ML20202A850
| ML20202A850 | |
| Person / Time | |
|---|---|
| Site: | University of Virginia |
| Issue date: | 01/11/1999 |
| From: | VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA |
| To: | |
| Shared Package | |
| ML20202A829 | List: |
| References | |
| NUDOCS 9901280295 | |
| Download: ML20202A850 (74) | |
Text
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APPENDIX A TECHNICAL SPECIFICATIONS :
I l
FOR THE UNIVERSITY OF VIRGINIA REACTOR FACILITY LICENSE No. R-66 DOCKET No. 50-62 As Revised to Facilitate Pennanent Reactor Shutdown, Decontamination and Decommissioning January 11*,1999 9901200295 990120 2 DR ADOCK 0500
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. UVAR Tech. Specs.
L TABI.E OF CONTENTS Ease 1.0. DEFINITION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Figure 1.1 Reactor Facility Boundary Areas . . . . . . . . . . . . . . . . . . 10 2.0. . SAFETY LIMIT AND LIMITING SAFETY SYSTEMS SETTINGS . . . . . . . 11 2.1. Safety Limit . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . .
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11 2.2. Limiting Safety System Settings ~ . . . . . . . . . . . . . . . . . . . . . . . . . . 15
- 3.0. LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . 16 t
3.1. Reactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
16
' 3.2. . Reactor Safety System . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . . . . 18 3.3. Reactor Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 i 3.4. Radioactive Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.5. Confinement ......................................23 3.6. Limitations on Experiments . . . , , . . . . . . . . . . . . . . . . . . . . . . . . 24 3.7. Operation with Fueled Experiments ....................... 26 3.8. Height of Water Above the Core in Natural Convection
- Mode of 0peration . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . 27 3.9. Rod-Drop Times . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.10. Emergency Removal of Decay Heat ....................... 29 3.11. Primary Coolant Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
- 4.0. SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 l
4.1. Shim Rods (Deleted) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
' 4.2. Reactor Safety System (Deleted) . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.3. Emergency Core Spray System (Deleted) . . . . . . . . . . . . . . . . . . . . 31 l 4.4. Area Radiation Monitoring Equipment . . . . . . . . . . . . . . . . . . . . . . 32 g . 4.5. . . Maintenance (Deleted) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
.4.6. Confinement (Deleted) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
'4.7. Airborne Effluents (Deleted) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.8. Primary Coolant Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 u 4.9 . Surveillance of Activity in Secondary System (Deleted) :. . . . . . . . . . 31 5.0. DESIGN FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 5.1. Reactor Fuel Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 ;
5.2. Reactor Building (Deleted) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 )
. ;l.
i' 5.3. Fuel Use and Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 i
!y 6.0, ADMINISTRATIVE CONTROLS . . . .......................39 l
'6.1. Organization . . . . . . . . . .. . .......................39 6.2. Radiation Safety, Reactor Safety w deactor Decommissioning Committees 41 l
. 6.3. Standard Operating Procedures ..........................48 l l 6.4. Review and Approval of Experiments . . . ..................49 l 6.5. Plant Operating Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 6.6. Required Actions ...................................53 l 6.7. Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 i
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UVAR Tech. Specs.
Note: Definitions have been added to Section 1 (in bold and italic print) for use dunny the time the reactor is permanentiv shutdown and subject to or in the process of decommissioning.
1.0. DEFINITIONS Administrative Controls: Administrative controls are those organizational and procedural requirements that are established by the reactor licensee management.
Applicability: As regards use of this term in the Technical Specifications, it is a statement that indicates which components are involved.
Basis: As regards use of this term in the Technical Specifications, it is a statement that provides the background or reason for the choice of specification (s), or references a particular portion of the Safety Analysis Report (SAR) that does.
Beampons: The beamports are the two 8-inch diameter neutron beamports that penetrate the shield on the south side of the UVAR pool.
Channel: A channel is the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter. (Also, see definition for measuring channel .
Channel Calibration: A channel calibration is an adjustment of the channel such that its output corresponds with acceptable range and accuracy to known input values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test.
Channel Check: A channel check is a qualitative verification of acceptable performance by observation of channel behavior, or comparison of the channel with other independent channels or systems measuring the same variable, where this capability exists.
Channel Test: A channel test is the introduction of a signal into a channel to verify that it is operable.
Confinement: Confinement means a closure on the overall facility that controls the movement of air into it and out through a controlled path.
Decommissioning: Decommissioning means to remove a facility or site safelyfmm service and reduce residual mdioactivity to a level that permits: (1) release of the pmperty for unrestncted use and termination of the license; or (2) release of the pmperty under restricted conditions and termination of the license (10CFR50.2). Decommissioning does not include storage or removal offuel, or non-radiological demolition activities.
Decontamination: Decontamination are the activities employed to reduce the levels of mdioactive and/or hazardous contamination in or on material, structures and equipment.
Design Features: The definition for design features is as defined in 10 CFR 50.36.
o UVAR Tech. Specs.
1 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 (lg = 1).
Exoeriment: Any operation, hardware, or target (excluding devices such as detectors, foils, activation samples in an irradiation facility, etc...) that is designed to investigate non-routine reactor characteristics or that is intended for reactor irradiation within the UVAR pool, on or in the beamport or irradiation facility, and that is not rigidly secured to a core or shield structure so as to be a part of their design.
Exoerimental Facility: An experimental facility is a structure or device associated with the reactor that is intended to guide, orient, position, manipulate, or otherwise facilitate a multiplicity of experiments of similar character.
Exoerimental Methods: Experimental Methods are written ano approved instructions which provide guidance to the reactor staff or experimenters for the completion of tasks specified in Experimental Procedures (EPs). While EPs, and changes thereto, are reviewed and approved by the Reactor Safety Committee (RSC), experimental methods are written and reviewed by reactor staff and/or experimenters and approved by a reuor supervisor or administrator. Newly developed experimental methods or changes to existing experimemal methods should be sent to the RSC as information items.
Exoerimental Procedures: Written procedures reviewed and approved by the Reactor Safety Committee which describe the manner in which experiments are run in conjunction with the UVAR, to assure reactor and radiological safety. Operational limits peculiar to the experiment are included in these procedures. Detailed implementation of experimental procedures may be made through the use of experimental methods.
Explosive Material: Explosive material is a solid or liquid that is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N.I. Sax, or is given an Identification of Reactivity (Stability) index 2, 3, or 4 by the National Fire Protection Association in its publication 704-M, " Identification System for Fire Hazards of Materials," also enumerated in the " Handbook for Laboratory Safety" published by the Chemical Rubber Co.
Forced Convection Mode: The reactor is in the Forced Convection Mode when the flow header is up and the primary pump is operating.
Fueled Exoeriment: A fueled experiment is an experiment that contains U-235, U-233 or Pu-239 in levels exceeding trace quantities. Reactor fuel elements are not included in this definition. (Also, see the definition for trace quantities and TS 3.7.).
Imnortant Process Variables: Important process variables are measurable parameters that individually or in combination reflect the basic physical condition of physical barriers. They may include fuel temperature, reactor power, reactor coolant flow rate, reactor coolant inlet or outlet temperature, pool level, or coolant pressure. (Also, see definition for safety limits)
Large Access Facilities: The large access facilities are the two large openings approximately 5 ft wide by 6 ft high that penetrate the shield on the south side of the UVAR pool.
UVAR Tcch. Specs.
Licensed Ooerator: A licensed operator is an individual authorized by the U.S. Nuclear Regulatory Commission to carry out the duties and responsibilities associated with operation of the UVAR. (Also, see definitions for Senior Reactor Operator and Reactor Ooerator).
Limitine conditions for Onerations: Limiting Conditions of Operation (LCOs) are those administratively established constraints on equipment and operational characteristics that shall be adhered to during operation of the facility. The LCOs are the lowest functional capability or performance level required for safe operation of the reactor.
Limiting Safety System Settings: Limiting Safety System Settings (LSSS) are those limiting values for settings of the safety channels by which point protective action must be initiated.
The LSSS are chosen so that automatic protective action will terminate the abnormal situation before a safety limit is reached. The calculation of the LSSS shall include the process uncertainty, the overall measurement uncertainty, and transient phenomena of the process instmmentation. To achieve operational flexibility, it is recommended that actual trip points, where possible, be set more conservatively than specification values.
Measured Value: The measured value of a parameter is the value of the variable as it appears on the output of a measuring channel.
Measurine Channel: A measuring channel is the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.
(Also, see definition for channel).
Methods: Methods are written and approved instructions which provide guidance to the reactor staff, and/or subcontractors working for reactor management, for the completion of tasks specified in Standard Operating Procedures (SOP's). While SOP's, and changes thereto, are reviewed and approved by the UVAR Decommissioning Committee (UDC),
methods are written and reviewed by the reactor staff and/or subcontractors working for reactor management, and approved by the reactor supervisor or reactor director. Newly developed methods, or changes to existing methods, should be sent to the UDC as information items.
Movable Experiment: A movable experiment is one where it is intended that all or part of the experiment may be inserted, removed, or manipulated in or near the core while the reactor is critical.
Natural Convection Mode: The reactor is in the Natural Convection Mode when the flow through the core is maintained by the buoyancy forces associated with the water being heated by the reactor.
Obiective: As regards use of this term in the Technical Specifications, it is a statement that indicates the purpose of the specifications.
On Call: To be on call refers to an individual who (1) hai, oeen specifically designated and the designation is known to the operator on duty, (2) keeps the operator on duty informed of where he may be contacted and the phone number, and (3) is capable of getting to the Reactor Facility within a reasonable time under normal conditions (e.g., approximately 3d min).
1
UVAR Tech. Specs.
Operable: A component or system n operable when it is capable of performing its intended function in a normal manner.
Ooerating: A component or system is operating when it is performing its intended function in a normal manner.
Protective Action: Protective action ir 'he initiation of a signal or the operation of equipment within the reactor safety system in response to a variable or condition of the reactor having reached a specific limit.
(1) channel level. At the protective instrument channel level, protective action is the generation and transmission of a trip signal indicating that a reactor variable has reached a specified limit.
(2) subsystem level. At the protective instrument subsystem level, protective action is the generation and transmission of a trip signal indicating that a specified limit has been reached.
NOTE: Protective auion at this level would lead to the operation of the safety shutdown equipment to immediately shut down the reactor.
(3) instrument system level. At the protective instrument system level, protective action is the generation and transmission of the command signal for the safety shutdown equipment to operate.
(4) safety system level. At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.
Reactor Facility: Rea: tor Facility refers to the immediate site-area surrounding and including the reactor building which houses the University of Virginia Reactor (UVAR).
The site boundary is demarcated by a chain linkfence and gates. (See Figure 1.1)
Egactivity Limits: Reactivity limits for experiments are quantities referenced to an average pool temperature of < 90* F with the effect of xenon poisoning on core reactivity accounted for if greater than or equal to 0.07$. The reactivity worth of samarium in the core will not be included in reactivity limits. The reference core condition will be known as the cold, xenon-free critical condition.
Reactivity Worth of an Exneriment: The reactivity worth of an experiment is the value of the reactivity change that results from the experiment being inserted into or removed from its intended position.
Reactor Ooerating: The reactor is operating whenever it is not secured or shutdown.
Reactor Ooeration: The reactor is in operation when not all of the shim rods are fully inserted and six or more fuel elements are loaded in the grid plate.
UVAR Tech. Specs.
Reactor Operator: An NRC-licensed reactor operator is an individad who is certified by the NRC and the reactor administration to manipulate the controls of the UVAR reactor.
Reactor Safety Systems: Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
Reactor Secured: The reactor is secured when:
(1) Either there is insufficient moderator available in the reactor to attain criticality or there !
is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderation and reflection, or i (2) The following conditions exits:
- a. All shim rods are fully inserted,
- b. The console key is in the OFF position and is removed from the lock, and
- c. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and
- d. No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum reactivity value allowed for a single experiment, or one dollar, whichever is smaller.
Reactor Shutdown: The reactor is shut down if it is subcritical by at least one dollar in the reference core condition with the reactivity worth of all installed experiments included.
Permanent Reactor Shutdown: A reactoris in a permanent shutdown state when all reactorfuel elements have been nmovedfmm the reactor gridplate and an administrative order is in place to prevent a nioading of the core.
Reactor Staff: The Reactor Director and all personnel administratively reporting to him.
Reference Core Condition: The condition of the core when it is at ambient temperature ,
(cold) and the reactivity worth of xenon is negligible (<0.30$). !
Regulating Rod: The regulating rod is a control rod of low reactivity worth fabricated from stainless steel and used primarily to maintain an intended power level. The regulating rod need not have scram capability. The rod may be controlled by the operator with a manual j switch or by the automatic servo-controller.
Reoortable Occurrence: A reportable occurrence is any of the conditions described in Section 6.6.2 of these specifications.
Research Reactor: A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, development, education, training, or experimental purposes, and that may have provisions for the production of radioisotopes.
Safety Limits: Safety Limits are limits on important process variables that are found to be necessary to reasonably protect the integrity of the principal physical barriers that guard against the uncontrolled release of radioactivity. The principal physical barrier is often the fuel cladding. (Also, see the definition for imoortant process variables).
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l UVAR Tech. Specs. l l
Scram Time: Scram time is the elapsed time between the initiation of a scram signal and a l specified movement of a control or safety device.
Secured Exneriment: A secured experiment is an experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces that are normal to the operating environment of the experiment or by forces that can arise as a result of credible malfunctions.
Senior Reactor Onerator: An NRC-licensed senior reactor operator is an individual who is certif!:;d by the NRC and the reactor administration to manipulate the controls of the UVAR reactor and to direct the activities of reactor operators.
Shall. should and mav: The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission, neither a i requirement nor a recommendation. !
Shim Rod: A shim rod is a control rod fabricated from borated stainless steel, which is used to compensate for fuel burnup, temperature, and poison effects. A shim rod is magnetically coupled to its drive unit allowing it to perform the function of a safety rod when the magnet is de-eneigized. (Also, see definition for regulatine rod). )
Shutdown Margin: Shutdown margin is the minimum shutdown reactivity necessary to l provide confidence that the reactor can be made suberitical by means of the control and I safety systems starting from any permissible operating condition and with the regulating rod and the most reactive shim rod in their most reactive position, and that the reactor will remain suberitical without further operator action.
Specification (s): As regards use of this term in the Technical Specifications, it is a statement that provides specific data, conditions, or limitations that bound a system or operation. This statement is the most important statement in the technical specifications agreement. Only the specifications statements are governing.
Standard Operating Procedarn: Written procedures reviewed and approved by the Reactor Safety Conunittee to assure reactor safety and compliance with federal regulations, which describe the manner by which the reactor staff will operate and msintain the UVAR. (Also, see TS 6.3).
Surveillance Reouirements: The definition for surveillance requirements is as defined in 10 CFn 50.36.
Surveillance Time Intervals:
Annually (interval not to exceed 15 months)
Semiannually (interval not to exceed 71/2 months)
Quarterly (interval not to exceed 4 months)
Monthly (interval not to exceed 6 weeks) l Weekly (interval not to exceed 10 days) i Daily (must be done during the calendar day) l UVAR Tech. Specs.
Trace Ouantities: As related to fissionable or fissile nuclides such as U-235, U-233 or Th-232 potentially present in environmental samples on which neutron activation analysis may j be attempted, trace quantities are taken to be negligibly-small concentration levels below 100 ,
parts-per-million (ppm). (Also, see the definition for Fueled Exoeriment).
Tried Exoeriment: A tried experiment is (1) an experiment previously performed in the UVAR or (2) an experiment for which the size, shape, composition, and location does not
}
differ significantly enough from an experiment previously performed in the UVAR to affect .
reactor safety.
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i True Value: The true value is the actual value of a parameter. i t
Unscheduled Shutdown: An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by the actuation of the reactor safety system, operator error, equipment i malfunction, or manual shutdown in response to conditions that could adversely affect safe l operation, not including shutdowns that occur during testing nor check-out operations.
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UVAR Tech. Specs. l l
2.0. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS 2.1. Safety Limits 2.1.1. Safety Limits in Forced Convection Mode of Ooeration Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the forced convection mode of operation. ;
These variables are:
l P = Reactor thermal power W = Reactor coolant flow rate ;
Ti = Reactor coolant inlet temperature L = Height of water above the core l I
Obiective: The objective is to ensure that the integrity of the fuel clad is maintained.
I Specification: In the forced convection mode of operation:
(1) The pool water level shall not be less than 19 ft above the top of the core.
(2) The reactor coolant inlet temperature shall not be greater than Ill*F. l (3) The true value of reactor coolant flow shall not be below 575 gpm.
(4) The combination of true values of reactor core power and reactor coolant flow shall be below the line defined by:
P = 0.24 + (4.5 x 10-3
- W)
P = 0 for W < 575; P in MW, W in gr m The allowed region of operation is shown by the t,nshaded region of Figure 2.1.
Hans Above 575 gpm in the region of full power operation, the criterion used to establish the safety limit was a burnout ratio of 1.49 including the worst variation in the manufacturer's tolerance and specification, hot channel factors and other appropriate uncertainties. The analysis is given in the LEU SAR.
Below 575 gpm buoyancy forces competing with forced convection may lead to flow instabilities in some of the channels and is therefore not allowed. The analysis of the loss of flow transient shows that during the transition from forced convection to natural convection following a loss of flow and reactor scram that the fuel temperature is well below the temperature at which fuel clad damage could occur.
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UVAR Tech. Specs.
2.1.2. Safety Limits in the Natural Convection Mode of Ooeration Apolicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the natural convection mode of operation.
These variables are:
P = Reactor thermal power T i = Reactor coolant inlet temperature Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. l Specification: In the natural convection mode of operation:
(1) The true value of reactor power shall not exceed 750 kW. l l
l (2) The reactor coolant inlet temperature shall not be greater than 111 F. l D_ans. The criterion for establishing a safety limit with natural convection flow is established as a fuel plate temperature. The analysis for natural convection flow shows that at 750 kW, the maximum fuel plate temperature is well below the temperature at which fuel clad damage could occur.
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I UVAR Tech. Specs.
2.1.3. Safety Limit for the Transition from Forced to Natural Convection Mode of Ooeration Apolicability: This specification applies to the condition when the reactor is in transition from forced convection flow to natural convection flow.
Obiective: The objective is to ensure that the integrity of the fuel clad is maintained.
Specification: The current to the control rod magnets must be off when the reactor is making a transition from forced to natural convection.
Basis: The safety analysis of the loss of coolant transient demonstrates that the fuel plate temperature is maintained well below the temperature at which fuel clad damage could occur during the transition from forced downflow through flow reversal to the establishment of natural convection provided that the loss of flow transient is accompanied by a scram.
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UVAR Tech Specs.
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2.2. I imitino Rafety Svetam S**tinom i
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Anoliemhility: These specifications apply to the set points for the safety channels monitoring l
. reactor thermal power, coolant flow rate, reactor coolant inlet temperature, and the height of ;
water above the core. ' l I
Obiective: The objective is to ensure that automatic protective action is initiated to prevent j the safety limit from being exceeded. !
I Specificatiana:
2.2.1. Forced Convertian Mode i i
For operation in the forced convection mode, the limiting safety system settings shall l be Reactor Thermal Power = 3.0 MWt (max) i Reactor Coolant Flow Rate = 900 gpm (min) i Reactor Coolant Inlet Temperature = 108'F (max)
Height of Water above Core = 19' 2" . (min) j Reactor Period = 3.3sec (min) l l
. 2.2.2. Natural Convactian Mode For operation in the natural convection mode, the limiting safety system settings shall .I be:
Reactor Power = 300 kWt (max)
Reactor Coolant Inlet Temperature = 108'F (max)
Reactor Period = 3.3 sec (min)
Bascr The analysis in the LEU SAR shows there is sufficient margin between these settings and the safety limit under the most adverse conditions of operation:
1 (2.2.1.) For the forced convection mode, the LEU SAR considers accidents with reactor power at 3.45 MW, a period of 3 seconds, pool inlet temperature of 111*F and a coolant flow of 837 gpm. The maximum fuel plate temperature calculated was considerably below the aluminum clad melting point. The LSSS specified above for this mode of operation are more conservative than the parameters used in the LEU SAR analysis.
(2.2.2.) With natural convection flow, there is no minimum coolant flow rate and no minimum height of water above the core so long as there is a path for flow (see Section 3.8 of these specifications). The LEU SAR shows that the maximum fuel plate temperature i under natural convection with initial power of 750 kW and pool inlet temperature of i
- lil'F was well below the aluminum clad melting point. The LSSS specified above for I this mode of operation are below the analyzed condition. j ,
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q UVAR Tech. Specs.
L l 3.0. LIMITING CONDITIONS FOR OPERATION 3.1. - Reactivity '
i l
Anchcability: This specification applies to the reactivity condition of the reactor and the i reactivity worth of control rods and experiments.
i l i Obiectives: The objectives are to ensure that the reactor can be shut down at all times and that the safety limit will not be exceeded.
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Specification: ' The reactor shall not be operated at powers in excess of 1 kW unless the ;
following conditions exist:
L (1) The minimum shutdown margin provided by shim rods, with secured experiments (see Section 1.0) in place and referred to the cold, xenon-free condition with the highest- t worth shim rod and the regulating rod fully withdrawn, is greater than 0.55$.
'(2) An experiment with a reactivity worth greater than 0.60$ must be a secured experiment.
(3) The total reactivity worth of the two experiments having the highest reactivity worth is less than 2.00$.
(4) The total reactivity worth of all experiments is less than 2.50$.
? (5) The maximum excess reactivity with fixed experiments in place and referred to cold, xenon-free condition shall be limited to 6.50$.
i Bass: Operation of the reactor at power levels below 1 kW to measure the reactivity worth l of untried experunents, and to measure the excess reactivity of new core loadings, is allowed j with procedums approved by the Reactor Safety Committee. Reactivity is measured in
'~
dollars from the reactor period, and as such is the quantity of safety significance. Reactivity limits expressed in $ are more appropriate for the Technical Specifications, since they are not dependent on the type of fuel used in the reactor or on the geometry of a particular core l loading. !
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- (1) The shutdown margin required by Specification 3.l(1) is necessary so that the reactor can be shut down from any operating condition and remain shut down after cooldown j and xenon decay, even if the highest worth shim rod should stick in the fully withdrawn l
- - position, and with no credit taken for the non-scrammable regulating rod. i l
- (2) The reactivity of 0.60$ in Specification 3.1(2) corresponds to an asymptotic 3-sec j
! period. If this period were sustamed without scramming the reactor until the reactor '
L power reaches the maximum tine value for the Limiting Safety System Setting (LSSS) for the High Power Scram (at which time the reactor scrams on high power), the resulting power overshoot would not exceed the safety limit for power vs. flow.
- (3) The reactivity of 2.00$ in Specification 3.l(3) is less than 2.16$ which corresponds to a !
. 6.9-msec period. Reactor Core DU-12/25 of the SPERT-1 series of tests had MTR plate type elements (
Reference:
Thompson and Beckerly, " Technology of Nuclear o l l
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UVAR Tech. Specs.
l Reactor Safety," Volume I, page 683 (1964)). A 6.9-msec period was nondestructive.
l The simultaneous failure of more than two experiments is considered unlikely. j (4) The total reactivity of 2.50$ in Specification 3.1(4) places a reasonable upper limit on the worth of all experiments. -
l (5) The limit of 6.50$ on excess reactivity is to allow for xenon override and operational '
l flexibility and to ensure that the operational reactor is reasonably similar in
. configuration to the reactor core analyzed in the SAR. In general, the excess reactivity :
! is limited by the shutdown margin requirement.
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UVAR Tech. Specs. I 3.2. Reactor Safety System Apolicability: This specification applies to the reactor safety system channels.
Objective: The objective is to stipulate the minimum number of reactor safety system channels that must be operable to ensure that the safety limit is not exceeded during normal operation.
Specification: The reactor shall not be operated unless the safety system channels described )
in Table 3.1 Safety System Channels are operable.
Bases The startup interlock, which requires a neutron count rate of at least 2 counts per second (CPS) before the reactor is operated, ensures that sufficient neutrons are available for ,
proper operation of the startup channel. j l
The pool-water temperature scram provides protection to ensure that if the limiting safw system setting is exceeded an immediate s'mtdown will occur to keep the fuel temperature below the safety limit. Power level scra as are provided to ensure that the reactor power is
)
maintained within the licensed limits and to protect against abnormally high fuel l temperatures. The manual scram allows the operator to shut down the reactor if an unsafe or abnormal condition arises. The period scram is provided to ensure that the power level i does not increase above that described in the SAR. I 1
Specifications on the pool-water level are included as safety measures in the event of a l serious loss of primary water. Reactor operations are terminated if a major leak occurs in the primary system. The analysis in the SAR shows the consequences resulting from loss of coolant.
The bridge radiation monitor gives warning of a high radiation level in the reactor room from failure of an experiment or from a significant drop in pool-water level.
A scram from loss of primary coolant flow, loss of power to the pump, or application of power to the pump when operating in the natural convection mode, protects the reactor from overheating.
Air pressure to the header above ambient results in a scram to:
- 1) Ensure that the header falls with loss of primary pump power when the reactor is operating in the forced convection mode.
- 2) Prevent raising the header when the reactor is in the natural convection mode.
- 3) Avoid producing additional Ar-41 by activating air introduced into the header.
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I TABLE 3.1 SAFETY SYSTEM CHANNELS i Minimum Operating Mode i Measuring Channel Set Point
Bridge radiation monitor 1 30 mr/hr Scram All modes !
Pool water temperature 1 108'F (max) Scram All modes ;
loss of power Scram Forced convection Power to primary pump 1 application of Natural Scram -
power convection !
Primary coolant flow I 900 gpm (min) Scram Forced convection i
Prevents f i
Startup count rate 1 2 cps (min) withdrawal of Reactor startup ;
any chim rod Manual button 1 Scram All modes i 3 MWt (max) Scram Forced convection Reactor power level 2 Natural 1 0.3 MWt (max) Scram 1 convection i Reactor period 1 3.3 sec (min) Scram All modes Air pressure to header 1 above ambient Scram All modes Values listed are limiting set points. For operational convenience, set points may be changed to more conservative values.
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l UVAR Toch. Specs.
3.3. Reactor Instrumentation Applicability: This specification applies to the instrumentation that must be operable for safe operation of the reactor.
Obiective: The objective is to require that sufficient information is available to the operator to ensure safe operation of the reactor.
Specification: The reactor shall not be operated unless the measuring channels described in Section 3.2 Reactor Safety Systems and in Table 3.2 Measurine Channels are operable.
Bases
- The neutron detectors and the core gamma monitor provide assurance that measurements of the reactor power level are adequately covered at both low and high power ranges.
The radiation monitors provide information to operating personnel of a decrease in pool-water level and of an impending or existing danger from radiation contamination or streaming, allowing ample time to take necessary precautions to initiate safety action.
The reactor room constant air monitor and reactor face monitor provide redundant measures of abnormal high radiation levels. Because the other measuring channels for determining the radiation levels are required for reactor operation, the reactor can be operated safely if these monitors are not functioning for short periods of time.
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I UVAR Tech. Specs. l 1
Table 3.2 Measuring Channels i
l Minimum No. Operating Mode in Which i Measuring Channel !
Operable Required l 1
i Linear power 1 All modes t Intermediate power (log N) and period 1 All modes l
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Core gamma monitor
- 1 Forced convection mode
- l Reactor room constant air monitor
- 1 All modes
- Bridge radiation monitor 1 All modes Reactor face monitor
- 1 All modes
- Pool-water level monitor 2 Forced convection mode Pool-water temperature 1 All modes Primary coolant flow 1 Forced convection mode Startup count rate 1 Reactor startup Reactor power level 2 All modes The reactor room constant air monitor, reactor face monitor, and wre gamma monitor may be out of service for a period not to exceed 7 days without requiring reactor shutdown. If the reactor face monitor cannot be repaired within 7 days, it may be replaced by a locally alarming monitor of similar range for up to 30 days without requiring a reactor shutdown.
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UVAR Tech. Specs. t i 3.4. Radioactive Effluents
' Applicability: This specification applies to the monitoring of radioactive effluents from the ,
Reactor Facility. Airborne and liquid effluents are discussed separately in the following ,
i sectio is.
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- 3.4.1. Airborne Effluents
- Obiective: The objective is to ensure that exposure to the public resulting from the release of Ar-41 and other airborne effluents to the environment will be below the limits of 10 CFR .
20 for unrestricted areas. >
- Specification The activity of gases released beyond the Reactor Facility's site boundary shall not exceed 10 CFR 20 limits. When a neutron beamport vented to the atmosphere is drained of water during reactor operations and until such time as the beamport has been refilled, the effluent shall be monitored by an instrument located in the effluent vent and the effluent vent will have sufficient flow to maintain releases within 10 CFR 20 limits.
Bases A basis for this specification is given by the analysis in the SAR. Compliance with federal regulation is another basis.
3.4.2. Liquid Effluents Obiective: The objective is to ensure that exposure to the public resulting from the release of radioactive effluents will be below the limits of 10 CFR 20 for unrestricted areas. 3 Soecification: The activity of liquids released beyond the Reactor Facility's site boundary i shall not exceed 10 CFR 20 limits.
Basis: The basis for this specification is compliance with federal regulations.
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UVAR Tech. Specs.
l 3.5. Confinement Applicability: This specification applies to the capability of isolating the UVAR's reactor room, when necessary.
Obiective: The objective is to prevent exposure to the public from exceeding the limits of 10 CFR 20 for unrestricted areas, resulting from airborne activity released into the UVAR's reactor room, by providing confinement.
1 Specification: The reactor shall not be operated unless the following equipment is operable.
Eauipment Function Truck door closed switch Scram reactor when truck door is not fully closed Ventilation duct doors Close and seal when Bridge Radiation Monitor alarms Personnel door Close and seal when Bridge Radiation Monitor alarms l
Emergency exit manhole Water level is high enough to form a water water level seal at least 6 in. in depth Basis: The basis for this specification is compliance with federal regulations.
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UVAR Tech. Specs.
. 3.6. I imimian on Emeriman'=
i N Applicability: These specifications apply to experiments installed in the reactor and its 1
,y experimental facilities,
' Obiective: The objective is to prevent damage to the reactor or excessive release of 4 radioactive materials in the event of an experiment failure. !
Specifications:
3.6.1. Pennivity 1
The reactor shall not be operated unless the following conditions exist: !
(1) The reactivity worth of all exj eriments shall be in conformance with specifications in . ;
Section 3.1. i t
(2) Movable experiments must be worth less than 0.13$. !
(3) Experiments worth more than 0.135 must be inserted or removed with the reactor shut j
. down except as noted in Item (4). l (4) Previously tried experiments with measured worth less than 0.50$ may be inserted or j
- removed with the reactor 2.70$ or more suberitical.
t i (5) .If an experiment worth more than 0.50$ is inserted in the reactor, a procedure approved !
by the Reactor Safety Committee shall be followed. !
l 3.6.2. Containers
- (1) All materials to be irradiated in the reactor shall be either corrosion resistant or I encapsulated within corrosion resistant containers.
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i (2) Irradiation containers to be used in the reactor in which a static pressure will exist or in l l which a pressure buildup is predicted shall be designed and tested for'a pressure l
- exceeding the maximum expected by a factor of 2.
, 3.6.3. Danoerous Materiale
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Explosive material shall not be allowed in the reactor unless specifically approved by tbn l Reactor Safety Committee. Experiments reviewed by the Reactor Safety Committee in '
which the material is potentially explosive, either while contained cr if it leaks from the l container, shall be designed to prevent damage to the reactor core or' to the control rods or l
' instrumentation, and to prevent any changes in reactivity. . l
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l UVAR Tech. Specs.
3.6.4. Cooling Cooling shall be provided to prevent the surface temperature of an experiment to be irradiated from exceeding the boiling point of the reactor pool water.
3.6.5. Precautions Experimental apparatus, material, or equipment to be inserted in the reactor, shall not be positioned so as to cause shadowing of the nuclear instrumentation, interference with the control rods, or other perturbations that may interfere with the safe operation of the reactor.
3.6.6. Cobalt Facility The Co-60 pins possessed under the UVAR Operating License when used and stored in the UVAR pool shall be at distances greater than 5 feet from the operating UVAR reactor.
Gamma irradiation facilities utilizing the Co-60 pins shall be designed to prevent physical damage to the Co-60 pins. H' hen the Co-60 pins are in the pool, UVAR pool water samples shall be subjected to gamma spectroscopy for the presence of Co-60 on a monthly frequency, (interval not to exceed six weeks) to assure that substantial leakage of Co-60 from the pins to reactor pool water does not occur.
l Bases: (TS 3.6.1 - 3.6.5) The limitations on experiments specified in TS 3.6.1 through TS !
3.6.5 are based on the irradiation program authorized by Amendment No. 3 to License No.
R-66 dated August 13, 1962. The reactivity ofless than 0.13$ that can be inserted or ,
removed with the reactor in operation is to accommodate experiments in the rabbits. l (Co-60 Facility) When the Co-60 pins are in the UVAR pool they shall be kept a safe distance away from the UVAR reactor when it is operated, to avoid neutron activation and possible failure of the pin cladding, which may result in leakage of Co-60 to the reactor pool water. The Co-60 pins and the gamma irradiation facilities in which they are used will not be used in conjunction with the UVAR.
The monthly reactor pool water sampling frequency, adopted to monitor possible Co-60 leakage from the pins, is the same as that used in the U.S. AEC Safety Evaluation that was performed for thesepins by the Division of Reactor Licensing on August 4,1971. This is a reasonable frequency, for the most likely damage to the pins would be caused by cladding corrosion leading to pin holes. Co-60 leakage under these circumstances would proceed very slowly, into a large pool of water. Therefore, a monthly water sampling and analysis frequency should be adequate to indicate contamination levels before they become significant. UVAR poolwater need not be sampled and analyzedfor Co-60 leakage of all Co-60 pins have been removedfrom the pool.
l UVAR Tech. Specs.
3.7. Oneration with Fueled Exneriments Applicability: This specification applies to the operation of the reactor with a fueled experiment within the reactor building.
Objective: The objective is to ensure that the confinement leak rate and fission product inventory in fueled experiments are within limits used in the safety analysis.
Soecifications:
3.7.1. Fueled Exneriments Generatine Power Above or Ecual to 1 W For fueled experiments in which the thermal power generated is greater than or equal to I watt (W), the reactor shall not be operated unless the following conditions are satisfied:
(1) The experiment must be in the reactor pool and under at least 15 ft of water.
(2) The thermal power (or fission rate) generated in the experiment is not greater than 100 W (3.2 x 102 fissions /sec).
(3) The calculated total energy produced by the experiment shall not exceed 600 W-years.
(4) The leak rate from the reactor room is not greater than 50% of contaimnent of volume in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> as measured within the previous 12 months.
3.7.2. Fueled Exneriments Generating Power Below 1 W Fueled experiments in which the thermal power generated is less than 1 W (3.2 x 105 fissions /sec):
(1) May be located anywhere in the reactor building.
(2) The calculated total energy produced by the experiment shall not exceed 600 W-years. l Bases In the event of the failure of a fueled experiment, with the subsequent release of fission products (100% noble gas,50% iodine,1 % solids), the 2-hour inhalation exposures to iodine and strontium 90 isotopes at the facility exclusion distance,70 meters, are less than the limits set by 10 CFR 20, using an averaging period of 1 year.
, The safety analyses for which results are used here are found in the SAR. The analysis l supporting Specification 3.7.2 assumes 100% exfiltration of fission products from the ,
reactor building in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The analysis supporting Specification 3.7.1 for the fueled !
experiments within the reactor pool assumes a fission product retention in the reactor room equivalent to 100% fission product exfiltration in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The specification provides suitable allowance for degradation between tests. The measurement of the exfiltration value is described in the SAR.
i UVAR Tech. Specs.
3.8. Height of Water Above the Core in Natural Convection Mode of Ooeration Applicability: This specification applies to the height of water above the reactor core when '
the reactor is operating with natural convection cooling. ,
i Ohiective: The objective is to ensure that there is a continuous path for circulation of water l when the reactor is operated in the natural convection mode.
Soecification: The reactor shall not be operated in the natural convection mode unless there is at least I ft of water above the core.
Bases: One foot of water above the core is sufficient to provide a continuous path for i natural convection cooling. For other than zero power operation, the radiation levels may 1 require a greater depth for shielding, in which case the regulations in 10 CFR 20 will l
govern. j
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l UVAR Tech. Specs.
3.9. Rod Drop Times Applicability: This specification applies to the time from the initiation of a scram to the time a rod starts to drop (magnet release time) as well as to the time it takes for a rod to drop from the fully withdrawn to the fully inserted position (free-drop time).
Obiective: The objective is to ensure that the reactr can be shut down within a specified period of time.
1 Specification: The reactor will not be operated unless (1) the magnet release time for each of the three shim rods is less than 50 msec and (2) the free-drop time for each of the three shim rods is less than 700 msec.
Bases Rod drop times as specified will ensure that the safety limit will not be exceeded in a ;
short period transient. The analysis is given in the SAR.
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UVAR Tech. Specs.
3.10, Fmergency Removal of Decay Heat This TS has been deleted because the reactor core has been permanently unloaded. The emergency decay heat removal system is designed to only cool elements located on the gridplate. Asfuel will never be placed on the gri& late again, this TS is no longer needed.
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UVAR Tech. Specs.
3.11. Prirnarv Coolant Canditian Anolicabnhtv: Technical SpecQlication 3.11 applies until all reactorfuel elements and Co-60 pins have been remondfmm the UVAR pool . Following their removal, TS 3.11 is not applicable during the permanent shutdown and decommissioning period. A substitute TS for thisperiodis unnecessary. This specification applies to the quality of the primary coolant in contact with the fuel cladding.
Obiectives: The objectives are (1) to nummize the possibility for corrosion of the cladding on the fuel elements and (2) to minimize neutron activation of dissolved materials.
Specifications:
4 3.11.1. Cond-tivity If reactor fuel elements or cobalt-60 pins are present in the UVAR pool, the conductivity of the pool water shall be no higher than 5 x 104 mhos/cm.
3.11.2. Water nH If reactor fuel elements or cobalt-60 pins are present in the UVAR pool, the water pH of the poolwater shall be between 5.0 and 7.5.
Bang:g: A small rate of corrosion continuously occurs in a water-metal system. To limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.
By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the as low as is reasonably i
achievable (ALARA) F nciple, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposures during maintenance and operations.
Following renomi of allfuel elements and Co-60 pinsfmm the pool, fuel claddmg and Co-60pinjacket corrosion due to impmperpoolwater conktions is no longerpossible.
Also, activation of dissolnd minemis in the poolwater cannot occur if the reactor does not
. operate. Consequendy, prunary water quahty con &tions can be relaxed and need not be specfed in the Technical Specylcations once allfuel element and cobalt pins are removed fmm the pool.
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UVAR Tech. Specs.
4.0. SURVEILLANCE REOUIREMENTS ,
i TS 4.1 thmugh 4.3 and 4.5 through 4.7, and 4.9, have all been deleted. Thejustification for each deletion is given below. TS 4.4 has been modiped with respect to its applicability. l t
4.1 This 15 has been deleted because suneillance requirements on shim md operation, rod dmp times, reactivity measurements and rodphysical condition either are not possible, l necessary or appropriate if the reactor has been permanently and completely unloaded from the core gridplate.
4.2 This TS has been deleted because a reactor safety system only is necessaryfor an operating or operable reactor. Safety system channel tests, checks, calibrations, and a :
core heat balance either are notpossible, necessary or appropriate if the reactor has been permanently and completely unloadedfrom the core gridplate.
4.3 This TS has been deleted because the emergency core spray system does not need to be checked and itspow rate measured if the reactor has been permanently and completely unloadedfrom the core gridplate.
4.4 l The wording as to the applicability of this TS has been changed to recognize that once the reactorfuel has been completely removedfrom the Facility, and the Co-60 pins are no longer stored in the UVAR pool, an area rodsation monitoring system wiH no longer be l needed because it will then be impossible to generate very high radiation levels. i l
4.5 This TS has been deleted because maintenance and suneiRance of reactor control or safety systems either is not pos.sible, necessary or appropriate if the reactor has been permanently and completely unloadedfmm the core gridplate.
4.6 This TS has been deleted because suneiRance of the reactor room closure equipment opembility is not necessary or appmpriate if the nactor has been permanently shut down and completely unloadedfrom the core gridplate. Fueled experiments cannot be run, and thepssion product levels in the fuel arefar below the levels in an operating reactor.
4.7 This TS has been deleted because suneillance of the airborne efpuent monitor of the ventdation ductfrom the groundpoor experimental area is not necessary or appropriate with the reactorpermanently and completely defuelled. No experiments producing airborne e,fluents in association with the reactor can be run.
4.8 The wording to this TS has been modiped, taking into account that it cannot be deleted until TS 3.11 no longer applies.
4.9 This TS has been deleted because secondary system coolant suneiRance is notpossible or needed if the reactor is permanently shut down. The surveiRance relies on the regular production of Na-24 in the primary coolant by an operating nactor. At this time, all Na-24 has decayed away. Also, with the reactor shutdown a leak in the heat exchanger would usult in secondary coolantpow into the primary coolant, and not the other way around as is the case when the reactor is being operated with the primary coolant pump on.
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UVAR Tech. Specs.
1 4.4. Area Radiation Monitoring Eauipment Apolicability: TS 4.4 anolies to the bridge radiation monitor referenced in Table 3.1. This TS will cease to avolv once all UVAR fuel elements have been removed from the Reactor Facility and all Co-60 pins have been taken from the UVAR o001 for shielded storage elsewhere.
Objectives: The objectives are to ensure that the radiation monitoring equipment is l operating and to verify appropriate alarm settings.
Specification-4.4.1 Daily Ooerability Verification l The operation of the radiation monitoring equipment and the position of their associated alarm set points shall be verified daily during periods when the reactor is in operation.
4.4.2. Seminnnnni Calibration j l
The calibmtion of the bridge radsation monitor referenced in Table 3.1 shall be performed semiannually until allfuel elements have been removedfrom the Reactor Facility and all Co-60 pins have been taken from the UVAR poolfor dry shielded storage or appropriate ,
disposal. j Bases: Surveillance of the monitoring equipment will provide assurance that sufficient warning of a potential radiation hazard is available.
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UVAR Tech. Specs.
- 4. 8. Primary Coolant Conddions Anglicabilitv: This specipcation applies to the surveillance ofpnmary water quality whenever eitherfuel elements and/or Co-60 pins are in storage in the UVAR pool.
Objective: The objective is to ensure that water quality does not deteriorate over extended periods of time should the reactor not be operated and eitherfuel elements and/or Co-60 pins be in storage in the UVARpool.
Soecitication: if the conductivity andpH of the primary coolant water is nquired to be maintained asper TS 3.11, then they shall be measured at least once every 2 weeks and veriped to be asfollows:
Conductivity: s 5 x 10~' mhos/cm pH: between 5.0 and 7.5 Bases: Section 3.11 of these specipcations ensures that the water quality is adequate during reactor operation. This section ensures that water quality is adequate whenever eitherfuel elements and/or Co-60 pins are in the UVAR pool and the reactor is not operated.
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y UVAR Tech. Specs.
5.0. DESIGN FEATURES 5.1. Reactor Fuel Specific =*inne
. Applicability: These specifications apply to UVAR low enriched uranium (LEU) fuel.
Obiective: The objective is to describe LEU fuel approved by the U.S. NRC for use in the UVAR.
Specifientiana:
5.1.1. Fuel Material UVAR LEU fuel is of a type described for use at U.S. research reactors by the U.S.
Nuclear Regulatory Commission (NUREG-1313 " Safety Evaluation Report Related to the Evalua^ ion of LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors").
The fuel meat is U 3Si2dispersed in an aluminum matrix and enriched to less than 20% U-235.
5.1.2. Flement Descrintion (1) Plate-type elements of the MTR type are used. The fuel " meat" is clad with aluminum alloy to form flat fuel plates. The active length of the fuel region in the fuel plates is approximately 24 inches and the width is approximately 2.5 inches. The LEU fuel plates are joined at their long-side edges to two side plates. The entire fuel plate assembly is joined at the bottom to a cylindrical nose piece that fits into the UVAR core gridplate. The overall fuel element dimensions are approximately 3 inches by 3 inches by 36 inches. Each fuel plate contains 12.5 grams of U-235.
(2) " Standard" LEU fuel elements are composed of 22 parallel flat fuel plates each, and contain 275 grams of U-235.
(3) " Control-rod" LEU elements are similar to the standard elements, with the exception that they have half as many fuel plates (the 11 center plates being removed to form a channel which is bounded by 0.125 inch thick aluminum plates). Control-rod elements acconunodate the control rods in the central channel. Their U-235 content is 137.5 grams.
(4) " Partial" LEU fuel elements are half-fueled elements composed of 11 LEU fuel plates and 11 unfuelled (dummy) plates. The U-235 content in these elements is 137.5 grams.
(5) "Special" LEU fuel elements have 22 fuel plates, of which 20 are removable. The maximum U-235 content in these elements is 275 grams and the minimum is 25 grams, i
UVAR Tech. Specs.
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5.1.3. Core Confiaurations i
A variety of UVAR core configurations may be used to acconunodate experiments, but the l loadings shall always be such that the minimum shutdown margin and excess reactivity !
specified in the UVAR Technical Specifications are not exceeded.
i Bass: The NRC has described LEU silicide-fuel suitable for use in U.S. research reactors l in NUREG-1313 " Safety Evaluation Report Related to the Evaluation of LEU Silicide j Aluminum-Dispersed Fuel for Use in Non-Power Reactors," [$36.00, from NTIS, !
Springfield Va. (703-487-4650)]. Also, Bretscher and Snelgrove from the Argonne National Laboratory documented LEU fuel test results in ANL/RERTR/TM-14, "The Whole-Core LEU U 3Si2-Al Fuel Demonstration in the 30-MW Oak Ridge Research Reactor." The LEU- }
SAR for the UVAR contains the safety analysis performed for the 22 flat-plate University of i Virginia fuel elements. The LEU elements were designed by EG&G, Idaho, and are i manufactured by the Babcock and Wilcox Company of Lynchburg, Virginia. ;
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UVAR Tecli. Specs. I 5.2. Reactor Building TS 5.2 has been deleted, for the specipcations on conynement, ventilation and reactor room free volume have been required to restrict leakage of radionuclides produced during reactor operation atpower. The UVAR is no longer operated. l 1
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UVAR Tech. Specs.
I 5.3. Fuel Use and Storage
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Applicability: With the exception of Technical Specittr=* inn 5.3.2. TS 5.3 aoplies until all l reactor fuel elements have been removed from the UVAR Farility. Following their !
removal. TS S.3 is not aoplicable during the per===ent sh"*4wn and decommissioning seriod. The specifications below apply to University of Virginia Reactor fuel used and/or l
stored at the University of Virginia Reactor Facility. !
Obiective: He objective is to describe reactor fuel which may be used, possessed and/or stored at the University of Virginia Reactor Facility as well as measures that avoid nuclear criticality or fuel-related accidents, i
Specifications-5.3.1. LEU Possession Lirgit A maximum of 11 kilograms of contained uranium-235 at less than 20% enrichment (which is defined as low emiched uranium, LEU) may be possessed and used at the Reactor Facility.
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5.3.2. Plutonium Possession I imit '
Allplutonium generated orpresent in UVAR LEU reactorfuel, start-up sources, irrudsation targets, puxfoils andpssion chamben may be possessed and used. Following removal of all reactorfuel elementsfrom the Reactor Facili~, only the plutonium present in start-up sourtes, sources, irradsasuon targets, puxfoils andpssion chambers may be possessed and used.
5.3.3. Storage Reactivity I imitatinn All reactor fuel elements, including fueled experiments and fuel devices not in the reactor, shall be stored in a geometric array where calculated km is no greater than 0.9 for all conditions of moderation and reflection using light water, except in esses where an approved fuel shipping container is used, in which case the fuel loading limitations specified in the certificate of compliance for the container shall apply.
5.3.4. Storage Cooling Reauirement !
Irradiated fuel elements and fueled devices shall be stored in an array that will permit sufficient natural convection cooling by water or air so that the fuel element or fueled device surface temperature will not exceed the boiling point of water. !
Bases: Section 5.4 of the American National Standard ANSI /ANS-15.1-1990, "The ;
Development of Technical Specifications for Research Reactors," was used as the overall basis for the above specifications. The limit given in specification 5.3.1 is based on an estimated i reasonable need for reactor fuel for use in the core and a spare fuel requirement determined by ;
DOE's expected spare fuel manufacturing schedule. The specification in 5.3.2 is based on the
. unavoidable production of small amounts of plutonium in reactor fuel, sources, irradiation l
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UVAR Tech. Specs. l I
,. targets, flux foils and fission chambers, as a consequence of normal reactor operation. Precise amounts
- J
' of plutonium produced, decayed or burned during reactor operation is hard to quantify, and this is not i
!- necessary for the small amounts of plutonium produced and contained in these aforementioned devices l l pose no undue reactor or radiation safety risks.
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6.0. ADMINISTRATIVE CONTROL.S 6.1. Organization l
Applicability: The specifications listed below in TS 6.1.1 through TS 6.1.4. apply to the organizational structure of the University of Virginia as it relates to the activities conducted at the Reactor Facility during the permanent shutdown and decommissioning period.
Objective: The objective is to describe the chain of command having responsibility for the safe maintenance, defueling, decontamination and decommissioning of the Reactor Facility.
At the various administration levels, the functions, assignments, responsibilities and associated I professional background, training and requalification requirements are listed, as applicable.
Specifications:
6.1.1. Structure The Reactor Facility shall be an integral part of the University of Virginia. The organizational structure of U.Va., relating to the Reactor Facility is shown in Figure 6.1. The Vice Pnsidentfor Research and Public Senice (Level 1) shall have overall responsibilityfor management of the Facility.
The Reactor Decommissioning Committee Chair shall be responsible for advising the Reactor Director (Level 2) on all matters pertaining to the decommissioning and decontamination of the University of Virginia Reactor Facility. The decommissioning committee members may include reactor staff from level 3, and employees from the Office of Environmental llealth and Safety.
6.1.2. Responsibility During the UVAR pennanent shutdown and decommissioning period, the Reactor Facility Director (Level 2) shall be responsible for overallfacility operation and the dinction of decommissioning activities at the Reactor Facility.
During periods when the Reactor Facility Directoris absent, the Director's responsibilities are automatically delegated to the Reactor Supervisor (Level 3).
The Reactor Facility Director shall have at least a bachelor's degree in science or engineering l and a minimum of 5 years of experience in the nuclear field. A graduate degree may fulfill 4 l years of experience on a one-for-one time basis.
The Reactor Supervisor shall be responsiblefor the day-to-day activities at the UVAR and ensuring that these are conducted in a safe manner and within the limits prescribed by the facility license. During periods when the Reactor Supervisor is absent, his responsibilities are delegated to a person at (level 4).
The Reactor Supervisor shall have the equivalent of a bachelor's degree in science or engineering and at least 2 years of experience in Reactor Operations at this facility, or an equivalent facility, or at least 6 years of experience in Reactor Operations. Equivalent education or experience may be substituted for a degree. Within nine months after being
. ~ . _ _ _ - _. . _ . _ . _ _ _ . . _ . . . _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ . _ . . ,
l UVAR Tech. Specs.
i assigned to the position, the Reactor Supervisor shall obtain and maintain a NRC Senior :
Reactor Opemtor license if nactorfuel elements an still at the Facility. A NRC Senior Reactor Operator license, or a Reactor Opemtor license, is not nquindfor level 3 a.ui 4 personnel once all nactorfuel elements have been shipped offsite.
i 6.1.3, Staffing l
A licensed Senior Reactor Operator shall supenise any movement of reactorfuel. One or i more health physicists, organizationally independent of the Reactor Staff as shown in Figure !
6.1, shall be responsible for radiological safety at the Reactor Facility. ;
6.1.4. Selection and Trainino of Permnnel The selection, training and requalification of Reactor Facility personnel shallfollow the
{
American National Standard for Selection and Training of Personnel for Research Reactors, l ANSllANS-15.4-1988, Sections 4-6, to the extent applicable to the decommissioning status of l thefacility. The selected eniteriafor the personnel will be contained in the NRC approved Opemtor Requalipcation Program, as amended. i i
Bases Sections 6.1,6.1.1,6.1.2,6.1.3 and 6.1.4 of the American National Standard l ANSIIANS 15.1-1990 "The Development of Technical Specifications for Research Reactors,"
describe a generic and generally acceptable organizational structure for U.S. research reactors. They provide the bases for TS 6.1 above. Some of the ANSIstandard ncommendations apply to operable or operating nactorfacilities, and an not necessarily validfor stqff hirsd to pet. brm decommissioning activities.
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UVAR Tech. Specs.
6.2 D-M*= Safety. barnar Safets ==A Reactor Dew ==iccie=i== Ce==inees 6.2.A. EmGedeals/dLQtapaistee Aeolicabdits: The specifications 6.2.A.] and 6.2A.2 apply to the expert group who wiR ,
pmvide oversight over the Reactor Sqfety and Reactor Decommissioning Committees. l l
. Objective: To describe the makeup, nsponsibilities, and authority of the Radiation Safety '
Committee as regards reactorpermanent shutdown and decommissioning. >
Spec (Reations:
I 6.2.A.J. Congpontion and QualgReation i
)
There shaR be a Radiation Sqfety Commstree (RaSC) to ensure that the Reactor Facility is shutdown and decommissioned in a sqfe manner within the terms ofits nactor and other licenses, reactor Technical Specifications and NRC pppmvedplans. The RaSC shall advise the Vice President and Provost and the Director of the Reactor Facility on sqfety and other concerns involving the decommissioning of the Reactor Facility.
The RaSC shaR include its Chairman, the Radiarian Sqfety Officer, the Director of the Reactor Facility, nyresentatives of the hospital administration, Nuclear Medicine, and Radiological Physics or Radiation Oncolcgy. Addutional members may be drawn from such areas as Environmental Health and Sqfety, Radiology, Pathology, Biology, Nursing, Nuclear Engineering, Microbiology, Physics, Obstectrics and Gynecology. Membership of the RaSC wiR change as pppointments are made by the Office of the President of the Univenity. However, the PnA=rian Sqfety Officer and the Reactor Director shaR have stanhng gppointments to the RaSC. CoRectively, the RaSC memben shaR represent a broad spectrum of expenise in the rubological sciences. The membership of the Committee shall be such so as to maintain a high degree of technicalpmfleiency in arwas niating to radiakon sqfety. He RaSC Chairman is the coordmatorfor aR university licenses involving the use of ruboacdve materials and radiation producing equipment.
The Radiadon Sqfety Committee is charged with ensuring that licensed materialis used sqfely and in compliance with NRC nguladons and institutionallicenses. The RaSC reviews changes to the Broad Scope and other licenses. The RaSC also identifies program problems and recommends solutions and nmedial actions. Some ofitsfunctions are carried out thmugh the use of subcommittees, such as the Reactor Sqfety Committee and the Reactor
' Decommissioning Committee. The RaSC wiR carry out mon ofitsfunctions relating to the Reactor Facility through these two subcommittees.
l 6.2.A.2. RaSC Charter and Rules '
(1) To establish a quorum, the ex-officio members and any 5 other Committee members must be pusent.
l (2) The Committee shaR meet as often as necessary to conduct its business by not less than l once in each calendar quarter.
(3) The Committee shall have a written chaner defining such matters as the authority of the Committee, the subjects within its purview, and other administrative provisions.
INAR Tech. Specs.
6.2.B. Reactor Safety Committee Appiscabdits: The specipcations 6.2.B.1 thmugh 6.2B.3 app ly Ic the expert gmup who will pmvide specfic reviews and audits of Reactor Facility opemtions while reactorfuel elements are on-stte.
Obiective: To describe the makeup, responsibilities, and authority of the Reactor Safety Committee.
I Specifications:
6.2.B.J. Comoosition and Oualificatinn l
Then shall be a Reactor Sqfety Committee (Re SC) to review and audit nactor opemtions and ensure that the Reactor Facility is opemted in a sqfe manner within the terms of the reactorlicense. However, reactor sqfety concerns will end once all reactorfuel elements have been permanently shippedfwm the Reactor Facility. At that time the needfor a Re SC shall cease, and any remaining mdation sqfety issues shall be nferred to and be addressed by the Univenity's Radiation Sqfety Commstree. The Technical Specipcation nquirementfor a Reactor Sqfety Committee shaR ceasefollowing the shipment of all reactorfuel elements off-site.
The Reactor Sqfety Committee shall be part (a subcommittee) of the Radiation Safety Committee (RaSC) and report to its Chairman, who is the coordinatorfor alllicenses involving the use of mdsoactive materials and radatson producing equipment at the Univenity of Virginia. The Reactor Sqfety Committee shall be composed of at leastfour memben, and shaR include the Radiarian Sqfety Officer of the Univenity and the Director of the Reactor Facility. The Reactor Director shaR be the sole nactor staff representative on the Committee. CoHectively, the committee memben shaR represent a bmad spectrum of expertise in the nsearth-reactorpeld. The membenkip of the Committee shaR be such so as to maintain a degree of technicalpwpciency in areas niating to reactor sqfety. The memben may be dmwnfmm within or outside the operating organization.
The ReSC shah advise the Vice Presidentfor Researth and Public Service and the Director of the Reactor Facility on reactor sqfety concerns with the operation of thefacility. ReSC reviews and audits are designed to uncover depciencies that affect reactor safety.
6.2.B.2. Charter and Rules I
(1) A quorum of the Committee shall consist of not less than the majority of the full committee. The Chair can designate another member from the Committee to preside in his absence.
(2) The Committee shall meet at least semiannually and shall be on call by the Chair.
Minutes of all meetings shall be disseminated as designated by the Chair.
(3) The Committee shall have a written charter defining such matters as the authority of the
, Committee, the subjects within its purview, and other administrative provisions as are required for effective functioning of the Committee.
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UVAR Tech. Specs.
6.2.R.3. Review and Audit Functions As a minimum the responsibilities of the Reactor Safety Committee include:
(1) Review and approval of untried experiments and tests that are significantly different from those previously used nr tested in the reactor, as determined by the Facility Director.
(2) Review and approval of changes to the reactor core, reactor systems or design features that may affect the safety of the reactor.
(3) Review and approve all proposed amendments to the reactor license, Technical Specifications, and changes to the standard operating procedures (Note: SOPS are discussed in Section 6.3 of these specifications).
(4) Review reportable occurrences and the actions taken to identify and correct the cause of the occurrences.
(5) Review si nificant F operating abnormalities or deviations from normal performance of facility equipment that affect reactor safety.
(6) Review reactor operation and audit the operational records annually for compliance with reactor procedures, Technical Specifications, and license provisions. Audits consist of spot checks of reactor staff compliance with SOP's, Technical Specifications and licenses.
Rases American National Standard ANSl/ANS-15.1-1990, The Development of Technical Specifications for Research Reactors," describes in Section 6.2 acceptable composition and qualification criteria for reactor safety committees and their review and audit functions.
Section 6.3 of the standard describes the organizational relationship of the group responsible for radiation safety to the reactor operations group. These sections of the standard are used as bases for the specifications listed above.
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UVAR Tech. Specs.
6.2. C. Reactor Decommissioning Committee Anchcabdsts: The specifleations 6.2.C.I through 6.2C.3 ppply to the expert gmup who will
- have nsponsibility and ovenightfor decommissioning planning and execution activities at ,
the Reactor Facility.
Obiective: To describe the makeup, nsponsibilities and authority of the Reactor ,
Decommissioning Committee.
Specincations: i 6.2. C.J. Congposition and Quaktication 1
There shall be a Reactor Decommissioning Committee (RDC) to plan the sqfe, legal and timsly decommissioning of the Reactor Facility. Collectively, the decommissioning committee memben shall nynsent a bmad spectrum of expertise in the nseanh-reactor and health-physicspelds, with experience in reactor management, mdiological sqfety, nsearch nactor decommissicning and university administration. Committee members may be dmwnfmm within or outside the Univenity of Virginia, including subcontracted companies. The ,
Committee shall be composed of at leastfour memben, and shallinclude the Radiation Sqfety Officer of the Univenity and the Dinctor of the Reactor Facility.
The Reactor Decommissioning Committee shall be part (subcommittee) of the Radiation Sqfety Committee, which reports to the Vice President and Provost. The Decommissioning Committee shall advise the Reactor Dinctor (Level 2) on all matters impacting the ,
decommissioning of the Reactor Facility.
6.2. C.2. Charter and Rules (1) A quorum of the Decommissioning Committee shall consist of not less than the mqjority of thefull committee. The RDC Chair can designate another memberfmm the Committee to pnside in his absence.
(2) The Reactor Decommissioning Committee shall meet at least quarterly and shall be on call by the Chair. Meeting minutes shall be disseminated as nquired by the RDC Charter.
1 (3) The Reactor Decommissioning Committee shall have a written charter defining such l matten as the authority of the Committee, the subjects within its purview, and other l administrative provisions as are nquiredfor effectivefunctioning of the Committee. \
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l 6.2. C.3. Decommissionine Comminee Functions i
As a minimum, the responsibilities of the Reactor Decommissioning Committeefollowing the termination of the Reactor Safety Committee shallinclude:
l (1) Review and approval ofproposed mqjorphysical changes to the Reactor Facility in
! accordance with 10CFRSO.59(a).
(2) Review and approval ofproposed changes to reactor licenses, Technical Specifications, NRC-approvedplans (such as the Emergency and Security Plans), as well as the UVAR
- Standani Operating Procedures (SOPS), with the exception of changes to the l organizational structure. [The responsibility and authorityfor the organizational structurefor the Reactor Facility resides with the Vice President and Provost.]
l (3) Review unusual and reportable occurrences as well as the actions taken by reactor management to identify and comet the cause of these occurrences.
(4) Annually audit reactor management recordsfor quality and compliance with licenses, Technical Specipcations, NRC regulations and inspections, as well as UVAR SOPS; and to recommend remedial actions to correct indentsped depciencies.
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i PRESIDENT UNIVERSITY OF VIRGINIA I
I I VICE PRESIDENT VICE PRESIDENT FOR RESEARCH & PUBLIC SERVICE AND PROVOST
{ LEVEL 1) l I
DIRECTOR CHAIR ENVIR. HEALTH & SAFETY RADIATION SAFETY COMM.
RADIATION SAFETY OFFICER DIRECTOR RADIATION SAFETY COMM. i
' REACTOR FACILITY 1 (LEVEL 2)
I
! I I HEALTH PHYSICS STAFF REACTOR SUPERVISOR CHAIR CHAIR (LEVEL 3) REACTOR SAFETY COMM. DECOMMISSIONING COMM.
l REACTOR STAFF REACTOR SAFETY COMM. DECOMMISSIONING COMM.
(LEVEL 4) p
- o a
m FIGURE 6.1 "A" ORGANIZATIONAL CHART
]
P UNIV. OF VIRGINIA NUCLEAR REACTOR FACILITY (PRIOR TO SHIPMENT OF ALL FUEL ELEMENTS OFF-SITE) ;
I
PRESIDENT UNIVERSITY OF VIRGINIA VICE PRESIDENT VICE PRESIDENT FOR RESEARCH & PUBLIC SERVICE AND PROVOST (LEVEL 1)
DIRECTOR CHAIR ENVIR. HEALTH & SAFETY RADIATION SAFETY COMM.
- RADIATION SAFETY OFFICER DIRECTOR l l RADIATION SAFETY COMM.
y REACTOR FACILITY -i (LEVEL 2)~
HEALTH PHYS;CS STAFF REACTOR SUPERVISOR CHAIR (LEVEL 3) DECOMMISSIONING COMM. $
E e
DECOMMISSIONING REACTOR m CONTRACTORS DECOMMISSIONING COMM. 3 (LEVEL 4) ,0 FIGURE 6.1 "B" ORGANIZATIONAL CHART UNIV. OF VIRGINIA NUCLEAR REACTOR FACILIT(
(AFTER SHIPMENT OF ALL FUEL ELEMENTS OFF-SITE)
UVAR Tech. Specs.
6.3. Standard Oneratine Pracadures Applicabihrv: The specification below concerns the procedural controls used to operate the University of Virginia Reactor (UVAR) and conduct experiments.
Obiective: The objective is the safe operation of the reactor in compliance with license conditions, federal regulations.
Specifications:
' 6.3.1. Items Covered by SOPS Written procedures, reviewed and approved by the Reactor Safety Committee shall be in effect and followed for the items listed below. These procedures shall be adequate to ensure the safe decommissioning of the reactor, but should not preclude the use of independent judgment and action should the situation require such.
(1) Startup, operation and shutdown of the reactor.
(2) Installation or removal of fuel elements, control rods, experiments, and experimental facilities facilities.
(3) Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system leaks, abnormal reactivity changes.
(4) Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.
(5) Preventative and corrective maintenance operations that could have an effect on reactor safety.
(6) Periodic surveillance, (7) Radiation control.
6.3.2. Chanoes to SOPS Substantive changes to approved procedures shall be made only with the approval of the Reactor Safety Committee (or the Reactor Decommissioning Committee after the ReSC ceases to exist.)
Basis: Section 6.4 of American National Standard ANSI /ANS 15.1-1990, "The Development
' of Technical Specifications for Research Reactors," suggests acceptable procedural controls to applied to opemaing U.S. research reactors.
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UVAR Tech. Specs.
6.4. Review and Approval of Exoeriments Applicability: Specifications 6.4.1 through 6.4.6 listed below apply to classes of experiments run in the UVAR core, in the UVAR pool, or which use UVAR-generated neutron and/or gamma-radiation beams. However, a partial listing of examples of experimental work covered under experiment classes for which broad approval may have been obtained and, therefore, for which individualized experimental procedures would DDI be required follows below:
(a) Samples to be irradiated in approved irradiation facilities, such as the neutron activation facilities, where the samples meet the criteria in TS 3.6 and TS 6.4.
(b) Samples to be irradiated in the neutron radiography facility beamport which are known not to be hazandous to reactor safety.
Obiective: The objective is the safe operation of the reactor and experiments, in accordance with license conditions and federal regulations. Experiments run in conjunction with the reactor should not adversely affect reactor and radiation safety. Notwithstanding the regard for safety, the requirement for review and approval of experiments shall not limit the flexibility of experimenters performing work covered under general written procedures, or for which unanalyzed safety issues do not exist, as determined by the Reactor Director.
Soecifications:
6.4.1. Exoerimental Procedures and Methods (1) Classes of experiments involving the UVAR, the UVAR pool or UVAR radiation beam facilities shall be carried out with established and approved written experimental procedures. The Reactor Safety Committee shall review all new classes of experiments prior to their initiation and approve written experimental procedures governing their operation.
(2) Written experimental methods that implement Reactor Safety Committee approved experimental procedures may be developed by the staff and/or experimenters, as needed.
Such experimental methods shall be approved by a Reactor Supervisor or the Reactor Director prior to use.
(3) The Reactor Director or the Reactor Safety Committee shall decide whether an experimental procedure is required. Usually, an experimental procedure will not be required if the work in question is already covered under an existing approved general experimental procedure or by a Standard Operating Procedure.
6.4.2. Reactivity limits As applicable, reactivity limits for experiments given in experimental procedures shall be based on analyses of maximum reactivity insenions that can be handled by the reactor or its control and safety systems without exceeding safety limits. Reactivity limits have been established in TS 3.6 Limitations on Exoeriments for maximum absolute reactivity worth of individual experiments and the sum of the absolute values of the worth of all experiments.
UVAR Tech. Specs.
6.4.3. Materials As applicable, special requirements shall be established in the e perimental procedures for
! significant amounts of special materials such as fissionable materials, explosives or metastable
! materials capable of significant energy release, or materials that are corrosive to reactor components or highly reactive with coolants. Requirements listed in experimental procedures l may range from detailed analyses to double encapsulation and prototype testing.
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l 6.4.4. Failure and Malfunctions (1) Credible failures of any experiments shall not result in the release or exposures in excess of the annual limits established in Title 10, Code of Federal Regulations, Part 20.
(2) Experiments shall be designed such that they will not contribute to the failure of other experiments, core components, or principal physical barriers to uncontrolled release of radioactivity. Similarly, no reactor transient shall cause an experiment to fail in such a way as to contribute to an accident.
6.4.5. Experimental Facility Soecific LCO Limiting Conditions of Operation limits unique to an experiment shall be specified, as necessary, in the written experimental procedures. Specific surveillance activities which may be required for experiments will also be addressed in the experimental procedures.
6.4.6. Deviations from Exnerimental Procedures (1) Changes to previously approved experiments and experimental procedures, determined by l the Reactor Director to be substantise, shall be made only after review and approval by the Reactor Safety Committee.
(2) Minor changes to experimental procedures may be made with the approval of the Reactor Director, who will determine that no new reactor safety concerns exist, and with the approval of the Reactor Health Physicist, who will assure that radiological safety requirements can be met.
Bases National Standard ANSI /ANS-15.1-1990, "The Development of Technical Specifications for Research Reactors," suggests acceptable provisions governing reactor-based experiments in sections 3.6 and 6.4. These sections served as bases for the above specification. In addition, examples are presented of activities involving the reactor which typically do not require individualized written procedures, because they are covered under a general procedure for an approved class of experiments, or covered by SOPS. It is unreasonable to require procedures with undue specificity when this would limit reasonable experimental flexibility and no unanalyzed safety issues exist. The Reactor Director has the
! resources and authority to determine when experimental procedures are required.
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UVAR Tech. Specs.
l 6.5 Plant Operating Reconis 1
Aeolicabastv: The specfications below apply to UVAR opemting records. ;
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l Obiective: The objective is to maintain and keep on file reactor operating recordsforfuare 1 i reference, andfor demonstration of compliance with license conditions andfederal l l regulations. l l
SpeciReations:
l 6.5.1. Records To Be n-a-t--A for at 1 ,- nye years l
l s i in addution to the requirements of applicable regulations, records of the items listed below l
\
shall be kept m a manner convementfor review. ,
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(1) Normal nactorfacility operation (for example, nactor logbooks, reactor checklists and inndsation nquestforms).
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' (2) Principal reactor systems maintenance records. l (3) ReportaNe occurrences.
1 (4) Equipment and component surveillance activity required by Technical Specyications.
, (5) Reactor Facility mdiation and contamination surveys. l i
\ (6) Experimentsperformed with the UVAR.
t i (7) Fuelinventories, tramifers of radioactive material to andfmm the R-66 license.
t j (8) Appmved changes to operasingpmceduns.
l (9) Records of meetings and audit reports of the Reactor Sqfety Committee.
(10) Records of meetings and audit reports of the Reactor Decommissioning Committee.
6.5.2. Records To Be Retained for One CertiMen*in= Cycle Records of ntraining and requalification oflicensed operators shall be maintained at all times the individualis employed or untillicensing is renewed. l l
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6.5.3. Records To Be Retained for the Life of the Facilits In addstion to the nquinments of applicable regulations, ncords (or logs) of the items listed below shall be kept in a manner convenientfor review and shall be ntained as indicated:
(1) Gaseous and liquid radioactive effluents nicasedfrom the Reactor Facility.
(2) Off-site (radsological) environmental monitoring surveys.
(3) Radiation exposuresfor allpersonnel monitored at the Reactor Facility.
(4) Updated, corncted and as-built drawings of the Reactor Facility.
Bais: American National Standard ANSI /ANS-15.1-1990, "The Development of Technical Specificationsfor Researth Reacton,"provides ncord-keeping guidance in Section 6.8. !
This is the basisfor the above specifications. l r
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6.6. Required Actions i
\ \
Applicabihtv. The specifications below apply to instances where radiologically unsafe ;
l situations have been, or were likely to have been, generated.
l Obiective: The objective is to report unsqfe conditions, study their causes and consequences, i determine their effect on the health and safety of personnel and the public, and take corrective '
- action to prevent recurrence. !
Specifications, j 6.6.1. Action To Be Taken in the Event of a nr:==rf': Occurrence l A reportable occurrence is any of thefollowing condutions: i (1) An observed inadequacy in the implementanon of either administrative orprocedural ;
controis, such that the inadequacy could have caused the existence or development of an l unrpfe condstion at the Reactor Facility. f (2) Abnormal and signficant degradation in reactorfuel, and/or claddmg, coolant :
\
boundary, or containment boundary (excluding minorleaks) wherr Applicable that l could result in exceeding prescribed radmtion-exposure limits ofpersonnel and/or l
I enrimnment. 1 1
(3) Mqfor damage to the Co-60 pins resulting in Co-60 concentrations in reactorpool water l in excess of1 x 10~' micro-curies /ml. l in the event of a reportable occumnce, thefollowing action shall be taken: '
i (a) Ongoing activkies shall cease untilthe occumace has been resolved.
(b) The Dinctor of the Reactor Facility or his designee shall be notyled as soon as possible and cometive action taken asforeseen in the procedures.
(c) A written report of the occumnce shall be made which shallinclude an analysis of the cause of the occumnce, the comctive action taken, and recommendationsfor measures to preclude or reduce the probability of reoccumnce. This report shall be submitted to the Director and the Reactor Sqfety Committee and/or the Radiation Safety Officerfor review.
(d) A report shall be submitted to the Nuclear Regulatory Commission in accordance with Section 6.7 of these specifications.
i' Bass National Standard ANSI /ANS-15.1-1990, "The Development of Technical Specifications l-for Research Reactors," describes in sections 6.6 and 6.7 acceptable specifications for required actions related to safety limits violations, actions to be taken upon their discovery, and reporting requirements. These form the bases for the above specifications.
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UVAR Tech. Specs.
6.7. Reoortine Reouirements Apolicability: The specifications 6.7.1 and 6.7.2 listed below apply to routine and special reports made by the University of Virginia Reactor Facility to the U.S. Nuclear Regulatory Commission.
Objective: The objective is to provide the licensing agency (NRC) with relevant information concerning normal and abnormal reactor operations which are necessary for the fulfillment of
, its mission to protect the public health and safety. A secondary objective is to comply with I
reporting requirements as given in the federal regulations, i
Specifications: In addition to federal regulatory requirements (for example, follow 10 CFR 20, 30.50,40.60, and 70.50, as applicable), reports should be made to the U.S. Nuclear Regulatory Commission as follows:
6.7.1. Reportine of Incidents (1) Immediate notification should be made by telephone, to the U.S. Nuclear Regulatory Commission IIeadquarters Operations Center of:
(a) Personnel total effective doce equivalent of 25 rem er more.
(b) The release of radioactive material, inside or outside of a restricted area, that results, or could result, over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, in personnel intake of 5 times the annual limit on intake specified in 10 CFR 20.
(2) A special report should be made by telephone as soon as possible, but no later than the next working day, to the U.S. Nuclear Regulatory Commission IIeadquarters Operations Center of:
(a) Personnel exposures or releases of radioactive material greater than the limits in 10 CFR 20.
(b) Reportable occurrences as defmed in Section 6.6.2 of these specifications.
(c) Violation of a safety limit.
(3) A special written report should be sent by mail within 14 days to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555 (a) Accidental off-site release of radioactivity above 10 CFR 20 limits, whether or not the release resulted in property damage, personal injury, or exposure.
(b) Reportable occurrence as defined in Section 6.6.2 of these specif6 cations.
l (c) Violation of a safety limit.
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UVAR Tech. Specs.
r (4) A special written report should be sent by mail within 30 days to the U.S. Nuclear ;
Regulatory Commission, Document Control Desk, Washington, D.C. 20555, of: !
(a) Accidental off-site release of radioactivity above 10CFR20 limits, whether or not the i release resulted in property damage, personnel injury, or exposure.
(b) Reportable occurrence as defined in Section 6.6.2 of these specifications.
(c) Changes in personnel serving as Vice Pnsident For Research and Public Service, the Radianon Sqfety Comminee Chair, Reactor Decommissioning Committee Chainnan, Reactor Sqfety Committee Chair, Reactor Facility Director, or Reactor Supervisor.
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(5) A written report should be sent within nine months after initial criticality of the reactor or within 90 days of completion of the startup test programs, whichever is ,
earlier, to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, upon receipt of a new facility license, an amendment to the license authorizing an increase in power level or the installation of a new core with l fuel elements of a design different design than previously used. The report will include the measured values of the operating conditions or characteristics of the l reactor under the new conditions, including:
5 (a) Total control rod reactivity worth. ;
(b) Reactivity worth of the single control rod of highest reactivity worth.
(c) Minimum shutdown margin both at ambient and operating temperatures.
6.7.2. Routine Annual Reports i A routine annual nport will be made by March 31 of each year on decommissioning and niated activiales completed during the pnvious calendaryear. The nport should l be sent to the U.S. Nuclear Regulatory Commission,-Document Control Desk, :
Washington, D.C. 20555, providing thefollowing iqfonnation:
1 (1) Reactor Facility unilimoion, (2) DeserQeion of univenity stqff assigned to decommissioning: numbers, background and nsponssbihties, l (3) TS comphance and nportable events, 1 .(4) Results of NRC inspections and licensing actions, (5) Summary nport on RDC meetings and audit. findings, (6) Health Physics Program (7) Annual waste content and volume sh& ped, 55 - l
, m-,,- . _ . , ,._- -- . , , , , , _.-_.-.m,, , -
UVAR Tech. Specs.
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, (8) Summary of the nature and amount of rodsoactive solid, liquid and airborne ;
effluents released or discharged to the environs beyond the effective control of l
the licensee, as measured or calculated at orprior to the point of such niease ,
or discharge, l (9) Results of environmental surveys and sampling outside the Reactor Facility, (10) Reactor Facility penonnel and visitor radiation exposure summary nport, including the dates and times of significant exposures (greater than 500 mrem for adults and 50 anmforpenons under 18 years of age),
- (11) Summary of rodsation and contamination surveys performed within the ;
Reactor Facility, (12) Status of decommissioningfunding and expenditures, (13) Descrsption of contractor companies operating on-site, (14) Summary of contracted tasks and timelincs, (15) Signficant Changes to the Reactor Facility, (16) Summary oflarge equ& ment transfen, (17) New and modfied SOPS having radiation sqfety sigmficance, (18) Status of emergency preparedness, (19) Figures on industrial accidents or incidents.
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l PROPOSED l l
UNIVERSITY OF VIRGINIA I l
l 8 l DOCKET NO. 50-62 l
AMENDED FACILITY OPERATING LICENSE i
l Amendment No.16 '
License No. R-66 !
I. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by the University of Virginia (the licensee) dated March 9,1977, as supplemented by filings dated December 18,1978, January 19, 1979, September 18,19'i9, July 15,1980, February 12,1981, August 19,1981, March 11,1982, March 19,1982, May 18,1982, June 7,1982 and August 27, l 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set :
forth in 10 CFR ChapterI;
~B . The facility will be managed, but not operated, in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;
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C. There is reasonable assurance (i) that the activities authorized by this amendment l can be conduced without endangering the health and safety of the public, and (ii) l that such activities will be conduced in compliance with the regulations of the ,
Commission; D. The licensee is technically and financially qualified to engage in the activities l authorized by this operating license in accordance with the rules and regulations of the Commission; E. The licensee is a nonprofit educational institution and will use the facility for the conduct of educational activities, and has satisfied the applicable provisions of 10 CFR 140," Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; F. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public and does not involve a significant hazards consideration; G. The issuance of this amendment is in accordance with 10 CFR 51 of the Commission's regulations and all applicable requirements have been satisfied; and
H. The possession and use of the byproduct and special nuclear material as authorized by this license, will be in accordance with the Commission's regulations in 10 CFR 30 and 70.
II. Facility Operating License No. R-66 is hereby amended in its entirety to read as follows:
A. This license applies to the light water-cooled and -moderated swimming pool nuclear reactor owned by the University of Virginia (the licensee), located on the campus of the University of Virginia at Charlottesville, Albemarle County, Virginia and described in the application for license renewal.
B. Subject to the conditions and requirements incorporated herein, the Commission, hereby, licenses the University of Virginia:
(1) Pursuant to Section 104c of the Act and 10 CFR 50," Licensing of Production and Utilization Facilities," only to possess, but not operate, the reactor as a utilization facility at the designated location near Charlottesville, Virginia, in accordance with the procedures and i
limitations described in the application and in this license.
(2) Pursuant to the Act and 10 CFR Part 70," Domestic Licensing of Special Nuclear Material," to possess up to a maximum of 12 kilograms of contained uranium-235 at various enrichments, up to a maximum of 16 grams of plutonium in the form of a sealed plutonium-beryllium neutron source previously used in connection with operation of the reactor, and to possess, but not separate , such special nuclear material as may have been i produced by the operation of the facility prior to its permanent shutdown.
Without exceeding the foregoing maximum possession limits, the maximum limits on specific enrichments of U-235 are as follows:
Maximum U-235 Kilograms % Enrichment Emm 11 < 20% Materials testing reactor (MTR)-type fuel 1 Any Fission chambers, flux foils, and other forms used in connection with operation of the reactor (3) Pursuant to the Act and 10 CFR Pan 30," Rules of General Applicability to Licensing of Byproduct Material" at the Reactor Facility, to possess, store and use 10,000 curies of cobalt 60; to possess and use 1.0 gram of neptunium 237; and to possess, but not separate, such byproduct materials as may have been produced by operation of the reactor prior to its permanent shutdown.".
(4) Pursuant to the Act and 10 CFR Part 70," Domestic Licensing of Special Nuclear Material," to possess, but not use a maximum of 1.0 kilogram of contained uranium-235 at greater than 20-percent enrichment.
C. This license shall be deemed to contain, and be subject to, the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now, or hereinafter in effect, and ;
l is subject to the additional conditions specified below:
(1) Mnimum Power Level l .- The University of Virginia will not load the reactor core and therefore not operate the reactor at any power level.
(2) Technical Specifications l
l The Technical Specifications contained in Appendix A, as revised through Amendment No. 23 are hereby incorporated in the license. The University of Virginia shall operate the facility in accordance with the Technical Specifications.
(3) Physical Security Plan The licensee shall maintain and fully implement all provisions of the Commission's approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of documents withheld from public disclosure pursuant to 10 CFR 2.790, entitled " University of Virginia Nuclear Reactor Facility Physical Security Plan (July 1980)," submitted by letter dated July 15,1980, as revised by letters dated February 26,1981 and July 29,1981.
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l This license is effective as of the date ofissuance and shall expire at midnight twenty years from j the date ofissuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
Cecil O. Thomas, Acting Chief Stand::rdization & Special Projects Branch Division of Licensing )
Enclosure:
Appendix A-Technical l Specifications, September 30,1982 Date ofIssuance: 9/30/82 - l l
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ATT 7.2 - RSC
, ATT 7.2 - Charter for the Radiation Safety Committee CHARTER FOR THE RADIATION SAFETY COMMITTEE
, UNIVERSITY OF VIRGINIA 7.2.1 Charge. The Radiation Safety Committee (RSC) shall:
7.2.1.1 Ensure that licensed material will be used safely. This includes review as necessary of training programs, equipment, facilities, equipment and procedures; 7.2.1.2 Ensure that licensed material is used in compliance with NRC regulations and the institutional license;
- 7.2.1.3 Ensure that the use of licensed material is consistent with the ALARA philosophy and program; 7.2.1.4 Establish a table of investigational levels for individual occupational radiation exposures; and 7.2.1.5 Identify program problems and solutions.
v 7.2.1.5 The RSC may carry out some ofits functions through the use of subcommittees, or through the Radiation Safety Officer Staff.
7,2.1.T Review changes to the Broad Scope license. If practical, all changes will be discussed prior to the license amendment request. If the nature of the amendment request is urgent, then the request will be discussed dusing the next RSC meeting.
7.2.2 Responsibilities. The RSC shall:
7.2.2.1 Be familiar with all pertinent NRC regulations, the license application, the license and amendments; 7.2.2.2 Review the training and experience of the proposed authorized users and the Radiation Safety Officer to determine that their qualifications are sufficient to enable the individuals to perform their duties safely and are in accordance with the regulations and the license; 7.2.2.3 Review on the basis of safety and approve or deny, consistent with the limitations of the regulations, the license, and the ALARA philosophy, all requests for authorization to use radioactive material within the institution; ATT72. SAM
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ATT 7.2 - RSC 7.2.2.4 Prescribe
- Fecial conditions that will be required during a proposed method of use of radioactive material such as requirements for bioassays, physical examinations of users, and special monitoring procedures; 7.2.2.5 Review quarterly the RSO's summary report of the occupational radiation exposure records of all personnel, giving attention to individuals or groups of workers whose occupational exposure appears excessive; 7.2.2.6 Establish a program to ensure that all persons whose duties may require them to work in or frequent areas where radioactive material s are used (e.g., nursing. security, housekeepino. physical plant) are appropriately instructed as required in paragraph 19.12 of 10 CFR 19; 7.2.2.7 Review at least annually the RSO's summary report of the entire radiation safety program to determine that all activities are being conducted safely, in accordance with NRC regulations and the conditions of the license, and consistent with the ALARA program and philosophy.
The review must include an examination of records, reports from the RSO. results of NRC inspections, written safety procedures, and the adequacy of the management control system; v
7.2.2.8 Recommend remedial actions to correct any deficiencies identified in the radiation safety program; 7,2.2.9 Maintain written minutes of all RSC meeting, including members in attendance and members absent, discussions, actions, recommendations, decisions, and numerical results of all votes taken when ballots are cast or requested by any member, and 7.2.2.10 Ensure that the byproduct material license is amended if required prior to any changes in facilities, equipment, policies, procedures. and personnel.
7.2.3 AdministrativeInformation.
7.2.3.1 The RSC shall meet as often as necessary to conduct its business but not less than once in each calendar quarter.
7.2.3.2 Membership of the Radiation Safety Committee The Radiation Safety Committee must include the membership of: its
" chairman, the Radiation Safety Officer. representatives of hospital administration, Nuclear Medicine, Radiological Physics or Radiation ATT72.5AM
- n. ~ upwaras7w.m ununw a w n u s ertrrbgvy vegg e.g -!
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Oncology and representatives from the mayor users of radioactive material.
- The currerd RSC is comprised of representatives of the following disciplines:
' a.' Radiological Physics I
- b. Environmental Health and Safety '
- c. Radiation Oncology i
- d. Radiology ;
- e. Pathology i
- f. siology
- g. Nursing i
- h. Physiology '
- l. Nuclear Engineering -
J. Microbiology
- k. Physics I. Hospital Administration
- m. Nuclear Medecine !
l n. Obstectrics and Gynecology l
- o. Student member l l
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~- Membership of the RSC will change as appointments are made by the
[ Office of the President of the University. However, the RSO and the l . management representative position are exdreio positions and will -
1 L receive standing appointments to the RSC.
7.2.3.3 All members shall have one vote. Votes shall be counted as being cast during an RSC meeting, or in the case of ballots, those retumed. Members not present at a meeting shall not have a vote. Ballots not returned shall not be counted. Votes by acclamation shall be decided by the Chairman and shall not require a count.
7.2.3.4 To establish a quorum, the ex-officio members and any 5 other -
Committee members must be present.
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, 7.2.3.5 Minutes of RSC meetings will be kept for a minimum of 10 years.
7.2.3.8 To the extent that they do not interfere with the mission of the RSC management may assign other responsibilities such as x-ray radiation safety quality assurance oversight, and research project ieview "and approval.
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ATT 7.2. RSC 7.2.3.7 The Radiation Safety Committee will review issues importard to v radiation safety. Such issues may be: new and different experiments involving radioactive material, the approval of new buildings and rooms for the use of radioactive materials, and the uses of new equipment that use or incorporate radioactive material. This safety-related review shall follow the general guidance described in ATT 7.6.
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ATT72. SAM j
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ATT 7.5 - Review of Safety Related items I
REVIEW OF RADIATION SAFETY RELATED ISSUES j RADIATION SAFETY COMMITTEE 7.5.1 Activities that involve the use of radiation or radioactive materials shall be reviewed by the Radiation Safety Committee in the following instances:
- a. When changes are made to the facilities described in the University's NRC license. '
- b. When changes are made to the procedures that are described in the University's NRC license.
- c. When tests or experiments are conducted that are substantially different than those that are normally conducted.
7.6.2 If any of the above occur, the Radiation Safety Committee shall review the changes to ensure that radiation safety related questions are addressed.
The following shall be addressed:
- a. Is the likelihood of an accident, or the consequences of an accidont increased?
- b. Is there a likelihood of a different type of accident, e.g., one not expected?
- c. Are any radiation safety precautions not adequate to protect the worker and public in the event of such an accident?
j 7.5.3 Records of allitems reviewed under this process shall be part of the Radiation Safety Committee meeting minutes and shall be kept by the Radiation Safety Officer. ;
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ATT75. SAM
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January 8,1999 FUNCTIONS AND RESPONSIBILITIES (CHARTER)
OF THE UNIVERSITY OF VIRGINIA REACTOR DECOMMISSIONING COMMITTEE L AUTIIORITY The Reactor Decommissioning Committee (hereafter called the Decommissioning Committee or Committee) is appointed by the Vice President and Provost of the University of Virginia and serves as a subcommittee of the Radiation Safety Committee.
The organizational chart for the University of Virginia Reactor Facility is shown in the attached Figure 6.1 taken from the UVAR Technical Specifications.
II. PURPOSE The purpose of the Decommissioning Committee is to review and audit licensed actisities and decommissioning work conducted at the Reactor Facility by the reactor staff and decommissioning work subcontractors so as to ensure that the UVAR and CAVALIER reactors are shut down, de-fueled, decontaminated and decommissioned in a manner consistent with public safety, federal regulations and within the terms of the reactor and materials licenses.
The general requirements and functions of the Decommissioning Committee are specified in Section 6.2.B of the Technical Specifications for the two reactors. The purpose of this Chaner is to provide guidance concerning the functions and responsibilities of the Committee as required by the reactor licenses (specifically, the Technical Specifications).
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r Page 2 of 5 III. ADMINISTRATIVE PROCEDURES A. Membership i As a minimum, the Reactor Decommissioning Committee is composed of at least four i members and includes the Radiation Safety Officer and the Reactor Director. The -
membership of the Committee shall be such that technical proficiency is maintained in ;
areas relating to radiation safety and decommissioning of nuclear facilities. A representative from among the decommissioning work subcontractors may be requested to l join the committee as an ex-officio member. The Chair of the Decommissioning
,. Committee (DC Chair) shall appoint a Secretary to the Committee charged with making,
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1 distributing and keeping written minutes of all meetings.
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B. Meetings i The Committee shall meet at least quarterly. The DC Chair may call additional meetings i at any time, provided that members are informed of the meeting by written notice, telephone call, e-mail message sent out at least one day prior to the meeting, one day prior to the meeting, or through a schedule established during a prior meeting. Typed minutes 1
of all meetings will be disseminated to the Vice President and Provost, the Vice President for Research and Public Service, all Committee members, and to any other people as requested by the Committee Chairman.
C Action on Recommendations Items requiring Committee approval will be reviewed and voted on at a meeting of a quorum of the Committee. A quorum of the Committee shall consist of not less than a majority of the full committee and shall include the Chairman or his designee ex-officio members do not have a vote. Normally, action to approve items brought before the Committee requires a favorable vote by a majority of the full committee. When a quorum I is not present, items may be approved if members provide prior written or electronic medium approva! of an item to the DC Chair, and a) the number of favorable votes by members present and absent is a majority of the entire committee, b) comments included
M Page 3 of 5 C Action on Recommendations (cont.)
in written approvals are discussed and resolved, and c) changes to the original item requested by the members present are more conservative than the original item.
However, items of a minor nature with respect to the decommissioning activities at the Reactor Facility may be approved by letter votes without a formal meeting of the Committee, provided no member of the Committee requests in writing that a meeting be f
held for the purpose of taking action on the item in question. Additionally, the Chairman may approve items of a minor nature without committee input. Items approved in this manner will be discussed and reaffirmed at the next subsequent meeting of the committee.
IV. RESPONSIBILITIES A. Audits Members of the Committee, or qualified personnel appointed by the Committee, will perform periodic audits of decommissioning activities as necessary to ensure quality and compliance with applicable regulations and license conditions.
B. Review and ApprovalofDecommissioning Plans, SOPS and DWPs The Committee shall review and approve Decommissioning Plans, and new Standard Operating Procedures (SOPS) and Decommissioning Work Procedures (DWP's) prior to first use.
Minor changes to SOPS and DWPs not requiring a change to the facility licenses, not adversely affecting the safety or cost oflicensed and decommissioning activities at the facility, and not violating the intent ofindependent oversight specified by the Committee in j this document or in the Technical Specifications, may be approved by the Reactor I Director without prior Committee approval. All such approvals shall be documented and subsequently reviewed by the Committee. 1 i
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l C Review and Approval of Changes to the Reactor Facility The Committee shall review and approve physical changes to the Reactor Facility during the reactor decommissioning period. Minor changes not involving unreviewed safety !
questions may be authorized by the Reactor Director. Such approvals shall be do::umented and subsequently reviewed by the Committee.
D. Review and Approval ofL)ocuments Requiring NRCApproval The Committee shall review and approve documents requiring NRC approval prior to their remittal to the NRC. The following documents, and changes thereto, require NRC approval:
- 1. Reactor Licenses and associated Facility Licenses
- 2. Technical Specifications
- 4. Security Plan
- 5. Changes to the Facility involving unreviewed safety questions (10CFR50.59) l
- 6. Decommissioning Plan j l
E. Review of Unusual and Reportable Occur?ences The Committee shall review unusual occurrences and, if warranted, recommend corrective actions. Unusual occurrences include the following:
- 1. Reportable occurrences as identified in the Technical Specifications l 2. Unexpected release of radioactive materials to the environment or releases in excess ofNRC limits
- 3. Significant abnormalities or deviations from normal performance of Facility equipment that may affect radiation or personnel safety 4, Possible items of noacompliance with license requirements as identified by NRC inspectors or other audits and proposed corrective actions.
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1 F. Record-Keeping and Reporting to the NRC l
The Committee shall verify that the Reactor StafTproperly collects and maintains I complete records of decommissioning activities. The Committee shall also I i
ascertain that annual or other necessary reports are prepared and remitted by the l staff to the NRC as per license and regulatory requirements.
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PRESIDENT UNIVERSITY OF VIRGINIA l
I I VICE PRESIDENT VICE PRESIDENT FOR RESEARCH & PUBLIC SERVICE AND PROVOST (LEVEL 1) l DIRECTOR CHAIR ENVIR. HEALTH & SAFETY ' "
RADIATION SAFETY COMM.
RADIATION SAFETY OFFICER DIRECTOR , RADIATION SAFETY COMM.
REACTOR FACILITY 1 (LEVEL 2)
, I
! I I HEALTH PHYSICS STAFF REACTOR SUPERVISOR CHAIR CHAIR (LEVEL 3) REACTOR SAFETY COMM. DECOMMISSIONING COMM.
REACTOR STAFF REACTOR SAFETY COMM.' DECOMMISSIONING COMM.
(LEVEL 4)
FIGURE 6.1 "A" ORGANIZATIONAL CHART UNIV. OF VIRGINIA NUCLEAR REACTOR FACILITY (PRIOR TO SHIPMENT OF ALL FUEL ELEMENTS OFF-SITE)
PRESIDENT UNIVERSITY OF VIRGINIA VICE PRESIDENT VICE PRESIDENT FOR RESEARCH & PUBLIC SERVICE AND PROVOST (LEVEL 1)
I DIRECTOR CHAIR ENVIR. HEALTH & SAFETY RADIAT!ON SAFETY COMM.
RADIATION SAFETY OFFICER DIRECTOR RADIATION SAFETY COMM.
REACTOR FACILITY H (LEVEL 2)
HEALTH PHYSICS STAFF REACTOR SUPERVISOR CHAIR (LEVEL 3) DECOMMISSIONING COMM.
DECOMMISSIONING REACTOR CONTRACTORS DECOMMISSIONING COMM.
I (LEVEL 4)
FIGURE 6.1 "B" ORGANIZATIONAL CHART UNIV. OF VIRGINlA NUCLEAR REACTOR FACILITY (AFTER SHIPMENT OF ALL FUEL ELEMENTS OFF-SITE)
_ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ __ _ - ._ _ - . _ _ _ _ - _ _ -- ____ _ __ _ _____ _ ___ - -___ _ _ _________ _ _