ML20072G680

Tech Specs 1.0 Through 6.0 Re Safety Limits & Limiting Safety Sys Settings,Limiting Conditions for Operation, Surveillance Requirements,Design Features & Controls

ML20072G680
Person / Time
Site:	05000112
Issue date:	05/18/1983
From:	OKLAHOMA, UNIV. OF, NORMAN, OK
To:
Shared Package
ML20072G622	List: ML20072G617 ML20072G642 ML20072G680
References
NUDOCS 8306280594
Download: ML20072G680 (33)

Text

- - . . . . - . . . - . . - - . - _ . - .. _ . - .

APPENDIX A LICENSE N0. R-53 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF OKLAHOMA REACTOR MODEL AGN-211P (S.N. 102)

May 18, 1983 8306280594 830620 PDR ADOCK 05000112 P PDR

- -j l

I l

i TABLE OF CONTENTS l

PAGE 1.0 DEFINITIONS 1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4 2.1 Safety Limits ..................... 4 2.2 Limiting Safety Systems . . . . . . . . . . . . . . . . 4 3.0 LIMITING CONDITIONS FOR OPERATION 6 3.1 Reactivity Limits ................... 6 3.2 Control and Safety Systems . . . . .. . . . . . . . . . 7 3.3 Limitations on Experiments . . . . . . . . . . . . . . 10 3.4 Shielding ....................... 11

~

4.0 SURVEILLANCE REQUIREMENTS 13 4.1 Reactivity Limits ................... 13 4.2 Control and Safety System . . . . . . . . . . . . . . . 14 5.0 DESIGN FEATURES 16 5.1 Ge n e ra l De s i gn . . . . . . . . . . . . . . . . . . . . . 16 5.2 Core Assembly ..................... 16 5.3 Fu e l El emen t . . . . . . . . . . . . . . . . . . . . . . 17 5.4 Safety and Control Rods . . . . . . . . . . . . . . . . 17 5.5 Shielding ....................... 19 5.6 Water Purification System . . . . . . . . . . . . . . . 19 5.7 Experimental Facilities . . . . . . . . . . . . . . . . 19 5.8 Instrumentation and Controls . . . . . . . . . . . . . . 20 5.9 Fuel Storage . . . . . . . . . . . . . . . . . . . . . . 20 5.10 Reactor Room . . . . . . . . . . . . . . . . . . . . . . 21 6.0 ADMINISTRATIVE CONTROLS 23 6.1 Organization ..................... 23 T

1.0 DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Conditions for Operation (LCO) are as defined in 50.36 of 10 CFR part 50.

1.1 Reactor Shutdown - The reactor shall be considered shutdown when all safety (2) and control rods (2) are fully inserted.

1.2 Reactor Operatinn - Reactor operation is any condition wherein the reactor is not shutdown; i.e., at least one rod is partially or fully out.

1.3 Measuring Channel - A measuring channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring or responding to the value of a process variable.

1.4 Safety Channel - A safety channel is a measuring channel in the reactor safety system.

1. 5

• Reactor Safety System - The reactor safety system is that combination of safety channels and associated circuitry which forms an automatic protective system for the reactor or provides information which requires manual protective action be initiated.

1.6 Reactor Component - A reactor component is any apparatus, device, or i material that is a normal part of the reactor assembly.

1.7 Operable - Operable means a component or system is capable of performing its intended function in its normal manner.

1.8 Operating - Operating means a component or system is performing its intended function in its normal manner.

1.9 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification may include comparison of the channel with other independent channels or methods measuring the same variable.

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

1.11 Channel Calibration - A channel calibration is an adjustment of the char.nel such that its . output responds, within acceptable range and accuracy, to known values of the parameter which the channel measures.

Calibration shall encompass the entire channel, including the equipment, actuation, alarm, or trip.

1.12 Reactor Utilization - Utilization of the reactor is defined by three (3) categories:

1.12 a - Irradiation of a sample in or near the core or by a beam brought out of the reactor.

1

1

.. \

l 1.12 b - Experiments is any use of the reactor or a component of the reactor for other than a sample irradiation. f 1.12 c - Operation of the reactor for the purpose of evaluating the reactor's performance, operator training, and demonstration.

1.13 Secured Experiment or Sample - Any experiment or sample, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor by mechanical means. The restraint shall exert sufficient force on the experiment or sample to overcome the expected effect of hydraulic, pneumatic, bouyant, or other forces which are normal to the operating environment of the experiment or sample or which might arise as a result of credible malfunctions.

1.14 Unsecured Experiment or Sample - Any experiment or sample, or component of an experiment is deemed to be unsecured whenever it is not secured as defined in 1.13 above.

1.15 Movable Experiment or Sample - A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.

1.16 Removable Experiment - A removable experiment is any experiment, experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life of the reactor.

1.17 Experimental Facilities - Experimental facilities are those portions of the reactor assembly that are used for the introduction of experiments or samples into or adjacent to the reactor core region or allow beams of radiation to exist from the reactor shielding. Experimental f acilities shall include the thermal column, glory hole, and access ports.

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

l 2

1.19 Static Reactivity Worth -The static reactivity worth of an experiment is the value or the reactivity change which is measureable by calibrated control or regulating rod comparison methods between two defined terminal postions or configurations of the experiment. For removable experiments, the terminal postions are fully removed from the reactor and fully inserted or installed in the normal functioning or intended position.

1.20 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N. I. Sax, Third Ed. (1968), or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its pubication 704-M,1966, " Identification System for Fire Hazards of Material," also enumerated in the " Handbook for Laboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.

1.21 Shutdown Margin - The shutdown margin is defined as the amount of negative reactivity by which the reactor is shutdown with the most reactive rod stuck out of the core.

1.22 Annually - Annually is defined as 12 months but not to exceed 14 months.

3

i

, 2.0 SAFETY LIMITS AND LIMITED SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability This specification applies to the maximum core temperature during ,

1 steady state or transient operation.

Objective To assure that the integrity of the fuel material is maintained and that fission products are retained in the core matrix to the extent possible.

Specification The maximum core temperature shall not exceed 392*F under any conditions.

Bases The polyethylene core material does not melt below 392*F and is expected to maintain its integrity and retain essentially all of the fission products at temperatures below 392 F. At 100 watts, the corresponding maximum core temperature would be well below 392 F thus assuring the integrity of the core and aiding the retention of fission products.

2.2 Limiting Safety System Settings Applicability This specification applies to the reactor safety systems which will limit the maximum core temperature.

4

\

o Objective To assure that automatic protective action is initiated to prevent

- the safety limit from being exceeded.

Specification The safety channels shall initiate a reactor scram at the following

' limiting safety system settings:

Channel Function LSSS Log Power Channel High Power 150 w'atts Low Power 10-7 watts Short Period 5 seconds Linear Power Channel High Power 120 watts Bases The high power LSSS in conjunction with the automatic safety systems and/or manual scram capabilities will assure that the safety limits will' not be exceeded during steady state or as a result of the most severe credible transient.

4 5

3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits Applicability This specification applies to the reactivity condition of the reactor and the reactivity _ worths of control rods and experiments.

Objective To assure that the reactor can be shut down at all times (normal and emergency) and that the safety limits will not be exceeded.

Specifications 3.11 The total excess reactivity (p) of any combination of core configuration and fuel temperature, including the static positive reactivity worth of any experiments or samples, shall not exceed 0.65%.

3.12 The reactor will operate with a maxmimum of 20 fuel elements in any core configuration.

a. Standard,12 fuel element, parallelepiped with a graphite or water reflector,

b. Flux trap, a maximum 20 fuel element configuration, with a one or two element void and graphite or water reflector.

3.13 The shutdown margin with the most reactive safety or control rod fully removed shall be at least 0.5% ak/k.

Bases The limitations on total core excess reactivity assure reactor periods of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down without exceeding 6

I any safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram.

The flux trap configuration adds considerably to the educational value of the reactor. This is important since the reactor is used most heavily in direct support of the undergraduate and graduate nuclear engineering program. Additionally, the flux trap offers a neutron flux in the trap equivalent to a power level of about 150 watts.

The total excess reactivity limit of 0.65% ak/k applies equally to all core configurations.

3.2 Control and Safety Systems Applicability These specifications apply to the reactor control and safety systems.

Objective To specify lowest acceptable level of performance, instrument set points, and the minimum number of operable components for the reactor control and safety systems.

Specification 3.21 The total scram insertion time of the safety rods and coarse control rod shall be less than 0.5 seconds.

3.22 The safety rods and coarse control rod shall be interlocked such that:

1. Only one safety rod can be withdrawn at a time.

2. The coarse control rod cannot be withdrawn unless both safety rods are fully withdrawn.

7

3.23 Nuclear safety channel instrumentation shall be operable in accordance with Table 3.1 whenever the reactor is not shut down.

3.24 Loss of electric power will result in a scram.

3.25 A neutron source must be present in the reflector.

3.26 The reactor will not operate at pool water temperatures less than 5*C.

Bases The specifications on scram insertion time, in conjunction with the safety system instrumentation and set points, assure safe reactor shutdown during the most severe foreseeable transients. Interlocks on control and safety rods assure an orderly approach to criticality and an adequate shutdown capability.

The neutron detector channels (nuclear safety channels 1 and 2) assure that reactor power levels are adequately monitored during reactor startup and operation. Requirements on minimum neutron levels will prevent reactor startup unless the log power channel is operable and responding, and will cause a scram in the event of instrumentation failure. The power level scrams initiate redundant automatic protective action at power low enough to assure safe shutdown without exceeding any safety limits. The period scram conservatively limits the rate of rise of reactor power to periods which are manually controllable and will automatically scram the reactor in the event of unexpected large reactivity additions.

The AGN-211's negative temperature coefficient of reactivity causes a reactivity increase with decreasing core temperature. The shield water temperature interlock will prevent reactor operation at temperatures below 5 C thereby limiting potential reactivity additions associated with 8

TABLE 3.1

~

REACTOR SAFETY SYSTEMS WHICH PRODUCE SCRAMS DEVICE (SAFETY CHANNEL) FUNCTION LIMITING SAFETY SYSTEM SETTINGS

! 1. Log Power Channel High Power Limit 150 watts (Nuclear Safety Channel-1)

Low Power Limit 10-7 watts Short Period Limit 5 seconds

2. Linear Power Channel High Power 120 watts (Nuclear Safety Channel-2)

3. Reactor Tank Water Level Low Pool Water 9 inches below reactor tank top

4. Pool Water Temperature Low Pool Water Temperature 5 C

5. Manual Scram Normal Shutdown Scram on operator decision

[

1 i

a 1

temperature decreases.

Water in the shield tank is an important component of the reactor shield and operation without the water may produce excessive radiation levels. The shield tank water level interlock will prevent reactor operation without adequate water levels in the shield tank.

The manual scram allows the operator to manually shut down the reactor if an unsafe or otherwise abnormal condition occurs that does not otherwise scram the reactor. A loss of electrical power de-energizes the safety and coarse control rod holding magnets causing a reactor scram, thus assuring safe and immediate shutdown in case of a power outages.

3.3 Limitations on Experiments Applicability 1

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

Objective To prevent damage to the reactor or excessive release of radioactive materials in the event of an experimental failure.

t Specifications

a. Experiments or samples containing materials which might damage reactor components by corrosion shall be doubly encapsulated,

b. Explosive materials shall not be inserted into experimental facilities of the reactor unless approved by the Reactor Safety Committee,

c. The calculated radioactive material content, including fission products of. any experiment shall be limited so that the complete release of all gaseous, particulate, or volatile components from the 10

experiment will not result in doses in excess of 10% of the equivalent annual doses stated in 10 CFR Part 20 for persons occupying restricted areas during the length of time required to evacuate the restricted area.

d. No part of the concrete biological shield will be removed or replaced to accommodate experiments without the written permission of the Reactor Safety Committee.

Bases These specifications are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from an experimental failure and to protect operating personnel and the public from excessive radiation doses in the event of an experimental failure.

3.4 Shielding Applicability This specification 3pplies to reactor shielding required during reactor operation.

Objective The objective is to protect facility personnel and the public from radiation exposure.

Specification The. following shielding requirements shall be fulfilled during i

reactor operation:

a. The reactor shield tank shall be filled with water to a height not less than 9 inches of the tno of the pool.

11

b. A gamraa shield equivalent to one inch of iron plate will be layed across the pool tank on either side of the control briage for operation at power levels above 50 watts.

Bases Thi facility shielding in conjunction with designated restricted radiation areas is designed to limit radiation dose to facility personnel and to the public to a level below 10 CFR 20 limits under operating conditions, 'and to a level below criterion 19, Appendix A,10 CFR 50 recommendations under accident conditions. Because the Reactor Laboratory has moderate student use, a very conservative exposure policy has always and will continue to be followed.

i 12

. ~

4.0 SURVEILLANCE REQUIREMENTS Actions specified in this section are not required to be performed if during the specified surveillance period the reactor has not been brought critical or is maintained in a shutdown condition extending beyond the specified surveillance period. However, the surveillance requirements must be fulfilled prior to subsequent startup of the reactor.

4.1 Reactivity Limits Applicability This specification applies to the surveillance requirements for reactivity l'imits.

Objective To assure the reactivity limits for Specification 3.21 are not exceeded.

Specification

a. Safety and control rod reactivity worths shall be measured annually,

b. The shutdown margin shall be determined annually,

c. The excess reactivity of the core configuration and experiments'will be determined during the initial phase of startup at or below 1 watt.

If the excess reactivity is greater than 0.65% ak/k, the reactor will be scramed and the excess reactivity of the core and experiments adjusted to conform with section 3.11.

Bases The control and safety rod reactivity worths are to be measured to assure that no degradation or unexpected changes have occurred which 13

could adversely affect reactor shutdown margin or total excess reactivity. The shutdown margin and total excess reactivity are determined to assure that the reactor can always be safely shutdown with one rod not functioning and that the maximum possible reactivity insertion will not result in reactor periods shorter than those that can be adequately terminated by either operator or automatic actions. Based on known experience with AGN reactors, significant changes in reactivity or rod worth are not expected with an 18-month period (we have experienced no change in rod worths over the past 20 years).

4.2 Control and Safety System Applicability This specification applies to the surveillance requirements of the reactor control and safety systems.

Specification

a. Safety and control rod scram times shall be measured annually.

b. Safety and control rods and drive shall be inspected for deterioration annually.

c. Channel tests of all safety channels listed in Table 3.1 (except safety channels 3 and 4) will be performed prior to the first reactor startup each day.

I

d. The period and power level measuring channels shall be calibrated and set points verified annually.

e. The shield water level interlock and shield water temperature interlocks shall be calibrated and tested annually.

f. Pool water conductivity will be measured prior to each reactor operation and a conductivity of less than 15 microomhos/cm will be 14-

maintained except for periods of maintenance not to exceed 30 days.

g. The pool water activity monitor will be in operation when ever the reactor is not shutdown and will provide an indication at the reactor console.

Bases The channel tests and checks required will assure that the safety channels and scram functions are operable. Based on operating experience with reactors of this type, scram measurements, channel calibrations, set point verifications, and inspections are of sufficient frequency to assure, with a high degree of confidence, that the safety system settings will be within acceptable drift tolerance for operation.

15

5.0: DESIGN FEATURES 5.1 General Design The AGN 211P nuclear reactor consists of the reactor unit and the control console. The reactor unit consists of a core-reflector assembly in a water-filled tank suspended by means of an aluminum frame attached to a steel I-beam bridge mounted at the top of the tank. The control rod actuator assemblies are mounted on the bridge with the absorber control rods mounted beneath. The reactor tank is a 3/8" thick 40" x 60" x 8' high tank shielded by a number of concrete blocks 20" x 20" x 40". A platform and railing are secured to the top of the reactor platform. The reactor tank is empty below floor level (sunk into the ground).

5.2 Core Assembly i The core assembly consists of 2-7/8" x 3-1/16" x 29" fuel elements.

The base of the element has an aluminum guide which fits into the aluminum grid plate. The grid plate is essentially identical to the one used in the Bulk Shielding Facility. The grid plate is supported from.

4 the I-beam bridge by an aluminum frame. The bridge also acts as a support for the ion chambers.

The elements are removable from above by grasping the element handle with the proper removal tool. The control rods fit into the core assembly in such a manner that a control rod cannot be removed without removing a fuel element.

The lattice is arranged so that each fuel element in the assembly is surrounded by approximately 1/8" of water. One watt operation results in a temperature rise at the center of the element of approximately 0.04 C, hence, at 100 watts a fuel temperature equilibrium of approximately 4 C 16

v -

over ambient is expected (this temperature increase will reduce the excess reactivity by about 0.08%).

5.3 Fuel Elements Each fuel element is 2-7/8" x 3-1/16" x 29". The center 10" of each element is composed of the homogeneous ~ mixture of bO2 in polyethylene (56 mg U-235/cm3) in the form of'20 micron diameter particles. The density of uranium dioxide is 318 mg/cm3, A fuel element is shown in Figure 1 and is constructed as follows: -

1. Bottom aluminum mounting section

2. Five inches of carbon reflector

3. Ten inches of fuel

4. Five inches of carbon reflector

5. Two inches of lead shadow shield

6. The top aluminum removal fitting 5.4 Safety and Control Rods The AGN 211P has a fine and a coarse control rod plus two safety rods. The safety rods are made of Boral, the coarse control rod of aluminum clad Cadmium and the fine control rod of stainless steel. The rods are in the form of 2-3/4" wide blades which fit into slots in the edge of four specially constructed fuel elements.

During operation, the rods are held in the out position by eletro-magnets. A scram signal deenergizes the holding magnets allowing the safety and coarse control rods to be accelerated into the core by both gravity and spring loading. The spring provides an initial acceleration of Sg which gives an injection time of less than 500 17

310 58 7

.'. MODEL 2H FUEL ELEMENT ASSEMBLY ..

.' a

STANDARD

?

• FUE ELEMENT

r. Handle

. L ,

CONTROL & SAFETY

, e Lead FUEL ELEMENT h,), j, *) Graphite i

I. ,, ,

ASSEMBLY I N s ,.. g s s S

-- - 1 f

I x M. N( \%%

Tie Rods i 9

s

' -  ; ~T Control or I I Safety Rod

, Fuel Blade

!. q ,,

j

,t _f

- l%i'. '" 'j 4 f /1 ,+ g ,

" ' ' \, 'l ,,

/

j

~

g ]

t

)

2 fh

\ ,i i I

j 1 .

7

.i  ! i

'w 1 Graphite

{'

l f"' hy

( [ \

's,.d....-

Dowel Hele

\M (not shown)

.'j l

M g'  !

,. ]p^ Fuel Element Tip ( .d )

j. 3 I Locating'

! . Cowel \.

l i' i li  !

I

e

! S

,h '

Grid g.

.l . . Fig. .L

- l8 .

-w - g

milliseconds. The fine rod is rigidly mounted to its carriage and is driven into the core by the automatic return system which drives the magnets to their initial postion. All rod operation is manual (no automaticcontrol).

The reactivity worth of the safety and control rods varies with core configuration from approximately 1.0 to 1.5% ak/k. The fine control rod is worth approximately 0.45% independent of core configuration.

5.5 Shielding Shielding of the core in the radial direction is provided by water plus 40" of ordinary concrete. Shielding in the vertical direction is provided by 5" of graphite, 2" of lead attached to the top of the fuel -

elements, and 4-1/2 feet of water.

~

5.6 Water Purification System The water purification system consists of a demineralizer and circulating pump to maintain the water purity. The water loop has a radiation monitor and conductivity meter, both outputs.being displayed at the console. The radiation monitor is connected to a scram circuit so that the reactor cannot be operated if activity in the water system exceeds the scram setting.

5.7 Experimental Facilities The AGN 211P is equipped with a 7/8" diameter horizontal glory hole through the core. Samples may be placed in the glory hole at varying positions in the core and reflector.

A horizontal access port through the reflector, and a horizontal beam port which faces the reflector are present.

The varied core and reflector configurations make it possible to 19

l

. test different types of reflector elements or to leave water filled voids in the core in which large samples may be placed for irradiation.

5.8 Instrumentation and Controls The neutron flux in the reactor is monitored by two neutron detectors located near the reactor core and corresponding flux indicators on the console. These detectors are connected respectively to a

, logarithmic micro-micreammeter and a linear'micreammeter. Strip chart

. recorders record the output of both channels.

Each of the indicators is connected to a scram circuit so that if any indicator reads above or below the~ respective limiting safety system settings, it de-energizes the holding magnets, and shuts down the reactor. A rate, or period scram, is connected to the logarithmic channel to prevent too rapid a power rise. Additional safety interlocks provide for reactor shutdown if the water level drops 9 inches below the

-top of the reactor tank.- the water temperature falls below 5'C, if the radioactivity level of the water rises above a preset level, or if the circulating pump loses power.

'The rods are controlled by manual hold-down switches, the two control rods having a normal and slo'w speed. The positions of the control rods are indicated to 0.01 cm on the console. Sequenced interlocks of the control rod drive switches require that only one safety rod be moved at a time, and that both safety rods be fully withdrawn before the coarse control rod can be moved.

- 5.9 Fuel Storage All fuel elements not in the reactor are normally stored in the fuel storage pit immediately adjacent to the reactor. Each fuel element is stored in a steel cylinder rigidly held in a subcritical configuration -

. 20 i

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

s .

when in an infinite water reflector. See Figure 2. Additional fuel storage in steel shipping casks is provided so that all 24 fuel elements may be teinporarily stored external to the core.

5.10- Reactor Room The reactor is located in Room 107A, in the Nuclear Engineering Laboratory, 905 Asp Avenue. This room has 20,000 square feet and is a ground floor room. The reactor pool is partially sunk into the floor.

The arrangement of the room is shown in Figure 3.

21

l

(, .

..; f.y

/ D / u ,

k .....t.. ._._.,,v 'Z

- i- - , ~_.__~ = -T N h :. - Q '

• $h' A'

-~

$: A h i

,. 1 , .

l DM iNb u _' 10 l f g $ cap n5gg 1 m ~n=1m

~ ..

3 a

a.m m n

9gg , ,

. h. , ' ~

-Qnye . .,  %, .a :3a

';:[ ,

~ '

' i., -L. .

i t

4

\ l  :

\

. _ \_ ,.. ',

J ---  % Am, j

. u e ~k 'r& o u go n 4Ki W

Fi- 5 a @ .- . :! -& 6 ..

K- Q -

gentpfmum iMDM ~TJh7 l :f

• s M MMM @ "'7'^d'. -

FIGURE 2. FUEL STOMGE UCKS 22

~

~

i i

I TO FIELCHOUSE

! O m = .y.

=

< _ - _ _m _ . . UP ,

g

=_= u w v.m 1

_. r_ m = u m _ b k l V 10 7A i I 10 3 10 8 REACTOR L ABORATORY 19 $j y 2 -

es o j

UP -

f I'*D' '

1079 h ~

J 10 5 l

,, I l / ~a h H i J =

ll l 0 i

l l Ll ,

h

' I l l -

E g

h

,,,, ios ~ ' *$ nc REACTOR I 10 6 q 2

- --- 9,

.d t.

e II

._ mm F I r

l } I 111 l'l

'I I

!{ p .

q (24 f it2 f ,; 41 4 11 5 11 6 11 7 tie I f :f i

lol r

- c TO E n ]I d FIELDHCUSE ,2 y T b~" ~-

c.: =i C

E

, 12 3 UP "

'f C I '

l I

S 1220 w - 122A H -i U

~I 1 i ll ) [ 12r 12 0 its f 126 q

122c e

[ iz? N : q 122e , [

l l

\t I~~J 1

p u---e--N 1 __ _

n R' Q

~~l ..________,l i '

m

{t rt 83 0 12 9

! i- I P 12C -o E L I48 147 8 146 14 4 .

F li h

.H a n > WJ 7 = in em i d Q - . _. ..U j t

y p g l-----

3s Iq

-- IL---

q-

, 3 3 ,,

/

L; -

, '- d 7 t r;

1r- lrl

'l m.

=

ll

=

is 1 e i; 1

s, l!

3, . I

?

r' k

131A 13 2 13 3 134 !l; .

i L__3 8 -- _ j i 140

L_3 9 {14 14 2 n

13 5 136 i. 13 7 -' '

1 1  :

d '-

h m =-, ,-

m -- - - .

i i Il .  %

e v ., , - . = . ,

4 ,i NI A

j

,! ,_\j ____,U M P .'

FIRST FLOOR ------

N l L___--___- J FIGURE 3. FLOOR PLAN OF REACTOR AREA IN NUCLEAR EDGINEERING BUILDING a.

o ,

6.0 ADMINISTRATIVE CONTROLS 6.1 Organization of Responsibility for the University of Oklahoma Nuclear Reactor - Administrative control of the reactor is diagrammed in Figure 4.

6.11 President - Chief Executive Officer of the University 6.12 Provost - Chief Academic Officer 6.13 V.P. for Administrative Affairs - Chief Administrative Officer 6.14 Dean of Engineering - Administrative Officer in charge of the College ,

4 of Engineering 6.15 Director, School of Aerospace, Mechanical and Nuclear Engineering (AMNE) - The Director is the head of the three degree programs, Aerospace, Mechanical and Nuclear Engineering. The reactor, while operated as a facility available to all appropriate university personnel, is deemed to be a facility for which the Director has administrative responsibility. The Director will provide proper financial and technical support to ensure operation of the facility in accordance with all requirements.

6.16_ Reactor Director - Responsible for *he general administration of the facility. In this capacity the Director shall have the authority and 1

responsibility for the facility and, within the limitations set forth by the facility license, shall make policy decisions on all phases of reactor operation, personnel training and is advised in all matters concerning reactor safety by the Reactor Safety Committee, the Radiation Safety Committee and the University Radiation Safety Officer. The 24

PRESIDENT UNIVERSITY OF OKLAHOMA VP ADMINISTRATIVE PROVOST AFFAIRS RADIATION ENGINEERING SAFETY DEAN 0FFICER l r J l I

' I RADIATION l DIRECTOR SAFETY ---g g SCHOOL COMMITTEE i g{E

, I I e j i l l I I I REACTOR I l- REACTOR SAFETY DIRECTOR COMMITTEE - - - - - -

REACTOR SUPERVISCR FIGURE 4. ADMINISTRATIVE ORGANIZATION OF UNIVERSITY OF OKLAHCMA REACTOR AGN 211P 1

REACTOR I STAFF  !

l 23 '

j

l l

I Reactor Director may serve as Reactor Supervisor.

6.17 Reactor Supervisor - Has over-all responsibility for the daily operation and maintenance of the reactor facility and training and supervision of reactor staff personnel. The Reactor Supervisor is responsible to the P.eactor Director.

All operations which include physical handling of the fuel will be directly supervised by a Senior Reactor Operator.

6.18 Reactor Safety Committee -

1. The Reactor Safety Committee (RSC) shall be composed of not less than five members appointed by the Vice-President for Administrative Affairs, two of whom will be the Radiation Safety Officer and the Reactor Director. -

2. A minimum of fifty percent (50%) of the membership shall constitute a quorum. A minimum quorum shall not be constituted with more than two

-committee members from the operating staff.

3. The committee shall have a maximum of three members appointed from the reactor operating staff.

4. Minutes of the meetin'g shall be taken and submitted to each member of the committee along with reports presented the committee.

5. Ad hoc subcommittees may be convened at the request of the chairman and will report their findings to the committee.

6. .The chairman shall be appointed by the Vice-President of l

Administrative Affairs. He shall be responsible to call the committee into session at least quarterly and shall make assignments as necessary to the members of the committee. The chairman will be responsible for the recording of the minutes and the dissimination of committee reports and minutes. The chairman or his designate shall have signature authority for the committee.

7. Committee members shall have duties and authority as assigned by the chairman.

8. The committee shall review and approve the comprehensive safety aspects of reactor operation. Review and approval by the RSC will include a determination that any previously unreviewed safety question does not pose a hazard to the reactor, reactor personnel, or the environment as defined in 10CFR 50. Its duties shall include, but not be limited to the following:

26

, ~

A) Review and approve safety standards associated with the use of the facility which are consistent with the technical specifications of the reactor.

B) Review and approve procedures as required by the technical specification for normal start-up, operations, shut-down, refueling or rearrangement of fuel, surveillance, requalification, emergencies, and security of the reactor.

C) Review and approve procedures for the modification and/or testing of replacement components associated with the reactor, reactor safety systems and reactor security systems.

D) Review and approve procedures for proposed experiments using the reactor or any of its irradiation facilities. A general class of sample irradiations will be covered under a single procedure that can be implemented by any senior reactor operator. The general procedure for sample irradiation shall be reviewed and approved by the R.S.C.

E) Review and approve standard procedures for audit and review of the surveillance, procedures required by the technical specifications, requalification program, emergency plan and security plan.

F) Audit performance at least annually of the reactor staff in the areas of reactor operation, facilities manual (procedures),

maintenance logs and test procedures, surveillance tests, requalification program, emergency plan and physical security plan.

G) Review all abnormal occurrences and violations of the technical specifications, evaluating the causes of such events, the corrective action taken and recommending measures to prevent the reoccurrence of the incident. Report their findings to all authorities as required by the technical specifications and University authority.

6.19. University Radiation Safety Officer (RS0) - Has over-all responsibility for radiological safety at the University and shall see that necessary radiation surveys are conducted in the Reactor and other associated laboratories.

The Radiological Safety Officer shall review the radiological s .ety aspects of reactor experiments and shall advise the Reactor D. ar in all matters relative to radiological safety of the reactor and its i

operations. The RSO will be a member of the Reactor Safety Committee 27

~

and of the Radiation Safety Committee.

6.20 Reactor Staff Qualifications - All Staff personnel associated with the reactor operations shall meet the minimum standard qualifications set forth in ANS 15.4 " Standard for Selection and Training of Personnel for Research Reactors." The Reactor Director shall see that Training and requalification programs are conducted as required.

6.21 Procedures - Detailed written procedures shall be provided for:

a. Normal Start-up

b. Normal Operation

c. Normal Shut-down

d. Refueling or rearrangement of fuel

e. Emergency procedures

f. Special maintenance which could conceivably affect reactor safety.

g. Surveillance procedures.

h. Experiment authorizations.

6.22 Record Retention - Records to be retained for a period of not less than six years are:

a. Operating logs

b. Check-out and shut-down forms

c. All data relevant to reactor performance

d. Maintenance operations

e. Facility radiation surveys

f. Records of experiments and irradiations performed 28

6 -

6.23 Records to be retained for the life of the facility -

i

a. Gaseous and liquid radioactive effluents intentionally released to the environs l

l b. Any off-site environmental monitoring surveys

c. Fuel inventories and transfers

d. Facility drawings

e. Training of personnel

f. Reactor Safety Committee minutes

g. Personnel dosimetry 6.24 Safety Limit Violation The following actions shall be taken in the event a Safety Limit is exceeded:

a. The reactor will be shut down immediately and reactor operation will not resume without the concurrence of the Reactor Safety Committee, the Reactor Director, the Director of AMNE, and the Nuclear Regulatory Commission.

b. The Safety Limit violation shall be reported by phone to the appropriate NRC Regional Office of Inspection and Enforcement, the Director of the NRC, and the Reactor Safety Committee no later than the next work day.

c. A Safety Limit Violation Written Report shall be prepared for review by the Reactor Safety Committee. This report shall describe the applicable circumstances preceding the violation, the effects of the violation upon facility components, systems or structures, and corrective action to prevent recurrence.,

d. The Safety Limit Violation Written Report shall be submitted to the NRC, and Reactor Safety Committee within 14 calendar days of the violation.

29

?

6.25 Reportable Occurrences Reportable occurrences, including causes, probable consequences, corrective actions and measures to prevent recurrence, shall be reported to the NRC.

a. Prompt Notification With Written Followup. The types of events listed shall be reported as expeditiously as possible by telephone or telegraph to the Director of the appropriate NRC Regional Office, or his designated representative no later than the first work day following the event, with a written followup report within 14 calendar days. Information provided shall contain narrative material to provide complete explanation of the circumstances surrounding the event.

(1) Failure of the reactor protection system subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reached the, setpoint specified as the limiting safety system setting in the technical specifications.

(2) Operation of the reactor when any parameter or operation subject to a limiting condition is less conservative than the limiting condition for operation established in the technical specifications.

(3) Abnormal accumulation of fission products in pool water.

(4) Reactivity balance anomalies involving:

a. exceeding excess static reactivity limit;

b. shutdown margin less conservative than specified in these technical specifications; 30

Ed o ,

(5) . Failure or malfunction of ~one (or more) component (s) which prevents, or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents.

_ (6) Personnel error or procedural inadequacy which prevents, or could prevent, by itself,:the fulfillment'of the functional requirements of systems required to cope with accidents.

(7) Errors discovered in the transient or accident analyses cr in the methods used for such analyses as described in the Safety Analysis Report or in the bases for the Technical Specifications that have' permitted reactor operation in a manner less conservative than assumed in the analys'es.

(8) Performance of structures,. systems, or components that requires remedial. action or. corrective measures to prevent operation in a manner less conservative than assumed in_the accident analyses in the Safety Analysis Report to technical specifications bases; or discovery during plant life of '.onditions not specifically considered in the Safety analysis Report or Technical Specifications that require remedial action or corrective measures to_ prevent the existence or development of an unsafe condition.

f 1

31 i

, . -...,.-..-m..,. _ . _ . .. - ._-..,_r,,-- m... _- - . r..-_._e.-.. , - . _ _ - , . , , . .__._-,_-.r,.. . .. , - . , . - . , , ry -.r-- - - - , . - - .