ML19241A754

Amend 17 to Tech Specs for License R-84,developed Per Reg Guide 1.16 & Proposed ANSI Std 15.18

ML19241A754
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	07/02/1979
From:	DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
To:
Shared Package
ML19241A748	List: ML19241A747 ML19241A754
References
NUDOCS 7907090354
Download: ML19241A754 (37)

Text

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A.1IENDMENT 4 l~ TO THE TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. R-34 FOR THE AFRRI TRIG A .'vIARK F REACTOR DOCKET NO. 50 '-')

These Tecnr.ical Scecificatiens ' ave ' eart

~ cdi +' d * ' te from .RC

+. ~ pidance, ec:f ca}:cr.s ret'later! Guide 1.S and salient f'eatures of T'S 'e '

'.* 21 t

ce their TRIG A, Decket No. 50 - 123, License No$xas ~R-33[ Ii 309 176 7007000$.5..

Included in :his dccument are the Technical Spet.ifications 2nd the "3ases" for *he Technical Scec:fications. These bases, which provide the technical suggcet for the incividual :ecnnical specifications, are incluced fcr informatien purposes caly They re n.ct ; tri cf the Technical igecifications, and they do not ecnstitute limitations or racuirements te which the licensee must achere. Reference NRC Regulatory Guide 1.16 and ANSI N373-1974.

1.0 DEFINITIONS

_ REACTOR OPERATING CONDITIONS .

1.1 REACTOR SHUTDOWN The reacter is snut dcwn when the reacter a suceritica.1 by at ' east ene dctlar of . eactivitJ.

1.3 REACTOR SECURED The -aactor is secured wnen an *.he fcilowing cenci-icas are satisfied:

3. The reactor is shut dcwn,

b. The conscle key switch is in *he "off' ;csiticn and the <ey :s removed from *he censole anc under the contrct of a licensed cpera ce Oc stored in a lccked stcrage rea. and

c. No werk is in ; reg ess invciving in-ccre fuel handling r refueling cperations, main:enance of the caseter cc its cent :t meenanisms, cc insertica cc withdrawal of in-ccre experiments, unless sufficient fuel is remcved to insure a 51.00 shutdown marg:n with :he mcs: reactive centrol red removed.

1.3 REACTOR OPERATION Reacter operaticn is any ecnditicn wherein :he reacter is not secured.

'. 4 COLD CRITIC AL The reacter is in the ccid critical ecnditicn wnen it is cri:ical at a ;cwer 'evel

'ess than 100 watts.

1.5 STEADY STATE MODE Cperaticn in the steady state mcde shan mean steady state cperaticn of the reacter either by manual cperaticn Of the centrol reds or by automatic cperation cf cne control red (servc cent:00 at pcwer levels up to . MW. utilizing the scrams in Table I and the intericeks in Table II.

309 177

1.S PULSE MODE Oper'. tion in the pulse mcde shall mean that the reactor is intentionally placed en a prompt critical excursicn by making a step insertion of reactivity above critical with the t ansient rod, utilizing the scrams in Table I and the interlocks in Table H.

1.7 SHUTDOWN MARGIN 4

Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence

• hat the reactor can be made subaritical by means of the centrcl and safety systems, starting from any permissible operating ecnditions and that the reactor will remain suberitical without further coerater action.

1.3 ABNORMAL OCCCRRENCE An 'Abncemal Occur ence" is defined for the purecses of the repceting require-ments of Section 208 of the Energv Reorganizaticn Act of 1974 (P.L.93-408) as an unscheduled incident er event which the Nuclear Regulatory Commissien determines is significant from the standpoint of puclic health cr safety.

1.3 a REPORTABLE OCCURENCE A re;crtacle occurrence is any of the following which cccurs during reacter coeration:

a. Operatien with any safety system setting less conservative than specified in Secticn 2.0, Limiting Safety System Settings;

b. Operation in viclaticn of a Limiting Ccnditicn for Cperation;

c. Failure of a recuired reactor cc exneriment safety system com;cnent which eculd render the system incapacle of performing its intended safety functicn;

d. Any unanticipated er unecntrolled change in retetivity g eater than One dellar;

e. An ecserved inacecuacy in the .mplementation of either administrative er peccedural controls such that the inadecuaci cculd have caused the existence ce devebpment of a ecnditicn which Nuld result in operation of the reacter cutside the specified safety limits; cr

f. Release of fission products from a ftel element.

2 309 178

1.10 EXPERISIENT Experiment shall mean (a) any apparatus, device, er material which is not a normal part of the core or experimental facilities, but which is inserted in these facilities cc is in line with a beam of radiation criginating from the reacter core; or (b) any cperation designed to measure reactor ,arameters or -

characteristics.

Lil EXPERISIENTAL FACILITIES The extosure facilities associated with the AFRRI-TRIG A reactor shall be:

a. Exposure R ' eom 41

b. Exposure Room 40

c. Pneumatic Transfer System

d. Reacter Pcci

e. Pertable 9aam Tube 1.12 STANDARD COGROL ROD A standard centrol red is a centrol red having an electo-mechanical drive and scram capabilities. It is withcrawn by an electromagnet / armature system.

1.13 TRANSIENT ROD The t ansient red is a cent ci red with scram capabilities that can be capidly ejected from the reactcc core to pecduce a pulse. It is activated by applying compressed air to a pisten.

1.14 FUEL ELE 31ENT A fuel element is a single TRIGA fuel red. ~

L15 CCRE POS' TION The ccre ;cs'tien refers to the locaticn of a fuel or centrol element in the g-id structure.

1.16 INSTRUSIENTED E' E3IENT An instrumentec element is a spe" "al =iement in which sheathed c.hrenei-alumel or < ;uivalent !!.ermoccuples are emoedded in Se :uel.

3 .

1.17 SAFETY LI31IT Safety limits are limits en impcrtant peceess variables which are found to be necessary.to reasonably protect the interity of certain phy ical barriers which guard agains: the uncontrolled release of radicactivity.

1.18 LIMITDiG SAFETY SYSTEM SETTING Limiting safety systems sett:ng is setting for automatic ptotective devices ,

related to these variables having significant safety functions.

1.19 OPERABLE A system, device, Or component shall be censidered Operable when it is ca7able of performing its in+. ended functions in a normal manner.

L20 REACTOR SAFETY SYSTEMS Reactor safety systems are these systems, including their associated input circuits, which are designed to initiate a reactor scram for the primary purpcse of protecting the reacter er to provide informaticn which may rec: ire manual protective action to be initiated.

1.21 MEASURED VALUE The measured value is the magnitude of r variable as it appears en the butput of an measuring channel.

1.22 MEASURING CHANNEL A measuring channel is that combination of senser, intercennecting cables or lines, amplifiers, and cutput device which are connected fer the purpcse of measuring the value of a variable.

1.23 SAFETY CHANNEL A safety channel is a measuring channel in the reactor safety system.

L24 CHANNEL CHECK A channel check is a verificaticn of accepcable perfcrmance by cbservaticn of channel behavice.

1.25 CHANNEL TEST A channel test is the introducticn cf a signalinto the channel to verify that it is cperable.

4

r 1.26 CHANNEL C ALIBRATION A channel calibration censists of using a known signal and adjusting channel output to resperd within limits specified by the manufacturer.

1.27 ON CALL A persen is censidered en callif:

s (1) the individual has been specifically designated and the operatcr knows o.

the designation; (2) the individual keeps the operater pested as to his whereabouts, and phone number; and (3) the individual is capabia of getting to the reacter facility within thirty (30) minutes under ncemal circumstances.

2.0 SAFETY LDIITS AND LDIITING SAFETY SYSTESI SETTINGS 2.1 SAFETY LDIIT-FUEL ELE 31ENT TESIPERATURE Acplicability This specificatica applies to the temperature of the reacter fuel.

Objective The cbjective is to define the maximum fuel element temperature that can be permitted with ccnfidence that no damage to the fuel element cladding will result.

Soecificatien The temperature in a standard TRIG A fuel element shall not exceed 500 C under any ccnditions of cperation.

Bas's The impcetant parameter for a TRIG A reacter is the fuel el".nent tempetature.

This parameter is well suited as a single scecification especially since it can be measured. A less in the integ-ity of the fuel element cladding could arise trem a build-up of excessive pressure between the fuel-mederator and the cladding if the fuel temperature exceeds 'he safety limit. The pressure is caused by the presence ef air, fissien product gases, anc nydrogen from the disscciaticn cf *he hycrogen ind zireenium in the fuel-moderator. The marnitude of this pressure is determined by the fuel-mcderator temperature and the ratio of hydrcgen to zirecnium in the alloy.

The safety Umit for the st'M'-" TRIGA :cel is based en data. 'nciucing -he large mass of experimental evidence cotained curing hign pedccmance reactcr

\

tests on :nis fuel. These data indicate that the stress in the cladding due to hydtcgen pressure from the dissceiatien of zirconium hydride will remain below the gtlimate stress pre cided that the temperature of the fuel does not exceed 1000 C and the fuel cladding is water cooled.

2.2 LDIITING SAFETY SYSTEM SETTINGS Acolicability This specificaticn acplies to the scram settings which prevent the safety limit from being reached.

Obiective The cbjective is to prevent the safety limits from being reached.

Scecificaticn The limiting safety system settings shall not exceed 600 C as measured in an instrumented fuel element. One system will utilize an instru:nented element in the "B" ring and the other will utilize an instrumented element in the "C" ring.

Basis The limiting safety system setting is a temperature which, if exceeded, shall cause a reaeter scram to be initiated preventing the safety limit fretn being exceeded. A setting of 500 C provides a safety margin of 400 C for standard TRIGA fuel elements.

During s*.eady state cperation, the flux peaks in the "C' ring and curing pulse cperation, the flux peaks in the '3" ring. By utilizing one element from each ring a wcrst case indication is reascnacly assured.

3.0 LDIITING CONDITIONS FOR OPERATION 3.1 STEADY STATE OPERAT!ON Acolicability This specificaticn applies to the energy generated in the reactor duri: g steady state cperation.

Scecificatiens The reactor steady state pcwe.- level shall not exceed '.1

. megawatts. T'.e ncrmal steady state cperating power limit of the reacter snail be 1.3 a

\u, -

- o/

megawatts. Fcr purposes of testing and calibratica, the reactor may be operated at power levels not to exceed 1.1 megawatts during the testing period.

Basis Thermal and hydraulic calculations indicate that TRIGA fuel may be safely operated up to power levels of at least 0.0 megawatts with natural convective coolirg.

3.2 REACTIVITY LIMITATIONS Apolicability These soecificatiens apply to the reactivity conditicn of the reactor and the reactivity wceths of control reds and exceriments. They apply for all modes of cperation.

Objective The objective is to assure that the reacter can be shut down at a!! times and to assure that the fuel temperature safety limit will not te exceeded.

Scecificatiens

a. The maximum available excess reactivity above ccid critical with er without all experiments in place shall be 55.00 (3.5% Sk/k). .-

b. The minimum shut dcwn margin provided by the remaining centrol reds with the mest reactive centrol red fully withdrawn er removed shall be 51.00 (0.~0% Sk/k).

Bases The shutdown margin assures that the reacter can be shut down from any operating cenditica even if the highest wceth centrol red shculd remain in the fully withdrawn position.

3.3 FCLSE MODE OPERATION Acclicacilit?

This soecificaticn acplies to the energy generated in the reacter as a result of a pulse insertica of reactivity.

Obiective The oofective is to assure that the fuel temcerature safety limit will not be exceeced. .

30/

Scecification The maximum step insertic.1 of reactivity shall be 2.5% Sk/k in the pulse mcde.

Basis Based upcn the Fuchs-Nordheim mathematical medel of the AFRRI-TRIGA reactor. an insertion of 2.5% Sk/k would result in a maximum average fuel .

temperature of less than 450 C, thereby staying within the limiting factces.

3.4 CONTROL AND SAFETY SYSTEM 3.4.1 Sc~am Time Acolicability This scecificatien applies to the time recuired fer the cent-cl rods to be fully inserted frcrr the time that a safety channel variable reaches the Safety System Setting.

Objective The cbjective is to achieve prcmpt shutdown of the reactor to prevent fuel damage.

Scecificaticn The time from scram initiation to full insertien of any centrol red shall not exceed one seccnd.

Basis This specificatien assures that the reacter will be promptly shut down when a scram signal is initiated. Experience and analysis indicate that for the range of ansients anticipated for a TRIGA reacter the specified scram time is adequate to assure the safety of the reacter.

3.4.2 Reacter Centrol System Acclicability

' tis scecificaticn applies to the channels monitoring the reacter ccre which must provide informatica to the reactor cperater during reacter cperatien.

Obiective The cojective is to recuire that sufficient inicemation is available to the cperator to amure safe operation of the reacter.

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3cecu!CatiCn T' e reac Or snall '.ot be cperated unless *he measuring enannels listed in *he following table are Operacle.

Min. No. Effective Mode Cceracle 3.5. ouisa Measurine Channel

.: uel Element Temcerature 2 X X I X Linear Power Level L:g : wer Level 1 X (2) ( ,i )

'barety vaannels ..

Numcer ::eracle sncwn in effer.tive cde :Olumn Basis

.. . e i . ...

. . y . , . . . ... . . . . a... y t.2_ ,

s 2_ ,. a.. a. . r. *s.-

. l . .n..emla. -

3.", a. s .- . n t '.. ." c ".s i . ..". r . .. . m' ". . r.

On this para. meter whien has a specified safet't limit. The ;cwer level monit:rs assure 2at de reactor ;ower level is idecuately menit: red fcr acch steacy state and pulsine ~)

mcces cf co. er2tien. The sc, ecifica icns en reactor o.cwer :evel incicatica re included in this sec:ica since the power level is reia:ed :: Se fuel temgerature.

3.4.3-Reacter Safety System Acclicacilit*7 This sceci:icaticn applies :c the reacter saf ar, system anannels.

Objective The cojective is :c speciff *te ninimum number of reacter safer / system channels :nat must :e cperacle fer safe :gera: ten. .

OCec1IlCat!Cn The react:r shall not be egerated anless :he safety sys: ems describec 'n T2bies I and II re cperacle.

k

Bases The fuel temperature and ; ser level scr2ms pecvide pectection to assure 't.at the reacter can be snut down before the safety limit en the fuel element tempera:=e will 'cc exceeded. The .anual scram allows :he cperater :c snu:

down the system if an unsafe er abncemal conditien occurs. In the event of failure of the power supply for :he safety chamcers, cperatica cf :he reac:ce withcut adequate ia.st umentatica is prevented. The preset timer insures that the reacter ;cwer level will reduce c a 1cw level af ter pulsing.

r v. a-v. . . :_.,t Ae r.. 2..n3 1 eA.a t Ie ,<InJn

. u.n<,,I 2e.3.C . nv2.. _u..

Mcde in Which Effective Set ?cint SS Pulse Orignating Channel 110% max ;cwer X X

1. Percent ?cwe."

X X

2. Scram Sar en Censcle

3. Preset'"! .er Less than :r ecual

c !5 seccnds X Water level will ..ct fall below .-

4. ?ccl Water Level X 14 feet accve :cp cf ccre X

3. Sc.2m Bu:~ca in E:qcsure 2coms

~ X X

5. Fuel Temperature (2 independent enannels) 500 C X X

7. Less of Icn Chamcer High Voltage en Safety Channels X X

3. Less of F acility E'.ectical X X Pcwer g g g e e g g a g ,

. 6 6 6 6 e e 6 e e e g e e g ge e m e g ee e e e e g g e g 6 6e e 6 6 6 6 e

ae

TABLE II - MINIMI'M INTERLOCXS Mode in Which Effective Acticn Prevented SS l'ulse

1. Simultaneous manual withdrawal of two standard centrol reds. X

2. Withdrawal of any centrol element except ,,

X transient ecd.

3. Pulse initiaticn at power levels g eater than 1 kw X The interlock to prevent the initiaticn of a pulse above 1 kW is to assure that the magnitt de of the pulse will not cause the fuel element temperature safet'J limits to be exceeded. The interlock to prevent withdrawal of the standard centrol reds in the pulse mode is to prevent the reacter from being pulsed while en a pcsitive period.

3.4.4-Facility Interlock System Acplicability This specificaticn applies to the intericcks which prevent accidental exposure of an individual in either exposure rcom. .

Obiective The cbjective is to provide sufficient warning and inter!ceks to prevent move-ment of the reactor ccre to the execsure rcom in which someone may be working.

Scecifications Facility inter!ceks shall be provided suen that:

a. The reactor cannot be operated unless the lead shielding deces within the reacter pcol are either fully cpened ce fully cicsed. ~

b. The reactor cannot be operated unless the excesure team plug dcce adjacent to the reactor ccre position is fully clcsed and the lead shielding doors are fully c!csed; er if the lead shielding dcces are fully opened, both execsure recm plug deces must be funy closed.

11 309 N

c. The reactor cannot be cperated unless a warning horn has sounded in an exposure rcom where operaticn er movement of the reactor would cause an unsafe ecnditicn.

Basis These intericcks physically prevent movement of the reactor from one exncsure rocm domain to the other without first sounding a warning so that an individual in the affected exposure room can cperate the " Emergency Step" system to .

alert the cperator and prevent core movement.

3.5 RADIATION NIONITORING SYSTE31 Acplicability This soccification applies to the radiation menitcring information which must be available to the reactor cperater during reac ce operation.

Obiective The objective is to assure that sufficient radiaticn monitcring informatien is available to the cperater to assure safe cperation of 2e recter.

Specification

_ _ . . The reactor. shall not be cperated unless the fc11owing radiation mc~nitcring systems are operacle:

a. A minimum of two area radiation mcnitors and a centinuous air meniter in the reacter ecom. The area menitor directly over the reactor pool surface and the centinucus air mcnitor shall produce audible and vuible alarm signals in the centrol ecom. The second area monitor in the reactor room shall have a visible alarm in the control room.

b. Two area radiatien menitors in the precaratien area en the wall directly opposite both Exposure Room 41 and Execsure Room #2 plug doors to serve as radiation streaming detectors, and provide a visible alarm in the centrol roo m.

c. A radiaticn detecter system centinuously sampling the effluent from the reacter building ventilating system exhaust stack and capable of pro-ducing a visible alarm in the centrol rcom.

309 188 12

Bases The radiation meniters crovide information to cperating persennel of any impending er existing danger from radiatien so that there will be sufficient time to evacuate the facility and take 2e necessary steps to prevent the spread of radioactivity to the surroundings.

3.6 ARGON-41 DISCHARGE LD1IT Acolicability ..

This specificatien applies to the concentration of Argen-41 that may be discharged from the TRIGA reacter facility.

Obiective To insure that the health and safety of the public is not endangered by the discharge of Argen-41 from the TRIGA reacter facility.

Soecificatien

a. An environmental monitoring pregaam shall be maintained to determine effects of the facility en the environs.

b. If a dosimeter reading for any envircnmenta. menitoring statien indicates that an excesure of 400 me above backgrouttd has been reacned during

- the year, reactor operations which generate and release rireasurable quantities of Argen-41 will be curtailed to 2 mw/hr per month fc>

the remainder of the year.

c. If dosimeter reading for any envirer. mental monitoring statien indicates that an exposure of 500 me abcve backg cund has been reached during the year, reacter cperaticns which generate and release measurable quantities of Argen-41 will be ceased for the remainder of the year.

Bases The basis for these exncsures is 10 CFR 20.105, ' Permissible levels of radiation in unrestricted areas." Calculaticns with regard to the cencentratien of AR-41 in air at locatiens around AFRRI indicate that even during sustained coerations the cencentratien will not appecach an MFC (ref SAR) for an unrestricted area. Therefore the centrolling factor is total dose.

3.7 POOL BULK W ATER TEMPERATURE Acolicabilit r This scecificaticn refers to cperation cf the reacter with respect to bulk poc1 temperature.

13 309 W9

Objective To insure the integ-ity of the resins in the water purification system.

Soecificatien The reactor win not be operated when the pool bulk water tempera'.ure cxceeds 60 C.

Basis Manufacturer's data states that the resins in the water curificatien system break down with sustaineC cperation in excess of S0 C.

3.3 ENGLNEERED SAFETY FEATURE - VENTILATION SYSTEM Acolicability This speelficatien applies to the cperatien of the fae!Iity ventilatica system.

Objective The objective is to assure tha". the ventilation system is in cperatica to mitigate the censequences of the pessible release of radicactive materials .esulting from reactor operation.

Scecification The reactor shall not be cperated unless the facility ventilation sys'em is coer-able except for periods of time necessary to permit recair of the sr; tem. In the event of a substantial release of air 0cene radicactivity, the ven'.latien svstem will be secured automatically by a signal from the reacter deck centinucus air menitor.

Easis During normal cperation of the ventilatica system, the ccncentraticn of Ar;cn-41in unrestricted areas is below the MPC. In the event of a substantial release of airectne radicactivity, the ventilation system will be secured automatically.

Therefore, cperatica cf the reactor with the ventilation system shut down for sheet periods of time to make recairs insures the same deg ee of centrol of release of radioactive materials. Mereever, radiation monitces within the building independent of these in the ventilatien system will give warning of high levels of radiatien t. hat might cccur during cperaticn with the ventilation system :ecured.

14 309 190

3.9 LIMITATIONS ON EXPERDIENTS Acclicability This spe.cification applies to exoeriments installed in the reactor and its experi-mental fccilities.

Objective The objective is to prevent damage to the reacter or excessive c. lease of radicactive materials in the event of an exNriment malfunction.

Specificatiens The follcuing limitations shall apply to the irradiation of materials (other than air):

a. If a pessibility exists that a release of racicactive gases er aeroscis may occur, the amount and type of material irradiated shall be limited to assure the yearly ecmpliance with Table II, Appendix B, of 10 CFR 20 assuming 100% of the gases or aercscis esespe.

b. Each fueled exteriment shall be lit 2ted such that the total inventory of icdine isotoces 121 througn 135 in the experiment is not Feater than 1.5 curies and the maximum Sr-90 inventory is act peater than 5 millicuries.

_ I_ c.. Known exclcsive materials shall not be ir adiated in the resetcr in quantities peater than 25 milligams. In additien, the pressure produced in the exoeriment centainer upca detonation of the explcsive shall have been ex::erimentally determined to be less than the design pressure of the centainer.

d. Samples shall be doubly encapsulated when the release of the centained material eculd cause ccercsion of the experimental facility.

e. The sum of the absciute reactivity worths of all cxceriments in the reacter and in the associated experimental facilit.es shall not exceec 53.00 (2.1% delta k/k). " his includes the total potential reactivity insertien which might result from exreciment malfunction. accidental exceriment Goeding er voiding and accidental removal er insertica of

~

exoerim ents.

f. In calculaticns regard:ng exoeriments, the fe" wing assumptiens shall be made:

1. If the effluent exhausts througn a :llter instal'ation designed for pester than 99% efficiency fer 0.3 micron particies, at least 10% of the aercscis produced can escape.

15 309 191

2. For materials whose boiling point is above 130 F and where vapors formed by boiling this material could escape only through a column of water above the ccre, at least 10% of these vapers can escape.

g If a capsule fails and releases material which could damage the reactor fuel ce structure by corrosion er other means, physical inspection shall be perfccmed to determine the consequences and need fer corrective action.

The results of the inspection and any corrective action taken shall be reviewed by the Physicist-in-Chrge and determined to be satisfactory ..

before cperation of the reacter is resumed.

Bases

a. This specificatien is intended tc reduce the likelihood that airborne activities in excess of the limits of Appendix 3 of 10 CFR Part 20 w&1 be released to the atmcsphere cutside the facility boundary.

b. The 1.5-curi' limitatien en iodine 131 through 135 assures that in the event of malfuncta .i of a fueled experiment leading to total re: ease c.' the iodine, the exposure dese at the exclusion area boundary will be less than that allowed by 10 CFR Part 20 for an unrest-icted area.

c. This soecificatien is intended to prevent damage to reacter components resulting from malfunction of an experiment involving excic;ive materials. _

d. This specification is intended to provide an additional safety facter where damage to the reacter and compcnents is possible if a capsule fails.

e. The maximum wcrth of experiments is limited so that their removal from the cold critical reactor will not resuit in the reactor achieving a power level high encugh to exceed the ccre temperature safety 1.

f. This specificatien is intended to insure that the limits of 10 CFR 20 Appendix B are not exceeded if an experiment malfunctions.

g. Operatien of the reacter with the reacter fuel ce structure damaged is pechibited to avoid release of 5ssien ;:reducts.

4.0 SURVEILLANCE REQUIRDIENTS .

4.1 GENERAL Acolicability This specificaticn acclies to the surveillance requirements of any system related to reacter safety. .

16 309 192

Objective The cojective is to verify the proper cperation of any system relatad to reactor safety.

Scecifications Any additions er modificatiens to the ventilation system, the core and its associated suppcet structure, the pec1, the pcci ecolant system, the red .

drive mechanism, er the reactor safety system shall be made and tested in acccedance with the specificaticns to which the systems were originally designed and fabricated ce to specifications approved by the RRFSC. A system shall not be considered cperable until after it is successfully tested.

Basis This specification relates to changes in reacter systr .s which eculd directly affect the safety of the reactor. As 1cng .s changes or replacements to these systems centinue to meet the original design specifications, then they meet the presently accepted cperating criteria.

4.2 SAFETY LDiIT - FUEL ELDIENT TEIPERATURE Acolicabill y .

This specification applies to the surveillance re<;uirements of the fuel element temperature measuring channel.

Cbjective The cbjective is to assure that the fuel element temperatures are properly menitcred.

Scecificaticra

a. Whenever a reacter scram caused by high tuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature safety 3mit was exceeded.

b. A calibratien of the temperature measuring channels shall be performed annually but at intervals not to exceed 14 mcnths.

c. A check of the fuel element te nrerature scram shall be made daily whenever the reactcr is Operatec.

_ Basis _

Cperaticnal excerience with the TRIG A system gives assurance that the thermeccuple measurements of tuel element temperatures have been 17 309 193

sufficiently reliable to assure accurate indicatien c' this parameter.

4.J LIMITING CONDIT ONS 72R OPERATION 4.3.1-Reactivity Recuirements A00llsaty11t1 These specificaticns apply to the surseillance requirements fer reactivity control of experiments and systems.

Objective Tne objective is to measure and verify the worth, performance, and cperacility of these systems affecting the reactivity of the reacter.

Scecificatiens

a. the reactivity worth of each centrol rod and the shutdown margin shall be determined annually but at intervals not to exceed 14 months

b. The reactivity wer? cf en entriment shall be estimated ce measured, as appecpriate, before rcactor power cperation with said experiment, the first time it is perfcrmed.

c. The centrol reds shali 56 visually inspected for deterioratien annually, not to exceec 14 months.

d. On each day that pulse mcde cperation of the reacter is planned, a functional pericemance check of the transient (pulse) red system shall be per'ormed. Semiannually , at intervals not to exceed eight months, the transient (pulse) red drive cylinder and the associated air suppl- system shall be insoected, cleaned and lubricated as necessary.

e. The core excess reactivity shall be measured at the beginning of each day of cperation or prior to each centinucus cperatien extend-ing mere than a day.

f. The power ccefficient of reactivity between 100Kw and 1 kW will be measured annually not to exceed 14 months.

Basis The reactivity wceth of the centrol reds is measured to assure that the recuired shutdown margin is available and to provide an accurate means fcr determining :he reactivity wceths of experiments inserted in the core.

13 309 194

Past experience with TRIGA reacters gives assurance that measurement of the reactivity wceth en an annual basis is adecuate to insure that no significant changes in the shutdown margin have occurred. The visual inspection of the centrol reds is made to evaluate ccrresien and wear characteristics caused by operation in the reacter.

4.3.2-Centrol and Safety Systems Aeolicabilit.y ..

These specificatiens apply to the surveillance requirements for measure-ments, tests, and calibrations of the control and safety systems.

Objective The cojective is to verify the performance and cperability of these systems and components which are directly related to reactor safety.

Scecificatiens

a. The centrol ecd drep times shall be measured semiannually but at intervals not to exceed 3 mcnths.

b. A Channel Test of each of the reacter safety system channels for the intended mcde _f cperation shall be performed prior to each day's cperatien er price to each cperation extending mere than enc day, except for the pool water level channel which snall be tested weekly.

c. A Channel Calibraticn shall be made of the power level monitoring channels by the calcrimetric method annually but at intervals not to exceed 14 months.

d. The water cenductivity shall be mcnitored at leau weekly during ncemal cperation. The water cenductivity shall not exceed ? micro-mho's per centimeter averaged over a month.

Bases Measurement of the scram time en an semiannual basis is a verificatien of the scram system electrcnics, and is an indicat:cn of the capability of the centrol rods to perform picperly. The channel tests will assure that the safety system channels are cperable en a daily basis or prict to an extended run. The power level channel calibration will assure that the reacter will be cperated at the ;rtper power levels. Transient centrei red checks and semiannual maintenance insure preger Operatien of this centrol red.

19 309 195

The water conductivity is centrolled to avoid 2ctivation cf minerals in the water and to avoid detericration of cladding, piping, and other systems.

4.3.3-Radiatien ?,1cnitoring System Acclicability This specification applies to the surveillance requirements for the area radiatica monitoring equipment and the centinucus air menitcring system. ,,

Objective The objective is to assure that the radiation menitoring equipment is cperating and to verify the apprcpriate alarm settings.

Scecificatien The area radiatien menitoring system and the centinucus air menitoring system shall be calibrated annually but at intervals not to exceed 14 menths and shall be verified to be opertble quarterly.

Basis Experience has shown that quarterly verificatien of area radiation and air monitcring system set points in cenjuncticn with annual calibration is ades ste to ectreet fer any variation in tne system due to a change of cperating character:stics over a long time span.

4.3.4-Ventilatien System Acolicability This specificaticn applies to the budding confinement ventilatica system.

Objective The cbjective is to assure the peccer cperation of the ventilatien system in centro 11ing releases of racicactive meterial into the uncentrolled envircnment.

Scecificatien It shall be verified mcnthly oy a visual check that the ventilation system is cperable.

20 309 196

Basis Exnerience accumulated over several years of oceration has demonstrated that the tests of the ventilation system en a monthly basis are sufficient to assure the proper cperaticn of the system and centrel of the release of radicactive material.

4.4 REACTOR FUEL ELDIENTS Acolicability This specification applies to the surveillance requirements for the fuel elem ents.

Cbiective The objective is to verify the integrity of me fuel element cladding.

Soecif! cations All fuel elements shall be inspected visually for damage or deterioration and measured fcr length and :cw at intervals separated by not '1cre tha7 500 ;ulses of insertien pester than $2.00 cr annually (not to exceed 14 months), which ever occurs first. Fuel elements incicating an elengation peater than IA0 inch, a lateral bending peater than UlS inch, er significant visible damage shall be censidered to be damaged and shall not be used in the reacter cere; Bases The frequency of inspection and measurement schedule is based en the parameters most likely to affect the fuel cladding of a pulsing reactor and utiliaing fuel elements wnose characteristics are well known.

The limit of transverse bend has been shown to result in no difficulty in disassembling the cere. Analysis cf the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching. Exnerience with TRIGA reacters has shown that :uel element bowing that could result in touching has cecurred without deletericus effects. The elongatien limit has been specified to assure that the cladding material wQ not be sucjected to stresses that eculd cause a loss of intepity in the fuel containment and to assure adequate ecclant flow.

5.0 DESIGN FEATURES 5.1 REACTOR FCEL 01 309 197

Aeolicability This specificatien applies to the fuel elements used in the reacter ccre.

Objective The cbjective is to assure that the fuel elements are designed and fabricated in such a manner as to permit their use with a high degee of reliability with respect to their physical and nuclear characteristics.

4 Scecificaticns The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:

a. Uranium centent: maximum of 9.0 Wt-% enriched to a nominal 20%

Uranium 235.

b. Hydrogen-to-circenium atom ratio (in the ZrH): Nominal 1.7 H atoms to 1.0 Z: atoms.

c. Cladding- 304 stainless steel, nominal 0.020 inch thick.

Basis A maximum uranium centent of 9 Wt-% in a standard TRIGA element is

- about-6% pester than the design value_of 3.5 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% reduces the safetf margin by at mest 10 %. The hydrogen-to-zircenium ratien of 1.7 will produce a maximum pressure within the cladding during an accident well below the rupture strength of the claddir.g.

5.2 REACTOR CORE Acolicabilit';

This scecificatien applies to the ecnfiguratien cf fuel and in-ccre experiments.

Cbiective The cbjective is to restrict the arrangement of fuel elements and exceriments so as to orevide assurance that excessive ;cwer densities will not be prcduced.

22 309 198

Scecificatien

a. The reactor ccre shall censist of standard TRIGA reactor fuel elements and a minimum of two (2) thermccouple instrumented TRIGA reactor fuel elements,

b. There shall be: four single cere pcsitiens cecupied by the three standard control rods and the transient rod: a neutrcn start-up scurce with holder; and positions for pessible in-ccre experiments.

4

c. The ecre shall be cooled by natural convection water flow.

d. In-ccre experiments shall not be placed in adjacent fuel positiens of the B-ring and/or C-ring.

e. Any burnable poisen used for the specific purpose of compensating fer fuel burnup cr icng term reactivity adjustments shall be an intepal part of the manufactured fuel elements.

Basis Standard TRIGA ccres have been in use for years and their safe operaticnal characteristics are well documented.

5.3 CONTROL ROCS Acclicability This specificatien applies to the centrol reds used in the reacter ecre.

Obiective The cbjective is to assure Sat the centrol rods are designed to permit their use with a high degee of reliability with rescret to their physical and nuclear characteristics.

Scecification

a. The standard centrol reds snall have scram capacility and centain torated paphite, B 4C power er beren and its compounds in solid fccm as a poisen in aluminum er stainless steel cladcing. These rods may have an aluminum follower.

b. The Tansient centrol red shall have scram capability and centain berated paphite 3 4C powder er beren and its compounds in a sclid fcrm as a poisen in an aluminum er stainless clad. This red may inecrporate an aluminum er poisen fc11cwer.

23 309 199

Basis The pcisen requirements for the centrcl rods are satisfied by using neut on abscrbing borated g"aphite, B C pcwder er beren and its compounds. These materials must be contained id a suitable cladding material, such as aluminum ce stainless steel, to insure mechanical stability during movement and to isclate the pcisen from the pool water environment. Scram capabilities are provided for rapid insertion of the control rods which is the primery safety feature of the reacter. The t ansient centrol red is designed fee a reacter ,,

pulse.

5.4 RADIATION MONITORING SYSTEM Acclicability This soecification describes the functions and essential compcaents of the area radiation monitoring equipment and the system for centincusly monitoring airbcene radicactivir/.

Objective The objective is to describe the radiation monitoring equipment that is available to the operater to assure safe operatien of the reacter.

Scecificaticn The radiatica monitoring equipment listed in the fellowing tacle ill be availaole for reacter cperaticn.

Radiatien Monitoring Channel and Functicn Area Radiation Monitor / gamma sensitive instruments)

Function - Menitor of radicactivity change in builcing, alarm and readout at centrol consele and initiate cicsing of ;csttive sealing dampers automatically.

Gas and Particulate Stack Radiaticn Monitor (gamma sensitive detector with air collection capabilit1)

Functicn - Monitor cencentraticn of radicacitve particulate activity and radicacitve gases in build.ing and exhaust, alarm and readout at centrol censcle.

Basis The adisticn mcnitoring system is intended to provide infccmation to cperating persennel of any impending er .isting danger from radiation se that there will be sufficient time to evacuate the facilit/ snd take the necessary steps to prevent the spread of radicactivit/ to the sur cundings.

24 309 200

5.5 FUEL STORAGE Acolicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective The cbjective is to assure that fuel 'whien is stored will not become crit; cal and will not reach an unsafe temperature.

Specification All fuel elem.nts not in the reactor core shall be stcred and handled in accordance with the provisions of 10 CFR 70. Irradiated fuel elements 'and fueled devices shall be stcred in an ar ay which will cermit sufficient natural convective cooling by water or air such that thr. fuel element or fueled device temperature will not exceed design vicues.

Basis The limits imposed by this specification are ecnservative and assure safe storage.

5.6 _

REACTOR BUILDING AND VENTILATION SYSTEM ,

Acolicaiblity This specification applies to the bC. ding which houses :he reactor.

Objective The objective is to restrict the amount of radicactivity rem 9 sed into the environment.

Soecifications

a. The reactor building, as a structurally indecei. dent building in the AFRRI complex. shall have its own ventilating system. The effluent from the reactor building ventilating system shall exhaust through absolute filters to a stack having a minimum elevation that is IS' above the roof level of the highest building in the AFRRI complex.

b. The reacter room shall contain a minimum free volu.ae of 22.000 cubic feet.

c. The ventilating system air ducts to the reactor rcom shall be equipped with positive sealing dampers activated by fail-safe centrols that will 25 3

clcse off the ventilation to the reactor room automatically upcn an alarm of the reacter room centinuous air monitor.

d. The reactor room shall be designed to restrict air leakage when the positive sealing dampers are cicsed.

Bases The facility is designed such that the ventilation system will .ccmally maintain a negatile pressure with respect to the atmosphere so that a.-.. will be no ,'

uncontrclied leakage to the environment. Th6 free air volume within the reactor building is confined when there is an emergency shuutown of the ventilation system, 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1-The Director, AFRRI shall be respcnsible for overall facility cperatien.

The Physicist-in-Charge of the Reactor (PIC) will ce respcnsible for day to day operatiens and for determination of applicability cf peccedures, experiment authcrizations, and maintainance cperations. During the absence of the PIC from tne facility for more than one working day, the Director. AFRRI, will designate, another SRO to discharge these respcnsibilities.

6.2 ORGANIZATION .

6.2.1 - The ceganizatien for facility management and operatien shall be as shown in figure 6.0-1

a. Minimum staff when the reactor is not secured shall include:

1. Senior Reacter Cperater (SRO) en call but not necessa rilyen site.

2. Radiation centrol technician cn call

3. Reacter Operator (RO) at controls

4. Another persen within the AFRR1 complex able to car y cut written emergency procecures, instructions of the cperator, ce to summen help in case the cperator becomes incapac:ated.

b. At least ene licensed cpeator shall be at the centrols when the reacter is not secured.

c. An individual that is designated by the licensee as cualified to implement routine raciatica protection procedures shall be present at the facility whenever the reactor is nc* secured or whenever any exneriment or excerimental facility is being servicec.

OS 309 202

d. All ecre alteraticns that eculd affect reactivity cf the eacter sha be supervised by a licensed Senice Reactcc Operator.

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C RI C'"CR CFIRACNC 5'".G7 7ip: e 6.2-L  ?ACI'd""? CRCA C.CCN 6.3 FACILITY STAFF GCALIFICATIONS 6.3.1 - (a) Physicist-in-Charge (PIC)

At the time cf appointment to this pcsition, the Physicist-in-Charge shall have a miniinum of five years of nuclear exterience. He shall have a baccalaureate or higner degee in an engineering er other scientific field. The degee T.ay fulfill up to fcur years of ex er:ence en a ene fer cne basis. Equivalent education er exterience may Oe substituted for a degee. 31ust pcssess an NRC Senic- Reacter Operator license.

(b) Reacter Operaticn Superviser (RCS)

At the time of appcintment to this pcsition, the RCS shall have a minimum of a high schocl diploma cc equivalent and should have four years of nue. ear exter ence. A maximum cf two 'rea-s cf '

exce-ience may be fulfilled by related academic cr technical training en a cne fer cne time basis. 11ust pcssess an NRC Senice Reactcc Cperater iicense.

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(c) Reactor Operator At the time of appointment to this positien, an ir.dividual shall have a high school diploma cc equivalent. Must pcssess an NRC Reacter Operater license.

6.4 TRAINING 6.4.1 - The PIC shall be respcn-ible fcr the facilit'f retraining and replacement

training pecgrams.

6.5 REVIEW AND AUDIT 6.5.1 Reacter and Radiatien Facility Safety Comm. m (RRFSC)

S.5.1.1 Function The Reactor and Radiatien Facility Safety Committee is directly respens-ible to the Directcc, AFRRL This committee will review all radiclegical health and safety matters cencerning the reacter and its associated equipment, the reacter, ccom, the reactor censole, the exnosure rooms, the pneumetic transfer system, the reced off preparation area, the fuel element shipping casks, the reacter fuel and it's stcrage area, in additien to the other AFRRI major radiation sources.

6.5.1.2 Comccsitien _and Cualificatiens (a) The RRFSC shall be compcsed of:

(1) Chairman: Representative of AFRRI Directorate (2) Member: Head, Raciaticn Scurces Divisicn cc Research Preg am Cccedinater (3) Member Head, Radiatien Safety Department (4) Member: PIC, Reacter (5) Member: Radic1cgical Safety Officers ce source facility cperaters from at least three cuiside facilities of which at least one must be a Department of Defense (DOD) facility, at least ene a non-DOD facility, and at least cne a reacter physicist or health physicist.

(6) Other knowledgeable incividuals appcinted by the Chairman, RRFSC.

(~) Voting ad hec members as invited by the Chairman. RRFSC, to assist in review ct particular pecblems.

23 309 204

(b) The minimum qualification for perscns en the RRFSC sna'l be 5 years of professional werk experience in the discipline or specific field they represent. A taccalaureate degee may fulfill 4 years of experience.

6.5.1.3 Alternates Alternate members may be appointed in writing by the RRFSC chairman to serve on a tempcrary basis. No more than two alternates nail ..

participate en a voting basis in RRFSC activities at any cne time.

6.5.L4 Meeting Femuency The RRFSC or a subecmmittee thereof shall meet at least fcur times a calendar year.The entire RRFSC snall meet at least semi-annually.

6.5.1.5 Cuccum A querum of the REFSC for review shall censist of the Chairman er his designated alternate and fcur other members, or alternate members. A majcrity of those present shall be regular members.

6.5.1.5 Review The RRFSC shall review: _

(a) Safety evaluatiens '.~ce (1) changes to precedures, eculpment er systems and (2) tests er e.geriments, conducted withcut NRC approval under provisicas of Section 50.59,10CFR, to verify that such actions did not constitute an unreviewed safety question.

(b) Preposed changes to :recedures, equipment or systems that change the criginal intent er use, and are non-censervative, or these that involve an unreviewed safety questien as defined in Section 50.59 of 10CFR (c) Propcsed tests cc ex eriments wnich are sig.ificantly different from previous apprcved tests cc e.geriments, or these that invcive an unreviewed safety questica as defined in See:icn 5 0.59,10CF R.

(d) Prepcsed changes in technical specifications ce licenses.

(e) Violatiens of appilcable statutes, ccdes, regulations, ceders, technical speci:icaticns, license recuirements, cr of internal peccecures or instructions having nuclear safety siznificance.

29 309 205

(f) Sipificant operating aoncemalities cc variations from normal and cxpected performance of facility ecuipment that affect nuclear safety.

Ig) Events which have been :eperted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC in writing.

(h) Audit reports.

4 6.5.1.7 Audit Audits of facility activities shall be performed under the cognizance of the RRFSC but in no case by the personnel respcnsible for the item audited at least ence every 13 months. Individual audits may be performed by cne individual who need not be an identified RRFSC member. These audits shall examine the operating records and encomcass.

(a) The ccnfermance of facility cperation to the technical specifi-ections and the license.

(b) The perfcrmance, training, and qualifications of the entire facility staff.

(c) The results of all actions taken to correct deficiencies cccur-ing in facility equipment, structures, systems, ce methods of cperatica

~ ~ - - ~~

that affect nuclear safety.

(d) The facility emergency plan and implementing precedures.

(e) The facility security plan and implementing precedures.

(f) Any other area of facility operatien censidered appropriate by the RRFSC ce the Director.

5.5.1.3 Authcrity The RRFSC shall repcet to the Directcc. AFRRI. and advise the ?!C en these areu of responsibility soecified in secticns 6.5.1.5 and 5.5.1.7.

S.S SAFETY LIMIT VIOLATIONS 6.5.1 The fclicwing actions shall be taken in the event a safety limit is exceeded:

(a) The reacter will be shutdown immeciately and reacter cperaticn will not be resumed without authcrication by the Commission.

30

)

(b) The safety limit violatien shall be reported to the Director of SRC Region 1 Office of Inspection and Enforcement (cr his designate),

the Director, AFRRI, and to the RRFSC not later than the next working day.

(c) A Safety Limit Violation Repcet shall be prepared. This report shall be reviewed by the RRFSC, and shall describe (1) applicable circumstances preceding the violation, (2) effects of the violatien upon facility components, structures, er systems, and (3) correctiv acticn taken to prevent recurrence.

(d) The Safety Limit Violation Repcrt shall be submitted to the Commission, the Directcc, AFRRI, and ne RRFSC within 14 days of the violation.

5.7 PROCEDURES 6.7.1 There shall be written cperating procedures that cov the following activities. They shall be approved by the Chief, Radiation Sources Division (RSD).

(a) Cenduct of ir adiations and experiments that could effect the cperation and safety of the reactor.

(b) Startuo, cperation, and shutdowns cf de reactor.

(c) Fuel movement and changes to the core and exteriments that can affect the reactivity.

(d) Preventive ce ccerective maintenance which could have an effect on the safety of the reactor.

(e) Surveillance, testing, and calibraticn of instruments, compenents, and systems invclving nuclear safety.

(f) Review and approval of changes to precedures.

(g) Personnel radiation protection censistent with 10CFR20.

(h) Implementation of the Securit/ Plan and Emergency Plan. .

(i) Administrative centrol of cperation and maintenance.

S .7.2 Though substantive changes to the acove precedures : hall be made only with approval by the Chief, RSD, tempcrary enanges to the procedures 2at do not change their original intent may be made by the PIC. All suen tampcracy changes shall be dccumented and subsecuently approved by the Chief, RSD within 14 days.

21 309 207

6.3 EXPERDIENTS 6.3.1 Before issuance of a reacter authcrization, experiments shall be reviewed by the following:

(1) Physicist-in-Charge (PIC) of the reacter (2) Head, Radiation Sources Division (3) Safety Department ..

(4) Reacter and Radiation Facility Safety Committee (RRFSC)

S.3.2 Prict to its perfccmance, an experiment must be included under cne of the following types of authorizations issued by the RRESC:

(1) Scecial Reacter Authcrization for new experiments cc experiments not included m a Reut:ne Reactor Autncrization. These exneriments shall be performed under the direct supervisien of the PIC of the reactor er his designee.

(2) Routine Reactor Authorization for exoeriments safely performed at least once. These experiments may oe performed at the ciscretion of the PIC and ecordinated with the Safety Department when appeceriate.

6.3.3. With the exception of the reutine measurement of reactor parameters, the reactor cperator snall operate the reactor under the specific authorizaticn of the P!C ce his designee. The Reacter Operations Superviser may authce: e 2e routine measurement of reactor parameters.

6.9 REPORTING REQUIRE.MENTS In additicn to the acplicable repceting requirements of Title 10, Ccde of Federal Regulations, the following reports shall be sucmitted to the Direetcc of the Appropriate NRC Regicnal Office unless otherwise noted.

6.9.1 Rcutine Reccets

a. Startuo Reccet- A summary repcet of plan startup and power escalat:en test:ng snall be suomitted fol'owing: (1) receipt of an operating license; (2) amendment of the license involving a planned increase in power level; (3) installaticn of fuel that has a different design; and (4) modifications that may have significantly altered de nuclear, thermal, or hydraulic performance of the plant. The report

'shall address each of the tests ;dentified :n Se FSAR anc shall in general include a dese iption of the measured values of the cperating conditicns ce charactet:stics obtained during the tests cccgsm and a ccmpariscn of these values with design ;redictions and specificaticts.

Any ecrrective actions that were required to 'cotain satisfactcry cperatien shall also be described. Any additional Ipecific deta!!5 required in license ccnditiens based en other commitments snall ce included in this report.

32 3

Stc-tup repcets shall be submitted within (1) 90 days follcwing completion of the startup test pecgram. (2) 90 days following resumption or commencement of power operaticn. (3) 3 months following initial c-iticality, wnichever is earliest. If the Startup Repcet does not cover all three events (i.e. initial criticality, completien of startup test pecg am, and resumption or commencement of power cperation), supplementary reports shall be submitted at least every three months until all three events have been completed.

Rcutine operating repcrts covering the

b. Annual Ooersting Recort-cperatien of tne umt curing the previcus calendar year should be suomitted prior to October 31 cf each year, covering the previcus fiscal year's cperation.

The annual operating reports made by licensees shall provide a comprehensive summary of the cperating excerience having safety significance that was gained during the year, even thcugh some repetition of previously reported infcrmation may be involved. References in the annual operating repcet to previcusly submitted reports shall be clear.

Each annual cperating report shall include:

(1) A brief narrative summary of (a) Changes in facility design, performance characteristics and coerating procedures related to reactor safety, that occur ed during the repceting period.

(b) Results of major surveillance tests and inspections.

(2) A tabulation showing the energv generated by the reactor en a monthly basis.

(3) List of the unscheduled shutdowns, inclucing the reascns therefore and ecerective action taken, if any.

(4) Discussien of the majcr safety related corrective maintenance performed during the pericd, including the effects, if any, on the safe operatien of the reacter, and tha reasons for the corrective maintenance rec,uired.

(5) A brief descriptien of:

(a) Each change to the facility to the extent that it changes a descripticn of the facilit';in the Safety Analysis Repcrt.

(b) Changes to the procedures as described in the Safety Analysis Repcet.

33 309 209

(c) Any new cr untried experiments or tests perfcrmed during the reocrting period that are not described in the Safety Analysis Report.

(6) A summary of the safety evaluation made fer each change, test. cr experiment not submitted for Commission approval pursuant to 10 CFR 50.59 which clearly shows the reascn leading to the conclusica that no unreviewed safety # ;stien existed and that no technical specification change was required. '

(7) A summary of the nature and amount of aircorne radioactive effluents released er discharged to the environs beyond the effective centrcl of the licensee as determined at er price to the point of such release or discham.

(a) Total estimated quantity cf radicactivity released (in curies) determined by an approved sampling and counting method.

(b) Total estimated quantity of Argen-41 released (in curies) during the repceting period based on data from an appropriate menitoring system.

(c) Estimated average atmcspherie diluted concentratien of Argen-41 released during the reperting period in terms of microcuries/cc and fracticn of the applicacle MPC value.

_ u.. _

_ (d) Total estimated quantity of radicactivity in particulate form with half lives creater than eight days (in curies) released during the repceting period as determined by an appropriate particulate menitoring system.

(e) An es+imate of the average concentratien of other significant radionuclides present in the gaseous waste dischargt in terms of microcuries/cc and fracticn of the applicable MPC value for the reporting period if the estimated release is greater than 20% cf the applicable MPC.

(8) A description of the results of any environmental radiclegical surveys performed cutside the facility.

6.3.2 Recortable Occur ences Repertable cccurrences, including causes, probable censequences, ccrrective acticns and measures to prevent recurrence. shall be repcete.d to the NRC.

Supplemental repcrts may be required to fully describe the .'inal resciution of the cccurrence. In case of corrected er succlemental repcets. In amended licensee event repcrt shall be completed and reference shall be made to the criginal repeat date.

34 309 210

a. Promot Notificaticn With Written Fellowuc. The types of events !!sted below shall be reportec as expediucusly as possible by telephene and confirmec by telegaph, mailg am, er facsimile transmission to the Director of the apprcpriate NRC Regional Office, er his designate no later than the first week day following the event, with a written followup epert within two weeks. The written followup report shall include, as a minimum, a completed copy of a licensee event repcet form. Informatien provided en the licensee event repcet form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrcunding the event.

(1) Failure of the reacter protection system, er cther systems subject to limiting safety system settings, to initiate the required pec-tective functicn at the time a monitored parameter reaches the setpcint specified as the limiting safety system setting in the technical specificatiens or failure to complete the required pro-tective functicn.

(2) Operaticn of the reacter er affected systems when any parameter ce cperation subject to a limiting concitien is less conservative than the limiting conditicn fcr operatien established in the technical specificaticns withcut taking permitted remedial action.

(3) Abnormal deg adation discovired in a nssien product bar-ier, i.e.,

fuel cladding, reactor ecclant boundary, ce reacter building. ,

(4) Reactivity balance anomalies invciving:

(a) disapeement between excected and actual critical positions of approximately 0.3% Ek/k; (b) exceeding excess reactivity limit; (c) shutdown margin less censervative than scecified in the technical specificatiens; (d) unexoected shcrt-term reactivity changes that cause a positive period of 10 seconds er less; (e) if sub-critical, an unplanned reactivity insertien cf mere than approximately 0.5% S</k or any unplanned criticality.

(5) Failure er malfunction of cne er mere components which prevents ce could prevent, by itself, the fulfillment of the functicnal require ments cf system (s) used to ecpe with accidents analy::ed in the SAR.

35 309 n\\

(6) Personnel errer er peccedural inadequacy which prevents, ce cc Id prevent.cy itself, the fulfillment of the functicnal requirements of systems required to ecpe with accidents analyzed in the SAR.

(7) Unscheduled cencitions arising from natural ce man-made events that, as a direct result of the event, require cperation of safety systems or other protective measures required by technical specifications.

(S) Errors discovered in the transient er accident analyses cc in the metheds used for such analyses as described in the safety analysis repcrt er in the bases for the technical specificaticns that have er could have permitted reactor cperation in a manner less cen-servative than assumed in the analyses.

(9) Performance of structures, systems, cc ecmponents that requires remedial action or ccerective measures to prevent cperation in a manner less censervative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of ccnditiens not specifically censidered in tne safety analysis report ce technical specifications that require remedial acticn ce ccerective measures to prevent the existence er develcpment of an u'Isafe cendition.

6.10 RECORD RETENTION C0.1 Reccrds to be retained for a period cf at least five years:

(a) Operating Icgs or data which sha11 identify:

(1) Completien of pre-startup checkout, startup, pcwer changes, and shutdown of the reacter.

(2) Insta11atien er removal of fuel elements, cent ci reds, er exneriments that could affect ccre reactivity.

(3) Insta11aticn er removal of jumpers, special tags, er notices or other temporary changes te bycass reacter safety circuitry.

(4) Red wceth measurements and other reactivity measure.q ents.

(b) Principal maintenance operaticns.

(c) Re;cetable :ccur ences.

~

(d) Surveillance activities required by technical specificatiens.

(e) Facility radiatien and centaminatien surveys.

(f) E:ceriments perfccmed wit", the reacter.

36 309 212

This requirement may be satisfied by the ncemal cperations 'eg bcci< plus:

(1) Records of radicactive material transferred frem the faciltiy as required by license.

(2) Reeceds required by the RRFSC for the performance cf new ce special experiments.

(g) Chenges to cperating peccedures. ,

6.10.2 Records to be Retained for the life of the Facility.

(a) Gaseous and liquid radicactive effluents released to the envirens.

(b) Appropriate eff-site environmental monitoring surveys.

(c) Fuel inventories and fuel t ansfers.

(d) Radiation ex osures fcr all persennel.

(e) Updated asau- t drawings cf the facility.

(f) Reccrds of transient er cperatienal cycles fer these components designed for a limited number of transients er cycles.

(g) Reeceds of training and q talificatien for members of the facility staff. -

(h) Reeceds of reviews performed for changes made to peccedures er equipment er reviews of tests and experiments pursuant to 10 CFR 50.59.

(i) Reeceds of meetings of the RRFSC.

w 37 309 20

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