ML20097F758
| ML20097F758 | |
| Person / Time | |
|---|---|
| Site: | Washington State University |
| Issue date: | 06/09/1992 |
| From: | WASHINGTON STATE UNIV., PULLMAN, WA |
| To: | |
| Shared Package | |
| ML20097F756 | List: |
| References | |
| NUDOCS 9206160007 | |
| Download: ML20097F758 (42) | |
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- APPENDIX A FACILrfY LICENSE NO. R 76 TECIINICAL SPECIFICATIONS AND BASES FOR THE WASliinGTON STATE UNIVERSITY MODIFIED TRIGA REACTOR DOCKET NO. 50-27 Updated Version of June 9,1992 9206160007 920609 PDR ADOCK 05000027 P.
PDR'
TALLE OF CONTENTS Page 1.0 DE FI NITIO NS.......................................
2 1.1 Reactor Operating Conditions........................
3 1.2 Reactor Experiment and Irradiations....................
3 1.3 Reactor Component..............................
3 1.4 Reactor Instrumentat on............................
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS...
6 2.1 Safety Limits - Fuel Element Temperature................
6 2.2 Limiting Safety System Settings.......................
6 3.0 LIMITING COND1 oCL'S OF OPERATION..................8 Q>
3.1 Steady-State Operation............................. 8 3.2 Reactivity Limitations..............................
8 3.3 Pulse Mode Operation.............................
8 3.4
- Maximum Excess Reactivity.........................
9 3.5 Core Configuration Limitation........................
9 3.6 Control and Safety System........................... 10 3.7r Radiation Monitoring System......................... 12 3.8 4 t Ar Discharge Limit............................... 13 3.9 Engineered Safety Feature - Ventilation System............. 13 3.10 Limitations on Experiments.......................,,
14 3.11 Limitations on Irradiations...........................
16 3.12 As Low As Reasonably Achievable (ALARA) Radioactive Ef fluent Releases..............,.................
16 3.13 Primary Coolant Conditions..........................
17 3.14 Sealed Sources in the Reactor Pool....................,.17 4.0 SURVEILLANCE REQUIREMENTS....................... 20
~ 4.1 G e ne ral.......................................
2 0 4.2 Safety Limit - Fuel Element Temperature................. 20 4.3 Limiting Conditions for Operation...................... 21 4.4 Reactor Fuel Elements............................
23 4.5 Primary Coolant Conditions.......................... 24 5.0 DESIGN FEATURES......................
25 5.1 R eactor Fuel....................................
25 5.2 -
Reactor Core................................... 26 5.3 Control Elements.............
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' 5.4 Radiation Monitoring System......................... 27 5.5 -
Fuel S torage.................................... 2 8 5.6 Reactor Building and Ventilation System.......
.......28 5.7 Reactor Pool Water System..
....,. 29 5.8 Physical Security.............
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TABLE OF CONTENTS (cent.)
Page 6.0 ADMINISTRATIVE CONTROL...........................
30 6.1.
Responsibility..................................
30
- 6. 2..
Org a n i za ti o n...................................
30 6.3 Facility Staff Qualifications.........................
30 6.4 Trai ni n g.....................................
30 6.5
- Reactor Safeguards Committee (RSC)..................
30 6.6 Quality Assurance...............................
33 6.7 -
Action to be Taken in the Event a Safety Limit is Exceeded....
33 6.8 Operating Procedures..............
34 6.9 Facility Operating Records..........................
34 6.10 Reporting Requirements............................
35 6.11 Written Communications....
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. Document Name:
WSU TECH SPECS APP A Requestor's ID:
'ARLENE Author's Name:
Gable / Wilson Document Comments:
Please retum this sheet when submitting corrections,
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TECHNICAL SPECIFICATIONS AND BASES FOR THE WASHINGTON STATE UNIVERSITY MODIFIED TRIGA REACIDR This document constitutes the Technical Specifications for Facility License No. R-76 and supersedes all prior Technical Specifications. Included in these Technical Specifications am the " Bases" to support the se.ection and significance of the specification. These bases are included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhen. Furthem1 ore, the dimensions, measurements, and other numerical values given in these specifications may differ slightly from actual values because of nomial construction and manufacturing tolerances, or normal degree of accuracy or mstrumentation.
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1.0 DEFINITIONS The following frequently used terms are herein explicitly defined to ensure unifonn inter-pretation of the Technical Specifications.
1.1 Reactor Operating Conditions Abnormal Occurrance: An abnonnal occurrence is defined for the purposes of the reponing requirements of Section 208 of the Energy Reorganization Act of 1974 (PL 93-438) as an unscheduled incident or event which the Nuclear Regulatory Commission determines is significant from the standpoint of public health or safety.
Cold Critical: The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperature both below 40'C.
Pulse Mode: Pulse mode operation shall mean any operation of the reactor with the mode selector switch i. '.he pulse position.
Reactor Ooeration: Reactor operation is any condition wherein the reactor is not secured.
Reactor Secured: The reactor is secured when all the following conditions are satisfied:
(1)
The reactor is shut down.
(2)
The console key switch is in the "off" position and the key is removed from the console and under the control of a licensed operator or stored in a locked storage area.
(3)
No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms, or insertion or withdrawal of in-core experiments.
Reactor Shutdown: The reactor is shut down when the n. actor is suberitical by at least 1.00$ of reactivity.
Reportable Occurrence: A reportable occurrence is any of the following that occur during reactor opemtion:
(1) operation with any safety system setting less conservative than specified in Section 2.2, " Limiting Safety System Settings" (2) operation in violation of a limiting condition for operation (3) operation with a required reactor or experiment safety system component in an inoperative or failed condition which could render the system incapable of performing its intended safety function (4) any unanticipated or uncontrolled change in reactivity greater than 1.00$
(5) an observed inadequacy in the implementation of either administrative or procedural controls, to such degree that the inadequacy could have caused the existence or development of a condition which could result in operation of the reactor outside the specified safety limits (6) release of fission products into the environment 2
Shutdown Margirl: Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that (1) the reactor can be made suberitical by means of the control and safety systems, starting from any pennissible operating conditions, and (2) the reactor will remain suberitical without further operator action.
Sleady4
.e Mode: Steady-state mode operation shall mean any operation of the reactor with the mode selector switch in the steady-state position.
1.2 Reactor Exneriments and Irradiations Esoeriment: Experiment shall mean: (1) any apparatus, device or material which is not a nonnal part of the core or expenmental facihties, but which is insened into these facilities or is in line with a beam of radiation originating from the reactcr, or (2) any operation designed to measure reactor parameters or characteristics.
Experimental Facilities: Experimental facilities shall mean beam ports, including extension tubes with shields, thennal columns with shields, vertical tubes, in-core irradiation baskets or tubes, pneumatic transfer systems, and any other in-pool irradiation facilities, Irradiation: Irradiation shall mean the inse: tion of any device or material that is not a nonnal part of the core or experimental facilities into an irradiation facility so that the device or material is exposed to a significant amount of the radiation available in that irradiation facility.
Irradiation Facilities: Any in-pool experimental facility that is not a nonnal part of the core and that is used to irradiate devices and materials.
Secured Exneriment: A secured experiment shall mean any experiment that is held finnly in place by a mechanical device or by gravity, that is not readily removable from the reactor, and that requires one of the following actions to permit removal:
(1) removal of mechanical fasteners (2) use of underwater handling tools (3) moving of shield blocks or beam port components 1.3 Reactor Comoonent FLIP Fuel: FLIP fuel is TRIGA fuel that contains a nominal 8.5 weight percent of uranium with a 2"U enrichment of about 70% and erbium, a burnable poison.
Euel Bundle: A fuel bundle is a cluster of three or four fuel rods fastened together in a square array by a top handle and bottom grid plate adapter.
Fuel Rod: A fuel rod is a single TRIGA-type fuel rod of either Standard or FLIP-type fuel.
Instrumented Fuel Rod: An instmmented fuel rod is a special fuel rod in which thennocouples have been embedded for the purpose of measuring the fuel temperatures during reactor operation.
hiixed Core: A mixed core is a core arrangement containing Standard and FLIP-type fuels with at least 22 FLIP fuel rods located in the central positions in the core.
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Onerational Core: An operational core is any arrangement of TRIGA fuel that is capable of operating at the maximum licensed power level and that satisfies all the requirements of the Technical Specifications.
Regulating Control Element: Regulating control element shall mean a low worth control element that may be positioned either manually or automatically by means of an electric motor-operated positioning system and that need not have a scram capability.
Standard Control Element: Standard control element shall mean any control element that has a scram capability, that is utilized to vary the reactivity of the core, and that is positioned by means of an electric motor-operated positioning system.
Standard Core: A standard core is any arrangement of all Standard fuel.
Standard Fud: Standard fuelis TRIGA fuel that contains a nominal 8.5 weight percent of i
uranium with a "U enrichment of less than 20%.
Transient Control Element: Transient control element shall mean any control element that has the capability of being rapidly withdrawn from the reactor core by means of a pneumatic drive, that is capable of being positioned by means of an electric motor-operated positioning system, and that ha: scram capabilities.
1.4 Reactor Instmmentation Channel Calibration: A channel calibration consists of comparing a measured value from the measuring channel with a corresponding known value of the parameter so that the measuring channel output can be adjusted to respond with acceptable accuracy to knowr.
values of the measured variables.
Channel Checi:: A channel check is a qualitative verification of acceptable performance by observation of channel behavior, This verification may include comparison with independent channels measuring the same variable or other measurements of the variables.
1 Channel Test: A channel test is the introduction of a signal into the channel to verify that it is operable.
Exoeriment Safety Systems: Experiment safety systems are those' systems, including their associated input circuits, that are designed to initiate a scram for the primary purpose of protecting an experiment or to provide infom1ation that requires manual protective action to be initiated.
Limiting Safety Systems Settinn: Limiting safety systems settings are the settings for automatic protective devices related to those variables having significant safety functions.
Measured Value: The measured value is the magnitude of that variable as it appears on the output of a measuring channel.
Measurine Channel: A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output devices that are connected for the purpose of measuring the value of a variable.
j; Ocerable: A system. device, or component shall be considered operable when it is capable l
of perfonning its intended functions in a normal manner.
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EIE (Peak-to-Measured Fuel Temocrature Ratioh The FTR is defined as the ratio tetween the maximum calculated fuel temperature in a given core arrangement to that mecsured by the instrumented fuel element. --
Reactor Safety Systems: Reactor safety systems are those systems, including their associated input cinaits, designed to initiate a scram for the primary purpose of 3rotecting the reactor or to provide information that icquires protective action to be initiatec Safety Chatmd: - A safety channel is a measuring channel in the reactor safety system.
Saferv Limits: Safety funits are limits on important process variables that are found to be necessary to reasonably protect the integrity of certain of he physical barriers which guard t
against the uncontrolled release of ra& _e.:vity.
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Fuel Element Temnerature Aonlicability: This specification applies to the temperature of the reactor fuel.
Objective: The objective is to define the maximum fuel temperature that can be permitted with confidence that a fuel cladding failure will not occur.
Soccificgions:
(1)
The naximum temperature in a Standard TRIGA fuel rod shall not excced 1000 C undu any condition of operation.
(2)
The maximum temperature in a FLIP-type TRIGA fuel rod shall not exceed 1150 C under any condition of operation.
Basrc: The important parameter for a TRIGA reactor is the fuel rod temperature. This parameter is well-suited as a singic specincation, especially since it can be measured, A loss it the integri;y of the fuel rud cladding could arise from a buildup of excessive pressu.e betwe.n ths. fuel moderator and the cladding if the fuel temperature exceeds the safety Emit. Thqec ure is caused by the presence of air, fission product gases, and hydroget from the disassociation of the hydrogen and zirconium in the fuel moderator.
The mag nitude of this pressure is determined by the fuel-moderator temperature and the ratio of L ydrogen to zircomum in the alloy. The safety limit for the TRIGA-FLIP fuel is based on data that indicate that the stirss in the cladding because of the hydrogen pressure from the disassociation of zirconium hydride will remain below the ultimate stress, provided the temperature of the fuel does not exceed 1150'C and the 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 because of hydrogen pressure from the disassociation of zirconium hydride will remain below the ultimate stress, provided that the temperature of the fuel does not exceed 1000aC and the fuel cladding is water cooled.*
2.2
- Limiting Safety System Settings Applicability: This specification applies to the settings that prevent the safety limit from being reached.
Objective: The objective is to prevent the safety limits from being reached.
SpecincMions: The limiting safety system settings shall be 500 C as measured in an instmmented fuel rod located in the central region of the core. For a mixed core, the instrumented rod shall be located in the region of the coie containing the FLIP-type fuel rods.
- G A-9064, Safety Analysis Repon for the Torrey Pines TRIG A Mark III Reactor, submitted under Docket No 50-227.
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Bases The limiting safety system setting is the measured instrumented fuel rod temperature that, if exceeded, shall inidate a scram to prevent the fuel temperature safety
- limit from being exceeded. Section 5.4 of the FLlP conversion safety analysis report f or the Washington State University (WSU) TRIGA reactor indicated that a 500*C safety system setting would limit the maximum possible steady-state fuel temperature in the FLIP fuel region to less than 800 C. This setting provides at least a 350*C margin of safety for FLIP fuel and at least a 200*C margin of safety for Standard fuel.
In the pulse mode of operation, the same limiting safety system setting will apply.
- However, the temperature channel will not limit the peak power generated during the pulse because of the relatively long response time of the temperature channel as compared with the width of a pulse. On the other hand, the temperature scram would limit the total amount of energy generated in a pulse by cutting off the " tail" of the energy transient in the event that the fuel temperature limit is exceeded. Thus, the fuel temperature scram provides an additional degree of safety in the pulse mode of operation to protect the fuelin the event of
. such conditions as sticking of the transient control element in the withdrawn position after a pulse.
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r 3.0 LIMITING CONDITIONS OF OPERATION 3.1 Steadv-State Ooeration Applicability: This specification applies to the energy generated in the reactor during steady-state operation.
Objective: The objective is to ensure that the fuel tcmperature safety limit will not be exceeded during steady state operation.
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- Specifications: The reactor power level shall not exceed 1.3 MW under any condition of operation.
Amendment No.13 Basis Themial and hydraulic calculations performed by the vendor indicate that TRIGA fuel may be safely operated up to power levels of at least 2.0 MW with natural convection cooling.
3.2 Reactlyjtv Limitations Aoplicability: These specifications apply to the reactivity condition of the reactor and the reactivity worth of control elements and experiments. They apply for all modes of operation.
Objecti-a: The objective is to ensure that the reactor can be shut down at all times and to ensure that the fuel temperature safety limit will not be exceedal.
SpecifLCations: The reactor shall not be operated unless the shutdown margin provided by control elements shall be 0.25$ or greater with:
(1) the highest wo;th nonsecured experiment in its most reactive state (2) the highest worth control element and the regulating element (if not scrammable) fully withdrawn s
(3) the reactor in the cold critical wndition without xenon D. asis: The value of the shutdown margin ensures that the reactor can be shut down from any operating condition even if the highest worth rod should remain in the fully withdrawn position. If the regulating rod is not scrammable,its worth is not used in detennining the shutdown reactivity.
3.3 Pulse Mode Ooeration Aoplicability: This specification applies to the peak fuel temperature in the reactor as a msult of a pulse insertion of reactivity.
Objective: The objective is to ensure that fuel element damage does not occur in any fuel rod during pulsing.
Specifications: The maximum reactivity inserted during pulse mode operation shall be such that the peak fuel temperature in any fuel rod in the core does not exceed 830'C The 8
maximum rufe allowable reactivity insertion shall be calculated annually for an existing core and prior to pulsing a new or modified core arrangement.
Ilasis: 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 which has a high creep strength and undergoes no phase changes at temperatures over 100(rC.
Ilowever, after extensive steady state operation at 1 MW, the hydrogen will redistribute due to migration from the central high temperature regions of the fuel to the cooler outer re ions. When the fuel is pulsed, the instantaneous temperature distribution is such that the hl hest values occur at the surface of the element and the lowest values occur at the center.
%c higher temperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that has now substantially increased above the nominal value. This produces hydrogen gas pressures considerably in excess of that expected for Zril.6. If the pulse l
insertion is such that the temperature of the fuel execch 874*C, then the pressure will be suf0cient to cae:c expansion of microscopic holes in the fuel that grow larger with each pulse. The expansion of the fuel stresses and distorts the fuel rod material which,in turn, can cause overall swelling and distortion of the cladding and entire fuel rod. The pulsing hmit of 830*C is obtained by exa.nining the equilibrium hydrogen pressure of zirconium hydride us..ction of temperature. The decrease in temperature from 874*C to 830*C reduces hydrogen pressure a factor of two, which rovides an acceptable safety factor.
This phenomenon does not ter the steady state safrt limit dere the total hydrogen in a fuel element does not change. Dus, the pressure exerted a 'he clad will not be signiGeantly affveted by the distribution of hydrogen, chin we element.
Amendment No.12 3.4 Maximum Excess Reactivity Apolicability: This specincation applies to the maximum excess reactivity, above cold critical, which may be loaded into the reactor core at any time.
Obiective: The objective is to ensure that the core analyzed in the safety analysis trport approximates the operational core within reasonab'e limits.
Sowincations: The maximum reactivity in excess of cold, xenon-free critical shall not exceed 5.6% Ak/k (8.00$).
Basis: Although maintaining a minimum shutdown margin at all times ensures that the reactor can be shut down, that specincation does not address the total reactivity available within the core. This specification, although over-constraining the reactor system, help 2.
ensure that the licensee's operational power densities, fuel t. mperatures, and temperatute peaks are maintained within the evaluated safety limits. The specified excess reactivity allows for power coefficients of reactivity, xenon poisoning, most experiments, and operational Hexibility.
3.5
ore Confieuration 1. imitation Apolicanility: This specification applies to mixed cores of FLIP and Standard types of fuel.
Obiective: The objective is to ensure that the fuel temperature safety limit will not be exceeded as a result of power peaking effects in a mixed core.
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Soecificah:
(1)
The FLlP fueled region in a mixed core shall contain at least 22 FLIP fuel rods in a contiguour block of fuel in the central region of the reactor core. Water holes in the FLIP region shall te limited to nonadjacent single-rod holes.
(2)
The IrIR as defined in Section 1.4 and as calculated by the method used in the FLIP conversion safety analysis report shall not exceed 1.5 for an operational core.
Bases: De limitation on the allowable core configuration as set forth in Section 4.1 of the t
FLIP fnel conversion t.afety analysis report limits power peaking effects. De limitation on power peaking effects ensures that the fuel temperature safety limit will not be exceeded in a mixed core.
A 50(PC safety system setting and a 1.5 P'IR limit the maximum possible steady-state fuel temperature in the FLIP region to less than 800#C.
3.6 Coultpl and Safety Systun 3.6.1 Scram Time Anollcability: Rb speciGcation applies to the time required for the scrammable control rods to te fully inserted from the instant that a safety channel variable reaches the safety system setting.
Objective: The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.
Specifications: The scram time from the instant that a safety system setting is exceeded to the instant that the slowest scrammable control rod reaches its fully inserted position shall not exceed 2 seconds. For purposes of this section, the above specification shall be considered to be satisfied when the sum of the response time of the slowest responding safety channel, plus the fall time of the slowest scrammable control rod, is less than or equal to 2 seconds.
Ilash: This specification ensures that the reacer will be promptly shut down when a scram signal is inithited. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to ensure the safety of the reactor.
3.6.2. Reactor Control System Aoplicabilitv: This specification applies to the information that must be available to the reactor operator during reactor operation.
Obiective: The objective is to require that sufficient infom1ation is available to the operator to ensure safe operation of the reactor.
Soecifications: The reactor shall not be o xrated in the specified mode of operation unless the measuring channels listed in Table 3.1 are operable.
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Table 3.1 Measuring Channels Min. no.
Effective inode Measuring channel onerable SS pulse Fuel element temperature 1
X X
Linear powerlevel 1
X Log power level 1
X Integrated pulse power 1
X Note: SS = steady state 13ases: Fuel temperature displayed at the control console gives continuous infonnation on this parameter, which has a specified safety limit. The power level monitors enr:4.;e that the reactor power level is adequately monitored for both steady state and pulsing unies of operation. *lhe specincations on reactor zwer level indication are included in this section snice the power level is related to the fue temperature.
3.6.3 Reactor Safety System Aoplicability: This specification applies to the reactor safety system channels.
Objective: The objective is to specify the minimum number of tractor safety system channels that must be operable for safe operation.
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Soecincations: The reactor shall not be operated unless the safety channels described in Table 3.2 are operable, 11
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i hble 3.2 Minimum Reactor Safety Channels Number operable in snecified mod,e Safety t'htinnel
" unction SS ~
l'ule e Fuel temperature hnif M temperature exceeds 5(XFC 1
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Power level Scram if power level exceeds 1100 kW l
l Manual scram Manually initiated scram 1
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Wide range Prevent initiation of a pulse above 1 kW l
Prevent control element withdrawal when neutron count is less than 2 cps 1
liigh voltage monitor Scram on loss of high voltage to pown channels 1
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- Pulse-mode switch Prevent withdrawal of standard c entrol and regulation elements in pulv mode 1
Preset timer Transient rod scram 15 secrads or less after pulse 1
Poollevel!
Alann if poollevel fallt telow 16 ft over the core 1
1 Transient rod centrol Prevent applicatio i of air unless fully inserted 1
Note: SS = steady state Amendment No.13 Bases: The fuel terncrature and p >wer level scrams provide protection to ensure that the reactor can be shut down before the safety limit on the fuel element temperature will be exceeded. The manual scram allows the o perator to shut down the system if an unsafe or
. abnonnal condition occurs. In ine event o: failure of the power supply for the safety chambers, operation of the reactor without adec uate instrumentation is prevented. The preset timer ensures that the reactor power leve' will reduce to a low les el after pulsing.
The interlock to prevent startup of the reactor with less than 2 eps ensures that sufficient neutrons are avadable for proper startup.
' The interlock to prevent the initiation of a pulse above 1 kW is to ensure that the magnitude of 'he pulse will not cause the fuel element temperature safety limits to be exceeded. The interlock to prevent withdrawal of the standard or regulating control elements in the pulse l
mode is to prevent the reactor from being pulsed whde on a positive period, The pool level alarm is intended to alert the operator to any significant decrease in the pool level.
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3.7 Radiation Monitorine System Applicability: This specification applies to the radiation monitoring information which must be available to :he reactor operator during reactor operation.
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i Oldesnyc: The objective is to ensure that sufnclent radiation monitoring is available to the operator to ensure safe operation of the reactor.
Specifications: The reactor shall not be operated unless the radiation monitoring channels listed in Table 3.3 are operable. Each channel shall have a readout in the control room and be capable of sounding an audible alann that can le heard ir the reactor control room.
Basis: The radiation monitors infonn operating personnel about any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.
Table 3.3 Minimum Monitoring Channels Channel
- Function No._
Area radiation monitor Monitor radiation level on the bridge 1
Area radiation monitor Monitor radiation level in the beam room 1
Continuous air monitor Monitor the activity of the pool room air 1
Exhaust gas monitor Monitor the Argon-41 activity in the exhaust 1
- For periods of time for mairtenance to the mdiation monitoring channels, the intent of this specification will be satisfied if they are te placed with portable gamma-sensitive instru-ments having their own alanns or that sha'l be kept under visual observation.
3.8 Argon-41 Discinggf.Limil 41 Arrlicattity: 'Ilis specification applies to the cont entration of Ar that may be discharged from the WSU TRIGA renctor facihty, i
Qbjective: To ensure that the heahh and safny of the public are not endangered by the discharge of 41 Ar from the WSU TRIGA waetor facility.
S2ecificatiRD: The concentration of 43Ar in the effluent gas from the facility, as diluted by 4
atmospheric air in the lee of the faci %ty as a result of the turbulent wake effect, shall not exceed 4 x 10 8 pel/mi averaged over one year.
Basis. The maximum allowable concentration of 41 Ar in air in unrestricted areas as specified in Appendix B Table 11 of 10 CFR 20 is 4 x 10 0 Ci/ml. Section 6.5 of the safety analysis repor1 for conversion of the W50 TRIGA reactor to Fl.lP fuel substantiates a 3.4 x 10' atmospheric dilution factor for a 4.4 mph windspeed. A somewhat more conservative value cf 4 x 104 has len selected for the calculation of 41 Ar dilution.
3.9 Illigineered Safety Featm - Ventilation System Applicability: This specification applies to the operation of the facility ventilation system.
Objective: The objective is to ensure that the ventilation system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation.
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Specifications: The reactor shall not be operated unless the facility ventilation system is operable, except for periods of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to pennit repair or testing of the ventilation system. In the event of a substantial release of airtome radioactivity within the facility, the ventilation system will be secured or operated in the dilution mode to prevent the release of a significant quantity of airbome radmactivity from the facility.
Basis: During nomial operation of the reactor and the ventilation system, the concentration of 41 Ar and other airborne radionuclides discharged from the facility is telow the applicable maximum pennissible concentration (MPC) values. In the event of a substantial release of airborne radioactivity within the facility, the ventilation system will be secured or operated in a dilution mode as appropriate. This action will permit minimizinj the concentration of airbome radioactive materials discharged to the environment until it. s within the appropri-ate MPC value. In addition, operation of the reactor with the ventilation system shut down for shon periods of time to make system repairs or tests does not compromise the control over the release of airborne radioactive materials. Moreover, radiation monitors within the building, independent of the ventilation system, will give warning of high levels of radiation that might occur during operation with the ventilation system secured.
3.10 Limitations on E ocriments t
Aeolicability: This specification applies to experiments installed in the reactor and its experimental facilities (denned in Section 1.2).
Objective: The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
Seccifications: The reactor shall not be operated unless the following conditions governing experiments exist.
(1)
Nonsecured experiments shall have reactivity worths less than 1.005.
(2)
The reactivity worth of any single experiment shall not exceed 2.00$.-
(3)
Total worth of all experiments will not exceed 5.00$.
(4)
Explosive materials, such as gunpowder, TNT, nitroglycerin, or PETN, in quantities greater than 25 mg shall not be irradiated in the reactor or expedmental facilities. Explosive materials in quantities less than 25 mg may be irradiated in the reactor or experimental facilities, provided the pressure procluced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.
(5)
Experiment materials, except fuel materials, which could off gas, sublime, volatilize, or produce aerosols under (a) normal operating conditions of the experiment or reactor, (b) credible accident conditions in the reactor, or (c) possible accident conditions in the experiment, shall be limited in activity so that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or
. the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the applicable limits of Appendix B of 10 CFR 20.
In calculations pur.,uant to item 4 above, the following assumptions shall be used:
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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 produced will escape.
If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 p particles, at least 10rfe of these particles can escape.
For materials whose boiling point is above 60 C and in cases where vapors fonned by boiling this material can escay only through an undisturbed column of water above the core, at least 10% of t1ese vapors can escape.
An atmospheric dilution factor of 4 x 103 for gaseous discharges from the facility.
(6)
Each fueled experiment shall be controlled so that the total inventory of iodine iso-topes 131 through 135 in the experiment is no greater than 1.5 Ci.
(7)
If a capsule fails and releases material that could damage the reactor fuel or structure by conusion or other means, that material shall te removed and physically inspected to determine the consequences and need for corn etive action. The results of the inspection and any corrective action taken shall be reviewed by the senior operator responsible for the operation and must be detennined to be satisfactory before operation of the reactor is resumed.
Bases:
(1)
This specincation is intended to provide assurance that the wonh of a single unse-cured experiment will be limited to such a value that the safety limit will not be exceeded if the positive worth of the experiment were to be suddenly inserted.
(2)
The maximum wonh of a single experiment is limited so that its removal from the cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since an experiment of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase so that the teactor protecuve systems would act tu prevent power levels from exceeding the safety limits.
(3)
The total worth of all experiments is limited to ensure that the reactor will remain suberitical in the event of a simultaneous removal of all of the experiments with one safety control element withdrawn.
(4)
This specification is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials.
(5)
This specification is intended to reduce the likelihood that radioactive ahborne particles in excess of the limits of Appendix B of 10 CFR 20 will be released to the atmosphere outside the facility.
(6)
The 1.5-Ci limitation on iodine isotopes 131 through 135 ensures that in the event of failure of a fue;ed experiment leading to total release m the iodine, the exposure i
15
m._ _._ ___ _
dose at the exclusion area boundary will be less than that allowed by 10 CFR 20 for an unrestricted area.
(7)
Operation of the reactcr with the reactor fuel or stmeture damaged is prohibited (to avoid release of fission products).
3.11 Limitations on irradiations Aoplicability: Ris specification applies to irradiations erfonned in the irradiation facilities contained in the reactor pool as defined in Section 1.2,p' Irradiation Facilities." Irradiations are a subclass of experiments that falls within the specifications herrinafter stated in this section The surveillance requirements for irradiations are given in Section 4.3.5(2).
Objective: The objective is to prevent damage to the reactor, excessive release of radio-active materials, or excessive personnel radiation exposure during the perfonnance of an irradiation.
Soccifications: A device or material shall not be irradiated in an irradiation facility under the classification of an irradiation unless all the following conditions exist:
(1)
The irradiation meets all the specifications of Section 3.10 for an experiment.
(2)
The expected radiation field produced in air by the device or sample upon removal from the reactor pool is not more than 10 rem /hr beta and gamma equivalent at I ft; otherwise,it shall be classed as an experiment.
(3) ne device or material is encapsulated in a suitable container.
(4) ne reactivity worth of the device or material is 0.25$ or less; otherwise, it shall be classed as an experiment.
(5)
The device or material does not remain in the tractor for more than a 15-day period; otherwise, it shall be classed as an experiment.
I} asis: This specification is intended to provide assurance that the special class of experi.
ments called irradiations will be performed in a manner that will not pennit any safety limit to be exceeded.
3.12 - Astow As Reasonablv' Achievable (ALARA) Radioactive Effluent fleleases Applicability: This specification applies to the measures required to ensuie that the radio-active efnuents released from the facility are in accordance with ALARA criteria.
operation of the %jective is to limit the annual opulation radiation exposure owing to theSU TRIGA reactor to a sma 1 percentage of the nonnallocal background Objective: The ob
?
exposure.
l Soecificatiotu:
(1)
In addition to the radiation monitoring specified in Section 5.4, an environmental radiation moni4% program shall be conducted to measure the integrated radiation exposure in an ! mund the environs of the facility on a quarterly basis.
l l
16 l
l
(2)
The annual radiation exposure due to reactor o wration, at the clo.est off site point of extended occupancy, shall not, on an annua' basis, exceed the aserage local off-site background radiation by more than 20%
(3)
Whenever practicable, the reactor shall be operated 4 in. or more from the thermal ralumn in order to minimize the production of dl Ar.
(4)
The total annual discharge of' tar into the environment shall not exceed 20 Ci per year.
(fj in the event of a significant fission product leak from a fuel nxi or a si;nificant airbome radioactive release from a sample being 1:Tadiated, as detectec. by the continuous air monitor, the reactor shall be shut down until the source of the leak is loc'ated an eliminated. Ilowever, the reactor may continue to be operated on a i
short tenn basis as needed to assist in detennining the source of the leakage.
(6)
Beiore discharge, the facility liquid effluents collected in the holdup tanks shall be analyr.ed for their beta gamma activity content. The total annual quantity of liquid effluents released (above background) shall not exceed 1 Ci per year.
Ilmis: The simplest and most reliable method of ensuring that ALARA release limits are accomplishing their objective of minirnal facility caused radiation exposure to the general public is to actually measure the integrated radiation exposure in the environment on and off the site.
3.13 Primarv Coolant conditions Apolicability: This specification applies to the quality of the primary coolant in contact with the fuel cladding.
Objectives: The objectives are (1) to minimize the possibility for corrosion of the cladding on the fuel elements, and (2) to mmimize neutron activation of dissolved materials.
Soecincations:
(1)
Conductivity of the pool water shall be no higher than 5 x 104 mhos/em.
(2)
The pil of the pool water shall be tatween 5.0 and 7.5.
Ilnsis: A small rate of corrosion continuously occurs in a water metal system. In order to limit this rate and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.
By limiting the concentrations of dissolved materials 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 vfill decrease personnel exposures during maintenance and operations.
3.14 Scaled Sources in the Reactor Pool Apatiatbility: This specincation applies to any and all se r.ed sources stored or used in the reactor pool.
17 4
)
1
Objective: %e objectives of this requirement are to ensuit that: 1) any scaled source or sourtes that are stored or used in the pool do not constitute any type of significant hazard to the operation of the tractor,2) any such sealed source or sources do not create a significant environmental or personal rudiation exposure hazani, and 3) any such scaled source or sourtes do not compromise the AaRA criteria of the facility.
Soccincations:
(1)
Scaled sourtes shall not at any time be stored or used closer than five (5) feet away from the face of an operating reactor core. The total activity of all scaled sou~ees stored in the pool shall not exceed 100,000 curies. All sealed source configurations shall be designed so that a loss of pool water accident will not preci pitate a scaled source iaapsulation integrity problem and the sources shall be stored in an appmpriate shield so as not to produce a significant radiation hazard in the event of a loss of reactor pool water accident.
(2)
All storage of seaW1 sourtesgoreater than 100 curies in the reactor pool shall be considered as an experiment and shall be reviewed and approved by the Reactor Safeguards Committee. A written operating procedure for the storage and use of scaled sources in the reactor pool shall be in effect.
(3)
The radionuclide content of the reactor pool water shall be monitored monthly at an inte: val not to exceed six (6) weeks in order to detect a significant leak in the sources stored in the reactor pool. if the specific radionuclide content of the pool wver for radionuclides from a sealed source stored in the tractor pool exceeds one-third (1/3) the 10 CFR 20 Appendix B. Table II, Column 2 value, steps shall be taken to isolate the source of the activity and to mitigate the problem.
Basis:
(1)
Limiting the proximity of sealed sources to five (5) or more feet away from the surface of the reactor core minimizes the effect of such sources on the reactor and the operation of the reactor upon the sources. The neutron flux at a distance of five (5) feet from the core surface is insignificant and thus could not cause activation of the sources and any associated sblelding. The 3resence of the sources in the pool would have no impact upon the D.D.A. which,s the mpture of the cladding on one fuel element, llowever, the presence of sources in the pool could contribute to the radiation hazard associated with a loss of pool water accident. The dose rate 25 feet above an unshicided core in the event of a loss of pool water accident would only be increased by less than 2% with the presence of 100,000 curies of *Co stored in the irradiation unit in the reactor pool.
(2)
Classifying the storage of sealed sources in the reactor pool as an experiment mandates that such ston>ge be reviewed by the Reactor Safeguards Committee.
l-l l
18 l.
(3)
The 10 CFR 20 Appendix II, Table 11 Column 2 limit for "Co is 3 x 102 pe#. At this limit the entire pool could be dumped into the WSU sewage system without taking advantage of the dilution factor associated with the discharge volume of the WSU sewage system. The detection limit for *Co in the reactor mol water depends upon toe system used but in the worst case would be at east 1 x lod pck, or 100 Pe#, or about one-thirtieth of the 10 CFR 20 limit stated above. Setting a limit of ten times the detection limit and one-third the discharge limits provide the facility with ample time to take corrective action in the event the limit is exceeded and does not compromise ALARA considerations.
l l
Ar mdment No.13 19
.i
r 4.0 SURVEILLANCE REQUIREMENTS 4.1 General Applicability: This specification applies to the surveillance requirements of any system related to reactor safety.
Objective: The objective is to verify the proper operation of any system related to reactor safety.
Specifications: Any additions, modifications, or maintenance to the ventilation system, the core and its associated su 1 port structure, the pool or its penetrations, the pool coolant sys.
tem, the control element c nye mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifications ap 3 roved by the Reactor Safeguants Committee. A system shall not be considered opera >le until after it has been successfully tested.
Basis: This specification relates to changes in reactor systems that could directly affect the safety of the reactor As long as changes or replacements to these systems contmue to meet the onginal design specifications, it can be assumed that they meet the presently accepted operating criteria.
4.2 Safety Limit - Fuel Element Temocrature Applicability: This speci0 cation applies to the surveillance requirements of the fuel element temperature measuring channel.
Objectise: The objective is to ensure that the fuel element temperatures are properly monitored.
l Specifications:
(1)
Whenever a reactor scram caused by high fuel element temperature occurs, the peak indicated fuel tem rature shall be examined to detennine whether the fuel element temperature safet limit was exceeded.
(2)
The fuel element temperature measuring channel shall be calibrated semiannually or at an interval not to exceed 8 months by the substitution of a thermocouple simula-L tor in place of the instmmented fuel element themmcouple.
L (3)
A channel check of the fuel element measuring channel shall be made each time the reactor is operated by comparing the indicated instmmented fuel element tempera-ture with previous values for the com configuration and power level.
- 1) mis:' Operational experience over the past 5 years with the TRIG A system gives assur-ance that the thermocouple measurements of fuel element temperature have been sufficiently reliable to ensure accurate indication of this parameter.
l 20
4.3 I imitine Conditions for Ooeration 4.3.1 Reactivity Requirements Applicability: These specifications apply to the sutveillance requirements for reactivity control of experiments and systems.
Objective: The objective is to measure and verify the wonh, perfom1ance, and operability of those systems affecting the reactivity of the reactor.
Soccifications:
(1)
The reactivity worth of each control rod and the shutdown margin shall be deter-mined annually but at intervals not to exceed 15 months.
(2)
The reactivity worth of an experinunt shall be estimated or measured, as appropri-ate', before reactor operation with said experiment.
(3)
The control rods shall be visually inspected for deterioration at intervals not to exceed 2 years.
(4)
The transient rod drive cylinder and associated air supply system shall le inspected, cleaned, and lubricated as necessary semiannually at miervals not to exceed 7.5 months.
(5)
The reactor shall be pulsed semi-i.dly to compare fuel temperature measurements and peak powerlevels with thos
' previous pulses of the same reactivity.
}
Basis. The reactivity worth of the control rods is measured to ensure that the required shut-down margin is available and to provide an accurate means for detennining the reactiv ty i
worths of experiments inserted in the core. l'ast experience with TRIGA reacters.,es assurance that measurement of the reactivity worth on an annual basis is adequate to ensure no significant changes in the shutdown margin. The visualinspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation in the reactor. The reactor is pulsed at suitable intervals and a comparison is made with previous similar pulses to detennine if changes in fuel or core characteristics are taking placc.
4.3.2 Control and Safety System Aeolicabilitv: These specifications apply to the surveillance requirements for measure-ments, tests, and calibrations of the control and safety systems.
Objective: The objective is to verify the perfonnance and operability of those systems and components which are dimetly related to reactor safety.
Soecifications:
(1)
The scram time shall be measured annually but at intervals not to exceed 15 months.
(2)
A channel check of each of the reactor safety system channels for the intended mode of operation shall be perfomied before each day's opcration or before each opera-tion extending more than 1 day, except for the pool level channel which shall be tested monthly.
4 21 m.
_________.__.____._._____..__._._____.________.__._h_____._._._______.______m___
(3)
A channel calibration shall be made of the power level monitoring channels by the calorimetric method annually, but at intervals not to exceed 15 months.
(4)
A channel test of each item in Tabic 3.2, other than measuring channels, shall be perfonned semlannually, but at intervals not to exceed 7.5 months.
Ilasis: 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 perfonn properly, The channel tests will ensure that the safety system channels are operable on a daily basis or before an extended run. The power level channel calibration will ensure that the reactor will be operated at the proper power levels. Transient control element checks and semiannual maintenance ensure proper operation of this control element.
4.3.3 Radiation Monitoring System Aoplicability: This specification applies to the surveillance monitoring for the, area momtonng equipment, Argon 41 monitonng system, and continuous air momtoring system.
Objectives: The objectives are to ensure that the radiation monitoring equipment is operating properly and capable of performing its intended function, and that the alarm points are set conectly.
Specincations: All radiation monitoring systems shall be verined to be operable at least i
monthly at an interval not to exceed 45 days. In addition, the following surveillance activities shall be perfonned on an annual basis at intervals not to exceed 15 months: 1) the area radiation monitoring system shall be calibrated using a certified source: 2) a calibration of the A-41 system shall be done using at least two different calibrated garr.ma-ray sources;
- 3) a calibration shall be perfonned on the CAM in terms of counts per unit time per unit of activity using calibrated beta sources.
Dasis: Experience has shown that monthly verification of Radiation Monitoring Systems' operability in conjunction with an annual more thorough surveillance is adequate to correct for any variations in the systems caused by a change of operating characteristics over a long t,imespan.
4.3.4 Ventilation System Anpheability: This specification applies to surveillance requirements for the pool room ventilation system.
Objective: The objective is to ensure the proper operation of the pool room ventilation system in the isolation and dilution modes, which would t,e used in controlling the release of sadioactive material to the uncontrolled environment in the event of an emergency.
Specincations: The operation of the pool room system shall be check monthly (at intervals not to exceed 6 weeks) by cycling the system from the " normal" to the " isolate" and
" dilution" modes of operation. The positions of the associated dampers, indicator display, and fan operation shall be visually checked to ensure correspondence between the device perfonnance and selected mode of operation. The nessure drop across the absolute filter g
m the pool ventilation system shall be measured at east twice a year. The absolu;e filter shall be changed at least every 2 years and whenever the pressure drop across the filter increases by 1 in, of water.
Amendment No.13 22
Dash: Experience has shnwn that the only reliable method of testing the ventilation is to cycle the system lato the various modes and visually check each ponion of the system for proper operation in that mode.
4.3.5 Experiment and hradiation Limits Applicability: This specification applies to the surveillanec requirements for experiments installed in the reactor and its experimental facilities and for irradia !ons performed in the irradiation facilities.
Specincations:
(1)-
A new experiment shall not be installed in the reactor or its experimental facilities until a hazards analysis has been performed and reviewed for compliance with
" Limitations on Experiments," Section 3.10, by the Reactor Safeguards Committee.
Minor mcdifications to a reviewed and appmved experiment may be made at the discretion of the senior operator responsible for the operation, provided that the hazards associated with the modhications have been reviewed and a determination has been made and documented that the modifications do not create a signincantly different, a new, or a greater hazard than the original approved experiment.
(2)
An irradiation of a new type of device or material shall not be performed until an analysis of the irradiation has been perfomied and reviewed for compliance with
" Limitations on Irradiations," Section 3.11, by a licensed senior operatc r qualified in health physics, or a licensed senior operator and a person quali0ed in health physics.
Basis: It has been demonstrated over a number of years that experiments and irradiations reviewed by the reactor staff and the Reactor Safeguards Committee, as appropriate, can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.
4.4 Reactor Fuel Elements Apolicability: This specification applies to the surveillance requirements for the fuel elements.
Objective: The objective is to verify the continuing integrity of the fuel element cladding.
Soecincations: All fuel elements shall be inspected visually for damage or deterioration and measured for length and bend at intervals not to exceed the sum of 3,500.00$ in pulse reactivity. The reactor shall not be operated with damaged fuel. A fuel element shall be considered damaged and must be removed from the core if:
(1) in measuring the transverse bend, its sagitta exceeds 0.125 in, over the length of the cladding (2) in ineasuring the elongation, its length exceeds its original length by 0.125 in.
(3) a clad defect exists as indicated by release of fission products 23
- -- u n.-.
Etsis: %e frequency ofinspection and measurement schedule is based on the parameters most likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levcis and utilizing fuel elements whose characteristics are well known.
The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal cf heat from touchin : fuel elements shows that there will be no hot spots resulting in damage to the fuel caused ay this touching. Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. He elongation limit has been specified to ensure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to ensure adequate coolant flow.
4.5 Primary coolant conditions Anolicability: This specification applies to the surveillance of primary water quality.
Obiective: He objective is to ensure that water quality does not deteriorate over extended periods of time if the reactor is not operated.
i Specification: The conductivity and pil of the primary coolant water shall be measured at least once every 2 weeks, and shall be as follows:
i (1) conductivity $ 5 x 104 mhos/cm (2) pli between 5.0 and 7.5 bisis: Section 3.3 ensures that the water quality is adequate during reactor operation.
Section 4.5 ensures that water quality is not permitted to deteriorate over extended periods of time even if the reactor does not operate.
24
5.0 DESIGN FEATURES 5.1 Reactor Fuel Applicabilitv: This specincation applies to the fuel elements used in the reactor core.
Objective: The objective is to ensure that the fuel eternents are of such a design and fabricated in such a manner as to pennit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Soccincations:
(1)
TRIGA-FLIP Fuel The individual unirradiated FLIP fuel elements shall have the following characteristics:
uranium content: maximum of 9 wt% enriched to nominal 70% 235U hydrogen to-zirconium ratio (in the Zrif ): between 1.5 and 1.7 t
natural erbium content (homogeneously distributed): between 1.1 and 1.6 wt%
claddi ; r 304 stainless steel, nominal 0.020 in. thick identification: top pieces of FLIP elements will have characteristic markings to allow visual identification of FLIP elements employed in mixed cores (1)
Standard TRIGA Fuel The individual unitradiated Standard TRIGA fuel elements shall have the following characteristics:
uranium content: maximum of 9.0 wt% enriched to less than 20% 235p hydrogen to zirconium atom ratio (in the Zrlix): between 1.5 and 1.8 cladding: 304 stainless steel, nominal 0.020 in, thick e
Basis: A maximum uranium content of 9.25 wt% in TRIGA FLIP elements is about 6%
preater than the design value of 8.5 wtE Such an increase in loading would result in an merease in power density of about 2%. Similarly, a minimum erbium content of 1.1 wt%
in an element is about 30% less than the design value. ' Itis variation would result in an increase in power dens"y of only about 6% An increase in local power density of 6%
reduces the safety margin by at most 10% The maximum hydrogen-to zirconium ratio of 1.75 could result in a maximum stress under accident conditions in the fuel element clad about a factor of 2 greater than the value resulting imm a hydrogen-to-zirconium ratio of 1.60. However, this increase in the clad stress during an accident would not exceed the ruptutf grength of the clad.
I 25 L
l A maximum uranium content of 9 wt% for the standard TRIG A elements is about 6%
);reater than the design value of 8.5 wth Such an increase in loading would result in an mercase in power density of 6% and reduces the safety margin by at most 10% %e maximum hydrogen to zirconium ratio of 1.8 could result in a maximum stress under i
accident conditions in the fuel element clad about a factor of 2 greater than the value resulting from a hydrogen to tirconNm ratio of 1.60. Ilowever, this increase in the clad stress during an accident would not exceed the rupture strength of the clad.
5.2 Reactor Com Applicability: His specification applies to the configuration of fuel and in-core experiments.
Objective: The objective is to ensure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assumnce that excessive power densities will not be produced.
Specifications:
(1)
The core shall tv an arrangement of TRIGA uranium zitronium-hydride fuel-moderator bundles positioned in the reactor grid plate.
(2)
%c TRIGA core assembly may be composed of Standard fuel FLIP fuel, or a combination thereof(mixed cores) provided that the FLIP fuel region contains at least 22 FLIP fuel rods located in a contiguous block in the central region of the core.
(3)
The reactor fueled with a mixture of fuel types shall not be operated with a core lattice position vacant in the FLIP fuel region. Water holes in the FLIP region shall be limited to single-rod holes. Vacant lattice positions in the core fuel region shall be occupied with fixrnes that will prevent the installation of a fuel bundle.
(4)
The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite, aluminum and water.
Ilnis: Standard TRIGA cores have been used for years and their characteristics are well-documented. -Mixed cores of FLIP and Standard fuel have been tested by General Atomics Co. and operated at a number of university reactors. Calculations, as well as measured aerformance of mixed cores in the WSU reactor, the Texas A&M reactor, and the Jniversity of Wisconsin reactor, have shown that such cores may be safely operated.
In mixed cores, it is necessary to arrange FLIP elements in a contipuous, central region of the core to control flux peaking and power generation peak values m individual elements.
Vacant core lattice positions in the Standard fuel region will contain experiments or an experimental facility to pmvent accidental fuel additions to the mactor core. Vacant core positions are not permitted in the FLIP fuel region as specified by Section 3.5.
The core will be assembled in the reactor grid plate which is located in a pool of light water.
Water in combination with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation requirements.
26
5.3 control Elemetus Applicability: This specification applies to the control elements used in the reactor core.
Obiective: The objective is to ensure that the control elements are of such a design as to pennit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Sacsi0 cations:
(1)
The standard control element shall have scram capability and contain borated A
graphi e, NC powder, or boron and its compounds in solid fonn as a poison in t
aluminum or stainless steel cladding.
(2)
The regulation control element need not have scram capability and shall be a stainless steel element or contain the materials as specified for standard control elements.
(3)
The transient control element shall have scram capability and contain borated graphite or boron and its compounds in a solid fonn as a poison in an aluminum or stainless steel clad. The transient element shall have an adjustable upper limit to allow a variation of reactivity insertions. This element may incorporate a nonfueled follower.
llasis: 'Ihe poison requirements for the control elements are satisf:ed by using neutron-absorbing borated ;raphite, II C powder, or boron and its com 30unds. Since the regulat-4 ing element nonna ly is a low worth element,its function couk be satisfied by using solid stainless steel. These materials must te contained in a suitable clad material, such as alumi-num or stainless steel, to ensure mechanical stability during movement and to isolate the poison from the pool water ervironment. Scram capabilities are provided for rapid inser-tion of the control element which is the primary safety feature of the reactor. He transient control element is assigned for a reactor pulse. The nuclear behavior of the nonfueled follower which may be incorporated into the transient element is similar to a void.
5.4 Radiation Monitorine System Aeolicability: This spe.!fication describes the functions and cWntia'l components of the area radiation monitoring equipment and the system for continuously monitoring airbome radioactivity.
Objective: The objective is to describe the radiation monitoring equipment that is available to the operator to ensure safe operation of the reactor.
Specifications:
(1).
ELLnction of Area Radiation Monitor (gamma-sensitiveinstruments): Monitor radiation fields in key locations, alann and readout at control console.
(2)
Function of Continuous Air Radiation Monitor (beta, gamma-sensitive detector with paniculate collection capability): Monitor radioactive paniculate activity in the pool room air, alann and readout at control console.
27
(3)
Function of Argon-41 Stack Monitor (gamma sensitive detector): Monitor 3Ar 4
content in reactor exhaust air, alam) and readout at console.
Amendment No.13 Ilasis: De radiation monitoring system is intended to provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioacti-vity to the surroundings.
5.5 Fuel Stontg-Applicability: Els specification applies to the storage of reactor fuel at times when it is not in the reactor core.
Objective: The objective is to ensure that fuel that is being stored will not become critical and will not reach an tmsafe temperature.
Soccifications:
(1)
All fuel elements shall be stored in a geometrical array where the kerr is less than 0.8 for all conditions of moderation.
(2)
Irradiated fuel elements and fueled devices shall be stored in an array, which will pennit sufficient natural convection cooling by water or air, so that the fuel element or fueled device temperature will not exceed design values.
Basis: The limits imposed by Specifications 5.5(1) and 5.5(2) are conservative and ensure safe storage.
5.6 Reactor Building and Ventilation System Applicability: This specification applies to the building that houses the tractor.
Obiective: The object ve is to ensure that provisions are made to restrict *he amount of i
radioactivity released into the environment.
Specifientions:
o (1)
The reactor shall be housed in a facility designed to restrict leakage. The minimum free volume in the facility shall be 109 cm.
3 (2)
The reactor building shall be equipped with a ventilation sy. stem designed to filter and exhaust air or other gases from the reactor building and release them from a stack at a minimum of 20 ft from ground level.
(3)
Emergency shutdown controls for the ventilation system shall be located outside the pool and control room areas and the system shall be designed to shut down in the event of a substantialirlease of airbome radioactivity within the facility.
(4)
. The pool room ventilation system shall have a dilution mode of operation in which air from the pool room is nuxed and diluted with outside air before being discharged from the facility.
28
Ibuis: De facility is designed so that the ventilation system will nonnally maintain a negative pressure with respect to the atmos 3 hex to minimize uncontrollable leakage to the environment. The free air volume within t ie reactor building is connned when there is an emergency shutdown of the ventilation system, limergency controls for startup, isolation, dilution, and normal operation of the ventilation system are located extemal to the control and pool rooms. Proper handling of aitterne radioactive materials (in emergency situations) can be effected with a minimum of exposure to operating personnel.
5.7 Reactor Pool Water Systems Apolicabilltv. His speci6 cation 1 ilies to the pool containing the reactor ar:d to the cooliag of the core by the pool water.
Oblective: The objective is to ensure that coolant water shall be available to provide ade-quate cooling of the reactor core and adequate radiation shielding.
Specifications:
(1)
The reactor core shall be cooled by natural convection water flow.
(2)
All piping extending more than 5 it below the su face of the pool shall have ade-quate provisions to prevent inadvertent siphonir g of the pool.
(3)
A pool level alarm shall be provided to indicate a loss of coolant if the pool level drops more than 2 ft below the nonnallevel.
(4)
The reactor shall not be operated w,h less than 15 ft of water above the top of the b
core.
Basis: This specification is based on thermal and hydraulic calculations which show that the TRIG A FLIP core can operate in a safe manner at power levels up to 2700 kW with natural convection flow of the coolant water. A comparicon between operation of the TRIGA FLIP and standard TRIGA MARK 111 has shown them to be safe for the above power level. Thennal and hydraulic characteristics of mited cores are essentially the same as those for TRIGA FLIP and standard cores.
In the event of accidental siphoning of pool water through system pipes, the pool water level will drop no more than 5 ft from tie top of the pool.
- Loss of coolant alann after 2 ft of loss requires corrective action. This alann is observed in the reactor control room, at the office, and at the campus police station.
5.8 Physical Security The Licensee shall maintain in effect and fully implement til provisions of the NRC staff-approved phypen security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of documents withheld from public disclosure pursuant to 10 CFR 2.70, collectively titled, Washington State University, Pullman, Washington TRIG A Reactor Security Plan."
29
6.0 ADMINISTRATIVE CONTROL 6.1 Responsibility The facility shall be under the direct control of a licensed Senior Reactor Operator (SRO) designated by the Director of die WSU Nuclear Radiation Center. The SRO shall be responsible to the Director for the overall facility operation including the safe operation and mamtenance of the facility and associated equipment. The SRO shall also be responsible for ensuring that all operations are conducted in a safe manner and within the limits mscritxxl by the facility license, Federal and State regulations, and requirements of the teactor Safeguards Committee.
6.2 Organization (1)
The reactor facility shall be an integral pan of the Nuclear Radiation Center of Washington State University. The organization of the facility management and operation shall be as shown in Figure 6.1. The responsibilities and authority of each member of the operating staff shall le defined b writing.
d)
When the reactor is not secured, the minimum staff shall consist of:
i (a)
Reactor Operator (RO) at the controls (may be the SRO)
(b)
Senior Reactor Operator (SRO) on call but not necessarily on site (c)'
another person present at the facility complex who is able to carry out prescribed written instructions 6.3 Facility Staff Oualifications Each member of the facility staff shall meet or exceed the minimum qualifications of ANS 15.4,"Standarti for the Selection and Training of Personnel for Research Reactors," for comparable positions.
6.4 Training The licensed Senior Reactor Operator designated by the Director as being responsible for the facility also shall be responsible for the facility's Requalif~ication Training Program and Operator Training Program.
6.5 Reactor Safeguards Committee (RSQ 6.5.1 Function The RSC shall function to provide an independent review and audit of the facility's activities including:
(1) reactor operations (2) radiological safety (3) general safety (4) testing and experiments (5) licensing and reports (6) quality assurance 30
President of WSit Dean of the Graduate School j
Director Nuclear Radiation Center Reactor Safeguards Committee l
I Review Audit
, Function Function Radiation Safety
---[----]
Committee Facility Director Rasctor Supervisor and Operating Staff Figure 6.1 Facility organization 31
l 6.5.2 Composition and Qualifications The RSC shall te composed of at least five members knowledgeable in fields that relate to nuclear reactor safety. 'The members of the Committee shallinclude one facility Senior Reactor Operator and WSU faculty and staff members designated to serve on the Committee in accordance with the procedures s xcified by the WSU committee manual, o
The University's Radiation Safety Supervisor s iall te an ex officio member of the Committee.
6.5.3 Operation 4
i The Reactor Safeguards Conunittee shall operate in accordance with a written charter, inelading provisio... Mr:
(1) meeting frequency: the full committee shall meet at least semiannually and a sulrommittee thereof shall meet at least semiannually (2) voting rules (3) quomms: chairman or his designate and two memters (4) method of submission and content of presentations to the conunittee (5) use of subcommittees (6) review, approval and dissemination of minutes 6.5.4 Reviews
'lhe responsibilities of the RSC or designated subcommittee thereof shall include, but are not limited to, the following:
(1) review and approval of all new experiments utilizing the reactor facility (2) review and ap 3roval of all proposed changes to the facility license by amendment, and to thelec inical Spxifications
.(3) review of the operation and operational records of the facility (4) review of significant operating abnomialities or deviations from nonnat and expected performance of facihty equipment that affect nuclear ufety (5) review and approval of all detenninations of whether a proposed change, test, or experiment would constitute a change in the Technical Specifications or an unre.iewed safety question as defined by 10 CFR 50 (6) review of reponable occunences ?.nd the repons filed with the Commissions for said occurrences (7) review and approval of all standard operating procedures and changes thereto (8)'
biennial review of all standard procedures, the facility emergency plan, and the facility security plan 32 1
. _.~
6.6.5 Audits The RSC or a subcommittee thereof shall audit reactor operations semlannually, but at intervals not to exceed 8 months. The s,emiannual audit shall include at least the following:
)
(1) veview of the reactor operating records i
(2) inspection of the reactor operating areas (3) review of t3nusual or abnonnal occurrences (4) radiation exposures at the facility and adjacent envitons 6.5.6 Records The activities of the RSC shall be documented by the secretary of the Committee and distributed as follows:
(1)
A written report of all audits performed under Section 6.5.5 shall be prepared and forwarded within 30 days in the Dean of the Graduate School and Facility Director.
(2)
A written repor* of all reviews perfonned under Section 6.5.4 shall be prepared and forwarded to the Facility Director within 30 days following the completion of the review.
_ (3)
The Accretary < 'se RSC shall maintain a Glc of the minutes of all meetings.
6.6 Ounlity Ar ar.irce-In accordance with Regulatory Guide 2.5 and ANSI 402, " Quality Assurance Pmgram Requirements for Researth Reactors," Section 2.17 the " facility shall not be required to prepare quality assurance documentation for the as built facilit >." Quality Assurance (QA) tenirements will still be limited to those specified in Section.17 as follows:
"All replacements, modification, and changes to systems having a safety irlated function shall be subjected to a QA review. Insofar as possible, the replacement, modification, or change shall be documented as meeting the requirements of the original system or component and have equal or better perfomsance or reliabihty."
"The required audit function shall be perfonned by the RSC specified in Section 6.5."
6,7 Action To Be Taken in the Event a Safety Limit is Exceeded in the event a safety. limit is exctedei (1)
The reactor shall ne shut down and reactor o J./ha shall not be resumed until authorized by the U.S. Nuclear Regulatory C6p ission (NRC).
(2)
An immediate report of the occurrence shall be made to the Chainnan of the Reactor Safeguards Commhtee, and reports shall be made to the NRC in accordance with Section 6.10 of these speelfications.
(3)
A report shall be prepared that shall include.n analysis of the causes and extent of possible resultant damage, ef6cacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This tepon shall bc 33
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submitted to the Reactor Safepuards Committee for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.
1 (4)
A report shall be made to the NRC in accordance with Section 6.10 of these specifications.
6.8 Qoerating Procedures Written operating procedures shall be adequate to ensure the safety of operation of the reactor, but shall not preclude the use ofirdependent judgment and action should the situation require such. Operating procedures shall be in effect for the following items:
(1) perfonning irradiations and experiments I
(2) startup, operation, and shutdown of the reactor (3) eniergency situations including provisions t'or building evacuation, earthquake, radiation emergcneies, fire or explosion, personal injury, civil disorder, and bomb threat (4) core changes and fuel movement (5) control element removal and replaceme.nt (6) performing preventive maintenance and calibration tests on the reactor and associated equipment (7) power equipment Substantiative changes to the above procedures shall be made only with the approval of the licensed SRO directly in charge of the facility. Temporary changes to the procedures that do not change their originalintent may be made by a licensed SRO. All such temporary changes shall be documented and subsequently reviewed by the licensed SRO directly in charge of the facility.
6.9_
Facility Operating Records in addition to the requiirments of applicable regulations, and in no way substituting for those requirements, records and logs shall be prepared for at least the following items and -
trtained for a period of at least 5 years for items (1) through (6) and indefinitely for items (7) through (11).
(1).
normal reactor operation
-(2) princioal maintenance activities (3) abnormal occurrences
'(4) equipment and component surveillance activities required by the Technical Specifications (5) experiments perfomed with the reactor (6) gaseous and liquid radioactive effluents released to the environs (7) off-site inventories and transfers (R) fuelinventories and transfers (9) facility radiation and contamination surveys (10) radiation exposures for all personnel (11) updated, corrected, and as-built drawings of the facility 34
6.10 Reportine Reuuironentu In addition to the requirements of applicable regulations, and in no way substituting for those requir-ments, trports shall te made to the Nuclear Regulatory Commission as follows:
(1)
A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and FAX to the Regional Administrator, USNRC, Region V Office, ard to the NRC Operations Center, of (a)
Any accidental irlease o? radioactivity above pennissible limits in unrestricted areas whethec or not the release resulted in propeny damage, personal injury.or exposute; (b)
Any violation of the safety liinir; (c)
Any reportable occurrence as defined in Section 1.1, " Reportable Occunence," of these specificadons.
(2)
A report within 10 days in writing to the Director Office of Nuclear Reactor Regulation USNRC, Washington, D.C. 20555, with a copy to the Regional Administrator, USNRC, Region V Office, of (a)
Any accidental release or radioactivity above pennissible limits in unrestricted areas whether or not the re' ease resulted in propeny damage, personal injury, or exposure. The written report (and, to the extent possible, the preliminary telephone or tehgraph report) shall describe, analyze, and evaluate safety implications,and outline the corrective measures taken or planned to prevent recurence of the event; (b)
Any violation of a safety limit;
-(c)
Any reportable occurrence as riefined in Sectian 1.1," Reportable Occurrence," of these specifications.
(3)
A report within 30 days in writing to the Director, Offict of Nuclear Reactor Regulation, USNRC, Washington, D.C. 20555, with a copy to the Regional Administrator, USNRC, Region V Office, of (a)
Any significant variation of meast. red values from a corresponding predicted or previously measured value of safety-connected operating characteristics occurring during operation of the reattor, (b)
Any signincant change in the transien' or accident an dysis as described in the Safety Analysis Report; (c)
Any significant changes in facility organization; (d)
Any observed inadequacies in the implementation of administrative or procedural controls.
35
(4)
A repon within 60 days after completion of stanup testing of the reactor (in writing to the Director, Office of Nuclear Reactor Regulation, USNRC, Washiagton, D.C.
20555) upon trecipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measu ed values of the operating conditions including:
(a)
An evaluation of facility perfonnance to date in comparison with design predictions and specifications; (b)
A reassessment of the safety analysis submitted with the license applient on in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.
(5)
An annual repon within 60 days follo,ving the 30th of June of each year (in writmg to the Director, Division of 1.icensing, USNRC, Washington, D.C. 20555) with a copy to the Regional Administrator, USNRC, Region V Office, providing the following information:
W)
A brief narrative summaty of (i) operating experience (including expeiiments perfomied), (ii) changes in facility design, perfonnance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (iii) results of surveillance tests ano inspectiens; (b)
TabulatNn of the energy output (in megawatt-days) of the reactor, hours reactor wa critical :he cumulative total energy output since initial criticality, and numte of pulses greater than 1.00$;
(c) ne number et emergency shutdowns and inadvertent scrums, including reasons for them; (d)
Discussion of the najor maintenance operations perfonned during the period, including the effect,if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required; (e)
A brief description, inch. ding a summary of the safety evaluations of changes in the facility or ia proceduits and of tests and experiments canied out pursuant to 10 CFR 5039; (f)
A summary of the nature and amount of radioactive effluents released or discharged to the environs beyoad the effective control of the licensee as measured at or before the point of such release or discharge:
36
4
([
Liquid Waste (summarized on a monthly basir)
Jfd (i) radioactivity discharged during the reporting period e
total estimated quantity of radioactivity released (in curies),
an estimation of the specific activity for each detectable radionuclide present if the specific activity of the re' eased material after dilution is greater tha,10 CFR 20, Appendix B limits, h?
summary of the total release (in cunes) of each nuclide 1
e deternined just above for the reporting period based on f
representative isotopic analysis,
~-
estimated average concentration of the released tr.dioactive e
material at the point of release for the reporting period in tenns of pCi/cc and fraction of the applicable MPC, (ii) total volume (in gallons) of effluent water (including dilutant) tele 6r,ed during each period of release Gaseous Waste (summarized on a monthly basis) y V
(i) radioactivity discharged during the reporting period (in curies) total estimated quantity of mdioactivity released (in curies) dete.rmined 'oy an appropriate sampling and counting x
- method, total estimated quantity oiM Ar reler. sed (in curies) during the reporting period based on data from an appropriate monitoring system, j
estimated average atmospheric diluted concentration of M Ar released during the rep Aing period in terms of pCi/cc and fraction of the applicable MPC value, total estimated quantity of radicar,ity in particulate form with half-lives greater than 8 days (in curies),eleased during (Se reporting period as determined by an appropriate particu-late nmnitoring system, average concentration of radioactive particulates with half-lives greater than 8 days released in pCi/cc during the report-ing period, and 37
an estimate of the average concentration of other significant radionuclides present in the gaseous waste discharge in tent' 'f pCi/cc and fraction of the applicable MPC value for the re ing period it the estimated release is greater than 20Vo of tne applicable Mm Solid Waste (summarized on an annual t: asis)
(i) total amount of solid waste packaged (in cubic feet),
(ii) total activity in solid waste (in curies),
(iii) the dates of shipment and disposition (if shipped off-site).
(g)
An annual summary of the radiation exposure received by facility personnel and visitors in terms of the average radiation exposure per individual and greater exposure per individual in the two groups. Each significant expo-sure in excess of the limits of 10 CFR 20 should be reported, including the time and date of the exposure as well as the circumstances that led up to the i
exposure; (h)
An annual summary of the radiation levels of contamination observed during routine surveys performed at the facility in terms of the average and highest levels; (i)
An annual summary of any environmental surveys perfomied mitside the facility.
6.11 Written Communications All written communications with the Nuclear Regulatory Commission shall be niade in accordance with the requirements of 10 CFR 50.4 " Written Communications."
Amendment No.13 38
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