ML082270383

Oregon State University Submission of Relicensing Technical Specifications Revision J

ML082270383
Person / Time
Site:	Oregon State University
Issue date:	08/11/2008
From:	Reese S Oregon State University
To:	Alexander Adams NRC/NRR/ADRO/DPR
References
Download: ML082270383 (58)

Text

OSU Oregon State UNIVERSITY

• Radiation Center Oregon State University, 100 Radiation Center, Corvallis, Oregon 97331-5903 T 541-737-2341 I F 541-737-0480 1http://ne.oregonstate.edu/facilities/radiationcenter August 11, 2008 Mr. Alexander Adams U. S. Nuclear Regulatory Commission Research and Test Reactors Branch A Office of Nuclear Reactor Regulation Mail Stop 012-G13 One White Flint North 11545 Rockville Pike Rockville, MD 20852-2738

Reference:

Oregon State University TRIGA Reactor (OSTR)

Docket No. 50-243, License No. R-106 Letter Submitted on August 6, 2008, Oregon State University Submission of Relicensing Technical Specifications Revision I.

Subject:

Oregon State University Submission of Relicensing Technical Specifications Revision J.

Mr. Adams:

We would like to submit to you draft revision J of the OSTR Technical Specifications for your review. Attachment I contains the draft revision J. Attachment II contains a list of differences between draft revisions I and J. If you have any questions, please call me at the number above. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: to Sincerely, Steve Reese Director Enclosure cc: Document Control, NRC Al Adams, NRC Craig Bassett, NRC John Cassady, OSU Rich Holdren, OSU Todd Palmer, OSU Todd Keller, OSU

/t [)-o--(

Attachment I Oregon State University Technical Specifications Revision J (Draft)

CHAPTER 14 DRAFT TECHNICAL SPECIFICATIONS (The Technical Specifications are contained in USNRC Operating License R-106, Appendix A)

THIS PAGE INTENTIONALLY LEFT BLANK APPENDIX A TO FACILITY LICENSE NO. R-106 TECHNICAL SPECIFICATIONS AND BASIS FOR THE, OREGON STATE UNIVERSITY TRIGA8 REACTOR DOCK ET NO. 50-243 Revision O.J Current through Amendment #XX Date of Issuance: XXXX XX, XXXX

THIS PAGE INTENTIONALLY LEFT BLANK TABLE OF CONTENTS 1 D E F IN IT IO N S ................................................................................................................. 1

1. 1 Au d it ............................................................................................................................ 1 1.2 C han n el: ....................................................................................................................... 1 1.3 C hannel C alibration ................................................................................................. 1 1.4 C hannel C heck ........................................................................................................ 1 1.5 C hann el Test ......................................................................................................... 1 1.6 Confinem ent .................................................. . . . ........................... 1 1.7 Control Rod ....................................................................... ...................... .....

1.8 Core Lattice Position . . . . . . . . . . . 1 1.9 E xcess R eactivity: .................................................... .......... ........ ....................... 2 1.10 Exp erim ent ......................................................................... 2 2..........................

1.11 Experim ent Safety System s ... . . .... . ........................................ 2 1.12 Fuel E lem ent ............................................................

1.13 Instrum ented Elem ent ... . ...................... ........... 2........................

2 1.14 M easured V alue ............................................... ........................................................ 2 1.15 Irradiation F acilities ....................................... . . ................................................. 2 1.16 O p erab le ........................................................... . . ............................................... 2 1.17 O perating ............................... .................................. 3 1.18 Operational Core ...................................... 3 1.19 P u lse M ode ............................................... . ............................................................. 3 1.20 Radiation Center Com plex ............................ .... . . .............................................. 3 1.21 Reactor Operation .. ................ 3 1.22 R eactor Safety Syste.s.........................................

........................................ 3 1.23 Reactor Se ...................................................................... 3 1.24 Reactor Shutdown)Thei rýctor is shut down when: .................................. .3 1.5RfrneCr Condition. ............................ "......4...

1.25 Reference.Core 1.26 Review :?i¢.**......... C. dIt.0....... I ................................................................. 4 Z*:

1 al C.....................................

... . ....................................... ..... 4 1.28 Scram time: ................... ...... ........................ .............................................. 4 1.29 Should, Shall, andMay.......................................... 4 1.30 Shutd n M argin .... ................................... .......................................... 4 1.32 sqUare-Wav Margin ......~l. ....................................... 4 1.31 Sh utd ,n Reactivity ....... ...................................................................................... 4 1.32 Square-W a\,c Mode ................................................................................................... 4 1.33 Steady-State M ode ............................................................................................... 4 1.34 Substantive Changes ............................................................................................... 4 1.35 Surveillance Intervals .......................................................................................... 4 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING ........................ 8 2.1 Safety Limit-Fuel Element Temperature ............................................................... 8 2.2 Lim ited Safety System Setting ................................................................................. 9 3 LIMITING CONDITIONS OF OPERATION .......................................................... 10 3.1 R eactor Core Parameters ....................................................................................... 10 3.1.1 Steady-state O peration ................................................................................... 10 i

3.1.2 Shutdown M argin ............................................................................................ 10 3.1.3 Core Excess Reactivity ................................................................................... 11 3.1.4 Pulse M ode Operation ................................................................................... 11 3.1.5 Core Configuration Lim itations ...................................................................... 12 3.1.6 Fuel Param eters .............................................................................................. 12 3.2 Reactor Control And Safety System ..................................................................... 12 3.2.1 Control Rods ................................................................................................... 12 3.2.2 Reactor M easuring Channels ...................................................................... . 13 3.2.3 Reactor Safety System ..................................................... ...................... 14 3.3 Reactor Primary Tank W ater ........................................... ,,..................................... 18 3.4 This section intentionally left blank ......................... ... ................................ 18 3.5 Ventilation System .............................................. ............. . . ............................ 18 3.6 This section intentionally left blank .............. . .............................. 20 3.7 Radiation M onitoring System s and Effluents ........................................................

. 20 3.7.1 Radiation M onitoring System s ...... ............... ........................ ................... 20 3.7.2 Effluents .............. .................... 20 3.8 Lim itations on Experim ents ....................................................................................... 21 3.8.1 Reactivity Lim its .............. .............................................................................. 21 3.8.2 M aterials .............................. . ...................................................................... 22 3.8.3 Failures and M alfunctions . i...... ................ ................................................. 22 3.9 This section intentionally left blank. ........ ................ .................................... .23 4 SURVEILLANCE REQUIREM ENTS .............................................. 24 4.1 Reactor Core Paramieters ................. .............................................. 2.4.................

24 4.2 Reactor Control and Safety System s ...... ............................................................ 25 4.3 Reactor Primary Tank Waer ............................... ..... 26 4.4 This section inte:tionally lefftblank ....................................................................... 27 4.5 VentilationSystem ... ... ....... .................................................................... 27 4.6 Thiss iftntionally left blank ................................................................ 27 4.7 PR diation M onitorin g System ;.............................................................................. 27 4.8 Iperim ental Limits. ....... ..................................................................................... 28 4.9 -1 sNsection inteh *0ionallyi left blank ................................................................ 28 5 DESIGN EATURES ............................................................................................... 29 5.1 Site and Facility Description ................................................................................ 29 5.2 Reactor Coolant S-stem ....................................................................................... 29 5.3 Reactor Core and, Fuel .......................................................................................... 30 5.3.1 Reactor Core ................................................................................................. 30 5.3.2 Control Rods ................................................................................................. 31 5.3.3 Reactor Fuel ............................................................................................ 32 5.4 Fuel Storage ............................................................................................................... 33 6 ADM INISTRATIVE CONTROLS .......................................................................... 35 6.1 Organization ......................................................................................................... 35 6.1.1 Structure ................................................................................................... 35 6.1.2 Responsibility .......................................................................................... 35 ii

6.1.3 Staffi ng ..................................................................................................... 36 6.1.4 Selection and Training of Personnel ........................................................ 38 6.2 R eview A nd Audit ................................................................................................ 38 6.2.1 Composition and Qualifications .............................................................. 38 6.2.2 Charter and Rules ..................................................................................... 38 6.2.3 R eview Function.................................................................................... 38 6.2.4 A udit Function ......................................................................................... 39 6.3 R adiation Safety ................................................................ .............................. 39 6.4 Procedures .......................................................................... 40 6.5 Required A ctions ........................................................................................ 41 6.5.1 Actions to Be Taken in Case of Safety Limit Violation .*................................... 41 6.5.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.6.2 Other than a Safety Limit Violation .................. ................ 41 6 .6 Rep orts ........................................................ ......... .................. 442..........................

2 6.6.1 Annual Operating Report . . ............. ........................ ...................... 42 6.6.2 Special ........................................

Reports ............................... ........... 43 6 .7 Records ............................................. 445..........................

5 6.7.1 Records to be Retained-:for a Period of at Least Five Years or for the Life of the Component Involved if Less thaiinFive Years ................................................... 45 6.7.2 Records to be Retained for at I-east One Tram iiugCycle ........................ 45 6.7.3 Records to be Retained for the Lifetime of Ifthe Reactor Facility ............. 45 L.ist of Tables and Figure Table 1 - Minim um M eas ring 1u1,n nels ........... 14 14..........................

Table 2 -M inim um Reactor t %C aialJs .............................. .................................... 15 T able 3 - Mini mumfi ter1ocks .. .................................................................................... 15 Table 4 -<MinimumRladia~tion MNoniitoring Channels ..................................................... 20 Figure 1 - Administrative Structure...................................... 36 Included in this document are the Technical Specifications and the "Bases" for the Technical Specifications. These bases, which provide the technical support for the individual Technical Specifications, are included for informational purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

iii

This page intentionally left blank iv

1 DEFINITIONS

1.1 Audit

An audit is a quantitative examination of records, procedures or other documents after implementation from which appropriate recommendations are made.

1.2 Channel

A channel is the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.

1.3 Channel Calibration: A channel calibration is an adjustment of thechannel such that its output corresponds with acceptable accuracy to known values ofthe parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall include a Channel Test.

1.4 Channel Check: A channel check is a qualitative-verification of acceptable performance by observation of channel behavior. This verification, w*here possible, shall incl-de corfiparison of the channel with other independent channels or systemsmeasunc ng the same variable.

1.5 Channel Test: A channel test is the introduction of a siýnal into the channel for verification that it is operable. ,

1.6 Confinement

Confinement is an enclosure 'of the overall fatciity that is designed to limit the release of effluents between the enclosure and its externaenvironment through controlled or defined pathways.

1.7 Control Rod: A, contrl rod isa device fabrin ed from neutron absorbing material or fuel or both which is used to establish neutron flux changes and to compensate for routine reactivity changes. A control rod may be~coupled'.tQ its drive unit allowing it to perform a safety function when the coupl iAgs disengagedi.Types of control rods shall include:

a. Regulating Rod (Reg Ro"d). The regulating rod is a control rod having an electric motor drive and scram capabilities. It may have a fueled-follower section.

Its position may be varied manually or by the servo-controller.

b. Shim/Safet* Rod: A shim safety rod is a control rod having an electric motor drive and sram capabilities. It may have a fueled-follower section.

c. Transient Rod: The transient rod is a control rod with scram capabilities that can be rapidly ejected from the reactor core to produce a pulse. It may have a voided-follower.

1.8 Core Lattice Position: The core lattice position is defined by a particular hole in the top grid plate of the core. It is specified by a letter indicating the specific ring in the grid plate and a number indicating a particular position within that ring.

I

1.9 Excess Reactivity: Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff=l) at reference core conditions.

1.10 Experiment: Any operation, hardware, or target (excluding devices such as detectors or foils) which is designed to investigate non-routine reactor characteristics or which is intended for irradiation within an irradiation facility. Hardware rigidly secured to a core or shield structure so as to be a part of their design to carry out experiments is not normally considered an experiment.

Specific experiments shall include:

a. Secured Experiment: A secured experiment is any experiment or component of an experiment that is held in a stationary JposItion relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneuinatcic buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions. `

b. Unsecured Experiment: An unsecured experinent is any experiment or component of an experiment that does not meet the definition of a secured experiment.

C. Movable Experiment: A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the core while the reactor is operating.

1.12 Fuel Element: A fuel elesmit Ls asingle TRIGA8 fuel rod.

1.13 Instrumented Element: An instrumented element is a special fuel element in which one or more thermocouples have been embedded for the purpose of measuring the fuel temperatures during react6or operation.

1.14 Irradiation Facilities: Irradiation facilities shall mean beam ports, including extension tubes with shields, thternal columns with shields, vertical tubes, rotating specimen rack, pneumatic transfer systeni, sample holding dummy fuel elements and any other in-tank irradiation facilities.

1.15 Measured Value: The measured value is the value of a parameter as it appears on the output of a channel.

1.16 Operable: A system or component shall be considered operable when it is capable of performing its intended function.

2

1.17 Operating: Operating means a component or system is performing its intended function.

1.18 Operational Core: An operational core shall be a fuel element core which operates within the licensed power level and satisfies all the requirements of the Technical Specifications.

1.19 Pulse Mode: Pulse mode shall mean any operation of the reactor with the mode selector switch in the pulse position.

1.20 Radiation Center Complex: The physical area defined by the Radiation Center Building and the fence surrounding the north, west, and east sides of the Reactor Building.

1.21 Reactor Operating: The reactor is operating whenever it is notsecured or shut down.

1.22 Reactor Safety Systems: Reactor safety systems are those systems, including their associated input channels, which are designed to initiate, automatically or manually, a reactor scram for the primary purpose of protecting the reactor.

1.23 Reactor Secured: The reactor is secured when:

a. Either there is insufficient moderator available in tie reactor to attain criticality or there is insufficient fissile material in the reactor to attain criticality under optimum available conditions of moderatio dand reflection; or,

b. All of the followingeii st:

1. The four (4) neutron absorbing control rods are fully inserted as required by technii&l specifications

2. The reactor Issiliut down; 3 No experiments or irradiation facilities in the core are being moved or serviced(Itat have, on movement or servicing, a reactivity worth

exceedinigthe maximum value of one dollar; and

4. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods.

1.24 Reactor Shutdown: The reactor is shut down when:

a. It is subcritical by at least one dollar both in the reference core condition and for all allowed ambient conditions, with the reactivity worth of all installed experiments and irradiation facilities included; and 3

b. The console key switch is in the "off' position and the key is removed from the console.

1.25 Reference Core Condition: The reference core condition is the reactivity condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible

(<0.30 dollars).

1.26 Review: A review is a qualitative examination of records, proceduresfor other documents prior to implementation from which appropriate recommendation"re made.

1.27 Safety Channel: A safety channel is a measuring channel in the reactor safety system.

1.28 Scram time: Scram time is the elapsed time between reaching a limitingr safety system set point and the instant that the slowest scrammable. control rod reaches its fully-inserted position.

1.29 Should, Shall, and May: The word "shall" is used to ,denote a requirement; the word "should" is used to denote a recommendation; and the word' "may" to denote permission, neither a requirement nor a recommendation.

1.30 Shutdown Margin: Shutdown margin'Shalli meanthe minimum shutdown reactivity necessary to provide confidence that the reactor can be mnae subcritical by means of the control and safety systems and will remaili*i subcritical wthout further operator action, starting from any permissible operating co0ilition With the most reactive rod is in its most reactive position.

1.31 Square-Wave Mode (S.-W. Mode): The square-wave mode shall mean any operation of the reactor with the mode selector switch >itthe square-wave position.

1.32 Steady-State Mode (S.-S. Mode): Steady-state mode shall mean operation of the reactor with the mioe selector swifch in the steady-state position.

1.33 Substantive Changes: Substantive changes are changes in the original intent or safety significance of an action orey,\,ct.

1.34 Surveillance Intervals: Allowable surveillance intervals shall not exceed the following:

a. Biennial - interval not to exceed 30 months

b. Annual - interval not to exceed 15 months

c. Semi-annual - interval not to exceed 7.5 months.

d. Quarterly - interval not to exceed 4 months.

4

e. Monthly - interval not to exceed 6 weeks.

f. Weekly - interval not to exceed 10 days.

5

THIS PAGE INTENTIONALLY LEFT BLANK

¢ 6

2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.1 Safety Limit-Fuel Element Temperature Applicability. This specification applies to the temperature of the reactor fuel.

Objective. The objective is to define the maximum fuel element Iemperature that can be permitted with confidence that no damage to the fuel elemit cl-Jadding shall result.

Specifications. The temperature in a TRIGA fuel elemenit ýsl11Int exceed 2,1000 F (1,150 C) under any mode of operation.

Basis. The important parameter for a TRIGA 0reactor is the fuel elemeit temperature.

This parameter is well suited as a single specification especially since it can be measured. A loss of the integrity of the fuel elemieht cl addig could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caised by the presence of air, fission product gases, and hydrogen from-the dissociation of the hydrogen and zirconium in the fuel-moderator. Themiagmiltde of this pressure is determined by the fuel-moderator temperature and the ratio o IIgkIve t0 zirconium in the alloy.

The safety limit for the 'TRIGAO fuel element is based on data which indicate that the stress in the claddi iidue to.the hydrogen pressure from the dissociation of zirconium hydride will remaiii below th ultimate stress, rovided the temperature of the fuel does not exceed 21000 F (1150' Q aiid the fuel cladding is water cooled (SAR 4.5.3.1).

7

2.2 Limiting Safety System Setting Applicability. This specification applies to the scram settings which prevent the safety limit from being reached.

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

Specifications The limiting safety system setting shall be equal to or less than 510°C (950'F) as measured in an instrumented fuel element. The instrumented fuel element shall be located in the B-ring.

Basis. During steady state operation, temperatures were calculated for the beginning-of-life normal core. Linear extrapolation of temperature and power indicates that an IFE power of 23.2 kW will produce an indicated power of 5100 C in the IFE at the midplane thermocouple location. The highest ratio of maximum to minimum power for elements in the B-ring was found to be 1.036, soiftelFEt F iss generating 23.2 kW, the maximum power in any B-ring element would be limiited to 23.1 x 1.036 = 24.03 kW. For a power of 24.03 kW, the maximum temperature anywhere in the hot channel fuel element will be 588°C. The accuracy of the temperature indication is +/-5°C. Even including the error in the temperature measurement, this value is well below the Safety Limit.

8

3 LIMITING CONDITIONS OF OPERATION 3.1 Reactor Core Parameters 3.1.1 Steady-state Operation Applicability. This specification applies to the energy generated in the reactor during steady-state operation.

Objective. The objective is to assure that the fuel temperature safety limit shall not be exceeded during steady-state operation.

Specifications. The reactor power level shall not exce'd 1.1 MW except for pulsing operations.

Basis. Thermal and hydraulic calculations indicatc that TRIGA8 fuel may be safely operated up to power levels of at least 1.9 MW with'natural convection cooling (SAR 4.5.3.3 and 4.5.3.3.9).

3.1.2 Shutdown Margin Applicability. These specifications apply the reaciv.ity condition of the reactor and the reactivity worths ofcontrol rods and exieriments. They apply for all modes of operation. .

Objective. The objective is to a,ýýureL that the reactor can be shut down at all times and to assure that the fuel temperature safet!jmhlt shall not be exceeded.

Speciffications. The reactor shall niot be operated unless the following conditions exist:

The shutdown margin provided by control rods shall be greater than $0.55 with:

a. Irradiation facilities and experiments in place and the total worth of all no1n1lsecured experiments in their most reactive state;

b. The most reactive control rod fully-withdrawn; and

c. The reactor in the reference core condition.

Basis. The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the most reactive control rod should remain in the fully-withdrawn position.

9

3.1.3 Core Excess Reactivity Applicability. This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments. It applies for all modes of operation.

Objective. The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit shall not be exceeded,.

Specifications. The maximum available excess reactivity based on the reference core condition shall not exceed $7.55.

Basis. An excess reactivity limit of $7.55 allows flexibility tohopera reactor in all core modes (NORMAL, ICIT and CLICIT) without the need to add or rpemoye fuel elements when changing between operating modes. The NORMAL core i'thmost reactive core. If operating a NORMAL core with the miinmirum shutdown nmargin of

$0.55 and typical control rod worths of $2.70 (Safety),,7S$21,70 (Shim), $2.70 (Regulating) and $4.00 (Transient) (section 4.2.2, (7ontrol Rods), the cialculated NRC core excess would be -$0.55+$2.70+$2.70+$2.70u $7.55. The shutdowal margin calculation assumes a) irradiation facilities and expemficfits in place and the total worth of all non-secured experiments in their most reactivc s ) the most reactive control rod fully-withdrawn and c) the reactor in the reference core cowniti. Activities such as changing out of the NORMAL core, moving away from, the reference state or adding negative worth experiments wllI iake core excess more negative and shutdown margin less positive. The only4activity which could result in requiring fuel movement to meet shutdown margin and cý 5ore excesiimits would be the unusual activity of adding an experiment withlarge positive rectivityi.wh 3.1.4 Pulse Mode Operation Applicability: This specifiation applies to the energy generated in the reactor as a result of a pulse : ihslertion of reactivity.

Objective. The objecitie is to assure that the fuel temperature safety limit shall not be exceeded.

Specifications. The reactivity to be inserted for pulse operation shall be determined and limited by a mechanical block and electrical interlock on the transient rod, such that the reactivity insertion shall not exceed $2.55.

Basis. The fuel temperature rise during a pulse transient has been estimated conservatively by adiabatic models. These models accurately predict pulse 10

characteristics for several core configurations and should be accepted with confidence, relying also on information concerning prompt neutron lifetime and prompt temperature coefficient of reactivity. These parameters have been established for these cores by calculations and have been confirmed in part by measurements at existing facilities. In addition, the calculations rely on flux profiles and corresponding power densities which have been calculated 3.1.5 This section intentionally left blank.

3.1.6 Fuel Parameters Applicability. This specification applies to all fuel elements.

Objective. The objective is to maintain integrit- 'f the fuel element cladding.

Specifications. The reactor shall not operate with dl iageLdtfl elements, except for the purpose of locating damaged fuel elements. A fuel e**iment shall be considered damaged and must be removed from-the core if:

a. The transverse bend exceeds 0.0625 inches over (he length of the cladding; ,

b. Its lengtl cceds its original" leength by Q01125 inches;

c. A"(dlddilng defectcexists as indicated by release of fission products; or

d. *,:yisual lnspI"*tion ideniies bulges, gross pitting, or corrosion.

Basis. Gross failure..r obvious visul deterioration of the fuel is sufficient to warrant declartion of the fuel as d'amaged. The elongation and bend limits are the values found acceptable to the USNRC INUREG-1537).

3.2 Reactor Control And Safety System 3.2.1 Control Rods, Applicability. This specification applies to the function of the control rods.

Objective. The objective is to determine that the control rods are operable.

Specification. The reactor shall not be operated unless the control rods are operable.

Control rods shall not be considered operable if:

11

a. Damage is apparent to the rod or rod drive assemblies; or

b. The scram time exceeds 2 seconds.

Basis. This specification assures that the reactor shall 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 assure the safety of the reactor (SAR 13.2.2.2.1).

3.2.2 Reactor Measuring Channels Applicability. This specification applies to the information which slhall be available to the reactor operator during reactor operation.

Objective. The objective is to specify the minimum number of measuring channels that shall be available to the operator to assure safe operation of tie reactor.

Specifications. The reactor shall not-be operated in the'specified mode unless the minimum number of measuring chanil2 lIsted in Table 1 are oPerable.

12

Table 1 - Minimum Measuring Channels Fuel Element Temperature 1 1 1 Linear Power Level 1 1 Log Power Level 1 1 Power Level 2 2 Nvt-Circuit - Ni1 -

(1) Any single Linear Power Level, Log.Power Level or Power Level imeasuring channel may be inoperable while the reactor is operating for the purpose of performing a channel check, test, or calibration.. .

(2) If any required measuring chaninels becomes inoperable while the reactor is operating for reasons other thai)tht 'identified in Technical Specification 3.2.2 (1) above, the channel shall be restored to operationwitlAin 5 minutes or the reactor shall be immediately shii~?dowi.

Basis. Fuel temperature displayed at the co1rol console gives continuous information on this parameterwhich has a specified safety limit. The power level monitors assure that the reactor power level is adequately monitored for both steady-state, square wave and pulse modes of operation. The specifications on reactor power level indication are included in this section, sin cethe power level is related to the fuel temperature. For footnote{ (1), taking tfheimeauring channels off-line for short durations for the purpose of a check, test or calibra ion' is considered acceptable because in some cases, the reactor must be op*rating in order to perform the check, test, or calibration. Additionally there exist two reduindant powerlevel indications operating at any given time while the third single channeliis off-line. For footnote (2), events which lead to these circumstances are self-revealing tOutihe operator. Furthermore, recognition of appropriate action on the part of the operator as aitreult of an instrument failure would make this consistent with TS 6.7.2.

3.2.3 Reactor Safety System Applicability. This specification applies to the reactor safety system channels.

Objective. The objective is to specify the minimum number of reactor safety system channels that shall be available to the operator to assure safe operation of the reactor.

13

Specifications. The reactor shall not be operated unless the minimum number of safety channels described in Table 2 and interlocks described in Table 3 are operable.

Table 2 - Minimum Reactor Safety Channels Fuel Element SCAM@ 10C Temperature SCA @51°C 1@,,:-<,- 1 Power Level SCRAM @ 1.1 MW(t) or less "2 2 Console Scram 1 -

Button Preset Timer Transient rod SCRAM @-15 15 -

sec after a pulse High Voltage SCRAM @ Ž25% of jnoinal operating voltage _,___1 Table 3 - Mlinimum Interlocks In~ock

~ jznctonEffectie

< /Modc lii,-doc ILiictO1 S-S. Pulse AQ Wide-Range Log Prvt control rod ivithdrawal Power Level (@less than 2 cps1 Channel @ less t.an 2 cps TransientRod Prevents appllcationi ofair unless Cylinder' full-,yuserted 1k1W Pulse . Pevets pulsifig above 1 kW I Interlock ShinPSafety, and Regulating Prevents simultaneous manual Rod Drive withdrawal of two rods Circuit Shim, Safety, j' and Regulating,' Prevents movement of any rod Rod Drive except transient rod Circuit Transient Rod Prevents pulse insertion of Cylinder reactivity greater than $2.55 Position________

14

(1) Any single Linear Power Level, Log Power Level or Power Level safety channel or interlock may be inoperable while the reactor is operating for the purpose of performing a channel check, test, or calibration.

(2) If any required safety channel or interlock becomes inoperable while the reactor is operating for reasons other than that identified in Technical Specification 3.2.3 (1) above, the channel shall be restored to operation within 5 minutes or the reactor shall be immediately shutdown.

Basis.

Fuel Element Temperature Scram: The fuel element temperature scram causes a scram in excess of the LSSS, which is 510°C. The supporting argument, f)r the safely limit of 1150'C are given in SAR 4.5.3.1. The LSSS is setýt0 less than hallf f1the safety limit. This is more than adequate to account for uncertainties in instrument response and core position of the instrumented fuel element. "

Power Level Scram: The set point forbcoth the safety and percent power channels are normally set to 106% of 1 MW(t), which is below the licensed p6wer of 1.1 MW(t).

The 6% difference allows for expected and observIed instrument fluctuations at the normal full operating power of 1 MW(t) to occur without *cramming the reactor unnecessarily. Conversely;I, -SAR13.2.2.2.2 shows that this set point is more than sufficient to prevent c:ceedi:c the reactivity insertion limit during non-pulsing operations and prevet; the operator from inad%ertently exceeding the licensed power.

Manual Scram: Tihe manial scram must bc functional at all times the reactor is in operatio... It s evalue for a scram set point. It is initiated by the reactor operator manually. I Preset Timer Scram: TIe preset timer ensures that the reactor power level will reduce to a low level after pulsing and preclude an unintentional restart or ramped increase to some equilibriumt ower.*

High Voltage Scramul The high voltage scram must be set to initiate a scram before the high voltage for any of the three detectors reaches 25% or less of the nominal operating voltage. The loss of operating voltage down to this level is an indication of detector failure. Many measuring channels and safety systems are fundamentally based upon accurate response of the detectors.

Wide-Range Log Power Level Channel Interlock: The rod withdrawal prohibit interlock prevents the operator from adding reactivity when the count rate on the wide-range log power channel falls below 2 cps. When this happens, the count rate is 15

insufficient to produce meaningful instrumentation response. If the operator were to insert reactivity under this condition, the period could quickly become very short and result in an inadvertent power excursion. A neutron source is added to the core to create sufficient instrument response that the operator can recognize and respond to changing conditions.

Transient Rod Cylinder Interlock: This interlock prevents the application of air to the transient rod unless the rod is fully inserted. This will prevent the'Operator from pulsing the reactor in steady-state mode.

1 kW Pulse Interlock: The 1-kW permissive interlock is desi]gned to prevent pulsing when wide range log power is above 1-kW. SAR 13.2.2.2.1 showsvtIat the peak temperature reached during an end-of-life core willl be*1, 15 OOC for an iIitia1 fuel temperature of 20 0 C. The methodology clearly shows that if the initial temperature was higher, the resulting peak temperature must be lower. However, there has not bden analysis or experiment to look at the relationship betieen'he atgenerated, within the fuel at power (i.e., > 1-kW) and heat generated on the surfacc ()f the fuel during a pulse.

Therefore, this interlock prevents the reactor from pulsinge t power levels which produce measurably significant increasesin fuel temperaturi-e Shim, Safety and Regulating Rod Drive Circuit:Vf*he single rod withdrawal interlock prevents the operator from removing multiple control rod.,simultaneously such that reactivity insertions from control rod manipulation is donie in a controlled manner. The analysis in SAR 13.2:2.2.2 and, 13.2.2.2.3 show that the reactivity insertion due to the removal rate of the most reactive-rod or all the control rods simultaneously is still well below the reactivity insetion design limit of $2.59.

Shim, Safety and Reg-tiulating Rod Drive Circuit: In pulse mode, it is necessary to limitAtie reactivity inse6ed to less thanthe design limit of $2.59 at the end of core life anailct*en SAR 13.2.2_21. This interlock ensures that all pulse reactivity is due to only the tiraihsient rod whilý in pulse mode. Otherwise, any control rod removal in pulse mode would add to the inscred reactivity of the transient rod and create an opportunity for exceeding theireactiv It iInsertion limit.

Transient Rod Cylindefr Position Interlock: For the transient rod cylinder interlock, SAR 13.2.2.2.1 shows that the designed limiting reactivity insertion for the fuel is $2.59 at the end of core life. This interlock limits transient rod reactivity insertions below this value. Furthermore, this interlock is designed such that if the electrical (i.e., limit switch) portion fails, a mechanical (i.e., metal bracket) will still keep the reactivity insertion below the criterion.

16

For footnote (1), taking these safety channels off-line for short durations for the purpose of a check, test or calibration is considered acceptable because in some cases, the reactor must be operating in order to perform the check test or calibration. Additionally there exist two redundant power level indications operating at any given time while the third single channel is off-line. For footnote (2), events which lead to these circumstances are self-revealing to the operator. Furthermore, recognition of appropriate action on the part of the operator as a result of an instrument failure would make this consistent with TS 6.7.2.

3.3 Reactor Primary Tank Water Applicability. This specification applies to the primary water of the reactor-tank.

Objective. The objective is to assure that there is an adequate amount of watcr in the reactor tank for fuel cooling and shielding purposes, and that the bulk temperature of the reactor tank water remains sufficiently low to guarantee ractor tank integrity.

Specifications. The reactor primary water !shallexhibit the f1'1owing parameters:

a. The tank water level shall be greater than 14 ifeet above the top of the core; *..... ,:,'!i!~~i'*:*

b. ThIýe builk tak vwater temperature shall be less than 120'F (49°C); and

c. The conductiity

• *of ......

k \iater shall be less than 5 ýtmhos/cm.

Basis.- The minimum height of'l14 Ifeet of water above the top of the core guarantees that theie-is s'-ufficient wateri for effective cooling of the fuel and that the radiation levels at the top of the reactor are within acceptable levels (SAR 4.3, 4.5.3, and 11.1.5.5). The bulk water temperature limit is necessary, according to the reactor manufacturer, to ensure that the aluminum reactor tank maintains its integrity and is not degraded (SAR 4.3). Experience at many research reactor facilities has shown that maintaining the conductivity within te specified limit provides acceptable control of corrosion (NUREG-1537). .

3.4 This section intentionally left blank.

3.5 Ventilation System Applicability. This specification applies to the operation of the facility ventilation system.

17

Objective. The objective is to assure that the ventilation system shall be in operation to mitigate the consequences of possible releases of radioactive materials resulting from reactor operation.

Specifications.

a. The reactor shall not be operated unless the facility ventilation system is operating and the reactor bay pressure is maintained negative with respect to surrounding areas, except for periods of timie ot to exceed two (2) hours to permit repair, maintenance or testing of the ventilation system.

b. The ventilation system shall be shutdown upon a high civity alarm from the Exhaust Particulate Radiation Monitor Basis. During normal operation of the ventilation system, the annual average ground concentration of4'at in unrestricted areas is well below the applicable effluent concentration limit in 10 CFR 20. In, addition, the worst-case maximum total effective dose equivalent is well below the applicable annual limit for individual members of the public. This has been shown to be true for scmarios where the ventilation system continues to operate during the maximum lhypothetical accident (MHA), where the ventilation system is secured during the MHA, and where the ventilation system and the confinement building are hot present during the MHA (SAR 13.2.1). Therefore, operation of the reactor for short periods while the ventilation system is shut down for repair or testing does riot compromise the control over the release of radioactive material to the unrestricted area nor should it cause occupational doses that exceed those limits given in 100CT 0 tS ASAR 111.1. 1.1:2)i JTheýtwo hour exception to permit repair, maintenaince or 1testIig should not 1 diminish'the effectiveness of the reactor top area radiation monitor or'th'e contindous air~particulate radiation monitor. The sampling loeations for both of these irionitors aire located directly above the core. Any fission product ceIlase should be detected in the same manner regardless of the status of the ventilation 'systemr because df the close proximity of the sampling point to the source term. Moreover, rqdiatio0n monitors in the building, independent of the ventilation system, will giveL waning of high levels of radiation that might occur during operation of the reactor while tIe ventilation system is secured (SAR 11. 1.4.2). The exhaust gas and particulate radiation monitors will be affected by the status of the ventilation system as they are designed to monitor the ventilation exhaust directly and are not in close proximity to the source term (i.e., reactor core). However, control of the release into the unrestricted area will be minimally compromised because the ventilation will be by definition off and the leak rate is negligible compared to the ventilation rate.

Furthermore, this situation is bounded by the MHA scenario A (i.e., without the reactor building) and C (i.e., ventilation off) in the SAR (SAR 13.2.1.1).

18

3.6 This section intentionally left blank.

3.7 Radiation Monitoring Systems and Effluents 3.7.1 Radiation Monitoring Systems Applicability. This specification applies to the radiation monitoring information which must be available to the reactor operator during reactor operation.>>.

Objective. The objective is to specify the minimum radiation monitoring channels that shall be available to the operator to assure safe operation'of th**reactor.

Specifications. The reactor shall not be operated un'less the minimurhiIUmber of radiation monitoring channels listed in Table 4 are operating.

Table 4 - Minimum Radiation Monitoring Channels e

~Radiaitlio Moitorn4 ChanneijksL Nmificr, Reactor Top Area Radiati i Montor 1t Continuous Air Particulate Radiation Monitor 1 Exhaust Gas Radiation Monitor 1 Exhaust Particulate, Radiation Monitor 1 Exception: Whenr a singlerequIred-radiation monitoring channel becomes inoperiableopoeratins may continue only if portable instruments, surveys, or arnalyses may be substituted for the normally installed monitor within one (1)

"ihour of discoveryfor periods jit to exceed one (1) month.

Thours'rdiationcyo of eilý'n Basis. The radiation monitors provide information to operating personnel regarding routine releases of radioactivity and any impending or existing danger from radiation.

Their operation will provide sufficient time to evacuate the facility or take the necessary steps to prevent the spread of radioactivity to the surroundings. Furthermore, calculations show that for both routine operations and under the three accident scenarios identified in SAR 13.2.1.1, predicted occupational and general public doses are below the applicable annual limits specified in 10 CFR 20 (SAR 11.1.1.1 and SAR 13.2.1).

That being the case, we have reasonable assurance that the applicable regulatory limits are being satisfied for the one hour period.

3.7.2 Effluents 19

Applicability. This specification applies to the release rate of 4 1Ar.

Objective. The objective is to ensure that the concentration of the 41 Arin the unrestricted areas shall be below the applicable effluent concentration value in 10 CFR 20.

Specifications. The annual average concentration of 41Ar discharged into the unrestricted area shall not exceed 4 x 10-6 /Ci/ml at the point of discharge.

Basis. If 4 1Ar is continuously discharged at 2.5 x 10-6 ptCiniV(miie., the concentration produced when the nitrogen purge of the rotating rack is off, allv'alves on the argon manifold are open, and all beam port valves are open),-measurements and calculations show that 41Ar released to the unrestricted areas under the w'orst-case weather conditions would result in an annual TEDE of 5 mrem (SARI 1..1.1.1). This is only 50% of the applicable limit of 10 mrem (Regulatory Guide 4.20). Therefore, an emission of 4 x 10-6 OCi/ml would correspond to an annual TEDE ofl8<mremw-hich is still 20%below the applicable limit.

3.8 Limitations on Experiments 'V 3.8.1 Reactivity Limits Applicability. This Ification applies to experimentsinstalled in the reactor and its irradiation facilit]es...

Objective. The objective is t p1revent damage to the reactor or excessive release of radioactive materials in the c\ nt ofain experiment failure.

Specific ations. The reactor shall not be operated unless the following conditions governing experiments exist:

a. The absolute value of the reactivity worth of any single unsecured experimenthall be less than $0.50; and

b. The SLIM of the absolute values of the reactivity worths of all experiments shallbe less than $2.55.

Basis. The reactivity limit of $0.50 for movable experiments is designed to prevent an inadvertent pulse from occurring and maintain a value below the shutdown margin.

Movable experiments are by their very nature experiments in a position where it is possible for a sample to be inserted or removed from the core while critical. That being said, Section 13.2.2.2.1 clearly shows that this value is still below the analyzed design limit of $2.59 for end of life fuel.

20

The reactivity worth limit of $2.55 for all experiments is designed to prevent an inadvertent pulse from exceeding the design limit of $2.59 for end of life fuel. This limit applies to movable, unsecured and secured experiments. Regardless of any other administrative or physical requirements, this limit has been shown in Section 13.2.2.2.1 to protect the reactor during the fuel's entire lifetime.

3.8.2 Materials Applicability. This specification applies to experiments nstalledi the reactor and its irradiation facilities.

Objective. The objective is to prevent damage to the reactor or excessive relfais of radioactive materials in the event of an experiment failure.

Specifications. The reactor shall notle operated unless the following conditions governing experiments exist: W

a. Exunosive materials, suchas gun d;er, TNT1, antroglycerin, or PETN, in quantities greater than 25 milligrams"1ha1ll not be irradiated in the reactor or irraditionffcilities. Explosive materials in quantities less than 25 milligrami s ma be,,tý%h irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally emonstrateI )tobeless than half the design pressure of the container.;

and]

b. Expe..i.**its con..taining corrosive materials shall be doubly encapsulated.

The failurŽ (f an encapsulation of material that could damage the reactor shall result In removal of the sample and physical inspection of

Žpotentially

• datmaged components.

Basis. This specilficatio is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials. Operation of the reactor with the reactor fuel or structure potential damages is prohibited to avoid potential release of fission products.

3.8.3 Failures and Malfunctions Applicability. This specification applies to experiments installed in the reactor and its irradiation facilities.

21

Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications. Where the possibility exists that the failure of an experiment under normal operating conditions of the experiment dr reactor, credible accident conditions in the reactor, or possible accident conditions in the experiment could release radioactive gases or aerosols to the reactor bay or the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor bay or the unrestricted area will not result in exceeding tilehapplicable dose limits in 10 CFR 20, assuming that:

a. 100% of the gases or aerosols escape from the experinent;

b. If the effluent from an irradiation facility exhausts throuia lioldup tank which closes automatically on high-radiation level, at least 100o4f the gaseous activity or aerosols produced will escape;

c. If the effluent from an irradiation facility, exhausts through a filter installation designed for -reater than 99% effilciency for 0.3 micron particles, at least 10% ol'tliese aerosols can escape; and

d. For materials whose boiling pointfis above 130 F and where vapors formed b\ oiling this material can escape only through an undisturbed col umin of water ibove the core, 10% of these vapors can escape.

Basis. This specification i sintlnde [to meta the purpose of 10 CFR 20 by reducing the likelihood that released /airbore radoactivity to the reactor bay or unrestricted area surrounding the OSTR will result M exceeding the total dose limits to an individual as specified in 10 CFR 20.`

3.9 This section intentionally left blank.

22

4 SURVEILLANCE REQUIREMENTS 4.0 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.

a. Surveillance requirements may be deferrededuring reactor shutdown (except Technical Specifications 4.3.a and 4.3.e); however,"they shall be completed prior to reactor startu1)Iunless reactor operation is required for performance of the surveillance. SLch surveillance shall be performed as soon as practicable after reactor staitup. Scheduled surveillanice, which cannot be performed with the reactor operating, may be deferred until a planned reactor shutdow:n.

b. Any additions, modificatibns, or aintenance to) the ventilation system, the core and its associated support structure, th e pool or its penetrations, the pool coolant system, the rod drive meclianism or the reactor safety syste~m shall bemade and tested in accordance with the specifications to which the systemsi were originally designed and fabricated or to specifications reviewed by the Reactor Operations Committee. A system shall not b*iecons>iderd Openab until after it is successfully tested.

Basis. This specification relahcjt c.ange sin reactor systems which could directly affect the safety of the reactor. As long ,, cIanges or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presentlyac"cepted operatihg criteria.

4.1 Reactor Core Parameters Applicability. This specification applies to the surveillance requirements for reactor core parameters.

Objective. The objective is to verify that the reactor does not exceed the authorized limits for power, shutdown margin, core excess reactivity, specifications for fuel element condition and verification of the total reactivity worth of each control rod.

Specifications.

23

a. A channel calibration shall be made of the power level monitoring channels by the calorimetric method annually.

b. The total reactivity worth of each control rod shall be measured annually or following any significant change (>$0.25) from a reference core.

c. The shutdown margin shall be determined prior to each day's operation, prior to each operation extending more than one day, or following any significant change (>$0.25) from a reference core.

d. The core excess reactivity shall be determined aMii'ually or following any significant change (>$0.25) from a reference core.

e. Twenty percent of the fuel elements ýcomprising the core shall be inspected visually for damage or deterioration and measured1or concentric or other swelling anniiallv such that,the entire core is inspected over a five year period. Annuailinspections shall be of non-repeating representative samples of fuel 'elements from each ring.

Basis. Experience has shown that the ideiified frequencies will ensure performance and operability for each of these systems or components. The value of a significant change in reactivity (>$0.25) is measurablef and will ýensuredequate coverage of the shutdown margin aftertaking into account t~he*accumulation of poisons. For inspection, looking at fuel elements from each ring annually will identify any developing fuel integrity issues throughout the core. Furthermore, the observed mechanism for non-conforming fuel at the OSTR has been exclusively concentric swell. Looking for swell will not only provide early indication '6f fuel non-conformance but it will significantly reduce the amoultof fiel mov:ements needed.

4.2 Reactor Control and Safety Systems Applicability. This specification applies to the surveillance requirements of reactor control and safety, systems .

Objective. The objetive is to verify performance and operability of those systems and components which are directly related to reactor safety.

Specifications.

a. The control rods and drives shall be visually inspected for damage or deterioration biennially.

b. The scram time shall be measured annually.

24

c. The transient rod drive cylinder and associated air supply system shall be inspected, cleaned and lubricated as necessary, semi-annually.

d. A channel check of each of the reactor safety system channels for the intended mode of operation shall be performed prior to each day's operation or prior to each operation extending more than one day.

e. A channel test of each item in Table 2 and 3 in section 3.2.3 shall be performed semi-annually.

f. A channel calibration of the fuel temperaturmneiai'suring channel shall be performed annually.

Basis. Experience has shown that the identified frequencies will ensure performance and operability for each of these systems or componeits.

4.3 Reactor Primary Tank Water Applicability. This specification appliesto the surveillance requirements for the reactor tank water.

Objective. The objective is to assure that the reac tr tank water level and the bulk water temperature monitoring systmns are operatin1g, and to verify appropriate alarm settings.

Specifications. '

a. A channel check o)f the reractor tank water level monitor shall be performred mronithly

b. A channel check of thereactor tank water temperature system, including a verificatiorl of the alarm set point, shall be performed prior to each day's operation or prior to each operation extending more than one day.

c. Ani operabIlity check of the reactor tank temperature alarm shall be perfonritd monthly.

d. A channel calibration of the reactor tank water temperature system shall be performed annually.

e. The reactor tank water conductivity shall be measured monthly.

Basis. Experience has shown that the frequencies of checks on systems which monitor reactor primary water level, temperature, and conductivity adequately keep the tank 25

water at the proper level and maintain water quality at such a level to minimize corrosion and maintain safety.

4.4 This section intentionally left blank.

4.5 Ventilation System Applicability. This specification applies to the reactor bay confinement ventilation system.

Objective. The objective is to assure the proper operation 6f the ventilation system in controlling releases of radioactive material to the unrestricted area.,

Specifications.

a. A channel check of the reactof ba\ coinenit ventilation system's ability to maintain a negative pressureI in t reactor bay with respect to surrounding areas shall be performed pr irto each day's operation or prior to each operation extending more thaone da

b. A channel test of the reactQr bay confinement \enition system's ability to be secured shall be performed annafi5y1:< atnyesby Basis. Experience has demonstrated that tests of the ventilation system on the prescribed daily and anIinual basis are sufficient to assure proper operation of the system and its control over releases of radioactive material.

4.6 This, section inientionall y left blank.

4.7 Radiation Monitoring Sysei Applicability. This specifi cation applies to the surveillance requirements for the area radiation monitoring equipment and the air monitoring systems.

Objective. The objecive is to assure that the radiation monitoring equipment is operating properly and to verify the appropriate alarm settings.

Specifications.

a. A channel check of the radiation monitoring systems in section 3.7.1 shall be performed prior to each day's operation or prior to each operation extending more than one day.

26

b. A channel test of the continuous air particulate, exhaust gas, and exhaust particulate radiation monitors shall be performed monthly.

c. A channel calibration of the radiation monitoring systems in section 3.7.1 shall be performed annually.

Basis. Experience has shown that an annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span.

4.8 Experimental Limits Applicability. This specification applies to the surveillance requirements for experiments installed in the reactor and its irradiatioin facilities.

Objective. The objective is to prevent the c6nduct of experiments which may damage the reactor or release excessive amounts of radioactive imnterials as a result of experiment failure.

Specifications.

a. The reactivity worth of an experiiý:iet sall be estimated or measured, as appropriate, before reactor option with said experiment.

b. An experiment shall not be installed in the reactor or its irradiation faclllties unlessSasfety analysil has been performed and reviewed for compliance witli Section -3.8by the Reactor Operations Committee in full ac dwith Section 6.2.3 of these Technical Specifications, and the proced s wi re established for this purpose.

Basis. ]E-perience has shown that experiments which are reviewed by the staff of the OSTR and~II eReactor Operations Committee can be conducted without endangering the safety of the reiator or exc eding the limits in the Technical Specifications.

4.9 This section intentionally left blank.

27

5 DESIGN FEATURES 5.1 Site and Facility Description Applicability. This specification applies to the Oregon State TRIGA Reactor site location and specific facility design features.

Objective. The objective is to specify the location of specific facility design features.

Specifications.

a. The restricted area is that area inside the fnCe surrdi.indig the reactor building and the reactor building itself. The unrestricted area is that area outside the fence surrounding the reactor building.

b. The reactor building houses the TR-IG reactor and is abutted to the Oregon State University Radiation CeNintr Building.

C. The reactor bay shall be equipped with ven.i.tion systems designed to exhaust air or other gascfrom the reactor buildingland release them from a stack at a minimum of 65 feet fr6m ground level.

d. Emergency, s]utdown controls for the ventilation systems shall be located in theteaictor control room.

Basis. The Radiationu enter, reactor building and site description are strictly defined (SAR 2.0). The facility Is4uiic~l SLIC1Sft:ta* the ventilation system will normally maintain ,anegative pressurenhe Reactor Building with respect to the outside atmosphere so that there, will bIno uncontrolled leakage to the unrestricted envio ent. Controlrsr startupand normal operation of the ventilation system are located in the reactor control room: Proper handling of airborne radioactive materials (in emergency, situations) can be conducted from the reactor control room with a minimum of ekposure to operating personnel (SAR 9.1 and 13.2.1).

5.2 Reactor Coolant System Applicability. This specification applies to the tank containing the reactor and to the cooling of the core by the tank water.

Objective. The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.

Specifications.

a. The reactor core shall be cooled by,natural convective water flow.

28

b. The tank water inlet and outlet pipes to the heat exchanger and to the demineralizer shall be equipped with siphon breaks not less than 14 feet above the top of the core.

c. A tank water level alarm shall be provided to indicate loss of coolant if the tank level drops 6 inches below normal level.

d. A bulk tank water temperature alarm shall be proyi'ded to indicate high bulk water temperature if the temperature exceeds 120'F (49 °C).

Basis.

a. This specification is based on thermal and hydraulic calculations which show that the TRIGA core caniioerate in a safe manner at 'power levels up to 1.9 MW with natural cohve:tlin flow of the coolant wat&er(SAR 4.5.3.3). <

b. In the event of accidental isiphoning of tank %ter through inlet and outlet pipes of the heat exchanger or demineralizer systeWm, the tank water level will drop to a level no less.tlhanJ14"feet from th, top of the core (SAR 5.2).

C. Loss-o)f-coolA alarm caused by a water level drop of no more than 6 incLhe:'rovIdes a timely warning so that corrective action can be initiatl. I his rm is located in the control room (SAR 5.2).

d. The bulk t tiemperature*alarm provides warning so that corrective action caan be inuitiated in a timely manner to protect the quality of the reactor tailk, The alari'is located in the control room (SAR 7.2.3.2).

5.3 Reactor Core and Fuel 5.3.1 Reactor Corei*

Applicability. This specification applies to the configuration of fuel and in-core experiments.

Objective. The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities shall not be produced.

Specifications.

29

a. The core assembly shall consist of TRIGA fuel elements.

b. The fuel shall be arranged in a close-packed configuration except for single element positions occupied by in-core experiments, irradiation facilities, graphite dummies, aluminum dummies, stainless steel dummies, control rods, and startup sources.

c. The reactor shall not be operated at power levels exeeding 1 kW with a core lattice position water filled, except for positIons on the periphery of the core assembly.

d. The reflector, excluding experiments and irradiation facilities, shall be water or a combination of graphite..ard water.

Basis.

a. Only TRIGA fuel is anticipated to ever be used (SAR 4.2).

b. In-core water-filled experiment positions have been demonstrated to be safe in the Gulf Mark II1 reactor. The largest values of flux peaking will be experienced in hydrogenou0in-core irradiation positions. Various non-hydrogenous experiments posiltined M element positions have been demonstrated to be safe in TRIGA fuel elemnent cores up to 2-MW operation (SAR 4.2).

c. For cases-where one in-core posit Ion is water filled, except in the core

.perip.....her.

reactor power level is reduced to 1 kW to ensure

iniiii

..... pea.k power generatibn levels in adjacent element positions (SAR C.1. The core will be ass~embled in the reactor grid plate which is located in a tank of light water. Water in combination with graphite reflectors can be Lised for neutron economy and the enhancement of irradiation facility rad IitIoiquirements (SAR 4.2).

5.3.2 Control Rods:

Applicability. This specification applies to the control rods used in the reactor core.

Objective. The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications.

30

a. The shim, safety, and regulating control rods shall have scram capability and contain borated graphite, B 4 C powder or boron, with its compounds in solid form as a poison, in aluminum or stainless steel cladding. These rods may incorporate fueled followers which have the same characteristics as the fuel region in which they are used.

b. The transient control rod shall have scram capability, and contain borated graphite or boron, with its compounds in a solid form as a poison in an aluminum or stainless steel cladding. The transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate an aluminum- or air-follower.:i-. ý Basis. The poison requirements for the control rods are satisfied by using'neutron absorbing borated graphite, B 4 C powder or boron as its compounds. These materials must be contained in a suitable clad material such as aluminum or stainless stel to ensure mechanical stability during movement and to i'okite the'poison from the tank water environment. Control rods that are fuel-followed Irovide additional reactivity to the core and increase the worth of the control rod. The us:of fueled-followers in the fueled region has the additional advantage*of reducing flux p*'Akin the water-filled regions vacated by the withdrawal of the"corntrol, rods. Scraiii pabilities are pro-ided for rapid insertion of the control rods which is the primary safety feature of the reactor.

The transient control rod is designed for rapid' Withdrawalfrom the reactor core which results in a reactor pulse. The nuclear behavior of the air- or aluminum-follower, which may be incorporated Iinto the tninsient rod, is similar to a void. A voided-follower may be required in certain oe~oadinigs to reduce flux peaking values (SAR 4.2.2).

5.3.3 Reactor Fuel Applicability. This specification applies to the fuel elements used in the reactor core.

Objective. The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their piysical and nuclear characteristics.

Specifications. TRIGA Fuel Elements The individual unirradiated fuel elements shall have the following characteristics:

1. Uranium content: maximum of 9 wt% enriched to a nominal 70% 235U;

2. Hydrogen-to-zirconium atom ratio (in the ZrHx): between 1.5 and 1.65;

3. Natural erbium content (homogeneously distributed): between 31

1.1 and 1.6 wt%;

4. Cladding: 304 stainless steel, nominal 0.020 inches thick; and

5. Identification: top pieces of fuel elements will have characteristic markings to allow visual identification of elements.

Basis. A maximum uranium content of 9 wt% in a TRIGA element is about 6% greater than the design value of 8.5 wt%. Such an increase in loading wo uld result in an increase in power density of about 2%. Similarly, a minimum"'rbium content of 1.1% in an element is about 30% less than the design value. This variation would result in an increase in power density of only about 6%. An increase in local power density of 6%

reduces the safety margin by, at most, 10%. The maximum hydrogen-to-zirconium ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element cladding of about a factor of two greater than te value resulting from a hydrogen-to-zirconium ratio of 1.60. However, this increase in the cladding stress during an accident would not exceed the rupture strength of the cladding,(SAR 4.2.1).

5.4 Fuel Storage Applicability. This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective. The objective S1, to xasure that fuel which is being stored shall not become critical and shall not reach an utsafe temperature.

Specifications.

a. All fuel elemrients shlall be stored in a geometrical array where the k-effective is less than 0.9 for all conditions of moderation.

b':,. Irradiated fuel elements and fuel devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that

,thetemperature of the fuel element or fueled device will not exceed the saf*ty limit.

Basis. The limits imposed are conservative and assure safe storage (NUREG-1537).

32

THIS PAGE INTENTIONALLY LEFt BLANK 33

6 ADMINISTRATIVE CONTROLS 6.1 Organization Individuals at the various management levels, in addition to being responsible for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license, technical specifications, and federal regulations.

6.1.1 Structure .

The reactor administration shall be related to the University as shown in Figure 1.

6.1.2 Responsibility The following specific organizational levels, anrd ir-sponsibilities shall exist:

a. Vice-President for Research (Level -1)j: ITh1" Vice-President for Research is responsible for university center and 1iniItute organizations representing Oregon State University.

b. Radiation Center Director (Lc el 2),': The Radiation Center Director reports to the Vice-President for Research and is accountable for ensuring that all regulatory requirements, including inplementation, are in accordancc wihill1requirements of the U5SNRC and the Code of Federal Regukitions.

c. Reactor kdinistratoi-( l 3): The Reactor Administrator reports to

/,the Radiation Center Dirfct'ot and is responsible for guidance, oversight, and technical support of reactor operations.

d. Senior Health Physicist (Level 3): The Senior Health Physicist reports to the Radiation Center Director and is responsible for directing the activities of health physics personnel including implementation of the radiationslafety program.

e. Reactor Supervisor (Level 3): The Reactor Supervisor reports to the Reactor Administrator and is responsible for directing the activities of the reactor operators and senior reactor operators and for the day-to-day operation and maintenance of the reactor.

f. Reactor Operator and Senior Reactor Operator (Level 4): The Reactor Operator and Senior Reactor Operator report to the Reactor Supervisor and are primarily involved in the manipulation of reactor controls, monitoring of instrumentation, and operation and maintenance of reactor related equipment.

34

Figure 1 - Administrative Structure

_ Normal administrative reporting channel Communication

--- lines 35

6.1.3 Staffing

a. The minimum staffing when the reactor is operating shall be:

1. A reactor operator or the Reactor Supervisor in the control room;

2. A second person present in the Radiation Center Complex able to carry out prescribed instructions; and

3. If neither of these two individuals is the Reactor Supervisor, the Reactor Supervisor shall be readily available on call. Readily available on call means an individual who:

i. Has been specifically designated and the designation is known to the operator on duty; ii. Can be rapidly contactetdlby phone by the operator on duty; and iii. Is capabk ofettling to the reactor facility within a reasonable t Imel1n1dr1rnormal conditions (e.g., 30 minutes or within a I15-mil(e adils

b. A list ofreactor facility personnel by name and telephone number shall bereadily available in the control room for use by the operator. The list shall inilude:

1J. "iRadiationCenter Director 2., Reactor.Administrator

3. Senior Health Physicist

4. Any Licensed Reactor or Senior Reactor Operator

c. Events requiring the direction of the Reactor Supervisor:

1. Initial startup and approach to power of the day;

2. All fuel or control-rod relocations within the reactor core region;

3. Relocation of any in-core experiment or irradiation facility with a reactivity worth greater than one dollar; and 36

4. Recovery from unplanned or unscheduled shutdown or significant power reduction.

6.1.4 Selection and Training of Personnel The selection, training and requalification of operations personnel should be in accordance with ANSI/ANS 15.4 - 1988; R1999, "Standard for the Selection and Training of Personnel for Research Reactors."

6.2 Review And Audit The Reactor Operations Committee (ROC) shall have primary responsibility for review and audit of the safety aspects of reactor facility operations. Minutes, findings or reports of the ROC shall be presented to Level 1 and 1-\el 2 management within ninety (90) days of completion.

6.2.1 ROC Composition and Qualifications An ROC of at least five (5) members kowledgeable in fields ich relate to reactor engineering and nuclear safety shall review andievaluate the safety aspects associated with the operation and use of the facility. "The ROC shiall-be Ippointed by Level 1 management.

6.2.2 ROC Rules The operations of the ROC s-hall hein accordance with written procedures including provisionsf'-,

for, a Meeting frequency (at least annually);

b. Voting rules-,

c. Quorums (5 members, no more than two voting members may be of the operating staff at any time);

d. Method of submission and content of presentation to the committee;

e. Use of subcommittees; and

f. Review, approval, and dissemination of minutes.

6.2.3 ROC Review Function 37

The responsibilities of the ROC, or designated Subcommittee thereof, include, but are not limited to, the following:

a. Review all changes made under 10 CFR 50.59

b. Review of all new procedures and substantive changes to existing procedures;

c. Review of proposed changes to the technical specifications, license or charter;

d. Review of violations of technical specifications, license, or violations of internal procedures or instructions havlrg safety significance;

e. Review of operating abnormalities having safety significance:;

f. Review of all events from reports require Ixn sections 6.6.1 and 6.7.2 of these Technical Specifications;

g. Review of audit reports.

6.2.4 ROC Audit Function The ROC or a Subcommittee thereof shall audit reactor operations at least annually.

The annual audit shall nclude at-least the following:

a. tacilityoperatioiis for conformance to the technical specifications and applicable icense or charter conditions; L the retraiii and t requalification program for the operating staff; C. %t he results of action taken to correct those deficiencies that may occur in theireactor facility equipment, systems, structures, or methods of operation that affect reactor safety; and

d. the Emergency Response Plan and implementing procedures.

6.3 Radiation Safety The Senior Health Physicist shall be responsible for implementation of the radiation safety program. The requirements of the radiation safety program are established in 10 CFR 20. The program should use the guidelines of the ANSI/ANS 15.11 - 1993; R2004, "Radiation Protection at Research Reactor Facilities".

38

6.4 Procedures Written operating procedures shall be adequate to assure the safety of operation of the reactor, but shall not preclude the use of independent judgment and action should the situation require such. Operating procedures shall be in effect for the following items:

a. Startup, operation and shutdown of the reactor;

b. Fuel loading, unloading, and movement within th, reactor;

c. Maintenance of major components of systems thuat could have an effect on reactor safety;

d. Surveillance checks, calibrations, and inspections required 1%the technical specifications or those that have an effect on reactor safety;

e. Radiation protection; '

f. Administrative controls' f6pperations and maintenance and for the conduct of irradiations and expciments, that could affect reactor safety or core reactivity; ,.

g. Shipping of rafdi6active materials;

h. Implementation of the Emergency Response Plan.

Substaiive changes totfhe above procedures shall be made only after review by the ROG . Except for radiation protectioni procedures, unsubstantive changes shall be approved prior to impler'efitation by the Reactor Administrator and documented by the Reactor Administrator witfli 120 days of implementation. Unsubstantive changes to radiation protrction procedures shall be approved prior to implementation by the SHP and documented by the .Senfior Health Physicist within 120 days of implementation.

Temporary deviations from the procedures may be made by the responsible senior reactor operator in order to deal with special or unusual circumstances or conditions.

Such deviations shall be documented and reported by the next working day to the Reactor Administrator.

6.5 Experiments Review and Approval Approved experiments shall be carried out in accordance with established and approved procedures. Procedures related to experiment review and approval shall include:

39

a. All new experiments or class of experiments shall be reviewed by the ROC and approved in writing by the Level 2 or designated alternates prior to initiation; and

b. Substantive changes to previously approved experiments shall be made only after review by the ROC and approved in writing by the Level 2 or designated alternates. Minor changes that do not significantly alter the experiment may be approved by Level 3 or higher.

6.6 Required Actions 6.6.1 Actions to Be Taken in Case of Safety Limit Violation In the event a safety limit (fuel temperature) is exceeded:

a. The reactor shall be shut down and reýctoi oeration shall not be resumed until authorized by the USNRC;

b. An immediate notification of the, occurrence sllIbe made to the Reactor Administrator, Radiation *CenterDlrector and CIairperson, ROC; and

c. A report, ýad any applicable followup report, shall be prepared and revied by tI (:ROC. The report shall describe the following:

1. applicable circumstances Ieading to the violation including, when knofxv' thecue C C -and*ieontributing factors;

2. -. effectsl o lhe violation upon reactor facility components, systems, or strteqad on the health and safety of personnel and the public; and

3. corrective action to be taken to prevent recurrence.

6.6.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2 Other than a Safety Limit Violation For all events which are required by regulations or Technical Specifications to be reported to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under Section 6.7.2, except a safety limit violation, the following actions shall be taken:

a. the reactor shall be secured and the Reactor Administrator and Director notified; 40

b. operations shall not resume unless authorized by the Reactor Administrator and Director;

c. the Reactor Operations Committee shall review the occurrence at their next scheduled meeting; and

d. a report shall be submitted to the NRC in accordance with Section 6.7.2 of these Technical Specifications.

6.7 Reports 6.7.1 Annual Operating Report An annual report shall be created and submitted by the Radiation Center Director to the USNRC by November 1 of each year consisting k '

a. a brief summary of operating experien,e icluding the energy produced by the reactor and the hours the reactor v ' itical;

b. the number of unplanned shutdowns, includingie'asons therefore; C. a tabulation o1f major preventati\e and corretive maintenance operations having safety significance;

d. a brief description, including a summary of the safety evaluations, of hanges in the ;facility orin procedures and of tests and experiments carried out putrsuant to 10 CFR 50.59; C~/ e. a summnar of the nature and amount of radioactive effluents released or discharged toithe environs beyond the effective control of the licensee as

,measured at or prior to the point of such release or discharge. The Ssummary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25 percent of the concentration allowe or recommended, a statement to this effect is sufficient;

f. a summarized result of environmental surveys performed outside the facility; and

g. a summary of exposures received by facility personnel and visitors where such exposures are greater than 25 percent of that allowed.

41

6.7.2 Special Reports In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be made by the Radiation Center Director to the NRC as follows:

a. a report not later than the following working day by telephone and confirmed in writing by facsimile to the NRC Operations Center, to be followed by a written report that describes the circumstances of the event within 14 days to the NRC Document Control Desk ofany of the following:

1. violation of the safety limit;

2. release of radioactivity from the site above allowed limits;

3. operation with actual safety system settings from required systems less conservative than the lihitfing safety system setting;

4. operation in violation of limiting 6onditions for operation unless prompt remedial'iýtioi is taken as pernitted'in Section 3;

5. a reactor safety system component malfunction that renders or could render the reactori *afety systerm incapable of performing its intended safety function. If the malfunction or condition is caused by maintenance, then no report is required;

6. anunianticipated or uncontrolled change in reactivity greater than

.one dollar. Reactor trips resulting from a known cause are excluded;

7. abnormal and significant degradation in reactor fuel or cladding, or both, coolant boundary, or confinement boundary (excluding ormin0 leaks) where applicable; or

8. *an observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

b. a report within 30 days in writing to the NRC, Document Control Desk, Washington, D.C. of:

1. Permanent changes in the facility organization involving Level 1-2 personnel; and 42

2. significant changes in the transient or accident analyses as described in the Safety Analysis Report.

43

6.8 Records 6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved if Less than Five Years

a. normal reactor operation (but not including supporting documents such as checklists, log sheets, etc., which shall be maintained for a period of at least one year);

b. principal maintenance activities;

c. reportable occurrences;

d. surveillance activities required by the Technical Specifications;.

e. reactor facility radiation and contaniination surveys;

f. experiments performed( wýith the reactor;

g. fuel inventories, receiptstind shipments;

h. approved changes to the operating proce(dues; and

i. Reactor Operations Committee meetings and audit reports.

6.8.2 Records to be Retained for at Least One Certification Cycle Pecords ot ng and requalification of certified reactor operators and senior

, reactor operators shall be retained at all times the individual is employed or until "the certification is renewed.

6.8.3 Records to be Retained for the Lifetime of the Reactor Facility

a. gaseous and liquid radioactive effluents released to the environs;

b. offsite environmental monitoring surveys;

c. radiation exposures for all personnel monitored;

d. drawings of the reactor facility; and e." Reviews and reports pertaining to a violation of the safety limit, the limiting safety system setting, or a limiting condition of operation.

44

This page intentionally left blank 3.

3 3.

7 45

Attachment II Oregon State University Differences between the proposed Technical Specifications draft revisions "I"and "J"

1. TS 4.2.e Existing: A channel test of each item in Table II in section 3.2.3 shall be performed semi-annually.

Proposed: A channel test of each item in Tables 2 and 3 in section 3.2.3 shall be performed semi-annually.

Basis: This was a typographical oversight. It should be consistent with the current TS.

2. TS 4.0.a Existing: Surveillance requirements may be deferred during reactor shutdown; Proposed: Surveillance requirements may be deferred during reactor shutdown (except Technical Specifications 4.3.a and 4.3.e);

Basis: It is appropriate to continue these two surveillances during shutdown as they are serve a safety function regardless of the status of the reactor.

3. TS 1.23.b.1 Existing: The four (4) neutron absorbing control rods are fully inserted or other safety devices are in shutdown position, as required by technical specifications Proposed: The four (4) neutron absorbing control rods are fully inserted as required by technical specifications; Basis: There are only four control rods in the core. No other safety devices exist.

4. TS 3.1.3

Existing: The maximum available excess reactivity based on the reference core condition shall not exceed $8.55.

Proposed: The maximum available excess reactivity based on the reference core condition shall not exceed $7.55.

Basis: This was a typographical error.

5. TS 6.1.1 Existing: The reactor administration shall be related to the University and USNRC structure as shown in Figure 1.

Proposed: The reactor administration shall be related to the University as shown in Figure 1.

Basis: This is consistent with the recommendations of ANSI/ANS-15.1-2007.

6. TS 6.1.3.b Existing: The list shall be updated annually and include:

Proposed: The list shall include:

Basis: This is consistent with the recommendations of ANSI/ANS-15.1-2007.

7. TS 6.2.2 Existing: The operations of the ROC shall be in accordance with a written charter including provisions for:

Proposed: The operations of the ROC shall be in accordance with written procedures including provisions for:

Basis: Procedures are the method for documenting structure and rules. This is also consistent with the recommendations of ANSI/ANS-15.1-2007.

8. TS 6.7.2.a.7 Existing: abnormal and significant degradation in reactor fuel or cladding, or both, coolant boundary, or containment boundary (excluding minor leaks) where applicable; or

Proposed: abnormal and significant degradation in reactor fuel or cladding, or both, coolant boundary, or containment boundary (excluding minor leaks) where applicable; or Basis: This was a typographical error.

9. TS 6.2.4 Existing:

The ROC or a Subcommittee thereof shall audit reactor operations at least annually.

The annual audit shall include at least the following:

a. Reactor operating records;

b. Operator Requalification Program;

c. Reportable Occurrences; and

d. Emergency Response Plan and implementing procedures.

Proposed:

The ROC or a Subcommittee thereof shall audit reactor operations at least annually.

The annual audit shall include at least the following:

a. facility operations for conformance to the technical specifications and applicable license or charter conditions;

b. the retraining and requalification program for the operating staff;

c. the results of action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operation that affect reactor safety; and

d. the Emergency Response Plan and implementing procedures.

Basis: This is consistent with the recommendations of ANSI/ANS-15.1-2007.