ML20195J156

Tech Specs for UTR-10 Reactor Facility at Iowa State Univ

ML20195J156
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Site:	University of Iowa
Issue date:	11/30/1988
From:	IOWA STATE UNIV., AMES, IA
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l TECHNICAL SPECIFICATIONS for the UTR-10 REACTOR FACILITY >

at IOWA STATE UNIVERSITY f

Docket No. 50-116 License No, R-59 l f

Original: August 1983 ,

Amendment 1: November 1988  ;

(Changes from 1983 original are highlighted in boldface) j I

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69 ' C '~ ' s f

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1.0 DEFINITIONS The terms Safety Limit, limiting Safety System Setting, and Limiting Condition for Operation are as defined in paragraph 50.36 of 10 CFR Part 50.

CHANNEL TEST - The introduction of a signal into the channel for verification that it is operable.

CHANNEL CAllBRATION The adjustment of the ch6hnel such that its output corresponds with acccotable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a Channel Test.

CHANNEL CHECK - A qualitative verification of acceptable performance by obsersation of channel behavior. This verification, where possible, shall include the comparison of the channel with other independent channels or systems measuring the same variable.

CONFINEMENT B0UNDARY - The surface surrounding the reactor facility defined by the interior partition walls of offices and laboratories on the north, east and south sides of the building and by the west interior wall which isolates the basement, first floor, and the west corridor of the second floor from the central bay.

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CONTROL R00 - A plate fabricated with Boral as the neutron absorbing material l which is used to establish neutron flux changes and to compensate for routine reactivity losses. This includes safety-type and regulating rods.

CORE - The portion of the reactor volume which includes the graphite reflector, core tanks, and control rods. The thermal column and shield tank duct are not included.

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DELAY TIME - The elapsed time between reaching a limiting safety system setpoint and the initial movement of a safety type rod.

DELAYED NEUTRON FRACTION - Yhen converting between absolute- and dollar value reactivity units, a beta of 0.00763 (changed from 0.00645) is used.

DROP TIME - The elapsed time between reaching a limiting safety system setpoint and the full insertion of a safety-type rod.

EXCESS REACTIVITY - That amount of reactivity that would exist if all control rods (control, regulating, etc.) were moved to the maximum reactive condition from the point where the reactor is exactly critical.

EXPERIMENT - Any operation, hardware, or target (excluding devices such as detectors, foils, etc.) which is designed to investigate non-routine reactor characteristics or which it intended for irradiation within the core region, on or in a beam port or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design.

MEASURED VALUE - The value of a parameter as it appears on the output of a channel.

MEASURING CHANNEL - The combination of sensor, line, amplifier and output devices which are connected for the purpose of measuring the value of a parameter.

MOVABLE EXPERIMENT - An experiment where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.

OPERABLE - A component or system is capable of performing its intended function.

OPERATING - A component or system is performing its intended function.

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REACTIVITY LlHITS - Those limits imposed on reactor core excess reactivity.

Quantities are referenced to a Reference Core Condition.

REACTIVITY WORTH OF AN EXPERIMENT - The maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.

REACTOR OPERATING - The reactor is operating whenever it is not secured or shutdown.

REACTOR OPERATOR (RO) - An individual who is licensed to manipulate the controls of a reactor.

REACTOR SECURED - A reactor is secured when:

(1) It contains insufficient fissile material or moderator present in the reactor to attain criticality under optimum availible conditions of moderation and reflection, or (2) A combination of the following:

a. The minimum number of neutron absorbing control rods are fully inserted or other safety devices i.re in shutdown position, as required by technical specifications, and

b. The magnet power keyswitch is in the off position and the key is removed from the lock, and

c. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and

d. No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or 0.763%

Ak/k (1.005) (unchanged from one dollar) whichever is smaller, 1-3

r REACTOR SHUTDOWN - The reactor is shutdown if it is subcritical by at least 0.763% Ak/k (1.00$) (unchanged from one dollar) in the Reference Core Condition and the reactivity worth of all experiments is accounted for.

REACTOR SAFETY SYSTEMS - Those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

READILY AVAILABLE ON CALL - Applies to an individual who:

(1) Has been specifically designated and the designation known to the operator on duty, and (2) Keeps the operator on duty informed of where he or she may be rapidly contacted (e.g., by phone, etc.), and (3) is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g., 30 minutes),

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! REFERENCE CORE CONDITION - The condition of the core when it is at ambient i temperature (cold)andthereactivityworthofxenonisnegligible,lessthan 0.23% Ak/k (0.30$) (unchanged from 0.30$).

i REGULATING ROD - A low-worth control rod used primarily to maintain an 4 intended power level that does not have scram capability. Its position may be

! varied manually or by the servo controller.

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SAFETY CHANNEL - A measuring or protective channel in the reactor safety system.

SAFETY-TYPE R0D - A rod that can be rapidly inserted by cutting off the l holding current in its electromagnetic clutch. This applies to safety #1, l safety #2, and shim safety.

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

SENIOR REACTOR OPERATOR (SRO) - An individual who is licensed to direct the activities of a Reactor Operator (RO) and to manipulate the controls or a reactor.

SHALL, SHOULD, AND MAY - The word "shall" is used to denoto a requirement, the word "should" to denote a recommendation, and the word "may" to denote permission, neither a requirement nor a recommendation.

SHUTDOWN MARGIN - The minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and safety systems starting from any permissible operating condition although the most reactive rod is in its most reactive position, and that the reactor will remain suberitical without further operator action.

TRVE VALVE The actual value of a parameter or variable.

UNSCHEDULED SHUTDOWN Any unplanned shutdown of the reactor caused by j actuation of the reactor safety system, operating error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or check out operations.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety limits 2.1.1 Applicability These specifications apply to the variables that affect thermal and hydraulic performance of the core.

2.1.2 Objective To assure fuel cladding integrity.

2.1.3 Specifications A. The true value of the steady power level under various flow conditions shall not exceed 15 kilowatts.

B. The true value of the primary coolant flow rate shall not be less than 3.5 gpm for periods greater than 5 minutes at all power levels greater than one kilowatt.

C. The true value of the primary coolant outlet temperature shall not exceed 180 0F 2.1.4 Bases Specifications A and B provide limits which protect the fuel cladding from damage due to excessive heat flux and surface temperature if the primary coolant pump fails. There is sufficient time for the operator to take corrective action before saturated pool boiling begins since the rate of temperature rise is approximately 9.8 O f per hour per kilowatt (SAR: 6.2); the l time to increase from the maximum allowable core inlet temperature of 160 Of to the boiling temperature when operating at 10 kilowatts would be i i

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approximately 32 minutes. Even if boiling did occur, the maximum critical i heat flux ratio (critical heat flux divided by the maximum heat flux in the  :

core) is so large (on the order of 1000) (changed from 500) that damage to the '

cledding would be very unlikely.  ;

Specification C provides a limit for core outlet coolant temperature under forced convection cooling. If the primary coolant flow rate was as low as 3.5 gpm and the core inlet temperature was 150 Of at 10 kilowatts, the  !

temperature rise across the core would be nearly 20 0 F As coolant I temperatures reach 180 0F (which is also the dump tank limit) and above, the corrosion rate increases, thus accelerating the loss of fuel plate cladding. i 2-2

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i 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS (continued)

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2.2 Limitina 3rfaly System Settinos  :

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! 2.2.1 Applicability 1 i i  !

! This specification applies to the setpoints of safety channels.

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! i 1 2.2.2 Objective  !

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To assure that automatic trip action is initiated and that the  ;

1 operator is warned to take protective action to prevent a safety limit from [

j being exceeded. '

j 2.2.3 Specifications l 4

The limiting safety system settings are the following:  !

l A. Maximum power level trip setpoint shall not exceed 12.5  ;

kilowatts.  !

B. Minimum primary coolant flow rate trip setpoint shall not be l less than 5 gpm.

l C. Maximum primary coolant outlet temperature trip setpoint shall l not exceed 170 0F f

2.2.4 Bases  ;

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The trip setpoints provide adequate margins for the limits specified  !

in 2.1.3. Trip setpoint A initiates automatic scram. Trip setpoints B and C  !

initiate alarms signaled by a horn and lighted annunciator. Operator  ;

i intervention in the non-scram trips provides timely response due to the slow variation of temperature even in the most adverse case discussed in 2.1.4.

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3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor fatt Parameters

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3.1.1 Applicability 1

l These specifications apply to the parameters which describe the reactivity condition of the core.

! 3.1.2 Objective To ensure that the reactor can not achieve prompt criticality and that 1

it can be safely shut down under any condition, j 3.1.3 Specifications l

l The reactor shall not be made critical unless the following conditions exist:

I A. The total core excess reactivity with or without experiments shall not exceed 0.50". Ak/k (0.65$) (changed from 0.78$).

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B. The minimum shutdown margin provided by control rodr. in the

reference core condition shall not be lets than 0.35% Ak/k l (0.46$) (changed from 0.62$).

! C. The fuel loading pattern and experiment apparatus inserted in the l core shall be approved by the Reactor Use Committee.

! 3.1.4 Bases 1

l Specifications A and B are based on values used in the power transient analysis (SAR: 6.3) where it is assumed that all of the excess reactivity, 3-1

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0.50% Ak/k (0.65$) (changed from 0.78$), is suddenly inserted as a positive step function. The safety system response is assumed to result in the minimum shutdown margin, 0.35% Ak/k (0.465) (changed from 0.62$), being supplied by  :

rapidly inserted safety type control rods, assuming the rod with the greatest worth is not available.

Specification C limits the changes in core configuration to those approved by the committee charged with review and approval of experiments.  ;

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3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.2 Reactor Control Lqd Safety System 3.2.1 Applicability These specifications apply to the reactor safety system and safety-related instrumentation.

3.2.2 Objective To specify the lowest acceptable level of performance or the minimum number of acceptable components for the reactor safety system and safety-related instrumentation.

3.2.3 Specifications The reactor shall not be made critical unless the following conditions exist:

A. The reactor safety channels and safety related measuring channels shall be operable in accordance with Table 3-1, including the minimum number of channels and the indicated maximum or minimum setpoints.

B. All three safety-type control rods shall be operable and have the following response time capabilities:

(1) Delay time shall not exceed 100 milliseconds.

(2) Drop time shall not exceed 600 milliseconds.

C. The reactivity insertion rate for a single control rod shall not exceed 0.019% Ak/k/sec (0.0255/sec) (changed from 0.03$).

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! I D. The dump valve shall be operable and shall be capable of reaching its normally opened position in not more than 600 i milliseconds after the scram signal is initiated. l I l E. The following bypasses may be applied to the channels indicated  !

provided the appropriate compensation is employed: l I' (1) During measurements of control rod worth, the startup l sequence for removal of safety type rods with no position f i indication may be altered if elapsed withdrawal times are (

!' observed as the rod that establishes criticality is  !

i i maneuvered. j i i i ,

(2) During measurements of reactor thermal power or control rod l worth, the signal from the multirange linear power channel (

neutron detector may be used exclusively for measurement j data recording if another detector of equivalent l characteristics is used as a substitute, i 2

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3.2.4 Bases r 1 i

! Specification A provides assurance that the reactor safety  !

) instrumentation channels which may be needed to shutdown the reactor are j operable. In addition, other channels which are important to safe operation

! because of interlock or alarm action are included. Each channel, along with the setpoint, minimum number required, and function, is listed in Table 31.

- The control rod withdrawal inhibit assures that the operator has an l

operable channel and appropriate neutron flux levels during startur..

- The integrity of the startup neutron source is protected, and excessive radiation levels are avoided by the coincident power and source / closure scrams.

- The period scram limits the rate of power level increase to values which are manually controllable without reaching excessive power levels or fuel 3-4

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temperatures.

-- The linear percent power scrams provide automatic protective action to [

prevent exceeding the safety limit (2.1.3 A) on reactor power.

-- The multirange linear power channel provides information to guide the  !

operator in establishing a set power level with greater precision than that <

available from other power level monitoring channels.  ;

-- The scram derived for the loss of high voltage to the neutron detectors l provides a conservative response to an instrumentation system failure. The ,

recommended operating voltage serves as the guide to detect a significant loss (

in power supply potential. (

- The alarm response to a fault in the scram circuit provides notice to the  !

operator that the scram bus may not be operable ifa subsequent fault develops.  !

l Operators are directed by prncedure to shut down the reactor when this alarm [

is noted.

-- The moderator level channel inhibits control rod withdrawal until the l mcderator reaches an appropriate level above the fuel plates during startup {

operation. This minimum level restricts variations in moderator level at J I

startup which could produce significant changes in reactivity balance and neutron detector response. (See also 3.3.4)

-- The moderator high level scram provides automatic shutdown and the }

subsequent draining of the moderator from the core tanks if the level exceeds  !

the setpoint. Accidental flooding of the graphite reflector and uncontrolled loss of coolant are avoided.

-- The shim safety position indicator channel must be operable to permit the operator to determine the excess reactivity from the critical rod position and rod calibration information.

-- The earthquake scram is provided to put the reactor in a shutdown condition before the protection system components are subjected to forcts which might make them inoperable.

- The manual scram and the magnet power keyswitch provide two methods for the reactor operator to manually shut down the reactor is an unsafe or abnormal condition should occur.

Specification B is based on values used in the power transient analysis (SAR: 6.3) where it is assumed that two safety-type control rods are 3-5

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inserted as a ramp function. The safety system is assurred to initiate rod l l

motion within 100 milliseconds after reaching the limiting safety system j setting and to have the rods full inserted within a total time (delay plus

! insertion) of 600 milliseconds.  !

i Specification C is based oil a conservative value used for many years l as a limit for reactors of the same type as the UTR 10. The limit assures a safe rate of power change during startup and during power ascensions. l Specification 0 assures that the moderator can be drained from the >

core tanks following a scram and provide backup shutdown action. When the dump valve opens in 600 milliseconds or less, the water s drained from the 6 core tanks in approximately 4 seconds. .

Specification E provides for bypasses of a startup interlock and a i normal instrumentation signal connection. I

-- The startup sequence requires a fixed order of safety rod removal: Safety [

1. 1 full out, safety #2 full out, then partial removal, depending on the excess l reactivity, of either the shim safety or regulating rods. To measure the l maximum worth of the shim safety, the startup sequence interlock may be bypassed to allow removal of a safety rod and then the shim safety. The remaining safety rod (neither safety rod is equipped with intermediate f position indication) can be safely maneuvered to the :ritical position by keeping cumulative withdrawal time. (

The normal connection of the multirange linear power channel can be bypassed with no reduction in the performance capability of the channel by  !

using another detector of equivalent characteristics, located in another but i comparable position with relation to the fuel region, as the signal source for [

power level information. The char.geover is completed at low power, and any [

change in calibration factor noted for later use at higher power levels. This [

bypass is used to obtain detector current information at high power levels for {

thermal power measurements and calibrations, and for conte ' rod worth l

measurements. L l

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l Table 3-1. Required Safety Channels and Safety-Related Channels.

Channel Setpoint Min. Operable Function NUCLEAR Log % power inhibits control Min. Counteate 20.1 mW 2 rod withdrawal.

Power level $1 W 2 Scram if all closures are not seated or source is not stored.

Period 25 seconds 2 Scram l Linear % power s12.5 kW 2 Scram Multirange linear power --

1 Power information.

High voltage loss to neutron detectors 290% V(a) 2 Scram Scram circuit failure Fault to gnd 1 Alarm PROCESS Moderator level (b) Inhibits rod Normal op level 242 inches 1 magnet current.

High level 555 inches 1 Scram i

Shim safety -- 1 Excess reactivity I position information.

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Earthquake 54 Richter 1 Scram MANUAL Manual scram switch --

1 Scram l

Magnet power keyswitch .. 1 Scram (a)Recomended operating voltage.

(b) Measured from the core tank base plate.

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3.0LIMITINGCONDITIONSFOROPERATION(Continued) 3.3 Coolant Systems 3.3.1 Applicability These specifications apply to the ninimum operating equipment and limits of operation for the cooling system.

3.3.2 Objective To ensure that the reactor fuel can be adequately cooled with water of l high quality.

3.3.3 Specifications j The reactor shall not be made critical unless the following conditions l exist:

A. The coolant system instrumentation channels shall be operable in accordance with Table 3 2, including the minimum number of channels and the indicated setpoints.

l l B. The primary coolant inlet tempe 4ture shall be maintained in the range from a lower value determined by specification 3.1.3 A to 160 0F , and the primary coolant outlet temperature shall not exceed 160 0F, C. The primary coolant flow rate shall be maintained in the range j from 5 to 15 gpm, except that the flow rate may be less than 5 gpm if the power level is less than one kilowatt and an approved experiment requires the reduced flow condition.

l D. The primary coolant temperature at the detonizer inlet shall not 3-8 I

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exceed 140 0F, and the detonizer flow shall be cut off until the l temperature is below the limit. I E. The primary coolant conductivity shall not exceed 2 micromhos per  !

centimeter, except for periods not to exceed 7 days when the l l value shall not exceed 10 micromhos per centimeter.

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f F. The radiation exposure rate observed by the detonizer column f

detector shall not exceed five times the nominal value measured during normal full power operation, or the exposure shall not  ;

j exceed 10 milliroentgens in one hour, whichever is the smaller i

{ value.  !

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, G. The net detection rate of confirmed fission product activity in the primary coolant shall not exceed the "decision" limit for j the detection system used in the analysis. I l

1 3.3.4 Bases

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) Specification A provides assurance that the cooling system  !

) instrumentation channels are operable. Each channel, along with the setpoint, f

minimum number required, and function is listed in Table 3 2.

l - The moderator (primary coolant) level channel inhibits control rod l withdrawal untti the moderator reaches an appropriate level above the fuel .

plates during the startup operation. This minimum level interlock ensures

ample coolant level to provide heat transfer for the fuel plates, and it also j restricts the variations in moderator level at startup which could produce  ;

significant changes in reactivity balancu and neutron detection rates (SAR: [

] 4.5.3).

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- The primary coolant inlet temperature channel permits compliance with [

4 specifications 3.1.3 A and 8 by initiating a low level alarm and providing the operato with information to establish a minimum coolant temperature which i avoids an excessive reactivity inventory.

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-- The primary coolant outlet temperature channel initiates an alarr: signal 3g i t

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at the high temperature setpoint of 160 0F, This provides an adequate margin j to avoid the the safety limit specified in 2.1.3 C.

- The primary coolant flow rate channel initiates a low flow alarm to warn i

] the operator to reduce power in compliance with safety limit 2.1.3 B, if the (

i power level is 4t or aoove one kilowatt (SAR: 6.2).

j The primary coolant conductivity channel initiates an alarm when the j specific conductance exceeds 2 micrombos per centimeter. Operation may f j continue at a higher level for a limited time as indicated in 3.3.3 E. [

J - The radiation equipment detector located near the deionizer initiates an j alarm when the exposure rate exceeds five times the nominal value observed during normal full power operation (See 3.7.4). j

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i Specification B is based on values of primary coolant temperature l which must be maintained to avoid violating the limit on excess reactivity {

(3.1.3 A) at the lower end of the range, and to avoid the high temperature  !

! safety limit 2.1.3 C (180 0F) which also is the limit on the dump tank (SAR:  !

4.3.2). f Specification C provides a range on the primary coolant flow rate I which will adequately cool the fuel plates and avoid safety limit 2.1.3 B, and f also provide flexibility for low power experiments which may require an f essentially stagnant coolant. It incorporates, through its lower limit of 5 gpm, an implied coolant leak detection provision since a significant loss of l

1 primary coolant (which is :1 eld in the process pit until analyzed) reduces the

{ suction head on the pump to the point where the minimum flow rate cannot be i maintained.  !

4 t Specification D provides a limit to prevent damage to the deionizer resins and possible transport of fractured resin beads past the filter and [

into the primary coolant stream. The flow through the detonizer will have to [

be restored at a temperature below the limit if the conductivity limit is  !

l approached, r l Specification E is based on experience at many facilities with similar l l coolant systems; this value is known to be a satisfactory upper limit for l normal operations. Trace mineral activation products do not exceed acceptable j limits and corrosion rates are negligibly low when the upper limit is not [

exceeded (SAR: 4.2.2, 4.5.3 and 6.1.2). Provision for conductivity transients f i

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t due to crud releases adds flexibility to the limit.

Specification F is based on the assumption that the increase in f

exposure level is due to either fission product activity or radioactive trace f

) minerals normally present in the primary coolant being concentrated in the l

deionizer column. The trip setpoint is based on local conditions and must be ,

determined so that it detects significant activity with respect to normal f detection rates without causing too frequent false alarms. Since the detonizer  !

j column is located near the boundary of the restricted area, the 1/ mR/h upper }

4 limit provides a conservative margin to avoid exceeding the requirements of i

paragraph 20.105 of 10 CFR Part 20 on radiation doses in unrestricted areas.

f Specification G provides a limit based on statistical hypothesis j I testing and it depends on the detection system being used to evaluate the  !

coolant sample. The term used in NCRP Report No. 58, pp. 275 279, is the

! "decision limit", and it can be used to det e line if the net detection rate of i the sample is statistically different from backcound at a confidence level of f i 95%, when equation (7.8) is used. The background in this case is taken to be  !

] the detection rate of samples without fission product activity. When the f I sample detection limit does exceed this limit, the leaking fuel assembly must be identified and removed from the reactor (see 3.7.3 C).

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) Table 3 2 Required Coolant System Instrumentation Channels. l l

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Channel Setpoint Min. Operable Function i

- t Moderatorlevel(a)  !

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Normal op level 142 inches 1 Inhibits rod t magnet current; I establishes  !

minimum coolant  !

level. l

! Primary coolant inlet as required to 1 Thermal power [

temperature satisfy 3.1.3 information l Primary coolant outlet temperature 5160 0F 1 Alarm ,

Primary coolant flow rate 25 gpm 1 Alarm

) Primary coolant

conductivity 52 micromhos/cm 1 Alarm j i As required to Radiatier level 1 Alarm j i Detonizer unit satisfy 3.3.3 F j l

l (a) Measured from the core tank base plate.

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3.0 LINITING CONDITIONS FOR OPERATION (continued) l 3.4 Confinement 3.4.1. Applicability l This specification applies to the operations that require confinement l

and to the equipment needed to achieve confinement, j 3.4.2 Objective i l

i To ensure that the confinement boundary can be secured when needed.

l 3.4.3 Specifications

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A. The reactor confinement boundary shall be operable whenever the reactor is operating.

B. The reactor confinement boundary shall be secured during fuel [

transfer operations, f I

3.4.4 Bases  !

Specification A is based on the assumption that the doors and windows l located in the building walls that define the confinement boundary may need to be secured dus to the accidental release of radioactive material generated during reactor operation.

l Specification B is based on the hypothetical accident (SAR: 6.4) that (

occurs during movement of a fuel assembly and the importance of hating the f

confinement boundary secured prior to the fuel transfer operation, j 3 - 13

3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.5 ygntilation lyittmi There is no forced air circulation system in the reactor room or the building housing it.

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i J i 3.0 LIMITING CONDITIONS FOR OPERATIfiN (Continued)  !

I 3.6 Emergency EggttI 1

3.6.1 Applicability

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, These specifications apply to the emergency power supply for the radiation monitoring system, j 4  ?

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, 3.6.2 Objective l

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I To specify the source of emergency electrical power and the minimum I operating time.  !

I 3.6.3 Specifications I

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] The reactor shall not be made critical unless the following conditions ,

i exist:  !

r I I I A. The battery powered standby AC power supply for the radiation monitoring system shall be operable and shall have the following

! operating time capabilities:

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(1) Operating time without the radiation evacuation horn being I activated shall be not less than eight hours, f

! (2) Operating time with the radiation evacuation horn being j activated shall be not less than two hours.

3.6.4 Bases i

l Specification A requires that the standby AC power system, which i consists of at least two lead acid strrage batteries, a charger transfer unit, 1

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and an inverter, be capable of providing a tripless switchover for supplying AC power to the radiation monitoring systems, and that the power source be able to sustain operation for the specified intervals. There are no systems, other than radiation monitoring, that need emergency power since the reactor is automatically shutdown when AC power failure occurs. The radiation evacuation horn imposes a large incremental load on the power source and severely reduces the operating time; however, the evacuacion signal, if needed, would be of sufficient duration to accomplish its intended purpose, l

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l l 3.0 LIMITING CONDITIONS FOR OPERATI0l' i';ontinued)  !

i 3.7 Radiation Monitorina Systems And Effluents f l

i 3.7.1 Applicability l

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i l These specifications apply to the radiation monitoring systems and to l

l the limits on effluent releases, i i

i 3.7.2 Objective  :

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To specify the minimum number of acceptable components or the lowest acceptable level of performance for the radiation monitoring systems and the ,

Ilmits for releases of effluents.

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3.7.3 Specifications k The reactor shall not be made critical unless the following conditions l l

exist: [

A. The radiation monitoring channels and components shall be l operable in accordance with Table 3 3, including the minimum {

number of channels or cogonents, and their setpoints.

B. The cumulative energy production of the reactor shall not exceed 4760 kilowatt.nours in any twelve month interval nor exceed 100 kilvwatt hours in any 7 day interval to limit the generattun and j l release of argon 41.

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C. If evidence exists that the Itait in 3.3.3 G will be exceeded, f the reactor shall be shutdown and the leakage source found and f eliminated however, the reactor may be operated intermittently  :

to assist indetermining the source of leakage. ,

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3.7.4 Bases Specification A provides assurance that the required radiation monitors are operable.

-- The air-particulate monitor is placed in service and operated continuously when designated experiments are being performed, viz., those which could produce airborne radioactivity. The alam setpoint is influen ed by the normal background reading while the reactor operates at the required power level and is based on the same reasoning as given for the deionizer monitor setpoint.

-- The radiation detector located near the deionizer initiates an alarm when the exposure rate exceeds five timos the nominal value observed during normal full power operation. The trip value is sufficient for significant radiation events, yet not too sensitive to produce frequent false alarms. (See also 3.3.3 F.) This monitor would be the first to sense a release of fission products into the coolant.

-- The radiation area monitors are placed aroand the perimeter of the reactor room. All four units are able to initiate an alarm signal at or above 5 mR/h I whenever the reactor console is energized. The south and west units initiate a radiation evacuation alarm at or above 50 mR/h when the reactor is in operation; when the console is not energized, the radiation-evacuation setpoint is 5 mR/h. The 5 mR/h limit is based on the minimum value permitted for criticality monitoring of SNH in storage and applies when the area is unattended, while the 50 mR/h limit is based on the radiation level associated with the emergency action level for the alert classification.

-- The doorway radiation monitor serves as a frisker to detect abnormal levels of radiation when a person passes the detector. The increasing aural signal alerts the reactor operator and the affected individual that further assessment must be initiated.

-- The radiation film badge (or its equivalent) provides radiation dose information at the perimeter wall of the reactor room and serves as a control for the film badges used by personnel in the restricted area.

Specification B provides a conservative limit on the generation and release of argon 41 and is based on measurements at this facility (SAR:

3 - 18

4.5.4). Argon-41 is the only significant radioactive effluent produced during normal operation of the reactor, and the limits prov4ded meet the requirements of paragraphs 20.103 and 20.106 of 10 CFR Part 20. The first part of specification B is based on the assumption that the reactor operates continuously at 10 kW for 476 hours0.00551 days <br />0.132 hours <br />7.87037e-4 weeks <br />1.81118e-4 months <br /> and that the dilution factor from diffusion of the air in the enclosure is only 10; for these conditions, the argon-41 concentration averaged over one year is about 50% of the value listed for unrestricted areas in Table II, Appendix B of 10 CFR Part 20. The 1econd part of specification B uses the assumptions that the reactor operates continuously at 10 kW for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> for one 40-hour week; these conditions yield an average concentration in the enclosure of 50% of the value listed for restricted areas in Table 1, Appendix B of 10 CFR Part 20.

Specification C allows a search for a leaking fuel element to be conducted by using the reactor to the extent needed to detect the source of fission products.

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Table 3-3. Required Radiation Monitoring Channels or Components.

Channel Setpoint Min. Operable Function i

Air-Particulate (a) l unit As required 1 Alarm Deionizer(b) unit As required to 1 Alarm satisfy 3.3.3 F Areaunits(c)(d) 5(50) mR/h 4 Alarm t

~

Doorway monitor --

1 Warn of abnormal radiation level.

Environmental 1 Integrated dose in res-Film badge or equivalent -- restricted area (a)This unit is activated whenever designated experiments are being performed.

(b)This unit serses as the fission product monitor as specified in 3.3.3 F.

(c)When either the north or east area monitoring units are inoperable, portable instruments may be substituted for periods up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(d)The normal setpoint is shown. The parenthetical value is the maximum setpoint to be used depending on local conditions. Use of higher than normal setpoints requires approval of the Reactor Manager. The south and west units monitor the fuel storage area and are reset to the normal value after reactor shutdown.

3 - 20

3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.8 Exoeriments 3.8.1 Applicability These specifications apply to the experiments installed in the reactor and its experiment facilities.

3.8.2 Objective To prevent damage to the reactor and excessive release of raoloactive material in the event of experiment failure, and to avoid exceeding any safety limit.

3.8.3 Specifications Experiments installed in the reactor shall meet the following i conditions:

A. Prior to initiation, each type of experiment utilizing the reactor shall be approved by the Reactor Use Committee.

8. Operational limits peculiar to an experiment shall be included i in instructions to the reactor operator.

C. The reactivity worth of any single experiment, or group of experiments, installed in the core shall be limited to -0.48%

Ak/k (-0.63$) to +0.14% Ak/k (+0.19$) (changed from -0.75$ to i

+0.22$).

D. Significant amounts of special materials used in experiments, including fissionable material, explosives or metastable materials capable of significant energy release, or materials 3 - 21 F

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l that are corrosive to reactor components or highly reactive with the coolant, shall conform to established special requirements.

E. Credible failure of any experiment shall not result in releases or exposures in excess of established limits nor in excess of the annual limits established in 10 CFR Part 20.

F. Experiments shall be designed so that they will not contribute to the failure of other experime.nts, core coniponents, or principle physical barriers to uncontrolled release of radioactivity.

Also, no credible reactor transient shall cause an experiment to fail in such a way as to contribute to an accident 3.8.4 Bases Specification A, B, 0, E, and F are based on requirements stated in the standard for The Development of Technical Specifications for Research Reactors, ANSI /ANS-15.1-1982.

Specificatior. C is based on the effect of the failure on an experiment, or group of experiments, on the reactivity of the reactor. In the case of an experiment failing with +0.14% Ak/k (+0.195) (changed from +0.22$)

inserted as a step function, the resulting period would be 30 seconds, which can be easily managed with control rod movement. If the -0.48% Ak/k (-0.63!)

(changed from -0.75$) step insertio, occurs because of experiment failure, the ,

reactor excess reactivity would drop to 0.02% Ak/k (0.026$) (changed from 0.03$), an amount sufficient to maintain criticality and to continue cperation i if necessary.  !

1 3 - 22

r 4.0 SURVEILLANCE REQUIREMENTS Surveillance tests, except those specifically required for safety when the reactor is shut down, may be deferred during reactor shutdown; however, they must be completed prior to reactor startup.

4.1 Reactor Q Ig Parameters 4.1.1 Applicability These specifications apply to the surveillance activities required for reactor core parameters.

4.1.2 Objective To specify the frequency and type of testing to assure that reactor core parameters conform to the specifications of section 3 of these Specifications.

4.1.3 Specifications A. The excess reactivity shall be measured at least annually and following significant core or control rod changes.

B. The shutdown margin shall be measured at least annually and following significant core or control rod changes.

4.1.4 Bases The measurements required in specifications A and B are sufficient to provide assurance that the reactor core parameters are main ained within the specifications 3.1.3 A and B since the fuel burnup rate is extremely low and important changes in the core parameters can be detected on a timely basis.

4-1

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Core or control rod changes are not considered to be complete until the excess

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reactivity and shutdown margin measurements are finished.  !

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4 4.0 SURVEILLANCE REQUIREMENTS (Continued) l 4.2 Reactor Control and Safety System 4.2.1 Applicability These specifications apply to the surveillance activities required for the reactor control and safety system, 4

4.2.2 Objective ,

To specify the frequency and type of testing or calibration to assure that the reactor control and safety system conforms to the specifications of section 3 of these Specifications.

! 4.2.3 Specifications 4

A. The reactivity worth of each control rod shall be measured at least i

annually and whenever the core configuration is changed by fuel assembly replacement or rearrangement.

B. The drop time and delay time of each control rod, except for the regulating rod, and the withdrawal time of each control red shall be measured at least annually and whenever any maintenance on a l control rod which may affect its motion is completed.

C. The operability of the control rods and the dump valve shall be

! tested daily when the reactor is operating.

D. An operability test, including trip action, of each safety channel listed in Table 3-1 that provides a scram function shall be completed prior to each reactor startup following a period when the reactor is secured in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> c'. at least weekly during continuous operating periods.

4-3

E. A calibration of the channels listed in Table 3-1 that can be calibrated shall be performed at least annually and whenever any maintenance on h channel which may affect its performance is completed.

F. The thermal power output of the reactor shall be measured at least annually.

4.2.4 Bases Specification A requires rod worth measurements for all rods at annual intervals; no significant changes in the worths of the contrni rods are likely to occur during that time. Since changes in the fuel loading in the vicinity of a control rod may cause a significant change in its worth, a measurement after the fuel change is appropriate.

The rod drop and delay time measurement intervals required in specification B verify the limits in specification 3.2.3 B and are appropriate to detect abnormal performance as can be shown by experience at this facility.

Withdrawal time measurements provide data to determine if specification 3.2.3 C is being violatea.

Specification C verifies the operability requirements in specification 3.2.3 8 and D during each day of uperation.

In specification D each channel capable of generating a scram signal is tested during the precritical procedure, prior to startup, so that the conditions of specification 3.2.3 A are satisfied.

Specification E requires calibration of safety and safety-related channels at an interval which is appropriate and justified by experience at this facility.

The annual verification of reactor thermal power output, as required by specification F, is appropriate and justified by experience at this facility.  ;

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4.0 SVRVEILLANCE REQUIREMENTS (Continued) 4.3 Coolant Systems 4.3.1 Applicability These specifications apply to the surveillance activities required for  :

the reactor coolant system. I 4.3.2 Objective To specify the frequency and type of testing or calibration to assure that the reactor coolant system conforms to the specifications of section 3 of these Specifications.

4.3.3 Specifications A. The coolant system instrumentation channels listed in Table 3-2 shall be calibrated at least annually and whenever any maintenance on a channel which may affect its performance is completed.

B. The primary coolant temperature, flow rate, conductivity, and -

radiation level at the deionizer shall be measured and recorded at t

startup and at least every four hours when the reactor is [

operating. '

C. A primary coolant sample shall be analyzed for radioactivity at least quarterly and whencver the exposure rate at the deionizer exceeds the limits of specit'ication 3.3.3 F.

4.3.4 Bases

Specification A requires calibration of the coolant system instrumentation channels at an interval which is appropriate and justified by 45

experience at this facility.

Specification B requires verification of the operating limits of specifications 3.3.3 8 - F at an interval which is appropriate and justified by experience at this facility.

Specification C relates to the mu .i.oring for fission products and other activated materials in primary c',lant samples. Experience at this facility shows that the sampling in', eval is appropriate.

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46

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1 4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.4 Confinement 4.4.1 Applicability

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This specification applies to the surveillance activities required for the reactor confinement.

4.4.2 Objective i To specify the frequency and type of testing to assure that the l reactor confinement conforms to the specifications of section 3 of these Specifications, i 4.4,3 Specification A. The doors and windows in the confinement boundary shall undergo i testing for normal closure at least once every quarter. '

4.4.4 Bases 1

This specification requires that the doors and windows in the confinement boundary be tested to verify that they can be closed when needed.

The testing interval is adequate to verify operability based on experience at  !

this facility.  !

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u c 4.0 SURVEILLANCE REQUIREMENTS (continued) 4.5 Ventilation Syjirjag This specification does not apply to this facility.

4-8

4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.6 Emergency E ntr 4.6.1 Applicability These specifications apply to the surveillance activities required for the emergency power system.

4.6.2 Objective To specify the frequency and type of testing to assure that the emergency power system conforms to the specifications of section 3 of these Specifications.

4.6.3 Specifications These surveillance activities are required for safety when the reactor is not being operated.

A. The battery-powered AC standby power supply shall be tested for switch-over action, and for voltage and specific gravity characteristics at least quarterly.

B. The batteries shall be tested for full discharge at least every three years.

4.6.4 Bases Specification A rrquires verification of operability of the standby power supply to complete the switch over from normal AC power to the batteries at an interval which is appropriate based on experience at this facility. The measured values of voltage and specific gravity give adequate warning of reduced battery performance within the testing interval.

49

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l A full discharge test of the batteries every three years, as required in specification B, is appropriate for the type of battery used in the power supply; the interval is well within the normal 4-5 year warranted life for conditions much more severe than those encountered in this application.

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4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.7 Radiation Monitorin_g System and Effluents 4.7.1 Applicability  !

These specifications apply to the surveillance activities required for f the radiation monitoring system and effluents released from the facility.

4.7.2 Objective I

To specify the frequency and type of testing to assure that the radiation monitoring system and effluent releases conform to the l specifications of section 3 of these Specifications.

4.7.3 Specifications These surveillance activities (except E) are required for safety when the reactor is not being operated.

A. A calibration of the channels listed in Table 3-3 that can be

, calibrated shall be performed at least annually and whenever any maintenance on a channel which may affect its performance is t completed. '

B. An operability test, including source checks, of the radiation monitoring channels listed in Table 3-3 shall be performed at least  :

j monthly.

C. The radiation levels at the area and deionizer units shall be j measured and recorded at startup and at least every four hours v: hen  ;

the reactor is operating.

O. The environmental film badge cited in Table 3-3 and smeer sur/eys l

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in and around the reactor enclosure shall be analyzed at least ,

quarterly.

E. The cumulative energy conversion shall be computed and recorded at  :

least quarterly, and it shall be computed on a weekly basis to ,

monitor short-term argon-41 releases.

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I 4.7.4 Bases j Based on experience at this facility and the average usage pattern of {

the reactor, specifications A-D are adequate to verify that the operations  !

conform to the specifications of 3.7.3.

Specification E requires verification that the cumulative energy f

j conversion limit of specification 3.7.3 8 is not exceeded; this is an indirect [

, method of monitoring the generation and release of argon 41. j 1

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4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.8 Exoeriments 4.8.1 Applicability 4

These specifications apply to the surveillance activities required for experiments installed in the reactor.

4.8.2 Objective To specify the frequency and type of testing to assure that the experiments conform to the specifications of section 3 of these Sperifications.

4.8.3 Specifications A. The identification and location of all installed experiments shall be recorded prior to each reactor startup.

B. Other specific surveillance activities shall be established during l the review and approval process specified in section 6.

f 4.8.4 Bases j

Specification A requires that the reactor operator verify that the installed experiments are approved on a most conservative frequency basis.

Specification B recognizes that detailed surveillance requirements will vary among experiments, and that the experiment review comittee specifies the appropriate type and frequency of surveillance.

4 - 13

i 5.0 DESIGN FEATURES 5.1 lilA And Facility Descriotion The reactor is housed in the Nuclear Engineering Laboratory, which is located on the west edge of the main campus of Iowa State University, in Ames, Iowa. The Nuclear Engineering Laboratory is a two-story, three-level building ,

of brick construction, built in 1934. The reactor, a model UTR-10, was ir, stalled and first operated in 1959. It is fueled with uranium enriched to approximately 19,75% (changed from 92%) in the U 235 isotope, moderated and cooled with light water, reflected with graphite, and operates at a maximum [

thermal power of 10 kilowatts. The reactor is located on the ground floor level, central bay area of the Laboratory structure. The central bay is  !

approximately 34 feet high and has a floor area of 37 feet by 56 feet of which I a space approximately 37 feet by 38 feet is allocated to the reactor. A wall surrounding this area is constructed of standard concrete block and reaches a height of 10 feet 4 inches on the north, east and south sides; the west boundary is a wall that reaches from the floor to the ceiling of the central bay region. The purpose of these walls is to limit access of unauthorized 7 personnel to the immediate vicinity of the reactor and to define the outer perimeter of the restricted area.

The enclosure surrounding the reactor includes the central section of l the building as defined by the interior partition walls of offices and  ;

laboratories on the north, east and south sides of the building and by the west interior wall which isolates the basement, first floor, and the west corridor of the second floor from the central bay. The purpose of the  !

enclosure is to act as as a confinement volume and to help limit the release i of radioactive materials to the environment. The enclosure volume is slightly l 1ess than 2500 cubic meters, and the average infiltration rate for the ,

building is estimated to result in two changes per hour. There is no contral  !

forced air circulation system in the building. l The enclosure has two outside doors, one in the east wall and a large {

overhead door opening to the south. All interior doors leading into the enclosure are of a standard type used in interior construction. Other  !

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5-1 1

significant penetrations into the enclosure consist of roof-level windows on the north and south sides which can be manually opr.ned or closed, as a group per side, in less that one minute per group.

52

5.0 DESIGN FEATURES (Continued) 5.2 Reactor Coolant System In normal operation, the primary coolant is pumped (18 psig) from the dump tank (capacity 220 gallons) through the heat exchanger (10 gpm, 80 0F) to the bottom of the core tanks, upward past the fuel plates, to the overflow pipe manifold and returned to the dump tank (0 psig. 87 Of at 10 kW and 10 gpm); approximately 92 gallons are contained in the piping and core tanks during operation. A quick-opening dump valve in the feed line to the core tanks is provided to allow draining of moderator (coolant) following a scram.

A low pressure (5 psig) steam heater and controller system for the dump tank and a deionizer/ filter system for the purification loop, which operates (1 gpm, <140 0F) in parallel with the main loop are provided (Ref: Drawing R1-D 130). The operating temperature may range from about 80 Of to no more than 160 0 F, with the lower end preferred to reduce the corrosion of aluminum.

Moderator level, inlet and outlet temperature, flow rate and conductivity sensors are installed at appropriate locations and connected to the process instrumentation system (Ref: Drawing R1-D-ll6). The primary coolant system is essentially all-aluminum in construction; the pump casing and impeller, some valve parts, the dump tank heater element, and process instrumentation sensor elements in contact with the water are stainless steel or similar corrosion-resistant materials.

The energy transferred through the heat exchanger is dissipated to the building sewer system by once-through cooling water obtained from the campus water main. Secondary cooling flow is induced by water main pressure, and the flow rate is set by a motor operated valve to control the amount of cooling in the heat exchanger resulting in core inlet temperature control. To prevent secondary water from entering the primary system if a tube-leak should occur, a pressure differential is maintained in the heat exchanger to allow primary water to enter the secondary system.

The process pit accommodates the equipment and instrumentation sensors for the process system (Ref: Drawing Rl-E 151). A sump, with a capacity of 9.5 gallons and a manually energized sump pump, can discharge liquids from 53

the process pit to another sump located in the basement floor. The basement sump also receives secondary coolant outlet water and acts like a dilution tank; it has a capacity of 123 gallons. Outflow from the basement sump passes through an overflow pipe connected to the building sewer system.

54

5.0 DESIGN FEATURES (Continued) t 5.3 Reactor Core. LLtl And Safety System Core A graphite reflector surrounds the core tanks, except a water

reflector of no less than 13 cm thickness is maintained above the fu

1 assemblies during reactor operation. The composition of the region between the core tanks (the coupling region) can be changed by removal of graphite blocks and insertion of other materials, and small-volume experiments can be ,

placed in the water gap between plates in the fuel assembly, or in the water i reflector above or beneath the fuel, subject to specifications 3.8.3

)

(Experiments). A rabbit tube, no larger than 10 cm outside diameter, [

penetrates the graphite reflector at the west face of the north core tank. A '

neutron source, providing a minimum of 1.0 E+6 neutrons /second is inserted into the coupling region by means of the source positioner during the startup i

operation (Ref: Drawings Rl-E-154 and RI-E-161).

Fuel  ;

4 Reactor fuel is contained in aluminum clad flat plates, similar to argonaut-type fuel. Fuel meat is U3Si2, enriched to 19.75% in the U 235

! isotope, dispersed in aluminum to achieve a uranium density of 3.47 g/cc. The fuel meat, 0.51 mm thick, is clad with 0.38 mm aluminua. Each fuel plate contains 12.5 grams of U 235. Thc core contains 12 assemblies each with ,

approximately 24 fuel plates depending upon the measured critical i configuration. Solid aluminum plates and assemblies with missing feel plates, ,

for experimental purposes, are used to adjust the core fuel loading for the I licensed excess reactivity of 0.50% Ak/k (0.655). (Ref: Drawing Rl A 121 1) l l

changed from:

Reactor fuel is contained in aluminum clad flat plates, similar to  !

argonaut-type fuel. Fuel meat is uranium enriched to approximately 92% in U-  !

5-5  ;

l 235 isotope contained as a solid solution nf uranium-aluminum in aluminum.

This matrix is clad with 0.508 m (0.020 inch) aluminum metalurgically bonded to the fuel meat. Each fully-loaded fuel plate contains no more than 23 grams of U-235. Half- and quarter-load fuel plates and solid aluminum plates are used to adjust the core loading (Ref: Drawings R1-C-121 R1-C-122, R1-C-236, and R1-C-147). A standard fuel element consists of twelve plates, or if a plate is omitted a spacer equal in thickness to a fuel plate is used it contains no Dore than 270 grams of U-235. (Ref: Drawing RD-D-131). Fuel elements are loaded six to each core tank, subject to specification 3.1.3 A or 3240 grams U-235 whichever is smaller (Ref R1-D-133).

Safety System Four Boral control rods, two safety, one shim safety, and one regulating, are positioned in the graphite external reflector adjacent to the outside face and near each outside corner of the core tanks assembly (Ref:

Drawings R1-R-212, R1-R-213, R1-R-214). Each control rod is connected by a stainless steel flat spring to a motor driven drum. Each safety rod is coupled to its drive mechanism by an electrically energized magnetic clutch.

Two safety rod drives have limit switches with console indicaters showing full withdrawal and full insertion. Shim saiety and regulating rod positions are disp 1 dyed on the control censole.

The moderator level measuring channel provides a signal (interlock) which permits control rod drive magnets to be energized only after a minimum moderator level setpoint is exceeded, and it provides a signal (scram) when the moderator level exceeds the high level setpoint.

A neutron Jensitive power level measuring channel with a functional range of 1. E-7 to 1.5 E+2 percent pcwer, based on 10 kilowatts thermal, provides a signal (interlock) which prevents withbwal operation of the control rod drive motors if the minimum power level (minimum count rate) setpoint is not exceeded, a signal (scram) when the one watt level is exceeded, and the neutron startup source is not in its storage position or all closures (two operating closures above the fuel, and one at the end of the thermal column) are not properly seated; this channel provides a signal to 56

the period channel to generate a signal (scram) when the period is less than the short period setpoint. These signals are derived from the log percent power channel.

A neutron sensitive power level measuring channel, with a functional range of 10 to 150 percent of 10 kilowatts, provides a signal (scram) when a high power level setpoint is exceeded. This signal is derived from the linear percent power channel, i

57

5.0 DESIGN FEATURES (Continued) 5.4 Fissionable Material Storaae Fueled experiments and fuel devices not in the reactor are stored in a dry fuel storage pit w nitored by radiation and intrusion detectors (Ref:

Drawing R1-E-194 and Physical Security Plan). The fuel storage array, under all conditions of moderation and reflection with light water, has an effective multiplication factor less than 0.9.

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I 6.0 ADMINISTRATIVE CONTROLS -

6.1 Oraanization 6.1.1 Structure I The organization for the management of the reactor facility shall be structured as indicated in Figure 6 1. Job titles are shown for illustration and may vary. Levels of authority indicated divide responsibility as follows: ,

i Level 1: Responsible for the facility license and site administration Level 2: Responsibic for the reactor facility operation and management Level 3: Responsible for daily operations.

The Reactor Use Committee is appointed by, and shall report to the University Radiation Safety Committee. Radiation safety personnel shall report to Level 2 or higher through an independent organizational channel.

i f.l.2 Responsibility The Executive Officer, Department of Nuclear Engineering, shall be j responsible for the facility license and site administration. l Individuals at the various management levels shown in Figure 6-1, in l addition to having the responsibility for tha policies and operation of the  !

faciliff, shall be responsthie for safeguarding the public and fMility l peisonnel from undue radiation uxposures and for adhering to all requirements [

of the Operating License and the Technical Specifications.  !

In all instances, responsibilities of one level may be assumed by f designated alternates, or by higher levels, conditional upon appropriate j qualifications.

I 6.1.3 Staffing I

r (1) The minimum staffing when the reactor is not secured shall be: l 61 l

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a. A licensed reactor operator in the control room.

b. A licensed senior reactor operator readily available on call,

c. A health physics-qualified individual readily available on call. l (2) Events requiring the direction of a senior reactor operator: l i

i

a. Recovery from unplanned or unscheduled shutdown (in this instance, documented verbal concurrence from a SRO is required).

b. Fuel transfer operations.

1 c. Any mainten, trice activity involving the reactor safety system that l l could cause : significant increase in the reactivity of the  !

! reactor.  !

d. Relocation cf any in-core experiment with a reactivity worth I

greater than 0.763% Ak/k (1.00$) (vachanged from one dollar), i (3) Events requiring the presence of a health physics qualified j individual:  ;

i 1

a. Fuel transfer operations.  !

i b. Installation, changing locations, or removal of an experiment that I involves removal of a shield olug or closure. I i c. Any maintenance activity invoising the rt.act',t ufety system that l l could cause an abnormal releast of radiaactis materials. l 1

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! 6.1.4 Selection and Training of Personnel l

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The selection, training and requalification of operations personne) l

shall meet or exceed the requirement of American National Standard for f

Selection and Training of Personnel for Research Reaccors, ANS!/ANS 15.41977, l l or its successor, and be in accordance with the Requalification Plan approved

{ by the Nuclear Regulatory Comission.  ;

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i 6.0 ADMINISTRATIVE CONTROLS (Continued) 6.2 Ceview And Audit The Reactor Use comittee (RUC) shall perform the independent review and l audit the safety aspects of reactor facility operations.  :

6.2.1 Composition and Qualifications The Reactor Use Comittee shall be composed of the Reactor Manager and a radiation health physicist, both ex officio (voting), and at least three other members having expertise in reactor technology. Comittee members shall r be appointed by the University Radiation Safety Comittee. (The Radiation '

Safety Comittee is composed of a representative from each of five colleges in ,

the university in which research in the physical and life sciences and in engineerii,9 is conducted, plus three members with specific expertise in I radiation protection. At least one of these members shall also represent [

university management. The college representatives are chosen from the ,

l Colleges of Agriculture, Engineering, Sciences and Humanities, Home Economics,  !

I and Veterinary Medicine. One of the three other members shall be the [

University Radiat;on Safety Officer (RS0). The chair of the comittee shall j be appointed by the Vice President for Academic Affairs. The terms on the l comittee for the RSO and chair are indefinite. All others are or three l years with reappointments being determined by the Vice President for Academic  !

Affairs.) f l

6.2.2 Charter and Rules

, (1) The Reactor Use Comittee shall meet at least semiannually and mors  !

frequently as circumstances warrant, consistent with effective monitoring of facility activities. Written records of its meetings f

j shall be kept and copies forwarded, in a timely manner, to the j University Radiation Safety Comittee. {

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i (2) A quorum shall be three members. Members of the operation staff i shall not be a voting majority.

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(3) Any action recomended by the Reactor Use Comittee that may l adversely affect the operations and/or safety of the University comunity shall be reported by the RUC chairman to the University j Radiation Safety Comittee which shall have veto power over such a {

recommendation, f

(4) The Reactor Use Committee may appoint one or more qualified )

individuals t erform the audit function.  !

i 6.2.3 Review Funct-i The following items shall be reviewed: i (1) Determinations that proposed changes in equipment, systems, tests,

! experiments, or procedures do not involve an unreviewed safety question.

}

l (2) All new procedures and mjor revisions thereto having safety I significance and proposed changes in reactor facility equipment, or systems having safety signiricance.

(3) All new experiments or classes of experiments that could affect reactivity or rerult in the release of radioactivity, l

i

(4) Proposed changes in the Technical Specifications or the Operating j License.

1 j (5) Violations of the Technical Specifications of the Operating i License. Violations of internal procedures or instructions having safety significance.

l l

6-4 l

l L__---___._____.._.,.. _ _ _ ,

i (6) Operating abnormalities having safety significance.

i i

(7) Reportable occurrences listed in 6.6.2.

i (8) Audit reports.

! 6.2.4 Audit Function The audit function shall include selective (but comprehensive) l, examination of operating records, logs, and other documents. Discussions with j cognizant personnel and observation of operations should also be used as j appropriate. In no case shall the individual imediately responsi9e for the i area, audit in that area. Deficiencies uncovered that affect reactor safety

) shall be reported imediately to the University Radiation Safety Comittee. A written report of the findings of the audit shall be submitted to the Reactor

{

Use Comittee within 30 days after completion of the audit. The following items shall be audited:

1 l (1) Facility operations for conformance te the Technical Specifications i and applicable Operating License conditions, at least one per calendar year (interval between audits not to exceed 15 months).

(2) The retraining and requalification program for the operating staff, l at least once every other calendar year (interval between audits not to exceed 30 months),

i (3) The results of action taken to corret.t those deficiencies that may occur in the reactor facility equipment, systems, structures, or

]

! methods of operations that affect reactor safety, at least once

{

per calendar year (interval between audits not to exceed 15 j months).

(4) The reactor facility Emergency and Physical Security Plans and i

i 6-5 I

I L _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .

i implementing procedures at least c:.ce every other calendar year (interval not to exceed 30 months),

i i

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i 6-6 J

6.0 ADMINISTRATIVE CONTROLS (Continued) 6.3 Procedures Written procedures shall be prepared, reviewed and approved prior to initiating any of the activities listed in this section. The procedures shall be reviewed by the Reactor Use Committee (see 6.2.3) and approved by the [

Reactor Manager or a designated alternate. These reviews and approvals shall  ;

be documented in a timely manner. Substantive changes to the procedures shall l be made effective only after documented review by the Reactor Use Committee .

and approval by the Reactor Manager or a designated alternate. Minor modifications to the original procedures which do not change their original  !

intent may be made, but the modifications must be approved by the Reactor Manager or a designated alternate within 14 days. Temporary deviations from l

the procedures may be made by the on duty SR0 in order to deal with special or i unusual circumstances or conditions. Such deviations shall be documented and j reported to the Reactor Manager or a designated alternate. Several of the following activities may be included in a single manual or set of procedures l

or divided among various maneals or procedures

'

(1) Startup, operation and shutdown of the reactor.

j (2) Fuel loading, unloading, and movement within the reactor, i  !

j (3) Routine maintenance of major components of systems that could have an effect on reactor safety. ,

(4) Surveillance tests and calibrations required by the Technical l Specifications or those that may have an effect on reactor safety. I (5) Personnel radiation protection consistent with applicable

! regulations, i

(6) Administrative controls for operations and maintenance and for the 6-7

\

conduct of irradiations and experiments that could affect reactor safety or core reactivity.

(7) Implementation of the Emergency and Physical Security Plans.

J W

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6-8

i 6.0 ADMINISIRATIVE CONTROLS (Continued) 6.4 Exneriment Review And Anoroval Approved experiments shall be carried out in accordance with established I and approved procedures.

4 (1) All new experiments or classes of experiments shall be revicwed by the Reactor Usa Committee and approved in writing by the Reactor Manager or a designated alternate prior to initiation. '

i (2) Substantive changes to previously approved experiments shall be made only after they are retiewed by the Reactor Use Committee and j approved in writing by the Reactor Manager or a designated alternate. Minor changes that do not significantly alter the t experiment may be approved by the Reactor Manager or a designated 1

alternate. i l

t I h i  ;

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l  !

{ 6-9

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6.0 ADMINISTRATIVE (Continued) 6.5 Raouired Actions

, i i  ;

6.5.1 Action to be Taken in case of a Safety Limit Violation [

d l

(1) The reactor shall be shut down and reactor operations shall not be i resumed until authorized by the Nuclear Regulatory Commission f (NRC). ,

(2) The safety limit violation shall be promptly reported to the l l Reactor Manager or a designated alternate, j j  !

) (3) The safety limit violation shall be reported to NRC. l i

t l (4) A safety limit violation report shall be prepared. The report, and l

any follow up report, shall be reviewed by the Reactor Use f

Committee and shall be submitted to the NRC when authorization is

}

1 sought to resume operation of the reactor. The report shall ,

J 1 describe the following: I l

i  !

a. Applicable circumstances leading to the violation, including, j j whon known, the cause and contributing fictors. I i b. Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of f I

personnel and the public.  ;

c. Corrective action to be taken to present recurrence. l

l l 6.5.2 Action to be Taken in the Event of an Occurrence of the Type  !

! Identified in 6.6.2(1)b. and 6.6.2(1)c.

l 4 t j (1) Reactor conditions shall be returned to normal or the reactor shall  !

I be shut down. If it is necessary to shut down the reactor to l i

i  !

6 - 10 [

] i 1 i f

correct the occurrence, operations shall not be resumed unless authorized by the Reactor Manager or a designated alternate.

(2) Occurrence shall be reported to the Reactor Manager or a designated alternate and to the NRC.

(3) Occurrence shall be reviewed by the Reactor Use Comittee at its next scheduled meeting.

k l

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6.0 ADMINISTRATIVE CONTROLS (Continued) 6.6 Reoorts 6.6.1 Operating Reports A routine operating report providing the following information shall be submitted to the Nuclear Regulatory Comission in accordance with the provisions of 10 CFR 50.59 at the end of each 12 month period:

(1) A narrative sumary of reactor operating experience including the energy produced by the reactor, j (2) The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence.

(3) Tabulation of major preventive and corrective maintenance operations having safety significance.

(4) Tabulation of major changes in the reactor facility and procedt res, j and tabulation of new tests or experiments, or both, that are

) significantly different from those performed previously and are not j described in the Safety Analysis Report, inc'iuding conclusions that no unreviewed safety questions were involved.

(5) A sumary of the nature and amount of radioactive effluents l released or discharged to the environs beyond the effective control

of the owner-operator as determined at or before the point of such release or discharge. The sumary shall include to the extent 4 practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or

! diffusion is less than 25 percent of the concentration allowed or j recomended, a statement to this effect is sufficient.

I 6 - 12 4

h i

(6) A sumarized result of any environmental surveys performed outside the facility.

l (7) A summary of exposures received by facility personnel and visitors where such exposures are greater than 25 percent of that allowed or

recommended.

4

.i 6 - 13

l l

6.6.2 Special Reports  !

(1) There will be a be a report not later than the following working day by telephone to the appropriate NRC Regional Office and -

confirmed in writing by telegraph or similar conveyance to the l appropriate NRC Regional Office with a copy to the Director of j Inspection and Enforcement to be followed by a written report that '

describes the circumstances of the event within 14 days of any of j l the following

l l t a

I

a. Violation of safety limits (see 6.5.1). l

b. Release of radioactivity from the site above allowed limits l

] (see6.5.2). j

c. Any of the following (see 6.5.2)

'

i  ;

) (1) Operation with actual safety system settings for required l l systems less conservative than the limiting safety system l l settings specified in the Technical Specifications [

l (ii) Operation in violation of limiting conditions for  !

j operation established in the Technical Specifications  !

{ unless prompt remedial action is taken.

f j (iii) A reactor safety system component malfunction which j

! renders or could render the system incapable of performing l l its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown. l 4

(iv) An unanticipated or uncontrolled change in reactivity (

) greater than the licensed excess reactivity, or 0.763%  !

Ak/k (1.00$) (unchanged from one dollar), whichever is  !

l smaller.

(v) Abnormal and significant degradation in reactor fuel, or f

cladding, or both, or coolant boundary which could result I

] in exceeding prescribed radiation exposure limits of j personnel or environment, or both, f j i 6 - 14

]

't n ,. ,_ ,- ,_ -. -- - - - - - . - - - - . - , - . . _ _ _ _ _ . . , , --

__ - - - - . - _ _ . , _ -. _ _ . . - _ - - . - , . - , . , - - - - - - , . , , _ . ~ . , - - , . - - - - - - - - - - - - . - - - . -

J (vi) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

(2) A written report within 30 days to the appropriate NRC Regional

Office with a copy to the Director of Inspection and Enforcement c concerning the following

a. Permanent changes in the organization involving Nuclear l Engineering Department Executive Officer, Reactor Manager, or j Radiation Safety Officer,

b. Significant changes in the transient or accident analysis as

, described in the Safety Analysis Report.

4 l

1 6 - 15

6.0 ADMINISTRATIVE CONTROLS 6,7 Records i

6.7.1 Records to be Retained for a Period of at least Five Years or the l Life of the Component if Less than Five Years )

(1) Normal reactor facility operation (out not including supporting  !

documents such as checklists, log sheets, etc., which shall be ,

maintained for a period of a least one year). f i

(2) Principal maintenance operations. {

I (3) Reportable occurrences.

(4) Surveillance activities required by the Technical Specifications. l (5) Reactor facility radiation and contamination surveys where required by applicable regulations. (

(6) Experiments performed with the reactor.

l l (7) Fuel inventories, receipts, and shipments, j l

(8) Approved changes in operating procedures, t

1 k l (9) Recorris of meetings and audit reports of the Reactor Use i Comittee. j 6'.7.2 Records to be Retained for at least One Training Cycle  !

i Retraining and requalification of licensed operators: Records of the most recent complete cycle shall be maintained at all times the individual is l employed. ,

l 6 - 16 )

l i

6.7.3 Records to be Retained for the Lifetime of the Reactor Facility Applicable annual reports, if they contain all of the required information, r.ay be used as records in this section.

(1) Gaseous and liquid radioactive effluents released to the environs.

(2) Off-site environmental monitoring surveys required by the Technical Specifications.

(3) Radiation exposure for all personnel monitored.

(4) Drawings of the reactor facility.

6 - 17

r v LEVEL 1 Univorelty Preeldent

'f 1

r 1 r ,

Vice Preeldent Vice President Buelness & Finance Academio Affairs 1 r r

1 r Deen of Rediation Safety Env ron ental Engineering Committee Health & Safety 1

1 1 r i f l Nuc E Dept Reactor Use l Exec Officer Committee k k l I

l l l , ,

, , l LEVEL 2 l L~ Radiation Sefety Officer

__d l Manager l i l

1 r r l

-- Health Physice Ftaff Reactor Operatione

' Staff

- - - Committee Membership 1

i Figure 6-1 Organization structure I

i .

l kt h l

l l.

I TECHNICAL SPECIFICATIONS j for the i UTR-10 REACTOR FACILITY i at [

IOWA STATE UNIVERSITY i

Docket No. 50-116 l License No. R-59 >

i I

Original: August 1983 i Amendment 1: November 1988 i i

i

i i

j 1.0 DEFINITIONS f

The terms Safety Limit, limiting Safety System Setting, and Limiting .

Condition for Operation are as defined in paragraph 50.36 of 10 CFR Part 50.

l CHANNEL TEST - The introduction of a signal into the channel for verification that it is operable. l CHANNEL CAllBRATION - The adjustment of the channel such that its output  !

corresponds with acceptable accuracy to known values of the parameter which  !

the channel measures. Calib ation shall encompass the entire channel,

, including equipment actuation, alarm, or trip and shall be deemed to include a 2

Channel Test. l j  !

(

f CHANNEL CHECK A qualitative verification of acceptable performance by [

observation of channel behavior. This verification, where possible, shall (

j include the comparison of the channel with other independent channels or j j systems measuring the same variable, j i

I,

(

) CONFINEMENT BOUNDARY - The surface surrounding the reactor facility defined by

{

the interior partition walls of offices and laboratories on the north, east '

and south sides of the building and by the west interior wall which isolates

{

j the basement, first floor, and the west corridor of the second floor from the

{

j central bay. j l

l CONTROL R00 - A plate fabricated with Boral as the neutron absorbing material

which is used to establish neutron flux changes and to compensate for routine j reactivity losses. This includes safety type and regulating rods. ,

l

CORE - The portion of the reactor volume which includes the graphite

! reflector, core tanks, and control rods. The thermal column and shield tank l duct are not included, i 1

l 1-1 i i

l l

t DELAY TIME - The elapsed time between reaching a limiting safety system setpoint and the initial movement of a safety-type rod.

DELAYED NEUTRON FRACTION - When converting between absolute and dollar value reactivity units, a beta of 0.00763 is used.

DROP TIME - The elapsed time between reaching a limiting safety system setpoint and the full insertion of a safety type rod. I EXCESS REACTIVITY - That amount of reactivity that would exist if all control rods (control, regulating, etc.) were moved to the maximum reactive condition from the point where the reactor is exactly critical. [

l EXPERIMENT - Any operation, hardware, or target (excluding devices such as ,

detectors, foils, etc.) which is designed to investigate non routine reactor (

characteristics or which is intended for irradiation within the core region. l on or in a beam port or irradiation facility and which is not rigidly secured l to a core or shielti structure so as to be a part of their design. l MEASURED VALUE - The value of a parameter as it appears on the output of a channel, f I

i MEASURING CHANNEL - The combination of sensor, line, amplifier and output devices which are connected for the purpose of measuring the value of a I

parameter.

MOVABLE EXPERIMENT - An experiment where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating. j r

l OPERABLE - A component or system is capable of performing its intended j function. l OPERATING A component or system is performing its intended function.

1-2

t 4

l i

i REACTIVITY LIMITS - Those limits imposed on reactor core excess reactivity, t

, Quantities are referenced to a Reference Core Condition. l 4

i

! REACTIVITY WORTH OF AN EXPERIMENT The maximum absolute value of the t l reactivity change that would occur as a result of intended or anticipated  !

changes or credible malfunctions that alter experiment position or 1 configuration, i 1

REACTOR OPERATING The reactor is operating whenever it is not secured or  !

shutdown.  !

i  ;

REACTOR OPERATOR (RO) - An individual who is licensed to manipulate the 6 control: of a reactor, j f

I REACTOR SECURED - A reactor is secured when:

(1) It contains insufficient fissile material or moderator present in the reactor to attain criticality under optimum availible '

j conditions of moderation and reflection, or  !

l (2) A combination of the following:  !

I a. The minimum number of neutron absorbing control rods are fully 7 inserted or other safety devices are in shutdown position, as  :

j required by technical specifications, and I

b. The magnet power keyswitch is in the off position and the key is removed from the lock, and  ;

! c. No work it in progress involving core fuel, core structure, I l installed control rods, or control rod drives unless they are  ;

l physically decoupled from the control rods, and

! d. No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth exceeding i

. the maximum value allowed for a single experiment or 0.7637. l l Ak/k (1.00$) whichever is smaller. l

! (

I

, f i 1-3 f

i I

i

REACTOR SHUTDOWN - The reactor is shutdown if it is suberitical by at least 0.763% Ak/k (1.00$) in the Refeience Core Condition and the reactivity worth of all experiments is accounted for.

REACTOR SAFETY SYSTEMS - Those systems, including their associated input chsnnels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

READILY AVAILABLE ON CALL - Applies to an individual who:

(1) Has been specifically designated and the designation known to the operator on duty, and (2) Keeps the operator on duty informed of where he or she may be rapidly contacted (e.g., by phone, etc.), and (3) Is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g., 30 minutes).

REFERENCE CORE CONDITION - The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible, less than 0.23% Ak/k (0.30$).

REGULATING R00 A low worth control rod used primarily to maintain an intended power level that does not have scram capability. Its position may be varied manually or by t!.e servc controller.

SAFETY CHANNEL - A measuring or protective channel in the reactor safety system.

SAFETY TYPE R00 - A rod that can be rapidly inserted by cutting off the holding current in its electromagnetic clutch. This applies to safety 81, safety #2, and shim-safety.

1-4

1 SECURED EXPERIMENT - Any experiment, experiment facility, or component of an

experiment that is held in a stationary position relative to the reactor by

mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, q buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.

1 2

SENIOR REACTOR OPERATOR (SRO) An individual who is licensed to direct the activities of a Reactor Operator (RO) and 's manipulate the controls or a reactor.

j SHALL, SHOULD, AND MAY - The word "shall" is used to denote a requirement, the

} word "should" to denote a recommendation, and the word "may" to denote permission, neither a requirement nor a recommendation.

4 j SHUTDOWN MARGIN - The minimum shutdown reactivity necessary to provide j confidence that the reactor can be made suberitical by means of the control

and safety systems starting from any permissible operating condition although the most reactive rod is in its most reactive position, and that the reactor will remain subtritical without further operator action.

TRUE VALUE - The actual value of a parameter or variable.

UNSCHEDULED SHUTDOWN Any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operating error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or check out operations.

1-5

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety limits 2.1.1 Applicability These specifications apply to the variables that affect thermal and hydraulic performance of the core.

2.1.2 Objective To assure fuel cladding integrity.

2.1.3 Specifications A. The true value of the steady power level under various flow conditions shall not exceed 15 kilowatts.

8. The true value of the primary coolant flow rate shall not be less than 3.5 gpm for periods greater than 5 minutes at all power levels greater than one kilowatt.

C. The true value of the primary coolant outlet temperature shall not exceed 180 0F.

2.1.4 Bases Specifications A and 8 provide limits which protect the fuel cladding from damage due to excessive heat flux and surface temperature if the primary coolant pump fails. There is sufficient time for the operator to take corrective action before saturated pool boiling begins since the rate of temperature rise is approximately 9.8 0F per hour per kilowatt (SAR: 6.2); the time to increase from the maximum allowable core inlet temperature of 160 0F to the boiling temperature when operating at 10 kilowatts would be 2-1

1 l

approximately 32 minutes. Even if boiling did occur, the maximum critical l heat flux ratio (critical heat flux divided by the maximum heat flux in the i core) is so large (on the order of 1000) that damage to the cladding would be I very unlikely.

Specification C provides a ilmit for core outlet coolant temperature l under forced convection cooling. If the primary coolant flow rate was as low I as 3.5 gpm and the core inlet temperature was 160 0F at 10 kilowatts, the temperature rise across the core would be nearly 20 0F. As coolant j temperatures reach 1800 F (which is also the dump tank limit) and above, the l corrosion rate increases, thus accelerating the loss of fuel plate cladding. l I

l l

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, 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS (continued) 4 2.2 Limitina safety System Settinas i

2.2.1 Applicability (

i This specification applies to the setpoints of safety channels. [

t l 2.2.2 Objective l t

i I a To assure that automatic trip action is initiated and that the j

! operator is warned to take protective action to prevent a safety limit from '

i being exceeded.

l t

j 2.2.3 Specifications l' 1

i J The limiting safety system settings are the following: j i

f A. Maximum power level trip setpoint shall not exceed 12.5 l kilowatts, t i

j ,

B. Minimum primary coolant flow rate trip setpoint shall not be f l

! less than 5 gpe. l i i

} C. Maximum primary coolant outlet temperature trip setpoint shall f j not exceed 170 Of.

I i 2.2.4 Bases  !

i

! The trip setpoints provide adequate margins for the limits specified j

in 2.1.3. Trip setpoint A initiates automatic scram. Trip setpoints B and C 1 initiate alarms signaled by a horn and lighted annunciator. Operator .

l intervention in the non-scram trips provides timely response due to the slow

, [

j variation of temperature even in the most adverse case discussed in 2.1.4. l l  ?

l 2-3  !

,! i

3.0 LINITING CONDITIONS FOR OPERATION 3.1 Reactor Car.t EAtaneters 3.1.1 Applicability These specifications apply to the parameters which describe the reactivity condition of the core.

3.1.2 Objective To ensure that the reactor can not achieve prompt criticality and that it can be safely shut down under any condition.

3.1.3 Specifications The reactor shall not be made critical unless the following conditions exist:

A. The total core excess reactivity with or without experiments shall not exceed 0.50% Ak/k (0.65$).

B. The minimum shutdown margin provided by control rods in the reference core condition shall not be less than 0.35% Ak/k (0.465).

C. The fuel loading pattern and experiment apparatus inserted in the core shall be approved by the Reactor Use Comittee.

3.1.4 Bases Specifications A and B are based on values used in the power transient analysis (SAR: 6.3) where it is assumed that all of the excess reactivity, 31

0.50% Ak/k (0.655), is suddenly inserted as a positive step function. The safety system response is assumed to result in the minimum shutdown margin, 0.35% Ak/k (0.46$), being supplied by rapidly inserted safety type control rods, assuming the rod with the greatest worth is not available.

Specification C limits the changes in core configuration to those ,

approved by the committee charged with review and approval of experiments, i

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3.0 LINITING CONDITIONS FOR OPERATION (Continued) i 3.2 Reactor Control And Safety System -

i l

3.2.1 Applicability i

{

These specifications apply to the reactor safety system and safety.  !

related instrumentation. L l

3.2.2 Objective i

}' To specify the lowest acceptable level of performance or the minimum

(

. number of acceptable components for the reactor safety system and safety.

, related instrumentation.

3.2.3 Specifications 1,

The reactor shall not be made critical unless the following conditions f i

exist:

i l A. The retetor safety channels and safety related measuring channels i j shall be operable in accordance with Table 31, including the (

j minimum number of channels and the indicatcd maximue or minimum (

) setpoints.

I  !

l B. All three safety type control rods shall be operable and have the following response time capabilities:

i

! (1) Delay time shall not exceed 100 milliseconds.

(2) Drop time shall not exceed 600 milliseconds.

j

! C. The reactivity insertion rate for a sir.gle control rod shall not l exceed 0.019% Ak/k/sec (0.0255/sec).

l I 3-3

(

D. The dump valve shall be operable and shall be capable of reaching its normally-opened position in not more than 600 milliseconds after the scram signal is initiated.

E. The following bypasses may be applied to the channels indicated provided the appropriate compensation is employed:

(1) During measurements of control rod worth, the startup sequence for removal of safety-type rods with no position ,

indication may be altered if elapsed withdrawal times are observed as tne rod that establishes criticality is maneuvered.

(2) During measurements of reactor thermal power or control rod worth, the signal from the multirange linear power channel I neutron detector may be used exclusively for measuremr-t data recording if another detector of equivalent characteristics is used as a substitute.

3.2.4 Bases Specification A provides assurance that the reactor safety

instrumentation channels which may be needed to shutdown the reactor are operable. In addition, o > .-hannels which are important to safe operation because of interlock or alarm action are included. Each channel, along with l the setpoint, minimum number required, and function, is listed in Table 3-1.

-- The control rod withdrawal inhibit assures that the operator has an 4 operaole channel and appropriate neutron flux levels during startup.

-- The integrity of the startup neutron source is protected, and excessive j radiation levels are avoided by the coincident power and source / closure scrams. ,

-- The period scram limits the rate of power level increase to values which are manually controllable without reaching excessive power levels or fuel i

3-4

temperatures.

-- The linear percent power scrams provide automatic protective action to prevent exceeding the safety limit (2.1.3 A) on reactor power.

-- The multirange linear power channel provides information to guide the operator in establishing a set power leve'l with greater precision than that available from other power level monitoring channels.

l

-- The scram derived for the loss of high voltage to the neutron detectors t

provides a conservative response to an instrumentation system failure. The recommended operating voltage serves as the guide to detect a significant loss in power supply potential.

-- The alarm response to a fault in the scram circuit provides notice to the operator that the scram bus may not be operable ifa subsequent fault develops.

Operators are directed by procedure to shut down the reactor when this alarm is noted.

-- The moderator level channel inhibits control rod withdrawal until the moderator reaches an appropriate level above the fuel plates during startup operation. This minimum level restricts variations in moderator level at startup which could produce significant changes in reactivity balance and neutron detector response. (See also 3.3.4)

-- The moderator high level scram provides automatic shutdown and the subsequent draining of the moderator from the core tanks if the level exceeds the setpoint. Accidental flooding of the graphite reflector and uncontrolled l loss of coolant are avoided.

-- The shim safety position indicator channel must be operable to permit the operator to determine the excess reactivity from the critical rod position and rod calibration information.

-- The earthquake scram is provided to put the reactor in a shutdown condition before the protection system components are subjected to forces which might make them inoperable.

-- The manual scram and the magnet power keyswitch provide two methods for the reactor operator to manually shut down the reactor is an . safe or abnormal condition should occur.

Specification B is based on values used in the power transient

! analysis (SAR: 6.3) where it is assumed that two safety-type control rods are l

\

3-5 1

inserted as a ramp function. The safety system is assumed to initiate rod motion within 100 milliseconds after reaching the limiting safety system setting and to have the rods full inserted within a total time (delay plus insertion) of 600 milliseconds.

Specification C is based on a conservative value used for many years as a limit for reactors of the same type as the UTR-10. The limit assures a safe rate of power change during startup and during power ascensions.

Specification D assures that the moderator can be drained from the core tanks following a scram and provide backup shutdown action. When the dump valve opens in 600 milliseconds or less, the water is drained from the core tanks in approximately 4 seconds.

Speciffcat;on E provides for bypasses of a starcup interlock and a normal instrumentation signal connection.

-- The startup sequence requires a fixed order of safety rod emoval: Safety

1. 1 full out, safety #2 full out, then partial removal, depending on the excess reactivity, of either the shim-safety or regulating rods. To measure the maximum worth of the shim safety, the startup sequence interlock may be bypassed to allow removal of a safety rod and then the shim-safety. The remaining safety rod (neither safety rod is equipped with intermediate position indication) can be safely maneuvered to the critical position by keeping cumulative withdrawal time.

-- The normal connection of the multirange linear power channel can be bypassed with no reduction in the performance capability of the channel by

using another detector of equivalent characteristics, located in another but comparable position with relation to the fuel region, as the signal source for power level information. The changeover is completed at low power, and any change in calibration factor noted for later use at higher power levels. This bypass is used to obtain detector current information at high power levels for

> thermal power measurements and calibrations, and for control rod worth measurements, i

3-6

- _= _ _ - - . _ _ . -.

Table 3-1. Required Safety Channels and Safety-Related Channels.

l Channel Setpoint Min. Operable Function NUCLEAR Log % power Inhibits control Min. Countrate 20.1 mW 2 rod withdrawal.

Power level s1 W 2 Scram if all closures are not seated or source is not stored.

Period 25 seconds 2 Scram Linear % power s12.5 kW 2 Scram Multirange linear power --

1 Power information.

High voltage loss to neutron detectors 290% V(a) 2 Scram Scram circuit failure Fault to gnd 1 Alarm PROCESS Moderator level (b) Inhibits rod Normal op level 242 inches 1 magnet current.

High level 555 inches 1 Scram Shim-safety --

1 Excess reactivity position information.

Earthquake s4 Richter 1 Scram MANUAL Manual scram switch --

1 Scram Magnet power keyswitch --

1 Scram (a) Recommended operating voltage.

(b) Measured from the core tank base plate.

3-7

i 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.3 Coolant Systems 3.3.1 Applicability These specifications apply to the minimum operating equipment and limits of operation for the cooling system.

3.3.2 Objective To ensure that the reacter fuel can be adequately cooled with water of high quality.

3.3.3 Specifications The reactor shall not be made critical unless the following conditions exist:

A. The coolant system instrumentation channels shall be operable in accordance with Table 3-2, including the minimum number of channels and the indicated setpoints.

B. The primtry coolant inlet temperature shall be maintained in the range from a lower value determined by specification 3.1.3 A to 160 0F, and the primary coolant outlet temperature shall not exceed 160 0F.

C. The primary coolant flow rate shall be maintained in the range from 5 to 15 gpm, except that the flow rate may be less than 5 gpm if the power level is less than one kilowatt and an approved experiment requires the reduced flow condition.

D. The primary coolant temperature at the deionizer inlet shall not 3-8

exceed 140 0 F, and the deionizer flow shall be cut off until the temperature is below the limit. I 1

E. The primary coolant conductivity shall not exceed 2 micrombos per centimeter, except for periods not to exceed 7 days when the value shall not exceed 10 micrombos per centimeter.

F. The radiation exposure rate observed by the deionizer column i detector shall not exceed five times the nominal value measured l during normal full power operation, or the exposure shall not exceed 10 milliroentgens in one hour, whichever is the smaller value.

G. The net detection rate of confirmed fission product activity in the primary coolant shall not exceed the "decision" limit for the detection system used in the analysis.

, 3.3.4 Bases Specification A p*ovides assurance that the cooling system instrumentation channels are operable. Each channel, along with the setpoint, minimum number required, and function is listed in Table 3-2.

-- The moderator (primary coolant) level channel inhibits control rod withdrawal until the moderator reaches an appropriate level above the fuel plates during the startup operation. This minimum level interlock ensures ample coolant level to provide heat transfer for the fuel plates, and it also restricts the variations in moderator level at startup which could produce significant changes in reactivity balance and neutron detection rates (SAR:

4.5.3).

-- The primary coolant inlet temperature channel permits compliance with specifications 3.1.3 A and B by initiating a low-level alarm and providing the operator with information to establish a minimum coolant temperature which avoids an excessive reactivity inventory.

-- The primary coolant outlet temperature channel initiates an alarm signal 3-9

at the high temperature setpoint of 160 0F. This provides an adequate margin to avoid the the safety limit specified in 2.1.3 C.

-- The primary coolant flow rate channel initiates a 10w41cw alarm to warn the operator to reduce power in compliance with safety limit 2.).3 s, if the power level is at or above one kilowatt (SAR: 6.2).

-- The primary coolant conductivity channel initiates an alarm when the specific conductance exceeds 2 micrombos per centimeter. Operation may continue at a higher level for a limited time as indicated in 3.3.3 E.

-- The radiation equipment detector located near the deionizer initiates an alarm when the exposure rate exceeds five times the nominal value observed

, during normal full power operation (Sec 3.7.4).

Specification B is based on values of primary coolant temperature which must be maintained to avoid violating the limit on excess reactivity 1 (3.1.3 A) at the lower end of the range, and to avoid the high temperature safety limit 2.1.3 C (180 0F ) which also is the limit on the dump tank (SAR:

4.3.2). ,

Specification C provides a range on the primary coolant flow rate J which will adequately cool the fuel plates and avoid safety limit 2.1.3 B, and also provide flexibility for low-power experiments which may require an essentially stagnant coolant. It incorporates, through its lower limit of 5 gpm, an implied coolant leak detection provision since a significant loss of primary coolant (which is held in the process pit until analyzed) reduces the suction head on the pump to the point where the minimum flow rate cannot be maintained.

Specification D provides a limit to prevent damage to the deionizer resins and possible transport of fractured resin beads past the filter and into the primary coolant stream. The flow through the deionizer will have to be restored at a temperature below the limit if the conductivity limit is approached.

Specification E is based on experience at many facilities with similar i

coolant systems; this value is known to be a satisfactory upper limit for l normal operations. Trace mineral activation products do not exceed acceptable limits and corrosion rates are negligibly low when the upper limit is not exceeded (SAR: 4.2.2, 4.5.3 and 6.1.2). Provision for conductivity transients 3 - 10 l

due to crud releases adds flexibility to the limit.

Specification F is based on the assumption that the increase in exposure level is due to either fission product activity or radioactive trace reinerals normally present in the primary coolant being concentrated in the deionizer column. The trip setpoint is based on local conditions and must be determined so that it detects significant activity with respect to normal detection rates without causing too-frequent false alarms. Since the deionizer column is located near the boundary of the restricted area, the 10 mR/h upper limit provides a conservative margin to avoid exceeding the requirements of paragraph 20.105 of 10 CFR Part 20 on radiation doses in unrestricted areas.

Specification G provides a limit based on statistical hypothesis

testing and it depends on the detection system being used to evaluate the coolant sample. The term used in NCRP Report No. 58, pp. 275-279, is the "decision limit", and it can be used to determine if the net detection rate of the sample is statistically different from background at a confidence level of 95%, when equation (7.8) is used. The background in this case is taken to be the detection rate of samples without fission product activity. When the sample detection limit does exceed this limit, the leaking fuel assembly must be identified and removed from the reactor (see 3.7.3 C).

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Table 3-E. Required Coolant System Instrumentation Channels.

Channel Setpoint Min. Operable Function Moderator level (a)

Normal op level 142 inches 1 Inhibits rod magnet current; i establishes minimum coolant level.

Primary coolant inlet as required to 1 Thermal power temperature satisfy 3.1.3 information Primary coolant outlet temperature s160 0F 1 Alarm Primary coolant flow rate 25 gpm 1 Alarm Primary coolant conductivity s2 micrombos/cm 1 Alarm  ;

Radiation level As required to 1 Alarm Deionizer unit satisfy 3.3.3 F (a)iieasured from the core tank base plate.

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1 3.0 LIMITING CONDITIONS FOR OPERATION (continued) 3.4 Confinenent 3.4.1 Applicability This specification applies to the operations that require confiremet and to the equipment needed to achieve confinement.

3.4.2 Objective To ensure that the confinement boundary can be secured when needed.

3.4.3 Specifications A. The reactor confinement boundary shall be operable whenever the reactor is operating.

B. The reactor confinement boundary shall be secured during fuel transfer operations.

3.4.4 Bases Specification A is based on the assumption that the doors and windows located in the building walls that define the confinement boundary may need to be secured due to the accidental release of radioactive material generated during reactor operation.

Specification B is based on the hypothetical accident (SAR: 6.4) that occurs dtring movement of a fuel assembly and the importance of having the confinement boundary secured prior to the fuel transfer operation.

3 - 13

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i 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) l 3.5 Ventilation Systems There is no forced-air circulation system in the reactor room or the building housing it.

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3.0 LIMITING CONDITION!, FOR OPERATION (Continued) 3.6 Emeraency E.gwat 3.6.1 Applicability These specifications apply to the emergency power supply for the radiation monitoring system.

3.6.2 Objective To specify the source of emergency electrical power and the minimum operating time.

3.6.3 Specifications The reactor shall not be made critical unless the following conditions exist:

A. The battery-powered standby AC power supply for the radiation monitoring system shall be operable and shall have the following operating time capabilities:

2 (1) Operating time without the radiation evacuation horn being activated shall be not less than eight hours.

! (2) Operating time with the radiation evacuation horn being activated shall be not less than two hours.

3.6.4 Bases Specificatinn A requires that the standby AC power system, which consists of at least two lead-acid storage batteries, a charger-transfer unit, 3 - 15

and an inverter, be capable of providing a tripless switchover for supplying AC power to the radiation monitoring systems, and that the power source be able to sustain operation for the specified intervals. There are no systems, other than radiation monitoring, that need emergency power since the reactor is automatically shutdown when AC power failure occurs. The radiation evacuation horn imposes a large incremental load on the power source and severely reduces the operating time; however, the evacuation signal, if needed, would be of sufficient duration to accomplish its intended purpose.

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! 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.7 Radiation Monitorina Systems And Effluents 3.7.1 Applicability These specifications apply to the radiation monitoring systems and to the limits on effluent releases.

3.7.2 Objective To specify the minimum number of acceptable components or the lowest acceptable level of performance for the radiation monitoring systems and the limits for releases of effluents.

3.7.3 Specifications The reactor shall not be made critical unless the following conditions exist:

A. The radiation monitoring channels and components shall be operable in accordance with Table 3-3, including the minimum number of channels or components, and their setpoints.

B. The cumulative energy production of the reactor shall not exceed

4760 kilowatt-hours in any twelve month interval nor exceed 100 kilowatt-hours in any 7 day interval to limit the generation and release of argon 41.

C. If evidence exists that the limit in 3.3.3 G will be exceeded, the reactor shall be shutdown and the leakage source found and eliminated; however, the reactor may be operated intermittently to assist indetermining the source of leakage.

3 17

3.7.4 Bases Specification A provides assurance that the required radiation monitors are operable.

-- The air-particulate monitor is placed in service and operated continuously when designated experiments are being performed, viz., those which could produce airborne radioactivity. The alarm setpoint is influenced by the normal background reading while the reactor operates at the required power level and is based on the same reasoning as given for the deionizer monitor setpoint.

-- The radiation detector located near the deionizer initiates an alarm when the exposure rate exceeds five times the nominal value observed during normal full power operation. The trip value is sufficient for significant radiation events, yet not too sensitive to produce frequent false alarms. (See also 3.3.3 F.) This monitor would be the first to sense a release of fission products into the coolant.

-- The radiation area monitors are placed around the perimeter of the reactor room. All four units are able to initiate an alarm signal at or above 5 mR/h whenever the reactor console is energized. The south and west units initiate a radiation evacuation alarm at or above 50 mR/h when the reactor is in operation; when the console is not energized, the radiation evacuation setpoint is 5 mR/h. The 5 mR/h limit is based on the minimum value permitted for criticality monitoring of SNM in storage and applies when the area is unattended, while the 50 mR/h limit is based on the radiation level associated with the emergency action level for the alert classification.

-- The doorway radiation monitor serves as a frisker to detect abnormal levels of radiation when a person passes the detector. The increasing aural signal alerts the reactor operator and the affected individual that further assessment must be initiated.

-- The radiation film badge (or its equivalent) provides radiation dose information at the perimeter wall of the reactor room and serves as a control for the film badges used by personnel in the restricted area.

Specification B provides a conservative limit on the generation and release of argon 41 and is based on measurements at this facility (SAR:

3 - 18

4.5.4). Argon-41 is the only significant radioactive effluent produced during normal operation of the reactor, and the limits provided meet the requirements of paragraphs 20.103 and 20.106 of 10 CFR Part 20. The first part of specification B is based on the assumption that the reactor operates continuously at 10 kW for 476 hours0.00551 days <br />0.132 hours <br />7.87037e-4 weeks <br />1.81118e-4 months <br /> and that the dilution factor from diffusion of the air in the enclosure is only 10; for these conditions, the argon-41 concentration averaged over one year is about 50% of the value listed 1

for unrestricted areas in Table II, Appendix B of 10 CFR Part 20. The second part of specification B uses the assumptions that the reactor operates continuously at 10 kW for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> for one 40-hour week; these conditions yield an average concentration in the enclosure of 50% of the value listed for restricted areas in Table I, Appendix B of 10 CFR Part 20.

Specification C allows a search for a leaking fuel element to be conducted by using the reactor to the extent needed to detect the source of fission products.

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Table 3-3. Required Radiation Monitoring Channels or Components.

Channel Setpoint Min. Operabir' Function Air-Particulate (a) unit As required 1 Alarm Deionizer(b) unit As required to 1 Alarm satisfy 3.3.3 F Area units (c)(d) 5(50) mR/h 4 Alarm Doorway monitor --

1 Warn of abnormal radiation level.

Environmental 1 Integrated dose in res-Film badge or equ* valent -- restricted area (a)This unit is activated whenever designated experiments are being performed.

(b)This unit serves as the fission product monitor as specified in 3.3.3 F.

(c)When either the north or east area monitoring units are inoparable, portable instruments may be substituted for periods up to 48 ho7rs.

(d)The normal setpoint is shown. The parenthetical value is the maximum setpoint to be used depending on local conditions. Use of higher than normal setpoints requires approval of the Reactor Manager. The south and west units monitor the fuel storage area and are reset to the normal value after reactor shutdown.

3 - 20

3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.8 Exoeriments 3.8.1 Applicability  ;

t These specifications apply to the experiments installed in the reactor and its experiment facilities.

p 3.8.2 Objective To prevent damage to the reactor and excessive release of radioactive material in the event of experiment failure, and to avoid exceeding any safety limit.

. 3.8.3 Specifications t Experiments installed in the reactor shall meet the following conditions:

l A. Prior to initiation, each type of experiment utilizing the reactor shall be approved by the Reactor Use Committee.

8. Operational limits peculiar to an experiment shall be included in instructions to the reactor operator.

C. The reactivity worth of any single experiment, or group of experiments, installed in the core shall be limited to 0.48%

Ak/k (-0.63$) to +0.14% Ak/k (+0.19$).

4 D. Significant amounts of special materials used in experiments, including fissionable material, explosives or metastable

materials capable of significant energy release, or materials that are corrosive to reactor components or highly reactive with 3 - 21

the coolant, shall conform to established special requirements.

E. Credible failure of any experiment shall not result in releases or exposures in excess of established limits nor in excess of the annual limits established in 10 CFR Part 20.

F. Experiments shall be designed so that they will not contribute to the failure of other experiments, core components, or principle  !

physical barriers to uncontrolled release of radioactivity.

Also, no credible reactor transient shall cause an experiment to fail in such a way as to contribute to an accident.

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.1 3.8.4 Bases Specification A, B, D, E, and F are based on requirements stated in the standard for The Development of Technical Specifications for Research I Reactors, ANSI /ANS-15.1-1982.

l Specification C is based on the effect of the failure on an l experiment, or group of experiments, on the reactivity of the reactor. In the  !

case of an experiment failing with +0.14% Ak/k (+0.19$) inserted as a step I function, the resulting period would be 30 seconds, which can be easily  !

managed with control rod movement. If the -0.48% Ak/k (-0.63$) step insertion  ;

i occurs because of experiment failure, the reactor excess reactivity would drop to 0.02% Ak/k (0.0265), an amount sufficient to maintain criticality and to continue operation if necessary. t i i i

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4.0 SURVEILLANCE REQUIREMENTS Surveillance tests, except those specifically required for safety when the reactor is shut down, may be deferred during reactor shutdown; however, j they must be completed prior to reactor startup.

4.1 Reactor C.QI.g Parameters 4.1.1 Applicability These specifications apply to the surveillance activities required for 4 reactor core parameters.

4.1.2 Objective To specify the frequency and type of testing to assure that reactor core parameters conform to the specifications of section 3 of these Specifications.

4.1.3 Specifications

! A. The excess reactivity shall be measured at least annually and following significant core or control rod changes.

l B. The shutdown margin shall be measured at least annually and j following significant core or control rod changes.

I l 4.1.4 Bases The measurements required in specifications A and B are sufficient to l provide assurance that the reactor core parameters are maintained within the specifications 3.1.3 A and B since the fuel burnup rate is extremely low and important changes in the coro parameters can be detected on a timely basis.

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1 Core or control rod changes are not considered to be complete until the excess

2. reactivity and shutdown margin measurements are finished.

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4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.2 Reactor control, ud Safety System 4.2.1 Applicability These specifications apply to the surveillance activities required for the reactor control and safety system.

4.2.2 Objective To specify the frequency and type of testing or calibration to assure that the reactor control and safety system conforms to the specifir tions of section 3 of these Specifications.

4.2.3 Specifications A. The reactivity worth of each control rod shall be measured at least annually and whenever the core configuration is changed by fuel assembly replacement or rearrangement.

B. The drop time and delay time of each control rod, except for the regulating rod, and the withdrawal time of each control rod shall be measured at least annually and whenever any maintenance on a control rod which may affect its motion is completed.

C. The operability of the control rods and the dump valve shall be tested daily when the reactor is operating.

D. An operability test, including trip action, of each safety channel listed in Table 3-1 that provides a scram function shall be completed prior to each reactor startup following a period when the reactor is secured in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or at least weekly during continuous operating periods.

4-3

E. A calibration of the channels listed in Table 3-1 that can be calibrated shall be performed at least annually and whenever any maintenance on a channel which may affect its performance is completed.

F. The thermal power output of the reactor shall be measured at least annually.

4.2.4 Bases Specification A requires rod worth measurements for all rods at annual intervals; no significant changes in the worths of the control rods are likely to occur during that time. Since changes in the fuel loading in the vicinity of a control rod may cause a significant change in its worth, a measurement after the fuel change is appropriate.

The rod drop and delay time measurement intervals required in specification B verify the limits in specification 3.2.3 B and are appropriate to detect abnormal performance as can be shown by experience at this facility.

Withdrawal time measurements provide data to determine if specification 3.2.3 C is being violated.

Specification C verifies the operability requirements in specification 3 2.3 8 and D during each day of operation.

In specification D each channel capable of generating a scram signal is tested during the precritical procedure, prior to startup, so that the conditions of specification 3.2.3 A are satisfied.

Specification E requires calibration of safety and safety related channels at an interval which is appropriate and justified by experience at this facility.

The annual verification of reactor thermal power output, as required by specification F, is appropriate and justified by experience at this facility.

44

4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.3 Coolant Systems 4.3.1 Applicability These specifications apply to the surveillance activities required for the reactor coolant system.

4.3.2 Objective To specify the frequency and type of testing or calibration to assure that the reactor coolant system conforms to the specifications of section 3 of these Specifications.

4.3.3 Specifications i A. The coolant system instrumentation channels listed in Table 3-2

shall be calibrated at least annually and whenever any maintenance on a channel which may affect its performance is completed.

l B. The primary coolant temperature, flow rate, conductivity, and

! radiation level at the deionizer shall be measured and recorded at l startup and at least every four hours when the reacto'* is j operating.

! C. A primary coolant sample shall be analyzed for radioactivity at 1 east quarterly and whenever the exposure rate at the detonizer l

i exceeds the limits of specification 3.3.3 F.

I i 4.3.4 Bases Specification A requires calibration of the coolant system l

l inst'umentation channels at an interval which is appropriate and justified by

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9 experience at this facility.

Specification B requires verification of the operating limits of specifications 3.3.3 8 - F at an interval which is appropriate and justified by experience at this facility.

Specification C relates to the monitoring for fission products and l other activated materials in primary coolant samples. Experience at this facility shows that the sampling interval is appropriate.

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I 4.0 SURVEILLANCE REQUIREMENTS (Continued)  :

4.4 Confinement C

4.4.1 Applicability l l

This specification applies to the surveillance activities required for the reactor confinement. [

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4.4.2 Objective l

To specify the frequency and type of testing to assure that the f reactor confinement conforms to the srecificatione. of section 3 of these Specifications. f 4.4.3 Specification A. The doors and windows in ;he confinement boundary shall undergo testing for normal closure at least once every quarter.

l 4.4.4 Bases  :

This specification requires that the doors and windows in the i confinement boundary be tested to verify that they can be closed when needed. l The testing interval is adequate to verify operability based on experience at I this facility.

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4.0 SURVEILLANCE REQUIREMENTS (continued) .

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4.5 ventilation Systems i

This specification does not apply to this facility. l t

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4 4.0 SURVEILLANCE REQUIREMENTS (Continued) i 4.6 Emergency Egn r

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4.6.1 Applicability l These s'pecifications apply to the surveillance act uities required for  !

I the emergency power system, i i

i 4.6.2 Objective  :

t To specify the frequency and type of testing to assure that the [

emergency power system conforms to the specifications of section 3 of these  !

Specifications.

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4.6.3 Specifications  !

These surveillance activities are required for safety when the reactor f is not being operated.

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j A. The battery powered AC standby power supply shall be tested for

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switch over action, and for voltage and specific gravity j l characteristics at least quarterly. g

! I B. The batteries shall be tested for full discharge at lent every l three yea:'s. .

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I 4.6.4 Bases f Specification A requires verification of operability of the standby power supply to complete the switch over from normal AC power to the batteries g at an interval which is appropriate based on experience at this facility. The measured values of voltage and spee.ific gravity give adequate warning of reduced battery performance within the testing interval.

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A full discharge test of the batteries every three years, as required in specification B, is appropriate for the type of battery used in the power supply; the interval is well within the normal 4-5 year warranted life for conditions much more severe than those encountered in this application.

4 - 10

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4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.7 Radiation Monitorina System And Effluents 4.7.1 Applicability ,

l These specifications apply to the surveillance E.ctivities required for  :

the radiation monitoring system and effluents released from the facility.

4.7.2 Objective i

, To specify the frequency and type of testing to assure that the i radiation monitoring system and effluent releases conform to the i specifications of section 3 of these Specifications.

4.7.3 Specifications  ;

1 4 These sur'reillance activities (except E) are required for safety when  !

the reactor is not being operated.

A. A calibration of the channels listed in Table 3 3 that can be f calibrated shall be performed it least annually and whenever any maintenance on a channel which may affect its performance is (

completed.

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B. An operability test, including source checks, of the radiation  !

monitoring channels listed in Table 3 3 shall be performed at least monthly, i l

C. The radiation levels at the area and deionizer units shall be f measured and recorded at startup and at least every four hours when [

the reactor is operating. l i i D. The environmental film badge cited in Table 3-3 and smear surveys 4 - 11

l in and around the reactor enclosure shall be analyzed at least quarterly.

E. The cumulative energy conversion shall be computed and recorded at least quarterly, and it shall be computed on a weekly basis to monitor short-term argon 41 releases, t

4.7.4 Bases l

Based on experience at this facility and the average usage pattern of the reactor, specifications A D are adequate to verify that the operations conform to the specifications of 3.7.3.

l Specification E requires verification that the cumulative energy

! conversion limit of specification 3.7.3 B is not exceeded; this is an indirect method of monitoring the generation and release of argon-41.

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4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.8 Experiments 4.8.1 Applicability These specifications apply to the surveillance activities required for experiments installed in the reactor.

4.8.2 Objective To specify the frequency and type of testing to assure that the experiments conform to the specifications of section 3 of these Specifications. <

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4.8.3 Sp:,,1fications A. The identification and location of all installed experiments shall be recorded prior to each reactor startup.

B. Other specific surveillance activities shall be estrblished during the review and approval process specified i:i section 6.

4.8.4 Bases Specification A requires that the reactor operator verify that the installed experiments are approved on a most conservative frequency basis.

Specification B recognizes that detailed surveillance requirements will vary among experiments, and that the experiment review committee specifies the approprie.te type and frequency of surveillance.

4 - 13

5.0 DESIGN FEATURES 5.1 Sitt And Facility Descriotion T M reactor is housed in the Nuclear Engineering Laboratory, which is located on the west edge of the main campus of Iowa State University, ir. Ames, Iowa. The Nuclear Engineering Laboratory is a two story, three-level building of brick construction, built in 1934. The reactor, a model UTP. 10, was installed and first operated in 1959. It is fueled with uranium enriched to approximately 19.75% in the U-235 isotope, moderated and cooled with light water, reflected with graphite, and operates at a maximum thermal power of 10 kilowatts. The reactor is located on the ground floor level, central bay area of the Laboratory structure. The central bay is approximately 34 feet high and has a floor area of 37 feet by 56 feet of which a space approximately 37 feet by 38 feet is allocated to the reactor. A wall surrounding this area is constructed of standard concrete block and reaches a height of 10 feet 4 inches on the north, east and south sides; the west boundary is a wall that reaches J s the floor to the ceiling of the central bay region. The purpose of these walls is to limit access of unauthorized personnel to the immediate vicinity of the reactor and to define the outer perimeter of the restricted area.

The enclosure surrounding the reactor includes the central section of the building as defined by the interior partition walls or office: cr.d laboratories on the north, east and south sides of the building and by the west interior wall which isolates the basement, first floor, and the west corridor of the second floor from the central bay. The purpose of the enclosure is to act as as a confinement volume and to help limit the release of radioactive materials to the environment. The enclosure volume is slightly less than 2500 cubic meters, and the average infiltration rate for the butiding is estimated to result in two changes per hour. There is no central forced air circulation system in the building.

The enclosure has two outside doors, one in the east wall and a large overhead door opening to the south. All interior doors leading into the enclosure are of a standard type used in interior construction. Other 5-1

significant penetrations into the enclosure consist of roof-level windows on the north and south sides which can be manually opened or closed, as a group per side, in less that one minute per group.

5-2

5.0 DESIGN FEATURES (Continued) 5.2 Reactor Coolant System In normal operation, the primary coolant is pumped (18 psig) from the dump tank (capacity 220 gallons) through the heat exchanger (10 gpm, 80 0F ) to the bottom of the core tanks, upward past the fuel plates, to the overflow pipe manifold and returned to the dump tank (0 psig, 87 0F at 10 kW and 10 gpm); approximately 92 gallons are contained in the piping and core tanks ,

during operation. A quick-opening dump valve in the feed line to the core tanks is provided to allow draining of moderator (coolant) following a scram.

A low pressure (5 psig) steam heater and controller system for the dump tank and a deionizer/ filter system for the purification loop, which operates (1 gpm, <140 0F ) in parallel with the main loop are provided (Ref: Drawing Rl D-130). The operating temperature may range from about 80 0F to no more than 160 0F, with the lower end preferred to reduce the corrosion of aluminum.

Moderator level, inlet and outlet temperature, flow rate and conductivity sensors are installed at appropriate locations and connected to the process ,

instrumentation system (Ref: Drawing RI-D-Il6). The primary coolant system is essentially all-aluminum in construction; the pump casing and impeller, some f

valve parts, the dump tank heater element, and process instrumentation sensor l

. elements in contact with the water are stainless steel or similar corrosion-resistant materials.

The energy transferred thre4.. the heat exchanger is dissipated to the building sewer system by once through cooling water obtained from the campus water main. Secondary cooling flow is induced by water main pressure, and the

! flow rate is set by a motor-operated valve to control the amount of cooling in  ;

the heat exchanger resulting in core inlet temperature control. To prevent l secondary water from entering the primary system if a tube-leak should occur, a pressure differential is maintained in the heat exchanger to allow primary water to enter the secondary system.

The process pit accommodates the equipment and instrumentation sensors

! for the process system (Ref: Drawing Rl-E-151). A sump, with a capacity of l 9.5 gallons and a manually energized sump pump, can discharge liquids from i

5-3

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the procass pit to another sump located in the basement f!oor. The basement  ;

sump also receives seconaary coolant outlet water and acts like a dilution i tank; it has a capacity of 123 gallons. Outflow from the basement sump passes i through an overflow pipe connected to the building sewer system.  ;

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5.0 DESIGN FEATURES (Continued) 5.3 Peactor h E g l and Safety System Core A graphite reflector surrounds the core tanks, except a water reflector of no less than 13 cm thickness is maintained above the fuel assemblies during reactor operation. The composition of the region between the core tanks (the coupling region) can be changed by removal of graphite blocks and insertion of other materials, and small-volume experiments can be placed in the water gap between plates in the fuel assembly, or in the water reflector above or beneath the fuel, subject to specifications 3.. 3 (Experiments). A rabbit tube, no larger than 10 cm outside diameter, penetrates the graphite reflector at the west face of the north core tank. A neutron source, providing a minimum of 1.0 E+6 neutrons /second is inserted into the coupling region by means of the source positioner during the startup operation (Ref: Drawings Rl-E-154 and Rl E-161).

Fuel Reactor fuel is contained in aluminum-clad flat plates, similar to argonaut-type fuel. Fuel meat is V3Si2, enriched to 19.75% in the U 235 isotope, dispersed in aluminum to achieve a uranium density of 3.47 g/cc. The fuel meat, 0.51 mm thick, is clad with 0.38 mm aluminum. Each fuel plate contains 12.5 grams of U 235. The core contains 12 assemblies each with approximately 24 fuel plates depending upon the measured critical configuration. Solid aluminum plates and assemblies with missing fuel plates, for experimental purposes, are used to adjust the core fuel loading for the licensed excess reactivity of 0.50% Ak/k (0.655). (Ref: Drawing Rl-A-121-1) 5-5 e

i Safety System Four Boral control rods, two safety, one shim safety, and one regulating, are positioned in the graphite external reflector adjacent to the outside face and near each outside corner of the core tanks assembly (Ref:

Drawings R1-R-212, R1-R-213 RI-R-214). Each control rod is connected by a stainless steel flat spring to a motor-driven drum. Each safety rod is coupled to its drive mechanism by an electrically energized magnetic clutch.

Two safety rod drives have limit switches with consola indicators showing full withdrawal and full insertion. Shim safety and regulating rod positions are displayed on the control console.

The moderator level measuring channel provides a signal (interlock) which permits control rod drive magnets to be energized only after a minimum moderator level setpoint is exceeded, and it provides a signal (scram) when the moderator level exceeds the high level setpoint.

A neutron-sensitive power level measuring channel with a functional range of 1. E-7 to 1.5 E+2 percent power, based on 10 kilowatts thermal, provides a signal (interlock) which prevents withdrawal operation of the control rod drive motors if the minimum power level (minimum count rate) setpoint is not exceeded, a signal (scram) when the one-watt level is exceeded, and the neutron startup source is not in its storage position or all closures (two operating closures above the fuel. and one at the end of the thermal column) are not properly seated; this channel provides a signal to the period channel to generate a signal (scram) when the period is less than the short period setpoint. These signals are derived from the log percent power channel.

A neutron-sensitive power level measuring channel, with a functional range of 10 to 150 percent of 10 kilowatts, provides a signal (scram) when a high power level cetpoint is exceeded. This signal is derived from the linear percent power channel.

5-6

5.0 DESIGN FEATURES (Continued) f 5.4 Fissionable Materid Storagt Fueled experiments and fuel devices not in the reactor are stored in a dry fuel storage pit monitored by radiation and intrusion detectors (Ref:

Drawing Rl-E-194 and Physical Security Plan). The fuel storage array, under all conditions of moderation and reflection with light water, has an effective multiplication factor less than 0.9.

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1 6.0 ADMINISTRATIVE CONTROLS 6.1 Oraanization 6.1.1 Structure The organization for the management of the reactor facility shall be structured as indicated in Figure 6-1. Job titles are shown for illustration and may vary. Levels of authority indicated divide responsibility as follows: ,

Level 1

Responsible for the facility license and site administration i

{ Level 2: Responsible for the reactor facility operation and management Level 3: Responsible for daily operations. ,

i The Reactor Use Comittee is appointed by, and shall report to the  !

University Radiation Safety Committee. Radiation safety personnel shall I report to Level 2 or higher through an independent organizational channel, f

6.1.2 Responsibility l t

l The Executive Officei Department of Nuclear Engineering, shall be l responsible for the facility license and site administration.

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Individuals at the various management levels shown in Figure 6 1, in  ;

l addition to having the responsibility for the policies and operation of the ,

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facility, shall be responsible for safeguarding the public and facility  !

personnel from undue radiation exposures and for adhering to all requirements of the Operating License and the Technical Specifications.

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< In all instances, responsibilities of one level may be assumed by designated alternates, or by higher levels, conditional upon appropriate qualifications. I i 6.1.3 Staffing l I

i (1) The minimum staffing when the reactor is not secured shall be: I 6-1 "

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a. A licensed reactor operator in the control room.

b. A licensed senior reactor operator readily evallable on call.  !

c. A health physics-qualified individual readily available on call. ,

(2) Events requiring *.~e n direction of a senior reactor operator:  !

a. Reco'fery from unplanned or unscheduled shutdown (in this instance, l

documented verbal concurrence from a SRO is required), t

b. Fuel transfer operations.  !

c. Any maintenance activity involving the reactor safety system that I could cause a significant increase in the reactivity of the  !

I reactor.

d. Relocation of any in-core experiment with a reactivity worth ,

I greater than 0.763% Ak/k (1.005).

(3) Events requiring the presence of a health physics qualified l individual-

a. Fuel transfer operations.  !

b. Installation, changing locations, or removal of an experiment that l Involves removal of a shield plug or closure. l

c. Any maintenance activity involving the reactor safety systert that [

> could cause an abnormal release of radioactive materials.

l 6.1.4 Selection and Training of Personnel i l j The selection, training and requalification of operations personnel  ;

shall meet or exceed the requirement of American National Standard for l Selection and Training of Personnel for Research Reactors, ANSI /ANS-15.4-1977, f or its successor, and be in accordance with the Requalification Plan approved f

l by the Nuclear Regulatory Commission.

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6.0 ADMINISTRATIVE CONTROLS (Continued) 6.2 Review And 81dit The Reactor Use comittee (RUC) shall perform the independent review and audit the safety asp 2 cts of reactor facility operations.

6.2.1 Con. position and Qualifications The Reactor Use Comittee shall be composed of the Reactor Manager and a radiation health physicist, both ex o/ficio (voting), and at least three other members having expertise in reactor technology. Comittee members shall be appointed by the University Radiation Safety Comittee. (The Radiation Safety Comittee is composed of a representative from each of five colleges in the university in which research in the physical and life sciences and in engineering is conducted, plus three members with specific expertise in radiation protection. At least one of these members shall also represent university management. The college re yesantatives are chosen from the Colleges of Agriculture, Engineering, Sciences and Humanities, Home Economics, and Veterinary Medicine. One of the three other members shall be the University Radiation Safety Officer (R50). The chair of the comittee shall be appointed by the Vice President for Academic Affairs. The terms on the comittee for the RSO and chair are indefinite. All others are for three years with reappointments being determined by the Vice President for Academic Affairs.)

6.2.2 Charter and Rules (1) The Reactor Use Comittee shall meet at least semiannually and more frequently as circumstances warrant, consistent with effective monitoring of facility activities. Written records of its meetings shall be kept and copies forwarded, in a timely manner, to the University Radiation Safety Comittee.

6-3 m I.

(2) A quorum shall be three members. Members of the operation staff shall not be a voting majority.

(3) Any action recomended by the Reactor Use Comittee that may adversely affect the operations and/or safety of the University comunity shall be reported by the RUC chairman to the University

[ Radiation Safety Comittee which shall have veto power over such a recommendation.

(4) The Reactor Use Comittee may appoint one or more qualified individuals to perform the audit function.

6.2.3 Review Function The following items shall be reviewed:

(1) Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question.

(2) All new procedures and major revisions thereto having safety significance and proposed changes in reactor facility equipment, or systems having safety significance.

(3) All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity.

(4) Proposed changes in the Technical Specifications or the Operating License.

(5) Violations of the Technical Specifications of the Operating License. Violations of internal procedures or instructions having safety significance.

6-4

1 (6) Operating abnormalities having safety significance.

(7) Reportable occurrences listed in 6.6.2.

I (8) Audit reports.

6.2.4 Audit Function The audit function shall include selective (but comprehensive) l examination of operating records, logs, and other documents. Discussions with

! cognizant personnel and observation of operations should also be used as j appropriate. In no case shall the individual immediately responsible for the area, audit in that area. Deficiencies uncovered that affect reactor safety shall be reported immediately to the University Radiation Safety Committee. A written report of the findings of the audit shall be submitted to the Reactor i Use Committee within 30 days after completion of the audit. The following l items shall be audited:

j (1) Facility operations for conformance to the Technical Specifications

! and applicable Operating License conditions, at least one per calendar year (interval between audits not to exceed 15 months).

(2) The retraining and requalification program for the operating staff, j at least once every other calendar year (interval between audits l not to exceed 30 months),

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! (3) The results of action taken to correct those deficiencies that may j occur in the reactor (acility equipment, systems, structures, or i methods of operations that affect reactor safety, at least once

! per calendar year (interval between audits not to exceed 15 l months).

1 (4) The reactor facility Emergency and Physical Security Plans and 6-5

troplementing procedures at least once every other calendar year (interval not to exceed 30 months).

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6.0 ADMINISTRATIVE CONTROLS (Continued) 6.3 Procedures Written procedures shall be prepared, reviewed and approved prior to i initiating any of the activities listed in this section. The procedures shall  !

I be reviewed by the Reactor Use Committee (see 6.2.3) and approved by the Reactor Manager or a designated alternate. These reviews and approvals shall ,

be documented in a timely manner. Substantive changes to the procedures shall j i

be made effective only after documented review by the Reactor Use Committee  ;

and approval by the Reactor Manager or a designated alternate. Minor l

! modifications to the original procedures which do not change their original 4 intent may be made, but the modifications must be approved by the Reactor j Manager or a designated alternate within 14 days. Temporary deviations from j the procedures may be made by the on-duty SRO in order to deal with special or j unusual circumstances or conditions. Such deviations shall be documented and

] reported to the Reactor Manager or a designated alternate. Several of the '

i following activities may be included in a single manual or set of procedures l or divided among various manuals or procedures-(1) Startup, operation and shutdown of the reactor. l

, t (2) Fuel loading, unloading, and movement within the reactor. i (3) Routine maintenance of major components of systems that could have an effect on reactor safety.

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(4) Surveillance tests and calibrations required by the Technical  !

Specifications or those that may have an effect on reactor safety. I i

1 (5) Personnel radiation protection consistent with applicable  !

regulations, i

! (6) Administrative controls for operations and maintenance and for the f i  !

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conduct of irradiations and experiments that could affect reactor safety or core reactivity.

(7) Implementation of the Emergency and Physical Security Plans.

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l 6.0 ADMINISTRATIVE CONTROLS (Continued)  !

l 6.4 Exneriment Review And Anoroval l Approved experiments shall be carried out in accordance with established and approved procedures.

(1) All new experiments or classes of experiments shall be reviewed by the Reactor Use Committee and approved in writing by the Reactor Manager or a designated alternate prior to initiation.

, (2) Substantive changes to previously approved experiments shall be i 1 made only after they are reviewed by the Reactor Use Comittee and  !

approved in writing by the Reactor Manager or a designated [

alternate. Minor changes that do not significantly alter the l experiment may be approved by the Reactor Manager or a drisignated I j alternate. l

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6.0 ADMINISTRATIVE (Continued) 6.5 Reauired Actions 6.5.1 Action to be Taken in case of a Safety Limit Violation (1) The reactor shall be shut down and reactor operations shall not be resumed until authorized by the Nuclear Regulatory Connission (NRC).

(2) The safety limit violation shall be promptly reported to the Reactor Manager or a designated alternate.

(3) 1he safety limit violation shall be reported to NRC.

(4) A safety limit violation report shall be prepared. The report, and any follow up report, shall be reviewed by the Reactor Use Comnittee and shall be submitted to the NRC when authorization is sought to resume operation of the reactor. The report shall describe the following:

a. Applicable circumstitices leading to the violation, including, when known, the cause and contributing factors,

b. Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public,

c. Corrective action to be taken to prevent recurrence.

6.5.2 Action to be Taken in the Event of an Occurrence of the Type Identified in 6.6.2(1)b and 6.6.2(1)c.

(1) Reactor conditions shall be returned to normal or the reactor shall be shut down. If it is necessary to shut down the reactor to 6 - 10

correct the occurrence, operations shall not be resumed unless authorized by the Reactor Manager or a designated alternate.

(2) Occurrence shall be reported to the Reactor Manager or a designated alternate and to the NRC.

(3) Occurrence shall be reviewed by the Reactor Use Comittee at its next scheduled meeting.

6 - 11

6.0 ADMINISTRATIVE CONTROLS (Continued) 6.6 Reoorts 6.6.1 Operating Reports A routine operating report providing the following information shall be submitted to the Nuclear Regulatory Comission in accordance with the provisions of 10 CFR 50.59 at the end of each 12 month period; (1) A narrative sumary of reactor operating experience including the energy produced by the reactor.

(2) The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence.

(3) Tabulation of major preventive and corrective maintenance operations having safety significance.

(4) Tabulation of major changes in the reactor facility and procedures, and tabulation of new tests or experiments, or both, that are significantly different from those performed previously aiid are not described in the Safety Analysis Report, including conclusions that no unreviewed safety questions were involved.

(5) A sumary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the owner operator as determined at or before the point of such release or discharge. The sumary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25 percent of the concentration allowed or l

recommended, a statement to this effect is sufficient.

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! (6) A sussnarized result of any environmental surveys performed outside the facility, i

i (7) A susenary of exposures received by facility personnel and visitors 1 where such exposures are greater than 25 percent of that allowed or 1

reconsnended.

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1 6.6.2 Special Reports (1) There will be a be a report not later than the following working

, day by telephone to the appropriate NRC Regional Office and j confirmed in writing by telegraph or similar conveyance to the l appropriate NRC Regional Office with a copy to the Director of j

, Inspection and Enforcement to be followed by a written report that [

describes the circumstances of the event within 14 days of any of I l the following-I

a. Violation of safety limits (see 6.5.1).  !

] b. Release of radioactivity from the site above allowed limits (see6.5.2).

l c. Any of the following (see 6.5.2): l i

l (1) Operation with actual safety system settings for required l i systems less conservative than the limiting safety system  !

settings specified in the Technical Specifications  !

(11) Operation in violation of limiting conditions for i l

1 operation established in the Technical Specifications

unless prompt remedial action is taken, j (iii) A reactor safety system component malfunction which 1 renders or could render the system incapable of performing f its intended safety function unless the malfunction or

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condition is discovered during maintenance tests or [

periods of reactor shutdown. [

(iv) An unanticipated or uncontrolled change in reactivity {

greater than the licensed excess teactivity, or 0.763%

Ak/k (1.005), whichever is smaller.

l (v) Abnormal and significant degradation in reactor fuel, or j cladding, or both, or coolant boundary which could result '

in exceeding prescribed radiation exposure limits of personnel or environment, or both, j (vi) An observed iradequacy in the implementation of 6 14 l

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administrative or procedural controls such that the inadequacy causes or could have caused the existence or  !

development of an unsafe conditic,n with regard to reactor j operations.

(2) A written report within 30 days to the appropriate NRC Regional j office with a copy to the Director of Inspection and Enforcement [

concerning the following: {

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a. Permanent changes in the organization involving Nuclear [

Engineering Department Executive Officer, Reactor Manager, or I

! Radiation Safety Officer.

l b. Significant changes in the transient or accident analysis as t

) described in the Safety Analysis Report, f

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r 6.0 ADMINISTRATIVE CONTROLS 6.7 Records 6.7.1 Rei.srds to be Retained for a Period of at least Five Years or the Life of the Component if Less than Five Years (1) Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc., which shall be maintained for a period of a least one year).

(2) Principal maintenance operations.

(3) Reportable occurrences.

(4) Surveillance activities required by the Technical Specifications.

(5) Reactor facility radiation and contamination surveys where required by applicable regulations.

(6) Experiments performed with the reactot (7) Fuel inventories, receipts, and shipments.

(8) Approved changes in operating procedures.

(9) Records of meetings and audit reports of the Reactor Use Connittee.

6.7.2 Records to be Retained for at least One Training Cycle Retraining and requalification of licensed operators: Records of the most recent complete cycle shall be maintained at all times the individual is employed.

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6.7.3 Records to be Retained for the Lifetime of the Reactor Facility l I

Applicable annual r. *

- if they contain all of the required i j information, may be used as ' t s, <! in this section. l L

i (1) Gaseous and liquid radioactive effluen's released to the environs.

! (2) Of f site environmental monitoring surveys required by the Technical  !

. Specifications.

l 4 t l (3) Radiation exp1sure for all personnel mor.itored,  !

(4) Drawings of the reactor facility. l I

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LEVEL 1 University Preefdent 1

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r 5 f Vice President Vice Pre ant Businese Et Finance Acedemi. Af felte 3 r 1 (

Director Redletion Safety Deen of En Ironmen,teI gg ,,

Engineering Committes g, I

i 3 r 1 r l Nuc E Dept Reactor Use l Exec Officer Committee I

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l Radiation LEVEL 2 L~ Sofety Ullicer Reactor ,

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1 LEVEL 3 6-- fleelth Phn'es sin',.

Renctor - -

f Opstoer a t,i,ons

- - - Committee Membership Figure 6-1 Organization structure

.