ML20125C564

Amend 4 to License R-106,to Permit Replacing & Upgrading of Electronics in Reactor Console,To Increase Operating Limits for Reactivity & to Change Title of Principal Officers

ML20125C564
Person / Time
Site:	Oregon State University
Issue date:	12/18/1979
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20125C563	List: ML20125C558 ML20125C564 ML20125C565
References
R-106-A-004, R-106-A-4, NUDOCS 8001100412
Download: ML20125C564 (11)

Text

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/[ UNITED STATES y .e g NUCLEAR REGULATORY COMMISSION j WASHINGTO N, D. C. 20665 s,

% *****/ OREGON' STATE UNIVERSITY DOCKET NO. 50-243 AMENDMENT TO FACILITY OPERATING LICENSE

Amendment ho. 4 l License No. R-106 i

j 1. The Nuclear Regulatory Comission (the Commission) has found that:

A. The application for amendment by Oregon State University (the licensee) dated April 16, 1979, as supplemented July 11, 1979,

, August 17, 1979, and October 10, 1979 complies with the stand- -

ards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities au-thorized by this amendment can be conducted without endan- ,

gering the health and safety of the public, and (ii) c. hat such activities will be conducted in compliance with the Lonrnis-sion's regulations; D. The issuance of this amendment will not be inimical to the

, common defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied; and F. Publication of notice of this amendment is not required since it does not involve a significant hazards consideration nor amendment of a license of the type described in 10 CFR Section 2.106 (a)(2).

4 90008229 8 001100 41 U

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2. Accordingly, the license is amended by changes to the Technical l Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License  !

No. R-106 is hereby amended to read as follows: ,

(2) Technical Specifications The Technical Specifications contained in Appendix A, as i revised through Amendment No. 4, are hereby incorporated I in the license. The licensee shall operate the facility '

in accordance with the Technical Specifications.

3. This license amendment is. effective as of the date of its issuance. -

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FOR THE NUCLEAR REGULATORY COMMISSION r O ~

, j a k' k .Wk Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the ,

Technical Specifications Date of Issuance: December 18, 1979 90008230 1

ATTACHMENT TO LICENSE AMENDMENT NO. 4 FACILITY OPERATING LICENSE NO. R-106 DOCKET NO. 50-243 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Pages .

9 - 12 -

15 24 i 30  ;

I 35 i

90008231 l

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3.3 PULSE MODE OPERATION i_,

Acolicabilitv. This specification applies to the er.ergy generated in tne reactor as a result of a pulse insei tion of reactivity.

Objective. The objective is to assure that the fuel temperature safety limit will not be exceeded.

Soecification. The reactivity to be inserted for pulse ::eration shall be cetermined and limited by a mechanical block and electrical interleck on the pulse rod, such that the reactivity insertion will not exceed 2.55 dollars. l Basis. The fuel temperature rise curing a pulse transient has been estimated conservatively by adiabatic models. This model accurately -

predicts pulse characteristics measured in an existing core of all Standard fuel. pulse characteristics for operatior.ai mixed cores and FLIP cores thus may be se'timated with confidence, relying also on infonnation concerning pro:pt neutron life time and prompt .

temperature feedback of reactivity. These parameters nave been .

established for mixed and full FLIP cores by calculations and have been confinr.ed in parts by measurement at existing facilities. In addition, the calculations rely on flux profiles and correspending power densities which have been calculated for a variety of opera-tional mixed and full FLIP cores in SAR I.

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In th3s manner it is estimated conservatively that reactivity up to 2.55 dollars in operational cores will produce pulse transients with maximum fuel temperatures no greater than 950*C in FLIF fuel and

@00*C in standard fuel; i.e., a _ safety margin of 200*C with_re.spect to the safety limit of the fuel is maintained in either case,' allowing fW' ~

any uncertiintibs in measurements and/or calculations.~

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3.4 CORE CONFIGURATION LIMITATI'ONS Aeolicabilitv. This specification applies to mixed cores of FLIP and Stancare types of fuel.

Objective . The objective is to assure that the fuel temoerature .

safety limit will not be exceeded due .to power peaking effects in a mixed core.

Soecification. The FLIP fueled region in an operational enra shall contain at least 80 FLIP fuel eierents in a contiaucus block of fuel in the central region of the reactor core. Single element 3:sitiens may be i left vacant or occupied by trer items as specified in Sec-ica 5.2.c '

and 5.2.d of tnese Technicai 5:etifications. '

Sasis. The limitation on a'.lcable core configurati r.s to th:se simi-iar to the enes consiceret r. Si::tions 3. and 5. of SA . : lirits power d

90008232

.g. 2_7. . # - . Sc. g ,4

peaking effects. ~The limitation oipower peaking effects insures that the fuel temoerature safety limit will not be exceeded in an operational core.

3. 5 CONTROL AND SAFETY SYSTEM 3.5.1 Scram Time =

Accli cability. This. specification applies to the time required for the scranrcaole control rods to be fully inserted from the instant '

that a safety channel variable reaches the Safety' System setting.

Obj ective. The objective is to achieve prompt shutdown of the reactor '

to prevent fuel damage.

Soecification. The scram time, measured from the instant the input' signal reacnes the value of the Safety System setting to the instant

- that the slowest scrannable control rod reaches its fully inserted .

position shall not exceed 2 seconds.

Bases. This specification assures that the reactor will be promptly l snutdown when a scram signal is initiated. Experience and analysis have 1ndicated that for the range of transients anticipated for a l TRIGA reactor, the specified scram time is adequate to assure the l

safety of the reactor. '

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3.5.2 Reactor Control System j Aeolicability. This specification applies to the information which l must ce available to the reactor operator during reactor operation.

Obiettive. The objective is to require that sufficient infonnation is available to the operator to assure safe operation of the reactor.

Scecification. The reactor shall not be operated in the specified i moce unless tne measuring channels listed in the following table are i l

operable.

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t Eff ective Mode Measurino Channel i S.S. < Pulse i S.W.

Fuel Element Temperature X X X X

Linear Power Level X Log Power Level X X X

i Percent Power Level l X

l Nyt-Ci rcui t - X i '

1 Peri od-Ci rcuit ,

X i X i Safety Power Level X 90008233 f, A=end: enc No. . 2; 4 l

Bases. Fuel temperature displayed at the control console gives con-t1nuous information on this paramater which has a specified safety limit. The power level monitors assure that the reactor power level is adequately monitored for both steady-state and pulsing modes of operation. The specifications on reactor power level indication are included in this section, since the power level is related to the s fuel temperature. The specifications on reactor period are included to ensure that safety limits are not exceeded.

3.5.3 Reactor Safety System Aeol icabili ty. This specification applies to the reactor safety ,

system enannels.

Ob_iective. The objective is 'to specify the minimum numbs or eactor -

safety system channels that must be operable for safe operatf5n. ,

Soecification. The reactor shall not be operated unless the safety channels described in Table I and interlocks described in Table II -

are operable.

Bases. The fuel temperature, power level, and period scrams provide protection to assure that the reactor ca.. be shut co ?. before the safety limit on the fuel element temperature will be exceeded. The

g; manual scram allows the operator to shut down the system if an unsafe

= or abnorcal condition occurs. The preset timer insures that the reactor power level will reduce to a low level after pulsing, The high voltage scram insures that the power measuring channels operate within their intended range as required for proper functioning of all power level scrams. ,

The interlock to prevent startup of the reactor at c:unt rates less than 2 cps assures that start up is not ini-f ated unless a reliable indication of the neutron flux level in the reactor core is available.

The interlock to prevent the initiation of a culse above 1 kw is to assure that the magnitude of the pulse will not cause the fuel element temperature. safety limits to be exceeded. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing the reactor in the steady-state mode.

The interlock to prevent withdrawal of the shim, safety or regulating red in the pulse ecde is to prevent the reactor from being pulsed while on a positive teriod. The interlock to preven: simultaneous withdrawal of two cor. trol rods is to limit reactivity insertion rate frem the standard c:rtr:1 rocs. The intericek en the transient red cylinder position :revents the pulse inserti: . cf re tnan 2.55 l d:llars of reactivii during the cuise or s:uare-wave mode.

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TABLE I Dhh Minimum Reactor Safety Cha,nnels Effective Moce  :

Safety Channel Function S.S. I Pulse iS.W.

Fuel Element Temperature SCRAM @ 510*C X X X .

Safety Power Level SCRAM @ 110%* X X l Percent Power Level SCRAM @ 110%* X X Console Scram Button SCRAM X X X Wide-Range Log Power SCRAM @ period no less than . -

x }

Level 3 sec. -

Preset Timer Transient rod SCRAM @ 15 sec

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X l l or less after pulse i High Voltage SCPAM @ 25% of nominal opera-X X X

. ting voltage g

• For the purpose of testing the full power. safety channels, the reactor ray be operated with the Linear Power Level and the Percent Power Level Scram

_ setpoints Jet not greater than 120% of rated steady state power during the EE testing period.  :

TABLE II  :

Minimum Interlocks ,

Effec:1ve Moce )

Interlock Function S.S. Pulse i S . a' . )

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'r.'ide-Range Log Power Prevents control rod X l Level Channel l withdrawal @ less than l 2 eps Transient Rod cylinder Prevents application of X air unless fully inserted 1 kw Pulse Interlock Prevents pulsing . X above 1 kw

! Shim, Safety and Prevents simultaneous X .

X l Regula-ing Rod Drive withdrawal of two rods '

Circui- ,

t .

! Shi., Safety, anc Prevents movement of X

' Re;ula:ing Rod Drive any rod excao: transien: -

Cir:ui- rod Trans itin: Red Cylinder .

Prevents pulse inser:icn of X !X reactivity greater tnan 52.55 !

lpcsition .

-I2-A= 2.:i=en: N o . J' , 4 90008235

Basis. The minimum height of 14 feet of water above the core guarantees that there is sufficient water for effective cooling of the fuel and that the radiation levels at the ::p c' the rea: tor are within acceptable levels (SAR). The bulk water tem:erature limit is necessary, according to the reactor manufacturer, to ensure that the aluminum reactor tank maintains its integrity and is not degradec.

3.8 LIMITATIONS ON EXPERIMENTS Apol.i ca bility. This specification applies to experiments installed

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in. tne reactor and its experimental facilities. ,

Obiective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an ex-periment failure. .

Soecifications. The reactor shall not be operated unless the -

following concitions governing experiments exist:

a. Non-secured experiments shall have reactivity wcrths less than 1 dollar.

b. The reactivity worth of any single experiment will be less than

2. 55 dollars. l

c. Total experiment worth of all experiments will not exceed 3.00 6

dollars.

d. Explosive materials, such as gunpowder, TNT, nitroglycerin, or PETN, in cuantities greater than 25 milligrams shall not be irradiated in the reactor or experimental fa.:ilities. Exel o-sive materials in cuantities less than 25 milligrams rr.ay be irradiated provided the pressure produced u:en cetonatier. of the explosive has been calculated and/or exoericentally de crstrated to be less than tne design pressure of the container. E7 EPTION:

Explosive materials not exceeding 0.014 lbs. equivalent cf Thi may be irradiated in the laboratory area adjacen: to the end of the OSTR tangential beamport for the purpose of neutron radi- -

ography.

e. Where the possibility exists that the failure of an experiment (except fueled experiments) under (1) normal operating ccnditions of the experiment or reactor, (2) credible accicent conditions in the reactor, or (3) possible accident cen:iticns in -he exceri-ment, could release radioactive gases or aerosols to the eac ce bay or the unrestricted area, the cuantity anc tj:e Of caterial shall be limited such that the air:orne concentratier, of radi:-

activity averag'ed over a year will not exceed - e limits :f Appendix B of 10 CFR Part 20, assuming 100% cf tne gases :r aerosols escape.

f. In calculations pursuant to d., above, the folle ing ass.m:ti:r.s shall'be used:

Amendment No. E, 4 l bbb 90008236

increase in loading would result in an increase in power density of about 2%. Similarly' a minimum erbium content of 1.1% in an element is about 30% less than the design value. This variation would result in an increase in power density of only about 6%.

An increase in local power 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. clad about a factor of two greater than the

. value resulting from a hydrogen-to-zirconium ratio of 1.60.

However, this increase in the clad stress during an accident would not exceed the rupture strength of the clad. When standard and FLIP fuel elements are used in mixed cores, visual identifi-cation of types of elements is necessary to verify correct fuel loadings. ,

b. A maximum uranium content of 9 wt-: in a standard TRIGA element .

. is about 6% greater than the design value of 8.5 wt-%. Such an -

increase in loading would result in an increase in power density of less than 6%. An increase in local power density of 6". . educes the safety margin by, at most,10%. The traximum hydrogen-to-zirconium ratio of 1.8 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than the value resulting from a hydrogen-to-zirconium ratio of 1.60. However, this increase in the clad stress during 7- an accident would not exceed the rupture strength of the clad.

When standard and FLIP fuel elements are used in mixed cores, visual identification of types of elements is necessary 'o verify correct fuel loadings. '

5.2 REACTOR CORE Acolicability. This specification applies to the configuration of fuel ano in-core experiments.

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

Soecifications.

a. The core shall be an arrangement of TRIGA uranitn-zirconium hydride fuel-moderator elements positioned in the reactor grid plate.

b. The TRIGA core assembly may consist of standard fuel elements, FLIP fuel elements, or a combinatien thereof -(mixed core). Any operational core. assembly involvino FLIP fuel shall have no less than 80 FLIP fuel elements, located in .a contiqueus, central region.

c. The fuel shall be arranged i, a close-sacked configuration exces: for single element cositions occupied by in-core experi-ments, experimental facilities, gra: nite dumies, aluminum du=ies, stainless steel du mies, c:ntrol rods, anc startup sources.

^:Sn:i:Snt No # , 4 90008237

6. ADMIf4ISTRATIVE C0liTROLS 6.1 ORGANIZATION

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a. The facility shall be under the direct centrol of the Reactor Administrator or a licensed senior operator designated by him to be in direct control. The Reactor A6fnistrator shall be responsible to the Vice President for Administrat. ion of Oregon State l University for safe operation and maintenance of the reactor and its associated equipment. The Reactor Administrater or his appointee shall be responsible for assuring that all operations are conducted in a safe manner and within the lirr.its prescribed

, by the facility license and the requirements of the Reactor Operation Cota .i ttee . He shall enfoice rules for the protection of personnel -

against radiation. f

b. The safety of operation of the OSTR shall be related to the -

University Administration as shown in the folicwing chart:

! President, Oregon  !

, = =. State University 6

Radiation Safety )

, Comittee OSU ----------- Vice President for ,

i s Administration I e 's % -

l i 's s i Radiation Center 's------

Heal h physics - - - - - - Adm ~5 i l Reactor Operation Assistant Reactor l

" " " ~ ~ "

Comi ttee Adninistrator Reactor Su::ervi ser OSTR Operations 30 90008238

.ee .=e= so . .e 4

2. These events reported as required b,y Sections 6.7.a.2 through 6.7.a.8.

c. A report within 30 days in writing to the NRC, Region V, Office of

. Inspection and Enforcement, with copies' to the NRC, Director Office of Inspection and Enforcement.

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1. Any significant variation of measured values from a corresponding predicted or previously measured value of safety-connected operating characteristics occurring during operation of the .

reactor;

2. Any significant change in the transient or accident analyses as described in the Safety Analysis Report;

3. Any changes in facility organization or personnel; and 4; Any observed inadequacies in the implementation of ad:ninistrative l or procedural controls.-

A report within 90 days after completion of starting testing of the d.

reactor (in writing to the NRC, Region V, Office of Inspection and Enforcecent and copies to NRC, Direc or, Office of Inspection and Enforcecent) upon receipt of a new facility license, or an amendment to the license authori:ing an increase in reactor power level, describing the measured values of the operating conditions or characteristics of the reactor under the new conditio.ns including:

1. An evaluation of facility performance to date in cor.arison with design predictions and specifications.

2. A reassessment of the sa'fety analysis subrnitted with the license application in light of ceasured operating charac-teristics when such measurements indicate that there may be substantial variance from prier analysis.

e. An annual report within 75 days following the 30th of June of each year l (in writing to the NRC, Region V, Office of Inspection and Enforcement, and copies to the NRC, Director, Office of Inspection and Enforcement).

1. A brief sumary of operating experierice including experiments perfomed and changes in facility design, perfomance charac-teristics and operating proced0res related to reactor safety occurring during the reporting period, and results of sur-veillance test and inspections.

2. A tabulation showing the energy generated by the reactor (in megawatt-hours), hours reactor was critical, and the cum-

_ ulative total energy output since jnitial criticality.

l 3. The number of emergency shutdowns and inadvertent scrams, including reasons therefore.

4 90008239 Discussion of the major maintenance ccerations perfomed during the period, including the e#fect, if any, on the safety of the operation of the reactor anc tne reasons for any cor-rec-ive maintenance required.

3 Amendment No. 4,' 4 D

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