ML12164A784

the Dow Chemical Company, Dtrr Revised Technical Specifications and Supplemental Information to the Dtrr Revised Response to RAI 41, Dated November 10, 2011, in Support of the License Renewal

ML12164A784
Person / Time
Site:	Dow Chemical Company
Issue date:	06/11/2012
From:	O'Connor P Dow Chemical Co
To:	Geoffrey Wertz Office of Nuclear Reactor Regulation
References
Download: ML12164A784 (59)

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The Dow Chemical Company Midland, Michigan 48667 June 11, 2012 Mr. Geoffrey Wertz Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Subject:

The Dow Chemical Company- License No. R-108; Docket No. 50-264 Enclosed are the DTRR Revised Technical Specifications and Supplemental Information to the DTRR revised response to RAI 41, dated November 10, 2011, in support of the license renewal. Also enclosed is a document that clarifies Technical Specification 3.7.2c. These documents are submitted as attachment 1, attachment 2, and attachment 3 respectively.

Should you have any questions or need additional information, please contact the Facility Director, Paul O'Connor, at 989-638-6185.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 11, 2012 Paul O'Connor, Ph.D.

Director Dow TRIGA Research Reactor Subscribed and sworn to before me this day of June, 2012

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Notary Publ c S SMichi n MYCommission Expires .

June 28, 2017 My Commission Expires: "A C  :

O i IIsoil IB 'W cc: Wayde Konze, R&D Director - Analytical Sciences Paul O'Connor, Director Siaka Yusuf, Reactor Supervisor

Attachment I The DTRR Revised Technical Specifications 6/11/2012

Appendix A To FACILITY LICENSE NO. R-108 DOCKET NO. 50-264 TECHNICAL SPECIFICATIONS AND BASES FOR THE DOW TRIGA RESEARCH REACTOR June/2012

TECHNICAL SPECIFICATIONS AND BASES FOR THE DOW TRIGA RESEARCH REACTOR

1. INTRODUCTION 1.1 Scope This document constitutes the Technical Specifications for the Facility License No. 108 as required by 10 CFR 50.36 and supersedes all prior Technical Specifications. This document includes the "Basis" to support the selection and significance of the specifications. Each basis is included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

1.2 Format These specifications are formatted in conformance to NUREG-1537 and ANSI/ANS15.1-2007 guidance.

1.3 Definitions 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.

Audit: An audit is a quantitative examination of records, procedures or other documents after implementation from which appropriate recommendations are made.

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.

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.

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

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Channel Test: A channel test is the introduction of a signal into the channel for verification that it is operable.

Control Rod: A control rod is a device fabricated from neutron absorbing material, which is used to establish neutron flux changes and to compensate for routine reactivity changes. A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged. Types of control rods shall include:

1. Regulating Rod (Reg. Rod): The regulating rod is a control rod having an electric motor drive and scram capabilities. Its position may be varied manually or by the servo-controller.

2. Shim/Safety Rod: A shim/safety rod is a control rod having an electric motor drive and scram capabilities.

Core Lattice Position: The core lattice position is defined by a particular hole in the top grid plate of the core. It is specified by a letter indicating the specific ring in the grid plate and a number indicating a particular position within that ring.

Excess Reactivity: Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff= 1) at reference core conditions.

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 reactor characteristics. Specific experiments shall include:

1. Movable Experiment: A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the core while the reactor is operating.

2. Modified Routine Experiment: 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 approved experiments in that the hazards are neither significantly different from nor greater than the hazards of the corresponding routine experiment.

3. Routine Experiment: A routine experiment is an approved 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.

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4. Secured Experiment: A secured experiment is any experiment 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 a result of credible malfunctions.

5. Special Experiment: Special experiments are experiments which are neither routine experiments nor modified routine experiments.

6. Unsecured Experiment: An unsecured experiment is any experiment or component of an experiment that does not meet the definition of a secured experiment.

Experimental Facilities: Experimental facilities shall include the rotary specimen rack, sample containers replacing fuel elements or dummy fuel elements in the core, pneumatic transfer systems, the central thimble, and the area surrounding the core.

Fuel Element: A fuel element is a single TRIGA fuel element.

Irradiation: Irradiation shall mean the insertion of any device or material that is not a part of the existing core or experimental facilities into an experimental facility so that the device or material is exposed to radiation available in that experimental facility.

Licensed Area: Rooms 51 (51, 51A, 51AA, 51B) and 52 of building 1602.

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

Operable: A system or component shall be considered operable when it is capable of performing its intended function.

Operating: Operating means a component or system is performing its intended function.

Operational Core: An operational core shall be a fuel element core, which operates within the licensed power level and satisfies all the requirements of the Technical Specifications.

Reactivity Worth of an Experiment: The reactivity worth of an experiment is the value of the reactivity change that results from the experiment, being inserted into or removed from its intended position.

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Reactor Operating: The reactor is operating whenever it is not secured or shut down.

Reactor Operator (RO): An individual who is licensed to manipulate the controls of a reactor.

Reactor Safety Systems: Reactor safety systems are those systems, including their associated input channels, which are designed to initiate, automatically or manually, a reactor scram for the primary purpose of protecting the reactor.

Reactor Secured: The reactor is secured when:

1. Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material in the reactor to attain criticality under optimum available conditions of moderation and reflection; or,

2. All of the following exist:

a. The three (3) neutron absorbing control rods are fully inserted as required by technical specifications,

b. The reactor is shutdown,

c. The console key switch is in the "off" position and the key is removed from the console,

d. No experiments are being moved or serviced that have, on movement, reactivity worth exceeding the maximum value allowed for a single experiment, or of one dollar, whichever is smaller, and

e. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods.

Reactor Shutdown: The reactor is shut down when it is subcritical by at least one dollar both in the reference core condition and for all allowed ambient conditions, with the reactivity worth of all installed experiments and irradiation facilities included.

Reference Core Condition: The reference core condition is the condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible

(<$ 0.30).

Review: A review is a qualitative examination of records, procedures or other documents prior to implementation from which appropriate recommendations are made.

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Safety Channel: A safety channel is a measuring channel in the reactor safety system.

Scram Time: Scram time is the elapsed time from the initiation of a scram signal to the time the slowest scrammable control rod is fully inserted.

Senior Reactor Operator (SRO): An individual who is licensed to direct the activities of ROs. Such an individual is also an RO.

Should, Shall, and May: The word "shall" is used to denote a requirement; the word "should" is used to denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recommendation.

Shutdown Margin: Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems and will remain subcritical without further operator action, starting from any permissible operating condition with the most reactive rod in its most reactive position.

Surveillance Intervals: Allowable surveillance intervals shall not exceed the following:

1. Quinquennial - interval not to exceed 70 months

2. Biennial - interval not to exceed 30 months

3. Annual - interval not to exceed 15 months

4. Semiannual - interval not to exceed 7.5 months

5. Quarterly - interval not to exceed 4 months

6. Monthly - interval not to exceed 6 weeks

7. Weekly - interval not to exceed 10 days Unscheduled Shutdown: An unscheduled shutdown is any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations.

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2. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Fuel Element Temperature Applicability 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 a fuel cladding failure will not occur.

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

Basis The important parameter for a TRIGA reactor is the fuel element temperature. If the fuel temperature exceeds the safety limit, a loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel moderator and the cladding. Since the pressure is caused by the presence of air, fission product gases, and hydrogen from the disassociation of fuel moderator, the magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium ratio in the alloy.

According to several reports on Training Reactor and Isotope Production, General Atomics (TRIGA)-type fuels (NUREG-1282; Simnad et al., 1976 and 1981; Simnad and West et al., 1986; West et al., 1986), for stainless steel-clad UZrH 1.65 LEU 8.5 w% TRIGA fuel, GA has shown and U.S. Nuclear Regulatory Commission (U.S. NRC) has accepted that the integrity of the fuel is not compromised if the peak fuel temperature is less than 1150 TC. For aluminum-clad UZrH 1.0 LEU 8 w% TRIGA fuel, the U.S. NRC has accepted that the peak fuel temperature should not exceed 500 °C (NUREG 1537, Appendix 14.1).

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2.2 Limiting Safety System Settings 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 prevent the safety limit from being reached.

Specification The LSSS shall not exceed 300 kW as measured by the calibrated power channels.

Basis Analysis of the Dow TRIGA Research Reactor (DTRR) at 300 kW resulted in a maximum fuel temperature of less than 350 °C following a loss of coolant after infinite hours of operation. Therefore setting the LSSS not to exceed 300 kW provides assurance that the safety limit of 500 'C will not be exceeded (SAR, as supplemented by letter dated December 6, 2011).

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3. LIMITING CONDITIONS FOR OPERATON (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.

Specifications

1. The shutdown margin provided by the control rods shall be more than $0.50 with:

a. Irradiation facilities and experiments in place and the total worth of all non-secured experiments in their most reactive state;

b. The most reactive control rod fully withdrawn; and

c. The reactor in the reference core condition.

2. The excess reactivity measured at less than 10 watts in the reference core condition, with experiments in their most reactive state, shall not be greater than $3.00.

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

4. There shall be a minimum of three operable control rods in the reactor core. A control rod is considered operable when:

a. There is no damage to the control rod or drive assembly; and

b. The scram time meets the requirement in Technical Specification 3.3, specification c.

Bases The value of the minimum shutdown margin assures that the reactor can be safely shut down with the most reactive control rod withdrawn. Assigning specifications to the maximum excess reactivity and maximum reactivity insertion rates, serve as additional restrictions on the shutdown margin and limits the maximum power excursion that could 8

take place in the event of failure of all of the power level safety circuits and administrative controls. The requirement for three operable control rods ensures that the reactor can meet the shutdown specifications (SAR, as supplemented, by letter dated December 6, 2011).

3.2 Reactor Fuel Parameters Limits Applicability This specification applies to all fuel elements.

Obiective The objective of this specification is to maintain integrity of the fuel elements.

Specification The reactor shall not operate with damaged fuel elements, except for the purpose of locating damaged fuel elements. A fuel element shall be considered damaged and must be removed from the core if:

a. The transverse bend exceeds 0.0625 inches (0.159 cm) for stainless steel-clad UZrH 1.65 and 0.125 inches (0.318 cm) for aluminum-clad UZrH 1.oover the length of the cladding;

b. Elongation exceeds 0.125 inches (0.318 cm) for stainless steel-clad UZrH1 .65 and 0.5 inches (1.27 cm) for aluminum-clad UZrH1.o;

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

d. U-235 Burn-up exceeds 50% initial concentration.

Bases Gross failure or obvious visual deterioration of the fuel is sufficient to warrant declaration of the fuel as damaged. The elongation, bend, and burn-up limits are values that have been found acceptable to the U.S. NRC (NUREG-1537).

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3.3 Reactor Control Rods and Safety Systems and Interlocks Applicability These specifications apply to the reactor control rods, safety system channels and interlocks.

Obiective The objective is to specify the minimum number of reactor control rods, the safety system channels and interlocks that shall be available to the operator to assure safe operation of the reactor.

Specification The reactor shall not be operated unless:

a. The safety channels and the interlocks listed in Table 3.3A are operable;

b. The measuring channels listed in Table 3.3B are operable; and

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

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TABLE 3.3A Specifications Minimum Reactor Safety Channels, Interlocks, and Set Points Scram Channels or Interlocks Minimum Operable Scram Set Point or Interlocks Reactor Power Level (NM1000 Not to exceed licensed power level

& NPP1000)' 2 (300kW)

Detector High Voltage 1 Loss of the High Voltage (NPP1000)

Detector High Voltage 1 Loss of the High Voltage (NM1000)

Manual Console Scram 1 Push Button Watchdog (DAC to CSC)

Communication Conflict 1 Not more than 10 sec delay Prevents control rod withdrawal Startup Count Rate (Interlock) 1 when the neutron count rate is less than 2 cps Prevents simultaneous manual Rod Drive Control (Interlock) 1 wi thdrawa l tw o co ntr l withdrawal of two control rods Prevents control rod withdrawal Reactor Period (Interlock) 1 when the period is less than 3 seconds Note: Bypassing of channels and interlocks in this table is not permitted.

1 Any single power level channel may be inoperable while the reactor is operating for the purpose of performing a channel check, channel test, or channel calibration.

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TABLE 3.3B Specifications Measuring Channel 2 Minimum Number Operable NM1000 1 NPP1000 1 Reactor Pool Water Radioactivity 1 Monitor Reactor Pool Water Temperature 1 Monitor Reactor Pool Water Level 1 2 if any required measuring channel becomes inoperable while the reactor is operating, for reasons other than identified in this TS, the channel shall be restored to operation within 5 minutes or the reactor shall be immediately shutdown.

Bases NUREG-1537 recommends at least two independent power level scram channels that provide diversity. This is accomplished by having one power level scram channel as an analog channel. The control rod scram time specification on the three control rods assures that the reactor can be shutdown promptly when a scram signal is initiated and that the reactor can meet the shutdown specifications (SAR, as supplemented, by letter dated December 6, 2011).

Uses of the specified reactor safety channels, set points, and interlocks given in table 3.3A assure protection against operation of the reactor outside the safety limit. The requirements for the specified measurement channels in table 3.3B provide assurance that important reactor operation parameters (power level, water radioactivity, water temperature, and water level) can be monitored during operation. The specification of maximum positive reactivity insertion rate helps assure that the Safety Limit is not exceeded (Dow SAR, as supplemented by U.S. NRC letter dated December 6, 2011).

For footnote 1, taking a single measuring channel off-line for a short duration for the purpose of a channel check, test or calibration is considered acceptable because in some cases, the reactor must be operating in order to perform the channel check, calibration or test. The redundant power level channel provides the scram function for the short period that the other power level channel is out of service. For footnote 2, events which lead to these circumstances are self-revealing to the operator.

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3.4 Reactor Coolant Systems 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.

Obiectives The objectives of this specification include minimization of corrosion of the cladding of the fuel elements and neutron activation of dissolved materials, detection of releases of radioactive materials into the coolant before such releases become significant, ensuring the presence of an adequate quantity of cooling and shielding water in the pool, and prevention of the thermal degradation of the ion exchange resin in the purification system.

Specifications

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

2. The pool water pH shall be in the range of 5.0 to 7.5.

3. The radioactivity of the reactor pool water shall not exceed the limits of 10 CFR 20 Appendix B Table 3 for radioisotopes with half-lives > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. The water shall cover the core of the reactor to a minimum of 15 feet above the core during operation of the reactor.

5. The bulk temperature of the coolant shall not exceed 60 °C during operation of the reactor.

6. There shall be an audible alarm on the coolant level set at 15 ft 10 in above the core.

Bases Increased levels of conductivity in aqueous systems indicate the presence of corrosion products and promote more corrosion. Experience with water quality control at many reactor facilities, including operation of the Dow TRIGA Research Reactor since 1967, has shown that maintenance within the specified limit provides acceptable corrosion control.

Maintaining low levels of dissolved electrolytes in the pool water also reduces the amount of induced radioactivity. Low levels of dissolved electrolytes in the pool water decrease the exposure of personnel to ionizing radiation during operation and maintenance.

The pool water conductivity is monitored continually, except during maintenance.

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Monitoring the pH of the pool water provides early detection of extreme values of pH which could enhance corrosion. Limiting the radioactivity to this level will help ensure that any disposal of pool water, either planned or inadvertent, will be within the limits of 10 CFR

20. This specification also provides verification of absence of fission product leakage.

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. This specification also maintains the height of water above the core used in thermal-hydraulic analyses (SAR, as supplemented by letter dated December 6, 2011).

Maintaining the bulk temperature of the coolant below the specified limit assures minimal thermal degradation of the ion exchange resin. This specification is consistent with the thermal-hydraulic analyses (SAR, as supplemented by letter dated December 6, 2011).

The alarm is audible in the control room as well as outside of the control room, and it alerts operating staff and other people in the building, when the coolant water level is low, to take appropriate action.

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3.5 Ventilation Applicability This specification applies to the reactor room ventilation.

Obiective 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 operating whenever the reactor is operated, fuel is manipulated, or radioactive materials with the potential of airborne releases are handled in the reactor room. The ventilation system is considered operable if:

a. The exhaust and the inlet fans are operating;

b. The external door (Door 10) is closed; and

c. The exhaust louvers are open.

Basis This specification ensures that the ventilation is operating and 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 and Effluents 3.6.1 Radiation Monitoring Systems Applicability These specifications apply to the radiation monitoring systems.

Obiective The objective isto specify the minimum radiation monitoring channels that shall be available to the operator to assure safe operation of the reactor.

Specification The reactor shall not be operated unless the minimum number of each radiation monitoring channel, listed in Table 3.6, are operating.

Table 3.6 Radiation Measuring Channels Number Continuous Air Monitor (CAM) 1 Area Radiation Monitor (ARM)' 1 Environmental Monitor (Film badges) 4 1When the area radiation monitor channel becomes inoperable, operations may continue only if a portable gamma-sensitive ion chamber is utilized as a temporary substitute, provided that the substitute can be observed by the reactor operator, can be installed within I hour of discovery, and not to exceed 60 days.

Bases The CAM provides information of existing levels of radiation and air-borne radioactive materials, including particulate alpha or beta emitters, which could endanger operating personnel or which could warn of possible malfunctions of the reactor or the experiments in the reactor.

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The ARM provides information of existing levels of radiation and airborne radioactive materials which could endanger operating personnel or which could warn of possible malfunctions of the reactor or the experiments in the reactor.

The film badges or any other environmental monitors, placed in the reactor room provide historical records of radiation exposures in the reactor room. One of the four film badges is placed in the control room. Experience at the DTRR showed that the 4 film badges are adequate for monitoring environmental radiation exposures.

For footnote 1, an analysis has shown that it takes more than 60 minutes for the radiation level in the reactor room to exceed the alarm set level of 2mR/hr (SAR, as supplemented by letter dated January 20, 2012, RAI No. 56). Therefore substituting an observable ion chamber within 1 hr assures that the reactor operator has the tool to observe actionable radiation level in the reactor room.

3.6.2 Effluents Applicability This specification applies to the release rate of 4 1Ar.

Obiective The objective is to ensure that the concentration of 41Ar in the unrestricted areas shall be below the applicable effluent concentration value in 10 CFR 20.

Specification The annual average concentration of 41Ar discharged into the unrestricted area shall not exceed 1 x10-9 ýiCi/ml.

Bases Operational experience at DTRR shows that the use of the Rabbit System, an irradiation facility, is the main contributor to the release of 4 1Ar. An analysis has shown that continuous operation of the Rabbit System will result in an annual release of 7.94x10-12 ýtCi/ml, with a corresponding annual TEDE to a member of the public in an unrestricted area of 0.056 mR/yr. (SAR, as supplemented by letter dated June 11, 2012, an additional information to RAI No. 41). The TS value of I x10-9 tCi/ml, which corresponds to a dose of 7.1 mR/yr at the nearest unrestricted area release point, is 30% lower than the 10 CFR 20 Appendix B effluent concentration value, and is therefore sufficient to meet the objective of this specification.

The amount of 41Ar effluent discharged, every year, will be calculated based on the actual number of hours the Rabbit System was operated and the result will be provided in the annual report.

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3.7 Experiments 3.7.1 Reactivity and Position Limits Applicability These specifications apply to experiments installed in the reactor and the irradiation 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.

Specification The reactor shall not be operated unless the following conditions governing experiments exist:

a. The sum of the total absolute value of reactivity worths of all experiments shall not exceed $1.00;

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

c. Experiments shall not occupy adjacent fuel element positions in the B- and C-rings fuel locations.

Bases These specifications are intended to limit the reactivity of the system so that the safety limit would not be exceeded even ifthe experiment were in the B-ring or C-ring, or if the contribution to the total reactivity by the experiment reactivity should be suddenly moved.

The reactivity worth limit of $1.00 for all experiments is intended to prevent the reactor from becoming prompt critical during experiments.

The reactivity limit of $0.75 for movable experiments is designed to prevent an inadvertent reactor pulse from occurring and maintain a reactivity value below the shutdown margin. The specification for the unsecured experiment ($0.75) is consistent with the reactivity insertion behavior analyses (SAR, as supplemented by letter dated December 6, 2011).

The prevention of experiments from occupying adjacent fuel locations in the B and C rings helps to limit the power excursions that may arise due to experiments and to be able to control the reactor within the limits imposed by the license.

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3.7.2 Materials Applicability These specifications apply to experiments installed in the reactor and the irradiation 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.

Specification The reactor shall not be operated unless the following conditions governing experiments exist:

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

b. 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 TNT equivalent 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 TNT equivalent 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 half the design pressure of the container. Such materials shall be packaged in the appropriate containers before being brought into the reactor room or shall be in the custody of duly authorized local, state, or federal officers; and

c. Experiments containing fissionable materials shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is less than 10 micro-curies and the maximum strontium-90 inventory is less than 35 nano-curies.

Basis The specification on corrosive, reactive, explosive and fissionable material is intended to prevent damage to reactor components resulting from failure of any experiment involving these materials. The double encapsulation requirement is also intended to prevent damage to the reactor or the experiments due to release of the listed materials.

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Explosive materials such as gunpowder, dynamite, TNT, nitroglycerin, or PETN require special handling. By limiting their quantities in any experiment to within the material specifications, and by controlling handling according to authorized local, state, and federal laws, potential damages to the reactor fuel or structure are avoided.

The specification on experiments involving fissionable material is intended to reduce the severity of the results of accidental release of airborne radioactive materials to the reactor room or to the atmosphere. By placing a limit on Iodine and Sr-90, the postulated maximum doses to the workers and to members of the public would be below the limits of 10 CFR 20 (SAR, as supplemented by letter dated January 20, 2012, RAI No. 56).

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3.7.3 Experiment Failure and Malfunctions AQplicability These specifications apply to experiments installed in the reactor and the irradiation 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.

Specification Where the possibility exists that the failure of an experiment under (a) normal operating conditions of the experiment or the reactor, (b) credible accident conditions in the reactor or (c) possible accident conditions in the experiment could release radioactive gases or aerosols to the reactor room or the unrestricted area, the quantity and type of material in the experiment shall be limited in activity such that the gaseous activity or radioactive aerosols in the reactor room or the unrestricted area, would not exceed the limits of 10 CFR 20, assuming that:

a. 100% of the gases or aerosols escape from the experiment;

b. 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;

c. 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;

d. 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; and

e. 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.

Bases The specification on aerosol escape 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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The specification on the effluents from any experiment is intended to reduce the severity of any possible release of these fission products which pose the greatest hazard to workers and the general public. These specifications on effluents require specific actions to determine the extent of damage following releases of materials. No theoretical calculations or evaluations are allowed.

The specification on undisturbed column of water above the core is to prevent 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.

The specification on physical inspection of failed experiment helps prevent unforeseen potential problems. Experience has shown that experiments which are reviewed by the staff and reactor operations committee (ROC) can be conducted without endangering the safety of the reactor or exceeding the limits in the technical specifications.

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4. SURVEILLANCE RQUIREMENTS 4.0 General Applicability This specification applies to surveillance requirements of any system related to reactor safety.

Obiective The objective is to verify the operability of any system related to reactor safety.

Specifications

1. Surveillance requirements may be deferred during reactor shutdown (except TS 4.4, items 1 and 2); however, they shall be completed prior to reactor startup unless reactor operation is required for performance of the surveillance. Such surveillance shall be performed as soon as practicable after reactor startup. Scheduled surveillance, which cannot be performed with the reactor operating, may be deferred until a planned reactor shutdown.

2. Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifications reviewed by the Reactor Operations Committee. Asystem shall not be considered operable until after it is successfully tested.

3. Required surveillances of the reactor control and safety systems, pool water level alarm and radiation monitoring systems shall be completed after maintenance of the respective items.

4. Required surveillances of the CAM, ARM, conductivity, pH and pool level shall not be deferred during reactor shutdown.

Basis These specifications relate to changes in reactor systems which could directly affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presently accepted operating criteria.

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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 reactor core excess reactivity, and the reactor shutdown margin shall be measured at least annually and after each time the core fuel is moved or following any change of reactivity by $0.25 or more from the reference core.

Basis Movement of the core fuel could change the reactivity of the core and thus affect both the core excess reactivity and the shutdown margin, as well as affecting the worth of the individual control rods. Evaluation of these parameters is therefore required after any such movement. Without any such movement the changes of these parameters over an extended period of time and operation of the reactor have been shown to be very small so that an annual measurement is sufficient to ensure compliance with the specifications of section 3.1.

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4.2 Reactor Fuel Parameters Applicability This specification shall apply to the fuel elements of the Dow TRIGA Research Reactor.

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

Specification Each fuel element shall be examined visually and for changes in transverse bend and length at least once each five years, with at least 20 percent of the fuel elements examined each year. If a damaged fuel element is identified, the entire inventory of fuel elements shall be inspected prior to subsequent operations.

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. Experience at the Dow TRIGA Research reactor and at other TRIGA reactors indicates that examination of a five-year cycle is adequate to detect problems, especially in TRIGA reactors that are not pulsed. A five-year cycle reduces the handling of the fuel elements and thus reduces the risk of accident or damage due to handling.

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4.3 Control and Safety Systems Applicability 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 Technical Specification 3.3.

Specifications

1. Control rod scram times shall be measured and reactivity insertion rates shall be calculated at least annually and whenever maintenance is performed or repairs are made that could affect the rods or control rod drives.

2. A channel calibration shall be performed for the NM1000 and NPP1000 power level channels by thermal power calibration at least annually.

3. A channel test shall be performed each day the reactor is operated and after any maintenance or repair for each of the six scram channels and each of the three interlocks listed in Table 3.3A.

4. The control rods shall be visually inspected at least biennially.

Bases Measurement of the control rod scram 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.

Variations of the indicated power level due to minor variations of either of the two neutron detectors would be readily evident during day-to-day operation. The specification for thermal calibration of the NM1000 and NPP1000 channels provide assurance that long-term drift of both neutron detectors would be detected and that the reactor shall be operated within the authorized power range.

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.

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Visual inspection of the control rods provides opportunity to evaluate any corrosion, distortion, or damage that might occur in time to avoid malfunction of the control rods.

Experience at the Dow TRIGA Reactor Facility since 1967 indicates that the surveillance specification is adequate to assure proper operation of the control rods. This surveillance complements the rod scram time measurements.

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4.4 Reactor Coolant Systems Applicability These specifications shall apply to the surveillance requirements for the reactor coolant system.

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

Specifications

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

2. A channel check of the pool water level shall be done weekly and before commencement of each day of operation.

3. A channel check of the temperature monitor shall be done during reactor operation and a channel test of the temperature monitor shall be done monthly.

4. A channel test of the pool water level alarm shall be done annually.

Bases Experience at the Dow TRIGA Research Reactor showed that these surveillance specifications on the conductivity, pH, and radioactivity are adequate to detect the onset of degradation of the quality of the pool water in a timely fashion. Evaluation of the radioactivity in the pool water allows the detection of fission product releases from damaged fuel elements or damaged experiments.

Experience also indicates that the surveillance specification on pool water level and pool water temperature are adequate to detect losses of pool water in a timely manner and to enable operators to take appropriate action when the coolant temperature approaches the specified limit. The monthly test of the temperature monitor is also necessary to assure operability of the temperature channel. A channel calibration is performed if required as a result of the channel test.

The pool water level alarm system is a robust unit and therefore the specification of an annual test is sufficient to assure operability of the pool water level alarm.

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4.5 Ventilation Applicability This specification applies to the surveillance of the ventilation system Objective The objective of these specifications is to ensure that the Technical Specification 3.5 is satisfied.

Specification A channel check of the ventilation system shall be performed prior to each day's operation, prior to fuel manipulation, or prior to handling radioactive materials with the potential of airborne releases in the reactor room.

Basis Experience has demonstrated that checks of the ventilation system on the prescribed basis are sufficient to assure proper operation of the system and its control over releases of radioactive material in the reactor room.

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

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

Specifications

1. A channel check shall be made for the CAM and the ARM before commencement of each day of operation, prior to manipulating fuel, or handling experiments or radioactive material which have a potential to become airborne.

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

3. A channel calibration shall be made for the CAM and the ARM at least annually.

4. The environmental monitors shall be changed and evaluated at least semi-annually.

Bases The specifications on the CAM and ARM ensure that they can perform the required functions and that deterioration of the instruments shall be detected in a timely manner when the reactor is operating, prior to manipulating fuel, or handling experiments or radioactive material which have a potential to become airborne. Experience with these instruments has shown that the surveillance intervals are adequate to provide the required assurance.

The continuous air monitor is further checked at least weekly, even if the reactor was not operating to ensure that it is performing its required function.

The frequency of changing and evaluating environmental monitors are also adequate to provide the required record based on past experience with these monitors.

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4.7 Experiments Applicability This specification applies to the surveillance of the experiments.

Objective The objective of these specifications is to ensure that the Technical Specification 3.7 is satisfied.

Specifications

1. The reactivity worth of an experiment shall be estimated or measured, as appropriate before reactor operation with said experiment.

2. An experiment shall not be installed in the reactor or its irradiation facilities unless a safety analysis has been reviewed and approved for compliance with Technical Specification 3.7 by the ROC.

3. ROC approved experiments shall be reviewed prior to irradiation by the Director or a designee.

4. Dose rate on contact for each sample shall be recorded when removed from the experimental facility.

Basis Experience has shown that these specifications verify that experiments can be conducted without endangering the safety of the reactor or exceeding the limits in the technical specifications.

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5. DESIGN FEATURES 5.1 Reactor Site and Building Applicability These specifications shall apply to the Dow TRIGA Research Reactor licensed area. The licensed area includes labs 51 and 52 of building 1602.

Obiectives The objectives of these specifications are to define the licensed area and characteristics of the reactor area.

Specifications

1. The minimum distance from the center of the reactor pool to the boundary of the restricted area shall be 75 feet.

2. The reactor shall be housed in a room with a minimum of 6000 cubic feet volume designed to restrict leakage.

3. 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.

4. Emergency shutdown controls for the ventilation systems shall be located in the reactor control room.

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.

The location of emergency ventilation shutdown controls in the control room assures quick and easy access to these controls by the operator.

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

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

Specifications

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

2. The water lines from the pool water to the heat exchanger shall have anti-siphon holes.

Basis Natural convention water flow has been demonstrated to provide sufficient cooling during reactor operations (SAR, as supplemented by letter dated December 6, 2011).

Anti-siphon holes prevent siphoning of the water out of the pool, should leaks develop in the water lines.

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

Obiective The objective of these specifications is to define certain characteristics of the reactor in order to assure that the design and accident analyses shall be correct.

Specifications

1. The critical core shall be an assembly of stainless-steel or aluminum-clad TRIGA fuel elements in light water.

2. The fuel shall be arranged in a close packed array for operation at full licensed power except for replacement of single individual fuel elements with in-core irradiation facilities or control rod guide tubes, or the start-up neutron source.

3. The aluminum-clad fuel elements shall be placed in the E or F ring of the core.

4. The control rods (Shim1, Shim2 and Regulating rod) 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.

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

6. The structural components of the core shall be limited to aluminum or stainless steel.

7. No fuel shall be inserted or removed from the core unless the reactor is subcritical by more than the worth of the most reactive fuel element.

8. No control rods shall be manually removed from the core for inspection unless it has been shown that the core is subcritical with all control rods fully withdrawn from the core.

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.

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Operation with standard U.S. NRC-approved TRIGA fuel in closed packed array ensures a conservative limitation with respect to the safety limit.

Placement of the aluminum-clad fuel element in the outer rings of the reactor core will help ensure that this element is not exposed to higher than average power levels, thus providing a greater degree of conservatism with respect to the Safety Limit for this one element.

The control rods perform their function through the absorption of neutrons, thus affecting the reactivity of the system.

Boron has been found to be a stable and effective material for this control.

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

The required conditions prior to any fuel movements ensure that an inadvertent criticality will not occur.

The required conditions prior to any control rod movements ensure that an inadvertent criticality will not occur.

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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.

Specifications

1. 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 0.9 under all conditions of moderation, and that will permit sufficient cooling by natural convection of water or air such that temperatures shall not exceed the safety limit.

2. Fuel storage shall be limited to in pool storage only.

Basis A value of keff of less than 0.9 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 fuel elements which has a potential to release radioactive materials.

Limiting fuel storage to in-pool storage only further assures safe storage and is a practice that has been found acceptable to the U.S. NRC (NUREG-1537).

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6. ADMINISTRATIVE CONTROLS 6.1 Organization Individuals at the various management levels, in addition to being responsible for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license, technical specifications, and federal regulations.

6.1.1 Structure The reactor administration shall be related to the Core Research and Development (R&D) of the Dow Chemical Company, Midland, as shown in Figure 6.1.

6.1.2 Responsibility The following specific organizational levels and responsibilities shall exist:

a. Dow Core R&D Director, Analytical Sciences (Level 1): The Dow Core R&D Director for Analytical Sciences is responsible for the Dow TRIGA Research Reactor's license;

b. Dow TRIGA Research Reactor (DTRR) Director (Level 2): The DTRR Director reports to the Dow Core R&D Director, Analytical Sciences, and is accountable for the facility's operation;

c. Reactor Supervisor (Level 3): The Reactor Supervisor, who must be an SRO, reports to the DTRR Director and is responsible for directing the activities of the reactor operators and the senior operators (including training, emergency, security and requalification programs) and for the day-to-day operations and maintenance of the reactor;

d. Radiation Safety Officer, RSO, (Level 3): The RSO reports to the Supervisor, The Dow Industrial Hygiene Expertise Center, and is responsible for directing the activities of health physics personnel including implementation of the radiation safety program; and

e. Reactor Operator and Senior Reactor Operator (Level 4): The Reactor Operators and Senior Reactor Operators report to the Reactor Supervisor and are primarily involved in the manipulation of reactor controls, monitoring of instrumentation, and operation and maintenance of reactor related equipment.

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Figure 6.1. Administration Line Management Reporting

..........................--> Com m unication Reporting 38

6.1.3 Staffing

1. The minimum staffing when the reactor is operating shall be:

a. A licensed Reactor Operator or the Reactor Supervisor in the control room;

b. A second person present in the 1602 Building able to carry out prescribed instructions; and

c. If neither of these two individuals is the Reactor Supervisor, the Reactor Supervisor shall be readily available on call. Readily available on call means an individual who:

I. Has been specifically designated and the designation is known to the operator on duty, I1. Can be rapidly contacted by phone by the operator on duty, and III. Is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g., 30 minutes or within a 15-mile radius);

2. A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator. The list shall include:

a. DTRR Director;

b. Reactor Supervisor;

c. Radiation Safety Officer; and

d. Any Licensed Reactor Operator or Senior Reactor Operator;

3. Events requiring the direction of the Reactor Supervisor:

a. Initial startup and approach to power of the day;

b. All fuel or control rod relocations and maintenance within the reactor core region;

c. Relocation of any in-core experiment or irradiation facility with a reactivity worth greater than $0.75; and

d. Recovery from unplanned or unscheduled shutdown or significant power reduction.

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

The selection, training and requalification of operations personnel should be in accordance with ANSI/ANS 15.4 - 1988; R1999, "Standard for the Selection and Training of Personnel for Research Reactors."

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

6.2.1 Composition and Qualification The ROC shall consist of at least four members who are knowledgeable in fields which relate to engineering and nuclear safety. The ROC composition shall include the Dow Core R&D Director, Analytical Sciences, (Level 1) who shall be designated the chair of the committee; Facility Director, (Level 2), the Reactor Supervisor (Level 3); the Radiation Safety Officer; and one or more persons who are competent in the field of reactor operations, radiation science, or reactor/radiation engineering. The ROC shall be appointed by Level 1 management.

6.2.2 ROC Rules The operations of the ROC shall be in accordance with written procedures including provisions for:

a. Quorums (majority of the members of the ROC, no more than one-half of the voting members present may be of the operating staff);

b. Meeting frequency (at least annually and as often as required to transact business);

c. Minutes of the meetings (shall be reviewed and approved at the following meeting and kept as records for the facility);

d. Voting rules (Members of the ROC may be polled by telephone or email for guidance and approvals); and

e. Communications (the ROC shall report at least twice per year to the Radiation Safety Committee).

6.2.3 ROC Review Function The ROC shall perform:

a. Review all changes made under 10 CFR 50.59;

b. Review of all new procedures and substantive changes to existing procedures; 41

c. Review of proposed changes to the technical specifications, license or charter;

d. Review of violations of technical specifications, license, or violations of internal procedures or instructions having safety significance;

e. Review of operating abnormalities having safety significance;

f. Review of all events from reports required by Technical Specifications 6.6.1 and 6.7.2; and

g. Review of audit reports.

6.2.4 ROC Audit Function The ROC shall audit reactor operations at least annually. The annual audit shall include at least the following:

a. facility operations for conformance to the technical specifications and applicable license or charter conditions;

b. the retraining and requalification program for the operating staff;

c. the results of action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operation that affect reactor safety;

d. the emergency response plan and implementation procedures;

e. 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; and

f. a written report of the audit shall be submitted to the ROC within three months of the audit.

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6.3 Radiation Safety The Radiation Safety Officer shall be responsible for implementing the radiation safety program for the Dow TRIGA Research Reactor. The requirements of the radiation safety program are established in 10 CFR 20. The program should use the guidelines of the ANSI/ANS-15.11-1993; R2004, "Radiation Protection at Research Reactor Facilities."

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6.4 Procedures Written operating procedures shall be adequate to assure the safety of operation of the reactor, but shall not preclude the use of independent judgment and action, should the situation require such. Operating procedures shall be in effect for the following items:

a. Startup, operation, and shutdown of the reactor;

b. Implementation of required plans such as the emergency plan and security plan;

c. Emergency and abnormal operating events, including facility shutdown;

d. Fuel loading, unloading and movement within the reactor;

e. Maintenance of major components of systems that could have an effect on reactor control and safety (control rod removal or installation);

f. Surveillance checks, tests, calibrations and inspections required by the technical specifications or those that have an effect on reactor safety;

g. Radiation protection;

h. Administrative controls for operations and maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity; and

i. Use, receipt, and transfer of by-product material held under the reactor license.

Substantive changes to the above procedures shall be made only after review by the ROC.

Except for radiation protection procedures, unsubstantive changes shall be approved, prior to implementation, by the Reactor Supervisor and documented by the Reactor Supervisor within 120 days of implementation. Unsubstantive changes to radiation protection procedures shall be approved, priorto implementation, by the RSO and documented by the RSO within 120 days of implementation.

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, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the next working day, to the Reactor Supervisor.

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6.5 Experimental Review and Approval Approved experiments shall be carried out in accordance with 10 CFR 20, 10 CFR 50.59 and the DTRR TS, operating and administrative procedures. Procedures related to experiment review and approval shall include:

a. All new experiments or a new class of experiments shall be reviewed and approved by the Reactor Operations Committee, and approved in writing by the Level 2 or designated alternates prior to initiation; and

b. Substantive changes to previously approved experiments or a previously approved class of experiments shall be made only after review and approval by the Reactor Operations Committee and approved in writing by the Level 2 or designated alternates prior to initiation; Minor changes that do not significantly alter the experiments may be approved by the Level 3 or higher.

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6.6 Required Actions 6.6.1. Actions to Be Taken in Case of Safety Limit Violation In the event a safety limit (fuel temperature) is exceeded:

a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the U.S. NRC; and

b. An immediate notification of the occurrence shall be made to the Reactor Supervisor, DTRR Director, Level 1, ROC, U.S. NRC; and

c. A report, and any applicable follow-up report, shall be prepared and reviewed by the ROC. The report shall describe the following:

i. Applicable circumstances leading to the violation including, when known, the cause and contributing factors; ii. Effects of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public; and iii. Corrective action to be taken to prevent recurrence.

6.6.2. Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2 Other than a Safety Limit Violation For all events which are required by regulations or Technical Specifications to be reported to the U.S. NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under Section 6.7.2, except a safety limit violation, the following actions shall be taken:

a. The reactor shall be secured and the Reactor Supervisor and Director notified;

b. Operations shall not resume unless authorized by the Reactor Supervisor and the Director;

c. The Reactor Operations Committee shall review the occurrence at their next scheduled meeting; and

d. A report shall be submitted to the U.S. NRC in accordance with Section 6.7.2 of these Technical Specifications.

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6.7 Reports 6.7.1. Annual Operating Reports A report shall be created and submitted, annually, to The Document Control Desk U.S. NRC, Washington, DC. The report shall include the following:

a. Status of the facility staff and licenses;

b. A narrative summary of reactor operating experience, including the energy produced by the reactor, or the hours the reactor was critical, or both;

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 they are allowed without prior authorization by the U.S. NRC 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);

g. A summary of the radiation exposures received by facility personnel and visitors where such exposures are greater than 25% of those allowed in 10 CFR 20; and

h. A summarized result of any environmental surveys performed outside the facility.

47

6.7.2. Special Reports In addition to the requirement of applicable regulations, and in no way substituting therefore, reports shall be made by the Level 1 manager to the U.S. NRC as follows:

1. A report not later than the following working day by telephone and confirmed in writing by facsimile to the U.S. NRC Headquarters Operations Center, and followed by a written report that describes the circumstances of the event within 14 days to the Document Control Desk, U.S. NRC, Washington, DC, 20555 of any of the following:

a. Violation of the safety limit;

b. Release of radioactivity from the site above limits;

c. Operation with actual safety system settings for required systems less conservative than the limiting safety system settings specified in Technical Specification 2.2;

d. Operation in violation of limiting conditions for operation established in the Technical Specifications;

e. 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 caused by maintenance, then no report is required;

f. Any unanticipated or uncontrolled change in reactivity greater than one dollar.

Reactor trips resulting from a known cause are excluded;

g. Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary; or

h. An observed inadequacy in the implementation of either administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

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2. A written report shall be sent, within 30 days, to the Document Control Desk, U.S. NRC, Washington, DC, 20555, of either:

a. Permanent changes in the facility staff involving the Level 1 and 2 personnel; or

b. Significant changes in the transient or accident analysis report as described in the Safety Analysis Report.

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6.8 Records 6.8.1. The following records shall be kept for a minimum period of five years or for the life of the component involved if less than five years:

1. Normal reactor operation (but not including supporting documents such as checklists, log sheets, etc., which shall be maintained for a period of at least one year);

2. Principal maintenance activities;

3. Reportable occurrences;

4. Fuel inventories, receipts, and shipments;

5. ROC meetings and audit reports;

6. Reactor facility radiation and contamination surveys;

7. Surveillance activities as required by the Technical Specifications; and

8. Approved changes in the operating procedures; and

9. Experiments performed by the reactor.

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6.8.2. Records to be Retained for at Least One Certification Cycle Records of the retraining and requalification of Reactor Operators and Senior Reactor Operators shall be retained for at least one complete requalification schedule and be maintained at all times the individual is employed or until the certification is renewed. For the purpose of this technical specification, a certification is an NRC issued operator license.

6.8.3. Records to be Retained for the Lifetime of the Reactor Facility

1. Gaseous and liquid radioactive effluents released to the environment;

2. Radiation exposure of all individuals monitored;

3. Offsite environmental monitoring surveys;

4. Drawings of the reactor facility; and

5. Reviews and reports pertaining to a violation of the safety limit, the limiting safety system setting, or a limiting condition of operation.

51

f /]

I "II Attachment 2 The DTRR Supplement to RAI-41 6/11/2012

Supplement to the response to RAI-41 Dose to an unrestricted public from 41Ar effluent release, calculated using a dose transport model, an additional information to the response to RAI-41 that was submitted November 11, 2011 Introduction The following calculation provides the dose, to a public individual in the nearest unrestricted area to the DTRR, from 4 1Ar. The calculation assumes that the reactor and the rabbit system were operated continuously for a year and the individual is standing there the whole time. The calculation demonstrates that, due to the small amount of rabbit tube volume, the amount of 41Ar generated and the subsequent dose to an individual in the nearest unrestricted area is significantly small.

Calculation The concentration of a radionuclide at a given distance from a point source is calculated by (Regulatory Guide 1.111, Nuclear Regulatory Commission, 1977):

h2 nil 2.032Q Zsecavg =N Ui C-

where, Zsec avg = Concentration of radionuclide in the air, averaged over a 22.5 degree sector (Ci/m 3) nij = Number of hours that wind is blowing in direction i (towards receptor) in wind speed group j [From wind speed and direction data]

N = Total number of hours of wind speed and direction data Q = Release rate of radionuclide (Ci/sec) [calculated to be 0.49 uCi/min, or 8.17e-9 Ci/sec],

from RAI 41, Uj = Average wind speed in wind speed group j (m/sec) [From wind speed and direction data]

= Vertical diffusion coefficient (m) [From Regulatory Guide 1.111 (U.S. Nuclear Regulatory Commission, 1977), dependent on stability class]

I L = Vertical plume spread with a volumetric correction for a release within the building wake cavity [equivalent to G, to conservatively not take credit for building wake effects]

x = Distance from release point to receptor (m) [23 m]

h = Release height (m) [Assumed to be a ground release (h=0) because vent height is less than two times the building height and the exit velocity is less than 5 times the horizontal wind velocity]

Using this formula and wind speed and direction data from Flint, MI as supplied with the CAP88-PC version 2.1 code (U.S. Environmental Protection Agency, 2000), the concentration of Ar-41 at the nearest fence, which is 75 feet to the west of the release point, is calculated to be 7.94e-12 Ci/m3 (7.94e-12 tCi/ml)

If a member of the public was continuously present at this location for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, 365 days per year, the dose to this individual from submersion would be calculated by:

D = Zsecavg x DCF x 8760 Where D = Dose from radionuclide (mremnyr)

Xsec avg = Concentration of radionuclide in the air, averaged over a3 22.5 degree sector (Ci/m3) 3 DCF = Dose Conversion Factor for Submersion (mremrhr/Ci/m ) [8.03 x 10 5 mrem/hr/Ci/m for Ar-41 from FGR 11 (U.S. Environmental Protection Agency, 1988)]

8760 = Hours per year The dose would be 0.056 mremlyr.

Attachment 3 The DTRR Clarification of the term fissionable material in TS 7.3.2c 6/11/2012

Clarification to Technical Specification 3.7.2c The term "fueled experiment", in the previous Technical Specification, was changed to

experiment containing fissionable material" to more realistically describe such an experiment.

The fissionable materials being referred to are typically analytical standards containing trace levels of fissionable materials. Such standards are used as relative standards when performing Neutron Activation Analysis for material characterization with the Dow TRIGA Research Reactor.