ML20092M307
| ML20092M307 | |
| Person / Time | |
|---|---|
| Site: | University of Virginia |
| Issue date: | 06/22/1984 |
| From: | VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA |
| To: | |
| Shared Package | |
| ML20092M276 | List: |
| References | |
| NUDOCS 8407020198 | |
| Download: ML20092M307 (38) | |
Text
, . - _ - _ . .- _ . -- . - . - - - . . .
I
?
I i
O l 1
l l
l t
6 FACILITY LICENSE R-123
?
?
TECHNICAL SPECIFICATIONS ,
FOR THE' !
UNIVERSITY OF VIRGINIA f CAVALIER REACTOR 1
O ~
i i !
i i
?
I
. I f
i r
k DOCKET NO. 50-396 l t
t O' i P
I a i
--..w- ,, , _ , . 4 ,,. + . . ~ , ._,,-,,y,...,----my,w.--,--,,pm,.7,,, ,,, , - -, _,y__ y,--me,_,yw ,..y-,, -,.c- ,,w,m,,-w-.rereyw-
(_S) TABLE OF CONTENTS ,
t
?.3&e.
1.0 DEFINITIONS. . . . . . . . . . . . . . . . . . . . . . . 1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS. . . . 4 2.1 Safety Limits . . . . . . . . . . . . . . . . . . . . 5
?
i 2.2 Limiting Safety System Settings . . . . . . . . . . 5 !
3.0 LIMITING CONDITIONS FOR OPERATION. . . . . . . . . . . . 6 t
3.1 Power Operation . . . . . . . . . . . . . . . . . . 6 3.2 Reactivity. . . . . . . . . . . . . . . . . . . . . l 7 '
3.3 Reactor Instrumentation . . . . . . . . . . . . . . 8 i 3.4 Reactor Safety System . . . . . . . . . . . . . . . 9 !
3.5 Limitations on Experiments. . . . . . . . . . . . . 11 ,
3.6 Operation With Fueled Experiments . . . . . . . . . 13 3.7 Rod Dro- Times. . . . . . . . . . . . . . . . . . . 14 j 3.8 Alternative Reactivity Insertion System (ARIS). . . 15 4.0 SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . . 15 i 4.1 Shim Roda . . . . . . . . . . . . . . . . . . . . . 15
(]) 4.2 Reactor Safety System . . . . . . . . . . . . . . . 16 -
4.3 Radiation Monitoring Equipment. . . . . . . . . . . 17 4.4 Maintenance . . . . . . . . . . . . . . . . . . . . 17 4.5 Alternative Reactivity Insertion System . . . . . . 18 ;
5.0 DESIGN FEATURES. . . . . . . . . . . . . . . . . . . . 19 L 5.1 Reactor Fuel . . . . . . . . . . . . . . . . . . . 19 5.2 Fuel Storage. . . . . . . . . . . . . . . . . . . . 20 i
6.0 ADMINISTRATIVE CONTRCLS. . . . . . . . . . . . . . . . . 21 r
6.1 Organization. . . . . . . . . . . . . . . . . . . . 21 !
6.2 Review and Audit. . . . . . . . . . . . . . . . . . 22 '
6.3 Operating Procedures. . . . . . . . . . . . . . . . 25 t 6.4 Required Actions. . . . . . . . . . . . . . . . . . 26 6.5 Plant Operating Records . . . . . . . . . . . . . . 28 6.6 Reporting Requirements. . . . . . .. . . . . . . . . 29 i
("% !
V
i
() 1.0 Definitions The terms Safety Limit (SL), " Limiting Safety System Setting" (LSSS), " Limiting Condition of. Operation" (LCO), " Surveillance I requirements," and " design features" are as defined in 10 CFR 50.36. j Channel Calibration: A channel calibration is an adjustment of the l channel so that its output responds, with acceptable range and accuracy, to known values of the parameter that the channel measures. Calibration shall encompass the entire channel, including equipment actuation, r
alarm, or trip.
Channel Check: A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification should include comparison of the channel with other independent channels or methods of measuring the same variable, where O this capability exists.
Channel Test: A channel test is the introduction of a signal into a channel to verify that it'is operable.
Experiment: An experiment is (1) any apparatus, device, or material placed in the reactor core region (in an experimental facility ~
associated with the reactor, or inline with a beam of radiation i
emanating from the reactor) or (2) any incore operation designed to measure reactor characteristics.
Experimental Facility: An experimental facility is any structure or device associated with the reactor that is intended to guide, orient, position, manipulate, or otherwise facilitate a multiplicity of experiments of similar character.
Explosive Material: Explosive material is any solid or liquid that is O(~N categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard 1
(O g in " Dangerous Properties of Industrial Materials" by N.I. Sax, or is given an Identification of Reactivity (stability) index of 2, 3, or 4 by ;
the National Fire Protection Association in its publication 704-M, i
" Identification System for Fire Hazards of Materials," also enumerated in the " Handbook for Laboratory Safety" published by the Chemical Rubber l Company.
i Fueled Experiment: A fueled experiment is any experiment that contains f U-235 or U-233 or Pu-239. This does not include the normal reactor core fuel elements.
- Measured Value: The measured value of the process variable is the value of the variable as it appears on the output of a measuring channel.
Measuring Channel: A measuring channel is the combination of sensor, lines, amplifiers, and output devices that are connected for the purpose of measuring the value of a process variable.
Movable Experiment: A movable experiment is one that may be inserted, l l
removed, or manipulated while the reactor is critical.
On Call: To be on call refers to an individual who (1) has been .
l specifically designated and the designation is known to the operator on duty, (2) keeps the operator on duty informed of where he may be contacted and the phone number, and (3) is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g.
l l approximately 30 minutes).
l Operable: A component or system is operable when it is capable of performing its intended function in a normal manner.
Operating: A component or system is operating when it is performing its
) intended function in a normal manner.
Reactivity Limits: Quantities are referenced to ambient tank water temperature with the effect of Xenon poisoning on the core activity l
2 L
(} accounted for if greater than or equal to 0.05% Ak/k. The reactivity worth of Samarium in the core will not be included in reactivity limits.
-The reference core condition will be known as the cold, xenon free critical condition.
Reactor Operation: The Reactor is in operation when not all of the shim rods are fully inserted and six or more fuel elements are loaded in the grid plate.
Reactor Safety System: The reactor safety system is that combination of measuring channels and associated circuitry that forms the automatic protective system of the reactor.
Reactor Secured: The reactor is secured when (1) all shim rods are fully inserted. (2).the console key is in the off position and is
,~
, removed from the lock, and (3) no work is in progress in core involving
\-
fuel or experiments or maintenance of the core structure, control rods, or control rod mechanisms.
Reactor Shutdown: The reactor is in a shutdown condition when all shim rods are fully inserted.
Reportable Occurrence: A reportable occurrence is any of the conditions
-described in Section 6.4.2 of these specifications.
Secured Experimenti A secured experiment is 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 sufficient to overcome those to which the experiment might be. subjected by hydraulic, pneumatic, buoyant or other forces that are normal for the operating environment of the experiment.
3
( Shim Rod: A shim rod is a control rod fabricated from borated stainless steel, which is used to compensate for fuel burnup, temperature, and l 1
poison effects. A shim rod is magnetically coupled to its drive unit l allowing it to perform the function of safety rod when the magnet is
[
de-energized.
i
- Surveillance Time Intervals
- Annual - Interval not to exceed 15 months Semi-annually - Interval not to exceed 7 1/2 months !
Quarterly - Interval not to exceed 4 months t
Monthly - Interval not to exceed 6 weeks Weekly - Interval not to exceed 10 days c l
Daily - must be done during the calcadar day f Tried Experiment: A tried experiment is (1) an experiment previously }
O performed in this reactor or (2) an experiment for which the size, !
shape, composition, and location does r.it differ significantly enough t
from~an experiment previously performed in this reactor to affect j i
reactor safety. l r True Value: The true value of a process variable is its actual value at [
any instant.
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS Applicability i The concepts of Safety Limits and Limiting Safety System Settings as ;
i normally developed with regard to fuel element integrity are not t
. strictly applicable for a low power system such as CAVALIER. The limitations on reactor power level due to radiation level and the need to handle irradiated fuel'are-far more restrictive than any limits based
{~}'
on fuel clad integrity. Therefore, limiting values are chosen rather 4 ;
b g w r-- , - - - -, n ,--,,,.--,.y e,,,. . . - - , ,.p, .w.,- . - w . . . , . . - - -. . ~ , - - - - .
conservatively at comparatively low levels according to the discussion provided in Section 2 of Chapter 3, SAR - CAVALIER.
2.1 Safety Limits Objective I To assure that the reactor is operated in a manner consistent with l maximizing safety for the operators and minimizing the chance for their i exposure, or the exposure of the public, to ionizing radiation.
Specification
{
Maximum Reactor Power Level 100 watts r
Moderator Tank Water Level >6.25 feet from top of core 6
Bases ,
The power level values were determined by the radiation levels above the 4
r; water level of the moderator tank as developed in Section 3.2 of the
(.)
CAVALIER SAR. The water height of 6.25 ft would lead to a dose rate of about 60 mR/hr above the reactor tank, at a power level of 100 watts, which produces a radiation level in control room work area which is significantly less than 60 mr/hr.
2.2 Limiting Safety System Settings (LSSS)
Objective
! To assure that automatic protective actions.are initiated to prevent a safety limit from being exceeded.
Specification (1) Maximum Reactor Power Level 80 watts (2) Moderator Water Level >7.25 feet from top of core Bases
() The reactor power value limits are slightly lower than those developed ,
in Chapter 3, Section 2, of the CAVALIER SAR. At a power level of 80 I watts with the water level at 7.25 ft above the core, radiation dose 5
() rates will be limited to <20 mr/hr above the reactor tank, and a dose rate of <1 mr/hr in the operating area and all normally accessible areas of the building. The American National Standards Institute Standard ANSI N18.9-1972 gives as a minimum requirement that the dose rates in unlimited access areas do not exceed the approved design values which are usually set at 10 to 50% of the current maximum permissible dose rate of 2.5 mrem /hr for plant personnel working a 40 hr week. The LSSS specified above will assure that the dose rates will not exceed these values.
3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Power Operation Applicability This specification applies to the average power rating of the CAVALIER.
Objective To assure that the reactor is operated in a manner consistent with maintenance of a low level of residual radioactivity'in the fuel elements.
Specification The Average Power Rating shall be less than 200 watt-hours / day where the
/
averaging period shall not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 Bases This rating will limit production of fission products to a level less than that analyzed in the Fission Product Released Section 9.4.4 of the CAVALIER SAR. This analysis indicates that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary after a very unlikely release of fission products from the fuel
(~h are within 10 CFR Part 20 averaged over a period of a year.
V 6
r
() 3.2 Reactivity Applicability These specifications apply to the reactivity condition of the reactor, and the reactivity worths of control rods and experiments.
Objective The objective is to assure that the reactor can be shut down at all times and that the safety limit will not be exceeded.
Specifications The following specifications apply to the reactivity conditions for reactor operation.
(1) The minimum shutdown margin provided by control rods with secured experiments in place and referred to the cold, xenon free condition with the highest worth control rod fully withdrawn, is greater than 0.4%
f'/
N Ak/k.
(2) Any experiment with a reactivity worth greater than 0.35% Ak/k must be a secured experiment.
(3) The total reactivity worth of all experiments is less than 1.6%
Ak/k and the reactivity worth of a single experiment is limited to 0.5% ,
i Ak/k.
(4) The excess reactivity including experiments in the core at any time shall be less than 1.6% Ak/k.
(5) The Alternate Reactivity Insertion System is operable.
These conditions must be met at all times with the following exceptions.
(a) The reactor may be operated up to 5 watts to measure the reactivity worth of experiments and the ARIS system must be operable.
(b) The reactor may be operated up to 60 watts to calibrate control i
rods after a major core configuration change to determine if
_ _ _ . _ _ _ _ _ _ . .___. - .. 7- _ _ _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _
O reciric tie 3.2.1 through 3.2.4 are met. The ARIS system must be operable during all operations.
Bases The shut down margin required by Specification 3.2(1) is necessary so that the reactor can be shut down from any operating condition and that it will remain shut down without further operator action.
The reactivity limitations in Specifications 3.2 (2) and (3) are based on the guidelines given in Regulatory Guide 2.2 as developed in
! the CAVALIER SAR. The reactivity worth limitations of specifications 3.2 (2) for a secured experiment and 3.2 (3) for any single experiment limit reactor period to prevent exceeding the Safety Limit.
The reactivity of 1.6% Ak/k in specificatica 3.2(4) corresponds to a 6.9 millisecond period. Reactor core DU-12/25 of the SPERT-I series
.O of tests had 12 plate fuel elements containing 168 grame of U-235 substantially similar to the CAVALIER fuel elements (Reference -
Thompson.and Beckerly, " Technology of Nuclear Reactor Safety," Volume I,
-page 683 (1964). A 6.9 millisecond period was non-destructive to the
, SPERT reactor when shut down immediately following the excursion. See Chapter 9 of the CAVALIER SAR.
The boron addition capability of the ARIS provides additional assurance that the reactor can be shut down and maintained subcritical I
in the event of all four control rods failing to respond to a scram signal. See section 9.4.6 of the CAVALIER SAR.
- 3.3 Reactor Instrumentation Applicability This application applies to the instrumentation which must be operable for; safe operation of the reactor.
8
Objective j r
The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.
t Specification The reactor shall not be operated unless the measuring channels !
described in the following table are operable and the information is displayed on the control console. !
i Measuring Minimum Operating Mode in Channel No. Operable Which Required '
Startup Count Rate 2 Reactor Startup !
Linear Power (Gamma-Ion Chamber) 1 All Modes Log N and Period (CIC) 1 All Modes !
Tank Top Radiation Monitor 1 All Modes Tank Water Level 1 All Modes Bases i The neutron detectors, and gamma monitors, provide assurance that measurements of the reactor power level are adequately covered at both low and high power ranges. The reactor tank water level indicator provides early warning of the possibility of a leak in the Moderator l Tank.
l The radiation monitor provides information to operating personnel of a l
[ decrease in tank water level, or of high reactor power, or of any impending or existing danger from radiation, contamination, or streaming
! allowing ample time to take necessary precautions to initiate safety.
l action.
3.4 Reactor Safety System Applicability This specification applies to the reactor safety system channels. ,
Objective The objective is to stipulate the minimum number of reactor safety ,
i 9
(} system channels that must be operable in. order to assure that the safety limit is not exceeded during normal operation.
Specification The reactor shall not be operated unless the safety system channels described in the following table are operable:
Operating Mode Measuring Minimum No. in Which Required'
. Channel Operable Function to be Operable Tcnk Water Level Monitor 1 Scram All Modes Tank Top Radiation Monitor 1 Scram All Modes
- Startup Count Rate 2 To prevent control Reactor Startup rod withdrawal when both channels read
. <2 CPS
' Manual Switch 1 Scram All Modes (1
--ccactor Power Level (CIC) 1 Scram All Modes Reactor Power Level (Gamma) 1 Scram All Modes Rsactor Period (CIC) 1 Scram All Modes at less than 5 second period Reactor Period (Gamma). 1 Scram All Modes at less than 5 second period Bases The startup interlock which requires a neutron count rate of at least 2 CPS on at least one startup count rate channel before the reactor is operated, assures that sufficient neutrons are available for proper operation of the startup channel. Power level scrams are provided to assure that the reactor power is maintained within the licensed limits.
i-The manual scram allows the operator to shut down the reactor if an unsafe or abnormal condition arises. The period scrams are provided to 10
() assure that the power level does not increase on a period less than 5 i seconds. This assures that the safety limit will not be exceeded as described in the CAVALIER SAR. One period scram specified is the power level channel using the compensated ion chamber and the other period scram utilizes a gamma sensitive chamber. Specifications on the tank water level scram are included as safety functions in the event of a serious loss of moderator tank water. Reactor operations are terminated
- when a major leak occurs in the tank. The analysis in Section 9.2 of the SAR for CAVALIER shows the consequences resulting from loss of this water but the area could be evacuated without difficulty before significant doses are received by personnel.
The tank-top radiation monitor provides a scram and gives warning in the event of a high radiat' ion level in the reactor room resulting from p
G failure of an experiment, from a significant drop in tank water level, or a higher than planned power level.
3.5 Limitations on Experiments Applicability This specification appites to experiments installed in the reactor and l
its experimental facilities.
Objective The objective is to prevent damage to the reactor or excessive release
[
of radioactive materials in the event of an experiment failure.
l Specifications The following limits on experiments shall be met at all times.
i
[ (1) The reactivity worths of all experiments shall be in conformance l
with specifications in Section 3.2.
)
(2) Movable experiment must be worth less than 0.1% Ak/k. '
11
(3) Experiments worth more than 0.1% Ak/k must be inserted or removed with the reactor shutdown except as noted in item (4).
(4) Previously tried experiments with measured worth less than 0.4%
Ak/k may be inserted or removed with the reactor 2% or more suberitical.
(5) If any experiment worth more than 0.4% Ak/k is to be inserted in the reactor, a procedure approved by the Reactor Safety Committee shall be followed.
(6) All materials to be irradiated in the reactor shall be either corrosion resistant or encapsulated within corrosion resistant containerr.
(7) Irradiation containers to be used in the reactor in which a static pressure will exist or in which a pressure buildup is predicted shall be
,_, designed and tested for a pressure exceeding the maximum expected by a .
factor of 2.
(8) Explosive material shall not be allowed in the reactor unless specifically approved by the Reactor Safety Committee. Experiments reviewed by the Reactor Safety Committee in which the material is potentially explosive, either while contained or if it leaks from the l container, shall be designed to prevent damage to the reactor core or to the control rods or instrumentation, and to prevent any changes in reactivity.
.(9) Experimental apparatus, material or equipment to be inserted in the reactor, shall not be positioned so as to cause shadowing of the nuclear instrumentation, interference with the control rods, or other perturbations that may interfere with the safe operation of the reactor.
(} Bases The above specified limitations on experiments are based on the guidance i
given in Regulatory Guide 2.2 as developed in Section 6 of the CAVALIER 12 r
r
() SAR and concern conservative requirements for protecting the reactor from materials to be used in experiments. The reactivity of less than '
O.1% Ak/k which can be inserted or removed with the reactor in operation !
in specification 3.5(2) can be compensated for by manual operation of a control rod.
3.6 Operation with Fueled Experiments Applicability [
This specification applies to a.a operation of the reactor with any I fueled experiment.
P Objective !
To assure that the fission product inventory in fueled experiments are within the limits used in the safety analysis. !
Specification The reactor shall not be operated with fueled experiments unless the'following conditions are satisfied. !
(1) The thermal power (or fission rate) generated in the experiment is 1-10 less than 1 watt (3.2x10 fission /second). [
(2) The total exposure of the experiment is not greater than the
- equivalent of 6 years continuous operation at 100 watts.
i Basis r
~
In the event of the failure of a fueled experiment, with the subsequent release of fission products (100% noble gas, 50% iodine, 1%
solids), the 2 -hour inhalation exposures to iodine and strontium 90 isotopes at the facility exclusion distance, 70 meters, are less than l the limits set by 10 CFR Part 20, using an averaging period of 1 year.
D-U 13 4' r w w -w ww w-P ~- ew- ww=w*- nww--,--- '%----t-- - - - -- v--- - -*-- y e - --w e-=-% -<--
() The analysis supporting this specification assumes 100% exfiltration of fission products from the reactor building in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The safety analysis is identical with that in Section 5.4 of the UVAR Safety Analysis Report for isotopes released to the reactor building in general (other than in the UVAR reactor room). The CAVALIER is in the same building as the UVAR. The UVAR Safety Analysis Report is on record with ,
the Commission: UVAR-18 (October, 1970), License NO. R-66, Docket No.
50-62.
3.7 Rod Drop Times Applicability-This specification applies to the time from the initiation of a scram to the time a rod starts to drop (release time), and to the time it takes for a rod to drop from the fully withdrawn to the fully inserted g+s
\"- position (free drop time).
Objective To aosure that the reactor can be shut down within a specified interval of time.
Specification The reactor shall not be operated unless:
(1) The release time for each of the shim rods is less than 100 milliseconds, and ,
(2) The free drop time for each of the shim rods is less than 700 i milliseconds.
Bases ,
Rod drop times as specified are sufficiently short to be consistent with
/~T i the reactor period and neutron level scram settings to assure that the V
safety limits will not be exceeded in a short period transient as shown i
in Section 9.3 of the CAVALIER-SAR.
i lh
() 3.8 Alternative Reactivity Insertion System (ARIS)
Applicability This specification applies to the boron solution in the ARIS tank and to the ARIS isolation valve.
Objective To assure that the ARIS is capable of providing an alternative means of reactor shutdown during all reactor operations.
Specification The reactor shall not be operated unless the following conditions exist:
(1) The volume of solution in the ARIS tank is greater than 24 gallons.
(2) The concentration of the baron is greater than 0.129 lb/ gal of solution. -
(3) The ARIS valve is unlocked.
b a Bases The boron solution in the ARIS tank will normally be kept at a volume of 25 gal. and a concentration of 0.144 lb of boron per gallon of solution.
The combination of 24 gal. with a concentration of 0.129 lb of boron per gallon of solution will yield a total negative reactivity addition of 3.2% Ak/k when uniformly mixed with the water in the moderator tank.
The requirement that the ARIS valve be unlocked will preclude unnecessary delay in the system initiation.
4.0 SURVEILLANCE REQUIREMENTS 4.1 Shim Rods Applicability This specification applies to the surveillance requirements for the shim
() rods.
. Objective To assure that the shim rods are capable of performing their function 15
and that no significant physical degradation in the rods has occurred.
d(~4
_ Specification (1) Shim rod drop times shall be measured semi-annually. Shim rod drop times shall also be measured if the control assembly is moved to a new position in the core or if maintenance is performed on the mechanism.
(2) The shim rod reactivity worths shall be measured whenever the rods are installed in a new core configuration.
Bases The reactivity worth of the shim rods is measured to assure that the required shutdown margin is available and to provide means for determining the reactivity worths of experiments inserted in the core.
4.2 Reactor Safety System Applicability This specification applies to the surveillance requirements for the safety system measuring channels and associated circuits of the reactor safety system.
Objective The objective is to assure that the safety system is operable and capable of preventing the safety limits from being exceeded.
Specification l
( (1) A channel test of each of the reactor safety system channels l shall be performed prior to each day's operation or prior to each operation extending more than one day.
(2) A channel check of each of the reactor safety channels shall be performed daily when the reactor is in operation.
l (3) A channel calibration of the reactor safety channels shall be performed semi-annually.
l
\ 16
r
(~) Bases
\_)
The daily channel tests and channel checks will assure that the safety channels are operable. The semi-annual calibration will permit any long-term drif t of the channels to be corrected.
4.3 Radiation Monitoring Applicability This specification applies to the radiation monitor required by Section i l
3.3 of these specifications. -
Objective The objective is to assure that the radiation monitor is operating and to verify the appropriate alarm setting.
Specification '
The operation of the radiation monitor and the position of its
/")
(_/
associated alarm set point shall be verified daily during periods when the reactor is in operation. Calibration of the radiation monitoring equipment shall be performed semi-annually.
Bases Surveillance of the monitor equipment will provide assurance that it is operable and that sufficient warning of a potential radiation hazard is available to permit corrective action before tolerances are exceeded.
4.4 Maintenance .
Applicability i
This specification applies to the surveillance requirements following maintenance of control or safety systems. '
Objective r i
The objective is to assure that a system is operable before being used O
! after maintenance has been performed.
L r
17 i
(' ') Specification
(./
Following maintenance or modification of a control or safety system t component, it shall be verified that the system is operable prior to its I
return to service.
r Bases I
The intent of the specification is to assure that work on the system or
,v component has been properly carried out and that the system or component :
has been properly reinstalled or reconnected.
4.5 Alternative Reactivity Insertion System (ARIS) ,
Applicability This specification applies to the alternative reactivity insertion i
system.
Obj ective N
To assure that the ARIS is operable and can provide sufficient reactivity to put the reactor in a suberitical condition. ;
Specification (1) Prior to each day's operation the volume of solution in the ARIS tank shall be verified, and che leak detection trap will be observed for signs of leakage.
(2) The concentration of boron in the solution shall be determined semiannually or after each make-up addition to the ARIS tank.
(3) A flow test from the ARIS tank to the flanged tee will be performed annually and the rssults compared to similar tests run at initial startup.
(4) The section of pipe from the flanged tee to the bottom of the moderator tank will be blown out with air annually.
g/
N.
S '
18
t j~.,
(.) Bases The daily verification and observation will provide a means of detecting leakage form the ARIS into the moderator tank which could cause unexpected reactivity fluctuations in the system. The concentration of the boron in the solution is determined periodically to assure that the ARIS is capable of providing a negative reactivity addition of 3.2%
Ak/k. The flow tests and air tests will demonstrate that the ARIS valve is operable and that the pipes are free of obstructions. '
5.0 DESIGN FEATURES 5.1 Reactor Fuel Applicability ,
This specification applies to the fuel elements used in the reactor
- core.
k_s)
Objective The objective is to assure that the fuel elements used in the CAVALIER are the same as those considered in the Safety Analysis Report.
Specification The fuel elements shall be of the materials testing reactor (MTR) type consisting of plates containing highly enriched uranium alloy fuel, clad with aluminum. There shall be 12 fuel plates containing 165 ( 3%)
grams of U-235 per element or 18 fuel plates containing 195 ( 3%),
grams of U-235 per element in the standard fuel elements. There shall be six fuel plates containing 82.5 ( 3%) grams of U-235, per element or nine fuel plates containing 98 ( 3%) grams of U-235, per element in the control rod fuel elements. Partially loaded fuel elements in which some
() of the fuel plates do not contain uranium may be used. An experimental element in which individual fuel plates can be removed or inserted may 19
g 4- -,
'~~. 7 p,. .c w - "
b e h '
1-
%g g,, S' '
,- c N.,s s s .-
. also'be used The mass of U ,235 listed above. refers to the initial s
,. N 't ' N \ -Ng
,,,i (zero burnup) loading <
~
Vari <E,eIhore configuratioYs consisting of any combination of the
-4_
+
above fuels eleme,nts may bs
- - -a -
% s used to accominodatei e'xperiments, but the
' s 3~. s.- -
g ,- ,y s i .
loadings shall31 ways be suchlihat the minimum shutdown margin and excess reactivity as specified in Section 3.2 of these specifications
- g. ,
are not exceeded. m Bases sE %
-g
,. % These same type fuel elementa have been run-l'n the UVAR reactor at 2MW
'- ,\ , .
c * $
26 for many years and would crea'te no safety problems for-the CAVALIER.
M i T x -
kg m- y 5.2 Fuel Storage w ,
['
g Applicability 1, ,ss . -
p . s ..
' " - ' x s s \\ .
p e mw M . 1 s ; --
This specification appliem~ to the storage of reactor fuel at times when
,x x-s-
it is not in 4he reactcr core. '
x + s
+
, i' 3 .y -N g. '" \ ' t ,,
4 Objective O" -
. 3
., .~
, 7,, ,
s ' g.
'M\ { The objective is to assureititat. fuel which is, being stored will not s s ., s
^
-s b'ecome appercritical and will n3t reach unsafe temperatures.
'S
,' *pecific
. s gtien 1, %,'( Q -
O di) N.;,iractorfuel.elehentst,ti hhe'r_eactor core shall be stored . g
" % ; ,. +
r u ._
in
.ys'a 3eometric' ,g,array'whereik'ff e is less than 0.9 for all conditions of i s.
Jmoderation. . . , ,
i
..s ' , <s ,
s
.""(2) Irradiated fuel,q(emente and fueled devices shall be stored in an '
, s n' 2\
s-
\[ .,
arrap v1.fch will permit sufficignt natural convection cooling by water s s or air svch that the fuell eleNth'irfueleddevicesurfacetemperature
\ .m,-
~
m.~
J. *
,W* -
c . 5 --
wiiL not ended the boiling point of water.
N, u-s e ' -
- ' Bas 3s '
l ,"
' \ ' N., - ;
Q, l. Wi(.hin these specificationb, the fuel can be stored safely under all s,j y ,
y conditions. the UVAR st'or33e facility was constructed to meet these t
\ -
, , g 20 41 6
..._y, . ,__ . . _ . _ . - -. , _ . . - . .
specifications and will be used to store the CAVALIER elements.
6.0 ADMINISTRATIVE CONTROLS 6.1 Organization
. 6.1.1 Structure The reactor facility shall be an integral part of the School of Engineering and Applied Science of the University of Virginia. The 5
organizational structure of UVA relating to the reactor facility is shown in Figure 6.1. The Chairman, Department of Nuclear Engineering will have overall. responsibility for management of the facility (Level 1).
6.1.2 Responsibility The Reactor Facility Director shall be responsible for the overall facility operation (Level 2). During periods when the Reactor Facility O' .
Director is absent, his responsibilities are delegated to the Reactor
! Supervisor (Level 3).
The Reactor Facility Director shall have at least a Bachelor of Science or Engineering degree and have a minimum of 5 years of nuclear experience. A graduate degree may fulfill 4 years of experience on a one-for-one time basis.
The Reactor Supervisor.shall be responsible for the day-to-day operation of the UVAR and CAVALIER and for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and the provisions of the Reactor Safety Committee.
During periods when the Reactor Supervisor is absent, his responsibilities are delegated to a person holding a Senior Reactor Operator license (Level 4).
The Reactor Supervisor shall have the equivalent of a Bachelor of
' Science or Engineering degree and have at least 2 years of experience in 21
, 1 q .,
1 .
- , o - . <
hl .
Reac or Operations at this facility,- or an equivalent facility, or at g y c,- . . :-
1 least 6,yn.ars ,
of cr.perience in Keactor Operations. Equivalent i
. / 1i
, educaticn. ore,xpehlencer>aybesubstituted~foradegree.' Within nine 3[9 .2, ' / s/g b l' 3 oj ' -! ;
, /, _ .aonths 'stter being assidded to the position, the Reactor Supervisor
.lb y i' y. +
il .. , I
. ,shall obtain;and maintain an NRC Senior Operator license.
- q'
' z y e i
)
Gs1.3 fStaffingi ~
,~ W- -
" J :
) ,
khen the reactor is operatingtthe following conditions will be met:
na ,/
i (1) ' A licenseid Senior Reactor, Oherator or a licensed Reactor Operator
,o
-, r. ,
A , ;/. ,
shall be4 present at the reacto'r controls.
a ? '
.n -
i 9 2: - . ,,, / ,
a g (2) A,[lidended Senior Reactor Operator'shall be on call, but not
,t
/
/ . /,
necessarily at the facility. '
/ 4 (3) At <1 east une'other pehson, not necessarily 'licdnsed-to operate the s
., . ,i 1 -<
+ 1 w reactor, shall be present at the faellity.
v fN, '
r <
s 1 . ,
1j _
" A
!(4) .4 ? k e,d.rangemen tN. ofL7thh. core or other'nenroutine .v ,
actions shall be t j . . A ,, , ; 3 f, , t<,
c'; ,
uupervised by'a licensed Senior Reactor Operator.
.. % ct
/
i' A /, e
' '? , /(5) IA/ health physicist who is organizationally' independent of the is j~ -
) ll . /
~
>^ '\ ) / [' _
. ReJLet'or Fac111t.y Operations groups, as shown 'in Figure 6.1, shall be
- ;,y
- V
. YJQ, ~ , -
5 y-y_ res, YonsiMa'foY rodiological safety at the fdc11'ity.
-v.-(.
c.
,t e .i" :> r ,
~'
6.3 Rtwiew and Audit lg t,
/, , >
.m ; .
j ! C7/~i
> , , :q
~
? ,.'
- '{>'- ;
s (+
- a. s' ']lp:"j There shall'be a-Reactor Safety Consaittee that, shall review and audit 3 , )_ : ,, p " '
G - ?'r.eactoy operadions to ensure that the facility is operated in a manner.
- m 9 .. q u 7R. -
,g k js ~
c.on.sistient - ~
with ;public safety and within1the terms of the facility
- y
.,s a
'y-)eya H, ,
.y
/
+
. lienise. - The Reactor Safety Committee shall_ report to the President of
~
a? ' n ,j ' a- ,
~'
l T '
- the Unisi ersity >,
and advise the t Chhirman, Department of Nuclear I
Engineer' $ riniNheReactorFacilityDirectoronthoseareasof
! ;Q L y, ; J . ., , ' l -
j A)l responsibil$t,yspedifiedbelow. /' i i
Q i .* *d, i ,- ^; ,z 'z .
- 8
- u i
J. .
't }, J , s /
y d , ,e , - ^?
,. i / ,~ -
I 2 I p' #
, .] y
,. .y. , / q g .f 6 5 j
, j ,
l ,t * /
' s '
i, '
) ,
O i
~ - a.~.,-. -- - .~ ~-- - -- - -
O O O l ,
PRESIDENT OF THE UNIVERSITY OF VIRGINIA i
DEAN, SCHOOL OF l RADIATION SAFETY ENGINEER 6NG AND
, COMMITTEE APPLIED SCIENCE R FETY - - - - ~ ~ CHAIRMAN PARTMENT OF NUCLEAR ENGINEERING
) l .
1 -
~ l
" L__________,
I l -
RADIATION SAFETY LEVEL 2 OFFICER OR -------------------- REACTOR FAcluTY DIRECTOR j HEALTil PHYSICIST 1
i
- LEVEL 3 REACTOR SUPERVISOR )
LEVEL 4 CHANNELS OF RESPONSIBILITY ~
REACTOR OPERATORS AND STAFF q- ----_= CHANNELS OF COMMUNICATION O
Figure 6.1 Organizational structure of UVA rela, ting to reactor facility
_ . . _ . . _ . _ . . _ _ _ _ _ . . _ _ _ _ _ _ . _ _ - _ _ . . . . _ _ . _ _ . . . ~ _ . . . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ ___._ _. _..~._ _ __.._ . . _ _ . _ _ _ _ _
l-(q
(_) 6.2.1 Composition and Qualification The Committee shall be composed of at least five members, one of whom shall be the Radiation Safety Officer of the University. No more than two members will be from the organization responsible for Reactor Operations. The membership of the Committee shall be such as to maintain a degree of technical proficiency in areas relating to reactor operation and reactor safety.
6.2.2 Charter and Rules (1) A quorum of the Committee shall consist of not less than a majority of the full committee and shall include the Chairman or his designee.
(2) The Committee shall meet at least semiannually and shall be on call by the Chairman. Minutes of all meetings shall be disseminated to responsible personnel as designated by the Committee Chairman.
(3) The Committee shall have a written statement defining such matters as the authority of the Committee, the subjects within its purview, and other such administrative provisions as are required for effective functioning of the Committee. j i
6.2.3 Review Function As a minimum the responsibilities of the Reactor Safety Committee include:
1 (1) review anJ spproval of untried experiments and tests that are significantly different from those previously used or tested in the reactor, as determined by the Facility Director.
(2) review and approval of changes to the reactor core, reactor systems or design feature that may affect the safety of the reactor.
-() (3) review and approve all proposed amendments to the facility license, Technical Specifications, and changes to the standard operating procedures (discussed in Section 6.3 of these specifications).
24 i- . . . .
____m. . _ _ _ _ _ _ -
() (4) review reportable occurrences and the actions taken to identify and correct the cause of the occurrences.
(5) review significant operating abnormalities or deviations from normal performance of facility equipment that affect reactor safety.
(6) review reactc operation and audit the operational records for -
r compliance with reactor procedures, Technical Specifications, and license provisions.
6.3 Operating Procedures Uritten procedures, reviewed and approved by the Reactor Safety Committee shall be in effect and followed for the items listed below.
These procedures shall be adequate to ensure the safe operation of the reactor, but should not preclude the use of independent judgment and action should the situation require such.
(1) startup, operation, and shutdown of the reactor.
(2) installation or removal of fuel elements, control rods, experimen'ts, and experimental facilities.
, (3) actions to be taken to correct specific and foreseen potential t malfunctions of systems or components, including responses to alarms, suspected system leaks and abnormal reactivity changes.
, (4) emergency conditions involving potential or actual release of ;
radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.
(5) preventive and corrective maintenance operations that could have an -
effect on reactor safety.
(6) periodic surveillance (including test and calibration) of reactor instrumentation and safety systems.
25
/]
(_j Radiation control procedures shall be maintained and made available to all operations personnel.
Su'ustantive changes to the approved procedures shall be made only with the approval of the Reactor Safety Committee. Changes that do not change the original intent of the procedures may be made with the approval of the Facility Director. All such minor changes to procedures shall be documented and subsequently' reviewed by the Reactor Safety Committee.
6.4 Required Actions 6.4.1 Action To Be Taken in the Event a Safety Limit is Exceeded In the event a safety limit is violated, the following actions shall be taken; (1) The reactor shall be shut down and reactor operations shall not be resumed until authorized by the Commission.
(2) The occurrence shall be reported to the Reactor Facility Director and the Chairman of the Reactor Safety Committee, or their designee, as soon as possible, but not later than the next work day. Reports shall be made to the Commission in accordance with Section 6.6 of these specifications.
(3) A written safety limit violation report shall be made that shall !
include an analysis of the causes of the violation and extent of resulting damage to facility components, systems, or structures; corrective actions taken; and recommendations for measures to preclude reoccurrence. This report shall be submitted to the Reactor Safety Committee for review.
26 c
(~) 6.4.2 Action To Be Taken in the Event of a Reportable Occurrence V-A reportable occurrence is any of the following conditions:
(1) any safety system setting less conservative than specified in Section 2.2 of these specifications.
(2) operating in violation of an LCO established in these specifications, unless prompt remedial action is taken.
(3) safety system component malfunctions or other component or system malfunctions during reactor operation that could, or threaten to, render the safety system incapable of performing its intended safety function, unless immediate shutdown of the reactor is initiated.
(4) an uncontrolled or unanticipated increase in reactivity in excess of 0.5% Ak/k.
(5) an observed inadequacy in the implementation of either f
administrative or procedural controls, such that the inadequacy could have caused the existence or development of an unsafe condition in connection with the operation of the reactor.
(6) abnormal and significant degradation in reactor fuel, and/or cladding, coolant boundary, or containment boundary (excluding minor leaks) where applicable that could result in exceeding prescribed
. radiation-exposure limits of personnel and/or environment.
l In the event of a reportable occurrence, the following action shall i
be taken:
(1) The Director of the Reactor Facility shall be notified as soon as possible and corrective action shall be taken before resuming the operation involved.
L 8
sustain an adequate knowledge and performance levels sufficient to operate i
the UVAR and CAVALIER reactors in a safe manner.
The program will cycle on a yearly schedule, running from 1 July to 30 June, with the yearly program terminating with the requalification
' examination.
(
I. Lectures and Drills The reactor operator and senior operator will attend a series ,
of lectures covering the following topics:
a) Normal Operating Procedures i b) Abnormal and Emergency Procedures c) Technical Specification ,
d) Plant Instrumentation r) Reactor Protective and Safety System f) Radiation Control and Safety ,
t g) Reactor Theory and, Plant Operating Characteristics h) Any Changes to Procedures or Equipment ,
If a lecture is missed, a make-up written or oral examination will be given and noted. The written examinations will be kept for i the qualification period. ;
Drills will be conducted during the requalification period to insure that the operators and senior operators are familiar with the procedures 4
4 9
.nacessary to prevant the endcngermint of tha personnt in the reactor building and aid an injured and/or contaminated perst - or persons. The drills to be held are:
a) Evacuation - once in fall and spring semester.
b) Injured and/or contaminated person - once a year.
II. Evaluation of Performance An annual evaluation will be made of each operator or senior operator while he or she performs a daily checklist and a reactor startup. The evaluation will be made by a senior operator or the operations manager and will be based on compliance with written procedures and good engineering
. practices. If an operator or senior operator is licensed on both the UVAR and CAVALIER reactors, he or she need only to be evaluated on one of the reactors. Any deficiencies in the operator or senior operator performances will be noted and appropriate corrective action will be taken and noted.
As a part of the evaluation, the operator or senior operator will discuss with the evaluator or simulate on the console, the procedures to be taken for emergency and/or abnormal conditions.
III. Requalification Examination '
The reactor operator examination will consist of a seven part test covering the following:
a) Principles of Reactor Operation b) Feature of Facility Design
-c)' General Operating Characteristics d) Instrumentation e) Safety System
'f) Operating Procedures g) Health Physics g
r The senior, operator examination will consist of the same areas listed for the operators test plus an additior.al section covering administrative
'1 controls. -
Each question of each section on the examination will be rated as to its total worth and the operator or senior operator must pass each section of the test with a grade of 70% or better. A 70% overall grade will be considered passing.
If an operator or senior operator receives less than 70% on one or more areas but pass the test overall, he or she will e-study the area (s) and will demonstrate his or her knowledge by taking a make-up written examination.
If an operator or senior operator scores less than 70% overall on the examination, he or she will retake the entire examination af ter a period of reviewing the material covered in the requalification lectures.
The individual administering the examination will be exempt i.om taking it.
l t .