ML20041F467

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Tech Specs Revision,Superseding & Clarifying Previous Submittals Submitted During License Renewal Review
ML20041F467
Person / Time
Site: University of Virginia
Issue date: 03/31/1982
From:
VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA
To:
Shared Package
ML20041F460 List:
References
NUDOCS 8203160528
Download: ML20041F467 (48)
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i APPENDIX A FACILITY LICENSE NO. R-66 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF VIRGINIA DOCKET NO. 50-62 DATE:

MARCH, 1982 8203160528 820311 DR ADOCK 05000062 p

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. s' TABLE OF CONTENTS-Pagg 1.0 DEFINITIONS

-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 5

2.1 Safety Limits 5

2.2 Limiting Safety System Settings -

8 3.0 LIMITING CONDITIONS FOR OPERATION 9

3.1 Reactivity 9

3.2 Reactor Safety System 11 3.3 Reactor Instrumentation 13 3.4 Radioactive Effluents 15 3.5 Confinement 16 3.6 Limitations on Experiments 17_

3.7 Ocaration with Fueled Experiments 19 i

3.8 licight of Water Above the Core in Natural 21 Convection bbde of Operation

3. 9 Rod Drop Times 22 3.10 Emergency Removal of Decay Heat 23 4.0 SURVEILLANCE REQUIREMENTS 24 4.1 Safety Rods 24 4.2 Reactor Safety System 25 4.3 Emergency Core Spray System 26 4.4 Radiation Monitoring Equipment 27 4.5 Maintenance 28 4.6 Con finement 29 4.7 Airborne Effluents 30 4.8 Reactor Fuci Dose Measurements 30a J

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TABLE OF' CONTENTS ~

(Continued)

Page 5.0 DESIGN FEATURES 31 5.1 Reactor Fuel 31 5.2 Reactor. Building 32

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5.3 Fuel Storage 33

,I 6.0 ADMINISTRATIVE CONTROLS 34

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6.1 Organization 34 6.2 Review 37 6.3 Operating Procedures 39 e

d 6.4 Action to be Taken in the Event a Safety Limit 40 is Exceeded 6.5_ Action to be Taken in the Event of an Reportable 41 Occurrence 6.6 Plant Operating Records 42 6.7 Reporting Requirements 43 LIST OF FIGURES Figure

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2.1 Safety Limits with Forced Convection Flow 6

6.1 Organizational Structure of the University 36

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of Virginia Relating to the Reactor Facility f

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1.0 DEFINITIONS v

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The terms Safety Limit (S?[], Limiting Safety System Setting (LSSS),

Limiting Condition of Operati'on (LCO), surveillance requirement), and design features are as defined in.10 CFR Part 50.36.

1.1 Reactor Shutdown - The reactor is in a shutdown condition when all shim rods are fully inseyted.

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a. Ali shim rods are fully inserted; P-

,<j b','The console key is in the OFF position and is removed from gf the lock;c e

c. No work is/In progress in-core involving fuel or experiments or maintenance of the core structure, control rods or control

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1.3 True Value - The true value of a process variable is its actual value at any instant.

1.4 Measured Value - The measured value of' the process variabic is

's the value of the variabic as it appears on. the output of a s.

s measuring channel.

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for the purpose of measuring the value of a process variable.

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nation of measuring channels and associated circuitry which forms

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the automatic protective system of the reactor, g

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1.7 Operable - A component or system is cperabic when it is capable

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i 1.8 Operating - A component or system is operating when it is per-forming its intended function in a normal manner.

1.9 Channel Check - A chainel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification should include comparison of the channel with other independent channels or methods of measuring the same variable, where this capability exists.

1.10 Channel Test - A channel test is the introduction of a signal into a channel to verify that it is operable.

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1.11 Channel Calibration - A channel calibration is an adjustment of the channel s.uch that its output responds, with. acceptable range and accuracy of the parameter which th'e', channel measures.

Calibration shall encompass the entire channel, includ-ing equipment actuation, alarm, or trip.

1.12 Reportabic Occurrence - A reportable occurrence is any of the j/

following:

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a. Any safety system setting less conservative than specified in the Limiting Safety System Settings section of these Technical Specifications.
b. Operating in violation of a Limiting Condition for _ Operation established in these Technical Specifications unless prompt remedial action is taken,
  1. 5 -
c. Safety system component malfunctions or other component or

, j system malfunctions during reactor operation which could, or threaten to, render the safety system incapable of performing its intended safety function, unless immediate shutdown of the reactor is, initiated.

c.

d. An uncontrolled or unanticipated increase in reactivity in excess of 0.005 Ak/k.
e. An observed inadequacy in the implementation of either admin-istrative or procedural controls, such that the inadequacy could have caused the existence or development of an unsafe condition in connection with the operation of the reactor.
f. Abnormal and dignificant degradation in reactor fuel, and/or cladding, coolant boundary, ~or containment boundary (excluding minor leaks) where applicable which could result in exceeding prescribed - radiation-exposure limits of personnel and/or environment.

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t 1.13 Experim int - An experiment is:

a. Any apparatus, device or mateiial placed in the reactor core region, in an experimental facility associated with the reactor, or in-line with a beam of radiation emanating from the reactor; or
b. Any in-core operation designed to measure reactor character-istics.

1.14 Tried Experiment - A tried experiment is:

a. An experiment previously performed in this reactor; or
b. An experiment for which the size, shape, compcsition and location

.does not differ significantly enough from an experiment previously performed in this reactor to affect reactor safety.

1.15 Beam Ports - The beam ports are the two 8-inch neutron beam ports which penetrate the shield on the south side of the pool.

1.16 Large Access Facilities - The large access facilities are the two large op:aings approximately 5-feet wide by 6-feet high which penetrate the shield on the south side of the pool.

1.17 Fuel,ed Experiment - A fueled experiment is any experiment which contains uranium 235, uranium 233 or plutonium 239.

This'does not include the normal reactor core fuel elements.

1.18 Secured Experiaent - A secured experiment is an experiment which fits into the reactor grid plate at the bottom and is held down on top by a hold down rod supported from the reactor bridge.

The hold down rod must be strength equivalent to or greater than a schedule 40, 1 1/2 inch diameter aluminum pipe.

1.19 Unsecured Experiment - Any experiment, experimental facility, or component of an experiment is deemed to be unsecured when it is not secured as defined in 1.18 above.

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1.21 Experimental Facility - An experimental f acility is any structure or device associated with the reactor which is intended to guide, orient, position, manipulate or otherwise facilitate a multi-plicity of experiments of similar character.

1.22 Reactor Operation - Reactor operation is when all of the shim rods are not fully inserted and six or more fuel elements are loaded in the grid plate.

1.23 Shim Rod - A shim rod is a control rod fabricated from borated stainless steel which is used to compensate for fuel burnup, tem-perature, and poison effects. A shim rod is magnetically coupled to its drive t.,it allowing it to perform the function of a safety rod when the magnet is de-energized.

1.24 Regulating Rod - The regulating rod is a control rod of low reactiv-ity worth fabricated from stainless steel and used to control reactor power. ' The rod may be controlled by the operator with a manual switch or by an automatic controller.

1.25 Reactivity Limits - Quantities are referenced to an average pool ~

temperature of (<90 F) with the effect of xenon poisoning on ~

_co_re.. activity accounted for if greater than or equal to 0.05%

Ak/k. 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.

1.26 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous l

Explosion llazard in " Dangerous Properties of Industrial Materials by N. I. Sax, or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its publication 704-M, "Identifi-cation System for Fire !!azards of Materials," also enumerated in the "llandbook for Laboratory Safety" published by the Chemical Rubber Co.

o s-1 2.0 SAFETY LD1ITS AND LIMITING SAFETY SYSTEit SETTINGS 2.1 Safety Limits 2.1.1 Safety Limits in Forced Convection Mode of Operation Applicability This specification applies to the interrelated variables assoc-lated with core thermal and hydraulic performance in the forced con-vection flow mode of operation. These variables are:

P = Reactor thermal power M = Reactor coolant flow rate Ty = Reactor coolant inlet temperature L = Height of water above the core Objective To assure that the integrity of the fuel clad is maintained.

Specificktion In die forced convection flow mode of operation,

a. The pool water icvel shall not be less than 19 feet above top of the core.
b. The reactor coolant inlet temperature shall not be greater than g

111 F.

c. The combination of true values of P and W shall be in the unshaded -

portion of Figure 2.1.

Bases Above 400 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 variations in the manufacturer's tolerances and specification, hot channel factors, and other appropriate uncertainties. The analysis is given in Section 9.4 of the SAR.

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In the region below 400 GPM where the flow coasts down to zero for selecting a safety limit is taken as a fuel plate The analysis of a loss of flow transient from 3.45 megawatts and 744 GPM of flow, shows that the maximum fuel plate temperat e

reached is 303 F which is well below the temperature at which fue damage could occar, T!te analysis is given in Section 9.7 of the SAR.

2.1.2 Safety Limits in Natural Convection Mode of Operation Applicability with core thermal and hydraulic performance i ae flow mode of operation. These variables are P = Reactor Therm.11 Power Ty = Reactor Coolant Inlet Temperature Objective To assure that the integrity of the fuel clad is maintained Specification In the natural convection flow mode of operation of P and T shall not exceed:

, the true valua y

P 750 kWt T

111 F g

Bases flow is established as a fuel plate temperature.The crite n ection Figure 2.l for forced convection flow during a transientThis is consistent with temperature is 259 F which is well below the temp The analysis uel plate clad damage could occur.

which fuel power is calculated to be 129 GPM.The flow rate with natural convection at this of Amendment 1 to the SAR.

The analysis is given in Chapter X i

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2.2 Limiting Safety System Settings Applicability

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This specification applies to the ser points for the safety channels monitoring reactor thermal power,- P, coglant flow rate, W, reactor coolan; inlet temperature, T;, and the height of water above the core, L.

Objective 4

To assure that automatic protective action is initiated in order to prevent a safety limit from being exceeded.

Specifications

1. For operation in the forced convection mode; the limiting safety system settings shall be as fo119ws:

P 3.0 MNt(max)

W 800 GPM (min)

T 108 F (max) y L

19 ft, 2 in (min)

2. For operation in the natural convection mode, the limiting safety system settings shall be as follows; P

300 KWt (max)

T 108 F (max) y Bases The analysis shows that there is sufficient margin between these i

settings and the Safety Limits under the nest adverse conditions of ope ration.

(See Section 9.5 of the SAR.)

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'.)

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3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity

. Applicability These specifications apply to the reactivity condition of the reactor, and the reactivity worths of control rods and experiments.

Objective The objective is to assure that the reactor can be shut down at all times and that the safety limit will not be exceeded.

Specifications The reactor shall not be operated at powers in excess of one kilowatt unless the following conditions exist:

a. The minimum shutdown margin provided by control rods, with secured experiments in place and referred to the cold xenon free condition with the highest worth control rod fully withdrawn, is greater than 0.4% A k/k.
b. Any experiment with a reactivity worth greater than 0.45% A k/k must be a secured experiment.
c. The total reactivity worth of die two experiments having the highest reactivity worth is less than 1.6% A k/k.
d. The total reactivity worth of all experiments is less than 2.0%

A k/k.

The maximum excess reactivity shall be limited such that the c.

shutdown margin is always greater than 0.4% Ak/k.

Bases The shutdown margin required by specification 3.1.a is necessary so that the reactor can be shut down from any operating condition and remain shut down after cool down and xenon decay even if one control rod should stick in the fully withdrawn position.

see Specification 1.18.

For definition of " secured experiment,"

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The reactivity of 0.45% A k/k in Spec,1fication 3.1.b corresponds to a three second period. An analysis that shows the peak power does not exceed the safety limit when the reactor power level is increasing on a three second period as the true value of the LSSS is reached is given in Section 9.6 of the SAR (UVAR-18) and in Section XI of Amendment I to UVAR-18.

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The reactivity of 1.6% 6 k/k in Specification 3.1.c corresponds to a 6.9 millisecond period.

Reactor Core DU-12/25 of the SPERT-1 series of tests had twelve plate fuel elements containing 168 grams of U-235 5

substantially similar to the UVAR fuel elements (

Reference:

Thompson and Beckerly, ' Technology of Nuclear Reactor Safety", Volume I, page 683 (1964)). A 6.9 millisecond period was non-destructive. The simultaneous failure of more than two experiments is considered unlikely.

The total reactivity of 2.0% A k/k in Specification 3.1.d plac.es a reasonable upper limit on the worth of all experiments.

Operation of the reactor at a power of less than one kilowatt is allowed to measure the reactivity worth of untried experiments in accordance with procedures approved by the Reactor Safety Committee and to measure the excess reactivity of new core loadings.

The excess reactivity will vary somewhat depending on the core configuration. The limiting factor is the specification on the shutdown margin.

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3.2 Reactor Safety System Applicability 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 in order to assure that the safety limit is not exceeded during normal operation.

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Specification The reactor shall not be operated unless the safety system channels described in the following table are operable:

Measuring Minimum Operating Mode Channel No. Operable Setpoint*

Function in Which Required Pool Water Level 2

19 ft 2 in Scram Forced Convection Monitor.

(min)

Mode Bridge Radiation 1

Scram All Modes-Monitor Pool Water 1

108 F(max)

Scram All Modes Temperature Power to Primary 1

Loss of Power Scram Forced Convection Coolant Pump Mode Application Scram Natural Convection of Power bbde Primary Coolant Flow 1

800 GPM (min)

Scram Fdrced Convection Mode Startup Count Rate 1

2 cps (min)

Preven ts Reactor Startup withdrawal of three shim rods Manual Button 1

Scram All Modes Reactor Power Level 2

3 MN (max)

Scram Forced Convection Mode 0.3 MN (max)

Natural Convection g

Mode Reactor Period 1

3 sec (min)

Scram All Modes Air Pressure to !!cader i

Scram All Modes

  • Values listed are limiting set points.

For operational convenience set points may be changed to more conservative values.

Bases The startup interlock which-requires a neutron count rate of at least 2 CPS before dhe reactor is operated, assures that sufficient neutrons are available for proper operation of the startup channel.

The pool water temperature scram provides protection to assure that if the limiting safety system setting is exceeded an immediate shutdown will occur to keep the fuel temperature below the safety limit.

Power level scrams are provided to assure that the reactor power is maintained within the licensed limits and to protect against abnormally high fuel tempe ratures.

The manual scram allows the operator to shut down the reactor if an unsafe or abnormal condition arises. The period scram is provided to assure that the power level does not increase on a period less than 3 seconds.

This assures that a Safety Limit will not be exceeded as described in Chapter XI of Amendment 1.

Specifications on the pool water level are included as safety measures in the event of a serious loss of primary system water.

Reactor operations are terminated if a major leak occurs in the primary system.

The analysis in Section 9.8 of the SAR shows the consequences resulting from loss of coolan t.

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 or of power to the pump both protect the reactor from overheating.

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3.3 Reactor Instrumentation.

Applicability This application applies to the instrumentation which must be operable for safe operation of the reactor.

Objective The objective is to require that sufficient information is available to the operator to assure 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 the following table are operable.

Measuring Nunimum Operating Abde in Channel No. Operable Which Required Linear Power 1

All Modes Inte rmedia te-- (Log-N) i and Period 1

'All Modes l

Core Gamma Monitor

  • 1 Forced Convection Reactor Room Constant
  • Air Monitor 1

All Modes

  • i Bridge Radiation >bnitor 1

All Modes -

Reactor Face Monitor

  • 1 All Modes
  • Pool Water Level Monitor 2

Forced Convection Mode Pool Unter Temperature 1

All Modec Primary Coolant Flow 1

Forced Convection Mode Start-Up Count Rate 1

Reactor Start-Up Reactor Power Level 2

All Modes

  • The reactor room constant air monitor, reactor face monitor, and core 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.. - - - -

Bases The neutron detectors,. provide. assurance tha,t meas,u,rements of_ the reactor power level is adequately covered,at both low and,high power _

range _s.

The radiation monitors provide information to operating personnel of a decrease in pool water level and of any 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.

Since other methods for determining the radiation levels are required for reactor operation, the reactor can be operated safely if the monitors are not functioning for short periods of time.

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3.4 RADIOACTIVE EFFLUENTS A pplicability This specification applies to the monitoring of radioactive effluents from the Reactor Facility.

Airborne and liquid effluents are discussed separately 'in the following sections.

1. Airborne Effluents Objective The objective is to assure that exposure to the public resulting from the release of A-41 and other airborne effluents will be well below the limits of 10 CFR 20 for unrestricted areas.

Specification When either of the neutron beam ports are drained the centrifugal blower which exhausts that area shal[~b'e in' operation and the airborne activity in the effluent shall be monitored by an instrument 1o' cated in the six inch exhaust

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Bases The basis for this specification is given by the analys ts in Chapter IX of Amendment 1 to the SAR.

2. Liquid Effluents Objective The objective is to assure that exposure to the public resulting from the release of radioactive effluents will be well below the limits of 10 CFR 20 for unrestricted areas.

Specification The activity of liquids released beyond the site boundary shall not exceed 10 CFR-Part 20 limits.

Bases The basis for this specification is given in Section 4.8 in the SAR.

3.5 Confinement Applicability This specification applies to the capability of isolating the reactor room when necessary.

Objective-The objective is to prevent the exposure to the publ.* c resulting from airborne activity released into the reactor room from exeeding the limits of 10 CFR 20 for unrestricted areas.

l Specification The reactor shall not be operated unless the following equipment is operable:

Equipment Function Truck Door Closed Switch Scram reactor when truck door is not fully closed.

Ventilation Exhaust Close and seal when Bridge Radiation Duct Doors bbnitor alarms.

Personnel Door-Close and seal when Bridge Radiation bbnitor alarms.

Emergency Exit Manhole Water -

Water level is high enough to form a Level water seal at least 6 inches in depth.

i Bases The bases for the proper - operation of these itens of equipment are given in Section 6.1 of the S AR.

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3.6 Limitations on Experiments.

Applicability This specification applies to experiments installed in the reactor and its experimental facilities.

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

Specifications The reactor shall not be operated unless the following conditions exist:

a. The reactivity worths of all experiments shall be in conformance with Specifications 3.1.
b. Movable experiments must be worth less than 0.1% 6 k/k.
c. Experiments worth more than 0.1% Ak/k'must be in.serted'or remIovkI[with' the reactor shut down except as noted in item d.
d. Previously tried experiments measured to be worth less than _0.4% A k/k may be inserted or removed with the reactor 2% or more suberitical.
e. If any experiment worth more than 0.4% Ak/k is inserted in the rea,ctor, a procedure approved by the Reactor Safety Committee shall be followed,
f. All materials to be irradiated in the reactor shall be either corrosion resistant or encapsulated within corrosion resistant containers.
g. Irradiation containers to be used in the reactor in which a static pressure will exist or in which a pressure buildup is predicted 1

shall be designed and tested for a pressure exceeding the maximum expected by a factor of 2.

h. Explosive material shall not be allowed in the reactor unless specifically approved by the Reactor Safety Committee.

Experimen ts reviewed by the Reactor Safety Committee in which the material,is potentially explosive, either while contained or if it leaks from the container, shall be designed to prevent damage to the reactor core or to the control rods or instrumentation, and to prevent any changes in reactivity.

2

i. 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.
j. Experimental apparatus, material or equipment to be inserted in the l

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.

Bases The limitations on experiments specified in items a-j are based on the irradiation program authorized by Amendme--

No. 3 to License No. R-66 dated August 13, 1962.

The reactivity of less i n 0.17, which can be inserted or removed with the reactor in operation L.

o accommodate experiments in the hydraulic rabbit.

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3.7 Operation with Fueled Experiments Applicability This specification applies to the operation of the reactor with any fueled experiment within the reactor building.

Objective To assure that the confinement leak rate and fission product inventory in fueled experiments are within limits used in the safety analysis.

Specification The reactor shall not be operated with fueled experiments unless the following conditions are satisfied:

1. For fueled experiments in which the thermal power generated is greater than 1 watt:

a) The experiment must be in the reatror pool and under at least 15 feet of water.

b) The thermal power (or fission rate) gggerated in the experiment is not greater than 100 watts (3.2x10 fissions /second).

c) The total exposure of the experiment is not greater than the equivalent of 6 years continuous operation at 100 watts.

d) The leak rate from the reactor room is not greater than 50% of containment 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.

2. For fueled experimeggs in which the thermal power generated is less than 1 watt (3.2x10 fissions /second) :

a) The experiment may be located anywhere in the reactor building.

b) The total exposure of the experiment is not greater than the equivalent of 6 years of continuous operation at 100 watts.. _ _ _ __

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 Part 20, using an averaging period of 1 year.

The safety analyses for which results are used here are found in the Safety Analysis Report, Section 5.4 The analysis 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 Section XII of Amendment 1.

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3.8 llelght of Water Above the Core in Natural Cenvection Mode of Operation Applicability This spec'ification applies to the height of water above the reactor core when the reactor is operating with natural convection cooling.

Objective To assure that there is a continuous path for circulation of water when the reactor is operated in the natural convection mode.

Specification The reactor shall not be operated in-the natural convection mode unless there is at least one foot of water above the core.

Bases One foot of water above the core is sufficient to provide a continuous path for natural convection cooling.

For other than zero power operation, the radiation levels may require a greater depth for shielding, in which case, the regulations in 10 CFR Part 20 will govern.

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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), and to the time it takes for a rod to drop from the fully withdrawn' to the fully inserted position (free drop time).

Objective To assure that the reactor can be shut down within a specified interval of time.

Specification The reactor will not be operated unless:

(a) The magnet release time for each of the three shim rods is less than 50 milliseconds, and (b) The free drop time for each of the three shim rods is less Chan 700 milliseconds.

Bases Rod drop times as specified will assure that the safety limits will not be exceeded in a short period transient. The analysis is given in Section 9.6 of' the SAR, and Chapter XI of Amendment 1.

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3.10 Emergency Removal of Decay-Heat Applicability This specification applies to the emergency removal of decay heat.

Objective The objective is. to assure that the flow rate from this system is sufficient to prevent overheating of the fuel elements subsequent to a total loss of primary water from the core.

Specification There shall be two separate emergency core spray systems, each capable of maintaining a flow rate of at least 10 gpm over the 64 fuel element positions for the first 30 minutes, and at least 7-1/2 gpm over the 64 fuel element positions for the next 60 minutes following a total loss of coolant.

Bases Either of the two spray systems as specified will provide sufficient cooling to maintain the fuel temperature below its melting point as demonstrated by the evaluation in Section 9.9 of the SAR.

4.0 SURVEILLANCE REQUIREMENTS 4.1 Shim Rods Applicability This specification applies to the surveillance requirements for the shim rods.

Objective To assure that the shim rods are capable of performing their function and that no significant physical degradation in the rods has occurred.

Specification

a. Shim rod drop times shall be measured at intervals not to exceed five months. Shim. rod drop times shall also be measured if the control assembly is moved to a new position in the core or if maintenance is performed on the mechanism.
b. The shim rod reactivity worths shall be measured whenever the rods are installed in a new core configuration,
c. The shim rods shall be visually inspected at intervals not to exceed thirteen months, and when rod drop times exceed the liniting conditions for operation, Section 3.9 of these specifications.

Bases The reactivity worth of the shim rods is measured to assure that the required shutdown margin is available and to provide means for determining the reactivity worth of experiments inserted in the core. The visual inspection of the ahim rods and measurement of their drop times are made to determine whether the shim rods are capable of performing properly.

m 4.2 Reactor Safety System Applicability This specification applies to the surveillance requirements for the reactor safety system of the reactor.

Objective The objective is to assure that the reactor safety system is operable as required by Specification 3.2.

Specification

a. A channel test of each of the reactor safety system measuring channels shall be performed prior to each day's operation or prior to each operation extending more than one day.
b. A channel check of each of the reactor safety system measuring channels shall be performed daily when the reactor is in operation.
c. A channel calibration of the reactor safety measuring channels shal) he per-formed at intervals not to exceed eight months.
d. The power range channels 1 and 2 shall be checked against a, primary system' heat balance at least once each week the reactor is in operation above 100 kilowatts in the forced convection mode.

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e. The following items which are listed in section 3.2 are not con-sidered to be reactor safety measuring channels: Power to primary coolant pump, manual button, header air pressure, and pool water level monitor.

Cperation of these systems will be checked prior to each days operation or prior to each operation extending more than one day.

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l Bases The daily channel tests and channel checks will assure that the safety channels are operable. The semi-annual calibration will permit any long-term drift of the channels to be corrected. The weekly calibration of the power measuring channels will correct for drift and assure operation within the requirements of the license.

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4.3 Emergency Core -Spray System Applicability This specification applies to the emergency core spray system.

Objective The objective is to assure that the spray systems are operable, and will deliver the specified flow rate of emergency coolant.

Specification

a. Whenever the reactor bridge is moved and replaced into position for forced convection operation, the remote coupler for each spray system shall be air pressure checked to assure that there is no leakage,
b. At intervals not to exceed thirteen months, measurements will be made to verify that each spray system will deliver at least 10 gpm for 30 minutes.

Bases The emergency spray system is an engineered safeguard. At the initial installation, each of the two core spray systems was checked to assure that it delivered the flow as specified in Section 3.10 of these speci-fications. Since there are no moving parts and no automatic electronic or mechanical machanisms subject to failure, a verification that-the remote couplers are engaged and not leaking will assure that the two core spray systens are operable.

The annual measurement of the flow rate will verify that each of the two core spray systems will deliver the flow as desired. The preoperational test of the core spray system demonstrated that water -

delivery is at least 10 gpm for 30 minutes and 71/2 gpm for the next 60 minutes.

Subsequent annual tests, which verify the 30 minute flow rate, are adequate to verify design perf rmance. The core spray system is described in o

Section 4.10 and the Safety Analysis is given in Section 9.9 of the SAR. The annual measurement of the flow rate is described. in Chapter IV of the Supple-ment to the SAR.

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4.4 Area Radiation Monitoring Equipment

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Applicability This specification applies to the - area radiation monitoring equipment required by Sections 3.2 and 3.3 of these specifications.

Objective The _ objective is to assure that the radiation monitoring equipment is operating and to verify appropriate alarm settings.

Specification 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. Calibration of the radiation monitoring equipment shall be performed at intervals not to exceed eight months.

Bases Surveillance of the monitoring equipment will provide assurance that sufficient warning of a potential radiation hazard is available.

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',s maintenance of control or safety systems.

Objective The objective is to assure that a system is operable before being used after maintenance has been performed.

Specification Following maintenance or modification of a control or safety ' system or component, it shall be verified that the system is operable prior to its return to service or during initial operation.

Bases The intent of the specification is to assure that work on the system or component has been properly carried out and that the system or component g,

js has been properly reinstalled or reconnected. Correct operation of some systems, such as power range monitors, cannot be verified unless the reactor is operating.

Operation of these systems will be verified during their.

initial operation following maintenance or modification.

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Objeetive The objective is to assure that the closure equipment to the reactor room -

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4.7 Airborne Effluents This specification applies to the surveillance of the instrument which monitors the airborne effluents from the ground floor experimental area.

Objective To ensure that the airborne effluent monitor is operating and properly calibrated.

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Specification

1. Prior to each day's operation or prior to each operation extending more than one day wher. either of the neutron beam ports are chained the centrifugal blower which exhausts that area shall be in operation, a channel check shall be performed on the airborne effluent monitor.
2. A calibration of the airborne effluent monitor will be pe-formed using

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a radioactive source at invervals not to exceed eight months.

Bases The daily channel check of the monitor will assure that it is operable.

The semi-ar.nual calibration with an external source will permit any long term drift to be corrected.

The analysis is given in Chapter IX of Amendment I to the SAR.

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4.8 Reactor Fuel Dose Measurements Applicability This specification applies to reactor fuel possessed under the Reactor Facility Licenses.

Objective The objective of this specification is to ensure that the maximum quantityf.of special nuclear material does not exceed the limits specified in the Facility licenses.

3 Specification A.

The amount of special nuclear material (SNM) possessed at the Reactor Facility will be determined as necessary to ensure that limits specified by the Facility licenses are not exceeded.

As an minimum a evaluation will be completed and documented eveg( six months.

B.

Fuel elements will be irradiated as a part of the core or shipped away from the Reactor Facility as necessary to ensure that the quanity of nonexempt SNM (as defined in 10 CFR Part 73) does not exceed that allowed by the Facility licenses.

If the amount of nonexempt 9NM excceds 5.0 kg the actions specified in the Security Plan will be implemented, C.

Whenever fuel elements which have not been irradiated as a.

part of the core for at least one month, adequate dose rate measurements of representative ' fuel' elemdnts will be made as necessary_ to determine which fuel elements have dose' rates higher than specified by 10 CFR Part 73.67 (b).

Basis The specification will' provide a high degree of ass'u'rance i

that the amount of SNM _and nonexempt SNM does not exceed the license limits.

The amount of nonexempt SNM wi'll" normally j

be maintained at less that 5.0 kg.

In the event that ~ this quantity is exceeded the Reactor Safety Committee will be informed and actions necessary to reduce the amount or

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other appropriate actions as defined in'the Physical Security Plan will be defined.

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5.0 DESIGN FEATURES S.1 Reactor Fuel The fuel elements shall be of the MTR type consisting of plates containing highly i

enriched uranium alloy fuel, clad with aluminum. There shall be twelve fuel plates containing 165 ( 3%) grams of uranium-235, or eighteen fuel plates t

containing 195 ( 3%) grams of uranium-235, in the standard fuel elements.

There shall be six fuel plates containing 82.5 (13%) grams of-uranium-235, or nine fuel plates containing 98 ( 3%) grams of uranium-235, in the control rod fuel elements.

Partially loaded fuct cJements in which some of the fuel plates do not contain uranium may be used.

The mass of uranitin-235 listed above refers to the initial (zero burnup) loading.

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a. The reactor shall be housed in a room designed to restrict leakage.
b. The reactor room shall be equipped with a ventilation system designed to exhaust air or other gases from the reactor room through a stack at a minimum of 37 feet.above ground level,
c. The minimum free volume of the reactor room shall be 60,000 cubic feet.

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5.3 Fuel Storage

a. All reactor fuel elements shall be stored in a geometric array where k,gg is less than 0.8 for all conditions of moderation.
b. Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device surface, temperature will not exceed the boiling point of water.

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1 DRAFT 6.0 ADMINISTRATIVE CONTROLS i

4 6.1 Organization

a. The Reactor Facility shall be an integral part of the School of Engineering and Applied Science of the University of Virginia.

The organizational structure of the University of Virginia relating to the Reactor Facility is shown in Figure 6.1.

b. The Reacter Facility Director shall be responsibic for the overall Facility operation.

During periods when the Reactor Facility Director is absent, his responsibilities are delegated to the Reactor Supervisor.

The Reactor Facility Director shall have a Bachelor of Science or Engineering degree and have a minimum of five years of nuclear experience.

A graduate degree'may fulfill four years of experience on a one for one time basis.

c. The Reactor Supervisor shall be responsible for the day-to-day operation of the UVAR reactor and for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and the provisions of the Reactor Safety Committee.

During periods when the Reactor Supervisor is absent, his responsibilities are delegated to a person holding a Senior Reactor Operators license.

The Reactor Supervisor shall have a Bachelor of Science or Engin-cering degree and have at least 2 years experience in Reactor Operations at this facility, or an equivalent facility, or at least l

6 years experiehce in Reactor Operations.

Equivalent education or experience may be substituted for a degree.

Within nine months after being assigned to the position, the Reactor Supervisor shall-obtain and maintain, a NRC Senior ' Operator license.

d. When the reactor is oper'atisg the following conditions will be met:

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(1) A licensed Senior. Reactor Operator or licensed Reactor Operator shall be present at the reactor controls.

(2) A licensed Senior Reactor Operator shall be on call, but not necessarily at the Facility. -

J (3) At least one other person, not necessarily licensed to operate the reactor, will be present at the facility.

All rearrangements of the core or other nonroutine actions shall e.

be supervised by a licensed Senior Reactor Operator, f.

A health physicist who is organizationally independent of the Reactor Facility Operations group, as shown in Figure 6.1, shall be responsible for radiological safety at the facility.

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UNIVERSITY OF VIRGINIA RADIATION SAFETY C0hMITTEE DEAN, SO100L OF ENGR.

AND APPLIED SCIENCE GIAIRMAN, DEPARTMENT E

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REACTOR SUPERVISOR REACTOR OPERATORS AND STAFF Figure 6.1 g

6.2 Review and Audit

a. There shall be a Reactor Safety Committee which shall review and audit reactor operations to assure that the facility is operated in a manner consistent with public safety and within the terms of the facility license.

The Reactor Safety Committee shall report to the President of_ the University and advise the, Chairman, Department of Nuclear Engineering and the Reactor Facility Director on those areas of responsibility specified below.

b. The Committee shall be composed of at 1 cast five members, one of whom shall be the Radiation Safety Officer of the University.

No more than two members will be from the organization responsible for Reactor Operations.

The membership of the Committee shall be such as to maintain a degree of technical proficiency in areas relating to reactor operation and reactor safety,

c. A quorum of the Committee shall consist of not less than a majorit/

of the full committee and shall include the Chairman or his designee.

d.

The Committee shall meet at least once every six months and on call by the Chariman.

Munutes of all meetings shall be disseminated to responsible personnel as designated by the Committee Chairman.

e. The Committee shall have a written statement defining such matters as the authority of the Committee, the subjects within its purview, and other such administ'rative provisions as are required for effective functioning of the Committee.

As a minimum the responsibilities of the Reactor Safety Committee include the following:

(1) Review and approval of untried experiments and tests which are significantly different from those previousiv used or 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 which may affect the safety of the reactor.

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(3) Review and approve all proposed amendments to the Facility License, Technical Specifications and changes to the Standard

' Operating procedures as 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 significant operating abnormalities or deviations from normal performance of Facility equipment that affect reactor safety.

(6) Review reactor operation and audit the operational records for compliance with reactor procedures, Technical Specifications and License provisions.

These audits shall be performed at least once each calendar year.

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6.3 Operating Procedures

a. 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 assure the safe oper-ation of the reactor, but should not preclude the use of independent judgement 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, (3) Actions to be taken to correct specific and foresect; potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system leaks, abnormal reactiv-ity changes.

(4) Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.

(5) Preventive and corrective maintenance operations which could have an effect on reactor safety.

(6) Periodic surveillance (including test and calibration) of reactor instrumentation and safety systems,

b. Radiation control procedures shall be maintained and made available to all operations personnel.

Substantive changes to the approved procedures shall be made only c.

with the approval of the Reactor Safety Committee.

Changes to the procedures which do not change their original intent may be made with the approval of the Facility Director. All such minor changes to procedures shall be documented and subsequently reviewed by the Reactor Safety Committee.

.A 6.4 Action to be Taken in the Event a Safety Limit is Exceeded In the event a safety limit is violated the following actions shall be.

taken:

a. The reactor shall be shut down and reactor operations shall not be resumed until authorized by the Commission;
b. The occurrence shall be reported to the Reactor Facility Director and the Chairman of the Reactor Safety Committee, or their designees as soon as possible but. not later than the next work day.

Reports shall be made to the Commissica in accordance with Section 6.7 of these specifications;

c. A wi ttten, Safety Limit violation report shall be made which shall include an analysis of the causes of the violation and extent of resulting damage to Facility components, systems or structures, corrective actions taken, and recommendations for measures to prevent the probability of reoccurrence.

This report shall be submitted to the Reactor Safety Committee for review.

6.5 Action to be Taken in the Event of-a Reportable Occurrence In the event of a reportable occurrence, as defined in Section 1.12 of i

these Technical Specifications, the following action shall be taken:

a. The Director of the Reactor Facility shall be notified as soon as possible and corrective action taken prior to resumption of the operation involved.
b. A written report of the occurrence shall be made which shall include an analysis of the cause of the occurrence, the corrective action taken and recommendations for measures to prevent or reduce the probability of reoccurrence.

This report shall be submitted to the Director and the Reactor Safety Committee for review.

c. A report shall be submitted to the Commission in accordance with Section 6.7 of these specifications.

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6.6 Plant Operating Records In addition to the requirements of applicable regulations, records and logs of the items listed below shall be kept in a manner convenient for review and shall be retained as indicated:

a. Records to be retained for a period of at least five years:
1. Normal plant operation.
2. Principal maintenance activities.
3. Experiments performed with the reactor.
4. Reportable occurrences.
5. Equipment and component surveillance activity.
6. Facility radiation and contamination surveys.
7. Transfer of radioactive material.
8. Changes to operating procedures.
b. Records to be retained for the life of the Facility:
1. Gaseous and liquid radioactive effluents released to the environs.
2. Off-site environmental monitoring surveys.
3. Fuel inventories and transfers.
4. Radiation exposures for all personnel.
5. Changes to reactor systems, components, or equipment which may affect reactor safety.
6. Updated, corrected and as-built drawings of the facility.
7. Minutes of Reactor Safety Committee meetings.

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6.7 Reporting Requirements In addition to the requirements of applicable regulations reports should be made to the Commission as follows:

a. A report as soon as possible, but no later than the next working day, to the_ Commission _ Region II Regional Office of:.
1. Any accidental off-site release of radioactivity above permissible limits, whether or not the release resulted in property damage, personal injury or exposure;
2. Any reportable occurrences as defined in Section 1.12 of these specifications; and
3. Any violation of a safety limit.
b. A report within 14 days (in writing to the Director, Division of Reactor Licensing, US NRC, Washington, D.C. 20545 with a copy to the Commission Region II Compliance Office) of:
1. Any accidental off-site release of radioactivity above permissible limits, whether or not the release resulted in property damage, personal injury or exposure.
2. Any reportable occurrence as defined in Section 1.12 of these specifications.
3. Any violation of a safety limit.

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c. A report within 30 days in writing to the Director, Division of Reactor Licensing, US NRC, Washington, D.C.

20545, with a copy to the Commission Region II Compliance Office of:

1. Any substantial variance from performance specifications contained in these Specifications or in the Safety Analysis Report.
2. Any significant change in the transient or accident analyses as described in the Safety Analysis Report.
3. Changes in personnel serving as Chairman of the Department of Nuclear Engineering, Reactor Facility Director, or Reactor Supervisor,
d. A report within nine months after initial criticality of the reactor or within 90 days of completion of the startup test programs, whichevor is earlier, to the Director, Division of Reactor Licensing, US NRC, Washington, D.C.

20545 upon receipt of a new facility license, an amendment to the license authorizing an increase in reactor power icvel or the inste11ation of a new core of a different design than previously used.

The report will include the measured values of the operating conditions or charac-teristics of the reactor under the new conditions, including:

1. T tal control rod reactivity worth.
2. Reactivity worth of the single control rod of highest reactivity worth.
3. Minimum shutdown margin both at ambient and operating temperatures.

A routine report will be made by March 31 of each year to the Director, e.

Division of Reactor Licensing, US NRC, Washington, D.C. 20545, with a copy to the Commission Region II Compliance Office providing the following information:

1. A nart?.tive summary of operating experience (including experiments performed) and of changes in facility design, performance charac-teristics and operating procedures related to the reactor safety occurring during the reporting period.
2. A tabulation showing the energy generated by the reactor (in megawatt hours) and the number of hours the reactor was critical each luarter during the year.
3. A report of the results of the safety related maintenance and in-spections.

The reasons for corrective maintenance of safety related items will be included.

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4. The number of. emergency. shutdowns and inadvertent scrams, including their reasons and corrective aetions taken.
5. A summary of changes to the Facility or procedur'es, which' affect reactor safety, and performance of tests or experiments carried out under the conditions of Section 50.59 of 10 CFR Part 50.
6. A summary of _ the nature and amount of radioactive-gaseous liquid and solid effluents released or discharged to. the environs beyond the effective control of the licensee as meas'ured or '

calculated at or prior to the point of such release or discharge.

7. A description of any environmental surveys performed outside the facility, and
8. A summary of radiation exposures received by facility personnel and visitors, including the dates and time of significant ex-posures (greater than 500 mrem for adults and 50 mrem for persons under.18 years of age) and a summary of the results of radiation and contamination surveys performed within thc facility.

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