ML11318A281

the Dow Chemical Company, Revised Submittal for the Technical Specification in Support of the License Renewal

ML11318A281
Person / Time
Site:	Dow Chemical Company
Issue date:	11/10/2011
From:	O'Connor P Dow Chemical Co
To:	Geoffrey Wertz Division of Policy and Rulemaking
References
Download: ML11318A281 (45)

Text

The Dow Chemical Company Midland, Michigan 48667 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 is the revised submittal for the technical specification in support of the license renewal.

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 November 10, 2011 Paul O'Connor, Ph.D.

Director Dow TRIGA Research Reactor Subscribed and sworn to before me this J day of November, 2011 "1111111161""',F

,,,". _o.. . . .... 4, Notary Pub County, Michigan My Commission Expires: June 28.20'7 cc: Wayde Konze, R&D Director - Analytical Sciences Paul O'Connor, Director Siaka Yusuf, Reactor Supervisor

TECHNICAL SPECIFICATIONS FOR THE DOW TRIGA RESEARCH REACTOR FACILITY LICENSE R-108 AMENDMENT 9 Effective with License renewal This document includes the Technical Specifications and the bases for the Technical Specifications. The bases provide the technical support for the individual Technical Specifications and are included for information purposes only. The bases are not part of the Technical Specifications and they do not constitute limitations or requirements to which the licensee shall adhere.

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.

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 which the channel measures. Calibration shall encompass the entire channel, including equipment, actuation, alarm, or trip and shall include a Channel Test.

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

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

1.6. Confinement - Confinement is an enclosure of the facility that is designed to limit the release of effluents between the enclosure and its environment through controlled or defined pathways 1.7. Control rod - A control rod is a device fabricated from neutron-absorbing material or fuel, or both, that is used to establish neutron flux changes and to compensate for routine reactivity losses. The control rods are coupled to the control rod drive systems in a way that allows the control rods to perform a safety function. The Regulating rod is a low worth control rod used to maintain an intended power level that need not have scram capability and may have a fueled follower. Its position may be varied manually or by a servo-controller. The DTRR Regulating rod is scrammable and it does not have fueled follower.

1.8. Core Configuration - The core configuration includes the number, type, or arrangement of fuel elements, reflector elements, and regulating/control/transient rods occupying the core grid.

1.9. Damaged Fuel- Damaged TRIGA fuel is defined for stainless steel-clad UZrH 1.65 as the fuel whose transverse bend exceeds 0.0625 inches (0.159 cm) over the length of the cladding and elongation exceeds 0.125 inches (0.318). For aluminum clad fuel UZrH1 o, ifthe transverse bend exceeds 0.125 inches (0.318) and elongation exceeds 0.5 in (1.27 cm), the fuel is considered damaged. Fuel is also defined as damaged when burn-up of U-235 exceeds 50% of the initial concentration. Also, if there is a clad defect which releases fission product, the fuel is defined as damaged.

1.10. Excess reactivity - Excess reactivity is that amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (kf=l) at reference core conditions 01

1.11. 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.

1.12. Experimental Facilities 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.

1.13. Licensed area - Rooms 51 (51, 51A, 51AA 51 B) and 52 of building 1602 1.14. Limiting Conditions for Operation - Limiting Conditions for Operation (LCO) are regulatory established constraints on equipment and operational characteristics which shall be adhered to during operation of the reactor.

1.15. Limiting Safety System Settings (LSSSs) - LSSSs are those limiting values for settings of safety channels by which point protective action must be initiated. The LSSSs are chosen so that automatic protective action will terminate the abnormal situation before a safety limit is reached.

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

1.17. Modified Routine Experiments - Modified routine experiments are experiments which have not been designated as routine experiments or which have not been performed previously, but are similar to routine approved experiments in that the hazards are neither significantly different from nor greater than the hazards of the corresponding routine experiment.

1.18. Movable Experiment - A movable experiment is an experiment intended to be moved in or near the core or into and out of the reactor while the reactor is operating.

1.19. Operable - A component or system is operable ifit is capable of performing its intended function.

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

1.21. Protective Action - protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.

1.22. Radiation Safety Committee (RSC) - The RSC is responsible for the administration of all Dow Midland location activities involving the use of radioactive materials and radiation sources including assuring compliance with US NRC regulations.

1.23. 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 removal from its intended position.

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

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

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1.26. Reactor Secured - The reactor is secured when (1) Eitherthere is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderation and reflection; (2) Orthe following conditions exist:

(a) the console switch is in the "off" position, the key is removed from the switch, and the key is in the control of a licensed reactor operator or stored in a locked storage area; and (b) sufficient control rods are inserted to assure that the reactor is subcritical by a margin greater than $1.00 at reference core conditions (c) 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 (d) no experiments in or near the core are being moved or serviced that have, on movement, a reactivity worth exceeding $0.75.

1.27. Reactor Shutdown - The reactor is shutdown if it is subcritical by at least one dollar in the reference core condition with the reactivity worth of all installed experiments included.

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

1.29. Reference Core Condition - The Reference Core Condition is that condition when the core is at ambient temperature (cold) and the reactivity worth of xenon in the fuel is negligible (less than $.30).

1.30. Research Reactor- A Research Reactor is a device designed to support a self sustaining nuclear chain reaction for research, development, education, training, or experimental purposes, and which may have provisions for the production of radioisotopes.

1.31. Reportable Occurrence - A Reportable Occurrence is any of the following a) Operation with actual safety system settings for required systems less conservative than the limiting safety system settings specified in Technical Specification 2.2.

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

c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is caused by maintenance, then no report is required.

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

Reactor trips resulting from a known cause are excluded.

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e) Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary.

f) 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..

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

1.32. 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.

1.33. 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 radioactivity. The principal physical barrier is the fuel element cladding.

1.34. Scram time - Scram Time is the elapsed time required to fully insert the control rods following the actuation of a SCRAM signal.

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

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

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

1.38. Special Experiments - Special experiments are experiments which are neither routine experiments nor modified routine experiments.

1.39. TRIGA 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 1.0-1.1 for aluminum-clad TRIGA elements and in the range of 1.6-1.7 for stainless-steel-clad TRIGA elements.

1.40. 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 LIMITS AND LIMITING SAFETY SYSTEM SETTINGS:

2.1. Safety Limit (SL)

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 no damage to the fuel element will result.

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

Basis 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 1.6 LEU 8.5 uranium weight percent TRIGA fuel, GA has shown and NRC has accepted that the integrity of the fuel is not compromised under the following cases and conditions:

For cladding temperature at or less than 500 degree C, peak fuel temperature at or less than 1150 degrees C For cladding temperature greater than 500 degrees C, peak fuel temperature at or less than 950 degrees C And for aluminum-clad UZrH1.0 LEU 8 w/o TRIGA fuel, NRC has accepted that the peak fuel temperature should not exceed 500 degrees C. (NUREG 1537, Appendix 14.1 page 4)

A loss-of-coolant analysis of the Dow TRIGA reactor showed that the maximum fuel temperature would be less than 3500C following infinite hours of operation at 300 kilowatts terminated by the instantaneous loss of water.

These temperatures are well below the region where the a+ d + g' to a + d phase change occurs in ZrH 1.1 (560 C).

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2.2. Limitinq Safety System Settings (LSSS)

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

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

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

Basis The LSSS which does not exceed 300 kW provides a considerable safety margin. One TRIGA reactor (General Atomics, Torrey Pines) showed a maximum fuel temperature of 350 0C during operation at 1500 kilowatts.

Analysis of the DTRR at 300kW showed a maximum fuel temperature of less than 3500C following a loss of coolant after infinite hours of operation.

Therefore setting the LSSS not to exceed 300kW further provides an assurance on the safety limit of 5000C e6

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

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

1. The reactor shall be shut down by more than $.50 with the most reactive control rod fully withdrawn, the other two control rods fully inserted at reference core conditions, including the reactivity worth of all experiments in their most reactive positions.

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.

Bases

1. The value of the minimum shutdown margin assures that the reactor can be safely shut down using only the two least reactive control rods.

2. The assignment of a specification to the maximum excess reactivity serves as an additional restriction on the shutdown margin and limits the maximum power excursion that could take place in the event of failure of all of the power level safety circuits and administrative controls.

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3.2. Reactor Core Configuration Applicability This specification applies to the core configuration.

Obiective The objective of this specification is to assure that the safety limit will not be exceeded due to power peaking effects.

Specifications

1. The critical core shall be an assembly of standard NRC-approved stainless-steel-clad 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 (1) replacement of single individual fuel elements with in-core irradiation facilities or control rod guide tubes and (2) the start-up neutron source.

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

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

5. 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.

6. A control rod shall not be manually removed from the core for inspection unless it is demonstrated that the core is subcritical excluding the worth of the control rods.

Bases

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

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

3. 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.

4. This specification further ensures conservative limitation with respect to release of fission products.

5. This specification ensures assure that the safety limit will not be exceeded due to movement of fuels

6. This specification ensures assure that the safety limit will not be exceeded due to movement of control rods for inspection.

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

Objective The objective of these specifications is to assure that all reactor control and safety systems and safety related instrumentation are operable to the minimum acceptable standards during operation of the reactor.

Specifications

1. There shall be a minimum of two scram capable safety channels, one of which must be analog.

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

3. 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.

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

5. The reactor shall not be operated unless the measuring channels listed in Table 3.3B are operable.

6. Positive reactivity insertion rate by control rod motion shall not exceed

$.20 per second.

7. Bypassing of channels and interlocks in table 3.3A is not permitted.

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Bases

1. Safety channels with scram capability utilizing analog circuitry have been proven acceptable by more than thirty years of experience.

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

3. 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.

4. Use of the specified reactor safety channels, set points, and interlocks given in table 3.3A assures protection against operation of the reactor outside the safety limits.

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

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

7. This further assures protection against operation of the reactor outside the safety limits.

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TABLE 3.3A.

MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS Scram Channels Scram Channel Minimum Operable Scram Set point Reactor Power Level 2 Not to exceed licensed NM1000 & NPP1000 power NPP1 000 Failure of the detector Detector High-Voltage Power Supply high voltage power supply NM1000 1 Failure of the detector Detector High-Voltage Power Supply high voltage power supply 1

Manual Scram Not applicable Watchdog (DAC to CSC) Not applicable Interlocks Interlock/Channel Function Startup Count-rate Prevent control rod withdrawal when the neutron count rate is less than 2 cps Rod Drive Control Prevent simultaneous manual withdrawal of two control elements by the control rod drive motors Reactor Period Prevent control rod withdrawal when the period is less than 3 seconds 011

TABLE 3.3A BASES FOR REACTOR SAFETY CHANNELS AND INTERLOCKS Scram Channels Scram Channel Bases Reactor Power Level Provides assurance that the reactor shall be shut down automatically before the safety limit can be exceeded Reactor Power Channel Provides assurance that the reactor Detector Power Supplies cannot be operated without power to the neutron detectors which provide input to the NM1000 and NPP1 000 power channels Manual Scram Allows the operator to shut the reactor down at any indication of unsafe or abnormal conditions Watchdog Ensures adequate communications between the Data Acquisition Computer (DAC) and the Control System Computer (CSC) units.

Interlocks Interlock/Channel Bases Startup Count-rate Provides assurance that the signal in the NM1000 channel is adequate to allow reliable indication of the state of the neutron chain reaction.

Rod Drive Control Limits the maximum positive reactivity insertion rate Reactor Period Prevents operation in a regime in which transients could cause the limiting safety system setting to be exceeded

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TABLE 3.3B MEASURING CHANNELS Measuring channel Minimum Number Operable NM1000 1 NPP1000 1 Water Radioactivity 1 Monitor Water Temperature Monitor TABLE 3.3B BASES FOR MEASURING CHANNELS Measuring Channel Basis NM1000 Provides assurance that the reactor power level can be adequately monitored.

NPP1000 Provides assurance that the reactor power level can be adequately monitored.

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

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

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3.4. Reactor Coolant System Applicability These specifications apply to the quality of the coolant in contact with the fuel cladding, to the level of the coolant in the pool, and to the bulk temperature of the coolant. The coolant is de-ionized water Obiectives The objectives of this specification are:

1. to minimize corrosion of the cladding of the fuel elements and minimize neutron activation of dissolved materials,

2. to detect releases of radioactive materials to the coolant before such releases become significant,

3. to ensure the presence of an adequate quantity of cooling and shielding water in the pool, and

4. to prevent thermal degradation of the ion exchange resin in the purification system.

Specifications

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

2. The pool water pH shall be in the range of 5.5 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 depth of 15 feet 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 15ft 1Oin above the core.

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Bases

1. 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 continuously, except during maintenance.

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

3. 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.

4. 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.

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

6. 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. Confinement Applicability 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.

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 operating if:

1. the exhaust and the inlet fans are on,

2. the external door (Door 10) is closed and

3. the exhaust louvers are open Basis This specification ensures that the confinement is configured to control any releases of radioactive material during fuel handling, reactor operation, or the handling of possible airborne radioactive material in the reactor room.

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

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

Specifications

1. A Continuous Air Monitor (CAM) (with readout meter and audible alarm) to measure radioactive particulates in the reactor room shall be operating during operation of the reactor.

2. The Area Monitor (AM) (with readout meter and audible alarm) in the reactor room shall be operating during operation of the reactor or when work is being done on or around the reactor core or experimental facilities. During short periods of repair to this monitor, not to exceed sixty days, reactor operations or work on or around the core or experimental facilities may continue while a portable gamma-ray sensitive ion chamber is utilized as a temporary substitute, provided that the substitute can be monitored by the reactor operator.

3. An environmental monitor such as a film badge, thermo-luminescent dosimeter or other device shall be placed in the reactor room.

Bases

1. 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.

2. The radiation monitors provide information of existing levels of radiation and air-borne radioactive materials which could endanger operating personnel or which could warn of possible malfunctions of the reactor or the experiments in the reactor.

3. The film badge or any other environmental monitor provides a historical record of radiation exposure in the reactor room.

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

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

Specifications

1. Operation of the reactor for any purpose shall require the review and approval of the appropriate persons or groups of persons, except that operation of the reactor for the purpose of performing routine checkouts, where written procedures exist for those operations, shall be authorized by the written procedures. An operation shall not be approved unless the evaluation allows the conclusion that the failure of an experiment will not lead to the direct failure of a fuel element or of any other experiment.

2. The sum of the total absolute value of reactivity worths of all 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.

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 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.

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6. Experiment materials, except fuel materials, which could off gas, sublime, volatilize or produce aerosols under (a) normal operating conditions of the experiment or the reactor, (b) credible accident conditions in the reactor or (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 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.

c. For materials whose boiling point is above 55 C and where vapors formed by boiling this material could escape only through an undisturbed column of water above the core, the assumption shall be used that 10% of these vapors escape.

7. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 10 micro curies and the maximum strontium-90 inventory is no greater than 35 nanocuries.

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

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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.

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

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

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

6. This specification is intended to reduce the severity of the results of accidental release of airborne radioactive materials to the reactor room or the atmosphere.

7. This specification is intended to reduce the severity of any possible release of these fission products which pose the greatest hazard to workers and the general public.

8. This specification requires specific actions to determine the extent of damage following releases of materials. No theoretical calculations or evaluations are allowed.

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

Experience has shown that experiments which are reviewed by the staff and reactor operations committee can be conducted without endangering the safety of the reactor or exceeding the limits in the technical specifications

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

biennially - not to exceed 30 months annually - not to exceed 15 months semi-annually - not to exceed seven and one-half months quarterly - not to exceed four months monthly - not to exceed six weeks weekly - not to exceed 10 days daily - shall be done before the commencement of operation each day of operation Established frequencies shall be maintained over the long term, so, for example, any monthly surveillance shall be performed at least 12 times during a calendar year of normal operation. If the reactor is not operated for a period of time exceeding any required surveillance interval, that surveillance task shall be performed before the next operation of the reactor. Any surveillance tasks which are missed more than once during such a shut-down interval shall be performed only once before operation of the reactor. Surveillance tasks scheduled daily or weekly which cannot be performed while the reactor is operating may be postponed during continuous operation of the reactor over extended times. Such postponed tasks shall be performed 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.

Required surveillances of the CAM and ARM shall not be deferred for extended reactor shutdown.

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.

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4.1. Reactor Core Parameters Applicability 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.

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

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

Specification Each fuel element shall be 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.

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

a) The transverse bend exceeds 0.0625 inches (0.159 cm) for stainless steel-clad UZrH 1.6 and 0.125 inches (0.318 cm) for alum inum-clad UZrH1 .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 UZrH.o .

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

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

• 23

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

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

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

3. A channel test shall be performed at least daily and after any maintenance or repair for each of the six scram channels and each of the three interlocks listed in table 3.3A.

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

Bases

1. Measurement of the control rod drop time and compliance with the specification indicates that the control rods can perform the safety function properly. Measurement of the control rod withdrawal speed ensures that the maximum reactivity addition rate specification will not be exceeded.

2. Variations of the indicated power level due to minor variations of either of the two neutron detectors would be readily evident during day-to-day operation. The specification for thermal calibration of the NM1 000 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.

3. The channel tests performed daily before operation and after any repair or maintenance provide timely assurance that the systems will operate properly during operation of the reactor.

4. Visual inspection of the control rods provides opportunity to evaluate any corrosion, distortion, or damage that might occur in time to avoid malfunction of the control rods. Experience at the Dow TRIGA Reactor Facility since 1967 indicates that the surveillance specification is adequate to assure proper operation of the control rods. This surveillance complements the rod drop time measurements.

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

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

Specifications

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

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

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

4. The pool water level alarm will be tested annually.

Bases

1. Experience at the Dow TRIGA Research Reactor shows that this specification is adequate to detect the onset of degradation of the quality of the pool water in a timely fashion. Evaluation of the radioactivity in the pool water allows the detection of fission product releases from damaged fuel elements or damaged experiments.

2. Experience indicates that this specification is adequate to detect losses of pool water in a timely manner.

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

4. Experience indicates that this specification is sufficient to assure operability of the pool water level alarm.

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4.5 Confinement Applicability This specification applies to the surveillance of the confinement system Obiective The objective of these specifications is to ensure that the specifications of section 3.5 are satisfied.

Specification The ventilation system shall be verified as operating and the external door (Door 10) shall be verified as 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 Experience has shown that this specification ensures that the confinement is configured to control any releases of radioactive material during fuel handling, reactor operation, or the handling of possible airborne radioactive material in the reactor room.

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

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

Specifications

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

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

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

Bases

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

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

3. Experience with these monitors has shown that the evaluation intervals are adequate to provide the required record.

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

Obiective The objective of these specifications is to ensure that the specifications of section 3.7 are satisfied.

Specification ROC approved experiments shall be reviewed prior to irradiation by the Director or a Designee. 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 Buildinq Applicability These specifications shall apply to the Dow TRIGA Research Reactor licensed area. The licensed area includes labs 51 and 52 of building 1602.

Oblectives The objectives of these specifications are to define the licensed area and characteristics of the confinement.

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 of about 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.

Bases

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

2. Restriction of leakage, in the event of a release of radioactive materials, can contain the materials and reduce exposure of the public.

3. Release of gases at a minimum height of 8 feet reduces the possibility of exposure of personnel to such gases.

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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.

Specification

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

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

2. This has been proved to be effective in preventing siphon of the water out of the pool should leaks develop in the water lines.

030

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.

Specification

1. The fuel shall be standard NRC-approved TRIGA fuel, 8.5/20

2. The control rods (Shim 1, 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.

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

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

5. The reactor shall be made subcritical by inserting all the control rods and removing two fuel elements from B or C rings before any other fuel shall be inserted or removed from the core.

6. The reactor shall be made subcritical by inserting all the control rods and removing two fuel elements from B or C rings before any control rod shall be removed for inspection or maintenance.

Bases

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

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

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

4. To ensure conservative control of the reactor structure.

5. To ensure that the reactor is subcritical by more than the worth of the most reactive fuel before any fuel movement.

6. To ensure that the reactor is subcritical by more than the worth of the most reactive fuel before control rod is removed.

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

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

Specification

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 that temperatures shall not exceed the Safety Limit.

2. Fuel storage is limited to in pool storage only.

Basis

1. 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 devices with subsequent release of radioactive materials.

2. The in-pool storage rack has been characterized for handling fuel according to specifications.

032

6. ADMINISTRATIVE CONTROLS 6.1.Orqanization The Dow TRIGA Research Reactor is owned and operated by The Dow Chemical Company. The reactor is administered and operated through the Analytical Sciences Laboratory of the Michigan Division of Dow Chemical USA and is located in 1602 Building of the Analytical Sciences Laboratory at the Midland, Michigan location of the Michigan Division.

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

The individual responsible for radiation safety is the Radiation Safety Officer for the reactor, who reports on matters of radiation safety to the Radiation Safety Committee (RSC) and to the Reactor Operations Committee ROC)

The Radiation Safety Committee oversees the radiation safety program and is responsible for its implementation.

The review and audit functions are performed by the Reactor Operations Committee (ROC)

The ROC is composed of at least four persons including a manager within Analytical Sciences Laboratory, the facility director, the Radiation Safety Officer, and the Reactor Supervisor.

6.1.2. Responsibility The Facility Director is responsible for reactor facility operations and the policies of the facility.

The day-to-day responsibility for the safe operation of the reactor shall rest 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 approval of the Facility Director and the Reactor Operations Committee, to assume the responsibilities of the Reactor Supervisor.

The Reactor Supervisor reports to the Facility Director regarding reactor operation and within the reactor organization to the Reactor Operations Committee.

The Radiation Safety Officer is assigned the responsibility for implementing the radiation protection program for the reactor facility using the guidelines of ANSI/ANS-15.11-1993 (R2004) [3].

033

Figure 6.1. Administration Ra d ia t io n .. . ... . .. ... ...

Safety ....................................

Officer Line Management Reporting

--

> Communication Reporting o34

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

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

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

c. a licensed Senior Reactor Operator in the facility or readily available on call and able to be at the facility within 30 minutes.

The following operations shall require the presence of the Reactor Supervisor

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 unplanned scrams, and

e. movement of any in-core experiment having an estimated reactivity value greater than $0.75.

f. initial startup and approach to power of the day.

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

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 shall be consistent with ANSI/ANS-1 5.4 sections 4 through 7(2007).

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

.35

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

6.2.1. Charter and Rules

a. This Committee shall consist of the Global R&D Director, who shall be designated the chair of this committee; Facility Director, the Radiation Safety Officer; the Reactor Supervisor; and one or more persons who are competent in the field of reactor operations, radiation science, or reactor/radiation engineering. Members and alternates shall be appointed by and report to the Global R&D Director. Qualified and approved alternates may serve in the absence of regular members.

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 SROs and ROs.

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

d. Minutes of the meetings shall be reviewed and approved at the following meeting and kept as records for the facility.

e. Members of the ROC may be polled by telephone or email for guidance and approvals.

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

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; 036

g. reportable occurrences;

h. proposed changes in equipment, systems, tests, or experiments or procedures with respect to 10 CFR 50.59,

i. audit reports.

6.2.3. Audit Function

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

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

c. 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.

d. A written report of the audit shall be submitted to the ROC within three months of the audit.

e. Deficiencies that affect reactor safety shall be reported to the Global R&D Director immediately.

f. The ROC shall also direct a biennial audit of the facility emergency plan, and the reactor operator requalification program.

6.2.4. Radiation Safety The Radiation Safety Officer is assigned the responsibility for implementing the radiation protection program for the reactor facility using the guidelines of ANSI/ANS-15.11-1993 (R2004).

037

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

a. reactor startup, routine operation, and shutdown;

b. emergency plan and security plan

c. emergency and abnormal operating events, including shutdown;

d. fuel loading or unloading;

e. control rod removal or installation;

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

g. preventive maintenance procedures.

h. operation of each experimental facility.

i. radiation safety procedures Temporary deviations from the procedures may be made by the responsible Senior Reactor Operator or higher individual in order to deal with special or unusual circumstances. Such deviations shall be documented and reported, 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 Operations Committee.

038

6.4. Experiment Review and Approval All new experiments or new class of experiments shall be reviewed and approved by the Reactor Operations Committee, and approved in writing by the Facility Director or designated alternates prior to initiation; Substantive changes to previously approved experiments or previously approved class of experiments shall be made only after review and approval by the Reactor Operations Committee and approved in writing by the Facility Director or designated alternates prior to initiation; Minor changes that do not significantly alter the experiments may be approved by the Reactor Supervisor or higher.

Experiments shall be classified as Routine, Modified Routine or Special.

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

b. Modified Routine Experiments shall have the written approval of the Reactor Supervisor or a designated Assistant Reactor Supervisor or higher. The written approval shall include documentation that the hazards 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 alternate) and the ROC.

Experiments which require the approval of the ROC through sections 6.2.2.a., 6.2.2.b., or 6.2.2.c. of the Technical Specifications are always Special Experiments.

Approved experiments shall be carried out in accordance with 10 CFR 20, TS 3.7, 10 CFR 50.59 and the DTRR operating and administrative procedures.

039

6.5 Required Actions 6.5.1. In case of Safety Limit violation:

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

b. the Safety Limit violation shall be immediately reported to the Facility Director or to a higher level; and

c. the Safety Limit violation shall be reported to the US NRC in accordance with section 6.6.2.; and

d. a report shall be prepared for the ROC describing the applicable circumstances leading to the 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 1.31:

a. reactor conditions shall be returned to normal ifthe condition was caused by known event such as an electrical transient.

or the reactor shall be shut down; ifthe reactor is shut down operation shall not be resumed unless authorized by the Facility Director or designated alternate; and

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

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

040

6.6. Reports 6.6.1. Operating Reports A report shall be submitted annually, starting with the first quarter 1991 performance of annual tasks, to the Radiation Safety Committee and to The Document Control Desk US NRC, Washington, DC, ,which shall include the following:

a) status of the facility staff, licenses, and training; b) a narrative summary of reactor operating experience, including the total megawatt-days of operation; c) tabulation of major changes in the reactor facility and procedures, and tabulation of new tests and experiments that are significantly different from those performed previously and are not described in the Safety Analysis Report, including a summary of the analyses leading to the conclusions that they are allowed without prior authorization by the Nuclear Regulatory Commission 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; ifthe estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, only a statement to this effect is needed); and g) a summary of the radiation exposures received by facility personnel and visitors where such exposures are greater than 25 % of those allowed or recommended in 10 CFR 20.

6.6.2. Special Reports

a. There shall be a report to NRC Headquarters Operations Center, not later than the following working day by telephone and confirmed in writing by facsimile or similar conveyance to The Document Control Desk, 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 occurrence (section 1.31).

b. There shall be a written report presented within 30 days to The Document Control Desk, US NRC, of: permanent changes in the facility staff involving the Levels 1, Levels 2 and levels 3; or significant changes in the transient or accident analysis report as described in the Safety Analysis Report.

e41

c. A written report shall be submitted to The Document Control Desk, US NRC, within 60 days after criticality of the reactor under conditions of a new facility license authorizing an increase in reactor power level, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions.

e 42

6.7 Records 6.7.1. The following records shall be kept for a minimum period of five years:

a. reactor operating logs;

b. irradiation request sheets;

c. checkout sheets;

d. maintenance records;

e. calibration records;

f. records of reportable occurrences;

g. fuel inventories, receipts, and shipments;

h. minutes of ROC meetings;

i. records of audits;

j. facility radiation and contamination surveys; and

k. surveillance activities as required by the Technical Specifications.

6.7.2 Records of the retraining and requalification of Reactor Operators and Senior Reactor Operators shall be retained for at least one complete requalification schedule and be maintained at all times the individual is employed or until the certification is renewed.

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

a. records of gaseous and liquid radioactive effluents released to the environment;

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

c. records of environmental dosimetry

d. drawings of the reactor facility.

e. notification of violations of Safety Limits, LSSS or LCOs

f. approved changes in the operating procedures 043