ML20154C276

Mods Supplementing Tech Specs for Renewal of License R-108

ML20154C276
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Site:	Dow Chemical Company
Issue date:	04/30/1988
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Download: ML20154C276 (44)

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I TECHNICAL SPECIFICATIONS FOR THE i DOW TRIGA RESEARCH REACTOR i

FACILITY LICENSE R 108 APRIL 1988 i

This document includes the Technical Specifications and the bases for i the Technical Specifications. The bases provide the technical support l for the individual Technical Specifications and are included for i 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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Amendment No. 5 8905180104 880429 PDR ADOCK 05000264 P DCD

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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 minimize occupational exposures to ionizing radiation and releases of radioactive materials to the environment.

l 1.2. Channel A channel is a combination of sensors, electronic circuits, and output devices connected by the appropriate communications network in order to measure and display the value of a parameter, 1.3. Channel Calibration - A channel calibration is an adjustment of a channel such that its output corresponds with acceptable accuracy to known values of the parameter i

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 channals 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 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.

1.7. Excess Reactivity . Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive position from the condition where the reactor is exactly critical.

I 1.8. Exceriment 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 reactor characteristics.

1.9. Exrerimental Facilities include the rotary specimen rack, vertical tubes, pneumatic transfer systems, the central l

thimble, and the area surrounding the core.

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1 Amendment No. $ l L

2 1.10. Limitine conditions for Oeeration . Limiting conditions for Operation (LCO);are administrative 1y established constraints on equipment and operational characteristics which shall be adhered to during operation of the reactor. '

1.11. Lig* tine safety system settine (Lsss) . An LSSS is the ac.aating 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 Exceriments . 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 hazards are neither significantly different from nor greater than the ha:arda of the corresponding routine experiment.

1.14. Movable Excerimeng . 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. 20erable . A component or system is operable if it is capable of performing its intended function.

1.16. Overstine . A component or system is operating if it is performing its intended function.

1.17. Radiation safety Committee (Rsci . The RSC is chartered by The Dow Chemical Company to be responsible for the license for the Dow TRICA Research Reactor facility.

1.18. Reactivity Limits . The reactivity limits are those limite imposed on reactor core excess reactivity. Quantitien are referenced to a Reference Core Condition.

1.19. Reactivity Vorth of an Exceriment .

The reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.

1.20. Resetor overatine . 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 scram.

Amendment No. 5

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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 1 attain criticality under optimum available conditions j of moderation and reflection, or

b) the consolo switch is in the "off" position, the key is l removed from the switch, and the key is it, the control 1

of a licensed reactor operator or stored in a locked storage area; and sufficient control rods are inserted to assure that the i i

reactor is suberitical by a margin greater than $1.00 t

cold, without xenon; and i

I 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 $0.75.

1.23. Reactor Shutdown . The reactor is shutdown if it is 4

suberitical by at least one dollar and the reactivity worth ,

of all experiments is accounted for.  ;

1.24. Resetor Operations Committee (ROC) . The ROC is charged with direct oversight of the reactor operations, including both review and audit functions.

", 1.25. Peactor Safeev systems . Reactor Safety Systems are those i

systems, including associated input channels, which are  !

j designed to initiate automatic reactor protection or to l provide information for initiation of manual protective i action. .

i 1.26. Reference Core condition . The Reference Core Condition is i

that condition when the core is at ambient temperature i (cold) and the reactivity worth of xenon in the fuel is

, negligible (less than $.30).

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Amendment No. 5

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1.27. Research Reactor . A Research Reactor is a device designed to support a self sustaining nuclear chain reaction for l research, development, education, training, or experimental i

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purposes, and which may have provisions for the production of radioisotopes, 1.28. Reportable Occurence . A Reportable Occurence is any of the i following which occurs during reactor operation:

i' a) Operation with actual safety. system settings for e required systems less conservative than the limiting i safety. system settings specified in Technical Specification 2.2. ,

b) Operation in violation of limiting conditions for j

2 operation established in the Technical Specifications. l 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 t greater than one dollar. Reactor trips resulting from a known cause are excluded, a  ;

, e) Abnormal and significant degradation in reactor fuel, j cladding, or coolant boundary which could result in .

! exceeding prescribed radiation exposure or release limits, j i

f) #a observed inadequacy in the inplementation of either administrative or procedural controls which could l result in operation of the reactor outside the limiting  ;

} conditions for operation.

e g) Release of radioactivity from the site above limits i j specified in 10CFR20.

I 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 reds are coupled to the control rod drive systems in a way that allows the j control rods to perform a safety function.

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Amendment No. 5 J

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1.30. Routine Experiment . A routine experiment is an experiment l which involves operations under conditions which have been  !

, extensively examined in the course of the reactor test i programs and which is not defined as any other kind of l experiment. Experiments and classes of experiments which j are to be considered as routine experiments must be so defined by the Reactor Operations Coraittee.

i 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 ,

4 radiaoctivity. The principal physical barrier is the fuel  !

element cladding.

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

1.33. Secured Exreriment . 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 i experiment might be subjected by hydraulic pneumatie, ,

buoyant, or other forces vM :h are normal to the operating  ;

snvironment of the experiv.n t, or by forces which can arise i as the result of credible a41functicns.

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1.34. Shall. Should. and lisy . The 'ard "shall" is used to denote a requirement, the word "should" denotes a tecommendation, i and the word "may" denotes permission, neither a  !

requirement nor a recommendation. 1 l  !

4 1.35. Shutdown Margin . Shutdown Margin is the reactivity existing when the most reactive control rod is fully  !

withdrawn from the core and the other control rods are  !

I fully inserted into the core. . I 1.36. grigial Experiments . Special experiments are experiments which are neither routine experiments nor modified routine experiments.

1.37. TRICA Fuel Element . A TRIGA fuel element is a sealed unit containing (U,Zr)Hx fuel for the reactor. The uranium is enriched to less than 20% in 235.U and the fraction of hydrogen is in the range of 1.01.1 for aluminwn. clad TRICA elements and in the range of 1.6 1.7 for stainless steel. I clad TRICA elemerts. i I '

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Amendment No. 5 l l

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2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1. Safety Limit (SL)

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

Obiective 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.

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

Basis A lcss 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 cir, 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 1 by the hydrogen content. Data indicate that the stress in the cladding due to hydrogen pressure from the dissociaticn of ZrHl .6 will remain below the ultimate stress provided that the fuel temperature does not I exceed luSO C and the fuel cladding temperature does not exceed 500 C. When the cladding-temperature can j equal the fuel temperature the fuel temperature design limit is 950 C (M. T. Simnad, G. A. Proj ect No. 4314, Report e-117-833, 1980).

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.

Amendment No. 5 1

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The thermal characteristics of aluminum clad TRIGA fuel i elements using ZrHi ,1 moderator have been analyzed )

($. C. Hawley and R. L. Kathren, NUREG/CR-2387, PNL-4028, Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors, 1982) A loss-of-coolant analysis j showed that in a typi i graphite-reflected Hark-I TRIGA reactor fueled with 60 aluminum-clad fuel l elements (Reed College) the maxii. fuel temperature l would be less than 150 C followine infinite operation  !

at 250 kilowatts terminated by the instantaneous loss of water. These temperatures are well below the region 1 where the a + 6 + y' to a + 6 phase change occurs in  !

ZrH1 ,1 (560 C). i l

2.2. Limitinz Safety System Settines (LSSS) l l

Applicability 2 This specification applies to the reactor scram setting which prevents the reactor fuel temperature from reaching the safety limit.

Obiective i

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

Specification l

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 l considerable safety margin. One TRICA reactor (General i Atomics, Torrey Pines) showed a maximum fuel temperature of 350 C during operation at 1500 kilowatts, while a 250-kilowatt TRIGA reactor (Reed College) showed a maximum' fuel temperature of less than 150 C (reported by S. C. Hawley, R. L. Kathren, '

NUREG/CR-2387, PNL-4028 (1982), Credible Accident i Analyses for TRIGA and TRIGA-Fueled Reactors). A '

portion of the safety margin could be used to account Amendment No. 5

_ _ _ - _ . - . -. -,-..,._m. _ . _ . _ _ _ _ , _ . - . _ _ _

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 1 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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Amendment No. 5 1

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3. LIMITING CONDITIONS FOR OPERATION (LCO) l 3.1. Reactivity Limits Aeolicability These specifications shall apply'to the reactor at all times that it is in operation.

Obiective 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 $.50 with the most reactive control rod fully withdrawn, the other two control rods fully inserted, cold, no xenon, including the reactivity worth of all experiments.

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 marsi n 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 i shutdown margin and limits the maximum power excursion 1 that could take place in the event of failure of all of i the power level safety circuits and administrative l

controls. .

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1 Amendment No. 5 1

3.2. Core Confieuration Apolicability This specification applies to the core configuration.

Obiective The objective of this specification _is to' assure that the safety limit will not be exceeded duq _to power peaking effects.

Soecifications The critical core shall be an assembly of standard NRC-approved stainless steel-clad or aluminum-clad TP. IGA 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 Operation with standard NRC approved TRIGA fuel in the standard configuration ensures a conservative i limitation with respect to the Safety Limit.

Placement of the aluminum clad fuel element in the outer rings of the reactor core will help ensure that J

this element is not exposed to higher than average j power levels, thus providing a greater degree of i conservatism with respect to the Safety Limit for this )

one element, j l

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l Amendment No. 5 e

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3.3. Reactor Control and Safety Systems Apolicability These specificatic apply to the reactor control and safety systems and .afety-related instrumentation that must be operating when the reactor is in operation.

Obiective 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.

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

Each of the three control rods shall drop from the fully withdrawn position to the fully inserted position in a time not to exceed one second.

The reactor safety channels and the interlocks s' 21 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.

1 Positive reactivity insertion rate by control rod l

motion shall not e.<ceed $.20 per second. j 1

Eases l The requirement for three operable control rods ensures l that the reactor can meet the shutdown specifications. l 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 points, and interlocks given in table 3.3A assures l protection against operation of the reactor outside the '

safety limits.

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Amendment No. 5 1

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.

Amendment No. 5

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TABLE 3.3A.

MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS Scram Channels Scram Channel Minimum Goerable Scram'Setooint Reactor Power Level 2 Not to exceed maximum licensed power Reactor Feriod 1 Not less than 7 seconds Wide-Range Linear Channel 1 Failure of the detector Detector Power Supply high voltage power sup. ply Wide-Range Log Channel 1 Failure of the detector i high-voltage power supply e Manual Scram 1 Not applicable Interlocks l Interlock / Channel Function Startup Countrate Prevent control rod '

withdrawal when the l

neutron count rate is less

i than 2 cps Rod Drive Control Prevent simultaneous manual withdrawal of two control elements by the  !

control rod drive motors I

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Amendment No. 5 I

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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 safecy 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 1 input to the wide-range _ linear power l channel and the vide-range log power j channel

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l Manual Scram Allows the operator to shut the j reactor down at any indication of unsafe or abnormal conditions Interlocks Interlock / Channel Bases Startup Countrate Provides assurance that the signal in the log power channel is adequate-to allow reliable indication of the l

state of the neutron chain reaction Rod Drive Control Limits the maximum positive reactivity insertion rate i

Amendment No. 5

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

Power-Level Channel 1 (Percent Power)

Water Radioactivity 1 Monitor Vater Temperature 1 Monitor  ;

TABLE 3.3B BASES FOR MEASURING CHANNELS l i

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Measurine Channel Basis Wide-Range Log N Provides assurance that the and Period Channel reactor power level and period can be adequately monitored.

Power level Channel Provides assurance that the reactor 1 (Linear) power level can be adequately l monitored. .  !

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

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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.

Amendment No. 5 l

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3.4 Coolant System Aeolicability ,

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.

Obiective_s The objectives of this specification are:

to minimize corrosion of the cladding of the fuel elemt.nts 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.

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

1 The pool water pH shall ?. 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.

r Amendment No. 5

l Eases Increased levels of conductivity in aqueous systems I 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 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 l in accordance with the ALARA prograra.

Monitoring the pH of the pool water provides early detection of extreme values of pH which could enhance corrosion.

Monitoring the radioactivity in the pool water serves l 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 ac.cordance 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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Amendment No. 5 l

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

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

Eppeification 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.

Amendment No. 5

3.6. Radiation Monitorine Systems Aeolicability These specifications apply to the radiation monitoring information available to the reactor operator during operation of the reactor.

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

i Soecifications A Continuous Air Monitor (CAM) (with readout meter and audible alarm) in the reactor room must be operating  !

during operation of the reactor. I i

l The Area Monitor (AM) (with readout meter and audible 1 alarm) in the reactor room must be operating during ,

operation of the reactor or when work is being done on i or around the reactor core or experimental facilities.  ;

During short periods of repair to this monitor, not to

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exceed thirty days, reactor operations or work on or i around the core or experimental facilities may continue '

while a portable gamma-sensitive ion chamber is l utili=ed 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. .

1 Amendment No. 5

. l 3.7. Exoeriments Apolicability ,

I These specifications apply to experiments installed in i the reactor and its experimental' facilities.

Obiective 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.  !

1 Specifications i l

1. Operation of the. reactor for any purpose shall l 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 er 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 voiding, or the removal or insertion of experiments.

3. Experiments having reactivity worths of greater than $0.75 shall be securely located or fastened to prevent inadvertent movement during reactor 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.

Amendment No. 5

5. Materials which could react in a way which could damage the components of the reactor (such as-gunpowder, dynamite, TNT, nitroglycerin,'or PETN) 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 authorized local, state, or federal officers.

6. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize or produce aerosols under (a) normal operating conditions of the experiment or the reactor, (b) credible i 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 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 automatically on high radiation levels, the assumption shall be used that 10% of the gaseous activity or aerosols produced will escape,

b. If the affluent from an experimental facility exhausts through a filter installation i

designed for greater than 994 efficiency ~for 1 0.3 micron particles, the assumption shall be used that 10% of the aerosols produced escape.

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Amendment No. 5 i

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c. For materials'whose boiling point is above 55.

C and where vapors formed by boilin8 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 ne greater than 5 mil 11 curies.

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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 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 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.

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3. This specification is intended to limit the power excursioet'which might be' induced by the changes in reactivity due-to inadvertent motion of an unsecured experiment. Such excursions could lead to an inability to control the reactor within the limits imposed by the license.

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

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.

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7. This specification is intended to reduce the '

severity of any possible release of these fission '

products which pose the greatest hazard to workers i and the general public. '

8. This specification requires specific actions to )

determine the extent of damage following releases i of materials. No theoretical calculations or I evaluations are allowed. l 1

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 REOUIREMENTS Allowable surveillance intervals shall not exceed the following:

biennially - not to exceed 30 months annually - not to exceed 15 months semi annually - not to exceed seven and one-half months monthly - not to exceed six weeks I weekly - not to exceed 10 days daily - must be done before the commencement of operation  ;

each day of operation '

1 Amendment No. 5 i

Established frequencies shall be maintained over the long term, so, for example, any monthly surveillance shalt be performed at least 12 times during a calendar year of normal operation. If the reacter 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 of the reactor. Surveillance tasks scheduled daily or weekly which cannot be performed while the reacror 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 Aeolicability These specifications apply to surveillance requirements for reactor core parameters.

Obiective 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 reactor core excess, 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 of the core and thus affect both the core excess reactivity and the shutdown margin, as well as affecting the worth of the individual control rods.

Evaluation of these parameters is therefore required after any such movement. Without any such movement the changes of these parameters over an extended period of time and operation of the reactor have been shown to be very small so that an annual measurement is sufficient to ensure compliance with the specifications of section 3.1.

Amendment No. 5

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4.2. Reactor Control and Safety Systems 62211cability These specifications apply ~ to the surveillance requirements of the reactor safety systems.

Obiective The objective of these specifications-is to ensure the operability of the reactor safety. systems as described in section 3.3.

Specifications

1. Control rod drive withdrawal speeds and control-rod I drop times shall be measur3d at least annually and  !

whenever maintenance is performed or repairs are i made that could affect the rods or control rod i drives.  ;

2. A channel calibration shall be performed .for the wide-range linear power channel by thermal power calibration at least annually.

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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 3.3A, and the log power channel. '

l 4 The control rods shall be visually inspected at least biennially. l Bases t'

l. 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 t

reactivity addition rate specification will not be exceeded.

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4 Amendment No. 5 4

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- , , , . . . . - - - - - , - - - , ,v.-- - . , - ,. ,,-,-n , , - - .n,e-- ...-,-,,.r.

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. JThe specification for thermal calibration of the wide-range linear channel provides assurance that long-term drift of-all 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 opportunity co 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 th'e rod drop time measurements, i

1 4.3. Coolant System '

Anoticability 1

i i

These specifications shall apply to the surveillar.ce requirements for the reactor coolant system.

Obiective The objective of these specifications is to ensure that the specifications of section 3.4 are satisfied, j

j Soecifications .

1. The conductivity, pH, and the radioactivity of the pool water shall be measured at least monthly. I

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

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3. The temperature of the coolant shall be monitored '

during operation of the reactor.

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Amendment No. 5 ,

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-27 l j Eases

1. Experience at the Dow TRICA Research Reactor shows l 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 J radioactivity in the~ pool water allows the detection of fission product releases from damaged fuel elements or damaged experiments.

2. Experience indicates that this' specification is adequate to detect losses of pool water by evaporation.

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

q 4.4. Radiation Monitorinz Systems Applicability t

These specifications apply to the surveillance ,

requirements for the Continuous Air Monitor (CAM) and the Area Monitor (AM), both located in the reactor i

room.

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Obiective 1

j The objective of these specifications is to ensure the  !

1 quality of the data presented by these two instruments.

Soccifications

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

4

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

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

! Amendment No. 5 l

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• I 4.5. Facility Soecific Surveillance -

-Aeolicability This specification shall apply to the_ fuel elements of' the Dow TRICA Research Reactor.

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

Soecification Each fuel element shall be visually examined annually, i

The res ator shall not be operated with damaged fu .i j except to detect and identify damaged fuel for removal.

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

a) The transverse bend excaeds 0.125 inch over the length of the claddins-

]

b) The length exceeds the original length by 0.125 inch. )

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

l 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

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Amendment No. 5  ;

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I 4.6. ALARA Apolicability This specification applies to the surveillance of all reactor operations that could result in occupational exposures to ionizing radiation or the release of radioactive materials to the environment.

Obiective The objective of this specification is to provide surveillance of all operations that could lead to occupational exposures to ionizing radiation or the release of radioactive materials to the environs.

Soecification The review of all operations shall include consideration of reasonable alternate operational modes which might reduce exposures to ionizing radiation or releases of radioactive materials.

Basis Experience has shown that experiments and operational requirements, in many cases, may be satisfied with a variety of combinations of facility options, power icvels, time delays, and effluent or staff radiation exposures.

The ALARA (As Low As Reasonably Achievable) principle shall be a part of overall reactor operation and detailed experiment planning.

Amendment No. 5 l

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

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

Soecifications The minimum distance from the center of the reactor i pool to the boundary of the exclusion area shall be 75 l feet.

The reactor shall be housed in a room of about 6000 cubic feet volume designed to restrict ieakage. i All air or other gas exhausted from the reactor room and from associated experimental facilities during reactor operation shall be released to the environment at a minimum of 8 feet above ground level.

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 the possibility of exposure of personnel to such gases.

Amendment No. 5

31. -

5.2. Reactor Coolant System Aeolicability This specification applies to the Dow TRICA Research Reactor, i 4  !

Obiective The objective of this specification.is to define the characteristics of the cooling system of this reactor, Soecification The reactor core shall be cooled by natural convective I j water flow.

j j Basis Experience has shown that TP. IGA 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  ;

I Aeolicabil(tv '

5 a

l These specifications shall be applicable to the Dov 4

TRICA Research Reactor.

{ Obiective l 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, Soecification i 1

! The fuel will be standard NRC approved TRICA fuel.

3 The control elements shall have scram capability and shall contain borated graphite, boron carbide powder, i

or boron and its components in solid form as a poison in an aluminum or stainless steel cladding.

} Amendment No. 5 l

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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 a would invalidate the findings of these analyses.

The control elements perform their function through the

,i absorption of neutrons, thus affecting the reactivity ,

of the system. Boron has been found to be a stable and  ;

effective material for this control. l j The reflector serves to conserve neutrons and to reduce the amount of fuel that must be in the core to maintain the chain reaction.

l 3 5.4. Fuel Storace ,i 4

Aeolicability This specification applies to the Dow TRIGA Research j

, Reactor fuel storage facilities.  ;

I Obiective The objective of this specification is the safe storage  ;

of fuel.

Specification j

j All fuel and fueled devices not in the core of the 1

reactor shall be stored in such a way that k.gg shall be less than 0.8 under all conditions of moderation,

, and that will permit sufficient cooling by natural i convection of water or air that temperatures shall not

exceed the Safety Limit.

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Basis

, A value of k egg of less than 0.0 precludes any 1

possibility of inadvertent establishment of a self-j sustaining nuclear chain reaction. Cooling which j maintains temperatures lower than the Safety Limit j prevents possible damage to the devices with subsequent

release of radioactive materials. l i i

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Amendment No. 5 j

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6. ADMINISTRATIVE CONTROLS 6.1. Orzanization 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 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. This structure cuts across the lines of management of The Dow Chemical Company. Th6 indisidual responsible for radiation safety is the Radiation Safety Officer for the reactor who reports on matters of radiation safety to the Radiattua Safety Committee and to the Reactor Operations Committee. The review and audit functions are performed by the Reactor Operations Committee which is composed of at least four persons including the Analytical Laboratery Research Manager, the Radiation Safety Officer, and the Reactor Supervisor.

6.1.2. Responsibility The day to-day responsibility for the safe operatior 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 equally qualified individuals, upon notification of the Facility 1 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 1 Operations Committee. I I

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Amendaent No. 5 a

34 Figure 6.1. Administration Otrector, Ofrector, Coreceste Micnigan Division R& O R&O Managtr. Otttctor, incustrial Hygiene Anaiytical C.Y; crate Laboratory RacialICS Safety Research Manager, Officer. Analytical Corporate Lacoratory Raciation Safety Licensed Offictr. SRCs TRIGA and Ros Ractatico l Safety Cctrattee Raciatien Safety '

Oft 1cer TRIGA

, Rtactor C;erations Committee

] Review Aucit .

, I l

l Reactor SRos

, Sucerviser and Ros 1 j l

1 Amendment No. 5 i i

T 35 6.1.3. Staffing The minimam 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 presencelof 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 unexplained scrams, and e, movement of any in core experiment having an estimated reactivity value greater than

$0.75.

A lirt of reactor facility personnel by name and telephone number shall be readily available in the  ;

control roca for use by the operator, including l management, radiation safety, and other operations persom el.

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Amendment No. 5 1

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.o 6.1.4. Selection and Training of Personnel j

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

i The selection, training, and requalification of  !

operations personnel shall be consistent with all current regulations.

1 Day to day changes in equipment, procedures, and specifications shall be cctmunicated to the facility staff as the cha.ges occur, 6.2. Review and Audit The review and audit functions shall be the responsibility ~  !

c f 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, '

radiation science, or reactor / radiation '

l engineering.

) b. A quorum shall consist of a majority of the  !

members of the ROC. No more than one-half of the voting members present shall be members of the day- ,

to day reactor operating staff,

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

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.

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f. The ROC shall report at least twice per year to

{ the Radiation Safety Committee.

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i Amendment No. 5 l

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s 6.2.2. Review Functions The ROC shall review and approve:

a. every experiment involving fissionable material;

b. experiments or operations which would require a change of core configuration, or a change in the equipment or apparatus associated with the reactor core or its irradiation facilities, or a new piece of apparatus being mounted in the reactor well; except that movement of the neutron source for the purpose of routinely 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 or 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

1. audit reports.

Amendment No. 5

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6.2.3. Audit Function l

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 i reactor facility equipment, systems, structures, or i j methods of operations that affect reactor safety.

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l 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 l

~

appointed by the ROC. At least one of the auditorr. i 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 l

the ROC within three months of the audit.

l Deficiencies that affect reactor safety shall be i J reported to the Facility Director immediately, i i

i b. The ROC shall direct an annual audit of the i facility emergency plan, security plan, and the  ;

reactor operator requalification program. This audit may consist of the annual review of these l' plans for the requalification program.

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i Amendment No. 5 u -. , , - . , . - _ , . _ = _ . ~ , - . - - _ , - - - . - , , . . - - - _ - . - , - . . _ . - . _ , . - - .,-.-, ..,., , ..

i l 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 i order to deal with special or unusua) circumstances.

Such deviations shall be documented and reported immediately to the Reactor Operations Committee.

6.4 Exceriment Review and Acoroval

a. Routine Experiments (as reviewed and defined by the ROC) shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supe rvisor.

I

b. Modified Routine Experiments shall have the written approval of the Reactor Supervisor or a designated ,

Assistant Reactor Supervisor. The written approval I shall include documentation that the bazards have been considered by the reviewer and been found appropriate for this form of experiment.

c. 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 j

, ROC. Experiments which require the approval of the ROC '

through sections 6.2.2.a., 6.2.2.b., or 6.2.2.c. of the Technical Specifications are always Special .

Experiments. l l

J Amendment No. 5  ;

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i 6.5, Recuired Actiqng 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 accordance with section 6.6.2.; and

d. a report shall be prepared for the ROC t 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 i alternate;

b. the occurrence shall be reported to the Facility i Director and to the US NRC as required per section 6.6.2.; and

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

1 Amendment No. 5 l 1

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e 41 6.6. Reoorts 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:

1 a) status of the facility staff, licenses, and j 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 (a) was applicable; i l

d) the unscheduled shutdowns and raasons 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; i 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 I

Amendment No. 5 I

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g) a summary of the radiation exposures received by facility personnel and visitors where such exposures are greater than 25 % of those allowed or recommended in 10 CFR 20, 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 occurence (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: ,

i 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 i of a new facility license authorizing an increase l in reactor power level, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions.

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Amendment No. 5 ,

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-43 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; l

g, fuel inventories, receipts, and shipments;

h. minutes of ROC meetings; l i. records of audits; l

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 I requalification schedule.

6.7.3. The following records shall be retained for the lifetime of the reactor: l i

i a. records of gaseous and liquid radioactive {

effluents released to the environment; i

b. records of the radiation exposure of all i individuals monitored; and

c. drawings of the reactor facility, l

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Amendment No. 5  !