ML19254D392

Tech Specs Amend 2

ML19254D392
Person / Time
Site:	Berkeley Research Reactor
Issue date:	09/28/1979
From:	Office of Nuclear Reactor Regulation
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ML19254D389	List: ML19254D388 ML19254D390 ML19254D392 ML19254D394 ML19254D395
References
NUDOCS 7910250426
Download: ML19254D392 (41)

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'.

Appendix A FACILITY LICENSE NO. R-101 TECHNICAL SPECIFICATIONS FOR THE TRIGA MARK III SERKELEY RESEARCH REACTOR

+

OF UNIVERSITY OF CALIFORNIA BERKELEY, CALIFORNIA DOCKET NO. 5 0-224 A=endment No. 2 Dated: September 28, 1979 79.10250

+er sa 1209 161 t

TABLE CF CONTENTS 1.0 Definitions Pace 1.1 Reacter Shutdewn 1

1.2 Reacter Secured 1

1.3 Reac:cr Cperatica 1

1.4 Cold Critical 1

1.5 Steady State Mode 1

1.6 Pulse Mode 1

1.7 Square Wave Mode 2

1.8 Shutdewn Margin 2

1.9 Abnormal Cccurrence 2

1.10 Repertable occurrence

~

2 1.11 Experiment 2

1.12 Experimental Facilities 2

1.13 Safety Red 3

1.14 Shim Rod 3

1.lS Transient Red 3

1.16 Regulation Rod 3

1.17 Fuel Sle=ent 3

1.18 Instrumented Element 3

1.19

.Cere Lattice Positien 3

1.20 Standard Core 3

1.21 Mixed Core 4

1.22 Cperational C re 4

1.23 Safety Limit 4

1.24 Limiting Safety System setting 4

1.2S Cperable 4

1.26 Reacter Safety Systems 4

1.27 Experi=ent safety Systems 4

1.28 Measured value 5

1.29 Measuring Channel 5

1.30 Safety Channel 5

1.31 Channel Check 5

1.32 Channel Test 5

1,.33 Channel Calibration 5

2.0 Safety Limits and 7'-4*ine safety System ca*e<ng, 6

2.1 Safety-Limit Fuel Element Tcaneratures 2.2 Li=i ting Safet y Sys:em Setting s in Stead y State Mode of 6

Op eration 3.0 Limitine cend:tions :cr operation 8

3.1 Steady State and Cquare Wave Mede Cperation B

3.2 Reactivity Limitatien 8

3.3 Pulse Mede Cperation 9

3.4 Cen r01 and Safety System 10 3.4.1 Scram Time 10 3.4.2 Rese::: Safety System 10 3.5 Radiation M: nit: ring System 11 3.5 Argen-41 2ischarge Limit 13 1209 1o,2

Pace Engineer d Safety Feature - Ventilation System 14 3.7 e

3.8 Linitatiens en Experiments 14 3.9 Irradiations 16 4.0 Serveillance Recuirements 18 4.1 General 18 4.2 Limiting Cenditions fer Cperatien 18 4.2.1 Reactivity Requirements 18 4.2.2 Centrol and safety System 19 20 4. 2. ~,

Radiatien Menitcring system 20 4.2.4 Ventilatien system 4.2.5 Experiment and Irradiatien Limits 21 4.3 Reactor Fuel Elements 22 23 5.0 Oesien Features 5.1 Reactor ruel 23 5.2 Reactor Core 24 24 5.3 centrol Rods Radiatica..cnitering system 25 w

5.4 26 5.5 Fuel Sterage 5.6 Reacter Building and Ventilation system 27 27 5.7 Reactor Pool Water systems 5.8 Reacter suppert Bridge 2S 29 5.9 Reactor Auxiliary Electrical S ystem 6.0 Administrative Centrols 30 30 6.1 Crganization 6.2 Review and Audit 30 6.3, Actien To se Taken in the Event a safety 32 Limit is Exceeded 6.4 Action To Be Taken in the Event of a 33 Reportable Occurrence 6.5 Operating Procedures 33 6.6 Facility Cperating Records 34 6.7 Reperting Requirements 34 120L9 163

100RORGLL Included in this document are the Technical Specifications and the "3ases" for

he Technical Specifications.

These bases, which previde the technical support

the individual technical specifications, are included for inferratien pur-puses only. They are not part of the Technical Specificati:ns, and they de not cens:itute limitations er requirements to which the licensee =ust adhere.

Refer-e..ce NRC Regulatory Guide 1.16 and ANSI N378-1974.

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FIACZR CFE3ATING CONDITICNS 1.1 REAC':CR SHUTDOWN The reactor is shut dcwn when the reacter is suberitical by at least ene dollar of reactivity.

1.2 FIACTOR SECUFID The reactor is secured when all the fellcwing cenditiens are satisfied:

a.

The reactor is shut down, b'.

The console key switch is in the "eff" ocsition and the key i s removed from t he con sole and stored in a lodc ed cabinet, and c.

No werk is in prog:ess involving in-core fuel hancling or refueling cperatiens, maintenance of the reactor er its centrol mechanisms.

1.3 REACM R CPERATICN Reacter operatien is any cendition wherein the reacter is net secured.

1.4 COLD CRITICAL

. The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures both below 40cc with equilibrium samarium present.

1.5 STIAOY STATE MCDE Steady state : de cperati:n shall rean aperatien of the reacter with the : de select:r swit:h in the steady state position.

.6 v._ _: _

.v.

Fulse =cde cperatien shall mean any cpera.icn of the reacter with the mcde select:r switch in the pulse pcsiti:n.

i2LD9 i64

?00RORBNAL 1.7 ECUApr WAvr MCCE Square wave nede operaticn shalJ mean any cperatien of the reacter with the nede selector switch in the square wave nede positicn.

1.8 SHU*DCWN MARGIN.

Shutdewn margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reacter can be made suberitical by means of the centrol and safety systems, starting fr:m any permissible operating conditiens and that the reacter will remain suberitical withcut further cperator action.

1.9 A3NCKv.A* CCCURRENCE An "Abncrmal occurrence" is defined fer the pu goses of the reperting requirements of Secticn 208 cf the Energy Recrganization Act of 1974 (P.L.93-438) as an unscheduled incident er event which the Nuclear Regulatory Ccmmissicn determines is signifi-cant f cm the standpoint of public health or safety.

1.10 REPCR~'A3LE CCCURRENCE A repertable occurrence is any of the following which occurs during reacter operatien:

a.

Operation wit h any saf ety system setting exceeding tint spe cified in Section 2.1.

b.

operatien in violation cf a Limiting Cendition for Operation; Failure of a required reacter er experiment safety system c.

ccmpenent which eculd render the system incapable of per-for=ing its intended safety function; d.

Any unanticipated or uncentrolled change in reactivity greater than ene dollar; An observed inadequacy in the implementation of either e.

administrative er precedural centrols,.such that the in-adequacy could have caused the existence er develcpment of a condition which could result in cperaticn of the reacter cutside the specified safety limits: and f.

Release of fission products fr = a fuel element.

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Ex eriment shall mean (a) any apparatus, device, er material which is net a normal, cart c' the c re Or experimental facilities, but whi:5 is inserted in these facilities : is in line with a :eam cf radiation criginating fr:m the reacter ::re; c

'b) any :perati:n desi:ned :: easure react:r arameters ::

1209 I0'"3

naracterist:rs.

?00RORCNAL

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.+..

I'xperimental f acilities shall mean beam ports, including extensien tubes with shields, thermal eclumns with shields, hchlraum, central tht:ble and other incere irradiation faci); ties, exposure recm, rotary specimen rack and pneumatic transfer systems.

FI.AC-CR CC.vPCf;*:;TS 1.13 SAFETY RCD The safety rod is a centrol red having an electric meter drive and scram capabilities.

It may have a fueled fellcwer section.

1.14 SHIM RCD The shim red is a centrol red having -

electric meter drive and sers= capabilities.

Its positica may :e varied manually er by a serve-centroller.

It may have a fueled follower secticn.

1.15 TRA :S E := RCD The transient red is a centrol red with scram capabilities that can be rapidly ejected f.cm the reactor core to produce a pulse.

It may have a voided fellower.

_r e. _n.

..!v. -r_D 3.36 n

r oa The regulating rod is a c:ntrol red that has scram capability and may have a fueled fellower.

Its positien may be varied manually 0

er by the servo-centrc13er.

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v A fuel element is a single TRICA fuel rod of either 8.5 wtt er 12 w:s standard type, i.e., 20% enriched in uranium-235.

1.18 INSTR!*ME '"E3 EI,F.v.Er An instrumented element is a special fuel element in which at least ene thermocouple is embedded in the fuel.

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The ecre lattice positien is that regien in the core n:rmally cc:upied by a fuel-element, a control red, an experiment, a refle:::: eierent, er an icnchamber.

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e.

. ar-n A standarf c:re ::ntains an arrangement of standard TRICA fuel in the react:r grid piste.

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1q09 1o,6 c

a P00R ORG NAL 1.21 MIXED COEI A mixed core contains an arrangement of standard 8.5 wtt and 12 wts TR A fuel elements.

1.22 CPEFA.TICNAL CCF2 An operatienal core may be a standard core, er mixed core, for which the ccre parameters of shutdown margin, fuel temperature, pcwer calibratien, and maximum allewable reactivity insertien have been determined to satisfy the requirement of the Technical Specifications.

REAC*CR INSTR"MINTATION 1.23 SA.*ITY LIMIT Safety limits are limits en important process variables which are found to be necessary to reasenably protect the integrity of certain of the physical barriers which guard against the uncontre.i d release of radicactivity.

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c.. u Limiting safety systems 2 -tting is a setting fer autematic pro-tective devices relatec to these variable s having significant safety functions.

1.25 CPERA3LE A system, devicc, er ccmpenent shall be censidered operable when it'is capable o perf rming its intended func*.lons in a normal manner.

1.26 REAC*CR SAFE"'Y SYS* EMS Reacter safety syste=s are these systems, including their associ-ated input circuits, which are designed to initiate a reactor scram fer the primary purpcse cf protecting the reactor or to p;cvide infer =atien which requires manual protective action to be initiated.

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xceriment safety systems are these systems, including their assecteued iaput circuits, which are designed to initiate a scra= fer the primary.rurpcse cf protecting a. experiser.: er

pr: vide inf rmatien whi:h requires manual pr:tective actien

be initiated.

1209 167

_4

P00RDnlNn 1,29 MIASUFID VALUE The measured value is the magnitude cf that variable as it appears on the eu.put of a measuring channel.

1.19 PIASURING CHA!! NIL A measuring channel is the ccmbinatien of senser, intercennecting cables er lines, amplifiers and other necessary eaectrical, mechani-cal and cutput devices which are cennected for the purpose of measuring the value of a variable.

1.30 sATI':Y CHANNEL A rafety channel is a measuring channel 'n the reacter safety system.

1.31 CHA!: NIL CHICX A channel check is a qualitative verification of acceptable performance by ebservatien of channel behavior.

1.32 CHANNIL TEST A channel test is the introduction of a signal into the channel to verify C.at it is coe:able.

1.33 CHANNEL CALISTATICN A channel calibration censists of ecmparing a measured value frem the measuring channel with a correspending known value of the parameter so that the measuring channel output can be adjusted to respond with acceptable accuracy to known values of the measured variable.

12.09 168

_3_

2.0 SAFE-'Y LIMITS A

-) LIMITING SArr'Y SYSTEM SE !NGS 2.1 SAFE-*Y LIMIT FUEL ELEMENT TEMPEFA""JFI Applicability This specificatien applies to the temperature of the reacter fuel.

Cbjective i

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.

Specifirations The temperature in a standard 8.5 wtt and 12 2-1 TRIGA fuel element shall not exceed 183CCF (ICC;oC) under any e nditiens cf operatien.

Bases

.. ~..

The important parameter for a TRIGA reacter is the fuel element te=perature.

This parameter is well suited as a single specifi-cation especially since it can be measured. A less in the integ-rity of the fuel element cladding could arise frca a build-up of excessive pressure between the fuel-mederator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fissien product gases, and hydrogen frem the dissociatien of the hydregen and irceniu=

in the fuel-moderater. The magnitude of this pressure is deter-mined by the fuel-moderator temperature and the ratic of hydrogen to zircenium in the alley.

Experiments perfor=ed by G. A. have shown that the measured pressure rise was only 24 psi at a calcu-lated te==erature of 11000C.(1)

I f

If we assumed that the cladding temperature is in equilibrium with,\\'

the fuel = eat, this pressure of 24 psi is far belew the rupture pressure of 450 psi at 8CCCC. (2) of ccurse it is unreasonable to f' expect that the cladding will be as het as the fuel = eat.

Calcu-lation shows (2) that the maximum clad temperature is about 2/3 i

of the maxi =um fuel te=perature.

Thus, even at a calculate. fuel temperature of 11CC C, the cladding does not exceed 7330C.

C 2.2 LIMITING SAFETY SYSTEM SETTINGS IN STEADY STATE MODE OF OPERATION

-. li cabi li t y Apo The specification applies to the alarm settings which indicate that the safety limit has beer. rssjhed during steady state and square wave modes of operation.

Objective The objective to prc ide an alarm so that operator action _an be e

taken to prevent the safety Jimit fuel element temperature of 1830*F" (1000*C) from being reached.

P00R ORGINAL

+

Specification The limiting safety system setting shall be 930*F (500*C) as measured in a B or C-ring instrumented fuel element.

Basis The limiting safety system setting is a temperature which, if exceeded, shall cause an alarm to be initiated prior to the safety limit being exceeded. A setting of 930*F (500*C) provides a safety margin of 900*F (500*C) for the standard TRIGA fuel elements. A part of the safety margin is used to account for the dif ference between the true and measured temperatures from the actual location of the thermocouple.

Locating the thermocouple in either the B or C-rings will reduce the difference to a small percentage. Calculatiore-have shown that if the thermocouple element were located on the peri-hery of the core, the true temperature at the hottest location wou':d differ from the measured temperature by less than a factor of two.

Therefore, placing the ther=ocouple in the B or C-rings will ensure that when the alarm setting of 930*F (500*C) is reached, there is a sufficient safety margin to permit operator action prior to reaching the safety limit.

(1) 3.

S'.

West, et al, " Kinetic 3ehavior of CRIGA Reacters", Ger.eral Atcmic Publication, GA-7882, March 31, 1967, page 2.

(2) C. C. Cof fer, et, " Characteristics of Large Reactivity Insertien in a High Perfcrmance TRIGA U-CrH Core", General Atcr.ic Fuhlication, GA-6216, April 12, 1965, pp. 23-26.

200RORGNAL 1209 170

3.0 LIMITING CCNOITICNS FOR CFERATION 3.1 S EADY STATE A!c SCCARI WAVE MODE OPERATICN Apelicability This specificatien applies to the energy generated in the reacter during steady state or square wave =cde cperation.

Cbiective The objective is to assure that the fuel temperature safety limit will not be exceeded during steady state er square wave ecde operatien.

Specification The reacter power level shall net exceed 1.3 megawatts under steady state er square wave mode operatien. The n:rmal steady state er square wave mode operating power level cf tha reactor shall be 1.0 megawatts.

However, for purposes of tc. sting and calibration, the reactor =ay be cperat ed at higher power levels not to exceed 1.3 megawatts during the testing period.

Bases Thermal and hydraulic calculations indicate that TRIGA fuel may be safely cperated up to power levels of at least 2.0 megawatts with natural convection cooling.--

3.2 FCACTIV!'"Y LIMITATICNS Apolicability These specifications apply to the reactivity ccndition of the reacter and the reactivity worths of ecntr:1 rods and experiments.

They apply for all modes of cperation.

Chiective

'The cbjective is to assure that the reacter can be shut down at all times and to assure.that the fuel te=perature safety limit wt11 not be exceeded.

Specificatien The reactor shall not be operated unless the shutdown margin previded by centrol rods shall be greater than 0.25 dollar with-the highest wcrth experiment in its = cst reactive state, a.

b.

the highest worth centrcl red fully withdrawn, the reacter in the cold critical conditien without xenen, c.

d.

the naximum excess reactivity above cold, clean critical plus samarium poisoning shall be 4.9% ak/k, and the maximum rate of reactivity insertion associated with novement e.

of a standard control rod shall be no greater than 0.15% ak/k per second.

P00R ORGINAL

+

Eases The value of the shutdown margin assures that the reactor a.

can be shut down frem any operating condition even if the highest worth centrol red should remain in the fully with-drawn position.

3.3 PULSI HCDI CPERATICN Applicability This specificatien applies to the enerpy generated in ene reactor as a result of a pulse insertien of reactivity.

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

Specificatien The reactivity to be inserted for pulse operation shall be a,

determined and limited by a mechanical step en the pulse red, such that the reactivity insertion will net exceed 3.0 dollars

('2.1% Ak/k),

Sis rod CQ one pulsing centrol rod nay be used in the core.

b.

shall contain al" dm:= or stainless steel clad, berated g-aphite poison.

Se pulse rod s M 1 be designed to release and fall upon initiation of a scrs= signal. Se w'-

verth of the poison section vith respect to vater shan be 2 9% Ak/k.

Se peak' neutron flux sha n be recorded for every pulse.

Peak c.

power levels durits pulsing that exceed 2600 t>egavalts shan be investigated to deter =ine the reason for the pulse na5nitude.

Conclusiens shall be sub=itted to the F.eacter Ea..ards Cc-4ttee for evaluation.

Fulsing vin be discontinued until resu=ption is approved by the C-- *ttee.

Bases Measurements perfcr:cd en the Berkeley Research Reactor TRIGA MARK III reacter using 8.5 wts fuel indicated that a pulse insertion of re-activity of 3.0 dollars resulted in a maximum temperature rise of approximately 45CCC. With an ambient water tenperature of approxi-mately 25c0, the maximum fuel temperature would be apprcximately 475 O resulting in a safety margin of 525cc fer standard 8.5 wts C

fuel. The maximum calculated temperature rise.n a 12 wtt fuel, underyteconditiensdescribedinSectica5.1isapprcximately 650 C.

The safety targin in this case is approximately 325cc, C

These margins allow a= ply for uncertainties due to the accuracy cf measurement er locatien of the instrumented fuel element er due to the extrapolation cf data frem the ERR reacter.

_9

3.4 CONTROL AND SAFETY SYSTEM 3.4.1 Scram Time Acclicability This specification applies to the time required fer the scram-mable control rods to be fully inserted frem the instant that a safety channel variable reaches t.5e Safety System Setting.

Chiective The objective is to achieve prempt shutdown of the reactor to prevent fuel damage.

Seecification The scram time measured from the instant a simulated signal reaches the value as described in Table I to the instant that th'e slowest scrammable control roc reaches its fully inserted position shall not exceed 2 seconds.

Sasis This specificatien assures that the reacter will be premptly shut down when a scram signal is initiated.

Experience and analysis have indicated that for the range of transients anti-cipated for a TRIGA reacter, the specified scram time is adequate to assure the safety cf the reactor.

3.4.2 Reacter Safety Svstem Arelicability This specificatien applies to the reactor safety system channels.

i l

n, W.

Freicff, " Analysis of Partial Refueling with 12 w:S Cranium Fuel fer the Berkeley Research Reacter", M.S Thesis, Department of Nuclear Engineering, Universi:, cf Califernia, Serkeley, 1977.

200R ORGINAL 1209 173

~

Cbjective The cbjective is to specify the minimum number of reactor safety system channels that must be cperable for safe operation.

Specification The reactor shall net be cperated unless the safety channels described in Table 1 are operable.

Bases The power level scrans provide protection to assure that the reacter can be shut down before the safe y limit en the fuel element temperature will be exceeded.

'.e manual and the r

magnet current key switch scrs=s allow the operator to shut down the system if an unsafe or abncrmal cec.dition occurs.

In the event of failure of the power supply for the safety chambers, cperatien of the reacter without adequate instrumentation is prevented. The preset timer insures that the reactor power level will reduce to a icw level after pulsing.

The interlock to crevent startup of the reactor at power levels less than 4 x 10-D watts, '<hich correspends to approximately 2 cps, assures that sufficient neutrons are available to initiate a self-sustaining chain reaction.

The interlock to prevent the initiation of a pulse above 1 kW is intended to assure that the magnitude of the pulse will not cause the fuel element te=perature safety limits to be exceeded. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted, is intended to prevent pulsing the reacter in the steady state mode. The interlock to prevent withdrawal of the shim, safety or regulating red in the pulse mode is designed te prevent changing the critical state of the reactor just before pulsing.

The earthquake scram will trigger a.shutdcwn upcn receipt of a herirental acceleration.

3.5 RADIATICN MONI'"CRING SYS*"FM A plicabilitv This specificati:n applies to the radiation menitoring infer =ation which must ce available to the reacter cperator during reacter cperation.

Chiective The cbjective is to assure that suffient radiation menit: ring inf:rmatica is available := the cperater to assure safe cperatica ci the react:r.

5:erificatien The reac:Or shall nct be Operated unless the radiati:n renit ring

hanne_s listed in the fell: ine ta:1e are cperable.

1209 174 TA2LE 1 Minimum Reactor Safety Channels Number Effective Mode Safety Channel Operable Function S.S.

Pulse Sc.w.

~

Linear 1

SCRAM @ 110".

X X

(Power Level)

Satety (Power Level) 1 SCRAM @ 1.1 megawatt X

X Console Scram Button 1 SCRAM X

X X

Log Power 1

SCRAM @ < 3 secondr X

Lineir and Safety 1

SCRAM en loss of X

X X

Detector supply voltage Power Supply Preset Timer 1

Transient red scram X

15 seconds or less after pulse r

Startup Channel 1

Prevent red withdrawal '

X at less than two neu-tron induced counts per secend Log Power 1

Prevent pulsing X

'above 1 kW Transient Red 1

Prevent applicatien X

Position of air unless fully inserted Red Drive Control 1

Prevent withdrawal X

of any rod except transient rod t

Fuel Element 1

Alarm @ 500*C X

X Temperature Earthquake Detecter 1

SCRAM X

X X

Macnet Current 1

SCRAM X

X X

Key Switch Pool Water Level 1

2 feet above top of X

X X

grid bridge Rod Drive Control 1

Prevent simultaneous X

X manual withdrawal of 2 rods Fool Bulk Water 1

Alarm @ > 120*F X

X X

Temperature 17"'

P00ROR:GINAL

- =-

s Radiation Monitcrine Channels Function Number

• Area Radiation Moniter*

Monitor radiaticn levels 1

within the reacter recm Area Radiatien Moniter*

Moniter radiatien levels 1

around reacter bay Continuous Air Particulate Moniter cencentration of 1

Moniter**

radioactive particulate activity within the reacter rocs Exhaust Gas ?

Ation M,9 niter

• Moniter radiatica levels 1

in the exhaust air stack Eases The radiatien monitors provide informatien to cperating persennel of any i=pending er existing danger frem radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread et radioactivity to the surroundings.

3.6 ARGCN-41 DISCEARGE LIMIT Applic bility This specification applies to the concentration of Arger *1 that may be discharged from the TRIGA reacter facility.

Cbjective To insure that the health and safety of the public is not endan-c.ered by the dischar e of Argen-41 frc= the TRIGA reactor facility.

m Specification

'The concentratien of Argen-41 at the facility stack exhaust shall not exceed 3.12 x 10-6 pCi/ml averaged over one year.

'Ter perieds of time fer maintenance to the radiation menitcring channels, the intent of this specificatien will be satisfied if they are replaced with pertable ga= a sensitive instruments having their cwn alarms er which shall be kept under risual observatien.

='rcr perieds of tire for maintenance te the radiation menitoring channels, the intent of this specificati:r w;11 he satisfied if they are replaced with portable beta-gamma sensitive instruments having their cwn alar s er which shall be kept ender visual cbser-eation. )DDRORGNAL 3209 176

Bases-As shown in the SAR, the continucus emission at the stack exhaust for the 2.4% ci provides a yearly dese of 0.13 rem in,still air, the time in which calm weather occurs. Also shown is that at the point of nearest west wind habitation, 100 m from the stack, on the centerline of a plume whose diameter of 40 m provides the largest possible dose, the same centinueus emissien would provide a yearly dose of 0.45 rem, for the 29% of the time that the westerly winds prevail. No credit is taken in the latter calcu-latien for plume dispersion or for finite wind velocity.

Thus, in the case of either still air er of prevailing westerlies, a person in continuous residence at the worst locatien eculd not receive more than the yearly dose allcwed under 10 CFR 20.

3.7 ENGINF. FRED SAFETY FEATURE - VEN"'ILATICN SYSm Applicability This specification applies to the operatien of the facility ventilatien system.

Chiective The objective is to assure that the ventilation system is in operatien to mitigate the censecuences of the possible release of radioactive materials resulting frem reacter operation.

Specificatien The reactor shall not be operated unless the normal ventilation system and glove box scrubber exhaust system are operable except for periods of time necessary to permit repair of the system.

In the event of a sub-stantial release of airborne radioactivity, the ventilation system will be secured automatically by a signal from an exhaust air radiation monitor.

Bases In the event of a substantial release of -airberne radicactivity, the ventilation systems will be secured automatically.

Therefore, operatien of the reacter with the ventilation systems shut dcwn for short perieds of time to make repairs insures the same degree cf centrcl of release of radicactive materials. Merecver, radi-atien menitors within the roc =, independent of these in the venti-latien systems, will give warning of high levels of radiation that might cecur during cperatien with the ventilatien syste=s secured.

3.S LIMITAT:CNS ON E:GERIME!. S T

Applicability This specificatien applies to experiments installed in the reacter and its experimental faci.ities.

P00R ORGINAL

~"~

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

Specifications The reacter shall not be operated unless the f=11cwing conditiens governing experirents exist.

The reactivity worth of any single experiment shall be less a.

than 3.0 dollars.

b.

Explosive materials, such as gunpowd., TNT, nitroglycerin, er PETN, in quantities greater than ;5 ailligrams shall not be irradiated in the reacter er exp rimental facilities.

Explosive =aterials in quantities lass than 25 milligra=s may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

c Experiment materials, except fuel materials, which could off-gas, sublime, volatilice, or produce aeresols under (1) normal cperating conditiens ef. the experiment or reacecr, (2) credible accident cenditions in the reactor, er (3) pcssible accident cenditiens in the experiment shall be limited in activity such that if 100% of the gaseous activity er radicactive aeresols produced escaped to the reacter rec = or the atmosphere, the airborne concentration of radicactivity averaged ever a year would not exceed the limit cf Appendix 3 of 10 CFR part 20.

d.

2n calculatiens pursuant to c. above, the following assu=p-tiens shall be used:

(1)

If the effluent frc= an experimental facility exhausts through a heldup tank which closes aute=atically on high radiation level, at least 10% cf the gaseous aerivity or aeresels produced will escape.

(2)

If the efflucnt frc= an experimental facility exhausts threugh a filter installatien designed for greater than 99% efficiency for 0.3 micren parti.les, at least 10% cf these vapers can escape.

(3)

Fc: materials whcse teiling point is abcve 13Ce? and where vapers fer:ed by boiling this material can escape enly thrcugh an undisturbed celumn of water above the cere, et least lot of these espers can escape.

1209 178 -

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e.

Each fueled experiment shall be controlled such that the total inventery of iodine isotepes 131 through 135 in the experiment is no greater than 1.5 curies.

f.

If a capsule fails and releases material which could damage the reactor fuel er structure by cerrosien or other means, removal and physical inspection shall be performed to deter-mine the conseqtences and need fer corrective acticn.

The results of the inspection and any corrective action taken shall be reviewed by the Reacter Superviser er his designated alternate and determined to be satisfactory before operation of the reacter is resumed.

Bates a.

The maxinum worth cf a single experi.ent is limited so that its removal frc= the cold critical reactor will not result in the reactor achieving a pcwer level high enough to exceed the core temperature safety limit.

b.

This specification is intended to prevent damage to reacter c ponents resulting frc= failure of an experiment involving explosive materials.

c.

This specificatien is intended t: reduce the likelihood that airborne activities in excess of.the limits of Appendix B of 10 CFR Part 20.will be released to the atmosphere outside the facility beundary.

d.

The 1.5 curie limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading lto total release of the iodine, the exposure dose at the exclusien area boundary will be less than that allcwed by 10 CFR Part 20 for an unrestricted a ea.

e.

Cperation of the reactor with the reactor fuel or structure damaged is prchibited to avoid release of fission products.

3.9 IFJaDIATICNS This specif'. cation applies to irradiotion performed in the experi-

= ental f scilities as defined in Section 1.12.

Irradiations are a subclass of experiments that fall within the specifications hereinafter stated in chis section. The surveillance requirements for irradiatiens are given in Section 4.2.5.

c Chiective The objective is := preve t damage to the reacter, excessive release cf radicactive materials, er excessive persennel radiatien exposure during the perfermance of an irradiatien.

P00RORGNa 1209 179

specifications A device or material shall not be irradiated in an irradiatien facility under the classificatica of an irradiatica t.21ess the following cer.ditions exist:

a.

The irradiatien =eets all the specif t:ations cf Section 4.2.5 for an experiment, b.

The expected radiaticn field produced by the device or sample upcn remeval frem the reactor is not more than 10 rem /hr at one fcot, otherwise it shall be. classed as an experiment.

c.

The device er material is encapsulated in a suitable centainer, d.

The reactivity worth of the device or material is 50.25 or less, otherwise it shall be classed as an experiment, and e.

The devica er material does not remain in the reactor for a period of over 15 days, otherwise it shall be classed as an experiment.

Bases This specification is intended to provide assurance that the special class of experiments called irradiatiens will be per-formed in a manner that will not permit any safety limit to be exceeded.

P00R ORBlW1 1209 180

4.0 SURVEILI.ANCE REQUIRE.V.E K S P00R BRGML=

Applicability This specificatien applies to the surveillance requirement of any system related to reacter safety.

Chiective The objective is to verify the proper operatica of any system related to reacter safety.

Specificatiens Any additions, mcdifications, er meinte.:mee to the ventilatica system, the cere and its asscciated sup, crt structure, the pool or its penetrations, the pocl ecc: int system, the red drive mechanism, or the reacter safetj system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated er to specifi-cations approvec by the Reacter Hazards Ccmmittee.

A system shall net be censidered cperable until after it is successfully tested.

Bases Th'is specification related to changes in reactor systems which could directly affect tha safety of the reactor.

As 1cng as changes or replacements to these systems continue to meet the criginal design specifications, then it can be assumed that they meet the presently accepted cperating criteria.

4.2 LI.MITING CCNDITICNS FOR OPERATICN 4.2.1 Reactivity Recuirements Acplicability These specificatiens apply to the surveillance requirements for reactivity centrol of experiments and systems.

Objective The cbjective is to measure and verify the worth, performance, and operability of these systems affecting the reactivity of the reactor.

Specificatiens a.

The reactivity werth of each control red and the shutdown margin shall be determined bi-annually but at intervals not to exceed 28 months.

b.

The reactivity worth cf an experiment shall be estimated er measured, as apprcpriate, befcre reacter cperation with said experiment.

c.

The centrol rods shall be visually inspected for deteriera-tien bi-annually but at intervals net to exceed 28 months. 1209 181

s d.

The tr=.nsient rod drive cylinder and associated air supply system shall be inspected, cleaned, and Itbricated as necessary semi-annually at intervals net to exceed 8 months The reactor shall be pulsed semi-annually to compare fuel e.

temperature measurements and peak power levels with those of previous pulses of the same reactivity value.

f.

Control rod drop times shall be verified to be less than one second.

If the rod drop time is found to be greater than one second, the rod shall not be considered operable.

Bases The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worths of experiments inserted in the core. Past experience with the BRR TRIGA Mark III reactors gives assurance that measurement of the reactivity worth on a bi-annual basis is adequate to insure no significant changes in the shutdown margin.

The visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation in the reactor. The reactor is pulsed at suitchle intervals and a eccparison =ade with previous similar pulses to determine if changes in fuel er core characteris-tics are taking p3 ace.

4.2.2 Centrol and Safety System Applicability Thesa specificatiens apply to the surveillanc a requirements for measurements, tests, and calibratiens of the etnt:01 and safety systems.

Objective The cbjective is to verify the performance and operability of those systems and components which are directly related to reacter safety.

Specifications a.

The scram time shall be measured annually but at intervals not to exceed 14 months.

b.

A Channel Test of each of the reacter safety system channels with scram capability for the intended nede cf cperation shall be performed prier to each day's cperation er prior to each operatica extending more than ene day.

c.

A Channel Calibration shall be made of the pcwer level moni-tering channels by the calerimetric method annually but at intervals not to exceed,14 months.

d.

A functional check shall be made of all reactor control and reac:or position interlocks described in Table 1..

1209 182

Sases Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the centrol reds to perform properly.

The channel tests will assure that the safety system channels are operable en a daily basis or pric: to an extended run.

The power level channel calibraticn will assure that the reacter will be operated at the preper pcwer levels.

Transient control red checks and semiannual maintenance insure proper operatica of this centrol red.

4.2.3 Radiatien Monitorine System Apelicability This specificatien applies to the surveillance requirements for the area radiatica menitoring equipment and the continuous air menitoring system.

Cb4ective The cbjective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.

Scecification The area radiation monitoring system and the centinuous air monitoring system shall be calibrated annually but at intervals to exceed 14 months and shall be verified to be operable at not weekly -intervals er after an extended period of reactor shutdown.

Basis Experience has shown that weekly verification of area radiation and air monitoring system set points in cenjunction with annual eslibration is adequate to correct for any variatien in the system due to a change of operating characteristics over a long time span.

4.2.4 Ventilatien System Acplicability This specificatien applies to the building confinement ventila-tien system.

Cbjective The cbjective is to assure the proper operation of the ventila-tien system in centrelling releases of radioactive saterial to the uncentrolled envircnment.

300RORGNAL 1209 183

Specification It shall be verified weekly er after an extended period of reactor shutdcwn that the ventilation system is operable.

3ases Experience accumulated over several years of cperation has demenstrated that the tests of the ventilatien system en a weekly basis are sufficient te assure the preper cperation of the system and centrol of the release cf radicactive material.

4.2.5 Experiment and Irradiation Limits Applicability

~

This specification applies to the surveillance rec;;irements for experiments installed in the reactor a..d its experimental facilites and for irradiatiens perfor:ed in the rradiation facilities.

Chiective The objective is to prevent the conduct of experiments or irradiatiens which may damage the reacter er release excessive amounts of'radicactive materials as a result f failure.

Specificatiens A new experiment shall not be installed in t ie reactor or a.

its experimental facilities until a ha:ards analysis has been performed and evaluated fer cc=pliance with the Limita-tiens en Experiaents, Section 3.8, by the Reactor Supervisor.

The new experident shall be submitted for co==ents to the Reactor Health Physicist prior to appreval by the Reactor Superviser. Certain experiments as defined by the Experiment Reviewal Precedure shall be referred to the Reactor Hazards Ccmmittee. The Experiment Reviewal Procedure and any changes to that procedure shall be approved by the Reactor Hazards Committee. Minor modifications te a reviewed and approved experiment may be made at the dis :retien of the senic: reactor cperator responsible for the cperaticn, provided that the ha:ards associated with the modificatiens have been reviewed and a determination made and documented that the modifications do not create a new, er a greater hazard than that in the original appreved experiment.

b.

An irradiation of a new type cf device or material shall not be perferred until an analysis of the irradiatien has been performed and reviewed for ce=pliance with the Limitatiens on Irradiations, Secticn 3.9, by the Reacte.. Supervisi.

1209 184 P00RORGNAL Bases It has been demcnstrated ever a number cf years of experience that experiments and irradiations reviewed by the Rea ter Staff er the Reactor Hazards Cc=mittee as appr:priate can be conducted without endangering the safety of the reacter or exceeding the limits in the Technical Specifications.

4.3 EE.AC CR FUEL ELEMIN"S Acclicability This specifica icn applies to the surveillance requirements fcr the fuel elements.

Objective The cbjective is to verify the continuing integrit; of the fuel elen.ent cladding.

s ecifications Each fuel element and fuel follower shall be checked for transverse bend and longitudinal elongation after the first 100 pulses of any magnitude and then af ter the first 500 pulses above 1.5% ak/k.

If any element distorts to or beyond the maximum limits during the first 500 pulses above the 1.5% ak/k, inspection of each element shall be made af ter the next series of 500 pulses above 1.5% ak/k. If no element distorts beyond maximum limits after a series of 500 pulses above 1.5% ak/k, the rember of pulses above 1.5% a k/k between inspections may be increased to 1000.

If an element is found to distort beyond maximum limits after a series of 1000 pulses above 1.5% ak/k, the next inspection interval shall be reduced to 500 pulses above 1.5% ak/k.

The limit of transverse bend shall be 0.125 inch over the total length of the element. The limit on longitudinal elongation shall be 0.125 inch. The reactor shall not be operated with elements which have been found to exceed these limits. Any element which is exhibiting a clad break as indicated by a measurable release of fission products shall be located and removed from service before continuation of routine operation.

Bases The frequency of inspection and measurement schedule is based on the parameters =cs: likely to effect the fuel cladding of a pulsing reactor cperated au moderate pulsing levels and utilizing fuel elements whose characteristics are well known.

The limit of transverse bend has been shown to result in no difficulty in disassembling the cere. Analysis of the removal of heat frem touching fuel elements shows that there will be no het spe ; resulting in damage to the fuel caused by this teuching.

Experience with TRIGA reactors has shown that fuel element beving that eculd result in touching has occurred without deletericus effects.

The elengati:n limit has been specified

= assure that the cladding material will not be subjected to stresses that could cause a less of integrity i.- the fuel cen-t ai..m e..: and t: assure adequate ::clant ficv.

?0DRORE1 m 185

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l This specificatien applies := the fuel elements used in the react:r cere.

Cbiective The cbjective is to assure that the fuel elements are of such a design and f abricated in such a manner as to permit their use with a high degree of reliability.with respect to their physical and nuclear characteristics.

Scecificatiens The core fuel elements may censist of an arbitrary mixture of the follcwing kinds of fuel:

1.

Fuel, when unirradiated, consisting of cir:enium-uranium hy-dride, in which urani m is centained to S.5 wts and is en-riched to 20s in uranium-235, and in which the hydrogen-to-metal ratio is 1.7.

2.

Fuel, when unirradiated, censisting cf =ircenium-uranium hydride, in which uranium is centained to 12 wtt and is enriched to 20% in uranium-235, and in which the hydregen-to-metal ratio is 1.7.

3.

Cladding:

3C4 stainless steel, ncminallY 0.02 inch thick Basis s

\\

The Safety Analysis Report shows that with a cere ce= posed exclusively of Type (1) fuel, peak (3-ring) fuel temperature was predicted to be 350cc r.reve ambient at 1.0 MW, r.nd less than 5000C during production of a 23 MW-s pulse. Both are well helcw the value of 1000CC which is known to i' safe fer such fuel.

The maximum power gradient in a cere ecmpesed of a mixture of Type (1) and Type (2) fuels, arises when Type (2) fuel fills the ll) 3 ring and Type (1) fuel fills all other cere pcsitions.

Analysis of this cenfiguration shews that the 3-ring peak fuel tempera-ture will be 4080C as compared with a prediction of 3200C for an all Type (1) cere, at a power of 1.0 MW.

Analysis of a 53.00 pulse prediers a peak 3-ring fuel temperature of 650cc as ec= pared with a value of 4900C indicated for the all Oype (1)

Purther analysis shews that the Type (2) fueled 3-ring cere.

cperates at a heat flux well helcw that necessary to produce film heiling. Thus, f:r this type cf mixed cere, all fuel temperatures are still well belcw ICCCCC, and in cther types cf mixed ::res, peak temperatures will be 10wer than in che case analysed.

" Anal s:s :f Partial Eefueling with '. w5 ::ani m Fuel for tne

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3erka'.e; 7esear.- Feaeter", MS Thesis, Cepartren: Of ?.; lear Ingineerin7, 2n.ers;t; Of Ca if:rnia, Serkeley, 197~

1209 186 23-

5.2 REACTOR CCRE Applicability This specification applies to the configuration of fuel and in-core experimants.

Cbjective

~

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

Specifications

e.. The cere shall be an arrangement of.'RIGA uranium-circenium hydride fuel-mederator elements pos :icned in the reacter grid plate.

b.

The TRIGA core assembly may be standard, 8.5 vtt or 12 wtt, or a cenbination thereof (mixed core).

c.

Single positions may be occupied by centrol reds, neutren startup source, ionization chamber, or by in-core experimental facilities. A maximum cf three separated experiment positions in the D threugh G rings, each occupying a maximum of three fuel element positions, may b'e arranged in the core.

Bases Standard TRIGA cores have been in use for years and their characteristics are well documented.

5.3 CCNTRCL RCDS Acplicability This specification applies to the centrol rods used in the reactor core.

Objective The cbjective is to assure that the centrol rods are of such a design as t0 permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specificaticn The shir er safety centrol reds shall have scram capability a.

and centain berated graphite, 3 C pcwder er beren and its 4

ccmpcunds in sclid ferm as a poisen in al ninum er stain' ess steel cladding.

These rods may inccrperate fueled follcwers.

P00RORGNLL a

387

b.

The regulating control rod has scram capability and is a stainless rod that contains the materials as specified for shim or safety control rods. This rod may incorporate a fueled follower, c.

The transient centrol rod shall have scram capability and contain berated graphite or beren and its ccepounds in a solid form as a poisen in an al==inum er stainless steel clad.

The transient red shall have an adjustable upper li=it to allcw a variatien of reactivity insertions. This rod may incorperate an aluminum er air follower.

Bases The pcisen requirements for the centrol rods are satisfied by using neutron absorbing berated graphite, 3 C powder er bcron 4

and its ecmpounds.

These materials must be centained in a suitabla clad =cterial, such as aluminum er stainless steel, to insure mechanical stability during movement and to isolate the poison frem the pool water environment. Control rods that are fuel follcwed provide additional reactivity to the core and increase the werth of the centrol red.

Scram capabilities are provided for rapid insertien of the centrol rods whien is the primary safety feature of the reactor. The transient control rod is designed for a re*-ter pulse.

The nuclear behavier of the air or aluminum follower which may be incorporated into the transient rod is similar to a void. A voided follower may be required in certain core leadings to reduce flux peaking values.

5.4 RADIATIC" MONITCRING SYS~EM Applicability This specification describes the functions and essential ecm-

'ponents of the area radiation monitoring equipment and the system for continuously monitoring airborne radioactivity.

Objective The cbjectivo is to describe the radiation monitoring equipment that is ava'.lable to the operator to assure safe cperation of the reacter.

Specificatien The radiatien monitoring equipment listed in the follcwing table will be available for reacter cperation.

Kadiacien Menitcrine Channel and Functice Area Radiaticn Mcnitor (gamma sensitive instruments)

Function - Monitor radiation fields in key iccations, alarm and readeut at centrol censole and readcut in recepti:n room.

[L 1209 18.8

s Centinuous Air Radiation Menitor (beta, gen =a sensitive detector with air particulate cellectica capability)

Punction - Monitor concentratien of radicactive

articulate activity in reacter roc =, alarm and readcut ac menitor and alarm at centrol censole and in reception rocm.

Gas Meniter (ga=ma sensitive detecter)

Function - Moniter concentratien of radicactive gases in building exhaust, alar = and readout at control censole and alarm in receptien reem.

Basis The radiation monitoring system is intended to previde infer-mation to operating personnel of any impending or existing danger from radiat.mn so that there will be sufficient time to evacuate tha facility and take the necessa.f steps to prevent the spread of radioactivity to the surroundings-.

5.5.

FtJIL STORAGZ A=plicability This specification applies to the stcrage of reacter fuel at times when it is not in the reacter cere.

Cbjective The, objective is to assure that fuel which is being stcred will net beceme critical and will not reach an unsafe te=perature.

Specificatiens All fuel elements shall be stored in a gec=etrical array a.

where the k-effective is less than O.8 for all conditiens of moderation.

b.

Irradiated fuel elements and fueled devices shall be stered in an array which will permit sufficient natural convec-tien ecoling by water er air such that the fuel element or fueled device temperature will not exceed design values.

Basis The limits impcsed by Specificatiens 5.5.a and 5.5.b are censervative and assure safe stcrage.

P00R ORIGINa

-v-1209 189

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.v....w Applicability This specification applies to the building which hcuses the recctor.

Cbjective The objective is to assure that previsiens are made to restrict the a cunt of release of radicactivity into the envir:r. ment.

Specificatiens a.

The reactor shall be housed in a facility design.

o le akage.

The minimum free volume in th,e facility shall be 300,000 cubic feet.

b.

The reactor building shall be equipped with a ventilation system designed to filter and exhaust air er other gases frem the reacter reem and release the: fr = a stack at a minimum of 40 feet from the highest ground level adjacent to the buildine c.

Emergency shutdown controls for the ventilation system shall be located in the reception rec: and the system shall be designed to shut down in the event of a sub-stantial release of fission products.

Bases The facility is designed such that the ventilation system will normally maintain a negative pressure with respect to the at=csphere so that there will be no uncontrolled leakage to the environment.

The free air volume within the reacter building is ccnfined when there is an emergency shutdown of the ventila-tien syste=.

Controls for emergency filtering and normal operation of the ventilation system are located in the reception rocm.

Proper handling cf airbor.e radioactive materials (in emergency situations) can be conducted frem the reception rocm with a minimum.of exposure to operating personnel.

5.7

.IAC-CR PCCL WATER SYSTEv.S Applicability This specification applies to the pcci centaining the reacter and to the c=eling Of the cere by the pcci water.

Objective Tne Ob ectite is to assure that ecolant water shall be available provide adequate ceci ng f the react:r cere and adequate radia-i:n shielding.

P00R 01El 1209 190

r specifications a.

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

b.

During reactor operation the pool water level shall be monitored and shall be at least 14 feet above the top grid plate.

if the A pool level alarm shall indicate less of coolant c.

pool level d: cps to approx'mately 2 feet above the tcp of.the i

cere.

d.

A pool temperature alarm shall indicate when the maximum bulk temperature exceeds 120*F.

e.

Loss of coolant alarm requires corrective action. This alarm is observed in the reactor control area and outside the reactor room in the reception room.

5.8 FIACM R SUFFCRT 3?lDGE Applicability This specification applies to the reactor support bridge.

'Cbjective The objective is to assure that the reacter support bridge be previded with electrical as well as rechanical steps to prevent bridge ever-travel and to assure that the bridge-travel speed is within a safe limit.

Specificatiens a.

Limit switches shall >* p cvided to step the drive meter of the suppert cridge when the bridge has reached either cf its two extreme limits of travel.

b.

Mechanical stops shall be mounted at bcth ends of the track to prevent bridge over-travel, The maximum speed of the bridge shall be 6.5 feet per.inute.

c.

Easis our past experience proves that the limit switches and the mechanical steps provided as well as the speed limit specified in 5.S.c are adequate.

?0DR ORGE1 1209 19l

s 5.9 REACTOR AUXILIARY ELECTRICAL SYSTEM Applicability This specification applies to the emergency generator system.

Objective The objective is to assure there is a source of electrical power in the event of loss of normal electrical power to the reactor laboratory and associated building.

Scecification The emergency generator shall be tested quar,terly (not to exceed four months) and verified to assuee and carry anticipated emergency reactor laboratory electrical loads and an equivalent maximum anticipated building load.

Basis The auxiliary electrical system is designed such that there will be sufficient electrical power available in emergency situations where loss of normal electrical power may occur.

1209 192.

t 6.0 AOMINISTFATIVE CO!CROI.S 6.1 ORGANI"ATICN The facility shall be under the direct control of the a.

Reacter Superviser er a licensed senior cperator designated by him to be in direct control. The Reacter Supervisor shall be ressensible to the Reacter Administrator for safe 6

operatien and maintenance of the reacter and its associated equip =ent.

The Reacter Superviser er his appointee shall review and approve all experiments and experimental pre-cedures prior to their use in the reacter. He shall enforce rules for the protection of personnel against radiation.

b.

The safety of cperation of the Berkeley Research Reacter shall be related to the 'Miversity Administration as shown in Figure 1.

6.2 REVIEW AND AUDIT a.

A Reacter Hazards Cc=mittee (REC) of at least three (3) mem-bers knowledgeable in fields which relate to Nuclear Safety shall review, evaluate, and approve safety standards asscei-ated with the operation and use of the f acility. The Reactor Health Physicist shall be an ex-officio member of the Reacter Hazards C:mmittee. The jurisdiction of the REC shall include all nuclear operatiens in the facility and gen-eral safety standards.

b.

The cperations of the Reacter Hazards Ccemittee shall be in ac:erdance with a written charter, including previsions for:

(1) Meeting frequency, (2) Voting rules, (2) Quorums, (4) Methed cf submissien and :ntent cf presentatier t: the C:=mittee, (5)

Use of subccamittees, and P00R ORIGIRL 1?_09 193

s s

Univ. cf Calif.

Berkeley Chanceller Reacter Associate Vice-Chair an Hazards Chanceller Cept. 2.'ucl.

ng.

--i Ocemittee Business Affairs 1

I u___-,

I i

i Office of Reacter

.nvironmental Administrater n e a.,,J1

& Safety Reactor Health Reacter Physicist Superviser (Member ?d:C) l Reacter Chief Reacter Cffice Cperater Secretary Reacter operators Line of Responsibility Li.e of C::cunicatien

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1209 194 l'

g (6)

Review, approval, and dissemination of minutes.

The RF.C or a Subec:.mittee thereof shall audit reacter c.

cperations at least quarterly but at intervals not to exceed four months.

d.

The respensibilities of the Cc=mittee er designated Subce=nittee thereof include, but are not limited to, the following:

(1)

Review and approval of experiments utilicing the reactor facilities, as described in the Exceriment seviewal Precedure, (2)

Review and approval of all prepcsed changes to the facility, precedures, and Techn cal Specificatic'c, as described in the charter, (3)

Review of the operation and operatienal records of the facility, J

(4)

Review of unusual or abnes. mal occurrences and incidents /

which are reportable under 10 CFR part 20 and 10 CFR Part 50, (5)

Determination of whether a proposed change, test, er experiment wculd cent tute an unreviewed safety questien er a change in the Technical Specificatiens, (6)

Review of abnormal performance of facility equipment and operating ancmalies, and (7) Review and approval of the implementation of the physical security plan and the emergency plan.

6.3 AC :cN a'o ssE TAxzN IN HI EVEN"F A SArr-'Y I.IMIT Is ExCri;rD In the event a safety limit is exceeded:

a.

The.:sm.tcr shall be shut dcwn and reacter operation shall not be resumed until authcriced by the NRC, b.

An i=ediate report of the occurrence shall be made to the Chairman, Reactor Hacards Cc=ittee, and repc s shall be made to the NRC in accordance with Section 6.7 cf these specificatiens, and c.

A repcrt shall be prepared which shal.' include an analysis cf the causes and extent cf pessible res _ltant damage, efficacy cf ccrrective action, and recc=e..dations fcr

.easures te prevent er reduce the prehability Of recur-rence.

This repert shall be submitted te the Reacter g

}h)r Macards Oc:=ittee fer revcew and then submitted to the NRC when autherication is scught to resu=e cperatien of ne reae:cr.

s AC :CN *O BE TAYIN IN THE EVENT CF A PIPCRTABLE CCOURRENCE 6.4 In the event of a reportable cecurrence, the folicwing action shall be taken:

The Reactor Supervisor or his designated alternate shall a.

be notified and corrective action taken with respect to the cperations involved, The Reactor Superviser or his designated alternate shall b

notify the Chairman of the Reacter Ha:ards Cc=mittee, A report shall be made to the Reactor Hazards Cc=mittee which shall include an analysis of the cause cf the efficacy of corrective ac:icn, and recommen-cccurrence, datiens for measures to prevent er reduce the probability of recurrence, and A report shall be made to the NRC in accordance with d.

Section 6.7 of these specifications.

6.5 CPERATING PROCEDURES for the Written operating procedures shall be in effect following items:

Testing and calibration of reacter cperating instrumenta-a.

tion and centrols,.trea radiation monitors, and air particulate monitors; b.

Reactor startup, cperaticn, and shutdown; Emergency and abnormal conditions, including provisions for c.

evacuation, reentry, recovery, and medical support; d.

Fuel element leading or unloading; Centrol red removal or replacement; e.

f.

Routine maintenance of the control rod drives and reacter safety and interlock systems er other routine maintenance that eculd have an effect en reactor safety; Actions to be taken to ccrrect specific malfunctions of g.

systems er ccmpenents, including respenses to alarms, and h.

Civil disturbances on or near the facility site.

Substantive changes to the abeve precedures will require the approval cf the Rea::cr Superviser and Reacter Administrater er the Reactcr Hazards C:=mittee as describad in 6.2.d(2).

Temporary changes te the precedures that do net change their

~"

?00R D a m 1209 196

g criginal intent may be made by the Reacter superviser er his designated alternate. All such temperary changes shall be documented and subsequently reviewed by the Reacter Administrater.

6.6 FACILI'"Y CPEFATINO RECOFOS In additien to the requirements of applicable regulations, and in ne way substituting therefer, records and 1 cgs shall be prepared of at least the fellcwing items and retained for a peried of at least five years fer items a through f and indefinitely fer items g threugh k.

4 Normal reacter eperation, t.

Principal maintenance activities, c.

Repertable occurrences, d.

Equipment and compenent surveillance activities required by the Technical Sp ecifications, e.

Experiments perfer=ed with the reacter, f.

Gasecus and Liqui? radicactive effluents released to the envirens, g.

Envirennental =enitoring surveys, h.

ruel inventories and transfers,

i. Facility radiation ind c5ntiminatico surveys, j.

Radiatica exposures fcr hil persennel, and k.

Updated, corrected, and as-built drawings of the facility.

6.7 REPORTING F20UI?2.v.E WS In additien to the requirements of applicable regulations, and in no way substituting therefer, reports shall be made to the NEC Region V, Office of Inspection and Enforcement as fellcws:

a.

A report within 24 hcurs by telephene and telegraph.

(1)

Any accidental release of radicactivity above per-missible limits in unrestricted areas whether er not the release resulted in prcperty damage, personal injury, er exposure; (2)

Any eiciatien of the safety limit; and (3)

Any reportable cerurrences as defined in Sectica 1.1C cf these specificatiens

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s b.

A report within 10 days in writing of:

(1)

Any accidental release of radicactivity above per-missible limits in unrestricted areas whether or not the release resulted in property damage, perscnal injury, er exposure.

'"he written repcrt (and, to the extent possible, the preliminary telephone or tele-graph report) shall describe, analyre, and evaluate safety implications, and cutline the ccrrective measures taken or planned to prevent recurrence of the event; (2)

Any violatien of a safety limit; and (3)

Any reportable occurrence as defined in Section 1.10 of these specifications.

A report within 30 days in writing of:

c.

(1)

Any significant variatien of reasured values frem a corresponding predicted or previcusly measured value of safety-cennected operating characteristics occurring during cperation cf the reacter; (2) Any significant change in the transient or accident analysis as described in the Safety Analysis Report; (3) Any changes in facility crganication; and (4) Any cbserved inadecuacies in the i plementation of administrative er procedural centrols.

6.7.1 A report within 90 days after ecmpletion, of startup testing of the reactor upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor pcwer level describing the measured values of the cperating conditiens er characteristics of the reactor under the new conditions including:

An evaluation of facility performance to date in ecm-a.

parisen with design predictions and specifications, and b.

A reassessment of the safety analysis s"'-

4 ~~ed with the license applicatien in light of measured cperating charac-teristics when neasurements indicate that there may be substantial variance frem prier analysis.

6.7.2 An annual repert cevering the cperatien of the u.it during the previcus calendar year subnitted pric: to March 31 of each year previding the fc11cwi..g infer a:icn 1209 198

a.

A brief narrative summary of (1) cperating experience (including experiments performed), (2) changes in facility design, performance characteristics, and cperating pre-cedures related to reacter safety and cccurring during the reperting peried, and (3) results of surveillance tests and inspecticns; b.

Tabulatien of the energy cutput (in megawatt days) of the reacter, hcurs reacter was critical, and the cumulative total energy cutput since initial criticality; c.

The number of emergency shutdowns and inadvertent scrams, including rehsens therefer; d.

Discussica of the major maintenance cperations performed during the period, including the ef fect, if any, en the safety of the cperatien of the reae c ar.d the reasons for any ccrrective maintenance required; e.

A brief description, including a summary of the safety evaluations,cf changes in the facility or in procedures and of tests and experiments carried out pursuant to Secticn 50.59 of 10 CFR Part 50; f.

A s --' y of the nature and amcunt of radicactive effluents released er discharged to the envirens beycnd the effective cen.rol of the licensee as measured at er prior to the point of cuch release er discharge.

Licuid Waste (1)

Radicactive waste packaged and shipped (s"--= riced on an annual basis)

(a)

Total volume shipped (in gallons)

(b)

Total activity shipped (in curies)

(c)

Cates of shipment and dispositica

(:)

Radicactive waste discharged direct to sewer during the reperting period (a) Tetal vclume discharged (in gallens)

(b)

Total radicactivity discharged (in curies)

(c)

Cate of discharge P00RBRGINA 1209 199

(d) Average cencentration at point of ralease (in mierecuries/cc) during the repcrting peried.

Casecus Waste (summariced en a monthly basis)

(1)

Radioactivity discharged during the reporting period (in curies)

(a) Total estimated quantity of radicactivity release (in curies) deter =ined by an appropriate sampling and counting method.

(b) Total estimated quantity of Argen-41 released (in curies) during the reporting period based en data frem an apprcpriate enitering system.

(c) Estimated average atmosphe:.c diluted concentratien of Argen-41 released during the reporting period in terms of micrecuries/cc and fracticn of the appli-cable MPC value.

(d) Total estimated quantity of radioactivity in par-ticulate form (in curies) released during the reperting period as determined by an apprcpriate particulate monitoring system.

(e) Average concentration of radicactive particulates released in microcuries/cc during the reperting period.

(f) An estimate of th average concentration of other significant radicuuclides present in the gaseous waste discharge in terms of microcuries/cc and fraction of the applicable MFC value for the re-perting period if the estimated release is greater than 20% of the applicable MPC.

Solid Waste (summari:ed on an annual basis)

(1)

Total amount of solid waste packaged (in cubic feet)

~

(2)

Total activity in solid waste (in curies)

(3)

The dates cf shipment and disposition (if shipped eff site).

g.

An annual summary of the radiatien exposure received by facility persennel and visiters in terms of the average radiation expc-sure per individual and greatest exposure per individual in the two groups. Zact. significant expcsure in excess of the limits cf 10 CFR 20 should be reported including the time and date of the expcsure as well as the name cf the individual and the circumstances leading cp to the expcsure.

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---

en

h.

An annual sunmary of the radiation levels and levels of centamination cbserved during reutine surveys perfor: ed at the f acility in te :ns of the average and highest levels.

1.

An annual su= nary of any environmental surveys performed outside the facility.

I jl 1209 201

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