ML20214T088

Tech Specs for Dow Triga Research Reactor,Facility License R-108

ML20214T088
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Site:	Dow Chemical Company
Issue date:	03/31/1987
From:	DOW CHEMICAL CO.
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NUDOCS 8706100114
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TECHNICAL SPECIFICATIONS-FOR THE DOW TRICA RESEARCH REACTOR FACILITY LICENSE R-108 MAY 1987 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 are included for information purposes only. The bases are not part of the Technical l Specifications and they do not constitute limitations or requirements

to which the licensee must adhere.

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1. DEFINITIONS 1.1.- ALARA - The ALARA (As Low As Reasonably Achievable) program is.a set of procedures which is intended to minimise occupational exposures to ionising radiation and releases of radioactive materials to the environment.

1.2. Channel - A channel is a combination of sensors, electronic circuits, and output devices connected by the appropriate e 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 output corresponds with acceptable accuracy to known. values of the parameter

< which the channel measures. Calibration shall encompass

! the entire channel, including equipment, actuation, alarm, or trip and shall include a Channel Test.

1.4.- 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.

1.5. Channel Test - A channel test is the introduction of a signal into a channel for verification of the operability j of the channel.

1.6. Confinement - Confinement is an enclosure of the facililty which controls the movement of air into and out of the facility through a controlled path.

I 1.7. Excess Reactivity - Excess reactivity is that amount of l reactivity that would exist if all control rods were moved

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

l 1.8. Experiment - An experiment is any device or material, not normally part of the reactor, which is introduced into the reactor for the purpose of exposure to radiation, or any operation which is designed to investigate non-routine l reactor characteristics.

1.9. Experimental Facilities include the rotary speciment rack, vertical tubes, pneumatic transfer systems, the central thimble, and the area surrounding the core.

' o 1.10. Limiting Conditions for Operation - Limiting Conditions for Operation (LCO) are administratively established constraints on equipment and operational characteristics which shall be adhered to during operation of the reactor.

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

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

1.13. Modified Routine Experiments - Modified routine experiments are 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 hasards are neither significantly different from nor greater than the hasards of the corresponding routine experiment.

I 1.14. Movable Experiment - A movable experiment is an experiment

, intended to be moved in or near the core or into and out of

the reactor while the reactor is operating.

! 1.15. Operable - A component or system is operable if it is 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

responsible for the license for the Dow TRIGA Research l Reactor facility.

1.18. 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 reactivity change that would occur as a result of inter.ded or anticipated changes or credible malfunctions that alter

experiment position or configuration.

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y 1.20. Reactor Operating - The reactor is operating whenever it is

.not secured or shutdown.

1.21. Reactor Safety Circuits - Reactor safety circuits are those circuits, including the associated input circuits, which are designed to initiate a reactor scraa.

1.22. Reactor Secured - The reactor is secured whenever:

a) it contains insufficient fissile material present in the reactor, adjacent experiments or control rods, to attain criticality under optinua 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 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 81.00 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 80.75.

1.23. Reactor Shutdown - The reactor is shutdown if it is subcritical by at least one dollar and the reactivity worth of all experiments is accounted for.

1.24. 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.

1.25. Reactor Safety Systems - Reactor Safety Systems are those systems, including associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

1.26. 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 8.30).

1.27. 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.

1.28. Reportable Occurence - A Reportable Occurence is any of the following which occurs during reactor operations a) Operation with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in Technical Specification 2.2.

b) Operation in violation of limiting conditions for operation established in the Technical Specifications.

c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.

d) Any unanticipated or uncontrolled change in reactivity greater than one dollar, e) Abnormal and significant degradation 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 administrative 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 specified in 10CFR20.

1.29. Rod, Control - A control 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.

r-1.30. 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 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 Reactor Operations Committee.

1.31. Safety Limit - A Safety Limit is a limit on an important process variable which is found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radisoctivity. The principal physical barrier is the fuel element cladding.

1.32. Scram Time - Scram Time is the time required to fully insert tliIe control rods following the actuation of a Limiting Safety System Setting.

1.33. Secured Experiment - A Secured Experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as the result of credible malfunctions.

1.34 Shall, Should, and May - The word 'shall" is used to denote a requirement, the word "should' denotes a recommendation, and the word 'may" denotes permission, neither a requirement nor a recommendation.

1.35. Shutdown Marsin - 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, i

1.36. Syecial Experiments - Special experiments are experirents wiich are neither routine experiments nor modified routine experiments.

1.37. TRICA Fuel Element - A TRIGA fuel element is a sealed unit

containing (U,Zr)H fuel for the reactor. The uranium is enrichedtolesstNan2095in235-Uandthefractionof 1 hydrogen is in the range of 1.0-1.1 for aluminum-clad TRICA

! elements and in the range of 1.6-1.7 for stainless-steel-clad TRIGA elements.

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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 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-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 dissociation of ZrH1 .6 will remain below the ultimate 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 equal the fuel temperature the fuel temperature design limit is 950 0 (M. T. Simnad, G.A. Project No. 4314, Reporte-117-833,1980).

Experience with operation of TRICA-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 TRICA fuel elements using ZrHi ,1 moderator have been analysed (S. C. Hawley and R. L. Kathren, NURE0/CR-2387, PNL-4028, credible Accident Analyses for TRICA and TRIGA-fueledReactors,1982). A loss-of-coolant analysis

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showed that in a typical graphite-reflected Mark TRIGA reactor fueled with 60 aluminum-clad fuel elements (Reed College) the maximum fuel temperature would be less than 150 C following infinite operation at 250 kilowatts terminated by the instantaneous loss of water. These temperatures are well below the region where the a + 6 + 7' to a + 6 phase change occurs in ZrHi ,1 (560 C).

2.2. Limitina Safety System Settings (LSSS)

Applicability This specification applies to the reactor scram setting which prevents the reactor fuel temperature from 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 300 kW as measured by the calibrated power channels.

Basis The LSSS which does not exceed 300 kW provides a considerable safety margin. 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. 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 tesperature and pressure effects.

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3. 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 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.

Specifications The reactor shall be shutdown by more than 8.50 with the most reactive control fully withdrawn, the other two control rods fully inserted, cold, no xenon, with all experiments accounted for.

The 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.

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 the power level safety circuits and administrative controls.

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3.2. Core Configuration a

Applicability This specification applies to the core configuration.

Objective The objective of this specification is to assure that

, the safety limit will not be exceeded due to power peaking effects.

Specifications The critical core shall be an assembly of standard NRC-approved stainless-steel-clad or aluminum-clad fuel elements in light water.

The fuel 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.

Bases I

Operation with standard NRC-spproved TRIGA fuel in the standard configuration ensures a conservative limitation with respect to the Safety Limit.

Placement of the aluminua-clad fuel element in the outer rings of the reactor core will help ensure that this element is not exposed to higher than average

. power levels, thus providing a greater degree of conscrvatism with respect to the Safety Limit for this one element.

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3.3. 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. l Objective ]

The objective of these specifications is to assure that 4

all reactor control and safety systems and' safety-related instrumentation are operable to minimum acceptable standards during operation of the reactor.

j Specifications 4

There shall be a minimum of three operable control rods in the reactor core. .

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$ 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.

5 The reactor safety channels and the interlocks shall be-operable in accordance with table 3.3A.

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 8.20 per second.

Bases

! The requirement for three operable control rods ensures that the reactor can meet the shutdown specifications.

The control rod drop time specification assures that the reactor can be shutdown promptly when a seras i 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 points, and interlocks given in table 3.3A assures protection against operation of the reactor outside the safety limits, i

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4 The requirement for the specified measurement circuits provides assurance that important reactor operation parameters can be monitored during operation.

The specification of maximum positive reactivity insertion rate helps assure that the Safety Limit is not exceeded.

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TABLE 3.3A.

MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS Scram Channels Scram Channel Minimum Operable Scram Setpoint Reactor Power Level 2 Not to exceed maximum licensed power Reactor Period 1 Not less than 7 seconds Wide-Range Linear Channel 1 Loss of power supply Detector Power Supply voltage to detector Wide-Range Log Channel 1 Loss of power supply Detector Power Supply voltage to detector Manual Scram 1 Not applicable Interlocks Interlock / Channel 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

TABLE 3.3A BASES FOR REACTOR SAFETY CHANNELS AND INTERLOCKS 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

-Interlock / Channel Bases i Startup Countrate Provides assurance that the signal l in the log power channel is adequate to allow reliable indication of the I

state of the neutron chain reaction

Rod Drive control Limits the maximum positive reactivity insertion rate i-

TABLE 3.3B MEASURING CHANNELS Measuring channel Minimum Number Operable Wide-range Log N 1 and Period Channel Power-Level Channel 1 (Linear)

Power-Level Channel 1 (PercentPower)

Water Radioactivity 1 Monitor Water Temperature 1 Monitor TABLE 3.3B BASES FOR MEASURING CHANNELS Measuring Channel Basis Wide-Range Log N Provides assurance that the y and Period Channel reactor power level and period can be adequately monitored.

Power-level Channel Provides assurance that the reactor (Linear) power level can be adequately monitored.

Power-level Channel Provides assurance that the reactor (Percent Power) power level can be adequately monitored.

Water Radioactivity Provides assurance that the water Monitor radioactivity level can be adequately monitored.

Water Temperature Provides assurance that the water Monitor temperature can be adequately monitored.

3.4. 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:

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 ion exchange resin in the purification system.

Specifications The conductivity of the pool water shall not exceed 5 pahos/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 pCi/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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Bases Increased levels of conductivity in aqueous systems both indicate the presence of corrosion products and promote more corrosion. Experience with water quality control at many reactor f acilities, including the past 19 years of operation of the Dow TRIGA Research Reactor, has shown that maintenance within the i specified limit provides acceptable control.

Maintaining low levels of. dissolved electrolytes in the pool water also reduces the maount of induced radioactivity, in turn decreasing the exposure of personnel to ionising radiation during operation and maintenance. Both of these results are'in accordance with the ALARA program.

Monitoring the pH of the pool water provides early j 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 ionising 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 ionising radiation in accordance with the ALARA program.

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

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3.5. Confinement Applicability This specification applies to the reactor room confinement.

Objective The objective of this specification is to mitigate the consequences of possible release of radioactive materials to unrestricted areas.

Specification The ventilation system shall be operable and the external door 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.

Basis 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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3.6. 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 resctor operator has adequate information to assure safe operation of the reactor.

l Specifications A Continuous Air Monitor (CAM) (with readout meter and audible alarm) in the reactor room must be operating l' 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, not to exceed one working week, reactor operations or work on or around the core or experimental facilities may continue while a portable gamma-sensitive ion chamber is utilised as a temporary substitute, provided that the substitute can be monitored by the reactor operator.

Bases 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. -

3.7. 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 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 $1.00. This includes the potential reactivity which might result from experimental malfunction, experiment flooding or i voiding, or the removal or insertion of l experiments.

I 3. Experiments having reactivity worths of greater than 80.75 shall be securely located or fastened to prevent inadvertent movement during reactor l 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.

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5. Materials which could react in a way which could damage the components of the reactor (such as gunpowder, dynamite, TNT, nitroglycerin,orPETN) shall not be irradiated in quantities greater than 25 milligrams 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 reaction. Such materials in quantities less ,

than 25 milligrams may be irradiated provided that the pressure produced in the experiment container 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 authorised local, state,

- or federal officers.

6. Experiment materials, except fuel materials, which could off-gas, sublime, volatilise or produce aerosols under (a) normal operating conditions of the experiment or the reactor, (b) credible L ' accident conditions in the reactor or (c) possible l 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 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 experimental facility exhausts through a holdup tank which closes L 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 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.

. 7. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is nc 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 one level of review of any proposed operation of the reactor in order to minimise the possibility of i 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 characterised 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. '

l This specification is intended to limit the power 3.

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.

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. 4. This specification is intended to 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 airborne radioactive materials to the reactor room or the atmosphere.

7. This specification is intended to reduce the severity of any possible release of these fission products which pose the greatest hasard to workers L 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:

l l biennially - not to exceed 30 months

! annually - not to exceed 15 months

! semi-annually - not to exceed seven and one-half months monthly - not to exceed six weeks c weekly - not to exceed 10 days l 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 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 l

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  !

following shutdown after the extended period 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.

4.1. Reactor Core Parameters Applicability These specifications apply to surveillance requirements for reactor core parameters.

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

Specification The reactivity worth of each control rod, the reacior

. core excess, and the reactor shutdown margin shall be measured st least annually and after each time the core fuel is moved.

Basis Movement of the core fuel could change the reactivity of the core and thus affect both the core excess

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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 l, changes of these parameters over an extended period of time and operation of the reactor have been shown to be j very small so that an annual measurement is sufficient j to ensure compliance with the specifications of section-l 3.1.

4.2. Reactor Control and Safety Systems

Applicability i

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

Specifications

1. Control rod drive withdrawal speeds and control rod 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.

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. Measurement of the control rod withdrawal speed ensures that the maximum reactivity addition rate specification will not be exceeded.

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 calibration of the wide-range linear channel provides assurance that long-tern drift of all three neutron detectors would be detected and that the reactor will be operated within the authorised 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 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 over the past 19 years indicates that the surveillance specification is adequate to assure proper operation of the control 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, pH, and the radioactivity of the pool water shall be measured at luast 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 i

water in a timely f ashion. Evaluation of the radioactivity in the pool water allows the 3

detection of fission product releases from damaged i fuel elements or damaged experiments.

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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 coolant temperature approaches the specified limit. )

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 at least annually.

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

Bases These specifications ensure that the named equipment can perform the req: Ired 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.

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4.5. 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.

The reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal.

A TRIGA fuel element shall be removed from the core if:

a) The transverse bend exceeds 0.125 inch over the length of the cladding, b) The length exceeds the original length by 0.125 inch.

c) A clad defect exists as indicated by release of fission products.

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.

4-4.6. ALARA Applicability 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.

Objective 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.

Specification The review of all operations shall include consideration of alternate operational modes which might reduce exposures to ionising 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 1

be a part of overall reactor operation and detailed i experiment planning.

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'5. DESIGN FEATURES 5.1. Reactor Site and Building Applicability These specifications shall apply to the Dow TRIGA Research Reactor.

Objectives The objectives of these specifications are to define the exclusion area and characteristics cf the confinement.

Specifications The minimum distance from the center of the reactor pool to the boundary of the exclusion area shall be 75

, feet.

The reactor shall be housed in a room designed to restrict leakage.

All air or other gas exhausted from the reactor room 2

and from associated experimental facilities during reactor operation shall be released to the environment

, at a minimum of 8 feet above ground level.

Bases The minimum distance from the pool to the boundary provides for dilution of effluents and for control of access to the reactor area.

Restriction of leakage, 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 I

the possibility of exposure of personnel to such gases.

5.2. Reactor Coolant System Applicability This specification applies to the Dow TRIGA Research Reactor.

Objective i

The objective of this specification is to define the characteristics of the cooling system of this reactor.

Specification The reactor core shall be cooled by natural convective water flow.

Basis Experience has shown that TRIGA reactors operating at power levels up to 1000 kilowatts can be cooled by natural convective water flow without damage of the fuel elements.

5.3. Reactor Core and Fuel Applicability These specifications shall be applicable to the Dow TRIGA Research Reactor.

Objective The 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.

Specification The fuel will be standard NRC-approved TRIGA fuel.

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.

The reflector (excluding experiments and experimental facilities) shall be a combination of graphite and water.

Bases The entire design and accident analysis is based upon the characteristics of TRIGA fuel. Any other material would invalidate the findings of these analyses.

The control elements perform their function through the absorbtion of neutrons, thus affecting the reactivity of the system. Boron has been found to be a stable and effective material for this control.

The reflector serves to conserve neutrons and to reduce the amount of fuel that must be in the core to maintain the chain reaction.

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1 5.4. Fuel Storage Applicability This specification applies to the Dow TRIGA Research Reactor fuel storage facilities.

Objective The objective of this specification is the safe storage of fuel.

Specification All fuel and fueled devices not in the core of the reactor shall be stored in such a way that keff 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.

Basis A value of keft of Isas than 0.8 precludes any possibility of inadvertent establishment of a self-sustaining nuclear chain reaction. Cooling which maintains temperatures lower than the Safety Limit prevents possible damage to the devices with subsequent release of radioactive materials.

l 6. ADMINISTRATIVE CONTROLS 6.1. Organisation The Dow TRICA 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 at 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 responsibilities include 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 i 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 organisation cuts across the lines of management of The Dow Chemical Company.

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 may appoint one or two equally-qualified individuals, upon notification of the l 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 organisation to the Reactor Operations Committee.

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Figure 6.1. Administration President Dow Chemical USA 9 y Director Michigan Division of Research General Manager

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. Corporate Michigan Division Laboratories Director of Research o a y Radiation Radiation Analytical Safety Officer - - -

Safety Committee Laboratories Chairman: Level 1 1

I g u g Reactor Operations Committee Chairman: Level 2  :

Facility Director o

Reactor Supervisor TRIGA Reactor Facility i

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• 6.1.3. Staffing 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

b. a second person present at the facility able to carry out prescribed written instructions, and

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:

a. manipulations of fuel in the core;

b. manual removal of control rods;

c. maintenance perforr.ed on the core or the control rods;

d. recovery from unexplained scrams, and

e. movement of any in-core experiment having an estimated reactivity value greater than

-80.75.

A list of reactor facility personnel by name and telephone number shall be readily available in the control roomfor use by the operator, including management, radiation safety, and other operations personnel.

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s 6.1.4. Selection and Training of Personnel

The Reactor Supervisor.is responsible for the-training and requalification of the facility Reactor Operators and Senior Reactor Operators.

The selection, training, and requalification of operations personnel shall be consistent with all current regulations.

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

6.2.1. Charter and Rules s

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, radiation science, or reactor / radiation l

engineering.

b. A quorum shall consist of a majority of the members of the R00. 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. 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 50.59, and would review and approve the j required safety analyses.

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

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,2.,c. t 6.2.2. Review Functions 1

c 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

,y core or'its irradiation facilities, or a new piece

, ;0 g 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 er operation which is of a type not previously approved by the Committee;

d. proposed changes in operating procedures, technical specifications, license, or charter;

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

f. operating abnormalities having safety significance;

g. reportable occurrences;

h. proposed changes in equipment, systems, tests, or experiments with respect to unreviewed safety questions; and

h. audit reports.

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

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6.2.3. Audit Function

a. The ROC shall direct an annual audit of the facility operations for conformance to the technical specifications, license, and operating procedures, 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.

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 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 plan, and the reactor opern+.or requalification program. This audit may consist of the annual review of these plans for the requalification program.

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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;

e. checkout, calibration and determination of operability of reactor operating instrumentation and controls, control rod drives 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. Experiment Review and Approval

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 l Assistant Reactor Supervisor. The sritten approval l

shall include documentation that the hasards have been

, considered by the reviewer and been found appropriate for this form of experiment.

d. Special Experiments, those experiments that are reither 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., 6.2.2.b., or 6.2.2.c. 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 immediately reported to the Facility Director or to a higher level;

c. The Safety Limit violation shall be reported to the US NRC in addordance with section 6.2.2.; 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 a Reportable Occurrence of the type identified in section 1.28:

a. reactor conditions shall be returned to normal or the reactor 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 6.2.2.; and

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

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L 6.6. Reports 6.6.1. Operating Reports

-A report shall be submitted annually, within 90 days of the anniversary of the license, 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, including a summary of the analyses leading to the conclusions that no unreviewed safety questions were involved and that 10 CFR 50.59 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 of individual 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 l g) a summary of the radiation exposures received by facility personnel and visitors, in the form indicated in 10 CFR 20.407(b), where such exposures are greater than 25 % of those allowed or recommended in 10 CFR 20.

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6.6.2. Special Reports

a. There shall be a report to US NRC Region III not later than the following working day by telephone-and confirmed in writing by telegraph or similar conveyance to the Director of Nuclear Reactor Regulation, US NRC, with copy to the Regional Administrator, Region III, US NRC to be followed by a written report that describes the event within 14 days of:

a violation of the Safety Limit; or a reportable occurance (section 1.28) .

b. There 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 involving level 1 or level 2 personnel; 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 III, US NRC, within 60 days after criticality of the reactor under conditions of a new facility license authorising an increase in reactor power level, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions.

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;

h. minutes of ROC meetings;

i. records of audits;

j. facility radiation and contamination surveys; and

k. surveillance activities as required by the Technical Specifications.

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:

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.

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ADDITIONAL INFORMATION DOW TRIGA RESEARCH REACTOR LICENSE R-108 DOCKET 50-264 MAY 1987

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1. Schematic plan view of'the reactor core sh~owing the position of

. the aluminum-clad and stainless-steel-clad fuel elements,.the control rods, the. source, the neutron detectors, and the experimental facilities.

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2. Is there an alarm installed to warn of loss of coolant condition in the reactor tank? Is there a credible method by which primary coolant can be introduced into the city water supply? If not, what design features prevent this event from occurring?

An alarm, equipped with lights and audible alarms, is installed in the reactor tank, to warn of loss of coolant.

There is no credible method by which primary coolant can be introduced into the city water supply.

Design features which prevent primary coolant from being introduced into the city water supply include:

Makeup water is supplied from deionized steam condensate, not from city water; and A source of water which can be temporarily connected to the primary side for the purpose of adding larger amounts of water is separated from the city water system by a) an antiflowback device in the line and b) the building repressurizing system, which provides an air break between the city water system and the building water supply.

2. Please analyze the failure of primary system components or piping in the basement area of 1602 Building. What is the maximum loss of primary coolant that could occur? In your analysis, please calculate the radiological consequences of coolant loss.

Failure of primary system components or piping in the basement area of 1602 Building would lead to the loss of at most 250 gallons of coolant, since anti-syphon holes one foot below the normal level'of the coolant prevent any further loss.

Loss of one foot of water from the pool would still leave the reactor core covered by at least 15 feet of water, which would provide cooling and shielding. There would be no radiological consequences in the reactor room of such a loss.

In the basement area, dams have been constructed which would prevent the loss of any of the water through the building drainage and sewer system. The dams are of such a height that up to 900 gallons of water could be so stored.

The radiological consequences of a. release of 250 gallons of primary coolant to the basement area would be negligible since the radioactivity of the pool water is low.

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4. What are the pressures on the primary and secondary sides of the heat exchanger? Is there an alarm to warn personnel if the pressure on the primary side of the heat exchanger exceeds the pressure on the secondary side? Is there a means to check the secondary discharge from the heat exchanger for radioactivity before it enters the sewer? Discuss the implications of a water leak between the primary and secondary sides of the heat exchanger.

The pressure en the primary side of the heat exchanger is a maximum of 11 psig. The pressure on the secondary side of the heat exchanger ranges from a minimum of 15 psig at full flow to a maximum of about 55 psig, the nominal supply pressure.

There is no alarm to warn personnel if the pressure on the primary side of the heat exchanger exceeds the pressure on the secondary side.

There is no means to check the secondary discharge from the heat exchanger for radioactivity before it enters the sewer.

A water leak between the primary and secondary sides of the heat exchanger would be from the secondary side to the primary side and would be detectable by an increase of conductivity and by the water level monitor, which has a high-level alarm as well as a low-level alarm.

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5. Schematic of the ventilation system for the reactor room and the 1602 Building I

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_130 CFM MANUAL AUTOMATIC LOUVERS AIR-OPERATED OUTSIDE LOUVERS BUILDING FILTERS n/ HEATING l

FAN INTAKE COIL LEAKAGE FROM 1602 BUILDING AIR SYSTEM

g s 6. What provisions are there for fire protection at the reactor facility?

Fire protection at the reactor facility is designed to provide protection against the types of fire hazards associated with the operation of a research laboratory:

fire extinguishers are provided at strategic locations; s smoke detector in the reactor console room is interfaced with the

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Dow Security Department console, manned at all times; the Dow Fire Department, a part of the Security Department, is staffed

- at all times with professional firefighters, who, during test exercises,-respond to the site with within a few minutes of an alarm, and who annually receive radiation training; and the Midland City Fire Department provides backup assistance.

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7. What provisions are there for emergency power at the reactor facility?

The only provision for emergency power at the facility is related to the security system, the smoke alarm, and the emergency lights.

Loss of electrical power during operation of the reactor would lead to automatic shutdown of the reactor. The low residual heat associated with this core, even at 300 kilowatts operation, would not lead to any damage of the core nor to any significant rise in the temperature of the pool water. The pool water normally circulates through the mechanism of natural convection, so there is no necessity to keep any pumps operating.

8. Describe your ALARA program and how it is implemented.

The Dow radiation protection program is designed to be consistent with the concept of ALARA - As Low As Reasonably Achievable.

The radiation protection program originates with the Radiation Safety Committee (RSC), a standing committee of Dow Chemical USA, which oversees the total radiation and radioactive material usage at the Midland location. The corporate Industrial Hygiene department is staffed by professional Health Physics and Industrial Hygiene specialists who monitor and oversee the program and serve as resource persons for all the users of radiation and radioactive materials at the Midland location.

Training, pre-startup reviews and evaluations of proposed programs, personnel dosimetry, administrative procedures, and regularly scheduled surveys of radiation and radioactive materials are used and documented as part of the radiation protection program.

Reports of exposures from the personnel monitoring program are recorded, sent to the persons involved, and sent to the managers of the persons involved. Management review of these reports, along with other aspects of the Dow Safety program, become part of the employee's annual Job Performance Review. A consistent series of reported exposures can lead to re-evaluation and re-design of the project.

The results of the radiation protection program are evident in the personnel dosimetry records - question 10.

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9. Please describe the administrative organization of your radiation protection program. Include details on the health physics training provided for reactor operations personnel and reactor users.

The health physics training for reactor operations personnel is provided by certified Health Physics personnel of the Dow Industrial Hygiene department during an initial three-hour training session; by a lengthy classroom and hands-on program from the Training Coordinator and other persons experienced in the handling of radioactive materials during reactor license training; and by an annual requalification program from both Health Physics persons and the Training Coordinator.

The health physics training for reactor users involves both the initial Health Physics / Industrial Hygiene training and hands-on training by experienced persons on the reactor staff, with annual retraining / review provided by the Health Physics / Industrial Hygiene group.

10. Please provide a summary of the annual personnel exposures at the facililty for the last five years. In light of your request to increase reactor power to 300 kW, do you anticipate any changes in personnel exposure?

TABLE 10.1. FIVE-YEAR

SUMMARY

OF ANNUAL PERSONNEL EXPOSURES YEAR NO MEASUR- 0.01R 0.02R 0.03R NUMBER OF ABLE AM'T. PERSONS 1982 8 - -

1 9 1983 7 1 1 -

9 1984 13 - - -

13 1985 9 2 1 -

12 1986 8 -

1 -

6 We do not expect to see any increase in personnel exposures due to 300 kW operation. Operation at 300 kW is expected to occur with about the same frequency and for about the same length of time as is the case for 100 kW operation now - a rather small part of the total reactor program. Exposure to the reactor radiation field occurs only when a person is standing over the pool during operation of the reactor - an uncommon occurrence. The radioactivity of sampics removed from the reactor after activation at up to 300 kW will not be greatly increased over that found after a longer operation at 100 kW.

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11. Please provide figures for the annual release of Ar-41 from your facility for the past five years. Please provide the assumptions made and calculations, or the specific reference, for the estimated release rate at 300 kW. What potential maximum exposure rate does this represent to people in unrestricted areas?

The reactor has been operated for about 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> equivalent at 95 kW per year for a number of years. The release of 41-Ar during such operation is dependent on the use of the pneumatic transfer system, which is estimated to occur during about 10 percent of the operation.

The annual release of 41-Ar from the pool (produced by activation of argon dissolved in the water) and the pneumatic transfer system is calculated to be about 0.1 Ci.

If the reactor is operated for the same equivalent time at 300 kilowatts the amount of 41-Ar released annually is expected to be about 0.2 01, calculated as follows:

1. The pneumatic transfer system contains about 5 mci of 41-Ar when the blower is not operating and the reactor has been operated at 100 kilowatts for more than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (static saturation) (Dow TRIGA Research Reactor Safety Analysis Report 1966). Thus the production rate of 41-Ar in the system will be 0.535 pCi per second at 100 kilowatts or 1.60 pCi/see at 300 kilowatts, in an air exhaust of 1000 cfm, giving a total of 0.07 Ci of 41-Ar released annually, assuming that the reactor is operated as above and that the pneumatic transfer system is operated ten percent of the time. This material would be released from a fume hood stack nine feet above the roof of the building.

2. The release of 41-Ar from the pool water (produced by the neutron activation of argon dissolved in the coolant water) is calculated to be at a rate of 7.37 x 107 nuclei per second when operating at 300 kilowatts (GA Report E-117-478, Safety Analysis Report for the Turkey 250 kilowatt TRIGA Reactor, 1975, modified for 300 kilowatts). This would amount to a release of 0.21 pCi/sce in an air exhaust of 1700 cfm or an annual release of .09 Ci when the reactor is operated as above. This material would be released through the reactor room ventilation system at a point eight feet above the ground on the east side of the building.

The maximum potential exposure rate to persons in the nearest unrestricted area (40 meters west of the fume hood exaust, 50 meters west of the reactor room exhaust) will be about 0.01 mR/hr with the reactor operating at 300 kilowatts and the pneumatic transfer system operating, using a dose factor for a semi-infinite cloud of 41-Ar given in table B-1, NRC REG CUIDE 1.109 (1977). The maximum annual potential exposure to persons in the unrestricted area would be less than 0.5 mR for the operating regime of the reactor given above.

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12. Please discuss the effect that increasing reactor power will have on the facility instrumentation.

The present instrumentation will be used at the 300 kilowatt power level. The three power-level channels will be re-calibrated by moving the neutron detectors to positions which will give full-scale response at the desired maximum power level. The ultimate calibration will be through a thermal calibration procedure identical to the current practice. All other instrumentation will remain the same as it is now.

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13. Drawing showing the relationship between the reactor room and areas of unrestricted access.

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DOW SECURITY FENCELINE -

6 REACTOR 1602 BUILDING 1 I I

14. Please describe the methods for and restrictions to irradiating experiments in the water volume near the core. Discuss the possibility of disturbing the natural convective cooling circulation of the core by the placement of experiments above the Core.

During the one set of experiments performed during the past six years involving irradiation in the water volume near the core, the experiments were placed in closed watertight plastic containers weighted with lead bricks, and lowered to the surface of the core using redundant retrieval equipment.

Experiments involving such irradiations are classified as special experiments and must have the evaluation and approval of the reactor supervisor and the Reactor Operation Committee. Restrictions generally include requirements that the experiment remain sealed, that no materials be released to the pool if the experiment should be breached, and that the experiment and its components be monitored for the presence of radiation and of loose radioactive material as it is removed from the pool. Specific restrictions may also apply, depending on the nature of the experiment.

The review process also considers the effect of the experiment on the convective flow of water through the core of the reactor. Convective cooling flow at any power level must not be disturbed in a way which would lead to overheating of part of the core. The size and the position of the proposed experiments will determine the effect on the flow of water.

15. What procedures are in place to assure that experiments do not stay in the reactor for periods greater than those authorized?

What assurances are there that the radioactivity released in a misplaced experiment accident will be less than the maximum hypothetical accident (MHA)?

Administrative procedures govern the documentation of the presence of samples in the experimental areas. Samples left in the core overnight are logged in the appropriate section of the shut-down checkout and the startup checkout the following day. Samples which are inserted in the core for gamma-ray irradiation, with the reactor otherwise secured, are also logged.

Review of all experiments assures that failure will not lead to the release of radioactivity approaching the maximum hypothetical accident (the failure of a single fuel rod after extended operation at full power). The review procedures are detailed in the technical specifications.

16. The SAR quotes a fission product activity of 4200 Curies for the fuel element considered in the MHA. If this number does not represent the fuel element with the maximum fission procuct inventory, please estimate the maximum fission product inventory that will occur after maximum expected operation.

The gaseous fission product activity, estimated above at 4200 curies, is calculated more exactly by S. C. Hawley and R. L. Kathren, NUREG/CR-2387, PNL-4028, CREDIBLE ACCIDENT ANALYSES FOR TRIGA Ah3 TRIGA-FUELED REACTORS, and is given in their table 4. These numbers b' were calculated for a core containing 50 elements operated at 1 megawatt for 365 days and are reported for the element with twice the average power density. Considering the Dow TRIGA Research Reactor with 78 fuel elements and a long-term operation at 300 kilowatts, the s fuel element with the maximum power density (twice the average) is expected to contain about 3300 Curies of krypton, xenon, and iodine .

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17. What are the maximum and average fuel temperatures at 300 kilowatts?.-

The calculated maximum and average fuel temperatures at 300 kilowatts are 260 C and 140 C, respectively, assuming that the maximum power density is twice the average power density.

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18. Please provide numbers for the table below considering the MHA.

Please provide the assumptions made and calculations, or the specific reference, that were used to arrive at your results.

Exposure Whole-body Thyroid and Location Immersion Dose Committed Dose 1-minute (occupational) 11 arem 5 rem exposure in the reactor room 1-hour (public) 0.4 mrem 153 mrem

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CALCULATION OF RELEASES OF RADI0 ACTIVE MATERIALS FOLLOWING A MAXIMUM HYPOTHETICAL ACCIDENT (MHA) INYOLVING THE DOW TRIGA RESEARCH REACTOR

1. Assumptions a) The reactor has been operating at 300 kilowatts for 365 consecutive 24-hour days.

b) The reactor contains 78 standard TRIGA fuel elements.

c) The wind is from the e3st at 7 miles per hour with moderately stable wind conditions.

d) The reactor room ventilation equipment is operating and remains so during the release.

i e) A catastrophic failure of the cladding of one of the B-ring elements (which has experienced maximum burn-up and thus contains the maximum inventory of fission products) leads to the release of the entire release fraction of fission products.

f) Only noble gas and iodine fission products are released in significant amounts.

g) The Continuous Air Monitor alarms within seconds of the release and the reactor is shutdown i==ediately, h) The entire release is extended over a period of ten minutes due to the turnover rate of air in the reactor room (about three air changes).

i) No credit is taken for retention in the pool water, plateout, or decay.

j) Gaseous materials are released eight feet from the ground 160 feet from the exclusion boundary.

k) The release fraction, following the practice of reference 1, is taken as 10-4

m .

2. References

1) Hawley, S. C., and R. L. Kathren, NUREG/CR-2387, PNL-4028, CREDIELE ACCIDENT ANALYSES FOR TRIGA AND TRICA-FUELED REACTORS (1982) 2)' NRC REG GUIDE 1.109, Table B-1: DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES (1977)

3) Holden, B. S., 1602 BUILDING NUCLEAR REACTOR DISPERSION MODELLING, Dow Chemical USA, Michigan Division (1987)

3. Noble gases Table 1 shows the results of the calculations of the dose at the fence-line using the assumptions given above. The' inventory of the noble gases, in curies, was taken from Table 4 of reference 1, modified by factors of 0.3 and 50/78 for the power level and number of fuel elements, respectively, for the Dow TRIGA Research Reactor versus the reference reactor. The concentrations at the fence-line, given in curies per cubic meter, are calculated from the results of Holden, reference 3, using program MIST with continuous emission from a building source. The dose factors, in arem/(Curie / cubic meter), are taken from table B-1, reference 2, modified for a ten-minute exposure.

Table 1.. Total Whole-body Doses at the Fence-line Following the MHA t

ISOTOPE RELEASES 8 ISOTCPE 1 t1/2 2 activity 3 C1/m3 4 dose 5 dose NAM ~ hrs curies factors mrom 1 83.. Kr 1.90000 23.8 4.118511e-88 1 8.000 2 85m Kr 4.40006 55.1 9.532359e-08 22245 8.e02 3 85 Kr 946e8.00000 8.9 1.594262e-09 3e6 e.000 4 87 Kr 1.30000 106.0 1.833481e-87 112557 8.021 5 88 Kr 2.80000 151.5 2.620568e-87 279491 8.073 6 89 Kr 8.e5000 186.2 3.221738e 315616 e.le2 7 90 Kr 8.90089 211.6 3.86016e-87 296663 0.109 .

8 133m Xe 55.20006 7.5 1.295338e-08 4772 8.000 9 133 Xe 127.20006 436.8 7.542853e-67 5596 8.004 18.135m Xe 8.30000 114.8 1.986185e-07 59321 8.912 11 135 Xe 9.10000 196.8 3.404414e-67 34414 0.012 The total dose of <0.4 mR at the fence-line is extremely conservative, given that the actual operating practice of the Dow TRIGA Research Reactor does not approach the assumed operation to maximum inventory of fission products, that the release fraction is conservative by a factor of about 6 (reference 1), and that no credit is taken for retention of materials in the pool water, for plate-out, or for decay.

'4. Iodines The total lifetime dose equivalent commitment to the thyroid following the MHA for the Dow TRIGA Research Reactor is calculated from the concentrations of the iodine isotopes 131 - 135 as calculated by Holden (reference 3) from the-initial inventories given in reference 1 as modified by factors of 0.3 and 50/78 for the power level and number of fuel elecents, respectively, for the Dow TRIGA Research Reactor versus the reference' reactor. The inhalation rate is taken as 1.2m3/ hour (reference 1) and the exposure is-taken for the entire release period, ten minutes. The dose factors are taken from Table E-7, Inhalation Dose Factors for Adults (mrem per pCi inhaled), NRC Reg Guide 1.109.-

Table 2. Total Committed Lifetime Doses of Iodines at the Fence-line Following the MHA Isotope Half- Inventory Concentration Dose Committed Name life, at Shutdown, at Fenceline, Factors,- Dose, hours Curles C1/ cubic meter mrem /Ci mrem k31I $94.4 207 3.587e-87 1.49ee-e3 107 132 I 2.3 319 5.520e-07 1.430e-05 2 133 I 28.3 371 6.417e-87 2.690e-04 35 134 I 8.9 488 8.450e-07 3.730 -06 1 135 I 6.7 425 7.350e-07 5.60ee-05 8 The total committed dose, 153 mrem, compares well with the estimate of 232 mrem (reference 1), as modified for the characteristics of the Dow TRIGA Research Reactor. Each of these calculations uses highly conservative estimates of the release fractions and of actual releases from the pool.

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19. Considering a core containing the aluminum-clad fuel element, please analyze and justify increasing your core K-excess to 3.008. I Also,-provide the assumptions cade and calculations, or the specific reference, for the reactor power transient analysis (SAR page 43), the instantaneous LOCA analysis (SAR page 46), and the loss of shielding analysis (SAR page 44).

The maximum core excess reactivity of $3.00 consists of the following contributions:

l 1. power'coeffittient - about $2.10 to achieve 300 kilowatts; l

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2. Xenon poisoning - not greater than 80.30; and

-3. control purposes - so the reactor can be operated at full licensed power, with Xenon, the regulating rod at its most sensitive position, about halfway withdrawn.

The aluminum-clad fuel element will be placed in an outer ring of the core where it will be subject to the minimum effect of an accidental transient.

TRIGA reactors containing all aluminum-clad fuel elements have been analyzed for pulsing at more than 82.00 but not at the 83.00 level.

The Dow TRIGA aluminum-clad fuel element, in an outer ring, would be subject to less stress and temperature rise following an accidental 83.00 pulse than are the B-ring aluminum-clad fuel elements subjected to S2.00 pulses.

o 23-References for reactor power transient analysis:

a) Dow TRIGA SAR 1967, p 47 b) S. C. Hawley and R. L. Kathren, Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors, NUREG/CR pp 15-21; table 3, page 19.

References for instantaneous LOCA analysis:

a) Ibid, pp 11-12 b) Dow TRIGA SAR 1967, p 46 Loss of shielding water analysis:

a) Ibid, p 54 The most recent data from General Atomics indicates that the table quoted (Dow TRIGA SAR 1967, p 54) and used as the basis for Table H.2.1, page 45, Dow TRIGA SAR 1986, should be significantly modified. This table as modified for the proposed 300 kilowatt operation of the Dow TRIGA Research Reactor is as follows:

Time After Direct Indirect Loss of Radiation, Radiation, shielding R/ hour R/ hour 10 seconds 3000 .78 1 day 360 .090 1 week 130 .042-1 month 35 .012 This table refers to exposures expected at the indicated times following instantaneous loss of all shielding water after operation of the reactor at 300 kilowatts for a time sufficient to produce the maximum inventory of radioactive fission products, of the order of one month.

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