' DEFINITIONS
. 1.1. ALARA - The ALARA (As Low As Reasonably Achievable) program is a seto' f procedures which is intended to ndnimize occupational exposures to ionizing radiation ;
and releases of radioactive materials to the environment. j 1.2. ' Channel - A channel is a combinadon of sensors, electronic circuits, and output devices connected by the appropriate communications network in order to measure and display . '
the value of a parameter.
- 1.3. channel Calibration - A channel calibration is an adjustment of a channel such that its. . 5 output corresponds with acceptable accuracy to known values of the paraneter which the !
channel measures.' Calibration shall encompass the entire channel, including equipment,
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i actuadon, alarm, or trip and shall include a Channel Test. l 1.4. Channel Check - A channel check is a qualitative verification of acceptable performance :
by observation of channel behavior. 'Ihe verification shall include comparison of the 1 g
channel with other independent channels or systems measuring the same variable, j whenever possible. !
1.5. Channel Test - A channel test is the introduction of a signal into a channel for verificadon - !
of the operability of the channel. l 1.6. Connnement - Confinement is an enclosure of the facility which controls the movement of i
- air into and out of the facility through a controlled path. (
1.7. Excess Reactivity - Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive position from the condition where the reactor is exactly critical. ,
1.8. Fxneriment - An experiment is any device or material, not normally part of the reactor, - l which is introduced into the reactor for the purpose of exposure to radiation,' or any -
operation which is designed to investigate non-routine reactor characteristics. .j i
- 1.9. Exnerimental Facilities include the rotar}' specimen rack, sample containers replacing fuel elements or dummy fuel elements in the core, pneumatic transfer systems, the -
central thimble, and the area surrounding the core.
1.10. Facility Director - Person with live management responsibility to whom the Reactor Supervisor reports. 'Ihe person aIso serves as chairperson of the Reactor Operations Committee. j i
1.11. 1.imitine conditions for Oneration ' Limidng Condidons for Operation (LCO) are administratively established constraints on equipment and operational characterlsdes which shall be adhered to durir:g operation of the reactor.-
1.12. 1.imitine Safety Svsum Settine U.SSS) - An LSSS is the actuating level for automatic protective devices ' elated to those variables having significant safety funcdons.
1.13. Measured Value - A measured value is the value of a parameter as it appears on the output of a channel, j l
l Amendment No. 7 3
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. . 6 1.14. Modilled Routine Experiments - Modified routine experiments arc experiments which '
have not been designated as routine experiments or which have not been performed '
previously, but are similar to routine experiments in that the hazards are neither ,
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significantly different from nor greater than the hazards of the corresponding roudne
. . experiment. l
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' 1.15. Movable Erneriment - A movable experiment is an experiment intended to be moved in - .j or near the core or into and out of the reactor while the reactor is operating. '!
., t 1.16. Oncrable - A component or system is operable if it is capable of perkaning its intended function. ,
1.17. Onerating - A component or system is operating ifit is performing its intended funcdon.
1.18. Rndintion Snfety Committee (RSC)- De RSC is responsible for the administration of all Dow Midland location activities involving the use of radioactive materials and .;
radiadon sources including assuring compliance with US NRC regulations.
l 1.19. Reactivity I imits - The reacdvity limits are those limits imposed on reactor core excess reacuvity. Quantitles are referenced to a Reference Core Condidon.
1.20. Reactivity Worth of an Exneriment - The reactivity worth of an experiment is the _
maximum absolute value of the reactivity change that would occur as a result of intended or antipated changes or credible malfuncdons that alter experiment position or j configuration.
1.21. Reactor Oncrnting - De reactor is operating whenever it is not secured or shutdown. :
i 1.22. Reactor Safety Circuits - Reactor safety circuits are those circuits, including the . l associated input circuits, which are designed to initiate a reactor scram.
1.23. Reactor Secured - The reactor is secured whenever:
a) !! contains insufficient fissile material present in the reactor, adjacent experiments '
or control rods, to attain cridcality under opdmum available conditions of moderatica and reflection, or .l b) the console switch is in the "off" position, the key is removed from the switch, and the key is in the control of a licensed reactor operator or stored in a locked storage area; and j sufficient control rods are inserted to assure that the reactor is subcritical by a margin greater than $1.00 cold, without xenon; and no work is f4 imStad involving core fuel, core structure, installed control rods or >
- control red tnt b4N those drives are physically disconnected from the control
- rods; and '
no experiments in or near the core are being moved or serviced that have, on movement, a reactivity worth e xceeding $0.75.
I Anwodnwat No. 7 ~
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~ 1.24. Reactor Shutdown - De reactor is shutdown ifit is subcritical by at least one dollar and ?
the reactivity worth of all experiments is accounted for.~
^ reactor operations, including both review and audit functions.
- 1.26. - Reactor snretv svstems - Reactor Safety Systems are those systems, including -
~~
associated input channels, which are designed to initiate automade reactor protection or ;
to provide informadon for initiation of manual protective action.~
l .27.. Reference Core condition - The Reference Core Condition is that condition when the core is at ambient temperature (cold) and the reactivity worth of xenon in the fuel is negligible (less than $.30).-
1.28. Research Reactor- A Research Reactor is a device designed to support a self-sustaining
'. nuclear chain reaction for research, development, education, training, or experimental purposes, and which may have provisions for the production of radioisotopes.
1.29. Renortable Occurrence - A Reportable Occurrence is any of the following which occurs during reactor operation:
a) Operation with actual safety-system settings for required systems less
. conservative than the limiting safety system settings specified in Technical
- Specificadon 2.2.
3 ' b) -
Operadon in violadon of limidng condidons for operation established in the o Technical Specificadons.
c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless '
the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.
. d) Any unanticipated or uncontrolled change in reactivity greater than one dollar.
Reactor trips resulting from a known cause are excluded.
c) Abnormal and significant degradadon in reactor fuel, cladding, or coolant boundary which could result in' exceeding prescribed radiation exposure or release limits.
f) An observed inadequacy in the implementation of either administradve or procedural controls which could result in operation of the reactor outside the -
limiting conditions for operation.
g) Release of radioactivity from the site above limits specilled in 10CFR20.
1.30. Rod Control - A c(mtrol rod is a device containing neutron absorbing material which is used to control the nuclear fission chain reaction. The control rods are coupled to the .
control rod drive systems in a way that allows the control rods to perform a safety -
function.
Amendment No. 7
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- i 1 1.31. Routine Funeriment - A routine experiment is an experiment which involves operations -
s under conditions which have been extensively examined in the course of the reactor test i~ programs and which is not defined as any other kind of experiment. Experiments and -
classes of experiments which are to be considered as routine experiments must be so defined by the Peactor Operations Committee.
l.32. ' Saferv i imit - A Safety Limit is a llmit on an important process variable which is found to L be necessary to reasonably protect the integrity of certain of the physical barriers which ' .i guard against the uncontrolled release of radioacdvity. 'Ihe principal physical barrier is j the fuel element cladding.
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.. . a 1.33. Scram time - Scram Time is the time required to fully insert the control rods following the ;
actuation of a Limiting Safety System Setting. j
.j 1.34. Secured Funeriment . A Secured Experiment is any experiment, experimental facility, or ; i component of an experiment that is held in a stationary position relative to the reactor by j mechanical means. The restraining forces must be substantially greater than those to 1 which the experiment might be subjected by. hydraulic, pneumatic, buoyant, or other .j forces which are normal to the operating environment of the experiment, or by forces L which can arise as the result of credible malfunctions.
l.35. Shn11. Shouhl. nnd May 'Ihe word "shall" is used to denote a requ'irement, the' word "should" denotes a recommendation, and the word "may" denotes permission, neither a .
- requirement nor a recommendation, _
1.36. ' Shutdown Marcin - Shutdown Margin is the reactivity existing when the most reactive control rod is fully withdrawn from the core and the other control rods are fully inserted into the core.
1.37. Snecini Exneriments - Special experiments are experiments which are neither routine -
experiments nor modified routine experiments. j i
1.38. TRIGA Fuel Element - A TRIGA fuel element is a sealed unit containing (U,Zr)H fuel - ')
for the reactor. 'Ihe uranium is enriched to less than 20% in 235-U and the fraction of - !
l hydrogen is in the range of 1.0-1.1 for aluminum-clad TRIGA elements and in the range of 1.6-1.7 for stainless-steel-clad TRIGA elements.
Amendnwat No.1 4
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? ,' 2 SAFETY I_IMITS AND I_IMITING SAFETY SYSTEM SETTINGS:
2.1, Enfety T Imit (SI3 ;
i Applicability j y .,
- lhls specification applies to the temperature of the reactor fuel. '!
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- Ohlective j i
'Ihe objective of this specification is to define the maximum fuel temperature that _- !
can be permitted with confidence that no damage to the fuel element will result. - .l
- i Snecifiention ;
- I The temperature in any fuel element in the Dow TRIGA Research Reactor sha11 !
not exceed 500 C under any conditions of operation.
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A loss in the integrity of the fuel element cladding could arise from a buildup of .
excessive pressure between the fuel and the cladding if the fuel temperature ,
exceeds the safety limit. '1he pressure is caused by the heating of air, fission :
- product gases, and hydrogen from the dissociation of t% fuel-moderator. The
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magnitude of this pressure is determined by the temperature of the fuel element .I and by the hydrogen content. Data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of ZrHi .6 will remain below the ultimate 'j stress provided that the fuel temperature does not exceed 1050 C and the fuel !
cladding temperature does not exceed 500 C. When the cladding temperature can ' l equal the fuel temperature the fuel temperature design limit is 950 C (M. T. ; ;
Simnad, G.A. Project No. 4314, Report e-Il7-833,1980).-
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Experience with operation of TRIGA fueled reactors at power levels up to 1500 ' ;
kW shows no damage to the fuel due to thermally-induced pressures.
The thermal characteristics of aluminum-clad TRIGA fuel elements using ZrHi .: J moderator have been analyzed (S. C. Hawley and R. L. Kathren, __
NUREG/CR-2387. PNL-4028, Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors,1982). A loss-of-coolant analysis showed that in a typical graphite-reflected Mark I TRIGA reactor fueled with 60 aluminum-clad fuel elements (Reed College) the maximum fuel temperature would be less than 150 C q following infinite operation at 250 kilowatts terminated by the instantaneous loss ~ 'j of water. These temperatures are well below the region where the a+ 6 + Y to a
+ 6 phase change occurs in ZrHi .: (560 C).
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i Amendment No. 7 '!
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- 2.2.1 imit1Ae Safetv Svstem Settinae (I RRR)
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F, ; his specification agiplies to.the reactor scram setting which prevents the reactor .
M, , fuel temperature from reaching the safety limit.
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' Ohlective ' -
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- De objective of this specificadon is to provide a' reactor scram to prevent thel
. safety limit from being reached.
Rnecifiention >
The LSSS shall not exceed 300 kW as measured by the calibrated power !
w channels.
l The LSSS which does not exceed 300 kW provides a considerable safety margin. .
One TRIGA reactor (General Atomics, Torrey Pines) showed a maximum fuel l temperature of 350 C during operation at 1500 kilowatts, while a 250-kilowatt -
- . TRIGA reactor (Reed College) showed a maximum fuel temperature ofless than'.. Li 150 C (reported by S. C. Hawley, R. L. Kathren, NUREG/CR-2387, PNL-4028 ,
(1982), Credible Accident Annivses for TRIGA and TRIGA-Fnalad Danctors). A ' .
portion of the safety margin could be used to account for variations of flux level (and thus the power density) at various parts of the core. De safety margin . ;
-should be ample to. compensate for other uncertainties, including power transients .
during otherwise steady-state operadon, and should be adequate to protect? g aluminum-clad fuel elements from cladding failure due to temperature and pressure effects. ;
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3.1IMITING CONDITIONS FOR OPERATION (LCO) 3.1. Renetivity I imits
>n Amlicability These specifications shall apply to the reactor at all times that it is in operation.
Oblective The purpose of the specification is to ensure that the reactor can be controlled and shut down at all times and that the safety limit will not be exceeded.
Snecifications The reactor shall be shutdown by more than $.50 with the most reactive control rod fully withdrawn, the other two control rods fully inserted, cold, no xenon, including the reactivity worth of all experiments.
'Ihe excess reactivity measured at less than 10 watts in the reference core condition, with or without experiments in place, shall not be greater than $3.00.
D.LT1 The value of the minimum shutdown margin assures that the reactor can be safely shut down using only the two least reactive control rods.
The assignment of a specification to the maximum excess reactivity serves as an additional restriction on the shutdown margin and limits the maximum power excursion that could take place in the event of failure of all of the power level safety circuits and administrative controls.
Amendment No. 7 7
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3.2. Core configuration .l t
- Annlienhility' L i This specification applies to the core configuration.
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. Ohlective The objective of this specification is to assure that the safety limit will not be exceeded due to power peaking effects. l Snecifications The critical core shall be an assembly of standard NRC-approved stainless-steel-clad or aluminum-clad TRIGA fuel elements in light water.
' 'Ihe fuc! shall be arranged in a close-packed array for operation at full licensed power except for (1) replacement of single individual fuel elements with in-core irradiation facilities or control rod guide tubes and (2) the start-up neutron source.
The aluminum-clad fuel element shall be placed in the E or F ring of the core, f Operation with standard NRC-approved TRIGA fuel in the standard configuration ;
ensures a conservative limitation with respect to the Safety Limit. ;
Placement of the aluminum-clad fuel element in the outer rings of the reactor core will help ensure that this element is not exposed to h]gher than average power ;
levels, thus providing a greater degree of conservatism with respect to the Safety i Limit for this one element. !
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Amendment No. 7
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- 3.3. De=ctor control and Safety Systems
. Annilenhillty .
'1hese specifications apply to the reactor control and safety systems and -
't safety-related instrumentation that must be operating when the reactor is in
. operation.'
Ohlective
. The objective of these specifications is to assure that all reactor control and safety .
systems and safety-related instrumentation are operable to minimum acceptable.-
standards during operation of the reactor.
Snecifications -
There shall be a minimum of one scram-capable analog safety channel.
There shall be a minimum of three operable control rods in the reactor core.
Each of the three control rods shall drop from the fully withdrawn position to the'-
fully inserted position in a time not to exceed one second.
The reactor safety channels and the interlocks shall be operable in accordance ' ,
with table 3.3A. ~;
1 The reactor shall not be operated unless the measuring channels listed in Table ~
3.3B are operable. :
Positive reactivity insertion rate by control rod motion shall not exceed $.20 per . -
second.
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Basci Safety channels with scram capability utilizing analog circuitry have been proven -
acceptable by more than thirty years of experience.
De requirement for three operable control rods ensures that the reactor can meet
- the shutdown specifications.
De control rod drop time specification assures that the reactor can be shutdown . .
promptly when a scram signal is initiated. De value of the control rod drop time is -
adequate to assure safety of the reactor.
Use of the specified reactor safety channels, set points. and interlocks given in table 3.3A assures protection against operadon of the mactor outside the safety limits.
The requirement for the specified measurement circuits provides assurance that important reactor operation parameters can be monitored during operation.
He specification of maximum positive reactivity insertion rate helps assure that the Safety Limit is not exceeded.
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Amendment No, 7 ;
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TABIE 3.3A.
- MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS
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Scram Channels f Scram Chnnnel Minimum Onerable Scram Setnoint Reactor Power Level 2 Not to exceed maximum licensed -
NM1000 & NPP1000 power NPP1000 1 Failure of the detector Detector liigh-Voltage Power Supply high-voltage power supply ;
NM1000 1 Failure of the detector ,
Detector fligh-Voltage Power Supply high-voltage power supply -
Manual Scram 1 Not applicable Watchdog (DAC to CSC) 1 Not applicable Interlocks Interlock /Chnnnel Function Startup Countrate Prevent control rod withdrawal when the neutron count rate is less than 2 cps Rod Drive Control Prevent simultaneous manual withdrawal of two control elements by the control rod drive motors Reactor Period Prevent control rod withdrawal when tim period is -
less than 3 seconds Amendment No. 7 h 2.
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TABLE 3.3A BASES FOR REACTOR SAFETY CHANNELS AND INTERLOCKS : ,
r; .
3 Scram Channels : l
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. Scram ('hnnnei Rases ,
. Reactor Power Level Provides assurance that the reactor will be' shut down automatically before the safety limit can be exceeded Reactor Power Channel' Provides assurance that the reactor
Detector Power Supplies cannot be operated without power to the neutron f detectors which provide input to the NM1000 and l NPP1000 power channels - ,
Allows the operator to shut the reactor down at any 0 Manual Scram indication of unsafe or abnormal conditions .
l Watchdog Ensures adequate communications between the Data Acquisition Computer (DAC) and the Control System Computer (CSC) units. a
.l Interlocks j 1
InterlockK'hnnnel Bjligg Startup Countrate Provides assurance that the signal in'the NM1000 channel is adequate to allow reliable indication of the state of the neutron chain reaction.
Rod Drive Control Limits the maximum positive reactivity insertion rate
. Reactor Period Prevents operation in a regime in which transients could cause the safety limit to be exceeded l
l 71 Amendment No, 7
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? ,2 if; TABLE 3.3B l MEASURING CHANNELS- '
a
> ' Measuring channel Minimum Number Operable .
. NM1000 -1 ;
NPP1000 1 Water Radioactivity 1-Monitor ;
Water Temperature 1.
- Monitor ;
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TABLE 3.3B -
i DASES FOR MEASURING CHANNELS .
f Mensuring Chnnnel M NM1000 Provides assurance that the reactor power level can be -
adequately monitored.'
NPP1000 Provides assurance that the reactor power level can be -
adequately monitored.
Water Radioactivity Provides assurance that the water Monitor radioactivity level can be adequately monitored. :
e i Water Temperature Provides assurance that the water Monitor temperature can be adequately monitored.
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- 3.4. Conlant Svatam j
' Annilenhilltv I
. Dese speclilcations apply to the quality of the coolant in contact with the fuel
- i. cladding, to the level of the coolant in the pool,' and to the bulk temperature of the -
coolant.
Ohlectives
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'Ihe objectives of this specification are:
to minimize corrosion of the cladding of the fuel elements and minimize neutroni l activation of dissolved materials, '
i to detect releases of radioactive materials to the coolant before such releases i i
become signlilcant, to ensure the presence of an adequate quantity of cooling and shielding water in .
~t he pool, and ~i to prevent thermal degradation of the ion exchange resin in the purification system, Snecifientlans '!
The conductivity or the yv.t water shall not exceed 5 p.mhos/cm averaged over one :
month.
The pool water pH shall be in the range of 4 to' 7.5.
The amount of radioactivity in the pool water shall not exceed 0.1 pC1/mL.
The water must cover the core of the reactor to a minimum depth of 15 feet during operation of the reactor.~
The bulk temperature of the coolant shall not exceed 60 C during operation of the reactor.
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Amendment No. 7 14
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B.aSci Increased levels of conductivity in aqueous systems indicate the presence of '
corrosion products and promote more corrosion. Experience with water quality control at many reactor facilities, including operation of the Dow TRIGA Research Reactor since 1967, has shown that maintenance within the specified limit provides acceptable control. Maintaining low levels of dissolved electrolytes in the pool water also reduces the amount of induced radioactivity, in turn decreasing the exposure of personnel to ionizing radiation during operadon and maintenance.
Both of these results are in accordance with the ALARA program.
Monitoring the pH of the pool water provides early detection of extreme values of .
pH which could enhance corrosion.
Monitoring the radioactivity in the pool water serves to provide early detection of .
possible cladding failures. Limitation of the radioactivity according to this specification decreases the exposure of personnel to ionizing radiation during operation and maintenance in accordance with the ALARA program.
Maintaining the specified depth of water in the pool provides shielding of the radioactive core which reduces the exposure of personnel to ionizing radiadon in accordance with the ALARA program.
Maintaining the bulk temperature of the ccolant below the specified limit assures l minimal thermal degradation of the ion exchacge resin.
k i-Amendment No. 7
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3.5. cnnfinement . ,
Annilenhilltv -
This specification applies to the reactor room confinement.
l Ohlective
'Ihe objective'of this specification is to mitigate the consequences of possible - <
release of radioactive materials to unrestricted areas.
5snecification The ventilation system shall be operable and the external door (Door 10) shall be closed whenever the reactor is operated, fuel is manipulated, or radioactive materials with the potential of airborne releases are handled in the reactor room.
Ibul5 ,
This specification ensures that the confinement is configured to control any releases of radioactive material during fuel handling, reactor operation, or the handling of possible airborne radioactive material in the reactor room.
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Amendment No. 7 16
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E -3.6.' Nkilwinn Monitoring Systems .1 s
' Annlicability '
. Dese specifications apply to the radiation monitoring information available to the l reactor operator during operation of the reactor. ,
i
~ Ohlective The objective of these specifications is to. ensure that the reactor operator has .
adequate information to assure safe operation of the reactor. ;
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Snecifications l
A Continuous Air Monitor (CAM) (with readout meter and audible alarm) to measure radioactive particulates in the reactor room must be operating during operation of the reactor.
I De Area Monitor (AM)(with readout meter and audible alarm) in the reactor room must be operating during operation of the reactor or when work is being done -j on or around the reactor core or experimental facilities. During short periods of repair to this monitor, not to exceed thirty days, reactor operations or work on or-around the core or experimental facilities may continue while a portable ._
gamma-sensitive ion chamber is utilized as a temporary substitute, provided thatf the substitute can be monitored by the reactor operator. !
Bascs l
The radiation monitors provide information of existing levels of radiation and air-borne radioactive materials which could endanger operating personnel or which -
could warn of possible malfunctions of the reactor or the experiments in the .,
reactor.
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. 3.7. Exnerimentn
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' Annlicability
& .'Ihese specifications apply to experiments installed in the reactor and'Its- f experimental facilities. 1 Ohlective
'Ihe objective of these specifications is to prevent damage to the reactor or ~l excessive release of radioactive materials in case of failure of an experiment. .
Sneciflentions ,
- 1. Operation of the reactor for any purpose shall require the review and -
approval of the appropriate persons or groups of persons, except that ,
operation of the reactor for the purpose of performing routine checkouts, where written procedures exist for those operadons, shall be authorized by -
the written procedures. An operation shall not be approved unless the evaluadon allows the conclusion that the failure of an experiment will not lead to the direct failure of a fuel element or of any other experiment. ;
- 2. The total absolute reactivity worth of in core experiments shall not exceed -l
. $1.00. 'Ihis includes the potential reactivity which might result from ,
experimental malfunction, experiment flooding or voiding, or the removal or insertion of experiments.- l
- 3. Experiments having reactivity worths of greater than $0.75 shall be securely . l
' located or fastened to prevent inadvertent movement during reactor j operation. ;
- 4. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials or !
liquid fissionable materials shall be doubly encapsulated.
- 5. Materials which could react in a way which could damage the components of1 the reactor (such as gunpowder, dynamite, TNT, nitroglycerin, or PETN) ,
shall not be irradiated in quantitles greater than 25 milligrams in the reactor 'i or experimental facilities without out-of-core tests whleh shall indicate that, - I with the containment provided, no damage to the reactor or its components shall occur upon reaction. L Such materials in quantities less than 25 -
milligrams may be irradiated provided that the pressure produced in the experiment c(mtainer upon reaction shall be calculated and/or experimentally, demonstrated to be less than the design pressure of the container. . Such materials must be packaged in the appropriate containers before being- ,
brought into the reactor room or must be in the custody of duly authorized i local, state, or federal officers. :i Amendment No. 7
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n.; 6. Experiment materials, except fuel materials, which could off-gas, sublime,.
- volatilize or produce aerosols under (a) normal operating conditions of the -
experiment or the reactor, (b) credible accident conditions in the reactor or -
p, (c) possible accident conditions in the experiment shall be limited in activity.
!- such that if 100% of the gaseous a(#1vity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of f radioactivity would not exceed the limits of Appendix B of 10 CFR Part 20.
The following assumptions should be used in calculations regarding experiments:
O a. If the emuent from an experimental facility exhausts through a . .
holdup tank which closes automatically on high radiation levels, the assumption shall be used that 10% of the gaseous activity or.
aerosols produced will escape - j
- b. If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron ;
particles, the assumption shall be used that 10% of the aerosols produced escape.
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- c. For materials whose boiling point is above 55 C and where vapors formed by boiling this material could escape only through an undisturbed column of water above the core, the assumption shall be .
used that 10% of these vapors escape.
t
- 7. Each fueled experiment shall be controlled such that the total inventory of.
kxline isotopes 131 through 135 in the experiment is no greater than 1.5 !
curies and the maximum strontium-90 inventory is no greater than 5 ~ l millicuries.
- 8. If an experiment container falls and releases material which could damage .l the reactor fuel or structure by corrosion or other means, physical inspection - !
shall be performed to determine the consequences and the need for corrective action. !
- 9. Experiments shall not occupy adjacent fuel-element positions in the B- and ,
C-rings. l l
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Amendment No. 7
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- 1. ' This specificadon 'si intended to provide at least one level of review of any- l proposed operadon of the reactor in order to_ minimize the possibility of : !
operations of the reactor which could be dangerous or in violation of -
administrative procedures or the technical specifications.1he exception is . -[
made in the case of those few very well characterized operations which are !
necessary for routine checkout of the reactor and its systems, provided that [
those operadons have been defined by. written procedures which have been !;
reviewed and approved by the Reactor Supervisor and the Reactor -
Operations Committee. '
2.' This specification is intended te limit the reactivity of the system so that the : !
Safety Limit would not be exceeded even if the contribution to the total. j reactivity by the experiment reactivity should be suddenly removed.
E 3. This specification is intended to limit the power excursions which might be .;
induced by the changes in reactivity due to inadvertent motion of an unsecured experiment. Such excursions could lead to an inability to control :
the reactor within the limits imposed by the license. ;
- 4. ' This specification is intended to reduce the possibility of damage to the . ;
reactor or the experiments due to release of the listed materials. 1
- 5. This specification is intended to reduce the possibility of damage to the '
reactor in case of accidental detonation of the listed materials. i r
- 6. This specification is intended to reduce the severity of the results of l accidental release of airborne radioactive materials to the reactor room or- A the atmosphere. l
- 7. This specification is intended to reduce the severity of any possible release : l of these ilssion products which pose the greatest hazard to workers and the -
. general public. '!
- 8. This specification requires specific actions to determine the extent of damage following releases of materials. ' No theoretical calculations or : ;
evaluations are allowed. l l
- 9. This specification prevents serious modification of the neutron' distribution -j which could affect the ability of the control rods to perform their intended a function of maintaining safe control of the reactor, j
- I Amendment No. 7 20 I
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14; SURVEII f ANCE REOUIREMENTS
-l
. Allowable surveillance intervals shall not exceed the following:
biennially- not to exceed 30 months annually - not to exceed 15 months . _
semi annually - not to exceed seven and one-half months .
quarterly - not to exceed four months ,
monthly - not to exceed six weeks weekly - not to exceed 10 days - . ;
daily - must be done before the commencement of operation each day of operation >
Established frequencies shall be maintained over the long term, so, for example, any monthly l surveillance shall be performed at least 12 times during a calendar year of normal operation. If the _
reactor is not operated for a period of time exceeding any required surveillance interval, that -
surveillance task shall be pufexmed before the next operation of the reactor. Any surveillance
?
tasks which are missed more than once during such a shut-down interval need be performed only once before operation of the reactor _ Surveillance tasks scheduled daily or weekly which cannot ;
be performed while the reactor is operating may be postponed during continuous operation of the ;
reactor over extended times. Such postponed tasks shall be performed followhg shutdown after l the extended perkxi of continuous operation before any further operation, where each task shall be ' ;
performed only once no matter how many times that task has been postponed.
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p Amendment No. 7 2I -
4.1. Reactor Core Parnmeters Armlienbility Dese specifications apply to surveillance requirements for reactor core parameters.
Obiective he objective of these specifications is to ensure that the specifications of section 3.1 are satisfied.
Snecification f
The reactivity worth of each control rod, the reactor core excess, and the reactor shutdown margin shall be measured at least annually and after each time the core fuelis moved.
Basis Movement of the core fuel could change the reactivity of the core and thus affect both the core excess reactivity and the shutdown margin, as well as affecting the worth of the individual control rods. Evaluation of these parameters is therefore required after any such movement. Without any such movement the changes of. -
these parameters over an extended period of time and operation of the reactor have been shown to be very small so that an annual measurement is sufficient to ensure compliance with the specifications of section 3.1. ;
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Amendment No. 7 )
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4.2. Renetor controt and Saferv Svstems Annilenhilltv
! Dese specifications apply to the surveillance requirements of the reactor safety systems.
Ohlective The objecdve of these specifications is to ensure the operability of the reactor safety systems as described in section 3.3.
Snecifications
- 1. Control rod drive withdrawal speeds and co' ntrol red drop times shall be measured at least annually and whenever maintenance is performed or repairs are made that could affect the rods or control rod drives.
- 2. A channel calibration shall be performed for the NM1000 power level channel by thermal power calibration at least annually.
- 3. A channel test shall be performed at least daily and after any maintenance
- or repair for each of the six scram channels and each of the three interlocks listed in table 3.3A.
- 4. De control rods shall be visually inspected at least biennially.
Basc1 :
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- 1. Measurement of the control rod drop time and compliance with the .
-l
! specificadon indicates that the control rods can perform the safety function properly. Measurement of the control rod withdrawal speed ensures that -
the maximum reactivity ack11 tion rate specification will not be exceeded. j
- 2. Variations of the indicated power level due to minor variations of either of the two neutron detectors would be readily evident during day-to-day )
operation. The specification for thermal calibradon of the NM1000 channel provides assurance that long-term drift of both neutron detectors would be . I detected and that the reactor will be operated within the authorized power ;
range.
- 3. . The channel tests performed daily before operation and after any repair or maintenance provide timely assurance that the systems will operate I properly during operation of the reactor. 'i
- 4. Visual inspection of the control rods provides opportunity.to evaluate any .
corrosion, distortion, or damage that might occur in time to avoid malfunction of the control rods. Experience at the Dow TRIGA Reactor Facility since 1967 indicates that the surveillance specification is adequate to assure proper operation of the control rods. This surveillance complements the rod :
drop time measurements. "i l
Amendment No. 7 23
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4'.3. Coolant Svstem : ;
-[
Annlicability .j
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These specifications shall apply to the surveillance requirements for the reactor. f coolant system.- ;
Ohlective r The objective of these specifications is to ensure that the specifications of section - <
3.4 are satistled.
Snecifications ' !
- 1. The conductivity, pH, and the radioactivity of the pool' water shall be measured at least monthly. ,l
- 2. The level of the water in the pool shall be determined to be adequate on a '
weekly basis, j
- 3. The temperature of the coolant shall be monitored during operation of the reactor, i Bases
- 1. Experience at the Dow TRIGA Research Reactor shows that this . .
specification is adequate to detect the onset of degradation of the quality of ;
the pool water in a timely fashion. Evaluation of the radioactivity in the pool.- ~
water allows the detection of fission product releases from damaged fuel elements or damaged experiments.
- 2. Experience indicates that this specification is adequate to detect losses of -
~
pool water by evaporation.' ,
- 3. This specification will enable operators to take appropriate action when the l -
coolant temperature approaches the specified limit.
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'l Amendn.ent No. 7 24 -
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~4.4. Rntilation Monitorine Systems ;
Ann 11cahility I
.j Dese specifications apply to the surveillance requirements for the Continuous' Air -
l Monitor,(CAM) and the Area Monitor (AM), both located in the reactor room.
Ohlective ,j De objective of these specifications is to ensure the quality of the' data presented . _
by these two instruments. ;
p Snecifications y .
7
~
Rases j Dese specifications ensure that the named equipment can perform the required ,
functions when the reactor is operating and that deterioration of the instruments - i
- will be detected in a timely manner. Experience with these instruments has . :
shown that the surveillance intervals are adequate to provide the required.. :
assurance. ,
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i Amendment No. 7 25 a -
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, i ix; o k C 4.5. Facility Snecific Surveillance k Ann 11cability
- + '
' This specification shall apply to thE fuel elements of the Dow TRIGA Research Reactor.
' Ohlective -
De objective of this specification is to ensure that the reactor is not operated with . .
- damaged fuel elements. 9 Snecification - ,
Each fuel element shall be examined visually and for changes in transverse ?
E bend and length at least once each five years, with at least 20 percent of -
the fuel elements examined each year. If a damaged fuel element is '
identified, the entire inventory of fuel elements will be inspected prior to. ,
subsequent operations. ,
he reactor shall not be operated with damaged fuel except to detect and identify -
damaged fuel for removal. A TRIGA fuel element shall be considred damaged and ,
removed from the coreif: .
.P a) The transverse bend exceeds 0.125 inch over the length of the cladding. 1j b) ne length exceeds the original length by 0.125 inch.
c) A clad defect exists as indicated by release of fission products. 5 i
Visual examination of the fuel elements allows early detection of signs of '
deterioration of the fuel elements, indicated by signs of changes of corrosion -
patterns or of swelling, bending, or clongation. Experience at the Dow TRIGA ,
Research reactor and at other TRIGA reactors indicates that examination of a five-year cycle is adequate to detect problems, especially in TRIGA reactors that are not pulsed. A five-year cycle reduces the handling of the fuel elements and thus reduces the risk of accident or damage due to handling.
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Amendment No. 7 26
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1 4'6. ALARA .
Annlicability This specification applies to the surveillance of all reactor operations that could '
result in occupational exposures to ionizing radiation or the release of radioactive materials to the environment. .
Obiective The objective of this specification is to provide surveillance of all operations that could lead to occupational exposures to ionizing radiation or the release of radioactive materials to the environs.
Snecification The review of all operations shall include consideration of reasonable alternate operational modes which might reduce exposures to ionizing radiation or releases of radioactive materials.
Basis Experience has shown that experiments and operational requirements, in many.
cases, may be satisfied with a variety of combinations of facility options, power levels, time delays, and effluent or staff radiation exposures. The ALARA (As Low As Reasonably Achievable) principle shall be a part of overall reactor operation and detailed experiment planning.
Amendment No. 7
+ 27
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- 5. DESIGN FEATURES h-_ ,
L 5.1. Haa' tor site nna Buildini
' Annilenhility
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' These specifications shall apply to the Dow TRIGA Research Reactor.
, Ohlectives i
- The objectives of these specifications are to define the exclusion area and -
characteristics of the confinement.
m Snecifications .;
The minimum distance from the center of the reactor pool'to the boundary of the i exclusion area shall be 75 feet.
The reactor shall be housed in a room of about 6000 cubic feet volume designed to restrict leakage. ,
All air or other gas exhausted from the reactor room and from associated; experimental facilities during reactor operation shall be released to the environment at a minimum of 8 feet above ground level. >
b ~
nases 1he minimum distance from the pool to the boundary provides for dilution of effluents and for control of access to the reactor area. ,
Restriction ofleakage, in the event of a release of radioactive materials, can -
contain the materials and reduce exposure of the public.
Release of gases at a minimum height of 8 feet reduces the possibility of exposure
.l of personnel to such gases. ]
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Amendment No. 7 28 1
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.u 5.2. Reactor Coolant Svstem Annllenhility This specification applies to the Dow TRIGA Research Reactor.
Ohlective
'Ihe objective of this specification is to define the characteristics of the cooling I
system of this reactor.
Knecifiention >
! The reactor core shall be cooled by natural convective water flow.
IlaSis
' Experience has shown that TRIGA reactors operating at power levels up to 1000 kilowatt 2 can be cooled by natural convective water flow without damage of the fuel elements.
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i Amendment No. 7
- 29 A
X 5.3. Reactor Core and Fuel
?
' Ann 11cnhility i
Dese specificatlors shall be applicable to the Dow TRIGA Research Reactor.
Ohlectiva . i i
ne objective of these specifications is to define certain characteristics of the reactor in order to assure that the design and accident analyses will be correct. :
Snecification >
The fuel will be standard NRC-approved TRIGA fuel.- I The control elements shall have scram capability and shall contain borated graphite, boron carbide powder, or boron and its components in solid form as a -
poison in an aluminum or stainless steel cladding.
De reflector (excluding experiments and experimental facilities) shall be a ,
combination of graphite and water.
Eases ,
ne entire design and accident analysis is based upon the characteristics of -l TRIGA fuel.' Any other material would invalidate the findings of these analyses.
?
The control elements perform their function through the absorption of neutrons,
- thus affecting the reactivity of the system. Boron has been found to be a stable and effective material for this control.'
he reflector serves to conserve neutrons and to reduce the amount of fuel that-must be in the core to maintain the chain reaction. I f
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Amendment No. 7 30 !
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f 5.4. Fuel Stornoe Apphcability his specification applies to the Dow TRIGA Research Reactor fuel storage facilitjes.
~ '
Ohlective he objective of this specification is the safe storage of fuel.
Snecification All fuel and fueled devices not in the core of the reactor shall be stored in such a :
way that k,n shall be less than 0.8 under all conditions of moderation, and that will '
permit sufficient cooling by natural convection of water or air that temperatures shall not exceed the Safety Limit.
Lula A value of kenofless than 0.8 precludes any possibility ofinadvertent establishment of a self-sustaining nuclear chain reaction. Cooling which -
maintains temperatures lower than the Safety Limit prevents possible damage to the devices with subseque it release of radioactive materials.
Amendment No. 7 31 +
._ _ , _ _ . - _m. . . . _ . _ _ - _ _ . . _
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6; ADMINISTRATIVE CONTROLS
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' 6.1. OrFanization ;
I The Dow TRIGA Research Reactor is owned and operated by he Dow Chemical ' Company..
=
. The reactor is administered and operated through the Analytical Sciences Laboratory of the l Michigan Division of Dow Chemical USA and is located in 1602 Building of the Analytical ;
Sciences Laboratory at the Midland, Michigan location of the Michigan Division.: :
l 6.1.1. Structure :
The structure of the administration of the reactor is shown in figure 6.1. This structure cuts across the lines of management of he Dow Chemical Companyc l De individual responsible for radiation safety is the Radiation Safety Officer for .
f the reactor who reports on matters of radiation safety to the Radiation Safety Committee arxl to the Reactor Operations Committee. De review and audit .
functions are performed by the Reactor Operations Committee which is composed l of at least four persons including the Facility Director, the Radiation Safety Officer, and the Reactor Supervisor, j 6.1.2. Responsibility he day-to-day responsibility for the safe operation of the reactor rests with the -
Reactor Supervisor who is a licensed Senior Reactor Operator appointed by the Facility Director. De Reactor Supervisor may appoint equally-qualified
- individuals, upon notification of the Facility Director and the Reactor Operations Committee, to assume the responsibilities of the Reactor Supervisor. De Reactor .
Supervisor reports in a management sense to the Facility Director and within the 1' reactor organization to the Reactor Operations Committee.
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1 Amendnwnt No. 7
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3 Figure 6.L Administration . ,
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l Director,IIcalth and . Director, l 1
.. Environmental - AnalyticalSciences Science Laboratory l 1
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l Manager, Industrial Facility l
liygiene Research and Technology 1 Direen Y
Chair, RSC L . ..
Chair, ROC Reactor Operations Committee (ROC) a Supervisor, jL JL I
. Industrial Reactor Supervisor Ilygiene 4
1 Radiation Safety . Licensed :
Officer SROs and ROs Line management responsibilities 5
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v-Amendment No. 7 B
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6.1.3.' Staffing
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- The' minimum staffing when the reactor is not secured shall be:
a; a licensed Reactor Operator or Senior Reactor Operator in the control room, and' ,
i i? . b. a second person present at the facility able to carry out prescribed written . l
- instructions, and j
, c. a licensed Senior Reactor Operator in the facility or readily available on call and able to be at the facility within 30 minutes; The following operations require the presence of the Reactor Supervisor or a ,
' designated alternate: !
'I
- c. manipulations of fuel in the core; ;
- b. manual removal of control rods;-
- c. main'enance t performed on the core or the control rods;
- d. recovery from unexplained scrams, and- ,
- e. movement of any in-core experiment having an estimated reachvity value ,
greater than $0.75. ,
A list of reactor facility personnel by name and 'elephone t number shall be readily '
available in the control room for use by the operator, including management, radiation safety, and other operations personnel.
i 6.1.4. Selection and Training of Personnel -l J
'Ihe Reactor Supervisor is responsible for the training and requalification of the i facility Reactor Operators and Senior Reactor Operators. !
The selection, training, and requalification of operations personnel shall be i consistent with all current regulations.
Day-to-day changes in equipment, procedures, and specifications shall be l communicated to the facility staff as the changes occur, !
3 Amendment No. 7
. 34
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4 6.2. Review nnd Audit
. The review and audit functions shall be the responsibility of the Reactor Operations Committee (ROC)..
6.2.1. Charter and Rules s
- a. His Committee shall consist of the Facility Director, who'shall be designated the chair of this committee; the Radiation Safety Officer; the Reactor Supervisor; and one or more persons who are competent in the field of reactor operations, radiation science, or reactor / radiation engineering.
- b. A quorum shall consist of a majority of the members of the ROC. No more than one-half of the voting members present shall be members of the day-to-day
~
reactor operating staff,'
- c. The Committee shall meet quarterly and as often as required to transact
, business,
- d. Minutes of the meetings shall be kept as records for the facility.
- e. In cases where quick action is necessary members of the ROC may be polled by telephone for guidance and approvals.
- f. He ROC shall report at least twice per year to the Radiation Safety Committee.
Amendment No. 7
- 35
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- 6.2.2.. Review Functions -
The ROC shall review and approve: ;
- a. every experiment involving fissionable material;
- b, experiments or operations which would require a change.of core configuration, or -
a change in the equipment or apparatus associated with the reactor core or its -
Irradiation facilities, or a new piece of apparatus being mounted in the reactor ..
well; except that movement of the neutron source for the purpose of routinely.
~
i checking the instrumentation, or the movement of the neutron detectors to )
establish the proper calibration of the associated channels shall not require -
s review by the ROC; - 1 3
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s
- c. any other experiment or operation which is of a type not previously approved by . .
- the Committee; l i
- d. proposed changes in operating procedures, technical specifications, license, or 'I
. charter;- -
't
- e. violations of technical specifications, of the license, of internal procedures, and of instructions having safety significance; .l
- f. operating abnormalities having safety significance; .;
- g. reportable occurrences; - ;
i
'h. proposed changes in equipment, systems, tests, or experiments with respect to -;
unreviewed safety questions; and .j
- l. audit reports.
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J Amendment No. 7
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6.2.3. Audit Function j a 4
- a. 'Ihe ROC shall direct an annual audit of the facility operations for conformance to l the technical specifications, license, and operating procedures, and for the l' results of actions taken to correct those deficiencies which may occur in the . -;
reactor facility cquipment, systems, structures, or methods of operations that affect reactor safety. . .l l
'Ihis audit may consist of examinations of any facility records, review of
- - procedures, and interviews oflicensed Reactor Operators and Senior Reactor
~
Operators. j
' The audit shall be performed by one or more persons appointed by the ROC. l At - !
least one of the auditors shall be familiar with reactor operations. No person: ;
directly responsible for any portion of the operation of the facility shall audit t
-l hat operation.
A written report of the audit shall be submitted to the ROC 'within three months of -
the audit.
Deficiencies that affect reactor safety shall be reported to the Facility Director immediately.
- b. The ROC shall direct an annual audit of the facility emergency plan, security j plan, and the reactor operator requalification program. This audit may consist of the annual review of these plans for the requalification program.
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i Amendment No. 7
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6.3. Procedures Written procedures shall be reviewed and approved by the ROC for a - reactor startup, routine operation, and shu'tdown;
- b. emergency and abnormal operating events, including shutdown;
- c. fuelloading or unloading;
- d. control rod removal or installation; .
- e. checkout, calibration and determination of operability of reactor operating instrumentation and controls, control rod driver and area radiation and air particulate monitors; and
- f. preventive maintenance procedures.
Temporary deviations from the procedures may be made by the responsible Senior Reactor Operator or higher individual in order to deal with special or unusual .
circumstances. Such deviations shall be documented and reported immediately to the Reactor Operations Committee.
6.4. Exneriment Review and Annroval-
- a. Routine Experiments (as reviewed and defined by the ROC) shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor,
- b. Modified Routine Experiments shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor. 'Ihe written approval shall include documentation that the hazards have been considered by the '
reviewer and been found appropriate for this form of experiment. -
- c. Si ccial Experiments, those experiments that are neither Routine Experiments not.
Modified Routine Experiments, shall have the approval of both the Reactor Supervisor (or designated Assistant Reactor Supervisor) and the ROC.
= Experiments which require the approval of the ROC through sections 6.2.2.a.,
6.2.2.b., or 6.2.2.c. of the Technical Specifications are always Special Experiments.
b Amendment No. 7
- 38
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6.5. ' Renuired Actions 6.5.1. In case of Safety Limit violation:
- a. the reactor shall be shut down undl resumed operadons are authorized by.the
- US NRC; and
- b. the Safety Limit violadon shall be immediately reponed to the Facility Director s
or to a higher level and
- c. the Safety Limit violation shall be reported to the US NRC in accordance with section 6.6.2.; and
- d. a report shall be prepared for the ROC describing the applicable circumstances leading to the violadon including, when known, the cause and contributing factors, describing the effect of the violation upon reactor facil!ty components,'
systems, or structures and on the health and safety of personnel and the public, and describing corrective action taken to prevent recurrence of the' violadon. -
6.5.2. In case of a Reportable Occurrence of the type idendfled in section 1.28:
- a. reactor conditions shall be returned to normal or the reac%r shall be shut down; if the reactor is shut down operation shall not be resumed unless authorized by the Facility Director or designated alternate;
- b. the occurrence shall be reported to the Facility Director and to the US NRC as required per section 6.6.2.; and
- c. the occurrence shall be reviewed by the ROC at the next scheduled meeting.
4 A:neadment No. 7
- 59
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6.6. Renorts
.6.6.1, Operating Reports ,
A report shall be submitted annually, starting with the first quarter 1991' _ i performance of annual tasks, to the Radiation Safety Committee and to the .;
Director, Office of Nuclear Reactor Regulation, US NRC, Washington, DC, with a copy to the Regional Administrator, US NRC Region III, which shall include the
. following:
a) status of the facility staff, licenses, and training;.
b) a narrative summary of reactor operating experience, including the total megawatt-days of operation;
- c) tabulation of major changes in the reactor facility and procedures, and tabulation ;
of new tests and experiments that are significantly different from those :
performed previously and are not described in the Safety Analysis Report, !
i l including a summary of the analyses leading to the conclusions that no -
unreviewed safety questions were involved and that 10 CFR 50.59 (a) was applicable; d) the unscheduled shutdowns and reasons for them including, where applicable,
. corrective action taken to preclude recurrence; e) tabulation of major preventive and corrective maintenance operations having .
safety significance; f) a summary of the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge (the summary shall include to the extent practicable an estimate ofindividual radionuclides present in the effluent; if the estimated average release after dilution or; diffusion is less than 25% of the concentration allowed or recommended, only a statement to this effect is needed); and g) a summary of the radiation exposures received by facility personnel and visitors where such exposures are greater than 25 % of those allowed or recommended in 10 CFR 20.
9 Amendment No. 7
6.6.2. - Special Reports
- a. %ere shall be a report to US NRC Region III not later than the following ,
worldng day by telephone and confirmed in writing by telegraph or similar - :
conveyance to the Director of Nuclear Reactor Regulation, US NRC, with copy t to the Regional Administrator, Region III, US NRC to be followed by a written t report that describes the event within 14 days of; a violation of the Safety Limit;or .
a reportable occurrence (section 1.28).
- b. Dere shall be~a written report presented within 30 days to the Director of Nuclear Reactor Regulation, US NRC, with copy to the Regional ,
Administrator, Region III, US NRC, of:
permanent changes in the facility staff involving the reactor supervisor or the facility director; or significant changes in the transient or accident analysis report as described in the Safety Analysis Report.
- c. A written report shall be submitted to the Director of the Office of Nuclear -
Reactor Regulation, US NRC, with copy to the Regional Administrator, Region ,
Ill, US NRC, within 60 days after criticality of the reactor under conditions of a new facility license authorizing an increase in reactor power level, describing i the measured values of the operating conditions or characteristics of the reactor under the new conditions.
5 5
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Amendment No. 7
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-I 6.7. Records l
6.7.1. The following records shall be kept for a minimum period of Hve years:
i
- a. reactor operating logs; l
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- b. Irradiation request sheets;
- c. checkout sheets;
- d. maintenance records;
- c. calibration records; I records of reportable occurrences;
- g. fuel inventories, receipts, and shipments;
- h. minutes of ROC meetings; L records of audits; 1 facility radiation and contamination surveys; and
- k. surveillance activities as required by the Technical SpeciHcations.
6.7.2 Records of the retraining and requalification of Reactor Operators and Senior Reactor Operators shall be retained for at least one complete requalification schedule.
6.7.3. 'lhe following records shall be retained for the lifetime of the reactor: ,
- a. records of gaseous and liquid radioactive effluents released to the environment;
- b. records of the radiation exposure of all individuals monitored; and
- c. drawings of the reactor facility.
i Amendment No. 7 1 42
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k Reactivity .
. excess 1, 8 limits 2,8 - ..
' maxirnum positive insertion rate 10 .
. worth of an experiment 2 ,
Reactor cordrol and safety systems 10 operating 2 research 3 .r safety circuits 2 -
safety systems 3 secured 2 - !
shutdown 3.
Reactor Operations Committee 3,37 audit function 38 +
charter and rules 37 Reactor Supervisor 37 .
Records 41 .
Reference Core Condition 3 Reflector 32 ..j Report operating 40 Reportable Occurrence 3 Reports . ,
special 41 !
Research Reactor 3 :
Review and Audit 37 Review function 37 ROC 3 -
Rod, control 4 4 Routine Experiment 4 ,
Routine Experiments 39 RSC 2 .,
Safety - . l related instrumentation 10 Safety Limit 4,6 ;
Scram ,
setting 7 time 4 '
Secured Experiment 4 )
.' Shutdown Margin 4 l minimurn 8 ' I
- Special Experiment 4 Special Experiments 39 i
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d Amendment No. 7 44 l
4
ALARA 1 surveillance 29
- " ' Audit function 38
' Calibration
- channel 1 thermal' power 24 Channel 1
' calibration 1
' area monitor 27- .
continuous air monitor 27.
. thermal power 24 -
!' check 1 test 1 area monitor 27 continuous air monitor 27 . ;
scrams and interlocks, surveillance 24 Check channel 1 Confinement '1 Control rod l inspection 24 l Control rods -
composition 32 ,
Core Configuration 9 j
< Excess Reactivity 1,8 !
Experiment 1 )
modified routine 2,39 )
movable 2 routine 4,39 secured 4 special 4,39 specifications 19 Experimental Facilities 1 Fuel element aluminum 9 surveillance 28 -
temperature design limit 6 TRIGA 5,9 Fuel storage 33 Irradiation facility incore 9 LCO 1 Limiting Conditions for Operation 1, 8 Limiting Safety System Setting 1,3,6,7 LSSS 1 Measured Value 2
- Minimum shutdown margin 8 Modified Routine Experirnents 2,39 Movable Experiment 2
- Occurrence, reportable 3 Operable 2 Operating 2 Procedures 38 Radiation monitoring system
- specifications 18 Radiation Safety Committee 2 4 Radiation Safety Officer 37 !
Amendment No. 7 43
VX-n 1
j Specification 36 administrative controls )
' audit function 38 i experiment review and approval 39 ;
operating report 40 j organization 34 procedures 38
. records 41 s required actions 39 review and audit 37 ;
review functions 37
'special reports 41 3: staffing 36 ;
- structure 35 ,
airborne radioactivity from experiments 20 aluminum-clad fuel element 9 analog safety channel 10 -
arrangement of fuel 9 -
confinement -
ventilation system 17 corrosive,' reactive, explosive, liquid fissionable materials 19 design features control element composition 32 core cooling 31 fuel storage 33.
pool to boundary 30 reactor room 30 reflector 32 release of gases 30 TRIGA fuel 32 experiment failure and inspection 20 experiments in the B and C rings 20 explosives, testing 19 fuel 9 fuel temperature (safety limit) 6 fueled experiments and radioactive inventory 20 limiting safety system setting 7 -
manual scram 12 maximurn control rod drop time 10 maximum core excess reactivity 8 maximum positive reactivity for experiments 19 maximum positive reactivity insertion rate 10 measuring
. water radioactivity 14 p measuring channel linear power 14 percent power 14 water temperature 14 wide-range Log and period 14 minimum shutdown margin 8 number of control rods 10
_ operability of measuring channels 10 operability of safety channels and interlocks 10 operation approval 19 pool water bulk temperature 15 conductivity 15 t
Amendment No. 7 4$ .
i e
n: '
depth over the core 15 pH 15 radioactivity 15 radiation monitoring system area monitor 18 continuous air monitor 18 Reactor Operations Committee 37 rod drive interlock 12
- scram setpoint power level 12 . <
reactor period 12 securing experiments 19 startup countrate interlock 12 surveillance ALARA 29 area monitor calibration 27 area monitor channel test 27 Continuous Air Monitor calibration 27 '
continuous air monitor channel test 27 control rod drive withdrawal speeds 24 control rod drop times 24 control rod inspection 24 control rod worth 23 fuel element examination 28 interlocks 24 log power channel 24 operation with damaged fuel 28 pool water conductivity 26 pool water level 26 pool water pH 26 pool water radioactivity 26 pool water temperature 26 reactor coolant system 26 reactor core excess 23 scram channels 24 shutdown margin 23 !
thermal power calibration 24 surveillarce intervals 22 watchdog 12 water 9 wide-range linear channel high voltage 12 wide-range log channel high voltage 12 Staffing minimum 36 Surveillance 22 Test channel 1 TRIGA Fuel Element 5,32 .
damage specifications 28 Ventilation system specification 17 Amendment No. 7 46
_ _ _ _ .