ML20058M098

Proposed Tech Specs for Compliance W/New 10CFR20 & Regulations

ML20058M098
Person / Time
Site:	Washington State University
Issue date:	11/16/1993
From:	WASHINGTON STATE UNIV., PULLMAN, WA
To:
Shared Package
ML20058M090	List: ML20058M084 ML20058M098
References
NUDOCS 9312200148
Download: ML20058M098 (41)

Text

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APPENDIX A FACILITY LICENSE.NO. R-76 '!

i TECHNICAL SPECIFICATIONS  !

AND BASES ~j t

FOR THE _;

WASHINGTON STATE UNIVERSITY -

l MODIFIED TRIGA REACTOR i

DOCKET NO. 50 j ,

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Updated Version of November 16,1993 4

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. . , Amendment No.14 1 9312200148 931212 PDR 8'  !

P-ADOCK 05000027  % i PDR E f

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TABLE OF CONTENTS j i

Page j 1.0 DEFINITION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2 ' l 1.1 Reactor Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . 3'  !

1.2 Reactot Experiment and Irradiations. . . . . . . . . . . . . . . . . . . . 3. i 1.3 Reactor Component. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.4 Reactor Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4  ;

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS. . . 6 ; i i

2.1 ' Safety Limits - Fuel Element T5mperature . . . . . . . . . . . . . . . . . 6 {

2.2 Limiting Safety System Settings. . . . . . . . . . . . . . . . . . . . . . . 16  !

3.0 LIMITING CONDITIONS OF OPERATION. . . . . . . . . . . . . . . . . . . 8  !

3.1 Steady-State Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 .

3.2 Reactivity Limitations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 .

3.3 Pulse Mode Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 -  !

3.4 Maximum Excess Reactivity. . . . . . . . . . . . . . . . . . . . . . . . . 9 :j 3.5 Core Configuration Limitation. . . . . . . . . . . . . . . . . . . . . . . . 9-3.6 Control and Safety System. . . . . . . . . . . . . . . . . . . . . . . . . . . 10 ; .

3.7 Radiation Monitoring Systen . . . . . . . . . . . . . . . . . . . . . . . . . - 12 ; i 3.8 U Ar Discharge Limit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 ~

3.9 Engineemd 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 - ,

Effluent Releases. . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 16 ,

3.13 Primary Coolant Conditions.. . . . . . . . . . . . . . . . . . . . . . . . . . . - 17  :

3.14 Sealed Sources in the Reactor Pool . . . . . . . . . . . . . . . . . . . . . .17 1 4.0 SURVEILLANCE REQUIREMENTS. . . . . . . . .. . . . . . . . . . . . . . ... 20 -  !

i 4.1 Ge neral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . 20 '

4.2 Safety Limit - Fuel Element Temperature . . . . . . . . . . . . . . . . . 20 '

4.3 Limiting Conditions for Operation. . . . . . . . . . . . . . . . . . . . . . 21.  :

4.4 Reactor Fuel Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 . i 4.5 Primary Coolant Conditions . . . . . . . . . . . . . . . . . . . . . . . . . 24 L i T

5.0 DESIG N FEATUR ES . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 25 - I 1

5.1 R eactor Fuel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 >

5.2 Reactor Core. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26' 5.3 Control Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 J 5.4 Radiation Monitoring System. . . . . . . . . . . . . . . . . . . . . . . . . 27 ,

5.5 Fuel Storage . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 28 '

5.6 Reactor Building and Ventilation System . . . . . .. . . . . . . . . . . . . 28 -

5.7 Reactor Pool Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . '29 - ,

5.8 Physical Security . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7 j i

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TABLE OF CONTENTS (cont.) l r .i Page [

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6.0 ADMINISTRATIVE CONTROL. . . . . . . . .. . . . . . . . . . . . . . . . . . . 30 ;j 6.1 Responsibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

30 l 6.2 Organ:zation .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . L30- j 6.3 Facility Staff Qualifications . . . . . . . ; . . . . . . . . . . . . . . . 30  !

6.4 Trai nin g. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 -

j 6.5 Reactor Safeguards Committee (RSC). . . . . . . . . . . . . . . . . . 30 t 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 j 6.9 Facility Operating Records. . . . . . . . . . . . . . . . . . . . . . . . . 34 [

6.10 Reporting Requirements. . . . . . . . . . . . . . . . . . . . . . . . . . 35  :

6.11 Written Communications . . . . . . . . . . . .. ............... - 38  :

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Amendment No.14

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TECHNICAL SPECIFICATIONS AND BASES FOR THE WASHINGTON STATE UNIVERSITY MODIFIED TRIGA REACTOR  ;

This document constitutes the Technical Specifications for Facility License No. R-76 and  !

supersedes all prior Technical Speci6 cations. Included in these Technical Specifications i are the " Bases" to support the selection and significance of the specification. These bases l are included for information purposes only. They are not part of the Technical l Specifications, and they do not constitute limitations or requirements to which the licensee I must adhere. Furthermore, the dimensions, measurements, and other numerical values i given in these specifications may differ slightly from actual values because of normal construction and manufacturing tolerances, or normal degme of accuracy or l instrumentation.

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Amendment No.14

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! 1.0 DEFINITIONS l

The following frequently used terms are herein explicitly defined to ensum unifonn inter-pretation of the Technical Specifications.

1.1 Reactor Ooeratine Conditions l

l Abnormal Occurrence: An abnormal 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 mactor with the mode selector switch in the pulse position.

Reactor Oneration: Reactor operation is any condition wherein the reactor is not secured.

I 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.

l (3) No work is in progress involving in-core fuel handling or refueling operations, l 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 reactor is suberitical by at least 1 1.00$ of reactivity.

Reoortable Occurrence: A reportable occurrence is any of the following that occur during mactor operation:

(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 of operation listed in Section 3.0 l (3) operation with a required reactor or experiment safety system component in an inoperative or failed condition which could render the system incapable of perfomiing its intended safety function l (4) any unanticipated or uncontrolled change in reactivity greater than 1.00$

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(5) a 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 l

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Shutdown Margin: 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 permissible opemting conditions, and (2) the reactor will mmain suberitical without further operator action.

t Steady-State Mode: Stead, _, ate mode operation shall mean any operation of the reactor with the mode selector switch in the steady-state position.

1.2 Reactor Exneriments and hradiations Exneriment: Experiment shall mean: (1) any apparatus, device or material which is not a normal part of the core or experimental facilities, but which is inserted into these facilities f or is in line with a beam of radiation originating from the reactor, or (2) any operation designed to measure reactor parameters or characteristics.

Experimental Facilities: Experimental facilities shall mean beam ports, including extension  :

tubes with shields, thermal 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 insertion of any device or material that is not a nomial 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

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Irradiation Facilities: Any in-pool experimental facility that is not a normal part of the core l

and that is used to irradiate devices and materials.

Secured Experiment: A secured experiment shall mean any experiment that is held fm' nly in l 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 pennit removal:

(1) removal of mechanical fasteners (2) use of underwater handling tools (3) moving of shield blocks or beam port components 1.3 Reactor Component FLIP Fuel: FLIP fuel is TRIGA fuel that contains a nominal 8.5 weight percent of uranium with a "U enrichment of about 70% and erbium, a burnable poison.

Fuel 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 instrumented fuel rod is a special fuel rod in which thermocouples have been embedded for the purpose of measuring the fuel temperatures during reactor operation.

Mixed 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.

Regulatine Control Element: Regulating control element shall mean a low worth control l 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 l positioned by means of an electric motor-operated positioning system.

Standard Com: A standard com is any arrangement of all-Standard fuel.

Standard Fuel: Standard fuel is TR. IGA fuel that contains a nominal 8.5 weight percent of uranium with a "U enrichment ofless than 20%

i 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 has scram capabilities.

1.4 Reactor Instrumentation l 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  ;

measur:ng channel output can be adjusted to respond with acceptable accuracy to known values of the measured variables.

. Channel Check: A channel check is a qualitative verification of acceptable peiformance 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.

Exneriment Safety Systems: Experiment safety systems are those systems, including their I

associated input circuits, that are designed to initiate a scram for the primary purpose of j protecting an experiment or to provide information that requires manual protective action to be initiated.

Limiting Safety Systems Setting: Limiting safety systems settings are the settings for  ;

automatic protective devices related to those variables having significant safety functions. l i

Measumd Value: The measured value is the magnitude of that variable as it appears on the output of a measuring channel.

l Measurine Channel: A measuring channel is the ccmbination of sensor, interconnecting cables or lines, amplifiers, and output devices that are connected for the purpose of '

measuring the value of a variable.

I Onerable: A system, device, or component shall be considered operable when it is capable of performing its intended functions in a normal manner.

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FFR (Peak-to-Measured-Fuel Temocrature Ration The PTR is defined as'the ratio between -

the maximum calculated fuel temperature in a given core arrangement to that measured by. l the instrumented fuel element.  ;

Reactor Safety Systems: Reactor safety systems are those systems, including their-  !

associated input circuits, designed to inidate a scram for the primary purpose of protecnng -

the reactor or to provide information that requires protecnve aedon to be initiated.

Safety Channel: A safety channelis a measuring channel in the reactor safety system. ' i Safety Limits: Safety limits are limits on important process variables that air found to be ^

necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivi;y.  :

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SE'ITINGS 2.1 Saferv Limit - Fuel Element Temocrature Acolicability: This specification applies to the tempemture of the mactor fuel.

Objective: The objective is to define the maximum fuel temperature that can be permitted with confidence that a fuel cladding failure will not occcr.

Specifications: -

(1) The maximum temperature in a Standan! TRIGA fuel rod shall not exceed 1000 C under 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.

Bases: The important parameter for a TRIGA reactor is the fuel rod temperature. This parameter is well-suited as a single specification, especially since it can be measured. A loss in the integrity of the fuel red cladding could arise from a buildup of excessive pressure between the fuel moderator and the cladding if the fuel temperature exceeds the -,

safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the disassociation of the hydrogen and zirconium in the fuel modemtor.

The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zimonium in the alloy. The safety limit for the TRIGA-FLIP fuel is based on data that indicate that the stress 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 1000'C and the fuel cladding is water cooled.*

2.2 Ljlpiting Safety System Settings Applicability: This specification applies to the settings that prevent the safety limit from I being reached.

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Objective: The objective is to prevent the safety limits from being reached.'

Soecifications: The limiting safety system settings shall be 500 C as measured in an instrumented fuel rod located in the central region of the core. For a mixed core, the instrumented rod shall be located in the mgion of the core containing the FLIP-type fuel rods.

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Bases: The limiting safety system setting is the measured instmmented fuel rod temperature that, if exceeded, shall initiate a scram to pmvent the fuel temperature safety limit from being exceeded. Section 5.4 of the FLIP conversion safety analysis report for the Washington State University (WSU) TRIGA reactorindicated 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 200aC 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 dunng the pulse because of the relatively long response time of the temperature channel as compared witlt 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 pmvides an additional degree of safety in the pulse mode of operation to protect the fuel in the event of such conditions as sticking of the transient control element in the withdrawn position after a pulse.

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3.0 LIMITING CONDITIONS OF OPERATION 3.1 Steady-State Ooeration j Applicability: This specification applies to the energy generated in the reactor during steady-state operation.

Objective: The objective is to ensure that the fuel temperature safety limit will not be exceeded during steady-state operation. i Soecifications: The reactor power level shall not exceed 1.3 MW under any condition of I operation. '

Basis Thermal 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  ;

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3.2 Reactivity Limitations Applicability: 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 i operation. l Obiective: 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 exceeded. I i

Soecifications: The reactor shall not be operated unless the shutdown margin provided by control elements shall be 0.25$ or greater with:

(1) the highest wonh nonsecured experiment in its most reactive state  !

(2) the highest worth control element and the regulating element (if not scrammable) I fully withdrawn  !

(3) the reactor in the cold critical condition without xenon Basis The value of the shutdown margin ensures that the reactor can be shut down from any operating condition even if the highest wonh rod should remain in the fully withdrawn ,

position. If the regulating rod is not scrammable,its worth is not used in determining the i shutdown reactivity.

3.3 Pulse Mode Operation Apnlicability: This specification applies to the peak fuel temperature in the reactor as a result of a pulse insertion of reactivity.

Obiective: 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 maximum safe allowable reactivity insertion shall be calculated annually for an existing core and prior to pulsing a new or modified core arrangement.

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Basis: 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 zimonium hydride which has a high crep strength and undergoes no phase changes at temperatures over 1000 C.

j However, after extensive steady-state operation at 1 MW, the hydrogen will ndistribute

due to migration from the central high temperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, the instantaneous temperature distribution is such that the highest values occur at the surface of the element and the lowest values occur at the center.

The higher temperatures in the outer regions occur in fuel with a hydrogen to zimonium ratio that has now substantially inemased above the nominal value. This produces hydrogen gas pressures considerably in excess of that expected for ZrH t.6. If the pulse insenion is such that the temperature of the fuel exceeds 874*C, then the pressure will be sufficient to cause expansion of microscopic holes in the fuel that grow larger with each j pulse. The expansion of the fuel stresses and distorts the fuel rod material which,in turn, l can cause overall swelling and distortion of the cladding and entire fuel rod. The pulsing-l limit of 830'C is obtained by examining the equilibrium hydrogen pressure of zirconium hydride as a function of temperatme. The decrease in temperature from 874 C to 830 C reduces hydrogen pressure by a factor of two, which provides an acceptable safety factor.

This phenomenon does not alter the steady-state safety limit since the total hydrogen in a fuel element does not change. Thus, the pressure exerted on the clad will not be significantly affected by the distribution of hydrogen within the element.

3.4 Maximum Excess Reactivity l Applicability: This specification applies to the maximum excess mactivity, above cold critical, which may be loaded into the reactor core at any time.

Objective: The objective is to ensure that the core analyzed in the safety analysis repon approximates the operational core within reasonable limits.

Soecifications: 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 -

mactor can be shut down, that specification does not address the total reactivity available -

within the core. This specification, although over-constraining the reactor system, helps ensure that the licensee's operational power densities, fuel temperatures, and temperature peaks are maintained within the evaluated safety limits. The specified excess reactivity allows for power coefficients of reactivity, xenon poisoning, most experiments, and l operational flexibility.

3.5 Core Conficuration Limitation Apolicability: 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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Specifications:

(1) The FLIP-fueled region in a mixed core shall contain at least 22 FLIP fuel rods in a contiguous block of fuel in the central region of the reactor core. Water holes in the FLIP region shall be limited to nonadjacent single-rod holes.

(2) The IYIR as defined in Section 1.4 and as calculated by the method used in the FLIP conversion safety analysis repon shall not exceed 1.5 for an operational core.

Bases The limitation on the allowable core conDguration as set forth in Section 4.1 of the FLIP fuel conversion safety analysis report limits power peaking effects. The limitation on i power peaking effects ensures that the fuel temperature safety limit will not be exceeded in a mixed core.

A 500oC safety system setting and a 1.5 PTR limit the maximum possible steady-state fuel temperature in the FLIP region to less than 800oC.

3.6 Control and Safety System 3.6.1 Scram Time Applicability: This specification applies to the time required for the scrammable control rods to be fully inserted from the instant that a safety channe.1 variable reaches the safety system settmg.

Objective: The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

Soecifications: 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. >

l Basis This specification ensures that the reactor will be promptly shut down when a scram signal is initiated. 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.

1 3.6.2. Reactor Control System Applicability: This specification applies to the inforrantion that must be available to the reactor operator dunng reactor operation.

Obiective

The objective is to require that sufficient information is available to the operator l to ensure safe operation of the reactor.

l l Specifications: The reactor shall not be operated 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 mode Measuring channel operable SS nulse Fuel element temperature 1 X X Linear powerlevel 1 X Log power leve1 1 X Integrated pulse power 1 X Note: SS = stead)-state Bases Fuel temperature displayed at the control console gives continuous information on this parameter, which has a specified safety limit. The power level monitors ensure that the -

reactor power level is adequately monitored for both steady-state and pulsing modes of operation. The specifications on reactor power level indication are included in this section since the power level is related to the fuel teraperature.

3.6.3 Reactor Safety System Anolicability: This specification applies to the reactor safety system channels.

Obiective: The objective is to specify the minimum number of reactor safety system channels that must be operable for safe operation.

l Specifications: The reactor shall not be operated unless the safety channels described in

Table 3.2 are operable.

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l Table 3.2 Minimum Reactor Safety Channels i Number operable j in specified mode Safety Channel Function SS Pulse  ;

Fuel temperature Scram if fuel temperature exceeds 500 C 1 1 Power level Scram if powerlevel exceeds 1100 kW 1 Manual scram Manually initiated scram 1 1 Wide range Prevent initiation of a pulse above 1 kW l Prevent control element withdrawal when neutron count is less than 2 cps 1 High-voltage monitor Scram on loss of high voltage to power channels 1 1 Pulse-mode switch Prevent withdrawal of standard control and regulation elements in pulse mode 1 Preset timer Transient rod scram 15 seconds or less after pulse 1 Poollevel Alarm if pool level falls below 16 ft over the core 1 1 Transient rod control Prevent application of air unless fully inserted 1 Note: SS = steady-state Bases The fuel temperature and power 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 operator to shut down the system if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the safety chambers, operation of the reactor without adequate instrumentation is prevented. The  !

preset timer ensures that the reactor power level will reduce to a low level after pulsing.

The interlock to prevent startup of the reactor with less than 2 cps ensures that sufficient neutrons are available for proper startup.

The interlock to prevent the initiation of a pulse above 1 kW is to ensure that the magnitude ,

of the 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 mode is to prevent the reactor from being pulsed while on a positive period. The pool level ,

alarm is intended to alert the operator to any significant decrease in the pool level. '

3.7 Radiation Monitorine System Applicability: This specification applies to the radiation monitoring information which must be available to the reactor operator during reactor operation.  !

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Obiective: The objective is to ensure that sufficient radiation monitoring is available to the operator to ensure safe operation of the reactor.

Soecifications: The reactor shall not be operated unless the radiation monitoring channels listed in Table 3.3 am operable. Each channel shall have a readout in the control room and be capable of sounding an audible alarm that can be heard in the reactor control room.

Luir The radiation monitors inform operating personnel about any impending or existing danger from radiation so that there will be sufficient tune 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 radiadon monitor Monitor radiation level on the bridge 1 Area radiation monitor Monitor rediation 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 maintenance to the radiation monitoring channels, the intent of this specification will be satisfied if they are replaced with portable gamma-sensitive instru-ments having their own alarms or that shall be kept under visual observation.

3.8 Argon-41 Discharge Limit Applicability: This specification applies to the concentration of 41Ar that may be discharged from the WSU TRIGA reactor facility.

Objective: To ensure that the health and safety of the public are not endangered by the discharge of 41Ar from the WSU TRIGA reactor facility.

Specification: The concentration of 41Ar in the effluent gas from the facility, as diluted by atmospheric air in the lee of the facility as a result of the turbulent wake effect, shall not exceed 1 x 10-8 pCi/ml averaged over one year. l Basis The maximum allowable concentration of 41Ar in air in unrestricted areas as specified in Appendix B, Table II of 10 CFR 20 is 1 x 10-8 pCi/ml. Section 6.5 of the l-safety analysis report for conversion of the WSU TRIGA reactor to FLIP fuel substantiates a 3.4 x 10-3 atmospheric dilution factor for a 4.4 mph windspeed. A somewhat more conservative value of 4 x 10-3 has been selected for the calculation of 41Ar dilution.

3.9 Engineered Safety Feature - Ventilation System Apolicability: Tnis 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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l 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 permit repair or testing of the l ventilation system. In the event of a substantial release of airborne radioactivity within the l facility, the ventilation system will be secured or operated in ce dilution mode to prevent  !

the release of a significant quantity of airborne radioactivity from the facility.

I his.tr During normal operation of the reactor and the ventilation system, the concentration of 41Ar and other airborne mdionuclides discharged from the facility is below the applicable ,

derived air concentration (DAC). In the event of a substantial release of airbome l l radioactivity within the facility, the ventilation system will be secured or operated in a dilution mode as appropriate. This action will permit minimizing the concentration of i airborne radioactive materials discharged to the environment until it is within the appropri-ate DAC value. In addition, operation of the reactor with the ventilation system shut down l for shon periods of time to make system repairs or tests does not compromise the control over the release of airbome 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 secumd.

3.10 Limitations on Exneriments Aonlicability: This specification applies to experiments installed in the reactor and its 1 experimental facilities (defined in Section 1.2). I Obiective: The objective is to pmvent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications: The reactor shall not be operated unless the following conditions governing  ;

experiments exist.  ;

(1) Nonsecured experiments shall have reactivity worths less than 1.00$.

(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 i quantities greater than 25 mg shall not be irradiated in the mactor or experimental  !

facilities. Explosive materials in quantities less than 25 mg may be irradiated in the l reactor or experimental facilities, provided the pressure produced upon detonation  !

l of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

1 (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 airbome concentration of radioactivity averaged over a year j would not exceed the applicable limits of Appendix B of 10 CFR 20.

In calculations pursuant to item 5 above, the following assumptions shall be used: l l

14 Amendment No.14

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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 l

aerosols produced will escape.

• If the effluent from an experimental facility exhausts through a filter installation l designed for greater than 99% efficiency for 0.3 p particles, at least 10% of these panicles can escape.

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• For materials whose boiling point is above 60 C and in cases where vapors formed l by boiling this material can escape only through an undisturbed column of water j above the cote, at least 10% of these vapors can escape.

• An atmospheric dilution factor of 4 x 10-3 for gaseous discharges from the facility, i (6) Each fueled experiment shall be controlled so that the total inventory ofiodine iso-I 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 reacter fuel or structure l by corrosion or other means, that material shall be removed and physically l inspected to determine the consequences and need for corrective action. The results 1

of the inspection and any cormctive 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.

l Bases:

i (1) This specification is intended to provide assurance that the worth 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 insened. -

(2) The maximum worth 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 reactor protective systems would act to 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 l 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 airbome 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 fueled experiment leading to total release of the iodine, the exposure 15 Amendment No.14

dose at the exclusion area boundary will be less than that allowed by 10 CFR 20 for .j an unrestricted area.  !

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(7) Operation of the reactor with the reactor fuel or structure damaged is prohibited (to avoid release of fission products). j i

3.11 Limitations on Irradiations  !

Applicability: This specification applies to irradiations performed in the irmdiation facilities  !

contained in the reactor pool as defined in Section 1.2, " Irradiation Facilities." Irradiations l are a subclass of experiments that falls within the specifications hereinafter stated in this l section. The surveillance requirements for irradiations are given in Section 4.3.5(2).

Obiective: The objective is to prevent damage to the reactor, excessive release of radio-active materials, or excessive personnel radiation exposure during the performance of an irradiation.

Specifications: 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 1 ft; otherwise, it shall be classed as an experiment. 1 (3) The device or material is encapsulated in a suitable container.

(4) The reactivity worth of the device or material is 0.25$ or less; otherwise, it shall be classed as an experiment.

l (5) The device or material does not remain in the reactor for more than a 15-day period; l otherwise, it shall be classed as an experiment, j l Basis This specification is intended to provide assurance that the special class of experi- ,

l ments called irradiations will be performed in a manner that will not permit any safety limit j to be exceeded.

l 3.12 As Low As Reasonably Achievable (ALARA) Radioactive Effluent Releases Aoplicability: This specification applies to the measures required to ensure that the radio-active effluents released from the facility are in accordance with ALARA criteria.

Obiective: The objective is to limit the annual population radiation exposure owing to the operation of the WSU TRIGA reactor to a small percentage of the normal local background exposure.

Snecifications:

(1) In addition to the radiation monitoring specified in Section 5.4, an environmental -

radiation monitoring program shall be conducted to measure the integrated radiation exposure in and around the environs of the facility on a quarterly basis.

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16 Amendment No.14 l

i (2) The annual radiation exposure due to reactor operation, at the closest off-site point i of extended occupancy, shall not, on an annual basis, exceed the average local off-site background radiation by more than 20E (3) Whenever practicable, the reactor shall be operated 4 in. or more from the thermal ,

column in onier to minimize the production of 41Ar.

(4) The total annual discharge of SAr into the environment shall not exceed 20 Ci per year.

l (5) In the event of a significant fission product leak from a fuel rod or a significant I 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 the leak is located and eliminated. However, the reactor may continue to be operated on a l short-term basis as needed to assist in determining the source of the leakage.

(6) Before discharge, the facility liquid effluents collected in the holdup tanks shall be analyzed for their beta-gamma activity content. The total annual quantity ofliquid effluents released (above background) shall not exceed 1 Ci per year.

, Basis The simplest and most reliable method of ensuring that ALARA mlease limits are l accomplishing their objective of minimal facility caused radiation exposure to the general ,

l public is to actually measure the integrated radiation exposure in the environment on and off {

the site. ,

3.13 Primanr 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 minimize neutron activation of dissolved materials.~

Snecifications:

(1) Conductivity of the pool water shall be no higher than 5 x 10-6 mhos/cm. )

(2) The pH of the pool water shall be between 5.0 and 7.5.

Basis: 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 l cleanup system is requimd. Experience with water quality control at many reactor facilities  !

has shown that maintenance within the specified limits provides acceptable control.

l 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 will decrease personnel exposures during maintenance and operations.

3.14 Sealed Sources in the Reactor Pool Applicability: This specification applies to any and all sealed sources stored or used in the reactor pool.

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Objective: The objectives of this mquirement are to ensure that: 1) any sealed source or sources that are stored or used in the pool do not constitute any type of significant hazard to the operation of the reactor,2) any such sealed source or soumes do not create a significant environmental or personal radiation exposum hazard, and 3) any such sealed soume or sources do not compromise the ALARA criteria of the facility.

Snecifications:

(1) Scaled sources shall not at any time be stomd or used closer than five (5) feet away from the face of an operating reactor core. The total activity of all sealed sources 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 precipitate a sealed '.

source incapsulation integrity problem and the soumes shall be stored in an appropriate shield so as not to produce a significant radiation hazani in the event of a loss of reactor pool water accident.

(2) All storage of scaled soumesgrreater 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 sealed sources in the mactor pool shall be in effect.

(3) The radionuclide content of the reactor pool water shall be monitored monthly at an interval 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 water for radionuclides from a sealed source stored in the reactor pool exceeds one-third (1/3) the 10 CFR 20 Appendix B, Table 3 value, steps shall be taken to isolate l the soume of the activity and to mitigate the problem.

Basis (1) Limiting the proximity of sealed souxes to five (5) or mom feet away fmm 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 neutmn 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 shielding. Thepresence of the sources in the pool would have no impact upon the D.B.A. which is the rupture of the cladding on one fuel element. However, the presence of sources in the pool could contribute to the radiation hazani associated with a loss of pool water accident. The dose mte 25 feet above an unshielded core in the event of a loss of pool water accident would only l 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 . 3 mandates that such storage be reviewed by the Reactor Safeguards Committee. )

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(3). ' The 10 CFR 20 Appendix B, Table 31imit for"Co is 3 x 105 pCi/mt. At this limit 'I the entim pool could be dumped into the WSU sewage system without taking i advantage of the dilution factor associated with the discharge volume of the WSU j sewage system. The detection limit for *Co in the reactor pool water depends upon . '

the system used but in the worst case would be at least 1 x 10-7 pCi/mt, or, i 100 pCi/t, or about one-thiree-hundredth of the 10 CFR 201imit. stated above. ~

Setting a limit of 100 times the detection limit and one-third the discharge limits - j provides the facility with ample time to take conective action in the event the limit is ; i exceeded and does not compromise ALARA considerations. j i

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i 4.0 SURVEILLANCE REQUIREMENTS f 4.1 General Applicability: This specification applies to the surveillance requimments of any system related to reactor safety. ,

Obiective: The objective is to verify the proper operation of any system related to reactor safety.

Soecifications: Any addidons, modifications, or maintenance to the ventilation system, the core and its associated support structum, the pool or its penetrations, the pool coolant sys-tem, the control element drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems wem originally designed and fabricated or to specifications approved by the Reactor Safeguards Committee. A system shall not be considered operable 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 continue to meet

the original design specifications, it can be assumed that they meet the presently accepted operating cntena.

4.2 Saferv Limit - Fuel Element Temocrature 1  !

Apolicability: This specification applies to the surveillance requirements of the fuel element i temperature measuring channel.

Obiective: The objective is to ensure that the fuel element temperatures are properly monitored.

Specifications:

(1) Whenever a reactor scram caused by high fuel element temperature occurs, the peak indicated fuel temperature shall be examined to determine whether the fuel element temperature safety limit was exceeded.

i (2) The fuel element temperature measuring channel shall be calibrated semiannually or at an inteival not to exceed 8 months by the substitution of a thermocouple simula-tor in place of the instrumented fuel element thermocouple. {

(3) A channel check of the fuel element measuring channel shall be made each time the  !

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reactor is operated by comparing the indicated instrumented fuel element tempera-ture with previous values for the core configuration and power level.

l Basis. Operational experience over the past 5 years with the TRIGA system gives assur-

j. ance that the thermocouple measurements of fuel element temperature have been sufficiently l reliable to ensure accurate indication of this parameter.

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20 Amendment No.14 l

4.3 Limiting Conditions for Oneration 4.3.1 Reactivity Requirements Applicability: These specifications apply to the surveillance requirements for reactivity control of experiments and systems.

Obiective: The objective is to measure and verify the worth, performance, and operability .

of those systems affecting the reactivity of the reactor.

Snecifications:

(1) The reactivity worth of each control rod and the shutdown margin shall bc deter-mined annually but at intervals not to exceed 15 months, (2) The reactivity worth of an experiment 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 be inspected, cleaned, and lubricated as necessary semiannually at intervals not to exceed 7.5 months.

(5) The reactor shall be pulsed semiannually to compare fuel temperature measurements and peak power levels with those of 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 determining the reactivity l

worths of experiments inserted in the core. Past experience with TRIGA reactors gives assurance that measurement of the reactivity worth on an annual basis is adequate to ensure -

no significant changes in the shutdown margin. The visual inspection 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 comparisor is made with previous similar pulses to determine if changes in fuel or core charac' eristics are taking place.

4.3.2 Control and Safety System l Annlicability: These specifications apply to the surveillance requirements for measure-l ments, tests, and calibrations of the control and safety systems.

Obiective: The objective is to verify the performance and operability of those systems and components which are directly related to reactor safety.

Specifications:

(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 performed before each day's operation or before each opera-tion extending more than 1 day, except for the pool level channel which shall be tested monthly.

21 Amendment No.14

(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 Table 3.2, other than measuring channels, shall be performed semiannually, but at intervals not to exceed 7.5 months.

Enty 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 pedorm pmperly. 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 Apolicability: This specification applies to the surveillance monitoring for the area monitoring equipment, Argon-41 monitoring system, and continuous air monitoring system.

Obiectives: The objectives are to ensum that the radiation monitoring equipment is operating properly and capable of pedorming its intended function, and that the alarm points are set cormetly.

Speci6 cations: All radiation monitoring systems shall be verified to be operable at least monthly at an interval not to exceed 45 days. In addition, the following surveillance activities shall be performed on an annual basis at intervals not to exceed 15 months: 1) the l area radiation monitoring system shall be calibrated using a certified source; 2) a calibration i

of the Ar-41 system shall be done using at least two different calibrated gamma-ray l sources; 3) a calibration shall be performed on the CAM in terms of counts per unit time per unit of activity using calibrated beta sources.

Basir 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 timespan.

4.3.4 Ventilation System Applicability: This specification applies to surveillance requirements for the pool room ventilation system.

Objective: The objective is to ensure the proper operadon of the pool room ventilation system in the isolation and dilution modes, which would be used in controlling the release of radioactive material to the uncontrolled environment in the event of an emergency.

Snecifications: 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 performance and selected mode of operation. 'lle pressure drop across the absolute filter in the pool ventilation system shall be measured at least twice a year. The absolute filter shall be changed at least every 2 years and whenever the pressure drop across the filter increases by 1 in. of water.

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Basis Experience has shown that the only reliable method of testing the ventilation is to -

cycle the system into the various modes and visually check each portion of the system for proper operation in that mode.

4.3.5 Experiment and Irradiation Limits Applicability: This specification applies to the surveifance requirements for experiments installed in the reactor and its experimental facilities and for irradiations performed in the inadiation facilities.

Specifications:

(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 modifications to a reviewed and approved experiment may be made at the discretion of the senior operator responsible for the operation, provided that the hazards associated with the modifications have been icviewed and a determination has been made and documented that the modifications do not create a significantly 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 performed and reviewed for compliance with

" Limitations on Irradiations," Section 3.11, by a licensed senior operator qualified in health physics, or a licensed senior operator and a person qualified 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 l

conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.

4.4 Reactor Fuel Elements Apolicabilily: This specification applies to the surveillance requirements for the fuel elements.

Obiective: The objective is to verify the continuing integrity of the fuel element cladding.

Specifications: All fuel elements shall be inspected visually for damage or deterioration and measured for length and bend et 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 l (1) in measuring the transverse bend,its sagitta exceeds 0.125 in. over the length of the cladding

(2) in measuring the elongation, its length exceeds its original length by 0.125 in.

(3) a clad defect exists as indicated by release of fission products i

23 Amendment No.14

Emil: The frequency of inspection and measurement schedule is based on the parameters mort likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levels 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 of heat fmm 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 msult in touching has occurred without deleterious effects. The elongation limit has been specified to ensum 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 1

Anplicability: This specification applies to the surveillance of primary water quality.

Objective: The objective is to ensure that water quality does not deteriorate over extended periods of time if the reactor is not operated.

Specification: The conductivity and pH of the primary coolant water shall be measured at least once every 2 weeks, and shall be as follows:

(1) conductivity s 5 x 104 mhos/cm (2) pIl between 5.0 and 7.5 Basis: Section 3.3 ensures that the water quality is adequate during reactor opemtion.  !

Section 4.5 ensures that water quality is not permitted to deteriorate over extended periods of time even if;he reactor does not operate.

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24 Aruendment No.14

5.0 DESIGN FEATURES ,!

5.1 Reactor Fuel l Anplicability: This specification applies to the fuel elements used in the reactor com.1 1

, . 1 Obiective: The objective is to ensure that the fuel elements are of sudh a design andl . j fabricated in such a manner as to permit their use with a high degree of mliability with .,

respect to their physical and nuclear characteristics.:

Specificanons: -

(1) TRIGA-FLIP Fuel - De individual unirradiated FLIP fuel elements shall have thel 1 'I followingcharacteristics: , ,

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• uranium content: maximum of 9 wt% enriched to nomirial 70% _235UA 4 0 e hydrogen-to-zirconium ratio Gn the ZrHx): between 1.5 and 1.7 j sy.

e natural erbium content (homogercously' distributed)f between 1.1 and l.6 wt%. a N l e cladding: 304 stainless steel, nominal d.020 in. thick 1 -

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• identification: top pieces of FLIP elements will have characteristic markings :.  ;

to allow visual identification of FLIP elements employbd in mixed cores .: ,

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. . 1 (2) Standard TRIGA Fuel - The individual unirradiatixi Standard TRIGA fuel elements 1. y shall have the following characteristics:: "

q e uranium content: maximum of 9.0 wt% enriched to less than 20%'235(y j

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e hydmgen-to-zirconium atom ratio (in the ZrHx): between 1.5 ahd 1.8)  :

e cladding: 304' stainless steel, nominal 0.020 in. ihick-l Hasis: A maximum uranium content of 9.25 wt% in TRIGA-FLIP clements is about 6%? .

'I gmater 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 wt%

in an element is about 30% less thr.n the design value. This variation would result in an inemase in power density of only about 6%.1An increase in locai power density of 6% ..

U reduces the safety margin b~ it most 10E The maximum hydrogen-to-zirconium ratio off 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 from a hydrogen-to-zirconium ratio of; 1.60. However, this increase in the clad stress during an accident would not exceed the -

rupture strength of the clad.

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A maximum uranium content of 9 wt% for the standard TRIGA elements is about 6%

greater than the design value of 8.5 wt%. Such an increase in loading would result in an -

l increase in power density of 6% and reduces the safety margin by at most 10%. The

! maximum hydrogen-to-zirconium ratio of 1.8 could result in a maximum stress under l accident conditions in the fuel element clad about a factor of 2 greater than the value msulting from a hydrogen-to-zirconium ratio of 1.60. However, this inemase in the clad t stmss during an accident would not exceed the rupture strength of the clad.

5.2 Reactor Com Anolicability: This specification applies to the configuration of fuel and in-com i

experiments.

Obiective: The objective is to ensure that provisions are made to restrict the anangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.

Sprifications:

, (1) The core sal' be an arrangement of TRIGA uranium-zirconium-hydride fuel-moderator bedles positioned in the reactor grid plate.

l (2) The TRIGA com 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 'uel mds located in a contiguous block in the central region of the CorC.

(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 l be limited to single-rod holes. Vacant lattice positions in the core fuel region shall e occupied with fixtums that will prevent the installation of a fuel bundle. j (4) The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite, aluminum and water. ,

Basis Standard TRIGA cores have been used for years and their characteristics are well- I 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 measumd  !

performance of mixed coms in the WSU reactor, the Texas A&M reactor, and the University of Wisconsin reactor, have shown that such cores may be safely operated.

In mixed cores, it is necessary to arrange FLIP elements in a contiguous, central region of the core to control flux peaking and power generation peak values in individual elements.

Vacant core lattice positions in the Standard fuel mgion will contain exs 'ments or an experimental facility to prevent accidental fuel additions to the reactor c m. Vacant core positions are not permitted in the FLIP fuel region as specified by Section 3.5.

The core will be assembled in the mactor 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.

26 Amendment No.14

4 5.3 Control Elements ,

Annlicability: 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

, permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

( Soecifications:

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! (1) The standard control element shall have scram capability and contain borated l graphite, B.tC powder, or boron and its compounds in solid form as a poison in 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 element.s.

1 (3) The transient control element shall have scram capability and contain borated graphite or boron and its compounds in a solid form as a poison in an aluminum or

, stainless steel clad. The transient element shall have an adjustable upper limit to l

allow a variation of reactivity insertions. This element may incorporate a nonfueled follower.

t Basis The poison requirements for the control elements are satisfied by using neutron-absorbing borated graphite, B4 C powder, or boron and its compounds. Since the regulat-ing element normally is a low wonh element, its function could be satisfied by using solid stainless steel. These materials must be contained in a suitable clad material, such as alumi-num or stainless steel, to ensua mechanical stability during movement and to isolate the poison from the pool water environment. Scram capabilities are provided for rapid inser-tion of the control element which is the primary safety feature of the reactor. The 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

~

Anolicability: This specification describes the functions and essential components of the l area radiation monitoring equipment and the system for cor,tinuously monitoring airborne )

radioactivity.

Obiective: The objective is to describe the radiation monitoring equipment that is available to the operator to ensure safe operation of the reactor.

Specifications:

(1) Function of Area Radiation Monitor (gamma-sensitiveinstruments): Monitor  !

radiation fields in key locations, alarm and readout at control console.

, (2) Function of Continuous Air Radiation Monitor (beta , gamma-sensitive detector with particulate collection capability): Monitor radioactive paniculate activity in the pool room air, alarm and readout at control console.

27 Amendment No.14

i (3) Function of Arcon-41 Stack Monitor (gamma-sensitive detector): Monitor 41Ar content in reactor exhaust air, alarm and readout at console.

Basis The radiation monitoring system is intended to provide information to operating personnel of any impending or exisdng danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spmad of radioacti-vity to the surroundings.

5.5 Fuel Storace Apolicability: This specification applies to the stomge of reactor fuel at times when it is not m the reactor core.

Objective: The objective is to ensure that fuel that is being stored will not become critical and will not reach an unsafe temperature.

Snecifications:

(1) All fuel elements shall be stored in a geometrical array where the keg is less than 0.8 for all conditions of moderation.

(2) hradiated fuel elements and fueled devices shall be stored in an array, which will pemiit 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 Systen)

Apolicability: This specification applies to the building that houses the reactor.

Objective: The objective is to ensure that provisions are made to restrict the amount of radioactivity released into the environment.

Specifications:

(1) The reactor shall be housed in a facility designed to restrict leakage. The minimum free volume in the facility shall be 109 cm3.

(2) The reactor building shall be equipped with a ventilation system designed to filter and exhaust air or other gases from the reactor building and release them from a stack at u ninimum 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 substantial release of airborne 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 mixed and diluted with outside air before being discharged from the facility.

Basis: The facility is designed so that the ventilation system will normally maintain a negative pressure with respect to the atmosphere to minimize uncontrollable leakage to the 28 Amendment No.14 t

environment. The free air volume v2hin the reactor building is confined when there is an emergency shutdown of the ventilation system. Emergency contmls for startup, isolation, dilution, and normal operation of the ventilation system are located extemal to the control and pool rooms. Proper handling of airborne radioactive materials (in emergency .

situations) can be effected with a minimum of exposure to operating personnel.

5.7 Reactor Pool Water Systems Applicability: This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.

Objective: The objective is to ensure that coolant water shall be available to pmvide ade-quate cooling of the reactor core and adequate radiation shielding.

Soecifications:

(1) The reactor core shall be cooled by natural convection water flow.

(2) All piping extending more than 5 ft below the surface of the pool shall have ade-quate provisions to prevent inadvenent siphoning 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 normal level.

(4) The reactor shall not be operated with less than 15 ft of water above the top of the -

Core.

Basis This specification is based on thermal and hydraulic calculations which show that the TRIGA-FLIP core can operate in a safe manner at power levels up to 2700 kW with natural convection flow of the coolant water. A comparison between operation of the TRIGA-FLIP and standard TRIGA MARK III has shown them to be safe for the above .

power level. Thermal and hydraulic characteristics of mixed 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 the top of the pool.

Loss of coolant alarm after 2 ft of loss requires corrective action. This alarm 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 all provisions of the NRC staff-approved physical 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 TRIGA Reactor Security Plan."

l 29 Amendment No.14 i

6.0 ADMINISTRATIVE CONTROL  ;

6.1 Resoonsibility The facility shall be under the direct control of a licensed Senior Reactor Operator (SRO) '

designated by the Director of the WSU Nuclear Radiation Center. The SRO shall be responsible to the Director for the overall facility operation including the safe opemtion and maintenance of the facility and associated equipment. The SRO shall also be msponsible for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, Federal and State regulations, and requirements of the Reactor Safeguards Committee.

6.2 Organization (1) The reactor facility shall be an integral part 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 ano authority of each member of the operating staff shall be defined in writing.

(2) When the reactor is not secured, the minimum staff shall consist of:

(a) Reactor Operator (RO) at the controls (may be the SRO)

(b) Senior Reactor Opentor (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 Trainine The licensed Senior Reactor Operator designated by the Director as being msponsible for the facility also shall be responsible for the facility's Requalification Training Program and Operator Training Program.

6.5 Reactor Safeguards Committee (RSC) 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 Amendment No.14

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. -i 0- j l

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-J President of WSU. )

i i

' Dean of .

- the Graduate School: .,

l I

m - Reactor Safeguards .

Nuclear Radiation Centerj Committee I-Review Audit -

' Function . Functior Radiation Safety Facility Director. - -l .---~'~ j -l I

commmee j

Reactor Supervisor :

and Operating Staff.

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Figure 6.1 Facility organization 31 Amendment No. '14 -

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6.5.2 Composition and Qualifications l  ;

l The RSC shall be composed of at least five members knowledgeable in fields that relate to nuclear reactor safety. The members of the Committee shall include one facility Senior Reactor Operator and WSU faculty and staff members designated to serve on the Commit-l tee in accordance with the procedures specified by the WSU conunittee manual. The Uni- 1 versity's Radiation Safety Director shall be an ex officio member of the Committee. l 6.5.3 Operation ,

The Reactor Safeguards Committee shall operate in accordance with a written charter, .

including provisions for: ,

! (1) meeting frequency: the full committee shall meet at least semiannually and a subcommittee themof shall meet at least semiannually l

l (2) voting rules l (3) quorums: chairman or his designate and two members I (4) method of submission and content of presentations to the committee l

I (5, use of subcommittees 1 (6) review, approval and dissemination of minutes 6.5.4 Reviews The responsibilities of the RSC or designated subcommittee thereof shall include, but are notlimited to, the following:

(1) review and approval of all new experiments utilizing the reactor facility (2) review and approval of all proposed changes to the facility license by amendment, and to the Technical Specifications (3) review of the operation and operational mcords of the facility (4) review of significant operating abnormalities or deviations from nomul and expected performance of facility equipment that affect nuclear safety (5) review and approval of all determinations of whether a proposed change, test, or- ,

experiment would constitute a change in the Technical Specifications or an unreviewed safety question as defined by 10 CFR 50 (6) review of reportable occurrences and the reports 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 (9) annual review of the radiation protection program 'l 32 Amendment No.14

c 6.6.5 Audits The RSC or a subcommittee thereof shall audit reactor operations semiannually, but at intervals not to exceed 8 months. The semiannual audit shall include at least the following:

(1) review of the reactor operating records (2) inspection of the reactor operating areas (3) review of unusual or abnormal occurrences (4) radiation exposures at the facility and adjacent environs 6.5.6 Reconis 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 to the Dean of the Graduate School and Facility Director.

(2) A written report of all reviews perfomied 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 secretary of the RSC shall maintain a file of the minutes of all meetings.

6.6 Ouality Assurance In accordance with Regulatory Guide 2.5 and ANSI 402, " Quality Assurance Program Requirements for Research Reactors," Section 2.17, the " facility shall not be required to prepare quality assurance documentation for the as-built facility." Quality Assurance ;QA) requirements will still be limited to those specified in Section 2.17 as follows:

"All replacements, modification, and changes to systems having a safety related 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 perfonnance or reliability."

"The required audit function shall be performed by the RSC specified in Section 6.5."

]

t 6.7 Action To Be Taken in the Event a Safety Limit Is Exceeded  ;

In the event a safety limit is exceeded:  ;

(1) The reactor shall be shut down and reactor operation shall not be resumed until I authorized by the U.S. Nuclear Regulatory Commission (NRC).  !

(2) An immediate repon of the occurrence shall be made to the Chainnan of the Reactor Safeguards Committee, and repons shall be made to the NRC in accordance with Section 6.10 of these specifications.

(3) A report shall be prepared that shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be 33 Amendment No.14

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submitted to the Reactor Safeguards 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 Operating Procedures Written operating procedures shall be adequate to ensure the safety of 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:

l (1) perfomiing irradiations and experiments  !

(2) startup, operation, and shutdown of the reactor (3) emergency situations including provisions for building evacuation, earthquake,

, radiation emergencies, fire or explosion, personal injury, civil disorder, and bomb threat (4) core changes and fuel movement j

! (5) control element removal and replacement l

l (6) perfomiing preventive maintenance and calibration tests on the reactor and I associated equipment (7) power calibration l Substantiative changes to the above procedures shall be made only with the approval of the I licensed SRO directly in charge of the facility. Temporary changes to the procedures that i i

do not change their original intent 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. l 6.9 Facility Ooerating Records 1

In addition to the requirements of applicable regulations, and in no way substituting for those requirements, records and logs shall be prepared for at least the following items and retained for a period of at least 5 years for items (1) through (6) and indefinitely for items (7) through (11).

(1) nonnal reactor operation (2) principali nintenance activities (3) abnormal occurrences (4) equipment and component surveillance activities required by the Technical Specifications (5) experiments performed with the reactor (6) gaseous and liquid radioactive effluents released to the envi ons (7) off-site inventories and transfers (8) 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 Amendment No.14

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6.10 Reporting Requirements ,

In addition to the requirements of applicable regulations, and in no way substituting for those requirements, repons shall be 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, and to the NRC Operations Center, of (a) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the mlease msulted in property damage, personalinjury,or exposure; (b) Any violation of the safetylimit; (c) Any reportable occurrence as defined in Section 1.1, " Reportable Occurrence," of these specifications.

(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 miease or radioactivity above permissible limits in unrestricted areas whether or not the release resulted in propeny damage, personal injury, or exposure. The written repon (and, to the extent posr:ble, the preliminary telephone or telegraph report) shall describe,'

analyze, and evaluate safety implications, and outline the corrective measures taken or planned to pmvent recurrence of the event; (b) Any violation of a safetylimit; (c) Any reportable occurrence as defined in Section 1.1, " Reportable-Occurrence," of these specifications.

(3) A mport within 30 days in writing to the Dimetor, Office 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 measured values from a corresponding predicted or previously measured value of safety-connected operating  ;

characteristics occurring during operation of the reactor, '

(b) Any signif: cant change in the transient or accident analysis as described in the Safety Analysis Repon; I

(c) Any significant changes in facility organization; (d) Any observed inadequacies in the implementation of administrative or precedural controls.

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35 Amendment No.14

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(4) A mpon within 60 days after completion of startup testing of the reactor (in writing to the Director, Office of Nuclear Reactor Regulation, USNRC, Washington, D.C.

20555) upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measured values of the operating conditions including:

(a) An evaluation of facility pedormance to date in comparison with design predictions and specifications; l (b) A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.

(5) An annual report within 60 days following the 30th of June of each year (in writing to the Dimctor, Division of Licensing, USNRC, Washington, D.C. 20555) with a

. copy to the Regional Administrator, USNRC, Region V Office, providing the

! following information:

(a) A brief nanative summary of (i) operating experience (including experiments pedormed), (ii) changes in facility design, performance characteristics, and operating procedures related to mactor safety and occurring during the reporting period, and (iii) results of surveillance tests and inspections; (b) Tabulation of the energy output (in megawatt-days) of the reactor, hours reactor was critical, the cumulative total energy output since initial criticality, and number of pulses greater than 1.00$;

(c) The number of emergency shutdowns and inadvertent scrams, including reasons for them; (d) Discussion of the major maintenance operations performed during the -

period, including the effect,if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required; j (e) A brief description, including a summary of the safety evaluations of I changes in the facility or in procedures and of tests and experiments carried l out pursuant to 10 CFR 50.59; (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 before the point of such release or discharge:

36 Amendment No.14

W r

o Liquid Waste (summarized on a monthly basis)

(i) monthly radioactivity discharged

• total estimated quantity of radioactivity released (in curies),

e an estimation of the specific quantity for each detectable radionuclide in the monthly release,

• fraction of 10 CFR 20 table 3, appendix B limit for each detectable radionuclide taking into account the dilution factor from the total volume of sewage released by the licensee into the sewage system, e sum of the fractions for each radionuclide reported above.

(ii) total quantity of radioactive material released by the facility into the sewage system during the year period of the report Gaseous Waste (summarized on a monthly basis)

(i) radioactivity discharged during the reporting period (in curies)

• total estimated quantity of mdioactivity released (in curies) determined by an appropriate sampling and counting method, e total estimated quantity of 42Ar released (in curies) during the reporting period based on data from an appropriate monitoring system, e estimated average atmospheric diluted concentration of d2Ar released during the reporting period in terms of pCi/ml and fraction of the applicable DAC value, e total estimated quantity of mdioactivity in particulate form with half-lives greater than 8 days (in curies) released during the reporting period as determined by an appropriate particu-late monitoring system, e average concentration of radioactive particulates with half-lives greater than 8 days released in Ci/ml during the report- l ing period, and  :

37 Amendment No.14

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• an estinute of the average concentration of other significant radionuclides present in the gaseous waste discharge in terms of pCi/ml and fraction of the applicable DAC value for l the Irporting period if the estimated release is greater than 20% of the applicable DAC. l Solid Waste (summarized on an annual basis)

(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 exposum 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 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 highestlevels; (i) An annual summary of any environmental surveys performed outside the facility.

6.11 Written Communications All written communications with the Nuclear Regulatory Commission shall be made in accordance with the requirements of 10 CFR 50.4 " Written Communications."

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38 Amendment No.14