ML082261409

Oregon State University, Submittal of Relicensing Technical Specifications, Revision 1

ML082261409
Person / Time
Site:	Oregon State University
Issue date:	08/06/2008
From:	Keller S Oregon State University
To:	Alexander Adams NRC/NRR/ADRO/DPR/RTRBA
References
Download: ML082261409 (55)

Text

Radiation Center Oregon State University, 100 Radiation Center, Corvallis, Oregon 97331-5903 T 541-737-2341 i F 541-737-0480 I http://ne.oregonstate.edu/facilities/radiationcenter Oregon State UNIVERSITY August 6, 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

Subject:

Oregon State University Submission of Relicensing Technical Specifications Revision I Mr. Adams:

Attachment I contains the most current revision of the technical specifications, Revision I. If you have any questions, please call me atthe number above. I declare under penalty of perjury that the foregoing is true and correct.

Executed on:t Sincerely, S. Todd Keller Acting Director Enclosure cc:

Document Control, NRC Al Adams, NRC Craig Bassett, NRC John Cassady, OSU Rich Holdren, OSU Todd.Palmer, OSU Steve Reese, OSU 14b2 0

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

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

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APPENDIX A TO FACILITY LICENSE NO. R-106 AND BASIS FOR T

Current through Amendment #XX Date of Issuance: XXXX XX, XXXX

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TABLE OF CONTENTS I DEFINITIONS..................................................................................

I

1. 1 Audit.........................................................................................

1 1.2 Channel:......................................................................................

1 1.3 Channel Calibration.........................................................................

1 1.4 Channel Check..............................................................................

1 1.5 Channel Test.................................................................................

1 1.6 Confinement...............................................

7.................................1.

1.7 Control Rod................................................................................

1..

1.8 Core Lattice Position......................................................................

1.

1.9 Excess Reactivity:.......................................................... 2 1.10 Experim ent......................................................... 2 1.11 Experiment Safety Systems............................................ 2 1.12 Fuel Elem ent....................................................... 2 1.13 Instrumented Element................................2 1.14 M easured V alue............................

........................ 2 1.15 Irradiation Facilities.................................................2 1.16 O perable.......................................................... 2 1.17 Operating................................................................. 3 1.18 Operational Core.........................................................

~

3 1.19 Pulse M ode.........................................................3 1.20 Radiation Center Com lx............................................ 3 1.21 Reactor O peratipon ' ~.................................I.................... 3 1.22 Reactor Safetv Systemns.......................................................................

3 1.23 R eactor Se'cured................................................... 3 1.24 Reactor Shutdown: Hi Theeactor is shu~t down when:..................................

3 1.25 Reference Core Conditionl................................................................

4 1.26 R eview............................................................. 4 1.27,Saft Channel.......I..........

........... 4 1.28 ~S'cram time:.................................................................................

4 1.29 Shud Shall, and May...................................................................

4 1.30 Shutdo~xy M argin.ý*................................................. 4 1.31 Shutdown Reactivity......................................................................

4 1.32 Square-Wa-ixie Mode.......................................................................

4 1.33 Steady-State Mo~ide...............................................................

i......... 4 1.34 Substantive Changes........................................................................

4 1.35 Surveillance Intervals..................................................

I................... 4 2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING.....................

8 2.1 Safety Limit-Fuel Element Temperature.................................................

8 2.2 Limited Safety System Setting..................................

9 3 LIMITING CONDITIONS OF OPERATION.............................................

10 3.1 Reactor Core Parameters...................................................................

10

.3.1.1 Steady-state Operation...............................................................

10 i

3.1.2 Shutdow n M argin.............................................................................................

10 3.1.3 Core Excess R eactivity...................................................................................

11 3.1.4 Pulse M ode O peration....................................................................................

11 3.1.5 Core Configuration Limitations...................................................................

12 3.1.6 Fuel Param eters...............................................................................................

12 3.2 Reactor Control And Safety System......................................................................

12 3.2.1 C ontrol R ods...................................................................................................

12 3.2.2 Reactor M easuring Channels..........................................

13 3.2.3 Reactor Safety System...............................................

.............................. 14 3.3 R eactor Prim ary Tank W ater.......................................

..................................... 18 3.4 This section intentionally left blank.........................

18 3.5 V entilation System

,..................... 18 3.6 This section intentionally left blank..

20 3.7 Radiation Monitoring Systems and Effluents....................................................

20 3.7.1 Radiation M onitoring System s..........

........... 20 3.7.2 E0.........................e0...............

3.8 Lim itations on Experim ents.

21 3.8.1 R eactivity Lim its..................................................

21 3.8.2 M aterials 22..

.... I........

3.8.3 Failures and M alf nction.......

......... 22 3.9 This section intentionally left blank.............

.................... 23 4

SURVEILLANCE REQUIREMENTS 24 4.1 Reactor Core Para i t 24 4.2 Reactor Contro I and Safet Systems.

.............. 25 4.3 Reactor PrimlaryJTank Water.............................................................

26 4.4 This section intentionally left blank...............

27 4.5 V n ia i n Systemn..

................................................ 2 4.6 ThiL scetioin.tentionallyReft blank......

.......... 27 4.7 T" Rdiation..............................*...................

27 4

R8 E xpetrim ental L imi t s *.......................................... e..................................................

28 4.9 This., c ntentaonally left blank.

28 5 DESIGN FEATURES.........................

29 5.1 w ater..................

S e..it..................Desc......................................

29 5.2 Reactor Cool*nt System.........................................

29 5.3 Reactor Core Fuel.............

.........................................................30 5.3.1.. e.. t......

e......................................

e.C........................................................

.... 30 5.3.2 Control Rods..........

.... 31 5.3.3 Reactor Fuel..........

..................................... 32 5.4 Fuel Storage................................................................................

3.

3 6

ADM INISTRATIVE CONTROLS...............................................................................

35 6.1 Organization..................................................................... 35 6.1.1 Structure.................................................

............ 35 6.1.2 R esponsibility...............................................................................................

35.

ii

6.1.3 Staffi ng......................................................................................................

36 6.1.4 Selection and Training of Personnel........................................................

38 6.2 R eview A nd A udit.................................................................................................

38 6.2.1 Composition and Qualifications..............................................................

38 6.2.2 Charter and R ules....................................................................................

38 6.2.3 R eview Function......................................................................................

38 6.2.4 A udit Function........................................................................................

39 6.3 Radiation Safety..............................................

39 6.4 P rocedures.........................................................................

39

.......................... 4 6.5 Required Actions............

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 OccUrrece of tlhe$Type Identified in Section 6.6.2 Other than a Safety Limit Violation...........................

.................... 41 6.6 R epo rlts.........................................................

7*.'

--42 6.6 Reports x

42 6.6.1 Annual Operating Report........

6.6.2 Special R eports.....................

42 6.7 R ecords................

44..:

44 6.7.1 Records to be Retainedfor a Period of at H", ast Five Years or for the Life of

, X $ *21¢

• , 1 1 ',&

the Component Involved if Less tl~anFi.e Years............... *.,.........................................

44 6.7.2 Records to be Retained fratl'st,.One Tralnin ycle........................ 44 6.7.3 Records to be Retained for~the LItitof the Reactor Facility..........

44 Nilst of Table'g,,and Figure Table 1 - M inim um M easuring Cliannels..........................................................................

14 Table 2 - M inimumn Reacto

.i, t y Cnannexs*.....................................................................

15 Table 3 7,,**'***'<

15.,

T a l 3 -

inim um hl~ terl ock s:k :¢*..

15 TableI'1 nM oiC Muo.C...........................

20 Figure t, dm inistrativ,.,uct ure ::...........................

d............................................

......... 36

• -*:1'::::,

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Included in this doc..'ument4ae the Technical Specifications and the "Bases" for the Technical Specificafion4s. 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.

111

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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*ofPf't channel such that its output corresponds with acceptable accuracy to known values ofitlb, parameter which the channel measures. Calibration shall encompass the entire channel, in luding equipment actuation, alarm, or trip and shall include a Channel Test.

',ng equipment 1.4 Channel Check: A channel check is a qualitatiy&\\'erification of accetbe performance by observation of channel behavior. This verification.xWhire possiJhle, shall includ,.&&oparison of the channel with other independent channels or systi m ieaiisunfig-the same varibl.e.

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

1.6 Confinement

Confinement is an enclosure o'tlit:e iaciliry that is designed to limit the release of effluents between the enclosure and its exte*iidl oniironment through controlled or defined pathways.

1.7 Control Rod: A'6ontiol rod is.*a device fabricated from neutron absorbing material or fuel or both which is used to est iish neIutron flux cha'g s and to compensate for routine reactivity changes. A conti,, o, may

,ounit.allowing it to perform a safety function when the coupl',ng i'stmsiengagedcl4Tyýpes of c6mrol rods shall include:

}*i),,,M Regulating a

.Regulating Rod (Reg 'Rod): The regulating rod is a control rod having an lectric otorie and 'scram capabilities. It may have a fueled-follower section.

'poSltlOn be varied manually or. by the servo-controller.

b.

Shim A shim safety rod is a control rod having an electric motor drive and~scram 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 (kff=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 -cicore or shield structure so as to be a part of their design to carry out experiments is not normally co*sidered an experiment.

Specific experiments shall include:

a.

Secured Experiment: A secured experiment*i is any experiment or component of an experiment that is held in a stationaryposition relative totie reactor by mechanical means. The restraining forces must be substantialI\\;greater than those to which the experiment might be suibjected by hydraulic, pneumatc i buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of crediNbl malfunctions.

b.

Unsecured Experiment:

nAhunmsecured experiment is any experiment or component of an experiment t*aatdoes not meet thedefimnition of a secured experiment.

c.

Movable Experi*ent: A movable experimentf is one where it is intended that the entire exgriment may be moved in or near the core or into and out of the core while the reactor 1S operating.

1.12 Fuel Element:

fuel eleimntLsa single TRIGA fuel rod.

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

1.14 Irradiation Facilities: Irirdiation facilities shall mean beam ports, including extension tubes with shields, tlheral, columns with shields, vertical tubes, rotating specimen rack, pneumatic transfer system, 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 Comple and the fence surrounding the no 1.21 Reactor Operating: The r 1.22 Reactor Safety Systems:

associated input channels, which scram for the primary purpose of 1.23 Reactor Secured: The rea

a.

Either there is ins there is insufficier optimum availabl ex: The physical area defined by the Radiation Center Building rth, west, and east sides of the R tor Building.

reactor is operating wheneye it is nor shut down.

Reactor safety systems" are those systems incuding their are designed to initiae, automatically or manual;yA'a reactor protecting the reactorq*g.-

ctor is secured when:

ufficient moderator:available in t.reactor to attain criticality or nt fissile material intIU{ei"keactor totai criticality under e conditions 8tmodrati 6{ideflection; or, i&'ekist:

4)sdeuotron absorbing control rods are fully inserted or other i ces aredin,,'sniuitown positioný, as required by technical ir is shut down; ments or irradiation facilities in the core are being moved or hat have, on movement or servicing, a reactivity worth the maximum value of one dollar; and

b.

All of 1.

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:

3

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

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 record.,"proceduires or other documents prior to implementation from which appropriate reconmendations are made.

1.27 Safety Channel: A safety channel is a measuri.ng channel in the reactor sf~tf system.

1.28 Scram time: Scram time is the elapsed time between reaching a limiting safety system set point and the instant that the slowest scraimmable control rouidreaches its fully-inserted position.

1.29 Should, Shall, and May: The word slial~l',liosued to deniotea. requirement; the word "should" is used to denote a recommendation; iand te word "may"to denote permission, neither.

a requirement nor a recommendation.

1.30 Shutdown MarginP' Shutdown margin shall mean the minimum shutdown reactivity necessary to providez fidri~ce thattithe reactor can be. made subcritical by means of the control and safety systems and wi11 reinainsubcritical without further operator action, starting from any permissible operatingL conditio ith.

6ij.

tive rod is in its most reactive position.

1.31 Square-Wave Mode (S.-W. Node):;, The square-wave mode shall mean any operation of the reacbtor'wlth the mode see~ctor switCni-in the square-wave position.

1.32 Steady-State \\Mode (S.-S. Mode): Steady-state mode shall mean operation of the reactor with the mode sele'g;1N switchlin the steady-state position.

1.33 Substantive Changes: Substantive changes are changes in the original intent or safety significance of an action or event.

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 4

C.

d.

e.

f.

Semi-annual - interval not to exceed 7.5 months.

Quarterly - interval not to exceed 4 months.

Monthly - interval not to exceed 6 weeks.

Weekly - interval not to exceed 10 days.

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THIS PAGE 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 elemen lremperature that can be permitted with confidence that no damage to the fuel element cladding shall result.

Specifications. The temperature in a TRIGA fuel elenment slhillI not exceed 2,1000 F (1,1500 C) under any mode of operation.

Basis. The important parameter for a TRIGA",,reactor is the fuel element,temperature.

This parameter is well suited as a single specicficatlon specially since it cantiPJ5 measured. A loss of the integrity of the fuel elemenit claidingcould arise from a build-up of excessive pressure between the fuel-moderator ýtiid the cladding if the fuel temperature exceeds the safety limit.. The pressure is caosed by the presence of air, fission product gases, and hydrogen f6in~the dissociatioi fCthe hydrogen and zirconium in the fuel-moderator. The imaniituLde of this prevssure is determined by the fuel-moderator temperature and the ratioof hytdrogen to zirconium in the alloy.

The safety limit for theTRIeGAO fuel element is based on data which indicate that the stress in the claddi*nig due to th :hydrogen ptrssure from the dissociation of zirconium.

hydride will remainjbelow theultimate stress.provided the temperature of the fuel does not exceed 2100° F (115(1 lid the fuel cladding is water cooled (SAR 4.5.3.1).

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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 q&, less than 510°C (9501F) as measured in an instrumented fuel element. The instf eried fuel element shall be located in the B-ring.

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Basis. During steady state operation, temperatures were"calculate~dffor the beginning-of-life normal core. Linear extrapolation of tempe ature~and power indicates that an IFE power of 23.2 kW will produce an indicatedpObwer of 5100C in the IFF';a'tthe midplane thermocouple location. The highest5tio of maximum to minimumpower for elements in the B-ring was found to be 1.036, so ifthdIFE's generating,*23.2 kW, the maximum power in any B-ring element would be li'fited to 23.1 x 1.036 = 24.03 kW. For a power of 24.03 kW, the maximum temperatueranywhere in the hot channel fuel element will be 588°C. The accuracy*bIothe temperatul'indication is +/-5C..Even including the error in the temperature measuremient, this valfuis well below the Safety Limit.

AP 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 exceed 11 MW excepti for pulsing operations.

Basis. Thermal and hydraulic calculations indicat hat TRIGA fuel may býsafely operated up to power levels of at least 1.9 MW with nattrial convection cooling (SAR 4.5.3.3 and 4.5.3.3.9).

3.1.2 Shutdown Margin Applicability. These specifications apply to

• reactvltcoition of the reactor and the reactivity worths of

°Po'hictri.trods and experients. I,Iy apply for all modes of operation.

4::*

• i*

Objective. The objective Is to rssnre that the re.actor can be shut down at all times and to assure that-the fuel tmnperaitreat L saf'etyin1ifh shall not be exceeded.

Specifications. The r oshall not,*e operated unless the following conditions exist:

The shutd2_own margin pro\\ ded by control rods shall be greater than $0.55 with:

a.

irradiation facilities and experiments in place and the total worth of all non*s*cured 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 exceede~d.

Specifications. The maximum available excess reactivity ba&&on the reference core condition shall not exceed $8.55.

Basis. An excess reactivity limit of $7.55 allows fle&ibil l1 e.;tl-e, reactor in all ilo ty to oprtAh raorial core modes (NORMAL, ICIT and CLICIT) without the need to add or remo.ye uel A'.&

lS, elements when changing between operating mod'es. The NORMAL core i eost reactive core. If operating a NORMAL core withth e,minimu shutdown imrgln of

$0.55 and typical control rod worths of $2.70 (Safet)*2.270 (Shim), $2.70 (Regulating) and $4.00 (Transient) (section 4.2.2,:Control Rods), th&-lculate RC core excess i

would be -$0.55+$2.70+$2.70+$2.70 *j755. The stw iargn calculation assumes a) irradiation facilities and expeimes.iian place

1. &total worth of all non-secured experiments in their most reactive st teiemost ctive control rod fully-withdrawn and c) the reactor in the referefibe core conhition' Activities such as changing out of the NORMA.core, movln~i away fromnthe reference state or adding negative worth exp,,eriments wifmake core excess more negative and shutdown margin less positive. The MRnlyactivityWhich could rdsultn requiring fuel movement to meet shutdown margin and core exces§"mits wouldti the unusual activity of adding an experiment. wyith:ýlarge positiye reactivywirm.

3.1~Iiuse Mode Opeation I

Appati i

[is speciMtion applies to the energy generated in the reactor as a result of a pu1%isertion..f reactivity.

Objective. The objetive 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 Objective. The objective is to maintain integrityA'f Specifications. The reactor shall not operate witl*° purpose of locating damaged fuel elements. A fuel damaged and must be removed fromthe core if:

a.

The transverse bend exce s625 for the cladding;

b.

Its h1 inches oW'eiTli'*6 length of the by Q*125 inches; by release of fission products; or

ross pitting, or corrosion.

C.

al deterioration of the fuel is sufficient to warrant The elongation and bend limits are the values found

.1537).

3.2 Reactor System 3.2.1 Control 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 whichlifWl1 be available to the reactor operator during reactor operation.

Objective. The objective is to specify the mimum nu*b~r of measuring -l els that shall be available to the operator to assure safe Specifications. The reactor shall notube operate minimum number of measuring chanfiM&*led the 12

Table 1 - Minimum Measuring Channels Measuring Channel Effective Mode S.-S.

Pulse S.-W.

Fuel Element Temperature 1

Linear Power Level 1

1 Log Power Level 1

Power Level 2

2 Nvt-Circuit (1)

Any single Linear Power Level, Log*,jber Level or Power Levl,.,neasuring channel may be inoperable while the reactor is operating for the pui'pose of performing a channel check, test, or caliboratlnV*

(2)

If any required measuring,cAfels becomes in peable while the reactor is operating for reasons other tfla1i'dugidentfied in Technical Specification 3.2.2 (1) above, the channel shall bewestorid~tooeratin within 5 minutes or the reactor shall be immediately shatdown,'F Basis. Fuel temperatgiie dis'pHitred at the control console gives continuous information on this parameterv'0ith has a spkcified safetýlimit. The power level monitors assure that the reactor power kvl is Wýdeqately monif: red for both steady-state, squarewave and pulse modes..of op#cfi-ýion'

,i~ficatilons on reactor power level indication are include his 4:ion, si*;:;:;

ncludeifi*.*tl1l's sec

ýnslnce: the power'6Ve'l is related to the fuel temperature. For footno "'(1), taking the.

measuring channels off-line for short durations for the purpose of ck as i s kIdered acceptable because in some cases, the reactor ofa,~ctest or cai t

si

ý cons, must beýo.erating in orde IVo perform the check, test, or calibration. Additionally there exist two Pietnndant power: I evel indications operating at any given time while the third single channd'3l{sfoff-line.

e4Obr footnote (2), events which lead to these circumstances are self-revealing tdifhepoperator. Furthermore, recognition of appropriate action on the part of the operator as a{resilt 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.

ThhI~ 2 - Minimum Re2etnr Safety Channek Effective Mode Safety I)Channel FunEction S.-S.* :Pulse, S§W.

Fuel Element SCRAM @ 510-C Temperature Power Level SCRAM @ 1.1 MW(t) or less 2

Console Scram SCRAM 1

Button

./

Preset Timer Transient rod SCRAM

'15.

se after a pulse High Voltage SCRAM @ Ž!25% of uoidhinal

, 1 1

operating voltage 1

Table 3 - Minimum Interlocks ?:

Effective Mode Interlock Function S.

Pulse.

S.W.

W ide-R ange Log}..

iits-con ro r w

a Power Level

Prevents control rod.ithdrawal 1

Cn),.@

less§'than 2 cps Channel 7'*!"........

TransientRod Pr~events applicati on air unless 1kW Pulse r

I t7i,rock

• irevents pulsing above 1 kW 1

Shim, Siafety, and Reg LiIating Preents simultaneous manual Rod Dri't*,

withdrawal of two rods Circuit-Shim, Safety, and Regulating Prevents movement of any rod Rod Drive except transient rod Circuit Transient Rod Prevents pulse insertion of Cylinder-Psitonde 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 elemnt te perature scradm causes a scram in excess of the LSSS, which is 510'C. Th8esupporing arguments for the safely limit of 1150'C are given in SAR 4.5.3.1. The4WS'SS is setito less than half, f e safety limit. This is more than adequate to account fnr Mramtles'In instrument, esponse and core position of the instrumented fuel element.

Power Level Scram: The set point ftoidZthh the safety at ow channels are normally set to 106% of 1 MW(t), whi i*,%,bw the licer of 1.1 MW(t).

The 6% difference allows for expected And obs*.;

strume ctuations at the normal full operating power of 1 MW(t) t6o' "r with o rming the reactor unnecessarily. ConverseY,1SNR 13 2 2 2 2*shows that his t point is more than sufficient to preve ceedigithe reactivity insertion limit during non-pulsing operations and prevertfthe operator from inad-Verently exceeding the licensed power.

Manual ScramcraThe mus, unctional at all times the reactor is in operatlor`ý It has'no**:*spcfie Vd*alue for a scram set point. It is initiated by the reactor operator manually. "..

Preset Timer Scram: Tlhe*1eseto 1imer ensures that the reactor power level will reduce to a low level`safter pulsing;,dd preclude an unintentional restart or ramped increase to some equilibi*,ow, i.ower.

• High Voltage Scram:* The high voltage scram must be set to initiate a scram before the high voltage for an/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%,perator from pulsing the reactor in steady-state mode.

1 kW Pulse Interlock: The 1-kW permissive interlocki ilesi,&d to prevent pulsing when wide range log power is above 1 -kW. SAR 13.2.2.1' 1 show'stiat the peak temperature reached during an end-of-life core willl,50°C for ani itial fuel temperature of 20'C. The methodology clearly,sh*5ws that if the initial tLemiperature was higher, the resulting peak temperature must le&rfower, HAowever, there has*tsbeen analysis or experiment to look at the relationship bet.eeJeat,,,enerated within the fuel at power (i.e., > 1-kW) and heat generated on the su-fac.of the fuel during a pulse.

Therefore, this interlock prevents the.reactor from pulsingat power levels which produce measurably significant incre '.0infuel temperatur

\\.

.l Shim, Safety and Regulating Rod Drive Circuit:tH he single rod withdrawal interlock prevents the operator from removing multiple cntrol'rO9iSmultaneously such that reactivity insertions froffont gol rod manipl5Iation is done in a controlled manner. The analysis in SAR 13?*2.2 a'fd0312.2.2.3 show that the reactivity insertion due to the removal rate of the&miost reactivorod or all thebcontrol rods simultaneously is still well below the reactivity insertion desIgn.*limit of 011

Shim, et ntlR ulanng ltoa Drive Circuit: In pulse mode, it is necessary to imit.itle reactivity isered to lessithan the design limit of $2.59 at the end of core life Wi SAR 13.2.. 2 his 1 rlock ensures that all pulse reactivity is due to only te sient rod whilcIdn pulge mode. Otherwise, any control rod removal in pulse mode woulI*atd to the inserted reactivity of the transient rod and create an opportunity for exceeding te*reactlVlt"Insertion limit.

Transient Rod Cylinder Position Interlock: For the transient rod cylinder interlock, SAR 13.2.2.2.1 sho&vs 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 a* adecquat amount of waLiT 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 i-ractor tank integrity.

Specifications. The reactor primary water shall,;exhibit the following parameters:

a.

The tank water level shall be gra tr h

.14.eet above the top of the core;

b.

The bIuL:k tank water temperatu e shall be less than 120'F (49'C); and

c..The condcRtIcIty% of tlief t1ikater shall be less than 5 i[mhos/cm.

Bas**. The minimurn ihight of'1,feet of water above the top of the core guarantees that there l sufficient waterfor effectiv \\e ;ooling 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 'atiiany research reactor facilities has shown that maintaining the conductivity within the specified limit provides acceptable control of corrosion (NUREG-1537).

y 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 tinie not to exceed two (2) hours to permit repair, maintenance or testni1g o rteventilation system.

b.

The ventilation system shall be shutdownupon a higha*a)1ivity alarm from the Exhaust Particulate Rad ition Monitor Basis. During normal operation of the ventilationsysfn the annun concentration of 41Ar in unrestricted areas is well belu6'it applicable effluent concentration limit in 10 CFR 20. In. addition, the worst case maximum total effective dose equivalent is well below the appljcahlannual limit for individual members of the public. This has been shown to be trueWfPOr.scenarlos where tl4\\,Atilation system continues to operate during the maximurinhyp\\

Iotti alaccident (MHA), where the ventilation system is secured during the M1,A,,and wner'dthe ventilation system and the confinement buililngarenotpresent durin We MHA (SAR 13.2.1). Therefore, operation of the re reac1t'(r for shiP*preriods while, the ventilation system is shut down for repair or testing does not compromise the coniil roover the release of radioactive material to the unrestricted area r should it cause occupational doses that exceed those limits given in 10, CFR ?0 (SAR'l 1..1 2).

lie&kwo hour exception to permit repair, maintenanc or tesgfin should not diminish the effectiveness of the reactor top area radiationi monitor or thecontinuo~is air Nparticulate radiation monitor. The sampling locartionsfor both of these monitors ae located directly above the core. Any fission product rellease should bedetected in the same manner regardless of the status of the ventilation system because of the close proximity of the sampling point to the source term; Moreover ', adiationii-onitors in the building, independent of the ventilation

system, lgivewarning of high levels of radiation that might occur during operation of the reactor while h n'ventilation system is secured (SAR 11.1.4.2). The exhaust gas and particulate radiat{ion 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 radiatio, Nonitoring channels that shall be available to the operator to assure safe operationof theeactor.

Specifications. The reactor shall not be operated unmessthe mlinmum number of radiation monitoring channels listed in Table 4 are *operating.

Table 4 - Minimum Radiation Mvlonitornfi4ngChannels Radiation Monitoring Chnnels iNumber Reactor Top Area Radiation ohnitor 1

Continuous Air Particulate °adlauMnýtomr, 1

Exhaust iGaRatlation MonitV 1

ExhsORT~articuk u ie*Radiation MOnitor 1

Exception: Whealn*sidg relqutired:radiation monitoring channel becomes injoperabe.operations,'may continue only if portable instruments, surveys, or

/analyses maý 1i:substitteld for the normally installed monitor within one (1) l.*iohour of discove or peril"t01not to exceed one (1) month.

Basis. ThIeiradiation moi trs provide information to operating personnel regarding routine releas es radioactivlity and any impending or existing danger from radiation.

• 'V:XPý,

.4-ýW*

Their operation'wi,:,.provi'de sufficient time to evacuate the facility or take the necessary steps to prevent the$pr, of radioactivity to the surroundings. Furthermore, calculations show ttg 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 41Ar.

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

Specifications. The annual average concentration of 4 1Ar discharged into the unrestricted area shall not exceed 4 x 10-6 MCi/ml at the point of di'Scharge.

Basis. If 41Ar is continuously discharged at 2.5 x 106Cthe concentration produced when the nitrogen purge of the rotating rack is~off, aI lXN lxes on the argon manifold are open, and all beam port valves are open)m easurem ens*and calculations show that Ar released to the unrestricted areas un me worstes a

conditions would result in an annual TEDE of 5 mrem (SAWR1 1.1.1.1..1). This is of the applicable limit of 10 mrem (Regulatory Guidel4j20). Th6refore, an emissioni6f 4 x 10 6 kCi/ml would correspond to an annual TEDE of*8fhremwicjih is still 20%Cbelow the applicable limit.

3.8 Limitations on Experiments

~l'l 3.8.1 Reactivity Limits Applcabilty. This applies to eperiments "istalled in the reactor and its Appicailiy.

hissp6elircationaple-i irradiation facilitiesA.

Objective. The objectiV s

.teptdamag¶ to the reactor or excessive release of radioactive,,m':tbrials in then of e

df riinent failure.

Specitmcations. The rettor shall*lnot,,be operated unless the following conditions governing experiments exst:

a.

absolutevalue of the reactivity worth of any single unsecured exjerimenifhall be less than $0.50; and

b.

The usmiin of the absolute values of the reactivity worths of all experiments shallfbe 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 nstMalled Mi the reactor and its irradiation facilities.

Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment fa ilurIL1I.e 4

Specifications. The reactor shall not+e operated unlesS'thoe:following conditions governing experiments exist:

a.

Explosive materials, such ~as gunpowdL1 er, TNT, nitroglycerin, or PETN, in quantities greater than 25 milligrams slhallnot be irradiated in the reactor or irradiation facilities. ExplO*sive materials in quantities less than 25 milhgramls ma b*e irradiated p rovided the pressure produced upon detonationof the explosive has been calculated and/or experimentally demonstrated to be less than half the design pressure of the container.;

1.

Experaits containingccorrosive materials shall be doubly encapsulated.

The fail,.ure:of an ene apsulation of material that could damage the reactor

\\:.

shall resufi'in removal of the sample and physical inspection of potentially damaged components.

Basis. This specifi-atiorii is intended to prevent damage to reactor components resulting from failure of an e\\eriment 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 or 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 r4doactivity in the reactor bay or the unrestricted area will not result in exceedingAhe applicable dose limits in 10 CFR 20, assuming that:

a.

100% of the gases or aerosols escape fro)

riment,

b.

If the effluent from an irradiation:facility exhausts througlMholdup tank which closes automatically on~iyhi radiatio'hilevel, at least 1%'6f the gaseous activity or aerosols produce1will escape;

c.

If the effluent from an.irradiation facilityhausts through a filter installation designed 004,greater than 99% efficlenc for 0.3 micron particles, at least 10% o'ftl69e-Eerosols can escape and

d.

For materials whose boilinpoint is aOoF)etI30F and where vapors formeda *y foing this materi'hfcan s ly through an undisturbed column of wate'above the core, 10% of these vapors can escape.

Basis. ThisýspecIfcato isiitendg6t to'ahNithe purpose of 10 CFR 20 by reducing the likelihood, fat rhaseo.airboneradioactivity to the reactor bay or unrestricted area surrodfnding t illresult "

su ui the OSTiI rl ineceeding the total dose limits to an individual as speciffead'in 10 CFR2'"~li**

3.9 Thi§SSection intentionally left blank.

ýMY*.,

• 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.

4 i

a.

Surveillance requirements may be defbed during reactor shutdown; however, they shall be completed prior to reactor startuphless reactor operation is required for performance of the surveillance. S 0

surveillance shall be performd asoon s tcable after rector startup. Scheduled surveillance, whic c-afrmot performedi the reactor operating, may be deferred until,a i.anned reactor shutdown.

b.

Any additions, modificautoni sor, maintenanceth veene tilation system, the core and its associated Sfu t

r tnhepool or its penetrations, the pool coolant system, th roddvv.ieV&ichanism or the reactor safety system shdllhe,-made and tested in accord'iace with the specifications to whichWthe syseims were originally designed and fabricated or to spe6fitations reviewed by the Reactor Operations Committee. A system shall no-%ee,,confered operabliwinil after it is successfully tested.

Basis. Thisspecrfiaton rmat* ao change sin reactor systems which could directly affect the saf6ty of the rea-i As Iongas changes or replacements to these systems continue to mnketthe original desiPýspc sthen it can be assumed that they meet the N..

P,11.*

presenylak.cepted operating criteria.

4.1 Reactor 6-0ie Param£e>rs Applicability. Thls'RspcIfication 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 determin edIanally or following any significant change (>$0.25) from a refere~?e core.

e.

Twenty percent of the fuel elements comprising the corsh'all be inspected visually for damage orl'eterioration and measure:tfor concentric or other swelling anuniAIlI{ysuch,,.

tdtthe entire core&is inspected over a five year period. Ar ntiatispections shall be of non-repeating representative samples of fuelOeTements from each ring.

Basis. Experience has shown that the i '* ti..th:ed frequiencle 'wi*01fsure performance and operability for each of these systemsqpr components. Th&value of a significant change in reactivity (>$0.25) is measural Vt coverage of the shutdown margin afterlaki~igg into account tlhiccumuaton of poisons. For inspection, looking at fuel elementfs fron aich ring annully will identify any developing fuel integrity issues throughout the coure. Furthermbre, the observed mechanism for non-conforming fuel at theSTR has been exclusi.ely concentric swell. Looking for swell W

will not onlyprovide ead 'indictiii O l

f ibn-conformance but it will significantly:

reduce th&

,,fuiel movementsn 666d.

4.2, ea'ctor Control a "Safety ms Applicabilit*V.. This speciffiation applies to the surveillance requirements of reactor control and saety.. systems-,*.,

Objective. The objlectiv:e 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 II in section 3 semi-annually.

f.

A channel calibration of the fuel temperature Ueas performed annually.

Basis. Experience has shown that the identified frequencies will ei operability for each of these systems or components.

4.3 Reactor Primary Tank Water Applicability. This specification apples&,to the surveillaftcerequn tank water.

shall be performed ga channel shall be and for the reactor

.afer level and the bulk water appropriate alarm settings.

temperature a.

water level monitor shall be

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

c.

Aop erabillty check of the reactor tank temperature alarm shall be perform 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.<of tle:,entilation system in controlling releases of radioactive material to the unrestfi'ýied are" ý te aea, I, Specifications.

INN,

a.

A channel check of the reactor ba\\ confinement ventilation sm's ability to maintain a negative pressurecr*fitleractor bay witl respect to surrounding areas shall be performed pi:to each day's operation or prior to each operatioin:,,ex tending more than ione day.

b.

A channel test of the reactor bay'%confinement yentilation system's ability to be secured shall be perioecan uall(y.,,

Basis. Experience has~demoiiated tat tests, of te veniilation system on the prescribed daily atidhnual ba'is are sufficientto assure proper operation of the system and its control over releases off dioactive mateiMal.

4.6 This.secti.n enni

• ,.I blan.-

4.*Ra'diation Monitoing System Applicabili-ty jhis specificdtion applies to the surveillance requirements for the area radiation montqring equipment and the air monitoring systems.

Objective. The object$i.hve 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 th, experiments installed in the reactor and its irrn Objective. The objective is to prevent the con the reactor or release excessive amounts of ra(

experiment failure.

Specifications.

a.

The reactivity worth of an ' xp(

%a.

appropria~~~before reactor ope

b.

Anepe, exriment shall not be insi facilite'.s

anls safety analys omnplian n *Ini.*,bt

. -aeco 1

tlon 6.2.3 'fh" procedae ch*W*,are establish, Basis. '*"Eperience has sl]In thatexperiment OSTR an &lieReactor Op ations Committee safety of the etor or exceeding the limits in 4.9 This se5t d

ionmentionallv left blank.

for as experiment.

led in the reactor or its irradiation has been performed and reviewed for he Reactor Operations Committee in full Technical Specifications, and the for this purpose.

s which are reviewed by the staff of the can be conducted without endangering the the Technical Specifications.

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 fence surro lng the reactor building and the reactor building itselifEhe Ounres icearea is that area outside the fence surrounding the r eactorD.

bui.d rectruidig'i.

b.

The reactor building houses the TRIGA reactor and is abutted,%6 the Oregon State University Radiation Cc i fteBuilding.

c.

The reactor bay shall bc equipped with vent itlion systems designed to exhaust air or other gases,,frioj) the reactor buin&'gind release them from a stack at a minimum of 65 feet frt 1f ground leyel.

d.

Emergencyhutdowncontro1s!f~

r the veirfilation systems shall be located.

in the,'ieactor-control room. -1%

Basis. The Radiation @*nter, reactor building a site description are strictly defined (SAR 2.0). The facility~i" l

geOl suhat.tlfie ventilation system will normally maintain Reactor tBuilding with respect to the outside atmospjere so that the*eill be no uncontrolled leakage to the unrestricted enyi.roinient. Controls'4W.r startup, normal operation of the ventilation system are located if.Ithe reactor control room. Proper handling of airborne radioactive materials.

(in emergge'situations) be conducted from the reactor control room with a minimum of'exPosure t operating personnel (SAR 9.1 and 13.2. 1).

5.2 Reactor Cool'af$ystem 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 pro, *,d'ddeto indicate high bulk water temperature if the temperature exceeds 120°F (49 °C).

Basis.

V

a.

This specification is based on thermal and hydraulic calhuations which show that the TRIGA core canikperate ina safe manner aPD r e up to 1.9 MW with natural coetl flowofthecoolant wtr' (SAR 4.5.3.3).

ON

b.

In the event of accidenftalsiphoning of tan 8.vater rough inlet and outlet pipes of the heat exchanef i@

demineralizer the tank water level will drop to a level no 8

,fet from thp of the core (SAR 5.2).

.1'

• 4

c.

Loss-odf coolfanT'&arm causediby a water level drop of no more than 6

inchles, irowdes a:titmely warnlrg, so that corrective action can be initiate'&{Uf his alrnm is located 1he control room (SAR 5.2).

d.T watefi1:emperatri ealarm provides warning so that corrective

• actlon,,Eh be initiated in a timely manner to protect the quality of the reactor tafk The"Jiffl is located in the control room (SAR 7.2.3.2).

5.3 Reactor oGre and FueIl 5.3.1 Reactor 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*;'.c ing I kW with a core lattice position water filled, except for positions on the periphery of the core assembly.

d.

The reflector, excluding e) water or a combination of Basis.

a.

Only TRIGA fuel is antic

b.

In-core water-filled expeir safe in the Gulf Mark 1IHI4ýf be experienced in hydroge non-hydrogenous experin, demonstradt1j:to be safe in operatidh (SXAI<,42).

c.

For cases'where.ýone in-coi erpe;, *te-imaxi mumingr(

z, -sagpeak powerggenerat i I 4.2 The core be assemble

,.tank of lign,,.ater. Water 1Used for nep'ron economy i*adihtionr.e urements (Sý 5.3.2 Control Rods, ition tci:lities, shall be (SAR 4.2).

been demonstrated to be 1ues of flux peaking will

ýn positions. Various ment positions have been cores up to 2-MW "A'15n is water filled, except in the core r power level is reduced to 1 kW to ensure Als in adjacent element positions (SAR d in the reactor grid plate which is located in a in combination with graphite reflectors can be and the enhancement of irradiation facility R 4.2).

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, B4C 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, 6oh' as a poison in an aluminum or stainless steel cladding. The transient rod shall have an adjustable upper limit to allow a variation of4edCtivity insertions. This rod may incorporate an aluminum-or airffillower Basis. The poison requirements for the control rod§ are satisfied by usnneutron absorbing borated graphite, B4C powder or bormo as its compounds. Thes, iaterials must be contained in a suitable clad material'suclh ýts aluminum or stainles stee to ensure mechanical stability during movement and tLo i 1te the'poison fro tank water environment. Control rods that are fuel-followe'iV: Vpvide additional reactivity to the core and increase the worth of the.control rod. The us, &f fueled-followers in the fueled region has the additional advantagec o I-freducing flux peking in the water-filled regions vacated by the withdrawal of theicontrol -Ads.

Scram capabilities are provided for rapid insertion of the control rods whibh is the pritn&M*safety feature of the reactor.

The transient control rodciPes&gned for rapidwithdrawaldfrom the reactor core which results in a reactor p.ulse. Tleticlear behavior of the air-or aluminum-follower, which may be incorporat~hto the tran'sient is simlar to a void. A voided-follower may be required in certain coreoadthgs to reduce u1]x peaking values (SAR 4.2.2).

5.3.3 Reactoie F,.uel AMAMpIbility. This specficatio. apples to the fuel elements used in the reactor core.

Obiective.Fhe objective ito assure that the fuel elements are of such a design and fabricated in sUh a manneras to permit their use with a high degree of reliability with respect to their pysical.and nuclear characteristics.

Specifications. TRIG"A 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 elem* ent is about 6% greater than the design value of 8.5 wt%. Such an increase in loading would2,result in an increase in power density of about 2%. Similarly, a minimunirbium content of 1.1 % in an element is about 30% less than the design value. This niatl*iwould result in an increase in power density of only about 6%. An increasein localIoer density of 6%

reduces the safety margin by, at most, 10%. The maximum, hydrogeni t(zirconium ratio of 1.65 could result in a maximum stress under accident conditions in thefuel element cladding of about a factor of two greater than thl i calue resulting from a hydirogen-to-zirconium ratio of 1.60. However, this increase 1Pthe cladding stress duringa accident would not exceed the rupture strength of the claddingh(S'i R 4.2. 1).

5.4 Fuel Storage Applicability. This specification appliesto tlie*morage of reactfr fuel at times when it is not in the reactor core.

>:'?':,A",*

Objective. The object,

Is toassure that fuel, which is being stored shall not become critical and shall not reach an unsafe temperature.

Specifications.

1a,.

Allfuet Ie:elements sshall be stored in a geometrical array where the k-

effectve s,ý less thanl 0.9,forQ,,

all conditions of moderation.

Irradiated. fuel elements and fuel devices shall be stored in an array which SWill permit sufficient natural convection cooling by water or air such that t lie temperature of the fuel element or fueled device will not exceed the satf n lmit*

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

32

THIS PAGE 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 Universi.ty ad USNRC structure as shown in Figure 1.

K 11 D11

1:1:-.

The following specific organizational levels, and r:esponsibilities shall exist:.<

a.

Vice-President for Research (Level 1): Tbhe Vice-President for Research is responsible for university center and institute organizations representing Oregon Stat~e University.

b.

Radiation Center Director (Levell'2)?:The Riadlation Center Director reports tt

.i*e Vlce-Presidenn*f6r Researci mand is accountable for ensuring that all, regulatory requirements, including implementation, are in accordance withiAll requirements of the USNRC and the Code of Federal Regulationr 3

c.

l.R*40ector Admiistrato(e 3): The Reactor Administrator reports to the lRhdation Center Director and is responsible for guidance, oversight, and technical suppo'i-tof reactor operations.

d.

Senior Health Physicist (Level 3): The Senior Health Physicist reports to

ýthe Radiation Center Director and is responsible for directing the actyities ofnhealth physics personnel including implementation of the radiation safety 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 IReactor Supervisor, the Reactor Supervisor shall be readily a"vailable on call. Readily available on call means an individa twh

1.

Has been specificall\\ designated and the designation is known to the opea tor on duty; ii.

Can be rapidly conta c b\\ phone by the operator on duty; and iii.

Is capable o t*ogeting to the reactor :facility within a reasonabletiIme under normal conutitions (e.g., 30 minutes or within a! 15-1m1e r-mdiu,).,

b.

A listof reactor facility personnel by name and telephone number shall be, ireadlly avai a11e in the control room for use by the operator. The list shall be update(ainnually and inTchde:

S1.

ý iiRadiation C6giier Director 2

1 ReaIc to r Administrator

3.

Senior Health Physicist 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 and audit of the safety aspects of reactor facility of the ROC shall be presented to Level 1 and he*

days of completion.

6.2.1 ROC Composition and Qualifications An ROC of at least five (5) members knw`e, ge, engineering and nuclear safety shall reveiw andN with the operation and use of the facility.,The R management.

6.2.2 ROC Charterdand Rules The operations of the ROcI'shGalI m11e ccor*....

or reports

)idh relate to reactor aspects associated nted by Level 1 with a written charter including P11 C.

!(' members, no more than two voting members may be of the 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 speciic.*tions, license or charter;

d.

Review of violations of technical specifications, llices or violations of internal procedures or instructions havifig safety significance;

e.

Review of operating abnormalities having Safety significanicý::;

f.

Review of all events from reports requIred ina sections 6.6.1 and 6.7.2 of these Technical Specifications;

.g.

Review of audit reports.

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

The annual audit shall incLude AtC0least the following:

a.

N:Reacto-r operating record*s;

,,b.

Operat, Req,,,uahlfi c1ton Program;

c.

%(Reportable occurrences; and

d.

Eiergencyi 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".

6.4 Procedures 38

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 the reactor;

c.

Maintenance of major components of systems thai could have an effect on reactor safety;

d.

Surveillance checks, calibrations, and lnsp ons re'l-red by the technical specifications or those thatla, n effect oneactor safety;

e.

Radiation protection;

f.

Administrative controls for operatioh~dns).i~

ainenance and for the conduct of irradiations and experiment"+fisiat could affect reactor safety or core reactivity;.

g.

Shipping of radioactive m'atef

,l.

h.

Implementiftibn. of the Emergeincy Response Plan.

Substantive changes to aove procedures shal be made only after review by the ROC. Except,for raia on ecTonoceures unsubstantive changes shall be approve'n

'.lm n

by the Raector Administrator and documented by the Reacto Q ministrai."'WlthIn l ays of implementation. Unsubstantive changes to radiffiih, protection prdc"lures stifl:lNl* approved prior to implementation by the SHP and do `Uf',ented by the Sdnior Hel'th Physicist within 120 days of implementation.

Temporary &VPae.tions fror.. he procedures may be made by the responsible senior reactor operatorIEnl.ordeor,f deal with special or unusual circumstances or conditions.

Such deviations s'f.,e 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:

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 39

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 0y,,

In the event a safety limit (fuel temperature) is exceeded:i d.."

a.

The reactor shall be shut down andalýeactor operation sh'al"4,hot be resumed until authorized by the USNRC*A,,I

,in"'

ý$

b.

An immediate notification of the occu.rentc shall be made to the Reactor Administrator, Radiation Center Directoi nd Chairperson, ROC; and

c.

A report, and any applicabldollo rt o

Wal.,.b prepared and reviewed by the ROC. The repo. shall descri l*he following:

1.

appdlitable circumstafice leading te violation including, when known H.the cause and ontributin factors;

2.

effects odhe violation ulon reactor facility components, systems, ori'sttu& resianadn, th&health and safety of personnel and the

3.

corective aclon to be taken to prevent recurrence.

6.6.2 Action tjo*Be Takeiiin the Event of an Occurrence of the Type Identified in Section 6.7.2 Oth;e, thani. a Safety Limit Violation For all events which',"dre 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;

b.

operations shall not resume unless authorized by the Reactor Administrator and Director; 40

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 Reports0 6.7.1 Annual Operating Report An annual report shall be created and submitted by the Rdiatiorn eter Director to the USNRC by November 1 of each year consisting of.

a.

a brief summary of operating exIperience including the ene.*prduced by the reactor and the hours the reactor was*,ritical;

b.

the number of unplanned shutdowns, incId;ing reasons therefore;

c.

a tabulation of major pr'enýative and correcn, intenance operations having safety significance;

d.

a brief degdiftion, includin a%,'ummary\\*fthe safety evaluations, of chane in ttin60 hility or in procedures of tests and experiments out pursuant to 10 CFR%ý505 9;

e.

,a

,summar f,, ealig&uanda,[6ount of radioactive effluents released or

,:diA:

iarged ttho-eenvirons beyond the effective control of the licensee as mear&>at or pfrtrto the point of such release or discharge. The summa~pall inry'U'-"t the extent practicable an estimate of individual radionuclides preseit in the effluent. If the estimated average release "4,after dilutido*for diffusion is less than 25 percent of the concentration

,allowed or recommended, a statement to this effect is sufficient;

f.

a summarized result of environmental surveys performed outside the facilify; and

g.

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

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:

41

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 of any of the following:

1.

violation of the safety limit;

2.

release of radioactivity from the site aboe allowed limits;

3.

operation with actual safety systef 1settings.foim required systems less conservative than the lim 'qg

':fety syste.

-settmg;

4.

operation in violation oimiting conditions for opefationg:unless prompt remedial actio'n istalen as permitted in Section"-3;

5.

a reactor safety system componeinPmalfunction that renders or could render tlieeactor safety system~incapable of performing its intended safety tfunc6tibn If the malfin>tbionor condition is caused by mainteBancethe".

o report "required;

6.

afi.ilqnticipated or nrole ange in reactivity greater than

'one dMIRa.

Reactor trips resulting from a known cause are

ý: cuded:'

-7.

abnlotmaa ahnsnrificanit degradation in reactor fuel or cladding, N or b boouidary, or containment boundary (excluding

• k, '**where applicable; or

8.

an.iNbservedi quacy in the implementation of administrative or prpcedural controls such that the inadequacy causes or could have 'caused the existence or development of an unsafe condition "t

witf regard to reactor operations.

b.

a rep.- rt 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

2.

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

42

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. bt'he Technical Specificaios,.

e.

reactor facility radiation and contamniiatio0,surveys;

f.

experiments performe dwith the reactor;

g.

fuel inventories, receipts, andSpimiients;:

h.

approved changes to the operatig procedures; and i e~aacor,,perations (Committ eeetngs and audit reports.

6.8.2 Recordsto be efalný*Onc Certification Cycle

,*'Records of retralning ar1iýrequalification of certified reactor operators and senior AOxl,ý'k*reactor h IranI be retifed at all times the individual is employed or until l-.ýcertification ignewedi 6.8.3 Records to be Retained for the Lifetime of the Reactor Facility

a.

gaseQLs 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.

43

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