ML19340B144

Revised Tech Specs for Triga Mark I at Univ of Texas

ML19340B144
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Site:	05000192
Issue date:	07/31/1980
From:	TEXAS, UNIV. OF, AUSTIN, TX
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i Facility License R-92 Technical Specifications for the TRIGA MARK I University of Texas ,

Austin, Texas Included in this document are the Technical Specifications and the " Bases" for the Technical Specifications. -These bases, whici provide.the technical

- support for.the individual technical specifications,'are included-for infor-mation purposes only. They=are.not part of the Technical Specifications, and they fdo not constitute limitations or requirements to which the licensee

- must adhere. - Reference NRC Regulatory Guide 1.16 and ANSI N378-1974.

Proposed Revision July 1980 8010210 Qf

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~ T TABLE OF CONTENTS Y ,

Page

. l .0 - Definitions 1.

1~.1 Reactor Shutdown -

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! 1.2 . Reactor Seetred 1 1.3 Reactor Operation 1

!. 1.4 Cold Critical 1 1.5- Steady State Mode 1

. 1. 6 - Pulse Mode 1

1. 7 - Shutdown Margin 1

-1.8a- Abnormal' Occurrence 2 1.8b Reportable Occurrence- 2 l 1.9 Experiment _ 2 l 1.10. Experimental Facilities 2' l 1.11 Shim Rod 3 1.12 Transient Rod 3 1.13 Regulating Rod. 3 l 1.14 Fuel Element 3 l.15 Instrumented Element 3

! '1.16 Standard Core 3 1.17 Operational Core 3

.l.18 Safety Limit .

3 1.19 Limiting' Safety System Setting 4

'1.20 .0perable 4 l 1.21 Reactor Safety. Systems-' 4 l 1.22 -Experiment Safety Systems 4 1.23 Measured thiue 4 l 1.24 . Measuring Channel 4 1.25 Safety Channel. 4

'1.26 Channel Check 24 1.27 Channel Test 4 1.28 Channel Calibration

• 5 12 . 0 . Safety Limits and Limiting Settings 6 l

! 2.1- -Fuel Element Temperature Safety Limit- 6-2.2 Limiting Safety System Setting 6

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' 3 D0 Limiting Conditions for operation a 3.~ 1 L ' Steady State-Operation 8

3. 2 _  ; Reactivity' Limitations 8 3 .' 3 '- Pulse Mode Operation. 9 3.4 Control and -Safety System 10 3.5 Radiation-Monitoring. System- 13 i 3.6 . Pool Water Systems ~

14 3.7  ; Limitations on. Experiments 14-

3. 8 - . Irradiations' 17

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Page D 4.'0 Surveillance Requirements 19 l

4.1 " General- 19 -

4.2 Limiting Conditions for Operation 19 l 4.3. Reactor Fuel Elements 23 25 :5 5.0 I Design Features -

5.1 Reactor Fuel 25 5.2 ' Reactor Core 25.

j. .5.3 . Control Rods' 26 5.4 . Radiation Monitoring-System 27

j. 5.5f Fuel Storage 28-5 ~. 6 - Reactor 1P.ool Water- Systems 28 5.7 . Reactor Building and-Ventilation' System 1 29 l

i 6.0 Administrative Controls 31

6.1 Organization- 31 u, 6.2 Review and Audit 31 6.3 ' Action to be-Taken in the Event a Safety 32 Limit 11s Exceeded.

6.4 Action to be Taken in the Event of an 33 Abnormal Occurrence.

6.5 .. Operating Procedures - 33 6.6~ '. Facility Operating Records- 34 6.7 Reporting Requirements 35 l

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-1.0~ DEFINITIONS REACTOR OPERATING CONDITIONS .

1.1 -REACTOR SHUTDOWN

-The reactor is shut- down when the reactor is suberitical by jat least one dollar (0.7% Ak/k) of reactivity.

1.2 REECTOR SECURED The reactor' is secured when all the _following conditions are satisfied:

a. The reactor is shutdown,

b. The console key switch is in the "off" position and the key is removed from the console and under the control of a licensed operator: or stored in a locked storage area, and'

c. No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms, or insertion or withdrawal of in-core experiments.

1.3 REACTOR OPERATION Reactor operation is any condition wherein the reactor is not secured.

1.4 COLD CRITICAL The reactor is in the cold critical condition when it is critical:with the fuel and bulk water temperatures both below 40 C.

-1.5 STEADY STATE MODE

-Steady state mode operation shall mean operation of the reactor with the mode selector switch in the steady-state position.

1.6- PULSE MODE Pulse mode operation shall mean any operation of the reactor -

.with the mode selector switch in the pulse position.

.l.7 . SHUTDOWN MARGIN-Shutdown margin shall mean the minimum shutdown reactivity necessary :to provide confidence lthat the reactor can be _ made subcritical by means of the control and safety systems, starting from any permissible operating conditions and that

'the reactor-will: remain suberitical without further operator action.

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I 'l.8a' ABNORMAL OCCURRENCE

'I L;. ' An " Abnormal' Occurrence" is defined for the purposes of the l i_ reporting requirements of. Section 208 of the Energy Reorgani-

,~ zation Act of 1974 (P.L.93-438) as an unscheduled incident .

(' or event which the Nuclear Regulatory Commission determines

, is significant from the standpoint of public health or safety.'

1.8b- REPORTABLE OCCURRENCE il.

{' A reportable occurrence is any of the following which occurs

! during~ reactor operation:

a. Operation with any safety system setting less conservative than specified in Section .2.2, Limiting Safety System, j Settings:

, * - -b. Operation in violation of a Limiting Condition for Operation;

c. Failure of'a required reactor or experiment safety system component which could render the system incapable of per-forming its' intended safety function; I

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d. Any unanticipated or uncontrolled change in reactivity greater than one dollar; lF l

i- e. An observed inadequacy in the implementation of either administrative or procedural . controls, such that the inadequacy could have caused the existence or development of a condition which could result in operation of the reactor outside the specified safety limits; and

f. Release of fission products from a fuel element.

REACTOR EXPERIMENTS 1.9 EXPERIMENT Experiment shall mean (a)'any apparatus, device,-or material

. - which is not a normal part of the core or experimental facilities, .

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~but which is inserted in'these facilities or is in line with a beam of radiation originating from the reactor core; or (b) any^ operation designed to measure reactor parameters.or characteristics..

[ 1.10' ' EXPERIMENTAL FACILITIES Experimental' facilities shall mean rotary specimen rack, pneumatic transfer systems,-in-pool irradiation facilities, central thimble, and vertical beam tubes.

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.f REACTOR COMPONENTS

= 1.11' SHIM ROD l

E A. shim'rodris a control rod having'an. electric motor drive and scram capabilities.

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( l.12 TRANSIENT ROD-The transient rod is, a control rod with scram capabilities that can be ejected rapidly from the reactor core to produce a pulse.

t 1.13 REGULATING ROD l

l The regulating rod is a. low worth control rod having an

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electric. motor and scram capabilities.

l 1.14 -FUEL ELEMENT A fuel- element ~ is a single TRIGA fuel rod of standard type (i.e. stainless steel clad; 8.5 wt% U - ZrH1.65;

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l nominal 20% enriched).

1.15 INSTRUMENTED ELEMENT l

! An instrumented element is a special fuel element in which-l at least one thermocouple is embedded in the fuel near the i horizontal center plane, f

! 1.16 STANDARD CORE l -

l A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate.

I 1.17 .0PERATIONAL CORE t

An operational core is a standard core for which the core parameters of shutdown margin, fuel temperature, power calibration, and maximum allowable reactivity insertion l

have been determined to satisfy the requirements of the

! Technical. Specifications.

. REACTOR INSTRUMENTATION

!! 1.181 SAFETY LIMIT l

~ Safety limits are 1'imits on'important process variables which

-are found to be necessary -to protect reasonably the integrity of 'certain; of the physical barriers which guard against the uncontrolled release of radioactivity.

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j 1.19 LIMITING SAFETY SYSTEM SETTING

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Limiting safety syster i setting is a setting for automatic protective devices related to those variables having signi--

-ficant safety functions.

1.20 OPERABLE A system, device', or component'shall be considered operable when it is capable' of performing its intended functions' in j a normal manner.

1 21 REACTOR SAFETY SYSTEMS I

' Reactor safety systems are those systems, including their associated input circuits,' which are designed to initiate -

a reactor scram for the primary purpose of protecting the l

reactor or to provide information which requires manual-protective-action to be initiated.

1.22 EXPERI!ENT SAFETY SYSTEMS Experiment' safety systems are those systems, including their associated input circuits, which are designed to initiate a scram for the-primary purpose of protecting an experiment or t to provide information which requires manual protective action to be initiated.

l 1.23. MEASURED VALUE

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l The measured value is the magnitude of that variable as it

(- appears on the output of a measuring channel.

1.24- MEASURING CHANNEL

/ measuring channel is the combination of sensor, intercon-necting cables or lines, amplifiers, and output device which l-are connected for the purpose of measuring the value of a l . variable.

1.25 SAFETY CHANNEL ~

A safety channel ~is a measuring channel in the reactor safety system.

1.26- CHANNEL CHECK-

~A channel check is a qualitative verification of acceptable-performance by'observationtof channel behavior.

1.27 CHANNEL TEST

~A channel test is the introduction of-a signal into the

channel to verify 1that it is operable.

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1.28 CHANNEL CALIBRATION l

A channel calibration consists of comparing a measured value from the measuring channel with a corresponding known value of the parameter so that the measuring channel output can l be adjusted to respond with acceptable accuracy to known values of the measured variable.

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0 2.0- SAFETY LIMITS AND LIMITING SYSTEM' SETTINGS L2.1- FUEL ELEMENT TEMPERATURE SAFETY LIMIT Applicability This specification applies to the temperature of the reactor fuel.

Objective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.

Specifications The temperature in a standard TRIGA fuel element shall not exceed 1000 C under any conditions of operation.

Bases The important parameter for a TRIGA reactor is the fuel j element . temperature . This paraneter is well suited as a l single specificiation especially since it can be measured.

l A loss in-the integrity of the fuel element cladding could l arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the r

presence of air, fission product gases, and hydrogen from l the dissociation of the hydrogen and-zirconium in the fuel-moderator. The magnitude of.this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the-alloy.

The safety limit for the standard TRIGA fuel is based on calculations and experimental evidence. The results indicate l 'that the stress in the cladding due to hydrogen pressure from the dissociation ot zirconium hydride will remain below the ultimate stress provided that~ the temperature of the fuel does not exceed 1150 C and the fuel cladding is below 500*C. The 1000 C safety limit represents a conservative limit to maintain clad integrity for all operating conditions.

2.2 LIMITING SAFETY SYSTEM SETTINGS Applicability-

-This~ specification applies to the scram settings which prevent the safety. limit from being reached.

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'i Objective The. obj ective is. to prevent the safety limits f rom being il reached.

b Specification

The l'miting i safety system setting .shall be 450 C as

.l measured in an instrumented fuel element. The instrumented element shall be located in the B or C ring of the core '

j: configuration.

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!! Basis The limiting' safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preven-ting the safety limit from being exceeded. A setting of 450 C provides a safety margin of at least 550 C for standard TRIGA fuel elements. A part of the safety margin is used to account for the difference between the true and measured temperatures resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the dif ference between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot. If the thermocouple element is located in a region of lower temperature, such as on the periphery.

of the core, the measured temperature will differ by a greater amount from that actually occurring at the core hot spot. Cal-culations indicate that, 'for this case, the true temperature at the hottest location in the core will differ from the measured temperature by no more than a factor of two. Thus, when the temperature in the thermocouple element reaches the i

trip setting of 450 C, the true temperature at the hottest j location would be no greater than 900 C,providing a margin to the safety limit of at least 100 C .for standard fuel elements. . These margins are suf ficient to account for the re-maining uncertainty in the accuracy of the fuel temperature measurement channel and any overshoot in reactor power re-sulting from a reactor transient during steady state mode operation.

l l In the pulse mode of operation, the same limiting safety L system setting will apply. However, the temperature channel ]

.will have no effect on limiting the peak powers generated-l because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode,'however, the temperature trip will'act to reduce

the amount of energy generated in the entire pulse transient by cutting.of the " tail" of the energy transient in the event the pulse rod remains stuck in the-fully withdrawn position.

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3. 0 - LIMITING CONDITIONS FOR OPERATION 3.1 STEADY STATE OPERATION Applicability

~ This specification applies to the energy generated in the j

reactor during steady state operation.

'Obiective

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'The objective is to assure that the fuel temperature safety

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limit will not be exceeded during steady state operation.

Specifications The reactor power level shall not exceed 275 kilowatts under any condition of operation. The normal steady state operating

!- power level of the reactor shall be 250 kilowatts. .However,.

for purposes of testing and calibration, the reactor may be operated at higher power levels not to exceed 275 kilowatts-during the testing period.

Bases Thermal and hydraulic calculations indicate that TRIGA fuel may be safely operated up to power levels of at least 2.0 l megawatts with natural convection cooling.

s L 3.2 REACTIVITY LIMITATIONS l Applicability l

These specifications apply tolthe reactivity condition of the reactor and the reactivity worths of control rods and experiments . They apply for all modes of operation.

Objective i The objective is to assure that the reactor can be shut down l at'all times and'to assure that the-fuel temperature safety limit will not'be exceeded.

Specifications j- .The reactor shall not be operated unless the shutdown margin-

- provided by control rods shall be greater than 0.2 % Ak/k ,

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.a. -the highest worth non-secured experiment in its most

-reactive state, 1-

, ,' b. E the most reactive control rod fully withdrawn,

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!j. c. the reactor in the cold critical condition without xenon, l!-

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d. a maximum excess reactivity. of 2.25% ok/k', .

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lI Ll- e. a maximum rate of reactivity insertion of a standard ll control rod no greater than 0.2% Ak/k/sec, and t

[! f. the total worth of the transient rod equal to or less it than 1.8% Ak/k.

I Bases l

The value of the shutdown margin arsures that.the reactor can be' shut down from any operating condition even if the l highest worth control rod should remain in the fully with-

! drawn position. Maximum core reactivity above cold clean critical, the standard rod reactivity insertion rate, and transient rod total worth, establish available limits on the core power, rate of change of power, and pulse power levels.

Each limiting parameter assures that the normally available reactivity and insertion rates cannot generate operating conditions that exceed fuel temperature safety limits.

3.3 FULSE MODE OPERATION Applicability This specification applies to the energy generated in the reactor as a result of a pulse insertion of reactivity.

' Objective i

! "The objective is to assure that the fuel temperature safety )

limit-will not be exceeded.

Specification The ' reactivity to be inserted for pulse operation shall be determined and limited by a mechanical block on the pulse rod, such_that the reactivity insertion will not exceed

. l.5% ak/ k, and shall be ' initiated from power levels less than 1 kilowatt.

Bases Experiments with pulsed operation of TRIGA reactors indicate that . insertions up to 3.5% Ak/k have not exceeded the fuel-9

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p-h temperature safety limit. For a 1.5% ak/k pulse the expected fuel temperature rise above ambient is approximately 400 C.

Thus at an ambient temperature of 50 C a safety margin of -

550 C is expected relative to the defined safety limit of 1000 C.

3.4 CONTROL AND SAFETY SYSTD1 3.4.1 Scram Time Applicability 1

This specification applies to the time required for the scrammable control rods to be inserted fully from the instant that a safety channel variable reaches the Safety System Setting.

Objective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

Specification The scram time measured from the instant a simulated signal reaches the value of a limiting safety system setting to the instant that the slowest scrammable control rod reaches its fully inserted position shall not exceed 2 seconds.

Basis This specification assures t: -he reactor will be shut down p romptly when a scram signa s initiated. Experiments and analysis have indicated that for the range of transients anticipated for a TRIGA reacror, the specified scram time is adequate to assure the safety of the react..e 3.4.2 Reactor Control System Applicability This specification applies to the information which must be available to the reactor operator during reactor operation.

Objective The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.

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Specification

.The reactor shall not'be operated unless the measuring

. channels listed in the following table are operable.

. Reactor Minimum Measuring Channels Effective Mode Measuring Channel- Operable Steady State Pulse Fuel Element Temperature 1 .X X Linear Power-Level 1 X Percent Power Level 1 X Integrated Pulse Power 1 X t

The minimum measuring and safety channels are required to be operable prior to reactor startup. However, failure of any channel while the reactor is in operation is permissible, provided that immediate cor-rective action is taken.

3.4.3 Reactor Safety System

-Applicability

. This specification applies to the reactor safety system channels.

Objective The objective is'to specify the minimum number of reactor safety system' channels that must be operable for safe operation.

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

Minimum Reactor. Safety Channels Number Effective Mode Safety Channel Operable Function Steady State Pulse

- Console Scram Button 1 Scram on operator X X demand Fuel Element Temp. 1 Scram at 450 C X X Linear. Power Level 1 Scram at 110% X X of full scale Percent Power Level- 1 Scram at 110% X of full power 11

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-i 4 Number Effective Mode i Safety Channel- Operable Function .Steadv State- Pulse

(( High Voltage 1- Scram on loss of X X p Scram Relay Power 1 Scram on loss of X X l Magnet Current 1 Scram on loss of X X I' Startup Channel l' Interlock, prevent X rod withdrawal with less than 2 l neutron cps Rod-Drive Control 1 Interlock, prevent X simultaneous with-

, drawal of any two j or more rods Rod Drive Control 1 Interlock, prevent X transient rod with-

! drawal unless all rods are down j -Rod Drive Control 1 Interlock, prevent X withdrawal of any rod except transient Rod Drive Control 1 Timer, drop tran- X sient rod within j 15 secs of pulse Reactor Pool Level 1 Alarm at < 4.."- X X Pool Bulk Water Temp. 1 Alarm at,_> 48 C X X Heat Exchanger Pressure Difference 1 Alarm at < 7kPa X X Bases The manual' scram allows the operator to shut down the system is an unsafe or_ abnormal condition occurs. The fuel temperature and power level scrams provide protection to assure that the reactor can be shut down before theLsafety limit on the fuel element temperature will be exceeded. In the event of. failure of the power supplies for high voltage, scram relays or magnet current, continued reactor operation is prevented without adequate measurement and control

~ instrumentation.

The interlock to prevent startup of the reactor at power levels less than 4 x 10-3 watts, which corresponds to approximately 2 cps, assures that sufficient neutrons are available for proper.startup. The inter-lock to prevent simultaneous withdrawal limits the maximum positive reactivity' insertion rate available for steady state operation.

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An interlock that prevents transient rod withdrawal during

! steady state operation protects against inadvertent pulse operation. The interlock to prevent withdrawal of the shim or regulating rod.in the pulse mode is designed to prevent changing the critical state of the reactor just' before pulsing. A

. preset timer insures that the reactor power level will reduce to a low level after pulsing.

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.3.5 RADIATION MONITORING SYSTEM Applicability This specification applies to the radiation monitoring infor-mation which must be available to the reactor operator during reactor operation.

Objective

! The objective is to assure that sufficient-radiation monitoring information is'available to the operator to assure safe opera-tion of the reactor.

Specification i The reactor shall not be operated unless.the radiation monitoring channels listed in the following table are operable.

Radiation Monitoring Channels Function Number 4

,i Area Radiation Monitor Monitor radiation 1

!< levels within the

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reactor room 1

'.- Continuous Air Radiation 1 Monitor, During maintenance to the radiation monitoring channels, the l intent of this specification will be satisfied by sub-stitution of portable gamma sensitive instruments _that are

- subject to visual observation or that provide audible alarms.

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!I t Bases i

The radiation nonitors provide 'information co operating personnel of any impending'or existing danger from radiation so that'there will be sufficient-time to evacuate the facility

and take the necessary steps to prevent the spread of' radio-activity- to the surroundings.  ;

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' 3. 6 '- ~ POOL WATER SYSTEMS Applicability

$ - This specification . applies to the operating condition of the

- reactor pool'and pool water systems.

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Objective i

- The objective is to assure that coolant water is available L

to: provide adequate cooling of the reactor core and adequate

- radiation shielding.

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Specifications Corrective action shall be taken or the reactor shall. be l shutdown if i

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a. less than 4.5 meters of water exists above the top grid-L- plate ,

the bulk pool temperature exceeds 48 C, or

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c. during heat exchanger operation,the chilled water l

outlet pressure is less than 7kPa greater than

the pool water inlet pressure to the heat exchanger.

. Bases-

a. Calculations and experiments at TRIGA facilities has ,

shown that 4.5 meters of water anove the reactor core is sufficient t'o provide reasonable rcdiation' levels

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j above the reactor pool.

I The bulk water temperature constraint assures that

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sufficient core cooling exists under all anticipated operating conditions.

c. :A pressure. difference at the heat exchanger chilled

. water outlet and pool water-inlet of 7kPa should be-

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. sufficient. to prevent loss of pool water to the secondary

1.  : chilled water system in the event of a leak between the two systems.

3.7: LIMITATIONS ON EXPERIMENTS

+ . Applicability This; specification applies to ' experiments installed in the Jreactor and.its experimentalifacilities..

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-- Objective-The objective is to prevent damage to the reactor or exces- .

sive release.of radioactive material'in the event of an

- experiment failure.

Specifications The reactor shall_ not be operated unless the following condi-tions governing experiments exist.

a. Non-secured experiments shall have reactivity worths less

.than 1 dollar,

b. The reactivity worth of any single experiment shall be less than 2.5 dollars.

c. The total reactivity worth of in-core experiments shall not exceed 3.00 dollars, including the potential re-activity which might result from experimental malfunction, experiment flooding or voiding, removal or insertion of experiments.

d. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, po-tentially explosive materials, and liquid fissionable materials shall be doubly encapsulated.

e. Explosive materials,.such as gunpowder, TNT, PETN, or nitro-glycerin, in cuantities greater than 25 milligrams shall not be irradiated in the. reactor or experimental facilities.

Explosive materials in quantities less than 25 milligrams may be irradiated provided the pressure produced upon detonation of_the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

f. Experiment materials, except fuel materials, which could off-gas,-sublime, volatilize, or produce aerosols under (1) normal ' operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment shall be limited in activity such that if 100% of the

. gaseous activity or radioactive' aerosols produced escaped to the reactor room or the atomsphere, the airborne con-centration of radioactivity averaged over a year would

> not exceed the limit of Appendix B of 10 CFR Part 20.

g. In: calculations pursuant c'o f. above, the following assumptions shall be used:

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(1) If the effluent from an experimental facility exhausts l

through a holdup tank which closes automatically on high_ radiation level, at least'10% of the gaseous activity or aerosols produced will escape.

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(2) If the effluent from an experimental facility

. exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of these vapors can escape.

.f (3) For materials whose boiling point is above 55 C and where_ vapors formed by boiling this material can escape only through an undisturbed _ column

! of water above the core, at least 10% of these vapors can escape.

h. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment _is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 milli-curies.

4 1. If a capsule fails and releases material which could i damage the reactor fuel or structure by corrosion or other means, removal and physical inspection shall be performed to. determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director,or his designated alternatg and determined to be satisfactory-before operation of the reactor is resumed.

Bases

a. This specification is intended to provide assurance that the worth of a single. unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were to be inserted suddenly.

b. The maximum worth of a_ single experiment is limited so that its removal from the cold critical reactor will not -

result in the reactor achieving a power level high enough

-to exceed the core temperature safety limit. Since ex-periments' of such worth must be fastened in place, its removal from the reactor operating at full power would result in a -relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

- c. The maximum worth of all experiments is also limited to a reactivity value such that the cold reactor will not-achieve a power level high enough to exceed the core temperature safety limit if the experiments were removed.

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1 e d. Double encapsulation is required to lessen the experimental-

_ hazards of some types of materials.

e. This ' specification is intended to prevent damage to reactor components resulting from failure of an'experi -

ment involving explosive. materials.

f. This specification is intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of- 10 CFR Part 20 will be released to the-atmosphere outside the facility boundary.

g. This specification. clarifies calculation conditions of part (f).

h'. The 1.5-curie limitation on iodine 131 through 135

. assures that in the' event of ' failure of a fueled experi-ment leading to total release of'the iodine, the exposure-dose at the exclusion area boundary will be less than that allowed by 10 CFR Part 20 for an unrestricted area.

i. Operation of the reactor with the reactor fuel or struc-ture damaged is prohibited to avoid release of fission

-products.

3.8 IRRADIATIONS Applicability This specification ~ applies to irradiations performed in the

-irradiation facilities contained in the reactor pool as de-fined in'Section 1.10. ' Irradiations .are a ' subclass of experi-ments that. fall within the' specifications hereinafter stated in this section.

Obiective The objective is to: prevent damage to the reactor, excessive

' _ release of radioactive materials, or excessive personnel-tradiation exposure 'during the performance of an irradiation.

Specifications

'A device or material shall not be irradiated in an irradiation facility under-the classification ,of-an irradiation unless the-

following con'ditions exist:

a. The irradiation. meets all the specifications of Section

.4.2.5;for an experiment,

~ b '. The expected radiation field produced by. the device or sample'upon removal 1from the' reactor is not more than 10 rem /hr at one foot, otherwise it shall be classed as an experiment..

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. c. The . device' or material is ' encapsulated in ' a ' suitable - container, The reactivity worth of. the device' or material is ' O.25 dollars or

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l: less, 'otherwise it shall be classed as an . experiment, and

<. . The device or-material'does not remain in'the creactor for a'-

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_. period of. over _15 days,zotherwise it shall be classed as an '

experiment.

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. Bases This specification is intended to provide assurante that the

- lspecial'elass of experiments called' irradiations will be per-j- formed in a manner _that will not permit any safety. limit to be.

- exceeded.

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4.0 SURVEILLANCE REQUIREMENTS 4.1 GENERAL Applicability This specification applies to the surveillance requirements of any . system related to -reactor safety.

Obj ec tive The objective is to verify the proper operation of 'any systemf related to reactor safety.

Specifications Any. additions, modifications, or maintenance to the ventila-tion system, the core and its associated support structure, the pool.or'its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications.to which the systems were originally designed and fabricated or to' specifications approved by the Reactor Committee.

A systen shall not be considered operable until af ter it is tested successfully.

Bases This specification relates to changes in reactor systems which could affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications,- then it can.be assumed that they meet the-presently accepted operating criteria.

4.2 LIMITING CONDITIONS FOR OPERATION 4.2.1 Reactivity Requirements Applicability These specifications apply to the surveillance requirements for reactivity control.of experiments and' systems.

' Objective The objective is to measure and verify the worth, performance, and operability of those systems affecting the reactivity of the reactor.

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4 "l Specifications

a. The reactivity worth of each control rod and the shut-down margin shall.be determined annually,but at intervals not to exceed 14 months.

'l b. The reactivity worth of an experiment shall be estimated J or measured, as appropriate, before reactor operation

., with said experiment.

'i

.c. The. control rods shall be inspected visually for deteriora-

. tiongat intervals not to exceed 2-years.

'f d. The transient rod drive cylinder and associated. air supply system shall' be inspected, cleaned, and lubricated semi-annually, but at intervals not to exceed 8 months.

e. The reactor shall Ire pulsed. semiannually to cc= pare fuel temperature measurements and peak power levels with those of previous pulses of 'the same reactivity value or the reactor. shall not be pulsed until such comparative pulse measurements are performed.

Bases The ' reactivity worth of the control rods is measured to

-assure that the required shutdown margin is available and to provide an accurate means forEdetermining the . reactivity .

worths of. experiments inserted in'the core. Long term ef fects

.ofLTRIGA reactor operation are such that measurement of the reactivity worth on an annual basis is adequate to insure no significant . changes in the shutdown margin. The visual inspection of the control rods.is made to' evaluate corrosion and wear characteristics caused by operation in the reactor.

The reactor is pulsed at suitable intervals and a comparison made with previous similar pulses to determine if changes -in fuel or core characteristics are taking place.

4 .' 2 . 2 - ' Control and Safety System Applicability-These specifications apply to-the surveillance requirements for measurements, tests, and calibrations of1the1 control and:

safe ty - syste ms.

Objective The objective ~is to verify the performance and operability of-

.those systems and components which are directlyl related to ,s reactor safety.

20 I

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- Specifications

!- a. The scram time shall be measured annually, but at intervals lnot to exceed 14 months.

b. A chaunct test of each of the reactor safety scram channels for tho' intended mode of operation shall be performed prior to each day's operation or prior to each operation extending more than one day.

c. A functional check of all safety channels ' listed in 1

. section 3.4.3 ~ shall be performed semiannually, but at intervals not to exceed 8 months.

d. A Channel Calibration shall be made of.the power level monitoring channels by the calorimetric method annually, but at intervals not to exceed 14 months.

9

e. A test of the power level safety circuits shall be performed semiannually, but at intervals not to exceed 8 months.

Bases thasurement of.the scram time on an annual basis is a check not only of the scram system electronics, but also is an

- indication of the capability of.the control rods to perform properly. The channel tests will cssure that the safety system channels are operable on a daily basis or prior to an extended run. The power level channel calibration will assure that the reactor will be operated at the proper

-power levels. Transient control ~ rod checks and semiannual maintenance insure proper operation of this control rod.

4.2.3 Radiation Monitoring System Applicability This specification applies to the surveillance requirements for the area radiation nonitoring equipment and the continuous air. monitoring system.

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

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Specification The area radiation monitoring system and the' continuous air monitoring system shall be calibrated semi-annually, but at intervals not to exceed 8 mon ~ths, and shall be verified to be operible at weekly intervals.

Basis Frequent inspection by weekly verification of area radiation and and air monitoring . system set points, in conjunction with semi-annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span.

4.2.4 Pool Water Systems Applicability s This specification applies to the reactor pool water tank, heat' exchanger cooling system and purification system.

Objective The objective is to maintain pool water level and quality to assure adequate core cooling, core radiation shielding, and control of water corrosion and activation properties.

! Specification l'

L a. Pool water conductivity shall be sampled at least l weekly. The conductivity.shall not' exceed 6 micro-mhos per centimeter averaged over a month.

f- b. The reactor normal pool level- shall be verified daily

[ when the heat ' exchanger chilled water inlet and outlet block valves are open.

! c. During periods when operators are not present in the

[.

laboratory, the heat exchanger primary pump will be shut

down and the' chilled water inlet and outlet block valves closed.

Bases

! The specification for pool water conductivity assures that

. potential corrosion products and water activation are con-trolled at acceptable levels.

The specifications on heat exchanger operation establish that reactor pool _-level conditions are monitored by qualified personnel to detect abnormal operation. .l l

22 i L A-

4.2.5- Experiment and Irradiation Limits.

Applicability

'This specification applies to the surveillance requirements for experiments installed in the reactor and its ' experimental

-facilities and for irradiations performed in the irradiation f acilitie s.

Objective

-The objective is to prevent the conduct of experiments or irradiations which nay damage the reactor or release excessive amounts of radioactive materials as a result of failure.

Specifications

.a. A new experiment shall not be installed in the reactor or its experimental facilities until a hazards analysis has been performed and reviewed for compliance with the Licdtations on Experiments, Section 3.7, by 'the Reactor Committee. Minor modifications to a reviewed and approved experiment may be made at the' discretion of the senior reactor operator responsible 'for the operation provided that the hazards associated with the modifications have been reviewed and a determination made and documented that the modifications-do not create a significantly different, a new, or a greater hazard than the original approved experiment.

b. An irradiation of a ner type of device or material shall not be performed until an analysis of the irradiation

~

has been performed and reviewed for ~ compliance with the Limitations on Irradiations, Section 3.7, by the Reactor

! Supervisor.

L l Bases The . procedure of administrative review of experiments and.

l irradiations by .the reactor staff and reactor committee as l

appropriate,~ monitors reactor utilization so that reactor i operation'is' conducted without endangering the safety of the reactor or exceeding the limits of the Technical Specifications.

4~. 3 REACTOR-' FUEL' ELEMENTS

(.

l Applicability.

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This specification applies to the surveillance requirements for the' fuel elements.

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Objective The objective is to verify the continuing integrity of the fuel element cladding.

Specifications All fuel elements shall be inspected visually for damage or deterioration and measured for length and bend after the first 100 pulses and then after each 500 pulses thereafter.

The maximum interval between inspections of fuel elements shall be no greater than two years. The reactor shall not be operated with damaged fuel. A fuel element shall be considered damaged and must be removed f rom the core if:

a. In measuring the tranverse bend, the bend exceeds the '

original bend by 1.6 millimeters, or

b. In measuring the elongation, its length exceeds the original length by 2.6 millimeters, or

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

Bases The f- of inspection and .surement schedule is based on th. .ameters most likely t. f t'e c t the fuel cladding of a pulsing reactor operated at ms _; rate pulsing levels and utilizing fuel elements whose characteristics are well known.

The limit of transverse bend has been shown to result in no difficulty in disassembling the ccre. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching. Experience with TRIGA reactors l has shown that f uel' element bowing that could result in touching has_ occurred without deleterious effects. The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coclant flow.

24

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! 5.0 DESIGN FEATURES 1

li 5.1 REACTOR IUEL 1

ll Applicability This specificiation applies to the fuel elements used in the reactor core.

lj -Objective il The objective is to assure that the fuel elements are of l

such a design and fabricated in such a manner as to permit their use with a high degree of _ reliability with respect ,

to their physical and nuclear characteristics.

Specifications Standard TRIGA fuel The individual unirradiated standard TRIGA fuel i elements shall have the following characteristics:

i

! (1) Uranium content: 8.5 Wt-% uranium enriched to i a nominal 20% Uranium-235.

l l (2) Zirconium hydride atom ratio: nominal 1.6 j' hydrogen-to-zirconium, ZrH .

l (3) Cladding: 304 st?.inless steel, nominal .51

( millimeters thick.

Bases The Design basis of the standard TRIGA core demonstrates that 250 kilowatt steady or 20 megawatt-sec pulse operatio .

l presents a conservative limitation with respect to safety l limits for the maximum temperature generated in the fuel.

l The fuel temperatures are not expected to exceed 450'C l during any condition of normal operation.

l

! 5.2- REACTOR CORE Applicability This specification applies to the configuration of fuel and in-core experiments.

25

Objective The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.

i; Specifications

a. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator elements positioned in the reactor grid plate.

b. The TRICA core assembly shall be the standard TRICA fuel.

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c. Single positions may be occupied by control rods, neutron startup source, ionization chamber, voids, in-core experimental facilities, or graphite reflector elements.

Bases Standard TRICA cores have been in use for years and their characteristics are well documented.

5.3 CONTROL RODS Applicability This specification applies to the control rods used in the reactor core.

Objective l The objectise is to assure that the control rods are of such l a design as to permit their use with a high degree or reli-ability with respect to their physical and nuclear characteris-tics.

Specification

a. The shim control rod shall have scram capability and contain borated graphite, B C4 powder,or boron and its l compounds in solid form as a poison in aluminum or stainless _ steel cladding.

b. The regulating control rod shall have scram capability and contain the materials as specified for the shim control rod.

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The transient control rod shall have scram capability c.

and contain' borated graphite or boron and its compounds '

in'a solid' form as a poison in an altiminum or stainless steel ~ clad. The transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions.

Bases The poison requirements;for the control rods are satisfied by using neutron absorbing borated graphite, B C 4 powder, or boron and its compounds. These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insura mechanical stability during movement and to isolate the poison' from the pool water environment. Scram capabilities are provided for rapid insertion of the control rods which is the . primary safety feature of the reactor. The transient control rod is designed for a reactor pulse, u

5.4 RADIATION MONITORING SYSTEM Applicabilitv-

.This' specification describes the functions and essential e components of the area. radiation mcnitoring equipment and' the system for continuously monitoring airborne radioactivity.

Objective The' objective is to describe the radiation monitoring equipment that.is available to the operator to assure safe operation of the reactor.

L Specification-Th2 radiation monitoring equipment listed in the following table will be available for reactor operation.

4 Radiation-Monitoring Channel and Function

' Area Radiation Monitor (gamma sensitive instruments)

Function - Monitor radiation fields in key i locations with alarm and readout at control console

. Continuous' Air Radiation Monitor (beta, gamma sensitive detector _with air particulate' collection capability)

Function - Monitor concentration of radioactive-particulate activity in the reactor room with-

alarm and readout available to operator.

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L l Basis The radiation monitoring system is intended to provide -

information co operating personnel of any impending or existing danger f rom radiation so that there will be l

L sufficient. time to evacuate the facility and take the

. necessary steps to. prevent the spread of radioactivity to the surroundings.

5.5 FUEL STORAGE

(

t Applicability This specification. applies to the storage of reactor.fu'el- ,

at times when it is not in the reactor core.

Objective.

[

The objective is to assure that fuel'which is being stored will not become ' critical and will not reach an unsafe tempe rature .

Specifications l

a. All fuel elements shall be stored in a geometrical array

.where the k-effective 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-temperature will not exceed

. design values.

Basis The limits imposed by Specifications 5.5a and 5.5b are conservative and assure safe storage.

- 5.6 ' REACTOR POOL WATER SYSTEMS

' Applicability

' This specification applies to the pool containing the reactor and'to the cooling of the core by the pool-water.

! 3 ~ 50biective The objective. is to. assure that' coolant water shall be

-available' to provide adequate cooling of the reactor core and adequate radiation shielding.

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S'pecifications -

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

b. Th'e' pool water inlet and. outlet pipes to. the water purification- system shall 'not extend more than 4.5 meters below.the top of the reactor pool when fuel is in the core.

c. ' Pool water inlet and outlet pipes to the heat exchanger shall not extend more than 4.5' meters below the top of the-

• reactor pool when fuel is in the core.

Bases

a. This specification is based on thermal and hydraulic calculations which show that a standard 63 element TRIGA-core can operate in a safe manner at power levels up to 1,700 kW with natural convection flow of the coolant water.

b. In the event of accidental pumping with no return of pool water through inlet and outlet pipes of the puri-fication system, the pool water level will drop'no more than 4.5 meters . f rom the top _ of the pool.

c. - In the event of accidental pumping with no return uof pool water through the- heat exchanger water system, the pool water level will' drop no more than 4.5 meters from the top of the pool.'

5.7 REACTOR FACILITY VENTILATION SYSTEM Applicability This specification applies to the room which houses the reactor.

Objective'

' The objective is to assura that provisions are made to restrict' the ' amount of release 'of radioactivity into the

~

environment.

Specifications

a. :The reactor shall be housed.in a closed room which has been modified to restrict leakage. The minimum volume in the reactor room shall be approximately 680! cubic -

. meters.

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b. The laboratory air ventilating syscem shall be contained and shall circulate' air within the confines

~

of the reactor room. Air and exhaust gases from the reactor room shall be released to the environment only as a result of leakage through doors, etc.

c. The pneumatic transfer system shall be a closed loop carbon dioxide filled system.

Bases The facility is designed such that the ventilation system In.the event that excessive airborne radioactivity is detected the ventilation system shall be shutdown to minin1ze transport of airborne materials. Analysis indicates that in the event of a major fuel element failure personnel would have sufficient time to

evacuate the facility before the 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week occupa-tional dose is exceeded.

i 30

6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION

a. The facility shall be under the direct control of. the Director or a licensed senior operator designated by him to be in direct control. The Director shall be responsible to the Dean of the College of Engineering and the Chairman of the Department of Ikchanical Engi-neering for safe operation and maintenance of the reactor an,d its associated equipment. The Director or his appointee shall review and approve all experiments and experimental procedures prior to their use in the reactor. He chall enforce rules for the protection of personnel f rom radia-tion.

b. The safety of operation of the Nuclear Engineering Teaching Laboratory shall be related to the University Administration as shown in the following chart.

Office, President University of Texas at Austin i

Dean Ns College of Engineering s i

eactor Radiation Chairman "" **

Safety Department of Committee Mechanical Engineering g I I I

l Director Nuclear Engineering k-  %

Teaching Laboratory s'

s I -

N Supervisor , /

Reactor Operations 6.2 REVIEW AND AUDIT

a. A Reactor Committee of at least three. (3) members knowledgeable in fields which relate to Nuclear Safety shall review, evaluate, and approve safety standards-associated with the operation and use of the facility.

The University Radiological Safety Officer shall be an ex-officio member . of _ the Reactor Committee. The juris-

' diction of the Committee shall include all nuclear opera-tions in the facility and general safety standards.

31

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= b. ' The operations of' the Reactor Committee shall be in' accordance with a written charter, including provisions for:

.(1) Meeting frequency, (2) Voting rules, (3) Quorums, (4) Method of submission and content of presentation to the Committee, (5)'Use of subcommittees, and (6) Review, approval, and dissemination of minutes.

c. The Committee thereof shall audit. reactor operations at least quarterly, but at intervals not to exceed

'four months.

d. The: responsibilities of the, Committee thereof include,

. but _are . net limited to, the following:

(1) Review and approval of experiments utilizing the reactor facilities;

. (2) Review and approval of all proposed changes to the facility.. procedures, and Technical Specifications; (3) Review of 'the operation and operational records of the facilityr (4) Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50; (5) Determination oflwhether a proposed change- te s't ,

or experiment would constitute an unreviewed safety question or a change in the-Technical Specifications; and.

5- .. .

. (6) Review.of chnormal performance of facility equipment ',

'and operating anomalies.

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- o. 3 - ' ACTION TO BE TAKEN IN THE EVENT A SAFETY-LIMIT IS EXCEEDED r

In.the event a safety lindt is exceeded: ,

l.

a. The reactor shall be shut down and reactor operation

.shall not be resumed until authorized by the NRC;

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b. An inmediate report of the occurrence shall be made to the Chairman, Reactor Committee, and reports shall be made to the NRC in accordance with Section 6.7 of these specifications: and

c. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recur-rence. This report shall be submitted to the Reactor Committee for review and tnen submitted to the NRC when authorization is sought to resume operation of the reactor.

6.4 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE In the event of a reportable occurrence, the following action shall be taken:

a. The Director or his designated alternate shall be notified and corrective action taken with respect to the operations involved;

b. The Director or his designated alternate shall notify the Chairman of the Reactor Committee;

c. A report shall be made to the Reactor Committee which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for j measures to prevent or reduce the probability of recur-rence; and l

d. A report shall be made to the NRC in accordance with Section 6.7 of these specifications.

l 6.5 OPERATING PROCEDURES Written operating procedures shall be adequate to assure the safety of operatien of the reactor, but shall not preclude the use of independent judgement and action should the situa-tion require such. Operating procedures shall be in effect for the following items:

a. Testing and calibration of reactor operating instru-mentation and controls, control rod drives, area radia-tion monitors, and air particulate monitors;

b. Reactor startup, operation, and shutdown;

c. Emergency and abnormal conditions, including provisions for evacuation, reentry, recovery, and medical support; 33

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~d.. Fuel elament and experiment loading or unloading;

e. Control rod removab or replacement; -

f. Routine maintenance of the control rod drives.and.

reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety;

g. Actions to be taken to correct specific and foreseen l-potential malfunctions of systems or components, including responses to alarms and abnormal reactivity

! ' changes ; and i h '. Civil disturbances on or near the facility site.

Substantive . changes to the above procedures shall be made-only with the approval of the Reactor Committee. ~ Temporary j changes to the-procedures that do not change their original intent may be made by the Director or his designated alternate.

All such temporary changes shall be documented and subsequently reviewed by the. Reactor Committee.

6 ~. 6 FACILITY OPERATING RECORDS In addition to the requirements of applicable regulations, and in no way substituting therefore, records and logs shall

[i be prepared of at least the following items and retained for

! a period of at least five years for items a through f and 1

indefinitelylfor items g through k.

'a. Normal reactor operation,

b. Principal maintenance activities,

c. Reportable occurrences,

~

'd. Equipment and component surveillance activities required

[ by~the Technical Specifications,

e. Experiments performed with the reactor,

f. Caseous and liquid radioactive' effluents released to

,the environs, g .~ 1Any offsite environmentalLmonitoring surveys, ,

-h. _ Fuel-inventories and transfers,

~

-i. Facility. radiation and contamination surveys, 4

j. Radiation exposures for'all personnel, and k; Updated, corrected, and as-built drawings'of'the facility.

34 -

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i 6.7 [ REPORTING REQUIREMENTS In addition.to the requirements of applicable regulations, -

and in no way substituting therefor, reports shall be made to the NRC Region IV, Of fice of Inspection and Enforcement

~

as follows.

a. A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph of:

(1) Any accidental release of radioactivity-above permissible limits in unrestricted areas whether  !

or not the release resulted in property damage,

. personal injury, or exposure; ,

(2) Any violation of the safety limit; and I (3) Any reportable occurrences as defined in Section 1.8 of these specificiations.

b. A report within 10 days in writing of:

(1)- Any_ accidental release of radioactivity above

_ permissible limits'in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure. The written report (and,to the extent possible. the preliminary 3

telephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned

j. to prevent reoccurrence of the event; (2) Any violation of a safety limit; and (3) Any reportable occurrence as defined in Section 1.8 of these specifications.

c. A report within 30 days in writing of:

(1) Any significant' variation of-measured values '

Jfrom a corresponding predicted or previously i measured value of safety-connected operating characteristics occurring during operation of the' reactor; (2)- Any significant change in the transient or .

accident analysis as described in the Safety Analysis Report; 35 I

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'(3) Any changes in facility organization; and

'(4 ) - Any observed inadequacies in the implementation of administrative or. procedural controls.

6.7.1 A report within 90 days after completion startup testing of the reactor,upon-receipt of a new facility license or .

an amendment to the license authorizing an increase in reactor. power level describing the measured values of the operating conditions.or characteristics of the reactor under the new conditions including:

a. An evaluation of facility performance to date in comparison with design predictions and specifications ;

and

b. A reassessment of the safety analysis submitted with the. license application in light of measured operating characteristics when such' measurements indicate that '

there may be substantial variance from prior analysis.

6.7.2 An' annual report covering the operation of the unit during the previous calendar year submitted prior to March 31 of each year providing the following information.

a. A .brief narrative summary of:(1). operating experience l (including experiments performed) . . (2) changes in f acility . design, performance characteristics, and ,

l operating procedures related to reactor safety and occurring during the reporting period, and (3) results of. surveillance tests and inspections;

b. Tabulation of the energy output (in megawatt days) of the reactor hours the reactor was critical, and the cumu-lative total energy output.since initial criticality:

c. The number of emergency shutdowns and inadvertent scrams, including reasons - therefore;

.d. Discussion of the major maintenance operations performed

! during the period, : including the effect, if any , -on'the

-safety of the operation of the: reactor and the reasons.

for any corrective maintenance required;

.e. =A brieffdescription, including a summary of.the. safety l- evaluations of. changes'insthe facility or in procedures.

l! ' and of tests and experimente carried out pursuant to Section.50.59 of 10 CFR Part 50; l

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! f. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as calculated or measured at or prior to the point of such release or discharge.

i l Liquid Waste (summarized on a monthly basis) l l (1) Radioactivity discharged during the reporting period.

(a) Total radioactivity released (in curies).

(b) The MPC used and the isotopic composition if l greater than 1 x 10-7 microcuries/cc for fission and activation products.

(c) Average concentration at point of release (in microcuries/cc) during the reporting period.

l (2) Total volume (in gallons) of effluent water (including diluent) released during each r period of release.

i I Gaseous Waste (summarized on a monthly basis) l Radioactivity discharged during the reporting period (in curies):

l (a) Total estimated quantity of radioactivity I released (in curies);

l

! (b) ' Total estimated quantity of Argon-41 released (in curies) during the reporting period; (c) Estimated average atmospheric-diluted concen-tration of Argon-41 released during the reporting period in terms of microcuries/cc and fraction of the applicable MFC value; (d) Total estimated quantity of radioactivity in particulate form with half lives greater than eight days (in curies) released during the reporting period; s

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37

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(e) Average. concento 2cion of radioactive particu-

- lates with halt lives greater than eight days

-released in microcuries/cc during the reporting period; and (f) An estimate of'the average concentration of l-other significant radionuclides present in the gaseous waste discharge in terms of microcuries/cc and fraction of the; applicable

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MFC.value.for the reporting period if the estimated release la greater than 20% of the applicable MPC.

l' Solid Waste (summarized on an annual basis)'

1 l -(l) ' Total amount of solid waste packaged (in cubic feet)

(2) Total activity in solid waste (in curies)

(3) The dates of shipment and disposition (if shipped off site).

g. An annual summary of.the radiation exposure received by facility personnel and visitors in terms of- the average l radiation exposure per individual and greatest exposure

~

per individual in the two groups. Each significant ex-posure'in excess of the limits of 10 CFR 20 should be reported including the time and date sof the exposure as well as the name~of the individual.and the circum-stances leading up to the exposure;

h. An annual summary of.the radiation levels and levels

! of contamination observed during routine surveys performed at the-facility-in terms of the average and i

highest levels; and l  ;

i. An annual summary of any environnental surveys performed outside the facility. '

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