ML20214C575

Proposed Tech Specs for Dow Triga Research Reactor

ML20214C575
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Site:	Dow Chemical Company
Issue date:	12/31/1986
From:	DOW CHEMICAL CO.
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ML20214C494	List: ML20214C492 ML20214C512 ML20214C535 ML20214C541 ML20214C549 ML20214C575
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TECHNICAL SPECIFICATIONS I

FOR THE DOW TRIGA RESEARCH REACTOR I

1986 I

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8611210156 861114 PDR ADOCK 05000264 P

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I 1.

DEFINITIONS 1.1.

ALARA - ALARA (As Low As Reasonably Achievable) is a device I

which helps to assure minimal exposure of Dow employees, contractors, and visitors to ionizing radiation and to minimize the uncontrolled release of radioactive materials through the establishment of procedures, through training programs, and through the development of proper attitudes and work habits.

1.2.

Channel - A channel is the combination of sensors, electronic circuits, and output devices connected by the appropriate communications network in order to measure and I

display the value of a parameter.

1.3.

Channel Test - A channel test is the introduction of a I

signal into a channel for verification of the operability of the channel.

1.4.

Channel Calibration - A channel calibration is the process I

of adjusting the parameters of a channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration I

shall encompass the entire channel, including equipment actuation, alarm, or trip and shall include a channel test.

1.5.

Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. The verification shall include comparison of the channel with other independent channels or systems measuring the same variable, whenever possible.

1.6.

Excess Reactivity - Excess reactivity is that amount of reactivity that would exist if all control rods were moved I

to the maximum reactive position from the condition where the reactor is exactly critical.

1.7.

Experiment - An experiment is (a) any device or material, not normally part of the reactor, which is introduced into the reactor for the purpose of exposure to radiation, (b) any operation which results in a change in the reactivity of the reactor, or (c) deliberate operations which produce changes in the core configuration or are likely to do so.

1.8.

Routine Experiment - A routine experiment is an experiment which involves operations under conditions which have been extensively examined in the course of the reactor test l

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I 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 I

defined by the Reactor Operations Committee. Normal operations of the reactor used for routine checkouts, authorized by approved written procedures, are routine experiments unless defined otherwise.

1.9.

Modified Routine Experiments - Modified routine experiments are experiments which have not been designated as routine I

experiments or which have not been performed previously, but are similar to routine experiments in that the hazards are neither significantly different from nor greater than the hazards of the corresponding routine experiment.

1.10.

Special Experiments - Special experiments are experiments which are neither routine experiments nor modified routine I

experiments. Deliberate movement of fuel elements and manual operation of control rods are always special experiments.

1.11.

TRICA Fuel Element - A TRIGA fuel element is a sealed unit containing the (U,Zr)H fuel for the reactor. The uranium is enriched to less thEn 20% in 235-U, and the fraction of I

hydrogen depends on the cladding of the fuel element. This type of fuel element was developed by General Atomics (now GA Technologies) for use in research and test reactors.

1.12.

Limiting Condition for Operation (LCO) - An LC0 is an administratively established constraint on equipment and operational characteristics which shall be observed during operation of the reactor.

1.13.

Limiting Safety System Setting (LSSS) - An LSSS is the actuating level for automatic protective devices related to those variables having significant safety functions.

I 1.14.

Measured Value - A measured value is the value of a parameter as it appears on the output of a channel.

1.15.

Operable - A component or system is operable if it is I

capable of performing its intended function.

1.16.

Operating - A component or system is operating if it is performing its intended function.

1.17.

Radiation Safety Committee (RSC) - The RSC is that group of people which is chartered by The Dow Chemical Company to be I

responsible for, among other things, the license for the Dow TRIGA Research Reactor facility.

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Reactivity Limits - The reactivity limits are those limits imposed on reactor core excess reactivity. Quantities are referenced to a Reference Core Condition.

1.19.

Reactivity Worth of an Experiment - The reactivity worth of an experiment is the maximum absolute value of the I

reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.

1.20.

Reactor Operating - The reactor is operating whenever it is not secured or shutdown.

1.21.

Reactor Secured - The reactor is secured whenever:

a) it contains insufficient fissile material or I

moderator present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or b) the console switch is in the "off" position, the key is removed from the switch, and the key is in I

the control of a licensed reactor operator or stored in a locked storage area; and sufficient control rods are inserted to assure that the reactor is suberitical by a margin greater than

$1.00 cold, without xenon; and no work is in progress involving core fuel, core structure, installed control rods or control rod drives unless 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 exceeding $1.00.

1.22.

Reactor Shutdown - The reactor is shutdown if it is subcritical by at least one dollar in the Reference Core I

Condition and the reactivity worth of all experiments is accounted for.

I 1.23.

Reactor Operations Committee (ROC) - The ROC is that group of people charged with direct oversight of the reactor operations, including both review and audit functions.

1.24.

Reactor Safety Systems - The reactor safety systems are those devices, including associated input channels and actuation elements, which are designed to initiate l

automatic reactor protection or to provide information for initiation of manual protective action.

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1.25.

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 I

negligible (less than S.30).

1.26.

Rod, Control - A control rod is a device containing neutron I

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.

1.27.

Scram - A scram is an unanticipated shutdown of the reactor which may be induced automatically or manually.

1.28.

Scram Time - Scram time is the interval between the induction of a scram signal and the complete insertion of all of the control rods.

1.29.

Shutdown Margin - Shutdown margin is the negative 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.

1.30.

Shall, Should, and May - The word "shall" is used to denote I

a requirement, the word "should" denotes a recommendation, and the word "may' denotes permission, neither a requirement nor a recommendation.

1.31.

Steady State Mode - The reactor is in the steady state mode when it is operating at criticality, neither supercritical (reactivity greater than zero) nor subcritical (reactivity less than zero).

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I 2.

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1.

Safety Limit (SL)

Applicability This specification applies to the temperature of the reactor fuel.

Objective The objective of this specification is to define the maximum fuel temperature that can be permitted with I

confidence that no damage to the fuel element will result.

Specification The temperature in any fuel element in the Dow TRIGA Research Reactor shall not exceed 500 C under any conditions of operation.

Basis 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. The pressure is caused by the heating of air, fission product gases, and hydrogen from the dissociation of the fuel-I moderator. The magnitude of this pressure is determined by the temperature of the fuel element and by the hydrogen content. Data indicate that the stress in the cladding due to hydrogen pressure from the

,I dissociation of ZrH will remain below the ultimate stress provided thai the fuel temperature does not exceed 1000 C and the fuel cladding is water cooled.

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. Analysis and I

measurements on TRIGA reactors shows that a power level of 1000 kW corresponds to a peak fuel temperature of about 400 C.

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Limiting Safety System Settings (LSSS)

Applicability This specification applies to the reactor scram setting which prevents the reactor fuel temperature from I

reaching the safety limit.

Objective The objective of this specification is to provide a reactor scram to prevent the safety limit from being reached.

Specification The LSSS shall not exceed 330 kW as measured by either of the two linear power level channels.

Basis The LSSS which does not exceed 330 kW provides a considerable safety margin. This value is one-third of that power level at which measurements have shown a fuel temperature of 400 C.

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 I

the core. The safety margin should be ample to compensate for other uncertainties, including power transients during otherwise steady-state operation, and should be adequate to protect aluminum-clad fuel elements from cladding failure due to temperature and pressure effects.

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LIMITING CONDITIONS FOR OPERATION (LCO) 3.1.

Reactivity Limits Applicability These specifications shall apply to the reactor at all times that it is in operation.

Objective The purpose of the objectives is to ensure that the reactor can be controlled and shut down at all times and that the safety limit will not be exceeded.

Specifications The shutdown margin shall be greater than S.50 in the reference core condition.

The excess reactivity measured at five watts in the reference core condition, with or without experiments in place, shall not be greater than $3.00.

Bases 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 I

the power level safety circuits and administrative controls.

3.2.

Reactor Control and Safety Systems Applicability these specifications apply to the reactor control and safety systems and safety-related instrumentation that must be operating when the reactor is in operation.

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I Objective I

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.

Specifications 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.

I The reactor safety channels and the interlocks shall be operable in accordance with table 3.2A.

Positive reactivity insertion rate by control rod motion shall not exceed.S.20 per second.

Bases The control rod drop time specification assures that the reactor can be shutdown promptly when a scram signal is initiated. The value of the control rod drop time is adequate to assure safety of the reactor.

Use of the specified reactor safety channels, set I

points, and interlocks given in table 3.2A assures protection against operation of the reactor outside the safety limits. Table 3.2b describes the bases for the specifications given in table 3.2A.

The specification of maximum positive reactivity insertion rate helps assure that the Safety Limit is I

not exceeded.

_g-TABLE 3.2A.

MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS Scram Channels Scram Channel Minimum Operable Scram Setpoint Reactor Power Level 2

Not to exceed 110% of I

full licensed steady-state power Reactor Period 1

Not less than 7 seconds Wide-Range Linear Channel 1

Loss of power supply Detector Power Supply voltage to detector Vide-Range Log Channel 1

Loss of power supply Detector Power Supply voltage to detector Manual Scram i

Not applicable Interlocks

_ Interlock /Chr.nnel 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 I

I TABLE 3.2B REACTOR SAFETY CHANNELS AND INTERLOCKS I

Scram Channels Scram Channel Bases Reactor Power Level Provides assurance that the reactor will be shut down automatically before the safety limit can be exceeded Reactor Period Prevents operation in a regime in which transients could cause the safety limit to be exceeded Reactor Power Channel Provides assurance that the reactor Detector Power Supplies cannot be operated without power to the neutron detectors which provide input to the wide-range linear power channel and the wide-range log power channel Manual Scram Allows the operator to shut the reactor down at any indication of unsafe or abnormal conditions Interlocks I

Interlock / Channel Bases Startup Countrate Provides assurance that the signal in the log power 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 3.3.

Coolant System Applicability These specifications apply to the quality of the coolant in contact with the fuel cladding, to the level of the coolant in the pool, and to the bulk temperature of the coolant.

Objectives The objectives of this specification are:

I to minimize corrosion of the cladding of the fuel elements and minimize neutron activation of dissolved materials, to detect releases of radioactive materials to the coolant before such releases become significant, to ensure the presence of an adequate quantity of cooling and shielding water in the pool and to prevent thermal degradation of the ica exchange resin in the purification system.

Specifications The conductivity of the pool water shall not exceed 5 pahos/cm averaged over one month.

The amount of radioactivity in the pool water shall not exceed 0.1 Ci/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.

Bases Increased levels of condue.tivity in aqueous systems both indicate the presence of corrosion products and promote more corrosion. Experience with water quality control at many reactor facilities, including the past 19 years of operation of the Dow TRIGA Research Reactor, has shown that maintenance within the specified limit provides acceptable control.

Maintaining low levels of dissolved electrolytes in the

I l pool water also reduces the amount of induced radioactivity, in turn decreasing the exposure of personnel to ionizing radiation during operation and maintenance. Both of 'these results are in accordance i

with the ALARA program.

Monitoring the radioactivity in the tool 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 radiation in accordance with the ALARA program.

Maintaining the bulk temperature of the coolant below the specified limit assures minimal thermal degradation of the ion exchange resin.

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3.4.

Radiation Monitoring Systems Applicability These specifications apply to the radiation monitoring information available to the reactor operator during operation of the reactor.

Objective The objective of these specifications is to ensure that the reactor operator has adequate information to assure safe operation of the reactor.

Specifications The Continuous Air Monitor (CAM) (with readout meter and audible alarm) in the reactor room must be l

operating during operation of the reactor.

The 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 on or around the reactor core or experimental facilities.

During short periods of repair to this monitor reactor operations or work on or around the core or experimental facilities may continue while a portable I

gamma-sensitive ion chamber is utilized as a temporary I

substitute.

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Bases The radiation monitors provide information of existing I

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.

3.5.

Experiments Applicability These specifications apply to experiments installed in the reactor and its experimental facilities.

Objective The objective of these specifications is to prevent damage to the reactor or excessive release of radioactive materials in case of failure of an experiment.

Specifications 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 I

of the reactor for the purpose of performing routine checkouts, where written procedures exist for those operations, shall be authorized by the written procedures. An operation shall not be approved unless the evaluation 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 S2.00.

This includes the potential reactivity which might result from experimental malfunction, experiment flooding or voiding, or the removal or insertion of I

experiments.

3.

Experiments having reactivity worths of greater than $1.00 shall be securely located or fastened to prevent inadvertent movement during reactor operation.

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Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials or liquid fissionable materials shall be doubly encapsulated.

5.

Explosive materials such as gunpowder, dynamite, I

TNT, nitroglycerin, or PETN in quantities greater than 25 milligrams shall not be irradiated in the reactor or experimental facilities without out-of-core tests which shall indicate that, with the containment provided, no damage to the reactor or its components shall occur upon detonation of the explosive. Explosive materials in quantities less than 25 milligrams may be irradiated without out-of-core tests provided that the pressure produced in the experiment container upon detonation of the explosive shall be calculated to be less than the design pressure of the container.

6.

Experiment materials, except fuel materials, which could off-gas, sublime, volatize or produce aerosols under (a) normal operating conditions of the experiment or the reactor, (b) credible accident conditions in the reactor or (c) 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 limits of Appendix B of 10 CFR Part 20.

The following assumptions should be used in calculations regarding experiments:

a.

If the effluent from an experiment 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.

b.

If the effluent from an experiment facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, the assu=ption shall be used that 10% of the aerosols produced escape.

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.

7.

Each fueled experiment shall be controlled such that the total inventory of iodine 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 millicuries.

8.

If an experiment container fails and releases material which could damage 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.

Bases 1.

This specification is intended to provide at least onc level of review of any proposed operation 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. The exception is made in the case of those few very well characterized I

operations which are necessary for routine checkout of the reactor and its systems, provided that those operations 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 to limit the reactivity of the system so that the Safety Limit would not be exceeded even if the contribution to the total reactivity by the experiment reactivity should be suddenly removed.

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.

I 4.

This specification is intended tc reduce the possibility of damage to the reactor or the experiments due to release of the listed materials.

5.

This specification is intended to reduce the possibility of damage to the reactor in case of accidental detonation of the listed materials.

6.

This specification is intended to reduce the severity of the results of accidental release of I

airborne radioactive materials to the reactor room or the atmosphere.

I 7.

This specification is intended to reduce the severity of any possible release of these fission 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.

9.

This specification prevents serious modification of the neutron distribution which could affect the ability of the control rods to perform their intended function of maintaining safe control of the reactor.

4. SURVEILLANCE REQUIREMENTS Allowable surveillance intervals shall not exceed the following:

biennially - not to exceed 30 months I

annually - not to exceed 15 months semi-annually - not to exceed seven and one-half months monthly - not to exceed six weeks weekly - not to exceed 10 days daily - must be done before the co=mencement of operation each day of operation Established frequencies shall be maintained over the long term, so, for example, any monthly 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 performed 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 I

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of the reactor. Surveillance tasks scheduled daily or weekly I

which cannot be performed while the reactor is operating may be l

postponed during continuous operation of the reactor over extended times. Such postponed tasks shall be performed following shutdown after the extended period of continuous operation before any further operation, where each task shall be l

performed only once no matter how many times that task has been l

postponed.

4.1.

Reactivity Limits Applicability l

l These specifications apply to surveillance requirements i

for reactivity limits.

I Objective The objective of these specifications is to ensure that the specifications of section 3.1 are satisfied.

Specification I

The reactivity worth of each control rod and the reactor shutdown margin shall be measured at least annually and after each time the core fuel is moved.

Basis Movement of the core fuel could change the reactivity l

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 l

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.

4.2.

Reactor Control and Safety Systems Applicability These specifications apply to the surveillance requirements of the reactor safety systems.

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Objective The objective of these specifications is to ensure the I

operability of the reactor safety systems as described in section 3.2.

Specifications 1.

Control rod drop times shall be measured at least annually and whenever maintenance is performed or repairs are made that could affect the rod drop times.

I 2.

A channel calibration shall be performed for the wide-range linear power 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 two interlocks listed in table 3.2A, and the log power channel.

4.

The control rods shall be visually inspected at least biennially.

Bases 1.

Measurement of the control rod drop time and compliance with the specification indicates that the control rods can perform the safety function properly.

2.

Variations of the indicated power level due to minor variations of any one of the three neutron detectors would be readily evident during day-to-day operation. The specification for thermal i

ll calibration of the wide-range linear channel l

provides assurance that long-term drift of all IE three neutron detectors would be 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 properly during operation of the reactor.

4.

Visual inspection of the control rods provides I

opportunity to evaluate any corrosion, distortion, I

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____ I or damage that might occur in time to avoid malfunction of the control rods. Experience at the Dow TRIGA Reactor Facility over the past 19 years I

indicates that the surveillance specification is I

adequate to assure proper operation of the control l

rods. This surveillance complements the rod drop time measurements.

4.3.

Coolant System Applicability These specifications shall apply to the surveillance requirements for the reactor coolant system.

Objective The objective of these specifications is to ensure that the specifications of section 3.3 are satisfied.

Specifications 1.

The conductivity and the radioactivity of the pool water shall be measured monthly.

2.

The level of the water in the pool shall be determined to be adequate on a weekly basis.

3.

The temperature of the coolant shall be monitored during operation of the reactor.

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 I

evaporation.

3.

This specification will enable operators to take appropriate action when the coolant temperature approaches the specified limit.

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4.4.

Radiation Monitoring Systems Applicability These specifications apply to the surveillance requirements for the Continuous Air Monitor (CAM) and the Area Monitor (AM), both located in the reactor room.

Objective The objective of these specifications is to ensure the quality of the data presented by these two instruments.

Specifications 1.

A channel calibration shall be made for the CAM and the AM annually.

2.

A channel test shall be made for the CAM and the AM weekly.

Bases These specifications ensure that the named equipment can perform the required functions when the reactor is operating and that deterioration of the instruments will be detected in a timely manner. Experience with these instruments has shown that the surveillance intervals are adequate to provide the required assurance.

4.0.

Facility Specific Surveillance Applicability This specification shall apply to the fuel elements of the Dow TRIGA Research Reactor.

Objective The objective of this specification is to ensure that the reactor is not operated with damaged fuel elements.

Specification Each fuel element shall be visually examined annually.

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I Basis 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 elongation.

5.

DESIGN FEATURES 5.1.

Reactor Fuel Applicability This specification applies to the fuel elements to be used in the reactor core.

Ob[ective The objective is to assure that the design and fabrication of the fuel elements is consistent with the fuel elements discussed in the Safety Analysis Report.

Specification The fuel will be standard TRIGA fuel as described in the Safety Analysis Report.

Basis Measurements on such standard TRIGA fuel elements have shown that operation of the Dow TRIGA Research Reactor within the Technical Specifications ensures a conservative limitation with respect to the Safety Limit.

5.2.

Fuel Storage Applicability This specification shall be applicable to the storage of fuel including fueled experiments and fuel devices.

Objective The objective of this specification shall be to prevent storage of any fuel or fueled devices in a manner which could allow the establishment of a critical assembly or the production of temperttures in excess of the Safety Limit.

I Specification All fuel and fueled devices shall be stored in such a way that k shall be less than 0.8 under all conditions *b moderation, and that will permit sufficient cooling by natural convection of water or air that temperatures shall not exceed the Safety Limit.

Basis At a k of less than 0.8 no assembly is critical, and neutronNultiplicationislow. Thus there is no danger that such an assembly will pose a danger due to a nuclear fission chain reaction. Attention to the cooling characteristics of air and water provides assurance that temperatures in the devices do not rise to dangerous values due to radioactive heating.

6.

ADMINISTRATIVE CONTROLS 6.1.

Organization The Dow TRIGA Research Reactor is owned and operated by The Dow Chemical Company. The reactor is administered and operated through the Analytical Laboratory of the Michigan Division of Dow Chemical USA and is located in 1602 Building of the Analytical Laboratory ac the Midland, Michigan location of the Michigan Division.

6.1.1.

Structure The structure of the administration of the reactor is shown in figure 6.1.

The individual at level 1 is the chairman of the Radiation Safety Committee.

The individual at level 2, the facility director, is the member of management within the Analytical Laboratory whose renponsibilities include

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activities at 1602 Building. The individual responsible for radiation safety is the Radiation Safety Officer for the reactor who reports on matters of radiation safety to the individuals at both level 1 and level 2.

The review and audit functions are performed by the Reactor Operations Committee which is composed of at least four persons including the individual at level 2, the Radiation Safety Officer, and the Reactor Supervisor. The structure of the reactor organization cuts across the lines of management of The Dow Chemical Company.

I Figure 6.1.

Administration Level 1 l

Chairman, Radiation Safety Committee A

Radiation Safety Committee A

Radiation Safety Officer Level 2 Facility Director;

Chairman, I

Reactor Operations Committee A

Reactor i

Operations I

Committee Audit Review 6.1.2.

Responsibility The 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. The Reactor Supervisor eay appoint one or two equally-qualified individuals, upon notification of the Facility Director and the Reactor Operations Committee, to assume the responsibilities of the Reactor Supervisor. The Reactor Supervisor reports in a management sense to the Facility Director and within the reactor organization to the Reactor Operations Committee.

6.1.3.

Staffing The minimum staffing when the reactor is not secured shall be:

a. a licensed Reactor Operator or Senicr Reactor Operator in the control room, and

b. a second person present at the facility, and

c. a licensed Senior Reactor Operator in the facility or readily available on call.

The following cperations require the presence of the Reactor Supervisor or a designated alternate:

a. manipulations of fuel in the core;

b. manual removal of control rods;

c. maintenance performed on the core or the control rods;

d. recovery from unexpir.ined scrams, and

e. movement of any in-core experiment having an estinated value greater than $1.00.

l 8.1.4.

Selection and Training of Personnel The Reactor Supervisor is responsible for the training and requalification of the facility Reactor Operators and Senior Reactor Operators.

A program shall be established for the purpose of training and documenting the training of the facility Reactor Operators and Senior Reactor Operators in a regular requalification program.

Day-to-day changes in equipment, procedures, and specifications shall be communicated to the facility staff as the changes occur.

6.2.

Review and Audit I

The review and audit functions shall be the responsibility of the Reactor Operations Committee (ROC).

6.2.1.

Charter and Rules

a. This 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.

b. A quorum shall consist of a majority of the members of the R00.

c. The Committee shall meet quarterly and as often as required to transact business.

d. Minutes of the meetings shall be kept as records I

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. The ROC will be responsible for determining whether a proposed change, test, or experiment would constitute an unreviewed safety question or a change of the technical specifications, as required by 10 CFR 50.

g. The ROC shall report at least twice per year to the Radiation Safety Committee.

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t 6.2.2.

Review Functions The ROC shall review:

a. every experiment involving fissionable material;

b. experiments 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 checking the instrumentation, or the movement of the neutron detectors to establish the proper calibration of the associated channels shall not require review by the ROC;

c. any other experiment which is of a type not previously approved by the Committee;

d. proposed changes in operating procedures;

e. violations of technical specifications, of the license, of internal procedures, and of instructions having safety significance;

f. operating abnormalities having safety I

significance;

g. reportable occurrences; and

h. audit reports.

Experiments reviewed by the ROC may be performed provided committee approval is granted and documented.

6.2.3.

Audit Function

a. The ROC shall direct an annual audit of the facility operations for conformance to the technical specifications and applicable license conditions, and for the results of actions taken to correct those deficiencies which may occur in the reactor facility equipment, systems, structures, or methods of operations that affect reactor safety.

This audit may consist of examinations of any facility records, review of procedures, and interviews of licensed Reactor Operators and Senior Reactor Operators.

I

_ The audit shall be performed by one or more persons appointed by the ROC. 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 that operation.

A written report of the audit shall be submitted to the ROC in a timely manner.

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 and security plans. This audit may consist of the annual review of these plans for the requalification program.

6.3.

Procedures Written procedures shall be reviewed and approved by the ROC for:

a.

reactor startup, routine operation, and shutdown; b.

emergency and abnormal operating events, including shutdown; c.

fuel loading or unloading; d.

control rod removal or installation; and e.

checkout, calibration and determination of operatility of reactor operating instrumentation and controls, control rod drives and area radiation and air particulate monitors.

6.4.

Experiment Review and Approval Experiments which involve the operation of the reactor a.

for the purpose of routine checkouts and which are described by written procedures approved by the ROC may be performed by any Reactor Operator or Senior Reactor Operator without further approval, b.

Routine Experiments (as reviewed and defined by the ROC) and experiments involving operation of the reactor for training or operating experience shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor.

, c.

Modified Routine Experiments shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor. The written approval shall include documentation that the hazards have been considered by the reviewer and been found appropriate for this form of experiment, d.

Special Experiments, those experiments that are neither Routine Experiments nor Modified Routine Experiments, shall have the approval of both the Reactor Supervisor (or designated Assistant Reactor Supervisor) and the R00. Experiments which require the approval of the ROC through sections 6.2.2.a. or 6.2.2.b. of the Technical Specifications are always Special Experiments.

6.5.

Required Actions 6.5.1.

In case of Safety Limit violation:

a. the reactor shall be shut down until resumed operations are authorized by the US NRC;

b. the Safety Limit violation shall be im=ediately reported to the Facility Director or to a higher level;

c. The Safety Limit violation shall be reported to the US NRC; and

d. a report shall be prepared for the ROC describing the applicable circumstances leading to the violation including, when known, the cause and contributing factors, describing the effect of the violation upon reactor facility 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 violation.

6.5.2.

In case of Reportable Events of the type identified in sections 6.6.2(1)b. and 6.6.2(1)c. :

a. reactor conditions shall be returned to normal or the reactor shall be shut down;

b. the occurrence shall be reported to the Facility Director and to the US h20 as required; and

c. the occurrence shall be reviewed by the ROC at the next scheduled meeting.

r 6.6.

Reports 6.6.1. ~ Operating Reports A report shall be submitted annually to the Radiation Safety Committee which shall include the I

status of the facility staff, licenses, and training; the usage of the reactor; scrams; major maintenance; and changes and inprovements involving the reactor control and monitoring systems.

6.6.2.

Special Reports

a. There shall be a report not later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to the US NRC to be followed by a written report that describes the event within 14 days of:

a violation of the Safety Limit:

release of radioactivity from the site above allowed limits; operation with actual safety-system settings for required systems less conservative than the I

limiting safety-system settings specified in the Technical Specifications; operation in violation of limiting conditions for operation established in the Technical Specifications unless prompt remedial action is taken; 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; or abnormal and significant degradation in reactor fuel, cladding, or c.colant boundary which could result in exceeding prescribed radiation exposure or release limits.

e b. There shall be a written report presented within 30 days to the US NRC of:

permanent changes in the facility involving level 1 or level 2 personnel; or I

significant changes in the transient or ar.cident analysis report as described in the Safety Analysis Report.

6.7.

Records 6.7.1.

The following records shall be kept for a minimum period of five years:

a.

reactor operating logs; b.

irradiation request sheets; c.

checkout sheets; d.

maintenance records; e.

calibration records; f.

records of reportable occurrences; g.

fuel inventories, receipts, and shipments; and h.

minutes of ROC meetings.

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.

The following records shall be retained for the lifetime of the reactor:

records of gaseous and liquid radioactive a.

effluents released to the environment; b.

records of the radiation exposure of all individuals monitored; and c.

drawings of the reactor facility.