Text

Dow Chemical Company - *Redacted* Response to NRC Request for Additional Information License Renewal Clarifications to March 9, 2011 Telephone Conference
	ML113460120
Person / Time
Site:	Dow Chemical Company
Issue date:	04/20/2011
From:	Krahenbuhl M; Dow Chemical Co
To:	Geoffrey Wertz; Division of Policy and Rulemaking
	Wertz G, NRR/DPR/PRLB, 301-415-0893
References
	TAC ME1595
	Download: ML113460120 (61)
	v • d • e

DOW CHEMICAL COMPANY TRIGA RESEARCH REACTOR LICENSE NO. R-108 DOCKET NO. 50-264 TECHNICAL RAI RESPONSES (DATED 4/20/2011)

REDACTED VERSION*

SECURITY-RELATED INFORMATION REMOVED

REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS

-440>0-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:

Dow Chemical Company-Response to the Request for Additional Information for the renewed license of the TRIGA research reactor. License No. R-108; Docket No. 50-264 Enclosed is the response to phone conference March 9, 2011. This response includes the original question followed by the modified or new response.

Should you have any questions or need additional information, please contact the undersigned at 989-638-6932.

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

Executed on April 20, 2011 Melinda Krahenbuhl, Ph.D.

Director Dow TRIGA Research Reactor Subscribed and sworn to before me this C) C) day of April, 2011 Notary I C MyCounty, Michigan My Commission Expires:

Stacy L. McKeon "tG*

MiDL4BY COUNTY, MICMIG ACTIN IN M IDAND COUNTY, MICHI1 UYCOMMISSION EXPiRES JUNE 28, 2011 cc: Sandeep S. Dhingra, R&D Director - Analytical Sciences AQýoýD

Technical Clarifications Requested during the Phone Meeting March. 9, 2011

1.

Remove the reference to heat exchangers to avoid confusion.

1.

NUREG-1537, Part 1, Section 1.2, "Summary and Conclusions on Principal Safety Considerations" requests the applicant to provide a summary of the principal safety.

considerations for its reactor. The DTRR safety analysis report (DTRR SAR) includes Section A.2 with the same title, but the contents only briefly discuss operational safety based on "TRIGA reactor fuel, current instrumentation, and operation controls." Please provide a summary of the principal safety considerations applicable to DTRR (control rods and their active/passive operation, the reactor tank structure, site selection, ventilation systems, etc.).

DTRR response:

The Safety Analysis report demonstrates and documents the systems and properties that provide for the continued safe operation of the DTRR.

The features include passively safe control rods that utilize electromagnets for withdraw, UZrH mix fuel with a negative temperature coefficient, natural convective heat transfer for cooling, adequate and well trained staff.

The maximum hypothetical accident does not lead to public exposure above the regulatory limits. Additionally, the DTRR has established emergency and security plans integrated with Michigan operations site. Audits and inspections performed regularly have not identified on-going deficiencies. The equipment, surveillances and instrumentation meet or exceed requirements established in the Code of Federal Regulations, ANSI and ANS standards and guidelines.

The reactor components, tank, fuel, control rods do not show signs of wear and continue to perform in the intended function.

6. Use Regulatory Guide 1.145 for accident scenario's instead of 1.111 which is for normal releases.

6.

NUREG-1537, Part 1, Section 2.3, "Meteorology' requests the applicant to indicate how the local (site) meteorology supports the dispersion calculations of airborne releases under normal and accident conditions. Please provide a description of the dispersion model based on this:

meteorological data.

DTRR response:

The off-site dose calculations are completed following the guidance in Regulatory Guide 1.145, (Nuclear Regulatory Commission, 1977) with the additional conservative assumptions that the wind is blowing towards the fenceline under very low dispersion atmospheric conditions at the time of the accident.

11. Include that we visually inspected the support grid and tank for corrosion and wear.

11.

NUREG-1537, Part 1, Section 4.2.5, "Core Support Structure" requests the applicant to provide design information pertaining to the core support structure. DTRR SAR, Chapter D, does not provide sufficient information. Please provide figures depicting the upper and lower core plates and provide the dimensions and locations of all penetrations that allow coolant to flow through them.

DTRR response:

Two aluminum grid plates fix the position of fuel elements, dummy elements and neutron source. Figure 4 is the schematic drawing of the upper grid plate with the position of the control rods, pneumatic transfer location and source indicated. The upper plate is 33/44 inch aluminum and 1.5 inch diameter holes to position the fuel, dummy elements, control rods, etc. The bottom plate is 33/44 inch aluminum and has holes to receive the end pins of the fuel and dummy elements. Thirty-six holes for the natural convection cooling, are found on the lower grid plate. The water passes through the upper grid plate by means of the gap between the triflute section of the fuel and the upper grid plate. The penetrations on the lower grid plate are in concentric rings with 7, 12, and 17 holes, respectively. Figure 5 is a photograph of the lower grid 1

plate. This photograph substantiates the as built drawings given by General Atomics, and previous Safety Analysis Reports submitted by The Dow Chemical Company.

Figure 5 is a photograph of the lower grid plate. The support grid plate and tank were inspected; there is no visible evidence of corrosion of wear.

F4-Rabbit Terminus

-S-AmBe Source D-Dumny element

AL-Aluminum Fuel element Safe-Safety control rod Shim-Shim Control rod

Reg-Regulating Control Rod' Figure 4. Upper grid plate 2

Bottom of the tank Ak I F Top of the core Figure 5. Lower grid plate.

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31. Scram Logic diagram needs to be sent again T86 I

K2 JK3 K4 100 K Magnet Power Supply --

1 K

= Hi+nd 1I z 'T Comm 10 Reg Shim Safe Lo Gnd Mag Mag Mag

TB5, Mag Vohtage T

Monitor Ground Fault Detector EXT NP-1000 SCRAM No1 DAC

%PWR2 HV2 Watch o--i i--o Relay (open) 6Dog2 c--J'-o Relay (closed)

TB6 Relays shown open (tripped) configuration Figure 8. SCRAM Logic for the DTRR "Operation and Maintenance manual microprocessor based instrumentation and Control for the ICI TRACERCO TRIGA reactor" General Atomics

33. Include that the smoke detectors are on a battery backup

33. NUREG-1537, Part 1, Section 8.1, "Normal Electrical Power" requests the applicant to provide schematic diagrams showing the basic distribution systems and circuits of the normal electrical power distribution system. DTRR SAR, Chapter H, does not provide sufficient information. Please provide schematic diagrams showing the basic distribution systems and circuits of the normal electrical power distribution system.

DTRR response:

The normal electrical distribution system for building 1602 is included as Figure 10. Panels which feed rooms 51,52 are M15 - LP "D" and LP "B". The 1 megawatt chiller dedicated to the reactor is circuit M10.

The smoke detectors are hardwired into the building and are not on a battery backup.

35. Tie the Foushee Document to our facility better.

35. NUREG-1537, Part 1, Section 9.2, "Handling and Storage of Reactor Fuel" requests the applicant to provide an analysis or a reference to an analysis that shows that subcriticality is assured under all conditions of fuel handling and storage. The DTRR SAR does not provide this information. Please provide an analysis that shows that K eff is maintained below 0.90 for all storage configurations and that dose limits are met.

4

DTRR response:

The DTRR does not currently store fuel outside of the core. For fuel inspection and control rod inspection the core is made subcritical by removing two fuel elements from the B ring and placing them in the in-pool storage racks. Subcriticality of the core is verified prior to fuel and control rod inspection. The in-pool storage racks have 10 positions with 2 inches separating the axial centerline of adjacent elements. This storage rack configuration is identical to the Keff evaluations described by F. Foushee (1966). As per the letter dated 3/1/1966 by Foushee, the Keff for one rack containing 10 storage positions with a 2 inches separating the axial' centerline is 0.5096. The Keff was calculated assuming the U-235 contained in the each element is Since the standard TRIGA fuel stored in possession has a U-235 load less than and the storage racks were designed and installed as described in the Foushee letter, the in-core storage for the DTRR is less than 0.90. The dose limits are met by storing the fuel in the pool with a required water level 15 feet above the core.

42.

Take out the (a)(6)

42. NUREG 1537, Section 10.3 "Experimental Review" states that the documentation of experiment review methodology should describe how 10 CFR Section 50.59 will be used to review all new experiments and changes to currently authorized experiments not described in a reactor's SAR. DTRR SAR, Section J.3.3.2 does not provide this information. Please provide a discussion regarding experiment review methodology.

DTRR response:

The DTRR is used for Neutron Activation. Experiments will continue to be reviewed as described in the Safety Analysis report section J.3 and Technical Specifications 6.2.2, using the Technical Specifications listed in TS 3.7. New experiments fixed or moveable will be reviewed using 10 CFR 50.59.

47. Elaborate on Jim's positions as EH&S and Radiation Safety Organization names and designations are not stagnant. Currently, The Radiation Safety officer position is administered as depicted in the following organizational chart.

Business Environmental Industrial Radiation Services Health and Hygiene Safety Safety Expertise Program Operations Center

49. Take out the (4)

49. NUREG-1537, Part 1, Section 12.1.3, "Staffing" requests the applicant to provide a description showing how the staffing requirements of 10 CFR Section 50.54 are met. DTRR SAR, Section L.1.2, states that the staffing meets all requirements of 10 CFR Section 50.54 but does not show how this is accomplished. Please show how the staffing requirements of 10 CFR Section 50.54 are met.

DTRR response:

As stated in L.1.2 the DTRR is "staffed with a minimum of two licensed senior reactor operators. These two licensed reactor operators are used to meet 10 CFR 50.54 (i) (i-1), (j), (k),(I), and (m)(1). A licensed operator shall be present to manipulate control rods or any operation which may affect the reactivity or power level of the reactor. A licensed operator shall be present at all times during the operation of the reactor. A licensed operator shall be on-call for the activities described in 10 CFR 50.54 (m)(1).

Therefore two operators met the requirements in 10 CFR 50.54.

50. Include the targeted subcriticality

50.

NUREG-1537, Part 1, Section 12.3, "Procedures" requests the applicant to provide an overview of procedures used at their reactor. DTRR SAR, Section L.3.12, "Control Rod Inspection and Removal," describes the rod drop time test but provides no information regarding control rod inspection and removal. Please describe the procedure used for control rod removal 5

and inspection, including the applicable console settings and how shutdown margin is maintained.

DTRR response:

Procedure 4.4.1 details the control rod removal and inspection procedure. This inspection is done after the annual fuel inspection is complete. As part of the Annual fuel inspection, two fuel elements are removed from the B ring to become subcritical.

Subcriticality is confirmed by measuring core excess with these elements removed. The core excess is measured as 0, and a power reading of less than 20 milliwatts.

The console remains operational to monitor power and water temperature. The control rods are disconnected from the bridge. Drive mechanisms are disconnected. One at a time, the control rods are withdrawn from the core and maneuvered for inspection and replaced to their original location. The drives are reattached and powered.

Full range of movement, and scram functions are tested prior to any operations. Each inspection is followed by Control Rod calibration.

56. Include analysis of the 10 microCi iodine fueled experiment.

DTRR response When a response to question 56 is submitted an analysis of the 10 microCi iodine fueled experiment will be included 61.3. Add surveillance for the tank water level alarm 61.3 ANSI/ANS-15.1-2007, Section 3.3, "Coolant Systems" recommends an LCO for monitoring pool leaks, loss-of-coolant, and isolation valve positions which were not found in the DTRR TS 3.4, Coolant System.

DTRR response A specification has been added requiring a pool water level alarm. An annual surveillance requirement has been added to TS 4.3 62.2. Add that we do surveillance prior to and after maintenance on the safety systems 62.2 DTRR TS 4.0, "Surveillance Requirements" does not specify which surveillances are required prior to, or following maintenance, inspection, and fuel movement activities.

DTRR response 62.2.

TS 4.0 has been changed to include a surveillance prior to and after maintenance on any of the safety systems. TS 4.1 requires at least annual surveillances of the reactivity worth of each control rod, reactor core excess, reactor shutdown margin. These surveillances are also required after core fuel movement.

62.5. add verified operable in daily checklist 62.5 DTRR TS 3.5, "Confinement" has no corresponding SR.

DTRR response 62.5. TS 3.5 requires that the ventilation systems shall be operational when the reactor is operated, fuel is manipulated, or radioactive materials with the potential of airborne releases are handled in the reactor room. A surveillance requirement for the confinement system has been add as TS 4.7 62.6. Add in section 4 62.6 DTRR TS 3.8, "Experiments" has no corresponding SR.

DTRR response A surveillance requirement for the experiments has been added as TS 4.8 64.1. Don't use terminology "exclusion area" use either restricted or controlled make sure site definition is consistent throughout EP and TS.

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64.

ANSI/ANS-15.1-2007, Section 5.0, "Design Features" provides information, identified below, regarding content and format that was not found in the DTRR TS. Please provide additional information for each of the following:

64.1 ANSI/ANS-15.1-2007, Section 5.1, "Site and Facility Description" recommends a description of the site and the facility expressly identifying the extent of the reactor license coverage.

DTRR response:

64.1. A definition of licensed area has been added to the TS definitions. TS 5.1 has been modified to remove the word exclusion. The Emergency Plan does not contain the word exclusion.

65.2. Add more description regarding Facility Directors position

65.

ANSI/ANS-15.1-2007, Section 6, "Administrative Controls" provides recommendations regarding content and format. DTRR TSs differences from these recommendations were noted. Please provide additional information for the following:

65.2 ANSI/ANS-15.1-2007, Section 6.1.2, "Responsibility" describes responsibilities for the operation and safeguarding of the public which was not fully described in DTRR TS 6.1.2.

Please describe the Facility Director's responsibilities and clarify what is meant by "management sense."

DTRR response 65.2. This has been removed from TS 6.1.2. The definition of Facility Director has been modified.

65.8. Add reference to 10 CFR 50.59 65.8 ANSI/ANS-15.1-2007, Section 6.5, "Experiment Review and Approval" recommends that experiments be carried out in accordance with approved procedures. DTRR TS 6.4 does not describe the process of experiment review and approval for new experiments.

DTRR response The following restriction has been added to TS 6.4. No experiment or class of experiments shall be performed without review and approval by the Reactor Operations Committee. Experiments are approved and classified by the ROC as Routine, Modified Routine or Special.

Experiments shall be reviewed with respect to 10 CFR part 20, TS 3.7 and 10 CFR 50.59

67. Include that we have visually inspected grid plate, tank, control rods and instruments and have found that they are able to continue in the specified function for the requested license period.

67. NUREG-1537, Part 1, Section 16.1, "Prior Use of Reactor Components" requests the applicant to provide information on prior use of items significant to safety, such as fuel cladding, reactivity control system, engineered safety features, and radiation monitoring systems. This means evaluating the continued serviceability of originally supplied components (e.g., for aging and wear); and also to consider the suitability of items supplied by other facilities. DTRR SAR Chapter P.1 does not provide sufficient detail regarding the prior use and continued use of items significant to safety.

DTRR response:

The DTRR equipment and instruments show no sign of failure. Equipment and instrumentation are calibrated, maintained, repaired and replaced as warranted. The tank is an open pool; therefore the grid.

support and inner tank are visible. Closer inspection with a camera did not locate evidence of corrosion or wear on the tank or structures located in the tank. The tank and structures were deemed serviceable for the license period requested.

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Requested changes to TS 4.7 Strike out surveillance and add to TS4.7 DTRR the proposed TS has been changed 4.0 add required surveillance before and after DTRR the proposed TS has been changed 4.3 add the pool water level alarm here DTRR the proposed TS has been changed 4.4 add changed and evaluate DTRR the proposed TS has been changed Figure 6.1 asked that we return to figure in amendment #7. Includes the ROC and RSC relationship.

DTRR the proposed TS has been changed 6.1.4 Add "2007" DTRR the proposed TS has been changed 6.2.3 Check for consistency with the requal plan The requalification plan does not require an audit therefore the TS and requalification plan are consistent.

6.4 reviewed and approved by ROC DTRR the proposed TS has been changed 6.6 c take out (a)

DTRR the proposed TS has been changed 6.7.3 Include safety limits DTRR the proposed TS has been changed 8

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.

0

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 which controls the movement of air into and out of the facility through a controlled path.

1.7 Core Configuration - Core configuration shall be an assembly of standard NRC-approved low enriched stainless-steel-clad or aluminum-clad TRIGA fuel elements in light water. The fuel shall be arranged in a close-packed array for operation at full licensed power except for (1) replacement of single individual fuel elements with in-core irradiation facilities or control rod guide tubes and (2) the start-up neutron source. The aluminum-clad fuel element shall be placed in the E or F ring of the core.

1.8 Damaged Fuel-Damaged TRIGA fuel is defined for stainless steel-clad UZrH1.65 as the sagitta exceeding 0.0625 in (0.159 cm) over the length of the cladding, elongation exceeding 0.125 in. (0.318). For aluminum clad fuel UZrH1.o sagitta exceeding 0.125 in. (0.318) and elongation exceeding 0.5 in (1.27 cm). Fuel is also defined as damaged when burn-up of U-235 exceeds 50% of the initial concentration 1.9.

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 (ke,=l) at reference core conditions 1.10.

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.

1.11.

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

1.12.

Facility Director - Person responsible for reactor facility operation, including implementation of security, emergency plans and operational procedures, communication within Dow and external to Dow regarding the operation of the reactor facility 1.13 Licensed area - Rooms 51 (51, 51 A, 51 AA 51 B) and 52 of building 1602 1.14.

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

1.15.

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

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

1.24.

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

1.25.

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

1.26.

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

02

1.27.

Reactor Secured - The reactor is secured whenever:

a) it contains insufficient fissile material present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or b) the console switch is in the "off" position, the key is removed from the switch, and the key is in the control of a licensed reactor operator or stored in a locked storage area; and sufficient control rods are inserted to assure that the reactor is subcritical by a margin greater than $1.00 at reference core conditions no work is in progress involving core fuel, core structure, installed control rods or control rod drives unless those drives are physically disconnected from the control rods; and no experiments in or near the core are being moved or serviced that have, on movement, a reactivity worth exceeding $0.75.

1.28.

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.

1.29.

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

1.30.

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.

1.31.

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).

1.32.

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.

1.33.

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.

03

d)

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

Reactor trips resulting from a known cause are excluded.

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 1 OCFR20.

1.34.

Rod, Control - A control rod is a device containing neutron absorbing material which is used to control the nuclear fission chain reaction. The control rods are coupled to the control rod drive systems in a way that allows the control rods to perform a safety function.

1.35.

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. Experiments and classes of experiments which are to be considered as routine experiments shall be so defined by the Reactor Operations Committee.

1.36.

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.

1.37.

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

1.38.

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.

1.39.

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.

1.40.

Shutdown Margin - 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.

1.41.

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

1.42.

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 04

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.

1.43 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 A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the heating of air, fission product gases, and hydrogen from the dissociation of the fuel-moderator. The magnitude of this pressure is determined by the temperature of the fuel element and by the hydrogen content. Data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of ZrH1.6 will remain below the ultimate stress provided that the fuel temperature does not exceed 1050 C and the fuel cladding temperature does not exceed 500 C. When the cladding temperature can equal the fuel temperature the fuel temperature design limit is 950 C (M. T. Simnad, G.A.

Project No. 4314, Report e-1 17-833, 1980).

Experience with operation of TRIGA-fueled reactors at power levels up to 1500 kW shows no damage to the fuel due to thermally-induced pressures.

The thermal characteristics of aluminum-clad TRIGA fuel elements using ZrH1.1 moderator have been analyzed (S. C. Hawley and R. L. Kathren, NUREG/CR-2387, PNL-4028, Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors, 1982). A loss-of-coolant analysis showed that in a typical graphite-reflected Mark I TRIGA reactor fueled with 60 aluminum-clad fuel elements (Reed College) the maximum fuel temperature would be less than 150 C following infinite operation at 250 kilowatts terminated by the instantaneous loss of water. These temperatures are well below the region where the a+ d + g' to a + d phase change occurs in ZrH1.1 (560 C).

.6

2.2. Limitina Safety System Settincis (LSSS)

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

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

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

Basis The LSSS which does not exceed 300 kW provides a considerable safety margin. One TRIGA reactor (General Atomics, Torrey Pines) showed a maximum fuel temperature of 350 C during operation at 1500 kilowatts, while a 250-kilowatt TRIGA reactor (Reed College) showed a maximum fuel temperature of less than 150 C (reported by S. C. Hawley, R. L. Kathren, NUREG/CR-2387, PNL-4028 (1982), Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors). A portion of the safety margin could be used to account for variations of flux level (and thus the power density) at various parts of the core. The safety margin should be ample to compensate for other uncertainties, including power transients during otherwise steady-state operation, and should be adequate to protect aluminum-clad fuel elements from cladding failure due to temperature and pressure effects.

7

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 The reactor shall be shutdown 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.

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

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

08

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

The fuel shall be arranged in a close-packed array for operation at full licensed power except for (1) replacement of single individual fuel elements with in-core irradiation facilities or control rod guide tubes and (2) the start-up neutron source.

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

Bases Operation with standard NRC-approved TRIGA fuel in the standard configuration ensures a conservative 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.

9

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 minimum acceptable standards during operation of the reactor.

Specifications There shall be a minimum of one scram-capable analog safety channel.

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

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

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

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

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

Bypassing of the channels and interlocks in table 3.3 is permitted for checks, calibrations, maintenance or measurement with written approval from the Reactor Operations Committee.

010

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

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

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

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

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

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

011

TABLE 3.3A.

MINIMUM REACTOR SAFETY CIRCUITS, INTERLOCKS, AND SET POINTS Scram Channels Scram Channel Minimum Operable Scram Setpoint Reactor Power Level NM1 000 & NPP1 000 2

NPP1000 Detector High-Voltage Power Supply NM1 000 Detector High-Voltage Power Supply 1

1 Not to exceed maximum licensed power Failure of the detector high-voltage power supply Failure of the detector high-voltage power supply Not applicable Not applicable Manual Scram Watchdog (DAC to CSC)

Interlocks Interlock/Channel Startup Countrate Rod Drive Control Reactor Period Function Prevent control rod withdrawal when the neutron count rate is less than 2 cps Prevent simultaneous manual withdrawal of two control elements by the control rod drive motors Prevent control rod withdrawal when the period is less than 3 seconds 912

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

Interlocks Interlock/Channel Startup Countrate Rod Drive Control Reactor Period Bases Provides assurance that the signal in the NM1 000 channel is adequate to allow reliable indication of the state of the neutron chain reaction.

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

13

TABLE 3.3B MEASURING CHANNELS Measuring channel NM1000 NPP1000 Water Radioactivity Monitor Minimum Number Operable 1

1 1

Water Temperature Monitor TABLE 3.3B BASES FOR MEASURING CHANNELS Measuring Channel Basis NM1 000 NPP1000 Water Radioactivity Monitor Water Temperature Monitor Provides assurance that the reactor power level can be adequately monitored.

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

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

Provides assurance that the water temperature can be adequately monitored

14

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

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

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

The pool water pH shall be in the range of 4 to 7.5.

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

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

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

There shall be an audible alarm on the coolant level.

15

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

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

Monitoring the radioactivity in the pool water serves to provide early detection of possible cladding failures. Limitation of the radioactivity according to this specification decreases the exposure of personnel to ionizing radiation during operation and maintenance in accordance with the ALARA program.

Maintaining the specified depth of water in the pool provides shielding of the radioactive core which reduces the exposure of personnel to ionizing radiation in accordance with the ALARA program.

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

916

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 operable and the external door (Door 10) shall be closed whenever the reactor is operated, fuel is manipulated, or radioactive materials with the potential of airborne releases are handled in the reactor room. The ventilation system is operable when the fan is on.

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.

17

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 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.

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-sensitive ion chamber is utilized as a temporary substitute, provided that the substitute can be monitored by the reactor operator.

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

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

.18

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.

.19

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

20

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 e 21

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 need 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.

. 22

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.

9 23

4.2. Reactor 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 NM1 000 and NPP1 000 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 NM1000 channel provides 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.

24

4.3. Coolant System 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.

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 by evaporation.

3.

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

. 25

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

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 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.

26

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

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

Specification Each fuel element shall be 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.

The reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal. A TRIGA fuel element shall be considered damaged and removed from the core if:

a)

The transverse bend exceeds 0.125 inch over the length of the cladding.

b)

The length exceeds the original length by 0.125 inch.

c)

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

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.

e 27

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

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

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

Basis Experience has shown that experiments and operational requirements, in many cases, may be satisfied with a variety of combinations of facility options, power levels, time delays, and effluent or staff radiation exposures. The ALARA (As Low As Reasonably Achievable) principle shall be a part of overall reactor operation and detailed experiment planning.

e 28

4.7 Confinement Applicability This specification applies to the surveillance of the confinement system Objective 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 operable 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.

29

4.8 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.

9 30

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.

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

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

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

All air or other gas exhausted from the reactor room and from associated experimental facilities during reactor operation shall be released to the environment at a minimum of 8 feet above ground level.

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

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

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

31

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

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

32

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 The fuel shall be standard NRC-approved TRIGA fuel.

The control elements shall have scram capability and shall contain borated graphite, boron carbide powder, or boron and its components in solid form as a poison in an aluminum or stainless steel cladding.

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

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

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

The control elements perform their function through the 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.

033

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

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

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

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

034

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. This structure cuts across the lines of management of The Dow Chemical Company. 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 and to the Reactor Operations Committee. 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 which is composed of at least four persons including a manager within Analytical Sciences Laboratory, the Radiation Safety Officer, and the Reactor Supervisor.

6.1.2.

Responsibility The Facility Director is responsible for reactor facility operation. 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 notification 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.

0 35

Figure 6.1. Administration Environmental, HAearth& SafetyR Analyti(

Operations

<Oaiton Safety

>ee CHAIR RSC CmiteL

~ Level 1\\4----

I r

II

~~

Fa~cili Supervisor, Reactor Operations Industrial, Committee

~Hygqienje

>Reactc I---------

Radiation Safety < -------

Licen

Officer, Line Management Reporting Communication Reporting irSu

.36

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 Senior Reactor Operator:

a. manipulations of fuel in the core;

b. manual removal of control rods;

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

d. recovery from unexplained scrams, and

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

f. initial startup and approach to power.

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.

9 37

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 members of the day-to-day reactor operating staff.

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

d. Minutes of the meetings shall be kept as records for the facility.

e. 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.

038

6.2.2. Review Functions The ROC shall review and approve:

a. every experiment involving fissionable material;

b. experiments or operations which would require a change of core configuration, or a change in the equipment or apparatus associated with the reactor core or its irradiation facilities, or a new piece of apparatus being mounted in the reactor well; except that movement of the neutron source for the purpose of routinely checking the instrumentation, or the movement of the neutron detectors to establish the proper calibration of the associated channels shall not require review by the ROC;

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

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

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

f. operating abnormalities having safety significance;

g. reportable occurrences;

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

i. audit reports.

. 39

6.2.3.

Audit Function

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

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

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

At least one of the auditors shall be familiar with reactor operations. No person directly responsible for any portion of the operation of the facility shall audit that operation.

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

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

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

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

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

.40

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

a. reactor startup, routine operation, and shutdown;

b. emergency and abnormal operating events, including shutdown;

c. fuel loading or unloading;

d. control rod removal or installation;

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

f. preventive maintenance procedures.

g. operation of each experimental facility Temporary deviations from the procedures may be made by the responsible Senior Reactor Operator or higher individual in order to deal with special or unusual circumstances. Such deviations shall be documented and reported immediately to the Reactor Operations Committee.

41

6.4. Experiment Review and Approval Approved experiments shall be carried out in accordance with 10 CFR part 20, TS 3.7 and 10 CFR 50.59.

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 Facility Director (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.

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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 if the condition was caused by known event such as an electrical transient.

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

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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; if the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, only a statement to this effect is needed); and 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.

044

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 1.29).

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 reactor supervisor or the facility director; or significant changes in the transient or accident analysis report as described in the Safety Analysis Report.

c. A written report shall be submitted to The 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.

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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. notice of violations, LSSS,LCOS

f. approved changes in the operating procedures

g. required safety limits

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Technical Specification Index ALARA 1 surveillance 29 Audit function 38 Calibration channel 1 thermal power 24 Channel 1 calibration 1 area monitor 27 continuous air monitor 27 thermal power 24 check 1 test 1 area monitor 27 continuous air monitor 27 scrams and interlocks, surveillance 24 Check channel 1 Confinement 1 Control rod inspection 24 Control rods composition 32 Core Configuration 9 Excess Reactivity 1, 8 Experiment 1 modified routine 2, 39 movable 2 routine 4, 39 secured 4 special 4, 39 specifications 19 Experimental Facilities 1 Fuel element aluminum 9 surveillance 28 temperature design limit 6 TRIGA 5, 9 Fuel storage 33 Irradiation facility incore 9 LCO 1 Limiting Conditions for Operation 1, 8 Limiting Safety System Setting 1, 3, 6, 7 LSSS 1 Measured Value 2 Minimum shutdown margin 8 Modified Routine Experiments 2, 39 Movable Experiment 2 Occurrence, reportable 3 Operable 2 Operating 2 Procedures 38 Radiation monitoring system 0 47

specifications 18 Radiation Safety Committee 2 Radiation Safety Officer 37 Reactivity excess 1, 8 limits 2, 8 maximum positive insertion rate 10 worth of an experiment 2 Reactor control and safety systems 10 operating 2 research 3 safety circuits 2 safety systems 3 secured 2 shutdown 3 Reactor Operations Committee 3, 37 audit function 38 charter and rules 37 Reactor Supervisor 37 Records 41 Reference Core Condition 3 Reflector 32 Report operating 40 Reportable Occurrence 3 Reports special 41 Research Reactor 3 Review and Audit 37 Review function 37 ROC 3 Rod, Control 4 Routine Experiment 4 Routine Experiments 39 RSC 2 Safety related instrumentation 10 Safety Limit 4, 6 Scram setting 7 time 4 Secured Experiment 4 Shutdown Margin 4 minimum 8 Special Experiment 4 Special Experiments 39

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Specification 36 administrative controls audit function 38 experiment review and approval 39 operating report 40 organization 34 procedures 38 records 41 required actions 39 review and audit 37 review functions 37 special reports 41 staffing 36 structure 35 airborne radioactivity from experiments 20 aluminum-clad fuel element 9 analog safety channel 10 arrangement of fuel 9 confinement ventilation system 17 corrosive, reactive, explosive, liquid fissionable materials 19 design features control element composition 32 core cooling 31 fuel storage 33 pool to boundary 30 reactor room 30 reflector 32 release of gases 30 TRIGA fuel 32 experiment failure and inspection 20 experiments in the B and C rings 20 explosives, testing 19 fuel 9 fuel temperature (safety limit) 6 fueled experiments and radioactive inventory 20 limiting safety system setting 7 manual scram 12 maximum control rod drop time 10 maximum core excess reactivity 8 maximum positive reactivity for experiments 19 maximum positive reactivity insertion rate 10 measuring water radioactivity 14 measuring channel linear power 14 percent power 14 water temperature 14 wide-range Log and period 14 minimum shutdown margin 8 number of control rods 10 operability of measuring channels 10 operability of safety channels and interlocks 10 operation approval 19 pool water bulk temperature 15 0 49

conductivity 15 depth over the core 15 pH 15 radioactivity 15 radiation monitoring system area monitor 18 continuous air monitor 18 Reactor Operations Committee 37 rod drive interlock 12 scram setpoint power level 12 reactor period 12 securing experiments 19 startup countrate interlock 12 surveillance ALARA 29 area monitor calibration 27 area monitor channel test 27 Continuous Air Monitor calibration 27 continuous air monitor channel test 27 control rod drive withdrawal speeds 24 control rod drop times 24 control rod inspection 24 control rod worth 23 fuel element examination 28 interlocks 24 log power channel 24 operation with damaged fuel 28 pool water conductivity 26 pool water level 26 pool water pH 26 pool water radioactivity 26 pool water temperature 26 reactor coolant system 26 reactor core excess 23 scram channels 24 shutdown margin 23 thermal power calibration 24 surveillance intervals 22 watchdog 12 water 9 wide-range linear channel high voltage 12 wide-range log channel high voltage 12 Staffing minimum 36 Surveillance 22 Test channel 1 TRIGA Fuel Element 5, 32 damage specifications 28 Ventilation system specification 17 9 50

v t e Letter Sequence Response to RAI
TAC:ME1595, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement, Supplement Administration Questions 17 June 2010 12 July 2010 1 September 2010 16 September 2010 2 November 2010 27 October 2010 21 March 2011 14 July 2011 11 July 2011 11 July 2011 6 September 2011 2 November 2011 26 June 2013 6 September 2013 6 March 2014 Answers 12 May 2011 10 August 2011 31 August 2011 12 October 2011 10 November 2011 20 April 2011 20 January 2012 20 January 2012 Results Approval, Approval, Approval Other: ML110270002, ML11318A281, ML113460038, ML120180131, ML120730278, ML120800032, ML120800036, ML12137A151, ML12137A171, ML13022A010, ML13094A006, ML13113A001
MONTHYEAR2011-07-14 [Table View]

ML113460120

Text

Subject:

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