ML20198L708

Amend 13 to Proposed TS for UTR-10 Reactor Facility at IA State Univ Possession Only License

ML20198L708
Person / Time
Site:	University of Iowa
Issue date:	12/28/1998
From:	IOWA STATE UNIV., AMES, IA
To:
Shared Package
ML20198L662	List: ML20198L657 ML20198L708
References
NUDOCS 9901050099
Download: ML20198L708 (42)

Text

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l TECHNICAL SPECIFICATIONS for the i

UTR-10 REACTOR FACILITY  :

at I IOWA STATE UNIVERSITY Docket No. 50-116

! License No. R-59 i

Possession Only License Amendment #13 l

Submitted: December 1998 99o105o099 es122e F PDR ADOCK 05000116 P PDR y

Technical Specifications Table of Contents  ;

1. 0 Defi nit ion s . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 2.0 Safety Limits & Limiting Safety System Settings. .............. ... ............ ...... . 2-1  ;

2.1 Sa fe ty Li mit s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 1 2.2 Limiting Safety System Settings ..................... ........... ..... . .. ................ 2-2.

3.0 Limiting Conditions for Operations. .... .............. ...... ................ ...........3-1  ;

l 3.1 Reactor Core Parameters . ...... .. . ...... . . . . . . .. ... . .... .. .... . . .... . . . . . . . . . . .. . . . ...... . 3 - 1 3.2 Reactor Control & Safety Systems .. . ......... ........... ...... .. .... .. .......... ... .... 3-2 3.3 Coolant S ystems . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . . . . . . . . 3 -3 3.4 Confi nement . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . ..............................3-4 l 3.5 Ventilation Systems ..... . ....... . ...... ....... . ... . . ..... ...... . ...............3-5 3.6 Emergency Power..... ... .......... .. . .... . . . .. . ..............................3-6 3.7 Radiation Monitoring. . .. . . . ... ... . .. . . . . . . ... ... . ... . . ... ..... ...... . . . . ... . . .. . . . . . 3-7 3.8 Experi ments. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................3-9 4.0 Surveillance Requirements ..... .. ...... ... .... . . . .. .................... .... ... .. . .. . . 4-1 4.1 Reactor Core Parameters .... . .. .... .. .. ... .. . . . . . . . . . . ....................4-1 4.2 Reactor Control & Safety Systems...... .. . ... . . . . . . . . . . . . . . . . . . . . . . . . .4-2 4.3 C oolant S ystems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4 - Confinement.............. .. . . . . . . . . . . . . . . . . . . . . . . . . . ..............................4-4 4.5 Ventilation Systems...... ... ..... ...... ... . . . . . . .......................4-5 4.6 Emergency Power......... .. . . ......... ......................................4-6 4.7 Radiation Monitoring. .. .... . .. . . .. .. . . . . . . . . . . . . . .. . ... . . . . . . . .. . . . . . . . . . . . . . . . 4-7 4.8 - Experi m e nt s. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . 4 -8 i

5.0 Design Features. . . . . . ... .. . .. . . .. . .. . . . .. . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . ......5-1  !

5.1 Site & Facility Description ........ ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Reactor Coolant System...... ... .. ... ..... .. ...............................5-2 ,

5.3 Reactor Core, Fuel and Safety System ... . .. ............ . .. . .. .. ...... ... .. . 5-3 5.4 Fissionable Material Storage.. ... .... ........ ... . . . . . . . . . . . . . . . . . . . . . . ... 5-5 6.0 Administrative Controls . . . .. . .... ... . . .. .. ... . . ...... .. . . . . . . . . . . . . . . . .. 6-1 6.1 Organizatio n . . . . . . ... . . . . .. . . . . . . . .. . . . . . . .. . . .. . . . . . . ....................6-1 6.2 Review & Audit..... . ......................................................6-3 i Proced u res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-

. 6.3 6.4 Experiment Review & Approval... .. .... . . . . . . . . . . . . . . . ..................6-6 l

6.5 Required Actio ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 6.6 Reports...................................................................... .6-8 6.7 Records .. . . . . . . . .... . .. ... . . . . . . . . . . . . . . . . . . . . . .. .. .6-10

-- Figure 6.1 Organizational Structure .............. . ..................... ..................6-11 4

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i l 1.0 DEFINITIONS i

l 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 CALIBRATION - The adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the j 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 ,

observation of channel behavior. This verification, where possible, shall include the l comparison of the channel with other independent channels or systems measuring the same variable.

CONFINEMENT . BOUNDARY - The surface surrounding the reactor facility defined by  ;

the interior partition walls of ofIices and laboratories on the north, east and south sides of i the building and by the west interior wall which isolates the basement, first floor, and the 1

west corridor of the second floor from the central bay.

CONFINEMENT SECURED - The confinement shall be considered secured when:

a. Doors 1-2, CC 102,101,114,113-1, CX112,112,111-1, CX111, CS107, CEI17-2, CC201B,112A-4 are closed or are attended by a person with the ability to close the door in the event of an emergency, and

b. Windows on the north, south, east and west sides of the penthe se, on the west i wall of room 112 A, on the south wall of room 101, on the east wall above door CX112, on the south wall of corridor CC201, on the west wall cf corridor CC212, on the nonh wall of corridor CC21 I are unbroken and cicsed or are attended by a person with the ability to close the window in the event of an emergency, and

c. The interior partitioned walls of the first floor offices and laboratones on the north, south, and east sides of the building; and the west interior wall which isolates the basement and first floor offices on the west side of the building from the central bay area; and the second floor north, south, east and west interior walls which isolate the second floor from the central bay area are intact and capable of performing as a non-pressure tight boundary, and i

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d. The roof covering the central bay area is intact and capable of performing as a non-pressure tight boundary.

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' CONTROL ROD - A plate fabricated with Boral as the neutron absorbing material 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.

DELAY TIMF - 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.

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 (excludmg 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, 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 l a part of their design.

MEASURED VALUE - The value of a parameter as it appears on the output of a channel, i 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 reactoris operating.  :

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

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

REACTIVITY LIMITS - Those limits imposed on reactor core excess reactivity.  :

I Quantities are references to a Reference Core Condition.

REACTIVITY WORTH OF AN EXPERIMENT - The maximum absolute value of the ,

reactivity change that would occur as a result ofintended or anticipated changes or credible malfunctions that alter experiment position or configuration.

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

I' shutdown.

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

1-l REACTOR SECURED - A reactor is secured when:

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

a. The minimum number of neutron absorbing control rods are fully

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

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j. c. No work is in progress involving core fuel, core structure, installed I

d. No experiments in or near the reactor are being moved or serviced that

have, on movement, a reactivity worth exceeding the maximum value j allowed for a single experiment or 0.763% Ak/k whichever is smaller.

I REACTOR. SHUTDOWN - The icactor is shutdown ifit is subcritical by at least 0.763%

Ak/k in the Reference Core Condition and the reactivity worth of all experiments is accounted for. i

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

' READILY AVAILABLE ON CALL - Applies to an individual who:

l (1) Has been specifically designated and the designation known to the operator on duty, and 1 1,

(2) Keeps the operator on duty informed of where he or she maybe rapidly ,

contacted (e.g., by phone, etc.), and

- (3) Is capable of getting to the reactor facility within a reasonable time under l l normal conditions (e.g.,30 minutes).

l ' 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. I i

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1 i,y-REGULATING ROD - 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 l by the servo-controller. j l

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

1 SAFETY-TYPE ROD - A rod that can be rapidly insened by cutting off the holdmg current in its electromagnetic clutch. This applies to safety #1, Safety #2, and shim-safety. ,

J 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.

i 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 denote 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 subcritical 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 subcritical without further operator action.-

' SURVEILLANCE TIME INTERVALS - The average over any extended period for each surveillance time interval shall be closer to the identified surveillance time than to the maximum allowed time, e.g., for the annual interval the average shall be closer to 12 ]

months than to 13 months.

l l Biennial- once every 2 years (interval not to exceed 25 months) I Annual- once every 12 months (interval not to exceed 13 months)

Bi-annual - once every 6 months (interval not to exceed 7 months)

Quanerly - once every 3 months (interval not to exceed 4 months) l Monthly - once every 30 days (interval not to exceed 6 weeks)

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~ Any extension of these intervals shall be occasional and for a valid reason and shall not affect the average as defined.TRUE VALUE - The actual value of a parameter or

' variable.

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$ ' UNSCHEDULED SHUTDOWN - Any unplanned shutdovm 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.

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' 2.0 SAFETY LIMITS AND SAFETY SYSTEM SETTINGS 2.1 Safety Limits 2.1.1- Applicability These specifications are not applicable.

The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized.

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2.0 SAFETY LIMITS AND SAFETY SYSTEM SETTINGS (Continued) 2.2 Limitina Safety System Settings l 2.2.1 Applicability These specifications are not applicable.

The UTR-10 reactor has been permanently de-fueled and is in possession only

' license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized.

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. 3.0 LIMITING CONDITIONS FOR OPERATIONS l

l 3.1 Reactor Core Parameters t

3.1.I ~ Applicability These specifications are not applicable.

l I..  : The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status.  !

Fuel handling operations are the only licensed activity allowed. Reactor operations I

are not authorized. '

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3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued) 3.2 Reactor Control and Safety System 3.2.1 Applicability.

These specifications are not applicable.  ;

The UTR-10 reactor has been permanently de-fueled and is in possession only  !

license (POL) status.-

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized.

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I 3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued) )

3.3 Coolant Systems l

3.3.1 Applicability  :

I These specifications are not applicable.  ;

i The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations i

are not authorized.

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I' 3.0 LIMITING CONDMONS FOR OPERATIONS (Continued) 3.4. Confinement 3.4.1' Applicability This specification applies to the operations that require confineinent and to the y equipment needed to achieve confinement.-

L 3.4.2 Objective E

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

. 3.4.3 Specifications L

l ~ A. The reactor confinement boundary shall be secured during fuel transfer operations.

i 3.4.4 Bases Specification A is based on the hypothetical accident (SAR: 6.4) that occurs during l movement of a fuel assembly and the importance of having the confinement boundary secured prior to the fuel transfer operation.

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i 3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued) l l

l-3.5 Ventilation Systems  :

l-There is no forced-air circulation system in the reactor room or the building housing it.

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, 3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued) 3.6 Emernency Power 3.6.l' Applicability l

These specifications apply to the emergency power supply for the radiation monitoring l system. _

l 3.6.2 Objective To specify the source of emergency electrical power and the minimum operating time. l h 3.6.3 Speqifications Fuel transfer operations shall not be nerformed unless the following conditions exist:

4 A. The battery-powered standby AC power supply for the radiation monitoring system '

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shall be operable and shall have the following operating time capabilities:

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

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

3.6.4 Bases ,

Specification A requires that the standby AC power system, which consists of at least '

- two lead-acid storage batteries, a charger transfer unit, and an invener, be capable of providing a tripless switchover for supplying AC power to the radiation monitoring ]

system, 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 i power (deletion). The radiation evacuation horn imposes a large incremental load on i 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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Y 3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued)

L 3.7 Radiation Monitorina Systems and Effluents 3.7.1 Applicability l These specifications apply to the radiation monitoring systems and to the limits on i effluent releases.

3.7.2 Objective l

t To specify the minimum number of acceptable components or the lowest acceptable

' level of performance for the radiation monitoring systems and the limits for release of effluents.-

L 3.7.3 Specifications Fuel transfer ooerations shall not be oerformed unless the following conditions exist:

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

components, and their setpoints.

i 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 i whenever fuel transfer operations are being performed which could produce  ;

o airborne radioactivity. The alarm setpoint is influenced by the normal j background reading. (Deletion.)

e (Paragraph deleted.)

• - The radiation area monitors are placed on the walls adiacent to the fuel star.agg pit (Sentence deleted.) The south and wesi units initiate an evacuation alarm at or above (phrase deleted) the radiation-evacuation setpoint of 5 mR/h The5 mR/h limit is based on the minimum value permitted for monitoring SNM in storage and applies when the area is unattended (phrase deleted). '

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

• The radiatio., 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.

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I 3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued)  :

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

l Channel Setpoint . Min. Operable Function Air. Particulate unit (a) As :equired 1 Alarm

- - - - - - - - - - - - - - - -(Table Entry Deleted) - - - - - - - - - - - - - - - - - - -

Area units 5 mR/h 2 ~ Alarm Doorway monitor -

1 Warn of abnormal radiation level.

Environmental film -

1 Integrated dose in restricted area.

(a) This unit is activated whenever fuel transfer ooerations are being performed.

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3.0 LIMITING CONDITIONS FOR OPERATIONS (Continued) 3.8 Exoeriments l 3.8.1 Applicability l

l- These specifications are not applicable.

f The UTR-10 reactor has been permanently de-fueled and is in possession only l- . license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations l are not authorized. <

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l~ 4.0 SURVEILLANCE REQUIREMENTS i l

4.1 Reactor Core Parameters 4.1.1 Applicability These specifications are not applicable. )

The UTR-10 reactor has been permanently de-fueled and is in possession only I license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized.

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4.0 ' SURVEILLANCE REQUIIEMENTS (Continued) )

4.2 Reactor Control and Safety System f 4.2.1 Applicability -  :

uTnese specifications are not applicable. l t

The UTR-10 reactor has been permanently de-fueled and is in possession only l license (POL) status.  ;

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized. ,

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4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.3 Coplant Systems P

4.3.1 Applicability These specifications are not applicable. -

The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized. ,

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l , 4.0 SURVEILLANCE REQUIREMENTS (Continued)

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4.4 Confinement l t I

. 4.4.1 Applicability This specification applies to the surveillance activities required for the reactor l- confinement.

4.4.2 Objective  ;

! - To specify the frequency and type of testing to assure that the reactor confinement  !

conforms to the specifications of section 3 of these Specifications.- 4 4.4.3 Specification  ;

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A. The doors and windows in the confinement boundary shall undergo testing for normal

! closure at least once every quarter, i

! 4.4.4 Bases 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 r i

verify operability based on experience at this facility.

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~- 4.0 SURVEILLANCE REQUIREMENTS (Continued) '

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4.5 Vantilation Systems

. This specification does not apply'to this facility.

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i 4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.6 Emergency Power 2 4.6.1 Applicability These specifications apply to the surveillance activities required for the emergency power system.

4.6.2 Objective l

To specify the frequency and type of testing to assure that the emergency power system l

conforms to the specifications of section 3 of these Specifications.

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

J A. The battery-operated AC standby power supply shall be tested for switchover action, l

and for voltage and specific gravity characteristics at least quarterly.

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

4.6.4 Bases 1 Specification A requires 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 batter performance within the testing interval.

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 Monitorina System and Effments 4.7.1 Applicability These specifications apply to the surveillance activities required for the radiation monitoring system and efUuents released from the facility.-

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

4.7.3 Specifications These surveillance activities 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 completed.

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

C. The environmental film badge cited in Table 3-3 and smear surveys in and around the reactor enclosure shall be analyzed at least quarterly.

4.7.4 Bases Based on experience at this facility and the average usage pattem of the reactor, specification A-C are adequate to verify that the operations conform to the specifications of 3.7.3 4-7

4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.8 Exoeriments 4.8.1 Applicability These specifications are not applicable.

The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operations are not authorized.

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5.0 DESIGN FEATURES p-5.1 Site aN Facility Description The reactor is housed in the Nuclear Engineering Laboratory, which is located on the  !

west edge of the campus ofIowa 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 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 appe ximately 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 nonh, 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 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 defmed by the interior panition walls of offices and laboratorie:: 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 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 building is estimated to restilt 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 significant penetrations into the enclosure consist of roof-level windows on the north and south sides which can be raanually opened or closed, as a group per side, in less than one minute per group.

9 5-1

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 F) 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 F at 10 kW and 10 gpm); approximately 92 gallons are contained in the piping and core tanks during operation. A quick-opening dump va!ve 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 *F) in parallel with the main loop, are provided (Ref: Drawing Rl-D-130). The operating temperature may range from about 80 F to no more than 160 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-116). The primary coolant system is essentially all-aluminum in construction; the pump casing and impeller, some valve pats, the dump tank heater element, and the 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 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 R1-E-151). A sump, with a capacity of 9.5 gallons and a manually energized sump pump, can discharge liquids from 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.

5-2

h 5.0 DESIGN FEATURES (Continued) 5.3 Reactor Core. Fuel 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 3.8.3 (Experiments).

A rabbit tube, no larger than 10 cm outside diameter, penetrates the graphite reflector at the west face of the nonh core tank. A neutron startup source, providing a minimum of 1.0 E+6 neutrons /second is insened into the coupling region by means of the source positioner during the startup operation (Ref: Drawings R1-E-154 and Rl-E-161).

Fuel 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 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. (Ref: Drawing RI-A-121-1)

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 RI-R-212, Rl-R-213, and 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 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 the minimum moderator level setpoint is exceeded, and it provides a signal (scram) when the moderator level exceeds the high level setpoint.

5-3

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5.0 DESIGN FEATURES (Continued)

A neutron-sensitive power level measuring channel with a functional range of 1.0 E-7 to

! 1.5 E+2 percent power, based on 10 kilowatts thermal, provides a signal (interlock)  !

t 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.

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' 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.

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aa 5.0 DESIGN FEATURES (Continued) 5.4 Fissionable Material Storane 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 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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6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Stmeture 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:

L Level 1: Responsible for the facility license and site administration.

l Level 2: Responsible 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.

l. 6.1.2 Responsibility L

l l The Facility Director shall be responsible for the facility license and site administration. The dean, College of Engineering, shall appoint persons, qualified in accordance with paragraph 6.1.4, to the Facility Director and Reactor Manager j positions.

I Individuals at the various management levels shown in Figure 6-1, in addition to having responsibility for the policies and operation of the facility, shall be

. responsible for safeguarding the public and facility personnel from undue radiation ll j exposures 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 designed alternates, or by higher levels, conditional upon appropriate qualifications.

6.1.3 Staffing (1) The minimum staffing during fuel transfer operations shall be:

~a. A licensed senior reactor operator in the control room.

b. A health physics-qualified individual in the control room.

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c. A aualified fuel transfer eauioment ooerator.

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6.0 ADMINISTRATIVE CONTROLS (Continued) i l

(2) Events requiring the direction of a senior reactor operator: i

a. Fuel transfer operations. I i

(3) Events requiring the presence of a health physics-qualified individual: l l a. Fuel transfer operations.

b. Any activity (deletion) that involves removal of a shield plug or closure.

( c. Any (deletion) activity (deletion) that could cause an abnormal release of  !

l radioactive materials.  !

6.1.4 Selection and Training of Personnel The selection, training and requalification of operations personnel shall meet or  !

exceed the requirements of American National Standard for Selection and Training l l of Personnel for Research Reactors, ANSI /ANS-15.4-1988, or its successor, meet i or exceed the requirements set forth in 10 CFR 55, and be in accordance with the  !

( Requalification Plan approved by the Nuclear Regulatory Commission.

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i l 6.0 ADMINISTRATIVE CONTROLS (Continued) t 6.2 Review and Audit

, The Reactor Use Committee (RUC) shall perform the independent review and audit of the safety aspects of reactor facility operations.

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

( members shall also represent university management. One of the three other members

shall be the University Radiation Safety Officer (RSO). The chair of the committee shall be appointed by the Provost. The terms on the committee for the RSO and chair are indefinite. All others are for three years with reappointments being determined by the Provost.)

5.2.2 Charter and Rules (1) The Reactor Use Committee shall meet at least semiannually and more frequently as circumstances warrant, consistent with effective monitoring of facility activities. Written records ofits meetings shall be kept and copies forwarded, in 1 l a t'.mely manner, to the University Radiation Safety Committee.

L l (2) A quomm shall be three members. Members of the operations staff shall not be a voting majority. Phone polling of members is allowed for final approval ofitems l

! discussed at a meeting or for approval of other items deemed " routine" by the Reactor Manager or the committee chair. Any member can veto the use of the phone poll and request a meeting of the committee.

l (3) Any action recommended by the Reactor Use Committee that may adversely affect the operations and/or safety of the University community shall be reponed by the RUC chairman to the University Radiation Safety Committee which shall have veto power over such a recommendation.

(4) The Reactor Use Committee may appoint one or more qualified individuals to perform the audit function.

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6.0 ADMINISTRATIVE CONTROLS (Continued) .

6.2.3 Review Function ,

The following items shall be reviewed (1) Determinations that proposed changes in equipment, systems, tests, (deletion)

or procedures do not involve an unreviewed safety question.

(2) All new procedures and major revisions thereto have safety significance and '

proposed changes in reactor facility equipment, or systems having safety L significance.

(Old #3 deleted.) i (3) Propor.ed changes in the Technical Specifications or the Operating License.

(4) Violations of the Technical Specifications or the Operating License. Violations ofinternal procedures or instructions having safety significance. ,

(Old #6 deleted.)

(5) Reponable occurrences listed in 6.6.2.

(6) Audit reports.

6.2.4 Audit Function The audit function shall include selective (but comprehensive) e:: amination of l operating records, logs, and other documents. Discussions with cognizant personnel and observation of operations should also be used as 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 Use Committee within 30 days after completion of the audit. The following items shall be audited. ,

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

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

(3) The results of action taken to correct those deficiencies that may occur in the ,

reactor facility equipment, systems, structures, or methods of operations that l affect safety, at least once per calendar year (interval between audits not to l exceed 15 months).

(4) The reactor facility Emergency and Physical Security Plans and implementing ,

procedures at least once every other calendar year (interval not to exceed 30 l months).

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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 be made effective only after documented ,

review by the Reactor Use Committee and approval by the Reactor Manger or a designated alternate. Minor modifications to the original procedure which do no change their original intent may be made, but the modification must be approved by the Reactor Manager or a designed alternate within 14 days. Temporary deviations from the procedures may be made by the on-duty SRO in order to deal with special or  :

unusual circumstances or conditions. Such deviations shall be documented and reported to the Reactor Manger or a designated alternate. Several of the following activities may be included in a single manual or set of procedures or divided among various manuals or procedures:

(Old #1 deleted)  ;

(1) Fuel element manipulations.

(Old #3 deleted)

(2) Surveillance tests and calibrations required by the Technical Specifications or those that may have an effect on safety. i (3) Personne'. nf.ation protection consistent with applicable regulations. l l

(4) Administrative controls for operations and maintenance and for the conduct of j irradiation and experiments that could affect safety.

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

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- 6.0 ADMINISTRATIVE CONTROLS (Continued) l 6.4 Experiment Review and Approval ,

This section is not applicable.

The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status. ,

! Fuel handling opera tions are the only licensed activity allowed. Experiments are not authorized.

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6.0 ADMINISTRATIVE CONTROLS (Continued) 6.5 Required Actions 6.5.1 Action to be Taken in Case of Safety Limit Violation This section is not applicable.

1 The UTR-10 reactor has been permanently de-fueled and is in possession only license (POL) status.

Fuel handling operations are the only licensed activity allowed. Reactor operation is not authorized.

6.5.2 Action to be Taken in the Event of an Occurrence of the Type Identified in 6.6.2(1)a and 6.6.2(1)h.

(Deleted old #1.)

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

i. (2) Occurrence shall be reviewed by the Reactor Use Committee at its next i

scheduled meeting.

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6.b ADMINISTRATIVE CONTROLS (Continued) l-

6.6' Reports i

6.6.1 Operating Reports l .I A routine operating report providing the following information shall be submitted l to the Nuclear Regulatory Commission in accordance with the provisions of 10 l CFR 50.59 at the end of each 12-month period

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. (Deleted old #1 and old #2.)

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

(2) Tabulation of major changes in the reactor facility and procedures that are ,

h significantly different from those performed previously and are not described I in the Safety Analysis Report, including conclusions that no unreviewed safety i questions were involved. i l - (3) A summary of the nature and amount of radioactive effluents released or l 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 summary shall include to the extent practicable an estimate ofindividual radionuclides  ;

o present in the effluent.

l l-(4) A summarized result of any envirocmental surveys performed outside the j facility. ,

l (5) A summary of exposures received by facility personnel and visitors where such l exposures are greater than 25 percent of that allowed or recommended. j 6.6.2 Special Reports l l

l_ (1) There shall be a report no later than the following working day by telephone to .

l_ the appropriate NRC Regional Office and confirmed in writing by telegraph or  ;

l similar conveyance to the Nuclear Regulatory Commission, in accordance with  ;

l instructions in 10 CFR 50.4, to be followed by a written report that describes the ,

circumstances of the event within 14 days of any of the following:

L . (Deleted old item a.)

a. Release of radioactivity from the site above allowed limits (see 6.5.2).  !

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6.0 ADMINISTRATIVE CONTROLS (Continued) j

b. Any of the following (see 6.5.2)

(Deleted old items: (i),(ii),(iii) and (iv).)

(i) Abnormal and significant degradation in reactor fuel, or cladding, or both, or coolant boundary which could result in ex4eeding prescribed radiation exposure limits of personnel or environment, o'r both.

(ii) 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 fuel l transfer operations.  ;

(2) A written report within 30 days to the Nuclear Regulatory Commission in i

accordance with instructions in 10 CFR 50.4, concerning the following:

a. Permanent changes in the organization involving the Facility Director, Reactor Manager, or Radiation Safety Officer.

b. Significant changes in the transient or accident analysis as described in the  !

Safety Analysis Report.  ;

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' 6.0 ' ADMINISTRATIVE CONTROLS (Continued) I L  !

6.7' Records i 6.7.1 Records to be Retained for a Period of at Least Five Years or for the Life of the l- Component ifLess than Five Years ,

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' (1) Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc., which shall be maintained for a period of at least -

one year).

- (2) Principal maintenance operations. j (3) Reportable occurrences.

(4) Surveillance activities required by the Technical Specifications t .

(5) Reactor facility radiation and contamination surveys where required by i applicable regulations..

(6) Experiments performed with the reactor.

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- (7) Fuel inventories, receipts, and shipments. ,

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(8) Approved changes in operating procedures.--

i (9) Records of meetings and audit reports of the Reactor Use Committee.

~ 6.7.2 Records to bh Retained for at Least One Training Cycle  !

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" Retraining and requalification oflicensed operators: Records of the most recent j complete cycle shall be maintained at all times the individual is employed.

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, may be used as records in this section.

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

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

(3) Radiation exposure for all personnel monitored.

I (4) Drawings of the reactor facility.

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6.0 ADMINISTRATIVE CONTROLS (cont.).

'1 LEVEL 1 L University President V

L- V t Provost Vice President l Business & Finance V V' V Dean of Radiatien Safety Director of Engineering Committee Environmental Health & Safety V V  ;

Facility Reactor Use .

Director - Committee l2 l

V V f LEVEL 2 l Radiation  !

Safety Oflicer I Reactor Manager l

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LEVEL 3 y 1 e tor l Health Physics Ope. :tions Staff l l Stiff  ;

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< 1 Committee membership  !

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