ML20043B411
| ML20043B411 | |
| Person / Time | |
|---|---|
| Site: | 05000113 |
| Issue date: | 05/22/1990 |
| From: | Holahan G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043B410 | List: |
| References | |
| R-052-A-015, R-52-A-15, NUDOCS 9005290278 | |
| Download: ML20043B411 (42) | |
Text
-
.. J
... q$ 58%q'o UNITED STATES
[#
~g NUCLEAR REGULATORY COMMISSION g
g E
W ASHING TON, D. C. 20555
[
t RENEWAL OF FACILITY-LICENSE DOCKET NO. 50-113 UNIVERSITY OF ARIZONA Amendment No.115-License No. R-5E'
- 1. _
The Nuclear Regulatory Comission (the Comission) has found that:
The application for renewal of Facility) License No.- R-52 filed by A.
the University-of Arizona (the licensee dated October 7.7, 1988, as-supplemented on July 17, 1989, September 15, 1989, and January 30, 1990, complies with the standards and requirements of the Atomic s
Energy Act of 1954,- as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B.
Construction'of the: facility was completed in substantial conformity with Construction-Permit.No. CPRR-30 dated November 21, 1958 and CPRR-111 dated January 13, 1971, the provisions of the Act, and the.
regulations of the Comission; C.
The facility will operate in conformity with the application, the-provisions of--the Act, and the regulations.of the 'Comission; D.
There is reasonable assurance:
(1)thattheactivitiesauthorized by this. license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; E.
The licensee is technically and financially qualified.to engage in the activities authorized by this operating license in accordance with the regulations of the Comission; F.
The licensee is a nonprofit educational institution and will use the facility-for the conduct of educational-activities, and has satisfied the applicable provisions of 10 CFR Part 140, " Financial Protection. Requirements and Indemnity Agreements," of the Comission's regulations; G.
The issuance of this license will not be inimical to the comon defense and security or to the health and safety of the public; H.
The issuance of this license is in accordance with 10 CFR Part 51 of the Comission's reguttians and all applicable requirements have been satisfied; and
/q
\\
' 9005290;78 900522 PDR-ADUCK 05000113 P
,3 n
I.
The receipt, possession and use of the byproduct and special nuclear materials as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30. and 70, including Sections 30.33, 70.23 and 70.31.
2.
Facility License No. R-52 is hereby emended in its entirety to read as follows:
h.
A.
The license applies to the TRIGA Mark I nuclear reactor (the facility) owned by the University of-Arizona (the licensee). The f acility is ' located on the licensee's site in Tucson, Arizona, and is described in the licensee's application for-renewal of the
-0
' license dated 0ctober 17; 1988, as supplemented on July 17, 1989, September 15,11989, and January 30, 1990.
B.
Subject to the conditions and requiremente incorporated herein, the Commission hereby licenses the University of Arizona:
(1) Pursuant to Section 104c of the Act and 10 CFR Part 50, "Do:r.estic Licensing of Production and Utilization Facilities,"
to possess, use, and operate the facility at the designated location in Tucson, Arizona, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70 " Domestic Licensing of Special Nuclear Material," to receive, possess and use up to 1
3.5 kilograms = of uranium-235 contained in uranium enriched to less than 20% in the isotope uranium-235 in connecticn with operation of the facility.
(3) Pursuant to the Act and 10 CFR Part 30 " Rules of General Applicability to Domestic Licensing of Byproduct Material," to
. receive, possess and use a 5-curie sealed americium-241-beryllium neutron source in connection with. operation of the facility.
(4) Pursuant to the Act and 10 CFR Parts 30 and.70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in Parts 20, 30, 50, 51, 55, 70 and 73 of 10 CFR Chapter I, to all applicable provisions of the Act, and to the rules, regulations and orders of the Commission now or hereafter in effect and to the additional conditions specified below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of 110 kilowatts (thermal).
a
=
g 4, st
. s
. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 15, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical-Specifications.
(3) Physical Security Plan -
The licensee shall maintain in effect and fully implement provisions of the Commission-approved physical security plan, including all amendments and revisions made pursuant to the authority of 10 CFR -50.90 and 10-CFR 50.54(p), which are part of the license. This plan, which contains information withheld from public disclosure under 10 CFR 2.790 is entitled " Physical Security Plan for the University of Arizona Research Reactor" dated December 8, 1989.
D.
This license is effective as of the date of issuance and shall expire twenty years from its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Y R h Gary M. Holahan, Acting Director Division of Reactor' Projects - III, i
. IV, V and $pecial Projects Office of Nuclear Reactor Regulation
Enclosure:
Appendix A. Technical Specifications
'Date of Issuance:
May 22, 1990
a A
.,- 4 r,
TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF ARIZONA
.TRIGA RESEARCH REACTOR y
FACILITY LICENSE R-52 This document includes the Technical Specifications and the bases for the n
_ Technical Specifications. The bases provide the technical support for the individual Technical Specifications and are included for information purposes only. The bases are not part of the Technical Specifications and they do not.
constitute limitations or requirements to which the licensee must adhere.
i Amendment No.15 l
__.j h;, e. 7 a,7 i
{
,;.u.
INDEX Page Number T
1.0 DEFINITIONS 1
- 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 5
2.1 Safety Limit - Fuel Temperature 5
-b 2.2, Limiting Safety System Setting - Steady State Reactor Power Level
'6 2.3 Limiting Safety System Setting - Pulse Mode Reactor Pcwer Level 7
3.0 LIMITING CONDITIONS FOR OPERATION 8
3.1 Reactivity Limits 8
3.2 High Power Operation 10 3.3 Pulse Operation i1 3.4 Reactor Instrumentation 12 3.5 Reactor Safety System 13 3.6 - Ventilation System 15 3.7 Experiments.
16 4.0 SURVEILLANCE REQUIREMENTS 17 4.1 Fuel 17 4.2 Control Rods 18 4.3 Reactor Safety System 19 4.4 Radiation Monitoring Equipment 20 4.5 Maintenance 21 4.6 Pool Water Conductivity 5.0 DESIGN FEATURES f
5.1 React < Fuel 23 5.2 Reactor 23ullding 24 5.3_ Fuel Storage 25 4
(
l i
. _.... _...;g _...
.,.2 a, j',.
{ '.' 'i 6.0 ADMINISTRATIVE CONTROLS '
26 6.1 Organization 26 6.2 Review 27 6.3.1 Operating Procedu:ca 28' 6.3.2 ALARA Program 28
. 6.4 Action to be Taken in the Event a Safety Limit is Exceeded 29 6.5 Action to be Taken in the Event of a Reportable Occurrence 30 6.6 Plant Operating Records 31 6.7 Reporting Requirements 32 6.8 Review of Experiments 35 O
e e
r
.4
e; i
i
^
.}.
1.0 DEFINITIONS Channel - A channel is a combination of sensors, electronic circuits, and output devices connected by the appropriate communications network in order to measure and display the value of a parameter.
P Channel Calibration - A channel calibration is an adjustment of a channel such that its output cor. responds with acceptable accuracy to known values of the parameter which the channel measures.
Calibration shall encompass the entite channel, including equipme,nt, actuation, alarm, or trip and shall include a Channel Test.
Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. The verification shall include comparison of the channel output with previous readings or performance or with other independent channels or systems measuring the same variable, whenever possible.
Channel Test - A channel test is the introduction of a signal into the channel for verification that it is operable.
Cold Critical - The reactor is in the cojd critical condition when it is critical with the fuel and bulk water temperatures the same (-20*C).
Experiment - An experiment is any device or material, not normally part of the reactor, which is -
introduced into the reactor for the purpose of exposure to radiation, or any open. tion which is designed to investigate non-routine reactor characteristics.
'Expe'rimental Facilities - Experimental facilities are the thermal column, pneumatic transfer systems, central thimble, rotary specimen rack, beam tube, and the in-core facilities.
Limitirig Conditions for Operation - Limiting Conditions for Operation (LCO) are administratively established constraints on equipment and operational characteristics which shall be adhered to during I
operation of the reactor.
Limiting Safety System Setting (LSSS) - The LSSS is the actuating level for automatic protective devices related to those variables having significant safety functions.
Manual Mode - The reactor is in the manual mode when the reactor mode selection switch is in the manual or automatic position. In this mode, reactor power is held constant or is changed on perieds of approximately one second or longer.
L Measured Value - The Measured Value is the value of a parameter as it appears on the output of a l
channel.
4 l'
l Movable Experiment - An experiment is movable when it is intended that all or part of the L
experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
l l
Operable - Operable means a component or system is capable of performing is intended function.
Operating - Operating means a component or system is performing its intended function.
l l
O 1
l
i.t i
.j t
- e Pulse Mode - The reactor is in the pulse mode when the reactor mode selection switch is in the
_ pulse position. In this mode, reactor power may be increased on periods less than one second by motion of the transient control rod.
Reactivity Worth of an Experiment - The reactivity worth of an experiment is the maximum value of the reactivity change that would occur as a result of planned changes or credible malfunctions that alter experiment position or configuratior..
Reactor Committee - The group of persons at the University who are assigned responsibility for review and audit of facility operation and review of changes and experiments in accordance with 10 CFR 50.59.-
A
- Reactor Operating - The reactor is operating whenever it is not secured or shutdown.
Reactor Safety Systems.- Reactor Safety Systems are those systems, including associated input channels, which are designed to initiate automatic reactor protection or to provide information for
. initiation of manual protective action.
Reactor Secured - The reactor is secured when:
a.
It contains insufficient fissile material or moderator present in the reactor, adjacent experiments or control rods,*to attain criticality under optimum available conditions of moderation and reflection, or b.
1.
The minimum number of neutron absorbing control rods are fully inserted or other safety devices are in shutdown position, as required by Technical Specifications, and 2.
The console key switch is in the off position and the key is removed from the lock, and 3.
No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and 4.
No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth of one dollar or more.
Reactor Shutdown - The reactor is in a shutdown condition when sufficient control rods are laserted to assure that it is suberitical by at least $1.00 of reactivity, h
Reportable Occurrence - A Reportable Occurrence is any of the following which occurs during reactor operation:
1 a)
Operation with actual safety-system settings for requi*ed systems less conservative than-the limiting safety-system settings specified in Techni s'. Specification 2.2.-
b)
Operation in violation of limiting conditions for operation establishd in the Technical Specifications.
L l
l
- 7,
~. n
= ' -
3-c)
- A reactor safety system component malfunction which renders or could render the reactor safety system > capable of performing its intended safety function unless the
~
malfunction or condition is discovered during maintenance tests or periods of reactor
- shutdown, d). '
Any unanticipated or uncontrolled change in reactivity greater than one' dollar, e)
Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary which could tualt in exceeding of prescribed radiation exposure or release limits, f)
An observed inadequacy in the implementation of either administrative or procedural controls which could result in operation of the reactor outside' the limiting conditions for operation.
g),
Release of radioactivity from the site above limits specified in 10CFR20.
Control Rod _ - A control rod is a device fabricated from neutron absorbing material or fuel which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.
Transient Rod - The transient rod is a control rod with scram capabilities that is capable of providing rapid reactivity insertion to produce a pulse.
Safety Limit - A Safety Limit is a limit on an important process variable which is found to ce necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity. The principal physical barrier is the fuel element cladding.
Secured Experiment - A Secured Experiment is any experiment, experimental 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 the result of credible malfunctions.
Shall, Should, and May - The word "shall* is used to denote a requirement, the word "shoulf denotes a recommeru:bu, and the word "may" denotes permission, neither.a requirement nor a recommendation.
Shutdown Margin - Shutdown Margin is the reactivity existing when the most reactive control rod is fully withdrawn from the core and the other control rods are fully inserted into the core.
Time Interval - The average over any extended period for each surveillance time item shall be the normal surveillance time, eg, for a two-year interval the average shall not exceed two years.
a)
Biennially - at two-year intervals (interval not to exceed 30 months) b)
Annually - at one-year intervals (interval not to exceed 15 months) c)
Semiannually - at 6-month intervals (interval not to exceed seven and one-hall aths)
.d)
Quarterly - at 3-month intervals (interval n:,t to exceed four months) e)
Monthly - at one-month intervals (ir.terval not to exceed six weeks)
- ~.. - - _ _ -. _ _... _ _ - _ _ _ _ _ _ _ - _. _ _ _ _ _ _ _ _
1 t
,g y:; i'lIO _ _s e' ;-
'- / g-
]
, =.
,4_.
. f)
Weekly ; at seven-day intervals (interval not to exceed ten days).
- 's) =
Daily - (must be done during the calendar day):
...-e.
Any extension of these intervals shall be o:casional and for a valid reason and shall not affect the-g" avenge as defined.
4 a
16 :-
-- Untried Experiment - An untried experiment is any experimer.t not previously performed in this reactor.
q
..t i.
t (i
I
- k ' 'q 9
{.
i 4
4 t
5
(
k
'j
'.\\.',
'1
.-i.
N f*
s
.r
., 3 9,
' 2.0 - SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETJir4GS 2.1 Safety igit - Fuel Temperature Applicability This specification applies to the reactor fuel temperature.
. Objective -
The objective is to define a fuel temperature below which it can be predicted with confidence that no damage to the fuel elements will occur.
Specification The temperature of the fuel shall not exceed 1000 oC under any conditions of operation.
Basis The recommen'ded limiting design basis parameter for TRIGA fuel is the fuel temperature. A fuel temperature safety limit of Il50*C for stainless-steel-clad U-ZrH,es TRIGA' fuel is i
re;inmended as a design value to preclude the loss of clad integrity when the clad temperature is below 500oC (Simnad, GA Report E-Il7-833, The U-Zr H Alloy: Its Properties and Use in TRIGA Fuel, Feb.1980, p. 4-1). The criterion for assuring the integrity of a taiga fuel eiement at the University of Arizona is that the fuel temperature be maintained below 1000*C, which is well below the recomrnended value. It has been shown by analysis and by measurements on other TRIGA reactors that a power level of 1000 kw corresponds to a peak fuel temperature of approximately 4000C. Pulsing with a reactivity input of $3.25 will give a peak fuel temperature of approximately 460oC,'
e
. ', i, -.
y 2.2 Limiting Safety System Setting - Steady State Reactor Power Level Applicability This specification applies to the reactor power level safety system setting for steady state
+
operation.-
Objective The obbe'ive is to assure that thi Safety Limit is not exceeded.
Specification The setting for the power level scram in steady state operation shall be no occater than 110 kw, Basis -
Calculations and measurements show that at 110 kwi the peak fuel temperature in the core will be less than approximately 150*C which is well below the safety criterion of 10000C and provides an ample safety margin to accommodate errors in measurement and anticipated operational transients.
e
.i
'/.
il /
'i-ys r
gy jf
~
V
^
' ~ ^ ^
^ ^
- 4 (
- ;-..,
.,[
14=
'~
7 is
- 2.3 ~ Limiting Safety System Setting - Pulse Mode Reactor Power Level Applicability This specification applies to the reactor power level safety system setting for pulse mode operation.'
Objective The objective is to assure that the fuel temperature specified by the Safety Limit is not exceeded in pulse mode operation.
Specification i
The setting for the peak power level scram in pulse mode operation shall be no greater than.
1100 Mw.
Basis Calculations and measurements show that at a peak power of 1100 $1w in pulse mode '
s operation, th( p';ak fuel temperature in the core will be less than approximately 400*C, This provides an Lt;;ple safety margin to accommodate errors in measurements and anticipated operational transients.
l l-l l
l..
j; i
l
~
~
n 1'
I l..
~..
1.
t g >.
{
24 1^
g.
3.0 - LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits
- Applicability These specliication: apply to the reactivity condition of the reae*or, and the reactivity worths
~
of control rods and experiments, and apply for all modes of reactor opera' ion.
Objective The objective is to assure that the reactor can be shut down at all times and to assure that the safety limit will not be exceeded.
Specifications
' a reactor shall not be operated unless the following conditions exist The shutdown margin referred to the cold xenon-free condition is greater than $0.50
.with the highest worth rod fully withdrawn and with the highest worth non-secured experiment in its most positive reactivity state, b.
Any experiment with a reactivity worth greater than $1.00 is secured so as to prevent unplanned reactivity removal from or insertion into the reactor; c.
The reactivity available to be inserted by the pulse rod is determined and is limited by a mechanical block to a maximum of $2.50.
d.
The reactivity worth of an individual experiment is not more than $3.00; l.
e.
The total of the abs'olute values of the reactivity worth of all experimenM in the reactor is less than $5.00; f.
A ramp or oscillating rod placed in the reactor cannot add more than $1.00 of reactivity; i
g.
The drop time of each standard control rod from the fully withdrawn position to 90 y
percent of full reactivity insertion is less than one seccnd; and l
h.
The neutron count rate on the startup channel is greater than one count per second.
i.
- The maximum reactivity insertion rate by control rod: for non-pulsed operation is less
{
l than $0.20/second.
e
~
i.
l l
I-t i
l ).
> w aw I
s
l y
9..
- J.
The maximum excess res:tivity does not exceed $3.25...
k.
The height of coolant water above the core is 14 feet or greater.
1.
. The bulk temperature of coolant wa.ter does not exceed 45' C.
rr..
The conductivity of the coolant water, averaged over 30 days, does not exceed 5 micromhos/cm.
Basis The shut down margin required by specification 3.la is necessary so that the reactor can be shutdown from any operating condition and remains shutdown after cooldown and xenon decav even if one control rod should stick in the fully withdrawn position.
Specification 3.lb is based on pulse measurements and analysis at the University of Arizona which indicate that as much as 53.00 reactivity could be inserted without increasing fuel temperature by more than 415'C By restricting each non-secured experiment to a reactivity worth of one dollar, an ample margin is provided.
Specifications 3.lc through 3.lf are intended to provide additional saargins between those values of reactivity changes encountered during the course of operations involving experiments and those values of reactivity which, if exceeded, might cause a safety limit to be exceeded, Specificatma 3.lg is intended to assure prompt shutdown of the reactor in the event a scram signai is receive,d.
Specification 3.th is hJended to assure that sufficient neutrons are available in the core to provide a signal at the output of the startup channel during approaches to criticality.
Specification 3.11 is based on analysis at the University of Arizona which demonstrates that with~ a reactivity ramp of S.20/seror.d starting at delayed critical at any power below 100 kilowatts, the maximum fuel temperature increase will be less than 750C, and thus will not exceed the Safety Limit.
i e
\\(,
g p
' ( e t-e '
e 30 -
3.2 High Power Operation '
s Applics'ility -
This specification applies to operation of the reactor at hi',h steady-state power.
Objective The objective is to prevent inadvertent pulse operation of the reactor while it is at a high power level.
Specification The reactor shall not be operated in the steady-state mode at power levels above 10 kw unless, in addition to the conditions of Section 3.1, the transient rod is fully withdrawn to the limit of its limiting switch.
Basis i
This specification is intended to prevent inadvertent pulse operation when the fuel temperature is above 500C (corresponding to a power level of 10 kw) as measured in the B-I ring. See Specification 3.3b.
4 P
1
?
4 3
I--
?
y,'
'i 1,.;
11 3.3 > Pulse Operation E
Applicability These specifications apply to operation of the reactor in the pulse mode, Objective The objective is to prevent the fuel temperature safety limit frorr being exceeded during pulse mode oper.ition.
Specifications The reactor shall not be operated in the pulse mode unless, in addition to the requirements of Section 3.1, the following conditions exist
- a. -
The transient rod is set such that the reactivity worth upon withdrawal is not greater than 52.50; and
[
b.
' The temperature of the fuel immediately prior to the pulse is essentially in equilibrium with the bulk water temperature. This is controlled by limiting the reactor power prior to pulsing.
Be's Specification 3.3a will maltain the maximum temperature of the fuel after a pulse below 400*C above the bulk pool temperature, and th'us well below the 1000oC fuel safety criterion.
e 9
N.
i K
-M.
S s__
s
'.D. s y,t '
12.
, g 3.4 - Reactor Instrumentation Applicability This specification applies to the information which must be available to the reactor operator during reactor operation.
Objective The objective is to require that sufficient information is availabie to the operator o assure-safe operation of the reactor.
Specification The reactor shall not be operated unless the measuring channels described in the following table are operable and observable from the control room:
Minimum Operating Mode Number in which Measuring Channel Operable Required.
Reactor Power Level (Linear)
I
' Steady State
- Wide-range Log Power 1
Steady State
- Level (Startup count rate)
Reactor Period 1
Steady State -
Reactor Power Level (high range) 1 Pulse Mode Reactor Tank Water Temperature 1
All Modes Area Radiation Monitors 2
All Modes Particulate Air Radiation Monitor 1
All Modes Reactor Water Activity Monitor 1
All Modes Basis The neutron detectors assure that measurements of the reactor power level are adequately -
covered at both low and high ranges.
The radiation monitors provide information to operating personnel of radiation above a preset level so that there will be sufficient time to evacuate the facility or take action to prevent the release of radioactivity to the surroundings.
i s
,, :. ;n y
? S..
q m
p.--
.. 33,
[
3.5 Reactor Safety System Applicability p-This specification applies to the reactor safety system channels and interlocks.
Objective The objective'is.to' require the minimum number of reactor safety system channels and interlocks that must be operable in order to assure that the safety limits and the LCO's are not exceeded.
Specification, The reactor shall not be operated unless the saf=ty system channels and interlocks described in tha following tab:es are operable.
Minimum Operating Mode l-Safety Systmem or Number in which Measuring Channel Operable
- Function, Required Setpoint Reactor Power Level 2
Scram Steady State not above 110 kw Reactor Period 1
Audible Alarm Steady State no shorter than 2 seconds
~ Peak Reactor Power 1
- Scram Pulse Mode
' not above 1100 Mw Manual Scram 1
Scram All Modes Pool Water Level-1
. Scram All Modes
' not less than 14 feet -
of water above core
-i
- Safety channel switched to "zero" or " calibrate" 2
Scram All Modes -
Timer after pulsing operation i
Scram Pulse Mode 15 seconds or less Power Failure 1
Scram All Modes loss of console power 3
8
,e.
3
'f:
no
'^
"o * ; 3 v.
e c:-
., -)
i Minimum Operating Mode Number in which Interlock Operable Function Required.
l 1
Startup Count Rate.
1 Prevent control Reactor Startup Interlock '
. rod withdrawal when neutron count rate is less than I per second
- j Transient Rod Interlock 1
Prevent withdrawal Steady State of a transient rod Mode when its shock ab-sorber anvil is -
not fully inserted -
N
)
" Simultaneous Control Rod 1
Prevent simultaneous All Modes l
Withdrawal Prohibit manual withdrawal of
. Interlock two control rods Reactor Power Level 1
Prevent transient Pulse Mode Interlock-rod withdrawal when power is greater than 10 kw Basis The powerxlevel scrams are provided in all modes of operation as protection against abnormally high foal temperatures and to assure that the reactor operation stays within the licensed limits. The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. The reactor period alarm alerts the operator to potential rapid transient power changes so Umiting actions may be taken. The pool water level scram assures--
that sufficient water shielding is above the core during reactor operation.
l E
The interlocks which prevent 'he withdrawal of the transient rod in ^ neady state mode and i
when the power level is greater than 10 kw prevent inadvertent p01ses. ~ The interlock to
, prevent startup of the reactor with less than one neutron per second 'ndicated on the startup channel assures that sufficient neutrons are available to provide indica:% on the measuring channels.
E, ~
5 e
l
t
^i-
~
4..
.a g_
j'.e
- i 15 -
3.6. -Ventilation System Applicability This specification applies to the operation of the reactor facility ventilation system.
.m a Ep Objective The objective is to assure that the ventilation system is in operation to mitipte the consequences of the possible release of radioactive materials resulting from reactor operation.
Specification L
The reactor shall not be operated unless the facility ventilation system is in operation with a minimum air withdrawal rate of 500 cfm except for periods of time not to exceed two days to permit repairs to the system. During such periods of repair:
o a.
The reactor shall not be operated at power levels above 10 kw and;
.[
b.
The reactor shall not be operated with experiments in place whose failure could result i
l in the release of radioactive gases or aerosols.
l.-
Basis i
It is shown in the Safety Analysis Report that operation of the ventilation system reduces doses in the reactor facility due to argon-41, and c!so in the event of a TRIGA fuel element L.
failure. The specifications governing operation of the reactor while the ventilation system is undergoing repairs limit the generation of argon-41'and also reduce the probability of fuel' element fe.ilure during such times..
i l
i' I
4 r i e
l
.4 e
1 4
- i i * *
-..-l-l
n
-i e.* ;..., ;
- 3.7 Experiments
-Applicability These specifications apply.to. experiments installed in the reactor and its experimental facilities.
j w
Objective The objective is to prevent damage to the reactor or excessive release of radioactive materials
~
in the event of an experiment failure.
Specifications-The reactor shall not be operated unless the following conditions exist :
fueled experiments shall be limited such that the total inventory of iodine isotopes 131 a.
through 135 in the experiment is not greater than 1.5 millicuries and the strontium-90 inventory is not greater than 5 millicuries; b.
experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, or liquid fissionable materiab shall be doubly encapsulated; and i
c.
known explosive materials shall not be irradiated in the reactor in quantities greater y
than 25 milligrams. In addition, the pressure produced in the experiment container upon detonation of the explosive shall have been determined experimentally, or by calculations, to be less than the design pressure of the container.
l
. Basi:,
The limits of Specification'3.7a prevent the dose in unrestricted areas'resulting from experiment failure from exceeding 10 CFR Part 20 limits. Calculations for the SAR
~
demonstrate that the maximum release in the event of a fuel element failure would not exceed 6.5 millicuries of iodine isotopes 131 through 135. Specifications 3.7b and 3.7c are provided to reduce the probability of damage to reactor components resulting from experiment failure.
l
[.
1 l
l h
O
(*
.. /
l' 17 4.0 SURVEILLANCE REQUIREMENTS 4.1-Fuel
' Applicability This specification applies to the surveillance requirements for the fuel elements.
Objective The objective is to assure that the dimensions of the fuel elements remain within acceptable
- limits, Specifications a.
All fuel elements shall be removed from the core and visually inspected for evidetae of deterioration of claddins, (including at least corrosion, erosion, wear, cracking, and weld integrity) at least once every five years, b.
The standard fuel elements shall be measured for length and bend at intervals separated by not more than 500 pulses of magnitude greater than $2.00 of reactivity, c.
A fuel element indicating an elongation greater than 1/4 inch over its original length or a lateral bending greater than 1/16 inch over its original bendin's shall be considered to be damaged and shall not be used in the core for f,urther operation.
d.
Fuel elements in.the B-and C-rings shall be measured for possible distortio.n in the event that there is indication that the Limiting Safety System Settings may have been exceeded.
Basis The most severa stresses induced in the fuel elements result from pulse operation with high reactivity input, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above ilmits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to cause rupture of a' fuel element.
e 0
w
=
e**
ll 6. -
p 18
,s 4.2 Control Rods _
Applicability.
This specification applies to the surveillance requirements for the control rods.
Objective The objective is to assure the operability of the control rods.
Specification a.
The reactivity worth of each control rod shall be determined annually.-
r b.
Control rod drop times shall be determined annually and after disassembly and reassembly of control rod drives or removal of control elements.
c.
The control rods shall be visually inspected for deterioration biennially.
d.
On each day that pulse mode operation 'of the reactor is planned, a functional performance check of the transient (pulse) rod system shall be performed prior to pulse mode operation, e.
Semiannually, the transient (pulse) rod drive cylinder and the associated air supply system shall be inspected, cleaned and lubricated as necessary.
f.
The maximum control rod reactivity insertion rates shall be determined annually.
Basis The reactivity worth of the control rods is measured to assure that the required shutdowm margin is available and to provide a means for determining the reactivity worths of-experiments inserted in the core. The visual inspection of the control rods and measurement of their drop times are made to determine whether the control rods are capable of performing their functions properly.
l
?
C, t. :;
.. 4 a.
e'
~
4.3 Reactor Safety System Applicability The specification applies to the surveillance requirements for the measuring channels of the reactor safety system.
}
Objective 1
1 The objective is to assure that the safety system will remain operable and will prevent the r?
fuel temperature safety limit from being exceeded.
Specification j
a.
A channel test of each of the reactor safety system channels required la the operating mode to be followed shall be performed prior to es;h day's operation or prior to each operation extending more than one day..
b.
A channel check of the power level measuring channels required in the operating mode to be followed shall be performed daily whenever the reactor is in operation.
c.
A channel calibration by the calorimetric method shall be performed for the reactor power level measuring channels annually.
~ Basis The daily tests and channel checks will assure that the safety channels are operable. The annual calibration and verifications will permit any long-term drift of the channels to be corrected.
O e
of e
~
^
~~
a
-..:, ue..-
- t 4.-g-z 20 4.4*
Radiation Monitoring Equipment Applicability
. This specification applies to the radiation monitoring equipment required by Section 3.4 of these specifications.
Objective The' objective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.
~
Specification The alarm set points for the radiation monitoring instrumentation shall be verified prior a.
to each day's run.
b.
The radiation monitoring equipment shall be calibrated annually.
Basis Verification of the alarm set points of radiation monittring instrumentation will assure that sufficient information to provide protection against radiation exposure is available.
j' l
l f
.w e,
a*
4.
p 4.5 ~ Maintenance Applicability i
This specifict. tion applies to the surveillance requirements following maintenance of a control or safety system.
Objective The objective is to assure that a system is operable before being used after maintenance has been performed.
Specifics. tion a) Following maintenance or modification of a control or safety system or component, it shall be verified that the system is operable prior to its return to service. A system shall not be considered operable until after it is successfully tested.
b) Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance wuh the specifications to which the systems were originally designed and fabricated or tc specifications approved by.the Reactor Committee.
c) A licensed reactor operator shall be present during maintenance of the reactor control and safety system.
l Basis This specification relates to changes in reactor systems which could directly affect the safety
. of the reacter. Changes or replacements to these systems which meet the original design specifications are considered to meet the presently accepted operating criteria.
'I L
t l
l l
p, p
o,..
g 4 22 4.6 Pool Water Conductivity t
Applicability This specific action applies to surveillance of pool water conductivity.
j Objective The objective h to assure that pool water mineral content is maintained at an acceptable level.
Specification The conductivity of bulk coolant water shall be verified to be within specified limits at least monthly.
Basis Based on experience, in which pool water conductivity changes slowly with time, observation at these intervals provides acceptable surveillanc? of conductivity to assure that accelerated fuel clad corrosion does not occur.
e i
e L
l l
6
-4 4:
g; -
f,~
23 5.0 DESIGN FEATURES I
5.1 Reactor Fuel Applicability -
This specification applies to the fuel elements used in the reactor core.
Objective j
The objective is to assure that the fuel elements are (,,och a design and fabricated in such a 1
manner as to permit their use with a high degree of reliability with respect to their '
mechanical integrity.
Specifications a.
Standard Fuel Element The standard fuel element shall be of the TRIGA type and shall contain uranium-zirconium hydride, clad in 0.020 inch of 304 stainless steel. It shall contain a maximum of 9.0 weight percent uranium which has a maximum enrichment less than 20 percent. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom, b.
Loading: The elements shall be placed in a closely packed array except for experimental facilities or for positions occupied by control rods, elements fully loaded with graphite, a neutron startup source, or single positions within the array filled with water.
Basis This type of fuel element has a long history of successful use.in TRIGA reactors.
e t
a
24 -
5.2 Reactor Building Applicability This specification applies to the facility which houses the reactor.
Objective i
The objective is to assure that provisions are made to restrict the radioactivity released into the environment.
Specifications a.
The reactor shall be housed in a closed room of a facility designed to restrict leakage, b.
The free volume of the reactor room shall be at least 6,000 cubic feet.
c.
All air or other gases exhausted from the reactor room during reactor operation shall be released at a minimum of 12 feet above ground level.
.d.
The reactor facility shall be equipped with a ventilation system capable of exhausting air or other gases from the reactor room from a stack at a minimum of 50 feet above ground level under emergency conditions.
Basis In order that the movement of air can be controlled, the facility contains no windows that can be opened. Under emergency conditions the room air is exhausted through a filter and discharged through a atack at a minimum of 50 feet above ground to provide dilution.
O F
y
,e s
' 5.3 Fuel Storage Applicability This specification applies to the storage of reactor. fuel at times when it is not in the reactor Core.
Objective The objective is to assure that fuel which is being stored will not become supercritical and will not reach unsafe temperatures.
Specifications All fuel elements shall be stored in a geometrical array where the value of k.fr is less a.-
than 0.9 for all conditions of moderation and reflection using light water, b.
Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed 800*C.
Basis Specification 5.3a assures that unplanned criticality will not occur in fuel storage racks.
Specification 5.3b is based on a fuel temperature limit of 950*C to assure fuel clad integrity when the clad temperatures can equal the fuel temperature (Simnad, G. A. Report E-Il7-833, February 1980, p.4-1)
D I
l' l
l l
l l.
l
-~
rc h-4
. 24 c 6.0 ; ADMINISTRA'HVE CONTROLS 6.1 Organization ~
l a.
The reactor facility shall be an integral part of the Nuclear and Energy Engineering Department of the College of Engineering and Mines at the University of Arizona as shown in the diagram below.
b.
The reactor facility shall be under the supervision of a licensed senior operator for the reactor. He shal be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by applicable federal regulations, by the facility license, and by the provisions of the Reactor Committee.
c.
There shall be a Health Physicist responsible for assuring the safety of reactor operations from the standpoint of radiation protection.
d.
An NRC-licensed operator must be present in the control room when the key switch is on. An operator and one other person authorized by the Reactor Supervisor must be present in the Reactor Laboratory whenever the reactor is not shut dowm.
l l-1 I
I; COLLEGE of ENGINEERING I
andMINES RADIATION CONTROL i
DEAN L-DIRECTOR i
i l
HEAD of NUCLEAR and ENERGY ENGINEERING ------,
RADIATION -
I DEPARTMENT
, CONTROL OFFICE l-m 7._._ _.1 r -- -,
I I
I I
I l
l REACTOR REACTOR REACTOR L~~
. HEALTH l
LA I Y
~ ~ '
SUPERVISOR COMMITTEE PHYSICIST D
't, REACTOR i
OPERATIONS e
8
L J q.,
1 i
+..
e.
,s.
6.2 - Review
- a. '
There shall be a Reactor Committee which shall review reactor operations to assure that the facility is operated in a manner consistent with public safety and within the terms of the facility license.
i b.
The res.remi :lity of the Committee includes, but is not limited to, the following-
' l.
Review and approval of experiments utilizing the reactor facilities; l
2.
Review and approval of all proposed changes to the facility, procedures, and Technical Specifications; 3.
Determination of whether a proposed change, test, or experiment would constitute an unreviewed safety question or a change in the Technical Specifications as required by 10 CFR 50.59, and review and approval of required safety analyses; 4.
Review of the operation and operational records of the facility; 3
5.
Review of abnormal performance of plant equipment and operating anomalies; r
and-6.
Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR 20 and 10 CFR 50.
7.
Review and audit of the retraining and requalification program for the operating staff.
8.
Biennial audit of the Emergency Plan.
c.
The' Committee shall be composed of at least five members, and shall include a health physicist and members competent in the field of reactor operations, radiation science, or reactor engineering. The membership of the Committee shall be such as to maintain a
'high level degree of technical proficiency.
d.
The Committee shall establish a written charter defining such matters as the authority-of the Committee, review and audit functions, and other such administrative provisions as are required for effective functioning of the Committee. Minutes of all meetings of the Committee shall be kept and submitted to committee members and to the Head of s
I the Department of Nuclear and Energy Engineering in a timely manner.
A quorum of the Committee shall consist of not less than three members of the
= e.
Committee and shall include the chairman or his designee.
f.
The Committee shall meet at least quarterly.
T
/
a-
- w.
s.
.., ~ $ -:
6.3.1 Operating Procedures Written procedures, reviewed and approved by the Reactor Committee, shall be in effect and followed for the following items. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgment and action should the situation require such, t
a.
Startup, operation, and shutdown of the reactor.
b, Installation or removal of fuel elements, control rods, experiments, and experimental facilities.
c.
Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system leaks, and abnormal reactivity changes, d.
Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.
e.
Maintenance procedures which could have an effect on reactor safety, f.
Periodic surveillance of reactor instrumentation and safety system, area monitors and continuous air monitors.
Substantive changes to the above procedures shall be made only with the approval of the Reactor Committee. Temporary changes to the procedures that do not chaage their original intent may be-made with the approval of the Reactor Laboratory Director. All such temporary changes to procedures shall be documented and subsequently reviewed by the Reactor Committee.
^
6.3.2 ALARA Program A program shall be established to assure that radiation exposures and releases are kept as low as reasonably achievable.
i e
l c
4 D
i 9
s, c:r:.
+-
e
. 6A Action to be Taken in the Event a Safety Limit is Exceeded in the event a safety limit is exceeded, or thought to have been exceeded:
a The reactor shall be shut down and reactor operation shall not be resumed until j
authorized by the NRC.
i b.
An immediate report of the occurrence shall be made to the Chairman of the Reactor H
Committee and reports shall be made to the NRC in accordance with Section 6.7 of these specifications.
c.
A report shall be made which shall include an analysis of the'causes and extent of possible resultant damage, efficiency of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall-be 3
submitted to the Reactor Committee for review, and a similar report submitted to the NRC when authorization to resume operation of the reactor is requested.
t 1
1 1
V O
o
Lei-4.
6.5 Action to be Taken in the Event of a Reportable Occurrence In the event of a Reportable Occurrence, the following action shall be taken:
a.
The Reactor Laboratory Director shall be notified and corrective action taken prior to resumption of the operation involved.
i b, A report shall be made which shall include an analysit of the cause of the occurrence, efficiency of corrective action and recommendations for measures to prevent or reduce
~
the probability of reoccurrence. This report shall be submitted to the. Reactor.
Committee for review.
i c.
. A report shall be submitted to the NRC in accordance with Section 6.7 of these SpCCilications.
s I
I e
l l
L 1
y
^
^
lG 33,
1 L
6.6 Plant Operating Records
'In addition to the requirements of applicable regulations, and in no way substituting therefor, records and logs of the following items shall be prepared and retained for a period of at least 5 years (except as otherwise specified in the Commission's regulations);
1,
. Normal plant operation (but not including supporting documents such as checklists, and
'~
a.
recorder charts, which shall be maintained for a period of at least one year);
b.
Principal maintenance activities; c.
Reportable Occurrences; d.
Equipment and component surveillance activities required by the Technical Specification; e.
Experiments performed with the reactor; Logs and records of the following items shall be prepared and retained for the life of the facility.
f.
Gaseous and liquid radioactive effluents released to the environs;
- g..
Off-site environmental monitoring surveys; h.
Fuel inventories and transfers; i.
Facility radiation and contamination surveys; j.
Radiation exposures for all personnel; and k.
Updated, corrected, and as-built drawings of the facility.
f W
7
i,;
P,
6.7 Reporting Requirements In addition to the requirements of applicable' regulations, and in no way substituting therefor, reports shall be made to the NRC as follows:
5e a.
A report'within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the USNRC RegionX Office of:
1.
' Any accidental offsite release of radioactivity above limits permitted by 10 CFR 20, whether or not the release resulted in property damage, personal hiery, or exposure; 2.
Any violation of a Safety Limit; and.
3.
Any reportable occurrences as defined in Section 1.0 (Reportable Occurrence) of these specifications in writing, b.
A written report within ten days to the U. S. Nuclear Regulatory Commission Attn:
Document Control Desk, Washington D.C. 20555, with a copy to the Director, Division of Reactor Safety.and Projects Region V. of:
1.
Any significant variation of measured values from a corresponding predicted.
- value of previously measured value of safety-connected operating' characteristics.
occurring during operation of the reactor; 2,
Incidents or conditions relating to operation of the facility which prevented or could have prevented the performance of engineered safety features as described in these specifications; 3.
Any reportable occurrences as defined in Section 1.0 of these specifications; and 4.
Any violation of a Safety Limit.
5.
Any accidental offsite release of radioactivity above limits permitted by 10 CFR 20, whether or not the release resulted in property damage, personal injury, or
(
exposure.
l l
L l'
e m-
. c. [....
.f 33 A written report within 30 days to the U.S. Nuclear Regulatory Commission, Aitn:
c.
Document Control Desk, Washington D.C. 20555, with a copy to the Director, Division of Reactor Safety and Projects, Region V, of:
1.
'Any substantial variance from performance specifications contained in these specifications or in the Safety Analysis Rt port; t
2, Any significant change in the transient or accident analysis as described in the Safety Analysis Report; 3.
Any changes in facility organization; and 4.
Any observed inadequacies in the implementation of administrative or procedural controls.
d.
A written report within 60 days after compiction of startup testing of the reactor to the U. S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington D.C.
20555, with a copy to the Director, Division of Reactor Safety and Projects, Region V, i
1.
An evaluation of facility performance to date in comparison with design predictions and specifications; and 2.
A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there.may be substantial variance from prior analysis.
' A written annual report within 60 days following the 30th of June each year to the U.S.
c.
Nuclear Regulatory Commission, Attn: D5cument Control Desk,' Washington D. C.
20555, with a copy to the Director, Division of Reactor Safety and Projects, Region V, 1.
A brief narrative summary of (1) operating experience (including experiments.
performed), (2) changes in facility design,' performance characteristics, and operat ng procedures related to reactor safety and occurring during the reporting i
period, and (3) results of surveillance tests and inspections; 2.
Tabulation of the energy output (in megawatt days) of the reactor, amount of pulse operation, hours reactor was critical, and the cumulative total energy output since initial criticality; 3.
The number of emergency shutdowns and inadvertent scrams, including reasons therefore; L'
4.
Discussion of the major maintenance operations performed during the period,
^
including the effect, if any, on the safety of the operation of the reactor, and the reasons for any corrective maintenance required; 5.
A brief description including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50; o
1 ee m
F
.. - t...
.e *
, 6.
A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge; Liquid Waste. (sumrnarized on a monthly basis) a.
Radioactivity discharged during the reporting period.
(1)' Total radioactivity released (in curies),
m (2)^ The MPC used and the ' isotopic composition if greater than 1 x 10-7 microcuries/cc for fission and activation products.
(3)
Total radioactivity (in curies), released by nuclide, during the reporting period, based on representative isotopic analysis.
(4)
Average concentration at point of release (in microcuries/ce) during the reporting period.
b.
Tctal volume (in gallons) of effluent water (including diluent) during periods of release.
Gaseous Waste (summarized on a monthly basis) a.
Radioactivity discharged during the reporting period (in curies) for.
(1).. Gases, d
(2)
Particulates with half lives greater than eight days, b.
The MPC used and the estimated activity (in curies discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis.
Solid Waste i
n.
The total amount of solid waste packaged (in cubic feet),
j b.
The total activity involved (in curies).
j c.
The dates of transfer or shipment and disposition j
.7.
A summary of radiation exposures received by facility personnel and visitors, including dates and time of significant exposures, and a summary of the results of radiation and contamination surveys performed within the facility; and i
8.
A description of any environmental surveys performed outside the facility.
l a
V}
.s
.s t.
6.8 Review of Experiments
~
= All proposed new experiments utilizing the reactor shall be evaluated by the
. a.
experimenter and the Reactor Committee. The evaluation shall be reviewed by a licensed Senior Operator of the facility (and the Health Physicist when appropriate) to assure compliance with the provisions of the utilization license, the Technical Specifications,10 CFR 20 and the requirements of 10 CFR 50.59. If, in his judgment, the experiment meets with the above provisions and does not constitute a threat to the integrity of the reactor, he shall submit it to the Reactor Supervisor for scheduling or to the. eactor Committee for final approval as indicated in Section 6.2 above. When R
pertinent, the evaluation shall include:
1.
The reactivity worth of the experiment; 2.
- The integrity of the experiment, inchding the effects of changes in temperature, pressure, or chemical composition; 3.
Any physical or chemical interaction that could occur with the reactor components; and 4.
Any radiation hazard that raay result from the activation of materials or from external beams.
5.
A determination that for the maximum planned or inadvertent pulse, no credible mechanism exists which could cause the experiment to fail.
b.
. Prior to performing an experiment not previously performed in the reactor, it shall be reviewed and approved in writing by the Reactor Committee. This review shall
- consider the following information:
1.
The purpose of the experiment;
- 2..
A procedure for the performance of the experiment; and 3.
The evaluation approved by a licensed Senior Operator, c.
- For the irradiation of materials, the applicant shall submit an "Irrad'ation Request" to the Reactor Supervisor. This request shall contain information on the target material including the amount, chemical form, and packaging. For routine irradiations (which do not contain known explosive materials and which do not constitute's significant threat to the integrity of the reactor or to the safety of individuals) the approval for the Reactor Conunittee may be made by the Reactor Supervisor.
1
g.
m,
.4 -'
- [
36 -
j
' d.
In evaluating experiments, the following assumptions should be used for the purpose of.
determining whether failure of the experiment would cause the appropriate limits of 10 CFR 20 to be exceeded:-
1.
If the possibility exists that airborne concentrations of radioactive gases or-aerosols may be released within the facility,100 percent of the gases or aerosols will escape;
' 2.
If the effluent exhausts through a filter installation designed for greater than 99 percent efficiency for 0.3 micron particles,10% of the particulates _will escape; and I
3.
For a material whose boiling point is above 130*F and where vapors formed by boiling this material could escape only through a column of water above the core, 10% of these vapors will escape.
s O
- - - - - - - - - - -