ML20199K543
| ML20199K543 | |
| Person / Time | |
|---|---|
| Site: | 05000128 |
| Issue date: | 01/15/1998 |
| From: | TEXAS A&M UNIV., COLLEGE STATION, TX |
| To: | |
| Shared Package | |
| ML20199K531 | List: |
| References | |
| NUDOCS 9802060206 | |
| Download: ML20199K543 (45) | |
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4 TECHNICAL SPECIFICATIONS FOR TiiE TEXAS ENGINEERING EXPERIMENT STATION TEXAS A&M UNIVERSITY SYSTEM NUCLEAR SCIENCE CENTER REACTOR FACILITY DOCKET NO. 50128 LICENSE NO. R 33 MARCH 1983 REVISED THROUGli AMENDMENT NO 14
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TECIINICAL SPECIFICATIO14S FOR TIIE NUCLEAR SCIENCE CENTER REACTOR FACILITY LICENSE NO. R 83 March 1983 Revised through Amendment No.14 TABIE OF CONE.NTS Pass 1.0 Definitions ........ .................. ., .............................. ..................................I 1.1 Abnormal c-x .= .... .........................................................I 1.2 ALARA............................................................................................................................!
1.3 Channel....................................................... ..............................................I 1.3.1 Channel Tc.x ........... ......... ............................................................I 1.3.2 Channel Calibra:lon.......= ..........................................................I 1.3.3 Channel check ............... ..............................................................I 1.4 continement....... ............................................................................................2 1.5 Core lattice Position........ ... .....................................................................2 1.6 Experiment........ . ............... .................................2 1.7 Experimental Facilities................. ... .........................................2 1.8 Expet iment S afety Systems.......................................................... .... .............................. 2 1.9 FLIP Core ......................... ...................................................................2 1.10 FuelBundie....................................................................................................... ... 2 1 11 - Fuel Elemert ........................ .......................................................2 1.12 Instrumented Element ........... ........................................................2 1.13 Limiung safety System Setting......... ..................................................................2 1.14 Measuring Channel ................................ .....................................................2 1.15 Measured val ue . ........... ........ .. ..... ... . ....................... ......... .. ... ...... ... .. . ..... ...... .... ..... 3 1.16 MixedCore.......................................................................................................................3 1.17 Movable Experi ment . ........................................ ................................. ..................... ......... ... 3 1.1s Operable..........................................................................................................................3 1.19 Operating ....... .................... ...........................................................3 1.20 Steedy State Operational 0 ore .................. ............................................................ ... ...... 3 1.21 Puhe Operational 00re......................... ..................................................................3 1.22 PulseMode..........................................................................................................3 1.23 Reactivity Worth of an Experiment............ . ............................................. .............. ........ 3 1.24 Reactor Coneole Secured ........................... .. ........... . ... ................................................ 3 1.25 Reactor Operating .... ....................... ....................................................................3 1.26- Reactor Safety Systems ........ ...... ...........................................;.................................. ......... 3 1.27 Reactor Secured....... ... .............. .... ... ... ...... .............. ... .. ..... ........ . ..... ....... ........ ..... ... 4 1.28 Reactor Shutdown ...... ........... ..................................................................4 1.29 Reportable Occurrence ....... ...... .......... ....... . J4 1.30 Rod Cocuol ............. ... .......... ................ ..................................................4 1.31 Rod Regulating ......................... . .............-.....................................................5 1.32 Rod Shim Safety ................. ..... ... ......................................... ... ... 5
- 1.33 - Rod Transient. ............... ................................... 5 1.34 Safety Channel. ...............................................................................................5 1.35 Safety Limit........ =.m . .................... . .. 5
9 P
E 1.36 Scram Ti me......... ............... ... ........ =.... ............................. $
l.37 Secured Experiment .......... ............. .... ........................... 5
' 38
. Shell, Should and May. =...................., =....................................... $
'l.39 Shutdown Margin............................. . = . . . . . . . . . . . . . . $'
1.40 StandaadCore...................................................................................................................$ ,
1.41 Steady Stste Mode............... ............. ....................... ..........................6 1.42 TrueValue......................................................................... .................................6 1,43 L Unscheduled S hutdown .......... . . ..... ........ .... .. . .... .. .. ... ...... ....... ... . .. . .. ... ..... ... . ... ...... ... ... .. .. .. . 6 2.0 hfety ' lmit and umiting Safety System Setting............................................................................... 7 ;
2.1 Safety unilt Fuel EleawM Temperature ............................ ............... ..................7 ;
2.2 ~ umiting Safety Sptem Setting, . .............................. 7 3.0 u miting Conditions for Operatio1.............. ..... ............................................................................... 9 'I 3.1 Reactor Core Parameters .. . . . . . . . . . . . . . . . . . . . . . . . . . .. 9 3.1.1 Steady State Operation ................................................. . ................9 ,
3.1.2 Pulse Mode Operation.............................- ......... ...................... ............... 9 3.1.3 Shutdon Margin.......... . ................10 3.1.4 Core Configuration umitation..... .......... x.. . . . .. ... .... . . . . . I 1 3.1.5 Maximum Excess Reactivity........... :. . . . . . ..................I1 3.2 Reector Con:roi and Safety Systems.; . ........: . : .. 12
.3.2.1 Reactor Control Systems......... ;...... . ........ .12 ;
3.2.2 Reactor Safety Systems .................................. .......... .................. 13 3.2.s ' Scram Time .............. ....... ................. .. . .. ... . . .. . 14 3.3 Confinement.......... ;.............................................., . . . . . . . . . . . . . . . . . . . . . . . . . . I4 3.3.I Operations that Require Confinement....................... ..................I4 i 3.3.2 Squipment to Achieve Confinement.. ............ ...............~.!$ 1 3.4 Ventilation System.. ... ......... . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . !$
3.5 Radiation Monitoring Systems and Elfluents.................... .................16 3.5.1 Radiation Monitoring .................................... . . ...... .. . .. . 16 3.5.2 : Mgon-41 Discharge umit............. ....................................................... 17-3.6 umitations on Experiments....................... ............ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.6.1 Reactivity u mits.................. .......... = . :17 3.6.2 Material Umitations= ............................................... . ....... ... .. I 8 3.6.3 Failures and Malfu+ - ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.7 As im As Reasonably Achievable (ALARA) RadioaWe Effluents Released................... 20 3.8 Primary Coolant Conditions ........... ..................... ... ............. . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.0 Surveillance Requirements.. .. =.......... .. . . . . . . . . . . . . . '22 4.I General .......... ........... ............. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... = 22 4.2 Reactor Core Parameters ......... ............................................ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 4.2.I Steady State Operation.... . ............................. =22 4.2.2 Pulse Mode Operation........... ........... .. .... ................ ...... 22 4.2.3 Shutdown Martin = ................................................... J3 4.2.4 Reactor Fuel Elements ......... ................ .............. 23 4.3 Reactor Control And Safety Systems..... ...... ... ...... ........ =24 ,
4.3.1 Reactor Control Systems...... . ..... ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 4.3.2 Ren aor Safety Systems .. ........................................., ....................... .... ............ .. 25 4.3.3 Scram Tina .......... .... . . ...... ................26 4.4 Equipment To Achieve Confinement: Ventilation System .... ........ ............... ..... . ... .... 26 4.5 - Radiation Monitoring Systems And El11uents ............. ........... .................... ............ .... .. 26 4.6 - E xperiments .. . .. . .. . .. ... .. . .. . ... ... .. ...... .... .. . .... ... ...... . . . . .... . .. . .. . . .... . . .. . . . ... . . .. . 2 7
-- 5.0 Des!gn Features ........... .......... . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . 28 5.I Reactor Fuel ... .. . . . . . . .. ... . . ...... .... . . . . . ... .. .. . . ... . . .. .... . . .. .. . . .. . ..... . .... ...............28
' $.2 Reactor Core ..............
. ....................................................... 29
.e.->r= .n, , , - . - - . , - - . , , , . + - , . .,,--y,, -.m,, ---,Srv--r= y---
4 5.3 Control Rods ....... .. ...................................................29 5.4 Radiation F.hhoring System ................................... ................. .. .... 30 5.5 Fuel Slotape ....... ..... x................. ..... ..... 31 l 5.6 Reacnor Building And Ventilation System. ......................................................31 5.7 Reactor Pool Water Systems ..... ........................................... . x32 5.8 Physical Security .................................. ....... ................33 6.0 Admi nise.rstin Controls .................... ... ........................................ ....... .... ................................ .. 3 4 6.1 0 ,anization x................................................. ....................34 6.1.1 Struct ure . .. ... . . .... .. ...... . ............................... .......... 34 6.1.2 Responsibility........ . e....................................... ,34 6.1.3 Stamng........................................................................................................34 6.1.4 Selection and Training of Pervu - ............ ................................3$
6.2 Review and Audit...............- ................. ...............................36 6.2.1 Reactor Safety Board ............................................... - .......... 36 6.2.2 RSB Charter and Rules ......... ........ ....................................... .. .. .... . .. . 36 6.2.3 RSB Review Function...................... .......................... =36 6.2.4 RSB Audit Function.................. .......... .................36 6.2.5 Audit of ALARA Program....... . ...................37 6.3 Operating Procedures.. .. ............... ... ...... 3 7 6.4 Experiments Review and Approval.......................... ......... ................38 6.5 Required Actions = .. . . . . . . . . . . . . . . . . . . . . . . . . . , = ..................38 6.5.1 Actioac tJ be Taken in the Event a Safety Limit is Exceeded................................. 38 6.5.2 Action to be Taken in the Event of A Reportable Occurrence............................ .. 38 6.6 Reporting Requirements ................................... ............. :39 6.6.I Annual Raport..: . .... . ..... ... . 3 9 6.6.2 SpecialReport6 .............................. .. .. 40 6.7 Rooords. ........ ............................ ... .. ...... . .. 4 1 6.7.1 Records to be Retained for a Period of at least Five Years or for the life of the Component Inwived ........... ................................................. .................41-6.7.2 Records to be Retained for at least One Training Cycle....................................... 41 6.7.3 Records to be Retained for the Lifetime of the Reactor Facility................ =41
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. NSC Technical Specifications, Version 01/15/98 TECIINICAL SPECIFICATIONS FOR Tile NUCLEAR SCIENCE CENTER REACTOR FACILITY LICENSE NO. R-83 Included in this document are the Technical Specifications and the " Bases" for the Technical
( Specifications. These bases, which provide the technical support for the individual technial specifications, are included for informationc.1 purposes only. They are not part of Technical Specifications and they do not constitute limitations or requirements to which the licensee must adhere.
1.0 Definitions 1.1 Abnormal Occurrence An " Abnormal Occurrence" is defined, for tiw purposes for tle reporting requirenwnts of Section 208 of the Energy Reorganization Act of 1974 (P.L.93-438) as an unscheduled incident or event which the Nuclear Regulatory Commission determines is significant from the standpoint of public health or safety.
1.2 ALARA The ALARA program (As low As Reasonably Achievable)is a program for maintaining occupational exposures to radiation and release of radioacthe effluents to the cmirons as low as reasonably achievabic.
1.3 Chan el A channel is the combination of sensors, lines, amplifiers, and output devices, w hich are connected for t$c purpose of measuring tlw vahw of a parameter.
1.3.1 ChannelTest A channel test is the introduction of a mignal into the channel for wrification that it is operable.
1.3.2 Channel Calibration A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter witich the channel measures.
Calibration shall encompass tiw entire channel, including equipment actuation, alarm, or trip arxl shall be deemed to include a channel test.
1.3.3 Channel Check A channel check is a qualitathe verification of acceptable performare:c by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systents measuring tbc same variable.
1 I
. NSC Technical Specincations, Version 01/15/96 1.4 Confinement ConAnement nuens a closure of the overall facility which controls the movement of alt into it and out through a controlled path.
1.5 - CoreIJttice Position
- The core lattlee position is that region in the core (approximately 3" x 3') over a grid plus hole, it may be occupied by a ibel bundle, an experiment, or a reflector element.
1.6 Experiment
- An operation, hardware, or target (excluding devices such as detectors, fails etc.) which is _
' designed lo investigate non-routine reactor characteristics or which its intended for irradiation within the pool, on or in a beam port or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design.
IJ Experimental Facilities Experimental facilities shall mean beam ports, including extenslor. tubes with shields, thermal columns with shields, vertical tubes, through tubes, in core irradiation baskets, irradiation cell, pneumatic transfer systemJ and in-pool irradiallon facilities.
1.8 Experiment Sahty Systems Experiment safety systems are those systems, including their associated input circuits, which are designed to initiate a scenm fw the primary purpose of protecting an experiment or to provide informatic n which requires r.ianual protecths action to be initiated.
1.9 FLIP Core A FLIP core is an arrangement of TRIGA FLIP fuel in a reactor grid plate.
1.10 Fuel Bundic A fuel bundle is a cluster of two, three or four elements and/or non-fueled elements secured in a -
square array by a top handle and a bottom grid plate adapter. Non fueled elements shall be fabricated from stainless steel, aluminum or graphite materials.
-1.11 Fuel Element A fbel element is a sigle 'IRIGA fuel rod of either standard or FLIP type.
1.12 lastrumented Element An instrumented element is a special fuel element in which a shanthod chromal alumel or equivalent thermocouple is embedded in the fuel near the horizontal center plane of the fbel element at a point approximately 0.3 inch from the center of the fuel body -
1.13 Limiting Safety System Setting
- . Ihe limiting safety system setting is the setting for automatic pruective devices related to those variables having significant safety functions.
1.14 Measuring Channel A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.
2
5
-. NSC Technical Specifications, Version 01/15/98 1.1$ Measured Value The measured value is the value of a parameter as it appears on the output of a channel.
1.16 Mixed Core A mixed core is an erir, N of Wandard TRIO A fuel elements wnh at least 3$ *!RIGA FLIP fuel elements located in a central contiguous region of the core.
1.17 Movable Experiment A movable experiment is one for which it is intended that the entire experiment may be moved in or near the core or into and out of the renmor while the reactor is operating.
1.18 Operable Operable means a component or system is capable of performing its intended function.
1.19 Operating Operating means a component or system is performing its intended fu alon, 1.20 Steady State OperrsmalCore A steady state operational core shall be a Wandard core, mixed core, or FLIP core for which the core paramcters of shutdown margin, fuel temperature and power calibration have been determined.
1.21 Pulse Operational Core i
A pulse operational core is a steady state operational core for which the maximum allowable pulse reactivity insertion has been determined.
1.22 Pulse Mode Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.
1.23 Reacthity Worth of an Experiment The reactivity worth of an experiment is the maximum absolute value of the reacthity change
. that would occur as a result ofinter led or anticipated changes or credible malfunctions that alter the experiment position or configuration.
1.24 Roamor Console Secured The reactor console is secured whenew.r all scrammable rods have been fully inserted and veriAed down and the console key has been removed from the console, 1.25 Reactor Operating The reactor is operating whenever it is not secured or shutdown.
1.26 Reactor Safay Systems Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protecthe action. Manual protecthe action is considered part of the reactor safety system.
3
4 NSC Technical Specincations, Wrsion 01/1$/93 1.27 Reedor Secured A reactoris sewted when:
a)
It contains insnut nasile matedal or moderator present la t experiments to attain criticality under optimum renection, or acent available condi b)
The reactor console is secured,and
- 1) No work is in progress imotving 2) o s,or core thet No expedments in or near the reactor m being moved or senio 1.2g movement, a reactivity worth exceeding the maximum value of one dolla ,
Reactor Shutdows .
1.29 including the reactMty worth of all experiments, is ree condition and onar, Reportable Occurrence A reportable occurrence is at of the following a) 3 tion:
which occurs during b) the limiting safety system settMgs specined in th ons 2.2.
ve than Operation in violation oflimiting conditionsec for specifiutions.
n ca operation establis c)
A reactor safety system component malfuncdoc which renders o reactor safety spiem incapable of performing ith intended safety func malfunedon 6hutdowns. or Where (Note: condition
=- .- is discovered dudng maintenance tests
_--ts or systems are pimided in addi i t on to those not Wi . reportable prmided that the minimum numbe s specified or required perform their intended . cats or systems reactor safety function.)
- d) c)
An unanticipated or uncontrollod change in reacthity gror.ter than on ,
Abnormal and signincant degradation in reactor an fuel or cladding boundary, or containment boundary (excluding minor leaks) wh could result in exceeding prescribed radiation exposure limits ofperson environment,or both.
f)
.. An observed inadequacy in the implementation of administrr ,e unsafe condition with regard to reactor operations.such an tha 1.30 Rod Control
_ A control rod is a device fabricated from neutron absorbing it.ate be coupled to its drive unit allowing disengaged.
ontrol rod oupling is -
itmayto perform a sa 4
~
. NSC Technical Specifications, Version 01/1$/98 1.31 Rod Regulating The regulating rod is a low worth control rod used primarily to maintain an intended power level that need not have scram capability and may have a fueled follower. Its position may te vailed manually or by the servo <ontroller.
1.32 Rod Shim Safety l
A shim-safety rod is a control rod having an electric motor drive and scram capabilities. It may have a fueled follower section.
1.33 Rod Transient The transient rod is a control rod with scram capabilities that is capable of prmiding rapid reactivity insertion to produce a pulse.
1.34 Safety Channel A safety channel is a me'.suring channel in the reactor safety syst .u 1.3$ Safety Limit Safety limits are limits on important process variables which are found to te necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioacthity.
1.36 Scram Time Scram time is the time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable control rods reaches its fully inserted position.
1.37 Secured Experiment A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relathe to tic reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, oe other forces which are normal to the operating emironment of the experiment, or by forces which can arise as a result of credible malfunctions.
1.38 Shall, Should and May The word "shall" is used to denote a requireo.at; the word "should" to denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recomuendation. In order to conform to this standard, the user shall conform to its requirements but not necessarily to its recommendations, 1.39 Shutdown Margin Shutdown margin shall mean the minimum shutdown reacthity necessary to prmide confidence that the reactor can be made suberitical by means of the control and safety systems starting from any permissible operating condition, if the most reacths rod is stuck in its most reacthe position, and that the reactor will remain subcritical without further operator action.
1.40 Standard Core A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate.
0
, NSC Technical Specincations, Version 01/15/98 l 1.41 Steady State Mode Steady state modt, operation shall mean operation of the reactor with the mode selector switch in the steady state position.
1.42 True Value .
The true value is the actual value of a parame.er.
1.43 Unschedul:d Shutdown An unscheduled shutdown is denned as any unplanned shutdown of the reactor caused by actuation of the rwctor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely aNoct safe operadon, not to include shutdowns which occur during testing or check out operations.
6
. NSC Tcchnical Specifications, Version 01/15/98 2.0 Safe y Limit and Uniting Safety System Setting 2.1 Safety Limit Fuel Element Temperature Applicabinty ,
This specification applies to the temperature of tlw reactor fuel.
QtscGibt ne objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.
Sectificallom a) The terrperature in a TRIGA FLIP fuel element shall not exceed 2100'F (ll50*C) under any conditions of operation.
b) The temperature in a standard TRIO A fuel element shall not exceed 1830'F (1000'C) under any conditions of operation.
Dattl Tlw important parameter for a TRIGA reactor is the fuel element temperature, his parameter is well suited as a single specificatic,n especially since it can be measured. A loss in the integrity of the fuel element claciding could arise from a buildup of excesshe pressure between tle fuel-moderator and the cladding if the fuel temperature 6xceeds the safety limit, ne magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of the hydrogen to zirconium in the alloy.
, The safety I'mit for the TRIGA FLIP fuel element is based on data which indicate that the stress in the cladding due to the hydrogen pressure from the dissociation of rirconium hydride will remain below tlw ultimate stress prmided tlw temperature of the fuel does not exceed 2100*F (l 150*C) and tie fuel cladding is water cooled.
The safety limit for the standard TRIGA fuel is based on data, including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress prmided that the temperature of the fuel does not exceed 1830*F (1000*C) and the fuel cladding is water cooled.
2.2 Limiting Safety System Setting Aeolicabihty This specification applies t - /; ram setting which prevents the safety limit from being reached.
Obiecthe ne objecthe is to prevent the safety limits from being reached.
Soccification The limiting safMy system setting shall be 975'F (525'C) as measured in an instrvrted fuel element. The instrumented element shall be located adjacent to the central bundle with tim exception of the corner positions.
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, NSC Technical Specifiutions, Version 01/15/98 The limiting safety system setting is a ne'nperature
. which, if exceeded, shall cause a reactor scram to be initiated prewnting the asfety limit from being exceeded A peak core temperature of 950*C in FLIP fuel and 800*C in standard fuel is / criteria established to provide a minimum anfety margin of 200*C for all moocs of M. A part of this margin is used to account for the difference between the maxinaur. measured temperatures resulting f>om ' 'e actuallocation of the thermocouple. If the therum Me element were lomted in the hottest position in the om, the difference between the true a measured temperatures would be only a few degress since the thermocouplejunction is at the nud-plane of the elennent and close to the anticipated hot spot. However, this position is normally not available due to the la ation of the transient rod. De location of the instrumented elements is therefore restricted to the positions _
cicmst to the central element. Calculations indicate that, for this case, the true temperature at the hottest location in the core will differ fWom the measured temperature by no more than 40%
%us, for the siesdy state mode of operation when the temperature in the thermocouple element reached the trip setting of $25*C, the true temperature a the hattest location in a standard core would be no greater than 632*C and 690*C in a mixed core, prmiding a safety margin of at least M8'C for standard fuel elements and 460*C for FLIP type elements. %cee margins are ample to account for the remaining uncertainty in the accuracy of the fuel temperature measurement channel and any overshoot in reactor power resulting from a teactor transient during steady state mode operation, in the pulse mode of operation, the same limiting safety s)*cm setting will apply. However, the temperature channel will have no effect on limiting peak powers generated because ofits relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).
In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting the " tail" of the energy transient in the event the pulse rod remains stuck in the fully withdrawn position.
=2-8
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o
. NSC Technical Specifications, Version 01/IS/93
' 3.0 uniting Conditions for Operation 3.1 ReactM Core Parameters 3.1.1 ' Steady State Operation Applicability Dis specifiution applies to the energy generated in the reactor during steady state operation.
Obiectin The objective is to assure that the fuel ;; .., i...i.ie safety limit will not be exceeded during steady state operation.
Specifications ne reactor power level shall not exceed 1.3 megawatts under any condition of operation. The normal steady state operating power level of the reactor shall be 1.0 megawatts. Howewr, for purposes of testing and calibration, the reactor niay be operated at higher power levels not to exceed 1.3 megawatts during the testing period, Bads nermal and hydraulic calculations indicate the TRIGA fuel may be safely operated up to power levels of at least 2.0 MW with natural convection cooling.
3.1.2 Pulse Mode Operation Anglicability This specification applies to the peak temperature generated in the fuel as the result of a pulse insertion of reactMty.
Obiective The objective is to assure that respective pulsing will not induce damage tot he reactor fuel.
SpedGGalipJ1 a) The reactMty to be inserted for pulse operation shall not exceed that amount which will produce a peak fuel temperature of 1526*F (830*C). In the pulse mode the pulse rod shall be limited by mechanical means or the .od extension physically shortened so that the reactMty intertion will not inadvenently exceed the maximum value.
b) Until the full FLIP fuel core has been calibrated, maximum pulse shall be limited to $2.00.
Best TRIGA fuel is fabricated with a nominal hydrogen to zirconium ratio of 1.6 for FLIP fuel and 1.65 for standard. This yields delta phase zirconium hydride 9
f 4
o NSC Technical Speci6 cations, Version 01/15/98 which has a high creep strength and de no phase changes at temperatures owr 1000*C. However, aner extensive steady st.de operation a MW the hytrogen will redistribute due to migration fross the contral hig temperature regions of the Ibel to the cooler outer regions When the fuelis pulsed, the instantaneous temposetore distribution is such that the highest values coeur at the swface of the element and the lowest val conter. The higher temperstwos la the outer regions occur la ibel with a hydrogen w zirconium ratio that has now substantially in nominal value, This produces hydrogen gas presswes = creased abow the -
" A in excess of the expected for ErHu. If the pulse insertion is such that the temperature of the ibel exoseds 874*C, then the pressure will be sufRcient to cause exp of microscopic holes in the fuel that grows with each pulse. The pulsing of 830*C is obtained by examining the equilibrium hydrogen presswe of zirconium hydride as a ibaction of", mm. The decreare in W.a.ic from 874*C to 230*C reduces hydrogen pressure by a factor of two, which acceptable safety factor. This phenomenon does not aher the safety limit since the total hydrogen in a fuel element does not change. Thus, the presswe exerted on the clad will not be signlAcantly affected by the distribution of hydrogen within the eleinent.
limit for each specinc core. The peaking facion fr element to the hattest spot in the core must be calculated for each cose connguration that is to be used, Temperature would then be enessured for smallpulseinsertions.
For new uncalibrated cores, the pulse insertions shall be lacreased by sma ircrements to a snaximum of $2.00 tc allow an extrapolation of peak temperatures, thereby establishing the maximum allowed pulse insertion for a given core. Following approval by the NRC staff of the calibration of the new core, the $2,00 restriction shall be renoved. I 3.1.3 Shutdown Margin Annue.hmy 1hese specincations apply to the reactivity condition of the reactor and the reactivity worths of control rods and erperiments. They apply for all modes of operation.
Obiective The objective is to assure that the reactor can be shutdown at all times and to answo that :he fuel temperatwe safety limit will not be exceeded, e .=:u .
The teactor shall not be operated unless the shutdown margin provided b control rods is greater than $0.2$ with:
a)
The highest worth non secured experiment in its most reactive state, b)
The highest worth control rod and the regulating rod (if not scrammable) fully withdranit, and
~
c)
The r' eactor lei thi cold conditlon without xenon.
10 1&
, NSC Technical SpeclAcetions, Version 01/15/98 Basis The value of the shutdown margin assures that the rendor can be shut down from any op. rating condition even if the highest woeth control rod should remain in the fully withdrawn position. If the regulating rod is not w -icits worth is not used in determ'.;ng the shutdown reactivity.
3.1.4 Core ConAguration Limitation Applicability This specincation applies to mixed cores of EIP and standard types of fuel and to fbil RIP cores.
Otdactive The chlective is to assure,that the fbel %mporature safety limit will not be exceeded due to power peaking effects in mixed cores and ElP cores.
Specincations a) Tie 11tlGA core annembly may be standard, EIP, or a combination thereof(mixed core) provided that any EIP fuel core be comprised of at least thirty-Ave (35) fbel elements, located in a contiguous, central region, b) The reactor shall not be taken critical with a core lattice position vacant except for positions on the M.:q of the core assembly.
Water holes in the inner fuel region shall be limited to single rod positions. Vacant core positions shall contain experiments or an experimental facility to prevent accidental fuel additions to the reactor core.
c) The instrumented element shall be located a4acent to the central .
bundle with the exception of the corner positions (
Reference:
2.2 Limiting Safety System Setting).
Baans a) In mixed cores, it is necessary to specify the minimum number of EIP elements and arrange them in a contiguous, central region of the core to control flux peaking and power generation values in individual elements.
b) Vacant core positions containing experimeres or an experimental facility will prevent accidental fbol additions to the reactor core. They will be permitted only on the periphery of the core or a single rod -
position to prewnt power peaking in regions of high pour density, c)
Reference:
2.2 Limiting Safety System Setting.
3.1.5 Maximum Excess ReactMay Anoticability '
This specification applies to the maximum excess reactMty, abow cold critical, which may be loaded into the reactor core at any time.-
._ - _ = = _
_ =. = =_=
11
NSC Technical Spectruations, Version 01/15/P3 Othdalit De objective is to ensure that the core analyzed in the safety analysis report
)
approximales the operational core within reasonable limits.
Snacii%tions ne maximum reactMty in excess of cold, menon free critical shall not exceed
$.$% Ak/k ($7.85).
Benis Although maintaining a minimum shutdown margin at all times ensures that the reactor can be shut down, that specification does not address the total reactivity available within tic core. His specification, although mer.
constraining the reacsor symem, helps ensure that the licensee's oport'.ional power densities, fbel temperatures, and temperature peaks are maintained within the evaluated safety limits. De specified excess reactivity allows for power coefficients of reactMty, xenon poisoning, most +k.a.4 and operational flexibility, 3.2 Reactor Control pad Safety Symems 3.2.1 Reactor Control Symems Applicability This specification applies to the information which must be available to the reacht operator during reactor operation.
Obiective De objectiw is to require that sumcient information is available to the operator to assure safe operation of the reactor..
Specifications De reactor shall not be operated unless the measuring channels timed in the following table are operable.
Ffective Afode Afeamring Chanme! Afin. No. Opemble SS Pulse Fuel Element Temperature 1 X X Linear Powerlevel 1 X Log Powerlast 1- X Integrated Pulse Power 1 X Bangs Fuel temperature displayed at the control console gives contimms information on this parame*ct, which has a specified safety limit. The power level monitors assure that the resctor power lewl is adequately monitored for both steady state and pulsing modes of 12
. NSC Technical Specifications, Version 01/15/98 operation. The specifications on reactor power level indication are included in this section, since tie power level is related to the fuel temperature.
3.2.2 Reactor Safety Systems l Applicabi!!!r .
Tids specification applies to the reacto. safety system circuits.
Obiective The objective is to specify the minimum numter of reactor safety system channels that must be operable for safe operation.
Sas;ifications The reactor shall not le operated unless the safety circuits described in the following table are operable.
Number Efecthe Ah de Safety IChannel Operable Function SS Palse Fuel Element Temperature i SCRAM # LSSS X X liigh Powerleel 2 SCRAM @ 125% X Corsole Scram Button i SCRAM X liigh Power imel Detector 2 SCRAM on loss of supply voltage X Power Supply Preset Timer 1 Trausient rod scram 15 seconds or less X after pulse Leg Power i Prevent withdrawal of shim safeties at X
<4 x 10'8 W Leg Power i Prevent Pulsing above I kW X Transient Rod position 1 Prevent application of air unless fully X ,
inserted Shim Safetics & Regulating i Prevent withdrawal X Rod Position RAE3 The fuel temperature and power level scrams provide protection to assure that the reactor can be shutdown tefore the safety limit on the fuel element temperature will be exceeded The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs in the event of failure of the power supply for a safety chamber, operation of the reactor without adequate instrumentation is prevented. The preset timer insures that the reactor power level will reduce to a low level after pulsing.
The interlock to prevent startup of the reactor at power lents less than 4 x 10'8 W which corresponds to approximately 2 eps assures that sufficient neutrons are available for proper startup.
The interlock to prevent the initiation of a pulse ateve I kW is to assure that the magnitude of the pulse will not cause the fuel element temperature safety lindts to te exceeded The interlock to prevent application of air to the transien: rod unless tlw 13
- 0
, NSC Technical Specincedons, Version 01/1$/98 cylinder is fully inserted is to pment pulsing of the reactor in steady state snode. The interlock to pment the withdrawal of the shim safeties or regulating rod in the pulse mode is to pment the reactor from being pulsed while on a positiw period.
/ 3.2.3 Scram Time Annlicability This :; _ "' =d= applies to the time required for the scrammable control rods to be fully laserted from the instant that the fuel temperature safety channel variable reaches the umiting Safety System Setting.
Oldsmin The objectiw is to achiew prompt shutdown of the reactor to pment fuel damage.
AsasiGestion The scram time rnessured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest serainma51e 3 control rod reaches its fhily inserted position shall not exceed 1.2 seconds.
Basis This specification assures that the reactor will be pmmptly shutdown when a scram '
signal is initiated. Experience and analysis how indicated that for the ranse of transients anticipated for a 1RIGA reactor, the speci6ed scram time is adequate to assure the safety of the reactor.
3.3 Connnement 3.3.1 Operations that Require ConAnement Applicability
. This specincetion applies to confinement requirements during operation of the reactor and the handling of radioactin materials.
QttAEths To mais.tain normal or emergency air flow into and out of the reactor building during operations that produce or could potentially pmduce airborne radioactivity.
Specification ConAnement of the reactor building will be required during the following operations.'
a) Reactor operating, b) Handling of radioactiw materials with the potential for airborne release.
- For periods of time for maintenance to the central exhaust ihn, entry doors to the reactor building will remain closed except for momentary opening for pemonnel entry or exit.
Basis a) This basis applies during the conduct of those activities defined as reactor 14
1 i i
,t NfC Technical SpectScadons, Venion 01/15Ng j
- operations. Arpn-41 is produced during eperation of the scactor in .
., experimental facilides and in tiw reactor pool, thus, sir control witida the !
building and the exhaust system in necessary to maintain proper airborne radiation lewis in the reactor building and release levels in the exhaust stack. ;
Other radioactivity releases to the reactor building must be considered during i reasser operation, such as Assion product releans Rom a leaking ibel element w a release horn Axed experiments in or near the core. !
t b) De handling of radioactiw materials can result in the cocidental or controlled t riense of alsturne radioactivity to the reactor building environment or direct l
, release to the building exhaust system. In these cases, the control of air into
- . and out of the reactor building is necessary. !
3.3.2 Equipment to Achlew ConAnement i ammmar This specinostion applies to the equipment and controls needed to pewide corSnement j of the reactor building.
i ~
L phiamin
- t. . 1 i The objectiw is to assure that a minimum of equipment is in operation to achieve :
I conAnement as spacined in 3.3.1 and that the control panel for this equipment is ' !
[-
available for normal and emotpacy situations. {
L spacincations :
a) The minimum equipment required to be in operation to achiew conAnoment of j: the reactor building shall be the contral exhaust fan.' !
- b) Controls for establishing the operation of the ventilation system during normal and emergency conditions shall be located in the reception room. '
J l - 'For periods of time for maintenance to the central exhaust ten, entry doors to the -
, reactor building will remain closed to assure closure except for the momentary opening ;
- fw personnel entry or exit. !
Bagni a) Operation of the contral exhaust fan will achiew conAnement of the reactor building during normal and emergency conditions when the controls for air input are set such that the central exhaust Asa capacity remains greater than the amount of air being delivered to the reactor building. De exhaust ihn has
- ssiAcient capacity to handle extra air intake to the building during momentary p opening of doors.
b) ne control panel for the ventilauon sysicin provides for manual selection of air input to the reactor building and the automatic or manual selection of air removal. De air supply and exhaust systems work together to maintain a
. small negatin prusure in the reactor building. These controls are loor.ted in
- the voception room for accessibility during emergency conditions.
3.4 Ventilation Syste 1 f
- AgqnicabilitY l _. --
i 15 a_- _ . _ _ - - . - , . - , - . - _ - - - , - , , . - -
, NSC Technic:1 Specifications, Version 01/15/98 This specification applies to the operation of the facility ventilation nstem.
Obicctive The objectin is to assure that the ventilation system is in operation to mitigate the consequen es of the possible release of radioactive materials resalting from reactor operation.
Specification
'lle reactor shat! not be operated unless tie facnity ventilation system is operable, except for periods of time necessary to permit repair of tlic system. In the cent of a substantial release of airborne radioacthity, the ventilation system uilt be secured automatically by signals from an W.haust air radiation monitor.
Fjlfil During normal operation of the ventilation system, the concentration of Argon-41 in unrestricted areas is below the Effluent Concentration (SAR,Section IX). In the event of a substantial release of airborne radioacthity, the ventilation s" stem will be secured automatkally. Therefore, operation of the reactor with the ventilation system shutdown for short periods of time to make repairs insures the same degree of control of release of radioactive materials. hforcover, radiation monitors within the building independent of those in the ventilation system will give warning of high levels of radiation that might occur during operation with the ventilation system secured.
3.5 Radiation Monitoring Systems and Effluents 3.5.1 Radiation Monitoring Anoticability This specification applies to the radiation monitoring information which must be available to the reactor operator during reactor operation.
Obiective Tle objectin is to assure that sufficient radiation monitoring information is available to the operator to assure safe opera: ion of the reactor.
Soccification
, The reactor shall not be operated unless the radiation monitoring channels listed in the following trale are operable.
2 >
l Radiation Monitoring Channels
- Function Number Area Radiation Monitor Monitor radiation lewis within 1
! ,, , . the reactor bay Continuous Air Radiation Monitor 1 Exhaust Gas Radiation Monitor Monitor radiation levels in the 1 exhaust air stack
_ Exhaust Particulate Radiation Monitor l
- For periods of time for maintenance to the radiation monitoring channels, the intent of this speci5 cation will be satisfied if they are replaced with portable gamma sensitive mstruments having their own alarms or which shall be kept under visual observation. If two of the above monitors are not operating, the reactor shall be shutdown.
16 s ___ _ _ _ _ - - - - I
. NSC Technical Specifications, Version 01/15/98 BaaGI -
The radiation monitors pmvide information to operating personnel of any impending or existing danger imm radiation so that there will be sumcient time to evacuate the facility and take the necessary steps to prevent the spread of radmactivity to the surroundings.
3.5.2 Argon 41 Discharge Limit Anolicabihty This specification applies to the concentration of Argon.41 that may be discharged from i the TRIGA reactor facility.
QhiMin To insure that the health and safety of the public is 'ot endangered by the discharge of Argon 41 from the 'IRIGA reactor facility.
Specification I The concentradon of Argon-41 la the emuent gas from the facility as diluted by -
atmospheric air in the lee of the facility due to the turbulent wake effect shall not exceed 4
1.0 x 10 pCi/mlawrsged over one year. '
Basis The maximum allowable comentration of Argon 41 in air in unrestricted areas as specified in Appendix B, Table II of 10CFR20 is 1.0 x 10* pCi/ml.Section IX of the .
SAR for the NSCR substantiates a 5.0 x 10'8 atmospheric dilution factor for a 2.0 mph wind speed. This dilution factor represents the conditions at the site building for a wind speed of 2.0 mph, which occurs less than is% of the ti:ne on an annual basis.
3.6 Limitations on Experiments 3.6.1 Reactivity Limits Applicability ;
This specification applies to the reactivity limits on experiments installed in the reactor and its expc.imental facilities. '
Obiective The objective is to assure control of the reactor during the handling of experiments ejacent toorin the reactor core.
Socctfications The reactor shall not be operated unless the following c alitions governing experiments exist.
a) Non-secured experiments shall have reactivity worths less than one dollar.
b) The reactivity worth of any single experiment shall be less than two dollars.
Ba3CE 17
- .- NSC Technical Specifications, Version 01/15/98 a) This'apecification is intended to provide assurance that the worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positNe worth of the experiment were suddenly inserted.
b) 'Ihe maximum worth of a single experiment is limited so that its removal from -
the cold critical reactor will not resuk in the reactor achieving a pmer level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at Aill power war d result in a relatively slow power increase such that the reactor protective systems would act to prevent high pow levels from being attained.
3.6.2 - MaterialLimitations Anplicability This specification applies to experiments installed in the reactor and its experimental 3 facilities.
Obiective The objective is to prevent damage to the reactor or excessive release of radioactivity by limiting materials quantity and radioactive material inventory of the experiment.
Specifications a) Explosive materials in quanti;ies greater than 5 pounds shall not be allowed '
within the reactor beilding. Irradiation of explosive materials shall be restricted as follows.
- 1) Explosive materials in quantities greater than 25 milligrams shall not be irradiated in the reactor pool. Explosive materials in quantities less ,
than 25 milligrams may be irradiated provided the pressure produced upon detonation of the erplosive has been calculated and/or -
experimentally demonstrated to be less than the design pressure of the
- container.
- 2) Explosive materials in quantities greater than 25 milligrarns shall be .
restricted from the reactor pool, the upper research level, the demineralimr room, cooling equipment room and the Interior of the poolcontainment structure.
- 3) Explosive materials in quantities greater than 5 poends shall not be irradiated in experimental facilities.
- 4) Cumulative expcsures for explosive materials in quantities greater than 25 milligrams shall not exceed 10 n/cm2 for neutron or 25 Roentgen for gamma exposures.
b) Each fueled experiment shall be controlled suhh that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 10 Ci.
Daacs a) This specification is inter.ded to prevent damage to the reactor or n.aor safety systems resulting from failure of an experiment involving explosive materials.
- 1) This specification is intended to prevent damage to the reactor core 18
- 1. --
4 NSC Technical Specincetions, Version 01/15/98 ano afety related ' reactor components located within the reactor pool in the ownt of failms of an experiment involving the irradiation ofi exploeiw materials. Limited quantities ofless than 25 milligrams and .
proper containmc.it of such eg provide the required safety for
~
in-poolirradiation.
- 2) This specincatiwt is intended to prevent damage to vital equipment by' restricting the quantity and locatior of explosive materials within the reactor building. Explosives in quantities exceeding 25 niilligrams are -
restricted froen areas containing the reactor bridge, nector console, pool water coolant and puriAcetion systems and mector safety related; i equipment.
- 3) . !he failure of an experiment involving the irradiation of up to 5 pounds of explosive material in an experinnental facility located external to the rencsor pool structure will not result in damage to the reactor or the reactor pml containment structure, r
- 4) This speci5 cation is intended to prevent any increase in the sensitivity of explosive materials due to radiation damage during exposures.
b) The 10 Ci limitation on lodine 131 through 135 assurcs that in the event of failure of a fbeled experiment leading to total release of the lodine, the exposure
. dose at the exclusion area boundary will be less than that showed by 10 CFR 20 for an unrestricted area. -
3.6.3 Failures and Malfunctions Anphcability This specification applies to experiments instal. led in the reactor and its experimental facilities.
Obtective
. Ihe objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
Specifications a) Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols pmduced escaped in ;
the reactor room or the .M..Q.a, the airborne concentration ofindioactivity averaged over a year would not exceed the limit of Appendix B of 10CFR20.
b) In calculations pursuant 6o a) above, the following assumptions shall be used:
b
- 1) If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10%
of the gaseous activity or aerosols pmduced will escape.
- 2) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles at least 10% of these vapors can escape.
19
,t NSC Technica: Specincations, Version 01/15/98 ~
- 3) For materials whose boiling point is above 130*F and where vapors
- formed by boiling this material can escape only through an undistu: bod column of water above the core, at least 10% of these -
vapors can escape.
, c) - If a capsule fails and releases material which could damage the rr. actor fbel or l structure by corrosion or other means, mmoval and physical inspection shall be -
- perfonned to deterraine the consequences and need for corrective action. *i he results of the inspection and any corrective action taken shall be reviewed by the Director (NSC) or his designated alternate and determined to be satisfactory befose operation of the reactor is resumed.
- Bases a) 'this specincation is intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of 10CFR20 will be released to the atmosphere outside the facility boundary of the NSC, .
. b) These - As are used to evaluate the potential airborne radioactivity {
release due to an experimen; failure.
c) Operation cf the reactor with reactor fuel or st ucture damage is prohibited to avoid release of fission products. Potential fank ge to reactor fbcl or structure .
must be brought to the attention of the Director (NSC) or his designated alternate for review to assure safe operation of the reactor.
3.7- As Im As Reasonably Achievable (ALARA) Radioactive Effluents Released Aspimainlity
. This specification applies to the measures required to ensure that the radioactive effluents 4 released from the facility are in accordance with ALARA criteria.
Obiective The objective is to limit the annual population radiation expose resulting from operation of the -
reactor to a small percentage of the normal local background exposure.
Specifications
- 1) In addition to the radiation monitoring spwN in Section 5.4, an environmental radiation monitoring program shall be conused to measure the ic .a4 radiation exposure in and around the environs of the facility on a quarterly basis.
- 2) The annual radiation exposure due to reactor operation, at the closest off-site point of extended occupancy, shall not exceed twice the average local off-site background radiation.
3). 'Ihe total annual discharge of Argon-41 into the environment may not exceed 30 Ci per year unless permitted by the RSBi
- 4) in the event of a significant fission product leak from a fuel rod or a significant airborne radioactive release from a sample being irradiated, as detected by the continuous air monitor, the reactor shall be shut down until the source of die leak is located and eliminated. However, the reactor may be continued to be operated on a short-term basis as needed to assist in determining the source of the leakage.
I 20
.- NSC Technical Specifwations, Version 01/15/98
- 5) L Before discharge, the facility liquid emuents collected in the holdup tanks shall be analyzed for the nature and concentration of radioactive emuents. The total annual -
t quantity ofliquid effluents _(above badground) shall not exceed 1 Ci per year.
Basis Ilic simplest and anost reliable method of ensuring that ALARA release limits are accomplishing .;
their objective of minimal facilitvaused radiation exposure to the general public is to actually measure the integrated radiab. exposure in the environment on and off the site. - 4 i'
3.8 Primary Coolant Conditions -
Annlicability
'llds specification applies to the quality of the primary coolant in contact with the fuel cladding.
% =1 The objectives are (1) to minimize the possibility fa conosion of the cladding on the fuel elements and (2) to minimize neutron activation ouissolved materials. -
Specifications
- 1) Conductivity of the bulk pool water shall be no highet than 5 x 104 mhos/cm for a period not to exceed two weeks
- 2) The pH of thc bulk pool water shall be betwien 5.5 and 8.0 Deviations of pH values outside this range shall not exmd a period of two weeks Baans A small rate of corrosion continuously ocx:urs in a water-inetal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel claMing, a water cleanup system is required. Experience with water quality control at many reactor facilition has shown that maintcaance within the specified limits provides acceptable control.
By limiting the concentrations of dissolved matenals in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposure during maintenance and operations.
e 1
21
.- NSC Technical Specitations, Version 01/15/98 4.0 Surveillance Requirements 4.1 General Applicabil:tv -
His specincation, applies to the surveillance requirements of any system related to reactor safety, Obiective The objective is to verify the proper operation of any system related to reactor safety, SPecincs %n3 -
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 with the ,!
wihlons to which the systems were originally designehnd fabricated or to specincations approved by the Reactor Safety Board. A system shall not be considered operable until aAer it is successfully tested, o
M This specincation relates to changes in reactor systems, which could directly affect the safety of tin reactor As long as changes or replacements to thcae systems continue to meet the original design specirntions, then it can be assumed that they meet the presently accepted operating criteria.
4,2 Reador Core Parameters ,
i 4.2.1 Steady State Operation Apolicability This specification applies to the survei llance requirement of the power level monitoring channels.
Obiective ne objective is to verify that the maximum power level of the reactor meets the license re,airements.
SocciDcation A channel calibration shall be made of the power level monitoring channels by the calorimetric method annually but at intervals not to exceed 14 month:
Basis
. The power level channel calibration will assure that the reactor will be operated at the paper powerlevel.
4.2.2 Pulse Mode Operation Acolic@ill'y nis specification appliu to the surveillance requirements for operation of the reactor in 22
4
, NSC Technical Specifications, Version 01/15/98 I
the pulse mode.
Obiective The objective is to verify that operation of the reactor in the pulse mode is proper and safe and to determine"if any significant changes in fuel chawteristics have occurred Sagetrication The reactor shall be pulsed semiannually to comparc fuel temperature measurements and core pulse energy with those of previous pulses of the same reactivity value or the z reactor shall not be declared operational for wlsing until such pulse measurements are '
performed.
Halis The reactor is pulsed at suitable intervals to make a comparison with previous similar pulses and to determine if changes in fuel or core characteristics are taking place. -
4.2.3 Shutdown Margin Anoticability This specificction applies to the surveillance requirement of control rod calibrations and chutdown margin.
Obiectnt The objective is to verify that the requirements for shutdown margins are met for opeisional cores.
Soccucation The reactivity worth of each controt rod and the shutdown margin shall be determined annually but at intervals not to exceed 14 months. <
Basil Tbc reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worth of experiments inserted in the core. Past experience with TRIGA reactots gives assurance that measurement of the reactivity worth on an annual basis is adequate to insure no significant changes in the shutdown margin.
4.2.4 Reac:or Fuel Elements Anoticability This speci". cation applies to the surveillance requirements for the fuel elements.
Qhicctne The objective is to verify the continuing integrity of the fuel element cladding and to ensure that no fuel damage has occurred Soccification 23
. NFC Technical Specificatioru, Version 01/15/98 a) The following fuel elements shall be inspected sisually for damage or deterioration and measured for length and bend annually, not to exceed 15 months.
- 1) At least four fuel elements which occupy the highest pulse t.mperature posi'tions in the core.
- 2) At least one-fiAh of the fuel elements used in operation of the nactor over the previous inspectio.: year.
- 3) The four elements in (1) above may be included in the inspection of fuel elements of(2)above, b) If any element is found to be damaged, the entire core will be inspected.
c) The reactor shall not be operated knowingly with damaged fuel, d) A fuel element shall be considered c.amaged and must be removoi from the -
coreif:
- 1) In measuring the transverse bend, the bend exceeds 0.12.5 inch over
) the length of the cladding.
- 2) In measuring the clongation, its length exceeds its original length by 0.125 inch, or
- 3) A clad defect exists as indicated by release of fission products.
Basis The frequency ofinspection and measurement schedule is based on the pararneters most likely to affect the fuel c. adding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose characteristics are well known. Experience has shown that temperature is the major contributor to fuel dam *8c.' Inspection of the four fuel elements which occupy the highest pulse temperature positions in the core prtnides surveillance for detection of the most probable fuel element damage should it occur.
Inspection of one Sfth of elements used in operation of the reactor provides surveillance of the lower temperature elements and over a five year period provid.:s for inspection of all elements.
The limit of transverse bend has been shown to result . no difficultyin dissonnbling fuel bundles. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching.
Experience with TRIGA reactors has shown that fuel element bowing that could result in tou,hing has occurred without deleterious effects. The clongation limit has teen -
specified to assure that the cladding material will not be subjected to stresses thtt could cause a loss ofintegrity in the fuel containment and to assure adequate coolant flow.
4.3 Reactor C'ntrol And Safety Systems 4.3.1 Reactor Control Systems Apolicability j These specifications apply to the surveillance requirements for reactor control sys: ems.
24
.: NSC Technical Specificati:ns, Version 01/15/98 l ' OL: ctive ne objective is to verify the condition and operability of system components affecting -
safe and proper control of the reactor.
SpeciAcations a) ne control rods shall be visually inspected for deterioration at intervals not to exceed 2 years..
b) The transient rod drive cylinder and associated air supply system shall be inspected, cleaned, and lubricated as necessary semiannually at intervals not to exceed 8 months.
BMil Tin visual inspection of the control rods is e .c evaluate corrosion and wear characteristics caused by operation of the r . Inspection and maintenance of the 1 transient rod drive assembly reduces the probsoility of failure of the system due to moisture induced cormsion of the pulse cylinder and piston rod assembly.
4.3.2 Reactor Safety Systems Apolicabahty nese specifications apply to the surveillance requirements for measurements, tests, and calibrations of the control and safety systems.
Obiecuve The objective is to verify the performance and operability of the systems and components which are directly related to reactor safety.
Soecifications a) A channel test of each of the reactor safety system channels for the intended mode of operation shall be performed prior to each day's operation or prior to each operation extending more than one day, except for the pool icvel channel -
which shallbe tested weekly, b) Whenever a reactor serain caused by high fuel element temperature occurs, an -
evaluation shall be conducted to determine whether the fuel element tempersture safety limit was exceeded c) A calibiation of the temperature measuring channels shall be performed -
senuannually but at intervais not to exceed 8 months.
d) A channel check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated by recording a measured value of a meaningful temperature indication.
Dnis Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run. Operational experience with the TRIGA system gives assurance that the thermocouple measurements of fuel element temperatures have been 25
. NSC Technical Specifications, Version 01/15/98 I sufficiently reliable to assure accurate indication of this parameter.
4.3.3 Scram Time Acolicability This specification applies Io the surveillance of control rod scram times.
Qht!G!YIl The objective is to verify that all scrammable control rods meet the scam time requirement.
Spectfication The scram time shall be measured annually but at intervals not to exceed 14 months.
DailE ,
Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly.
4.4 Equipment To Achieve Confinement: Ventilation Sptem Apolicability This specification applies to the building confinement ventilation system.
Obiective The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled emironment.
Soccification It shall be verified weekly that the ventilation system is operable.
Basil Experience accumulated over several years of operation has demonstrated that the tests of the ventilation system on a weekly basis are sufficient to assure the proper operation of the syttem and control of the release of radioactin material.
4.5 Radiation Monitoring Systems And Efiluents Apolicability This spectfication applies to the surveillance requirements for the area radiation monitoring equipmer t and the continuous air monitoring system.
Obiective The objective is to assure that the radiation monitoring equipment is operating and to verify the appropri'.te alarm settings.
Soccification 26
. NSC Technical Specifications, Version 01/15/98 i
The area radiation monitoring system and the continuod air mr.'titoring system shall be calibrated annually but at intervals not to exceed 14 months and shall be verified to be operable et weekly intervals.
Basis Experience has shown that weekly verification of area radiation and air monitoring system set points in coq) unction with annual calibration is adoquate to correct for any variation in the system due to a change of operating characteristics mer a long time span.
4.6 Experiments <
Acolicability This spectfication applies to the serveillance requirements for experiments installed in the reactor and its experimental facilities and for irradiations performed in the irradiation facilitics.
Obiective .
The c,bjective is to present the conduct of experiments or irradiations which may damage the reactor or release excessive amounts of radioactive materials as a result of failure.
)
Soccifications a) A new experimcat shall not be installed in the reactor or its experimental facilitics until a hazard analysis has been performed and miewed for compliance with the Limitatic..s on Experiments, Section 3.9, by the Reactor Safety Board. Minor modifications to a miewed and apprmed experiment may be made at the discretion of the senior reactor operator responsibic for the operation prmided that the hazards associated wit h the modifications have been miewed and a determination made and documented that the modifications do not create a significantly different, a new, or a greater safety risk than the original approved experiment.
b) The performance of an experiment classified as an apprmtd experiment shall not be performed until it has been mieu M for compliance by a licensed senior operator and a person qualified in health physics.
c) The reactivity worth of an experiment shall be estimated or measured, as appropriate, tefore reactor operation with said experiment.
Bass It has been demonstrated over a number of years of experience that experiments and irradiations reviewed by the Reactor Staff and the Reactor Safety Board as appropriate can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.
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. NSC Technical Specifications, Version 01/15/98 i
5.0 Design Fe.tures -
5.1 - Reactor Fuel Applicabahty -
'Ihis specification applies to the fuel elements used in the reactor core.
Obiective
~
The objective is to assure that the fuel elements are of such a dedgn and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Specifications s) - TRIGA-FLIP Fuel .
The individual unirradiated FLIP fuel elements shall have the following characteristics:
- 2) Hydrogen-to-zirconium atom ratio (in the ZrHx ): nominal 1.6 H atoms to 1.0 Zr atoms.
- 3) Natural erbium content (homogeneously distributed): nominal 1.5 Wt%
- 4) Cladding: 304 stainicas steel, nominal 0.020 inch thick.
- 5) Identification: Top pieces of FLID elements will have characteristic markings to allow visual identification of FLIP elements employed in mixed cores.
b) Standard TRIGA fuel The individual unitradiated standard TRIGA fuel elements shall have the following characteristics:
- 1) Uranium content: maximum of 9.0 Wt% enriched to a nominal 20% Uranium-235,
- 2) Hydrogen-to-zirconium abm ratio (in the ZrHx): nominal 1.7 H atoms to 1.0 Zr atoms.
3)- Cladding: 304 stainless steel, nominal 0.020 inch thick.
Bases a) A maximum uranium content of 9 Wt% in a TRIGA-FLIP element is about 6% greater than the design value of 8.5 Wt% Such an increase in loading would result in an increase in power density of about 2% Similarly, a minimum erbium content of 1.1%
in an element is about 30% less than the design value. This variation would result in an increase in power density of only about 6% An increase in local power dens $ of 6%
reduces the safety margin by at most ten percent. The maximum hydrogen-to-zirconium ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than the value resultin- from a hydrogen-to-28
, NSC TechairJ Specifications, Version 01/15/98 zirconium ratio of 1.60. Honver, this increase in the clad stress during an accident would not exceed the ruptwe strength of the clad.
When standard and EIP fuel elements are used in mixed cores, visual identification of
' types of elements is m:y to verify correct fuel loadings. The accidental rotation of fuel bundles containirig standard and EIP elements can be detected by visual -
inspection. Should this occur, however, studies of a single EIP element accidentally rotated into a standard fuel region indicate an insubstantial increase in power geaeration in the EIP element, b) A msximum uranium content of 9 Wt% in a standard TRIGA element is about 6%
greater than the desit ,n value of 8.5 Wt% Such an increase in loading would result in an increase in power density ofless than 6% An increase in local power density or 6% (
reduces the safety margin by at most 10% The maximum hydrogen 4o zirconium ratio of 1.8 will produce a maximum pressure within the clad during an accident well below the rupture strength of the clad.
5.2 Reactor Core Apolicability This spectfication applies to the configuration of fuel and in core experiments.
Obiective The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.
Soccifications -
a) The core shall be an arrangement of TRIGA uranium zirconium hydride fuel moderator bundles positioned in the reactor grid plate.
b) - The reficctor, excluding experiments and experimental facilities, shall be water or a combination of graphite and water or D,0.
I
-BatGE a) Standard TRIGA cores have been in use for years and their characteristics are well documented. EIP cores have been operated at General Atomics and the Puerto Rico
' Nuclear Center and their operational characteristics are availabic. Ceneral Atomics has also performed a series of experiments using standard and RIP fuel in mixed cores. In addition, stadies performed at Texas A&M for a variety of mixed core arrangements and opentional experience with mixed cores indicate that such loadings would safely satisfy alloperational requirements.
b) The core will be assembled in the reactor grid plate which is located in a pool oflight water. Water in combination with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation requirements.
5.3 Control Rods Aeolicability This specification applies to the control rods used in the reacter core.
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. NSC Technical Specificatisns, Version 01/15/98 Obeective The objectiw is to assure that the control rods are of such a design as to permit their use with a high degree of rcliability with respect to their physical and nuclear characteristics.
Specification
, a) The shim safety control rods shall have scram capability and contain barated graphite, l "
B4CNwder or boron an#. !'s compounds in solid fcnn as a poison in aluminum or stainless steel c! adding. T;eese rods may incorporate fueled followers which have the same characteristics as the fuel region in which they are used, b) The regulatir*g control rod need not haw acrum capability and shall be a stainless rod or
( contain the materials as specified for shim-safety contml rods 'Zhis rod may incorporate l
a fueled follower.
c) The transient control rod shall have scram capability and contain borated graphite or boron and its compounds in solid form as a poison in an aluminum or stamless steel clad. The transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate an alurr.inum or air follower.
11889E The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B4C powder or boron and its compounds, Since the regulating rod normally is a low -
worth sod, its function could be satisfied by using a solid stainless steel rod. These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insure mechantcal stability during movement and to isolate the poison from the pool water environment.
Contro' rods that are fuel followed provide additional reactivity to the core and increase the worth of the control rod. The use of fueled followers in the FLIP region has the additional advantage of reducing flux peaking in the water filled regions vacated by the withdraw 31 of the control rods. Scram capabilities are presided for rapid insertion of the control rods which is the primary safety feature of the reactor. The transient control rod is designed for a reactor pulse.
The nuclear behavior of the air or aluminum follower which may be incorporated into the transient rod is similar to a void. A voided follower may be required in certain core loadings to reduce flux peaking values.
5.4 Radiation Monitoring Sptem Applicability This specification describes the functions and essential components of the area radiation monitoring equipment and the system for continuously monitoring airborne radioactivity.
Obiective The objective is to describe the radiation monitoring equipment that is available to the operator to assure safe operation of the reactor.
Soccifl:atiotl The radiation monitoring equipment listed in the following tatle will have these characteristics.
Radiation Monitoring Channel end Function Arca Radiation Monitor (gamma sensitive instruments) 30
L
. NSC Technical Specifications, Version 01/15/9E Function: Monitor radiation fields in key locations, alarm and readout at control console and readout in reception room.
Continuous Air Radiation Monitor (beta-gamma sensitive detector with air collection capability)
Function: Monitor co'ncentration of radioactive particulate activity in building, alarm and readout at control console, and readout in reception room.
g Gas and Particulate Stack Radiation Monitors (gamma and beta-gamma sensitive 1_ detectors with air collection capability)
Function: Monitor concentration of radioacthe particulate activi:y sal radioactive gases in building exhaust, alarm and readout at control console and readout in recepdon room.
Ilalia 1( The radiation monitoring system is intended to prmide information to operating personnel of any impending or existing danger from radiation so that there will be sufHcient time to evacuate the facility and take the necessary steps to prevent the spread of radioactisity to the surroundings.
5.5 Fuct Storage Apolicability This specification applies to the stonge of reactor fuel at times wl l it is not in the reactor core.
I obiecths The objective is to assure that fuel which is being stored will not become critical and wili not reach an unsafe temperature.
Specifications a) All fuel elements shall be stored in a geometrical array for which the k-effective is less than 0.8 for all conditions of moderation.
b) Irradiated fuel elements and fueled detices shall be stored in an array which will permit sufficient natural comtetion cooling by water or air such that the fuel element or fueled device temperature will not ecced design values.
Datis The limits imposed by Specifv ations 5.5.a and 5.5.b are conservative and assure safe storage.
5.6 Reactor Building And Ventilmtion System Apolicability This specification cpplies to the building which houses the reactor.
Obiective The objective is to assure that provisions are made to restrict the amount of relem ' af radioactivity into the emironment.
Spgifications 2 a) The reactor shall be housed in a facility designed to restrict leakage. The minimum free 31
o
^
, NSC Technical Specifications, Version 01/15/98 volume in the facility shall be 180,000 cubic feet.
b) The reactor building shall be equipyxl with a ventilation system designed to filter and exhaust air or other gases from the reactor building and release them from a stack at a minimum of 85 feet from ground level.
c) Emergency shutdown controls for the natilation system shall be located in the reception
- room and the system shall be designed to shut down in the event of a substantial release of fission products.
Dama The facility is designed such that the ventilation system will normally maintain a negative pressure with respect to the atii.v.pl.cs, so that there will be no uncontrolled leakage to tie environment. De free air volume within the reactor building is confined when there is an emergency shutdown of the ventilation system. Controls for startup, emergency filtering, and normal operation of the watilation system are located in the reception room. Proper handling of aliborne radioactive materials (in enwrge.ncy situations) can be conducted fiom the reception room with a minimum of exposure to operating personnel.
5.7 - Reactor Pool Water Systems Anolienbility This specification applies to the pool containing the reactor and to the cooling of the core by the pool water, OhiCSlh5 The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.
Specifications a) The reactor core shall be cooled by natural convective water flow, b) The pool water inlet and outlet pipe to the demineralizer shall not extend more than 15 feet below the top of the reactor pool when fuel is in the core.
- t. . Diffuser and skimnwr pumps shall be located no more than 15 feet below the top of the reactor pool.
- 4) Pool water inlet and outlet pipes to the heat exchanger shall have emergency covers within the reac:or pool for manual shut off in case of pool water loss die to external pipe system failure.
c) A pool level alarm shall indicate loss of coolant if the pool level drops approximately 10% below operatinglevel.
Basa a) His specincation is based on thermal and hydrauhc calculations which show that the
$ TRIGA FLIP core can operate continuously in a safe maaner at power levels up to 2,700 kW with natural conwetion flow and sufficient bulk pool cooling. A comparison of operation of the TRIGA FLIP and standard TRIGA Mark III has shown them to be safe for the abose power level. Thermal and hydraulic characteristics of mixed cores are 32
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... - NSC Technical Specifications, Version 01/15/98 f essentially the same as that for 'IRIOA-FLIP and standard cores, b) In the event of accidental siphoning of pool water through inlet and outlet pipes of the -
demineralizer system, the pool water lewl will drop no more than 15 feet from the top of
. the pool.
c) In the ewnt of pipe faltare and siphoning of pool water through the skimmer and diffuser water systems, the pool water level will drop no more than 15 feet from the top of the pool.
d)- Inlet and outlet coolant lines to the pool heat exchanger terminate at the bottom of the pool. In the event of pipe failure, these lines must be manually scaled from within the reactor pool. Covers for these lines will be stored in the reactor pool. The time required to uncover the reactor core due to failure of a single pool coolant pipe system is 17 minutes, e) less of coolant alarm after 10% loss requires corrective action. This alarm is observed in the reactor control suom and the scaption room.
5.8 Physical Security The licensee shall maintain in effect and fully implement all provisions of the NRC staff approwd physical security plan, including amendments and clenges made pursuant to the authority of 10 CFR 50.54 (p). The appreved security plan consists of documents withheld from public disclosure pursuant to 10 CFR 2.70, collectively titled " Texas A&M University System, Nuclear Science Center Reactor Security Plan."
(-
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e NSC Tectuucal Specifications, Version 01/15/98 3 6.0 Adelaistrative Controls 6.1 Organization 6.1.1 Structure a) A line management orgatJzational structure shall prmide for the administration and operation of the reactor facility. 'Ihe Dept.ty Director of the Texas Engineering Experiment Station (TEES) Snd the Director of the Nuclear Science Center (NSC) shall have line management responsibility for adhenng to the terms and conditions of the Nuclear Science Center Reactor (NSCR) license and technical specifications and for safeguarding the public and facility personnel from undue radiation exposure The facility shall be under the direct control of the Director (NSC) or a licensed senior reactor operator designated by him to be in airect control.
b) Management ievels Lesel 1: Deputy Director TEES (Licensee): Responsible for the NSCR fr.cility liceuse.
Level 2: Director (NSC): Responsible for reactor facility operation and shall report to Lestl 1.
level 3: Senior Reactor Operator on Duty: Responsibic for the day-to-day operation of the NSCR or shift operation and shall report to Level 2.
Level 4: Reactor Operating Stas: Licensed reactor operators and senior reactor operators and trainees. These indisiduals shall report to Int! 3.
r.) Radiation Safety A qualified, health physicist shall be assigned responsibility for implementation of the radiation protection program at the NSCR. The indisidual shall report to Lesti i management. ,
d) Reactor Safety Board (RSB)
Responsible to tbc Licensee for providing an independent review and audit of the safety aspects of the NSCR.
6.1.2 Responsibility _
Responsibility for the safe operation of the reactor facility shall be in accordance with the line organization c,st.iblished in 6.1.1.
6.1.3 Stalling a) The minimum staffing when the reactor is not secured shall be as follows:
- 1) At least two individuals wiil be present at the facility complex and will consist of a licensed senior reactor operator and either a licensed reactor operator or operator trainee. During periods of reactor maintenance as specified in 1.27 (b) the reactor operator or the 34
. NSC Technical Specifications, Version 01/15/98 operator trainee may be replaced by maintenance personnel.'
- 2) A licensed scactor operator or senior reactor operator will be in the control room.
- 3) The' Director (NSC) or his designated alternate will be readily available for emergencies (i.e., capable of getting to the reactor facility within a reasonable time).
- 4) At least one member of the health physics support group will be readily available to provide advice and technical assistance in the area of radiation protection.
- The licensed senior reactor operator may be permitted as the only operations person present at the facility to perform a pre-startup check of the reactor or perform general reactor maintenance not specified in 1.27 (b).
b) A list of reactor facility personnel by name and telephone number shall be readily available for use in the control room. The list shall include:
- 1) Administrative personnel
- 2) Radiation safety personnel
- 3) Other operations persormel c) ne following designated individuals shall direct the events listed:
- 1) The Director (NSC) or his designated alternate shall direct any loading of fuel or control rods within the reactor core region.
- 2) The Director (NSC) or his designated alternate shall direct any loading of an in< ore experiment with a reactivity worth greater than one dollar.
- 3) The senior reactor operator on duty shall direct the recovery from an unplanned or unscheduled shutdown other than a safety limit violation.
6.1.4 Selection and Training of Personnel The xlection and training of operations personnel shall be in accordance with the following:
a) Responsibility: The Director (NSC) or his designated alternate is responsible for the training and requalification of the facility reactor operators and senior reactor operators.
b) Requalification Program 1.
Purpose:
To insure that all operating personnel maintain procciency at a level equal to or greater than that required for initial licensing.
- 2) Scope: Scheduled lectures, written examinations and evaluated console manipulations will be used to insure operator proficiency is maintair.ed.
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. NSC Technical Specificatisas, Version 01/15/98 6.2 Review and Audit 6.2.1 Reactor Safety Board The Reactor Safety Board (RSB) shall consist of at least three 3) voting members knowledgeable in fields which relate to nuclear safety. The RSB shall eview, evaluate and make recommendations on safety standards associated with the ope.ational use of- i the facility. NSCR operations and health physics shall be represented as *.x officio members on the RSB. The review and advisory functions of the RSB shalt include .
NSCR operations, radiation protection, and Le facility license.
6.2.2 - RSB Charter and Rules The operations of the RSB shall be in acwid nce with a written charter, including provisions for; a) Meeting frequency: not less than once per calendar year and as frequent as circumstances warrant consistent with effective monitoring of facility activities.
b) Voting rules c) Quorums ds Use of subcommittees e) Review, approval and dissemination of minutes 6.2.3 RSB Review Function The responsibilities of the RSB or a designated subcommittee thereofinclude, but are not limited to the following.
a) Review and evaluation of whether a proposed change, test, or experiment would constitute an unreviewed safety question or a change in Technical Specifications.
b) Review of nov procedures, major revisions of procedures, and proposed
- changes in reactor facility equipment or systems ! ving safety significance.
c) Review of new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity, d) Review of proposd 4. lunges in technical spec 3ications, license, or charter, c) Review of the NSCR radiation protection program.
f) Review of violations of technical specifications license, or charter, and violations ofinternal procedures or instructions having safety significance.
g) Review of operating abnormalities havir g safety significance.
h) Review of reportable occtnences listed in 6.6.2.
i) Revictv of audit repor'.s.
6.2.4 RSB Audit Function The RSB or a subcoir.mittee thereof shall audit reactor operations and radiation
. 36
. NSC Technical Specifications, Version 01/15/98 protection programs at least quarterly, but at intervals not to exceed four months.
Audits shall include but are not limited to the following:
a) Facility operations, including radiation protection, for conformance to the tecLdcal specifications, applicable license conditions, and standard operating procedures at least once per calendar year (interval between audits not to exceed 15 months),
b) The retraining and requalification program for the operating staff at least once per calendar year (inten al between audits not to exceed 15 months).
c1 The facility security plan and records at least once per calendar year (inten al between audits not to exceed 15 months).
d) The reactor facilic emergency plan and implementing procedures at Icast once per calendar year (interval between audits not to exceed 15 months).
6.2.5 Audit of ALARA Program The licensee or his designated alternate (excluding anyene whose normaljob function is within the NSCR) shall conduct an audit of the reactor facility ALARA program at least once per calendar year (interval between audits not to exceed 15 months). The results of the audit shall be transmitted by the licensee to the RSB at the next scheduled meeting. [
6.3 Operating Procedures Written operating procedures shall be prepared, reviewed, and approsed prior to initiating any of the activities lit i in this section. The procedures shall be, reviewed and approved by, the Director (NSC) is designated alternate and the Reactor Safety Board and shall be documented in a timely manne.. Procedures shall be adequate to assure the safe operation of the reactor but shall not preclude the use ofindependent judgment and action should the situation require such.
Operating procedures shall be in effect for the following items:
a) Startup, op. 4 tion, and shutdown of the reactor.
b) Fuel loading, enloading, and movement within the reactor.
c) Control rod removal or replacement.
- d. Routine maintenance of the c ntrol rod, drives and reactor safety and interlock systems or other routine maintenance t. at could have an effect on reactor safety.
c) Testing and calibration of reactor instrumentation and controls, control rod drives, area radiation monitors, and facility air monitors.
f) Administrative controls for operations, maintenance, and conduct ofirradiations and experiments, that could affect reactor safety or core reacthity.
g) Implementation of required plans such as emergency or security plans.
h) Actions to be taken to correct specifih and foreseen potential malfunctions of systems, including responses to alarms and abnormal reactisity changes.
Substantive changes to the above procedures shall be made effecthc caly after documented review and approval by the Director (NSC) and the Reactor Safety Board minor modifications or temporary changes to the original procedures which do not change their original intent may be 37
. NSC Technical Specifications, Version 01/15/98 made by the Director (NSC) or his designated alternate. All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Board.
6.4 Experiments Review and Apprmal Apprmtd experiments shall be carried out in accordance with establiskd and apprmed procedures.
a) All new experiments or class of experiments shall be reviewed by the RSB (C:cction 6.2.3) and implementation approved in writing by the Director (NSC) or his designated alternate.
b) Substantive changes to previously approved experiments shall be made only after review by the RSB and implementation approved in writing by the Director (NSC) or his designated alternate. Minor changes that do not significantly alter the experiment may be approved by the Director (NSC) or his designated alternate.
6.5 Required Actions 6.5,1 Action to be Taken in the Event a Safety Limit is Exceeded In the event a safety limit is exceeded a) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.
b) An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Board, and reports shall be made to the NRC in accordance with Section 6.6.2 of these specifications, and c) A report shall be prepared which shall include an analysis of the cause and extent of possible resultant damage, efficacy of correctist action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safety Board for resiew and then submitted to the NRC when authorization is sought to resume operation of the reactor.
6.5.2 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) Reactor conditions shall be returned to normal or the reactor shall be shut down. Ifit is necessary to shut down the reactor to contet the occurrence, operations shall not be resumed unless authorized by the Director (NSC) or his designated alternate.
b) The Director (NSC) or his designated alternate shall be notified and corrective action taken with respect w the operations invoked.
c) The Director (NSC) or his designated alternate shall notify the Chairman of .ae Reactor Safety Board.
d) A report shall be made to the Reactor Safety Board which shall include an analysis of the cause of the occurrence, etTicacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence, and 38
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." NSC Technical Specifications, Version 01/15/98 e) A report shall be made to the NRC in accordance with Section 6.6.2 of these specifications.
f) Occurrence shall be reviewed by the GB at their next scheduled meetin3 6.6 Reporting Requirements 6.6.1 Annual Report An annual report covering the operation of the reactor facility during the previous calendar year shall be submitted to the NRC prior to March 31 of each year providing the followinginformation; a) A brief narrative summary of(1) operating experience (including experiments
_ performed), (2) changes in facility design, performance characteristics, and operating procedures : elated to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections; b) Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality; c) The number of emergency shutdowns and inadvertent scrams, including reasons therefore; d) Discussion of the mqjor 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 requited; c) A brief description, including a summary of the safety evaluations of clumges in the facility or in procedures and of tests and experiments carried out pursuant to Sectio 7 50.59 of 10 CFR Part 50; .
f) 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. If the estimated awrage release aAer dilution or 6ffusion is less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient.
- 1) Liquid Waste (summarized on a monthly basis)
(a) Radioactivity discharged during the reporting period.
(1) Total radioactivity released (in Curies).
(2) The Effluent Concentration used and the isotopic
- composition if greater than 1 x 10-7 pCuries/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 pCuries/cc) during the reporting period.
(b) Total volume (in gallons) of effluent water (including 39
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- NSC Technical Specifications, Version 01/15/18 dilutent) during periods of release.
- 2) Gaseous Waste (sunmarized on a monthly basis)
(a) Radioacthity discharged during the reporting period (in Curies) for; .-
(1) Argon-41 (2) Particulates with half lives greater than eight days.
- 3) Solid Waste (a) The total amount of solid waste transferred (in cubic feet),
(b) The total acthity involved (in Curies).
(c) The dates of shipment and disposition (if shipped off site),
g) A summary of radiation exposures received by facility personnel and visitors, including dates and time where such exposures are greater than 25% of that allowed or recommended, h) A description and summary of any emironnental suntys performed outside the facility, ,
6.6.2 Special Reports in addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC headquarters, Office ofInspection and Enforcement as follows:
a) There shall be a report not later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to be followed by a written report that describes the circumstances of the event within 14 days of any of the following:
(1) Violation of safety limits (Sec 6.5.1).
(2) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal irglury, or exposure; (3) Any reportable occurrences as defined in Section 1,28 of these Specifications. The written report (and, to the extent possible, the preliminary telephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prev ~it reoccurrence of the event; (b) A written report within 30 days of:
(1) Personnel cha.tges in the facility organitation imching Level I and inti 2.
(2) Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
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4 NSC Technical Specifications, Version 01/15/98 6.7 Records Records of facility operations in the form oflogs, data sheets or oller suitable forms shall be retained for the period indicated as follows:
6.7.1 Records to be Retained for a Period of at 1. cast Five Years or for tiw Life of tle Component imotved a) Normal reactor facility operation b) Principal nu.intenance operations .
c) Reportable occurrences d) Surveillance activities required by the Technical Specifications e) Reactor facility radiation and contamination surveys where required by applicable regulations f) Experiments performed with the reactor g) Fuel inventories, receipts, and shipments h) Approved changes in operating procedures i) Records of meeting and audit reports of the RSB 6.7.2 Records to be Retained for at Least One Training Cycle Retraining and requalification of certified operations personnel. Records of the most recent complete cycle shall be maintained at all times the indisidual is employed.
6.7.3 Records to be Retained for the Lifetime of the Reactor Facility a) Gaseous and liquid radioactiw effluents released to the emirons. -'
b) Off-site emironmental monitoring surveys required by the Technical Specifications.
c) Radiation exposure for all personnel monitored, d) Drawings of the reactor facility.
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