ML20085D193

Univ SAR Amend Pages for Rev 1.01

ML20085D193
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Site:	University of Texas at Austin
Issue date:	05/31/1991
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TRICA Reactor Facility Nuclear Engineering Teaching Iaboratory The University of Texas at Austin SAFETY ANALYSIS REPORT Submitted May 1991 hevision-1.01 Revision P tes Chapter 1 $/91 Chapter / S/91 Chapter 2 5/91 Chapter 8 S/91 Chapter 3 S/91 Chapter 9 S/91 Chanter 4 5/91 Chapter 10 5/91 Chapter 5 5/91 Chapter 11 5/91 Cnapter 6 5/91 Pat Replacements - October 1991 2-25, 73, ?? 23, 27, 28, 4-69, 8-1, 7, 10, 11, 16, 18, 22, 5-12, 14, 9-11, 9 15 6-13, 15, 10-2, 6, 9, 10, 31 1

9110150331 911G04 POR ADOCK 09000602 A PDR

UT TRIGA1 Safety Analysis Report Revision 5/91 The following list of pages replace pages of Revision 5/91 of the Safety Analysis Report.

he Comments 2-25 Revise figure caption date to 1940 4-69 Adi 2 sentences- to end of paragraph ' 2 5-12,14 Add paragraph 3, move last paragraph to page 5-14 6 13- Ravise wording of last sentence paragraph 2 6-15 Revise figure 7- 5 Correctspellinggfpartstoports,-line46 7-22 Correct 103 to-19 in line 8 7 23 Correct subscript in equation 13 7-27 Cortect value in line 8 to 2.12 x IQ'O recalculate results and revise result in line 15.

7-28 Revise results to correspond to values from page 7 27 8-1 - Add reference 'l] to paragraph 1 87 Revise. references [2,364]; revise wording of first sentence of_last paragraph 8 10 Revise reference annotation to (3,5)

-8 11 Revise reference annotation to (5) 8 16 Revise reference annotation to [6]

8-18 Revise reference annotation to (7l 8 22 Revise reference list and add new itema 1 and 3' 9-11 Revise entire page for cl rity and to correct results and units. (no'effect on other pages) 9-15 Correct mci to C1, 2 places in last paragraph.

10 2 Replace page to correct reference to Univ. Safecy Ofitee-10-6 'Chunge_1 hour to 1/2 hour, last sentence of first paragraph 10-9 Add item (d) An unanticipated...

,- 10 Add item (c) Events that._,

Change training cycle-to license cycle 10-31 Revise list in section 10.7.5 i

1-l l

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SAR 5/91 i

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i . ~.- - s na RESEARCll CEliTER AREA 1940 Figure 2-15a l

2-25

SAR $/91 Stepping motors operate on phase-switched direct current power.

The motor shaft advances 200 steps per revolution (1.8 degrees per step). Since current is maintained on the motor windings when the motor is not bein6 stepped, a high h,lding torque is maintained.

The torque vs speed characteristic of a stepping motor is greatly dependent on the drive circuit used to step the motor. To optimize the torque characteristic vs motor frame size, a Translator Modult was selected to drive the stepping notor. Tt.i s combination of stepping motor and translator module produces the optimum torque at the operating speeds of_ the control rod drives. Characteristic data for t'Te drive indicate a possible travel rate of 33 1pm (1.40 cm/s). Meesurements of the actual rate provide a speed of 27 ipm (1.14 cm/s). )

4.4.8.3. Transient Rod Drive Assembiv.

The safety transient contcol rod on pulsing TRICA Mark 11 reactors is operated with a pneumatic rod drive (see Figutes 4 32 and 4-33).

Operation of the transient rod drive is controlled from the reactor console.

The transient rod is a scrammable rod operated in both puls9 and stead-state modes of reac tor operation. During non pulse operation, the transient rod will fuu tion as an alternate safety rod with air continuously supplied to the rod. Rod position is thus controlled by operation of an electric motor that posir%s the air - drive- cylinder.

The position of the transient control rod ,. Its associated reactivity worth will generally dictate removal of the to1 as the first step of a startup for steady state operation. Rod withdrawal speed is about 2Sipn (1.19cm/s).

Tlu transient rod drive is mounted on a steel frame that bolts to the bridge. Any value from zero to a maximum of 15 in. (38.1 cm.) of rod may be withdrawn from the core; administrative control is exercised to restrict its travel ro as not to execed the maximum licensed step insertion of reactivity ($3.14 or 2 2% Sk/k).

The transient rod drive is a single-e.cting pneumatic cylinder with its piston attached to the transient rod through a connecting rod assembly. The piston rod passes through an air seal at the lower end of the cylinder. Compressed air is supplied to the lower end of the cylinder from an accumulator tank when a three-way solenoid valve locatad in the_ piping between the accumulator and cylinder is energized.

The compressed air drives the piston upward in the cylinder and causes the rapid withdrawal of the transient rod from the core. As the piston rises, the air trapped above it is pushed out through vents at the upper end of the cylinder. At the end of its travel, the piston strikes the anvil of an oil filled hydraulic shock absorber, which has a spring return, and which decelerates the piston at a controlled rate over its last 2 in. (5 cm.) of travel. When the solenoid is de-energized, the valve cuts off the compressed air supply and exhausts the pressure in the cylinder, thus allowing the piston to drop by gravity to its original position and restore the transient rod to its fully inserted position in the reactor core.

4-69

SAR 5/91 l

Numerous water system parameters are measured by local pressure or temperature sensors in the system lines. Both temper ature and pror.sure probe are located on the inlet and outlet lines of the pool water sido and chilled water side of the heat exchanger. A local indication of flow in the coolant loop is provided by the pressure drop across a venturi in the flow path. Purification loop flow is measured by an in line flow meter. Water pressure beforo and after the filter in the purification loop is measured for indicatloa of filter condicion.

Water quality is measured by two conductivity cells in the purification loop. The cells are located on inlet and outlet lines of the demineralizer that readout locc.lly in the control room. Typical conductivity celle are composed of two parts, titanium electrodes shielded by ryton for conductivit-f measurement, aad a thermister for xmperature compensation. A Wheatstone bridge circuit on the purification skid is connected to the cells. A switch allows selection of either inlet or outlet conductivity.

A connection from the purification system to the domestic water system provides makeup water to replenish pool evaporation losses. The makeup water connection includes features that isolate the two water systems. Two of these features are valves for flow control and a cuick release connection. The valvas include a check valve for limiting flow direction and one or more block valves for stopping water flow, Use of the quick release connection allows physical separation of the two systems except during periods in which the makeup process is operating.

5. 3 WATER SYFTDi DESIGN EVALUATION The water system including the reactor pool and the external cooling and purification loops have similar design features as used in many other operating TRIGA facilities. The demonstrated capability and in.:egrity of this system provides assurance that the coolant system will perform its function properly and safely.

Availability of pool water for cooling and vertical shieldin6 is assurcu by designing the system with siphon breaks on suction lines and discharge lines within 2 meters of the normal pool level. Greater losses of pool water are extremely improbable, although they could conceivably be initiated by rupture of the reactor tank. As t.hown in the loss of pool water accident analysis, even with complete loss of pool water fuel clad integrity is not threatened.

Adequacy of reactor cooling is assured by the large amount of cooling c pacity inherent in the reactor pool volume as well as the capacity of the external cooling circuit which can dissipate heat at a rate equivalcut to 1000 kW steady-stace operation. If available heat exchanger capacity is diminished to 900 kW and initial pool temperature is 100*F, the reactor can be operated for more than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> be fore the bulk pool temperature reaches 120*F. The actual time would be considerably longer since as bulk pool temperature increases heat enchanger heat removal capacity incceases. Without external cooling or other heat loss the bulk pool temperature will rise about 20. 7'c af ter one hour of operation at a steady-state power level of 1000 kW(t).

5-12

1 SAR 5/91 Heat removal capacity and thus pool heat rate is specified by analysis of a tube and shell heat exchanger. At a flow rate of 400 gal / min (25.2 liters /sec) of chilled water at 48'F (8.89'C) a heat removal rate of 1140 kW is expected. The presence of fouling in the heat exchanger is considtred minimal based on the purity of the two heat exchanger fluids. Capacity is reduced to 1070 kW for a fouling factor of ,0004, The heat t.ransfer and hydraulic parameters are shown in Table 5 2.

Experience with this purification equipmert in 0 9er TRICA systems has shown that coolant conductivity can be easily maint ained at levels of less than five micrombos per centimeter using the mate rials contained in the coolant system design. Furthermore, this experJence has shown that no apparent corrosion of fuel clad or other components will occur if the coaductivity of the warer does not exceed fiva micrombos per centimeter when average,1 over a 30 day period.

Control of radioactivity in the coolant is provided by the purification system. Should radioactivity be released from a clad leak or rupture of an experiment, detection of the *elease would be signaled by the continuous air monitor or by the rea 'or room area monitors.

Based on coolant transport t1me calculations in the safety analysis section, these monitors should register an increase in coolant radicactivity within approximately 60 secon?- of the time of radioactivity release . The transport time is es imated from the time for the coolant exposed in the core to reach the surface of the water where the continuous air monitor will detect a release of radioactivity from the pool water. An alternate indication of radioactive release is prouided if a water activity monitor is installed or by a GM detector area monitor, 9

5-14

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SAR 5/91-i p.cactor ~ cont rol in che pulsing mode conalsts- of _ establishing

. critical' ty at ' a flux level below Ik in: the MANUAL. mode . This is accomplished by *he use of:the motor-driven control rods, leaving the transient rod either fully or partially inserted. The mode selector switch 's then depressed. The MODE selection switches automatically-connect the pulsing chamber to monitor and record peak flux (nv) and erergy release (nvtr. r ilsing esn be initiated from either the critical or suberitical reactor state.

In a square-wave operation, the reactor is first brought to

-criticality below 1 kW, leaving the transient rod partially in the core.

All of the manual instrumentation is in operation. The transient rod is ejected froin the core by means of the transient rod FIRE pushbutton.

When the power level reaches the deuand level, it is maintained much the same as in the automatic mode. Two rods are used, the transient rod to achieve power and the regulating rod to maintain power. ,

6.1.5. Reactor Scram and Sudgyn_liy.;Ltsm, A reactor protective action [4] interrupts the magnet current and results in the immediate in wrtion of all rods under any of - the

-following:

-n. liigh neutron fluxes from either NP-1000 or NPP1000,

b. liigh-voltage failure on the NM.1000, NP 1000,-or NPiOOO.

c. liigh fuel cemperature (one out af two).

d. Manual scram,

e. Peak neutron flux or energy (pulse mode),

f. Minimwn period (available for use as- des, Ired),

g. External safety switches (for experiments),

h. L.oss of electrical power to the control console

1. Watchdo6 circuits for each . computer to monitor computn status by updating timers.

All scram conditions are automatically indicated on the CRT monitor. A manual scram will als9 insert the control rodr. and may be used for a normal fast shutdown of the reactor. The scram circuit s,fety function is an inaependent system that dependu on wiring

. ladependent of the digital control system functionn Several conditions of the digital processing system will cause the scram mode condition. Among these are the loss of communication between the two computers, a database timeout condition or failure of a digital input scanner. By updating dual programmable timers, watchdog circuits at periodic intervals, determine the execution status of key elements of the computer digital program.

6-13

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! kI I a IJX;IC DI ACRAM FOR CONTROI, SYSTEM Figuro 6-7 6 l$

SAR 5/91 Ventilation of the reactor bay ir provided by two modes of system operation. One mode is for standard operation with recirculation of air. The other mode is an exhaust operation with high volume flow that has no air recirculation. Design during exhaust mode operation is a rate of air exchange in excess of two per hour. Total volume for the room exceens 4120 cubic meters. Norwil operation of the ventilation system uses a roof stack for the exhaust of air from the reactor bay.

Air filtration in the ventilation system is to be of typical design for normal HVAC operation with no special provisions. Schematics of the ventilation system for the reactor bay area and a logic diagram of the ventilation control system sensors and controls are provided in Figure 7-3.

Control of air confinement within the reactor bay is provided by differential pressure control betseen the reactor bay and a representative ambient external measurement pc,at. Additional measurement points in ventilation zones adj acen t to the reactor b.~

maintain the differential pressere between the reac tor bay and adj ace, v_

access areas. The differential pressure control is intended to function in both standard and exhaust operation modes of the ventilation system.

Isolation of the reactor bay is provided by ventilation dampers.

These dampers will shut in response to either manual or automatic signal actuation. An automatic signal will initiate shutdown of the ventilation system by closure of the dampers if a set point for nirborne particulate radioactivity exceeds a setpoint. Protective switches within the ventilation system will cause the air. fans to respond to the position change of the dampers. Damper design is for fail-safe operr. tion so that loss of control power will isolate the reactor bay.

Dampers locations are in the vicinity of the duct penetrations into the reactor bny. An isolation damper is in each of two supply air ducts.

Une return air duet with two sections contains two isolation dampers, one in each section. A pait of return air ducts also contain a damper in each duct.

The separate air purge system is designed to exh raus t air that may contain radionuclide products by a low volume system. The primary nuclide ot interest is argon-41. Figure 74 shows a. schematic of the argon purge system and its control logic. Air from potential sources of neutron activation such as beam tubes, sample transfer systems, and releases from exchanges at the pool water surface are subject to confinement ar9i isolation by the system. Filtration of air in the system will Irelude prefilter and h!ch ef ficiency particulate filter.

Design provisions allow for the addition of charcoal filters if experiment conditions should require the additional protection. Sample ports in the turbulent flow stream of the purge system exhaust provide for measurement of exhaust activities. Actuation of the isolation damper in the argon purge systen is by ma nur.1 operation ~ of the fan control switch.

A schematic of each ventilation system is showr. In Figure 7-5.

7-5

2' SAR 5/91-w - width of reactor pool (200 cm),

1. - len6th of reactor. pool (300 cm), l h - height of pool water (750 cm).

Exposure time, t, is'about 2.2 seconds and cycle time, T, is about 4.6 x-30 seconds. ,

Argon atoms exchanged at the water air _ interface depend on a water thickness. depth that_ is ~.4 mall relative' to the pool- dimensions and, '

the re fo re , a small-- fraction of the available. saturated . argon is exchanged with the air. During the time required-for the pool watar_to circulate once tt' rough the reactor . core , about one hour and twenty minutes, the argon equilibriurn concentration should deplete to the lowest solubility value for equilibriwn concentration. The argon release as a . function of temperature and sulubility - thus approaches zero. This depletion occurs as the activity of the argon radioisotope increases but is substantially complete as _ the argon 41 activity reaches hal f the equilibrium value at about -110 minutes.

• Evaluation of the water-air interface exchange rate for argon is rela' ed to an air and water thickness depth that depends on the argon atom diffusion coefficient. The total exchange rate then is a function of the pool surface area, A, . and an effective release vc,lume Vt '. The two-teras are related by-

'i l As fjV' t t , ( 13 ) .. i where

1. 1 is a surface exchange . coefficient (cm/secy ,) and ft ,j is the f raction of atoms exchanged from volume i to j (sec' Estimatus of the surface exchange coefficient (i.e., the 6as in a l' unit volume that is exchanged at the surface per unit time per unit l surface area) fer argon very considerably.

One method of arriving at a value for this parameter is through the diffusion coefficient of. the gas in water. The mean square distance

. traversed by a molecule is p

< AX >2 ,

2Dr (14) where D - dif fusion coefficient (cm 2 /sec),

j; t- time (sec). ,

i l_ The exchange coefficient is asswned to be evaluated for 1 sec as

! p - ( < AX >2 )1\2 je

( 2D/t )l\2 7 22

SAR 5/91 10'b cm2 /sec, The diffusion coefficient at 40*C is about 1.1  :

and, if one a s susne s that only one-hal f of the atgon atoms within one dif fusion length of the surface escape.

B -

l\2 (2 x 1.1 x 10'D)l\2 -

2.35 x 10-3 cm/sec Values for the surface exchange coefficient have been reported by Dorsey [11 for air, 02 , and N2 The values for these three gaseu are all about equal. Austuning arson behaves as do these gases, a value is obtained of p -

5.7 x 10'3 cm/sec.

Meaaurements have been made of the argon Al activity in a TRIGA -

Mark 111 r eac ar pool and from t he data acquired f rom these measurements it was posulble to construct a value for the xurface exchart,e coci!!clent. This value at 40*C is about 2.9 x 10'0 cm/ nee.

During equilibrium conditions and assumina, no ditference in the raten of escape fractions for Ar 40 and Ar 41, the number or argon atoms that escape from t.he water into the air equals the number of argan atow that enter the water f rom the air, i.e.,

fjV'Ni t t - f j -. i Vj' Nj , (15) where Nj - 2.1 x 10 17 argon atoms /cm of air -N 40 ,

I5 argon atoms /cm3 of water -N 40 N i - 7.1 x 10 Solving for fj_,g V' gives f j .,1 Vj' >

f t.,j Vj ' (Nt /Nj) (16)

The following calculations were performed to evaluate tm rate of Ar-41 escaping from the reactor pool water into the room enclosure. The calculations show that the Ar-41 decays while in the water, and most of

r. he rad!ation is safely absorbed in the wat e r. The changes in Ar-41 concentration in the reactor, in the pool water external to the reactor, and in the air of the room enclosure are given by 0I dNt Vt -V1 4N"o40 t , gl41 (y , yl j yM , gel y1) ,

g241y}

dt (17) dN,0I V7 -- - - # N2 41 1

V2 + V101 (Nt 0I N2 01) d* *

( f7. t'l2 41 V' 41 C18 ;

2 ' f3 >2N3 73 )

OI dN 3 41 '

41 3-+2N 3 41 V 3 ') ' N3 4l V3 - - (f243 N2 V2 f (\ V3 ' 4) dt (19)  :

7-23 l

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7,4.2. ~ Eyftlyglan of Arutt1.AL1tlEMS J

Tho . re1 case of , argon.41- f rom ' the_ facillty -is diluted by the ventilatica exhaust _ rate,.'asswaed to represent two air chargos per heur, and_ averaged for a 5 day, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> operation sendule at full power. .The release concentration from the pool averaged for one year-is,

,24 (2.12 x 10'8) - 5.1 x 10*9 pC1/cm3 ,

only -20% of 'the experin.cnt f acility argon 41 is assumed to exhaust since experiments will replace sorne or most_ of the exposed air, .'

6 (,20)-( 24)/2.29 x 10 6 - 1.3 x 10~7 pC1,'cm3 .

Total estimated release is 1.3 x 10*7 pC1/cm3 .

The whole body gamma ray dopo rate to al person ' immersed in a- acini'-

infinite cloud of radioactive gases can be approximated by D + 900 EAD- (29) where E - the photon energy, 1.3 Mov AD - effective exposure concentr eton Gj/m3 The concentration downwind from the point at wotch t' e activity - in discharged from the building is AD-Aqi(x), (30) wheroES -- the dilution factor at the distance x, (soc /m ),3 3

A activity concentration in the dischar6e (Ci/m );

q - the building exh4.ist rates'(m 3/nec).

4 If It is ' as.sumed that the discharge is at the ~ roof lino, thn

, - dilution factor in.the lee of the building (x - 0), is give.n [6] by: ,

((0) -

1/csu  :, (31) whero c - a constant (0,5),

s - building crogs sectional area normal -to the wind direction (m ),

u - wind velocity-(m/sec).

2 A' minimum cross sectional area in assumed of 234 m (60 x 42. f t) _ond, for a wind velocity of 4 m/sec, p(0) - 1/(0.5 x 4 x 234) - 2.1 x 10'3 sec/m 3 ,

(32) 7-27 5

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SAR 5/91 The averaged dose rate at the exhaust stack is D - 900 x 1.3 (1.3 x 10~7) - 1.5 x 10"' rads /hr, an average of 0.15 mrad /hr 1: 'ne stack or D - 1. 5 x 10 "' (2.3 x 2.1 x 10'3) - 7.2 x 10'Irads/hr, an average of 12 prad/hr at ground level.

At the limiting exhaust rate, 640 cm3/sec, to ignore the argon 41 decay. the source term is 27 pC1/sec for the beam ports which souid increase the dose rate 4.5 timca to 3 . ') prad/hr. If core experiment facilities, such as the conter tube and rota v specimen rack, are vented at the same exhaust rate the releases would increase by 102 pCi/sec.

Ilowe ve r , these exhaust rate conditions represent limiting conditions, not actual release rates. For the exhaust manifold rate of 3.15 cm3/sec ,

the release rate would be 50 pC1/sec for the two corn experiment facilitics. The rotary specimen rack is the primary source of the activity.

Venting of experimont facilities, especially the rotacy specimen rack, will req tire monitoring it the release rate or replacement of the att with gases such as nitrogen or carbon dioxide to control release concentrations of air activity. However, normal operating conditions do nt vent the rotary specimen rack, although this is an optional operating cond it lon. The pneumatic transfer system, by comparison, routinely contains nitrogen or carbon dioxide gas to limit the releases within the room that contains the access terminal.

Actual dose values for thw argon-41 release may vary. The ocam port release estimate is less than u,8 2 rad /hr whien is equivalent to 10

mrad /yr. Lower neutron fluxes, smaller air volumes, shorter operation i n' times and larger dilution factors will assure that releases do not

i. exc 'ed annual release constraints. Monitoring the exhaust will verify that nther release points such as the core axperiment facilities do not cause the total to exceed preset limits.

7-28

SAR 5/91 Chapter 8 EXPERIMENT AND 1RRADI AT10N FACILITIES the experimental and irradiation facilities [1} of the TRICA Mark Il reactor are extensive and versatile. Physical arcoss and observation of the core are passibic at all times through the vertical water shitid.

Experimental tubes can cantly be installed in the core region to provide ,

facilities for high-level irradiations or in-core experiments. Areas outside the core and reflector are available for larger e x;>e r ime n t equipment or facilitten.

8.1. STANDARD EXPERIMENT FACILITIES 8.1.1. Guta LIMtt&

The reactor is equipped with a central thimble for access to *ho e point of maximum f i t.x in the coro. The central thimble connists of an a l urn i nurn tube that fits through the center hole of the top and bot torn grid platon. D irnenn i nns o f t '.o tube are 1. ') in. o.d. (3.Al cro.) and

) 1.33 in. i.d. (3.38 cm.). lloles in the tube assure that it in normally filled with water. Water is expo'. led from the tube by comprensed air.

Ex pe r i n.c n t s with the central thimble include irradiations of small samples and the exposure of materials to a collimat.ed beam of neutrons or gamma rays.

8.1.2. Eu Laty_SrtchcLRach A rotary, enultiple position (40) specimen rack located in a well in the top of the graphite reflector provides for the large scale production of radioisotopen and for thu activation and irradiation of multiplo samplen. All positions in this rack are exposed to neutron fluxes of comparable intensity. Specimen positions are 1.23 in. (3.1B _

cm ) in diameter by 10.80 in. (27.4 cm.) in depth. Samples are loaded from the top of the r eactor through a water-tight tube into the rotary rack using a specimen lifting device or pneumatic prensure for incertion and removal o f: samples frma the sample rack positions. The rotary specimen rock can he turned frorn the top of the reactor by manual operation or by a m W r drive. Figure 8-1 shows the rack.

8.1 3. Pptumali(LSpttimdLIVht a pneumatic transfer system permits applications wi th short-lived radioisotopes. The in core terminus of thin system in nors 11y located in the outer r tng of fuel element positions, a region of hf gh neutron flux. The <. ample capsule (rabbit) la conveyed to a recolver nender station via 1,2$ in, o.d. (3.18 cm.) aluminun tubing. Effective space in the s pe e l te n transfer capsules la 0.68 in. (1.7 cm,) diameter by 6.5 in. (11.6 cm.) height. An optional t.rans f e r box may he employed to permit the sample to be sent and received from up to three d'fferent receiver-nender utations. A schematic of the pneumatic irrad!ation terminal and air flow control valves in nhown in Figure 8 7.

8-1

SAR 5/91 8.1.5. Jiva1ua.Lton of Malcrlfti s in Expe_Ilment Faclll1 Lea

't h e following information is a guide for evaluating experiment materials in order to prevent the introduction of materials that could damage the reactor or its components. A careful evaluation [2,3) of proposed experiment materials shall be performed to classify the experiment as an approved experiment. Guidelines for the following types of experiment materials are provided: materials which require doubla encapsulation, explosive materials and the ir confi neirent , fueled experiments, and materials which could be sources of airbocne radioactivity. The limits referenced in this section are technical specification requirements to prevent the occurrence of a serious safety hazard.

8.1.5.1. Qmible Encapsulation Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, and 1Iquid fissionable materials shall be doubly encapsulated. Chemical th zard Information in the "llandbook of Laboratory Safety," [4] shall be used as a guide for classifying the first three categories of materials.

A table of isotopes shall be used as a guide for the last category.

The Chemical llazard Information lists several categories of information about hazardous materials. One category of significant interest has the title " Relative llazard to Health from Concentrated

' ho r t -Te rm Exposure."

, This category identifies materials that are highly corrosive. From a conservative standpoint, ti.ose materials which are corrosive to human tissue shall be consi& ctd to be corrosive for the purposes of double e nc a p sM a t ion . References also exist to other t o .,. i c or hazardous material listings for details about a particular material.

'Ivo categories determine t.he explosive and flammability.. properties of mate rials . Cne titled "N.F.P. Hazerd Identification Signals" and another titled " tammable Lirits in f.ir," t'ecify materials that can represent safety hazards. The table identifies degree of explosive potential, chemical react ' . y and flammability limits.

Material ratings for "N.F.P.A. Mr_zard Identification" consist of a scale of 0 (low) to 4 (high), regarding their health, fire, and renetivity hazards while under fire conditions. In this table, fire ratings indic,te the degree of susceptibilit.y for a part.icular material to burn. Further, a reactivity rating indicates the degree of susceptibility for a particular material to release energy. The code rating scale for flammability specifies limits as a functian of the air volume percent.

Materials having reactivity ratings greater than two or fire ratings equal to four should be doubly encapsulated. The encapsulation ,

I is to protect against *he energy release and corrosion prsperties.

Materials having a reactivity rating of 1 or a fla.umability rating greater than 1 should be evaluated individually to determine if d"uble encapsulation is warranted.

8-7

LAR $/91 Annuming an internal pressure of 1057 atm (15,538 psi), maximum values of d/t are displayed in Table 8-2 for the encapsulation matorials o t' Table 8 .1 The figures indicate that a polyethylene vial is not a practical container since its wall thicknens must be 4.5 times the diameter. Both the alumi nu:n and the stainless steel made 3,a r i s f ac t ory containers.

8.1.5.3. EudsMspedatemi Each fueled experimert shall he controlled such that the total inventory of iodine isotr. pes 131 through 135 in the experiment is no greater than 750 millicuries and the maximusa stront iurn inventory is no greater than 2.5 millicuries. These restrictions are used to limit the fisulon prodect re lo. a.o in the event of an experiment container rupture.

In terms of harard to husnans , the iodine iaotopes represent the most harmful short lived isotopes. The values are justified by making the following assumptions:

a. Half of the total todino and st.rontium inventories are comprised of 1 131 and Sr 90, the isotopes of each element which ha.o the lowest permitted concentratica levels.

b. The lodine / strontium inventories are evenly distributed throughout the reactor room, which has a volume of 4.83 x 103m3

c. At two air changes per hour, the isotope inve r.t o r y is removed from the room in one half hour.

o. Average indoor concentration over an occupational year.

Cticulat.lons based on these assumptions r e .S u l t in indoor concentrations below the occupational concentration limits of each tsotope, as shown in Table 8 3. Assu.ning the building wake dilution factor to be 5.3 x 105 sec/it3 Calculations indicate that ground Icvel concentrations outside the building are below the reference leveln [3,5]

of each isotope Table 8-3 Calculated Isotope Release Values Units 1 131 Sr 90 indoor pCL/ml 1.9 x 10~0 fi Sx 10-11 occupattonal 11 nit pCi/ml 2.0 x 10'8 7.0 x 10 0 outdoot pC1/ml 2.2 x 10'II 1.$ x 10"I'*

reference level pC1/ml 2.0 x 10-10 5.0 x 10-12 8-10

SAR 5/91 i

8.1.5.4 M thmMxpniment Re kne_2 Experiment materials, except fuel materials, which could off. gas, sublime, volatize, or produce aerosols under one of the following conditions shall be limited to occupational levels for airborne radioactivity concentration, s specified in 10CFR20 (5), when averaging over a year;

a. Normal operating conditions of the experiment or reactor,

b. Credible accident conditions in the reactor,

c. Possibic accident conditions in the experiment.

The radioactivity release is based on the assumption that 100% of the gaseous activity or radioactive aerosols produced escape to the reactor room or the atmosphere.

When considering materials for experimentation, the type of radiation emitted by the altborne radioactive products should be noted.

For example, alpha-emitters, while not considerably hazardous outaide the body, may cause significant damage to human tissue if they are inhaled. Although conce.tration- limits in 10CFR20 reflect these considerations, the experimenter has a responsibility to know what hazards may exist during and after the experiment.

The following assumptions shall be used to calculating the airborne reactivity concentration -

a. If the effluent from an experimental facility exhausts ,

through a holdup tank vSich closes automatically on high radiation Icvel, at least 10% of the gaseous activity or aerorols produced will escape.

b. If the effluent from an experimental facility exhaust through a filter installation designed for greater that 991

(:fficiency for 0.25 micron particles, at leact 10% of thcne vapors will escape,

c. For materials whose boiling point is above 55*C and where vapors formed by bolling this raterial can escape only ,

through an undisturbed column of water above the core, at least 10% of these vapors will escape These three assumptions contain phrases stating that at 1 cast 10%

of the vapors escape from the specified, engineered safety systems that are part of the experiment design. For the purposes of calculation, the assumptions are intended to provido a c ont.e rva t ive estimate of the amount of radioactivity that will be released.

Calculations of the airborne radioactivity concentration depend on the reactor room effective volume, building wake dilution factor, and the vont11ation nystem ficw rates. The reactor room effectivo volume, 4.83 x 103 m3, the building wake dilution factor, 5.31 x 105 sec/ft3, and the ventilation system flow ratec are facility design features.

Calculations have been done to demountrate the fraction of airborne release that will occur interior and exterior to the building for d i f i'e re nt ventilation flow rates, i

8-11

. . . . - - -. .= .

SAR 5/91 Pulse release:

3 (40m )

• 10'5 Ci/m3 - 1 x 10'3 Ci (2) '

Equilibrium release (24 hr/dy) (30 dy/mo) 4.5 x 10-5 C1/hr x 10 2 C1 (3)

Projection of the dose from the cobalt +60 isotope in the deionir.er resin has been done by calculation with Microshield (6)(version 3.12).

Table 8-4 Microshield Data (6) ,

CASE: Dose at deionizer, 30 millicuries cobalt-60, one meter GEOMETRY: Cylindrical source from side - cylindrical shields Distance.to detector... . . . .. . . ... . . , .100. cm.

Source length... .... .. .... . ..... .. .. ..... 86. "

Source cylinder radius... ... . .., . . . . 23. "

Dose point height-from base. . . .. ... . . ..... , 42. "

Thickneus of shield., .... ... .... ... . .. . .. 0.480 "

Air Cap..., . ....... ..._.... .... ... ...... . . 76.520 "

Source1Volu.ae: 142924, cubic centimeters HATERIAL DENSITIES (g/cc);

Material Source Shield 63r EaD Air .001220 Iron- 7.860 Vater 1.0 BUILDUP FACTOR: based on TAYLOR method.

Using the Characteristics of the materials in the source region.

INTEGRATION PARAMETERS:

Number of lateral angle segments (Ntheta).. .. . 5 Number of azimuthal angle segments (Mpsi). ... . 5 Number of radial segments-(Nradius). ..... . .. , 5 SOURCE JUCLIDES:

Co-60: 3.0000e-02-curies RESULTS :

Group Energy Activity Dose point flux Dose rate

1. ( Me v ), _ _ _ '(photons /sec) MeVf(sq gm)/see - (mr/b r) __.

I 1.3359 ' 110e+09 7.591e+03 1.370e+0L 2 1.1797 110e+09 6.736e+03 1.252e+0!

3 .6953 t.811e+05 7.741e-01 1.594e-03 TCTALS: 2.220e+09 1.433e+04 2.522evol 8-16

SAR 5/91 i

l The linear absorption coefficient depends on the mass attenuation coefficient and the density, p - (p/p )p For water and concrete the respective mass attenuation coefficients and densities are as follows:

p/p p water 0.060 cm 2/g 1.0 g/cm 3 conc re te 0.0567 cm2 /g 2.9 g/cm 3 Calculating px for the distances in water and concrete determines the shield attenuation and build up effects.

In water, 2 3 px -

(0.0600 cm /g) (1.0g/cm ) (405 cm) ,

- 24.3 .

In concrete, px - (0.0567cm2 /g) (2.9g/cm 3) (122 cm) ,

- 20.1 The product, px, not only determiaes the shield attenuttion but also relates to the scattering buildup within a shield thick enough to cause multiple scattering.

The build-up factor, B, is calculated from the er.pression Al e "1M x + Aze' 2x .

where A t, A 2 , 01. ^'d 02 are constants (71 For water the buildup factor constants have the following values (E) 1.35):

At - 8.5 A2

- 7.5 at

- -0.093 a2

- 0.064 TLu; , the build-up factor may be expressed as B -

8.5e&O.093(24.3) , 7,7, 0.064(24.3) ,

- 81.5 - 1.5 ,

- 80 .

8-18

SAR 5/91 l

Chapret 8 References

1. Me cha nLg.aLJ2pe ra t ion and Maintenance Mattual, University of Texas TRICA Mark II Reactor, February 1989, General Atomics (with supplements).

2 " Review of Experiments for Research Reactors", ANSI /ANS 15.6 -

1974 (N401).

3. Technical Bulletins Nos 1 30, The University of Texas at Austin TRICA Reactor Facility, (some incomplete).

4. CRC Hanstbook of Laboratory Safety, Second Edition, CP,C Press Inc.,

Boca Raton Florida 1971.

5. " Standards for Protection Against Radiation," Part 20, chapter 10, U.S. Code of Federal Regulations, 1991.

6. tilGROSHIELD Users Manual, Version 2, Grove Engineering, Rockdale, Maryland, 1985.

7. A. B. Chilton, J. K. Shultis, R. E. Faw, Principles of Radd.i_q lga ShLt.bling , Prentice Hall Inc., 1984 8-22

4 1 e kk ifs

• 1 '? sgh*

IMAGE EVAL.UATION TEST TARGET (MT-3)

/ ,/// e /[ Y @ g

' *f, y,,

+ 4ks, q?

i

+ 'e l,Q 'if Eld Edd

[][jf EM i,i [w L24

{lj]=i.8 ijn =

I.25 1.4 i.6

}

4 150mm - ~-*

6" #

4 p>u % ++ A s,y, +g,,%r

. yc -

%u4+ 1 0  ; 9.

1. p

' h_. .. . -_s ud .

Q

r SAR 5/91 4

dNg q-

-~~ - _ ~ . N r i (1) dt V t with the solution, t- .

q <N r  !

Nf (t) - (1 o'AC), (2)

V A where, Ng - number of atoms on the filter, Nr - number of atoms within the room, q/V, (, A - facility and material constants.

Figurn 93 represents a plot of equction 2 for the particulate monitor with typical background conditions and the following assumptions:

6.5 x 10' cm3/mla q/V - - 1.35 x 10 6/ min .

4.83 x 109 cm 3 A -

(In2)/(2 houra) -

.346/hr .

< - 98 percent At equilibrium, the saturation condition determinas the number of atoms. Fora filter count rate of 2000 cpm with a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> h.ilf life isotope the results are Nt - ( t-= ) - - 9.9 x 105 atoms.

.35 In(2) ht on file.er

.\nd q cN r Ng(t-=) - - --- ,

V A Nr./c 9.9 x 105 /.98 (.346/hr)

Nr " A~

(q/v) (1.35 x 10-6/ min) (60 min /hr)

- 4'.32 x 10 9atoms or .89 atoms /cm 3 At this concentration t e filtegcount an activity of 2.5 x 10~

rate of 2000 cpm corresponds to pCi/cm for a two hour half-life isotope.

9-11 1

cm.-

1 4 ,

SAR 5/91 4

Monitor position to sample reactor room air is within 5 meters of the pool at t'ae pool 4.ccess level. The location will sample air activity in the vicinity of the reactor pool. Leakage of fission products from the fuel into the. water then into +.he room would occur at the room air to pool water interface. Background measurements of air particulate activities between Sept. 90 and Sept. 91 provide a record of '

the naturally occurring count rate levels for the Ludlum Model 333-2 in room 1.104 of the _ NETL facility. These data indicate that count rates of 4000 cpm to 6000 cpm will occur several times each year as a result of weather conditions that effect vertical air stability such as, frontal lines . temperature itiversions - and storm systems. A set point at

-5000 cpm will provide an alert level with an occasional alert for a natural occurring condition.

The count rate alarm set point will assume beta energies of 0.3 ,

and detector efficiency of 30%. The set point may now be calculated:

dis / min 0 a - (4.44 x 10'3 3

) (6.5 x 10 cm /3 min) (120 min) (0.3) cm a 10,400 counts / min.

Refer to Table 9-1. Most of the isotopes listed in the 1 day column have beta energies greater than 0.5 MeV. Since the detector efficiency increases with the incident beta energy, a more representative estimate for the detector efficiency may be 50%. The set point may be calculated as follows:

dis / min 0 a -

(4.44 x 10'3 ) (6,5 t 10 cm /3 min) (120 min) (0.5) 3 em

= -17,300 counts / min.

A particle accumulation time, t, of two hours may be considered, as shown in Figure 9 3.

9.5.2. Argon-41 Monitor-

~

L Set points for the argon *1 continuous air monitor should warn of excessive . radiation levels for effluent release and occupational exposure. This radiation monitor will operate whenever the reactor system and the at.tillary air purse system are operating.

As specified ir 10 CFR 20. - the reference concentration of argon 41 is-1 x 10'8 pCi/cm . Dividing this number by the purge exhaust system flow rate and by the building wake dilution factor yields the averagg annual3 concentration limit for release at the stack, which 1.g 2 x 10'3 501/cm An alarm set point at ten times this level, 2 x 10' pci/cm '

will warn of'an excessive daily release. In the event of a gaseous fission product release in interference will occur in the argon-41 count due to betas emitted by isotopes of krypton and xenon, refer to Table 9-2.

9-15

.. . . .-~

i - .

SAR 5/91 s

4 I.

i ne ent..r,it y o f te. .. it am.t.n Organisatton

,- 1 I- Offi e of the Presadent  !

, the university of rosa.

at av tin

} 7_ ,

vice ere.ident for : - taiversity safety 3

i Susaness affaire j i Office

) L_ ,

.  ! l l-1 IRadiationSafety  ! Radiation Safety lcomalttee ] officer-f aecutive Vice President and Provost I

Sean of the College of Engineertnq y

j F.

. Ottector of NETL I

f" Mechanical chairman ofEngineerinq the Dept. $ i 4

7 . _ _

Nuclear teactor 4 Reactor Comalttee j l Supervisor s a -

l j suclear taq tneering teach ang Laborat ory

, organtaatton 1

0 Ontactor of Muclear Inqtneerinq  ; Radiation te ac n t nq *.a tka r t ory {$4fetyOfftcer 4

1 l l

lf6ealth Physics f**d

Administrative

• Staff a

fAsssstant OLrector j Reactor supervisor i

Cert A f Led operators

{nosearcastafr l a

ADMINISTRATION Figure 10-1 10-2

sAR $M1 10.1.3.1. R dfing all activities that reqvite the presence or lie nw cert tiied l

operators will also require the presence in the facility cenplex of a )

second person capable of performing prescribed written instruccions l Unexpected absence of a second person for greater than two hours will be i acceptable if immediate action i. s taken to obtain a replacement. A designated license certified seniet operator will be readily available on call during all periods in which activities requiring a certified operator are being performed. The person on call will be considered available if the time to initiate a call request and respond on site is less than 1/2 hour.

• Movement of fuel or control rods and relocaticn of experiments '

with greater than one dollar reactivity worth will require t%e presence of a . license certified senior operator. Other activities, such as initial startup, recovery from unscheduled shutdowns and modifications to instrument r,ystems, control systems, safety systems, radiation measurement equipment or engineered safety features, will require concurrence and documentation by a license certified senior operator.

Operation of reactor controls, movement of reactor experiments, '

maintenance of instrument control', safety, and radiation measurement systems will require the presence of a license certified operator, A license certified operator will be present in the control room whenever the reactor is not shut down by more than one dollar of reactivity or the control and system console panel is not secured. l The staff required for performing experiments with - the reactor will be determined by a classification system specified for the experiments. Requirements will_ range from the presence of a certified operator for some routine experiments to the presence of a senior operator and the experimenter for other less routine experiments. Some other activities that occur in the area of the reactor will require knowledge of a 1icense - certified operator, but not necessarily the presence of the operator, Such activities will include maintenance, handling of radioactive materials and experiment preparation.

10.1.3.2. Procedures s.

Written procedures shall govern many of the activities associated with - reactor operation.- Preparation of the procedures and minor modifications of the -procedures will be by certified operators.

Substantive changes or major modifications to procedures, and prepared procedures will be submitt-d to the-Nuclear Reactor Committee for review and approval. Temporary deviations from the procedures may be made by the reactor supervisor or designated senior operator provided changes of

. substance are reported for review and approval.

Activities subject to written procedures will include routine startup, shutdown and operation of the reactor: fuel loading, unloading and movement within the reactor; and routiae maintenance of major l- components of systems that could have an effect on reactor safety, i

I-l~

10-6 I

l

- - - _ _ _ _ ~ - . _ _ _ _ . - - _ , _ _ _ . _ _ _. _ __. . , m _ .

m -m ._ m__.. . _ _ _ __ _ _ -- >_ _ _ . _ .. _ _ . _ . - . _ _ .. _ _ _

SAR 5/91 4

10 .1. .+ . 4 , Qther Reportable OccuIrences Other events that will be considered reportable events are listed j in this vction. A return to normal operation or curtailed operation until authorized by management will occur. Appropriate reports shall be submitted to license authorities. (Note: L'he r e components or systems are provided in addition to those required by the technical specifications, the failure of cortponents or systems is not considered repor:able provided that the m inimwn nwnbe r of components or systems specified or required perform their intended reactor safety function.)

j a. Operation with actual s a fe t y - s y s t ern settings for required j systems less conservative than the limiting safety system j settings specified in tne technical specifications.

i b. Opetation in violation of limiting conditions for nperation i established in the technical specifications unless prompt i- remedial action is taken.

l: c. A reactor safety system component malfunction which renders f or could render the reactor safety system ineapable of ll perfortting its intended safety function. unless the i malfunc e: ion or condition is discovered during maintenance

tests or periods of reactor shutdowns.

I d; An unantic; pated or uncontrolled change in reactivity j_ greater-than one dollar. Reactor trips resulting from a '

q- known cause are excluded.

e. ~ Abnormal and significant degradation in reactor fuel, or '

cladding, or both, coolant boundary,. or confinement _ boundary j

j. (excluding _ minor leaks) where:-applicable which could result iL in - exceeding -prescribed radiation exposure limits of l _ personnel _ or environment, or both.

f. An observed

nadequacy . in implementation of adrainistrative

• i or procedura' controls such that the inadequacy cauces, or

! could have cau:ed, the existence or development of an unsafe i condition with regard to reactor operations.

10.I.4.5. Other Renorts

• j.

'. A written report _ within 30

days - to the chartering or licensing authorities of

a. Permanent _ changes _ in the facility organization involving i Director or Supervisor.

I b. -Significant changes in the transient.or accident analysis as l described in the Safety Analysis Report.

t

10-l.5.. Records Records of- the'- following- activities shall be. maintaired und i retained for the periods specified below [4). The records may be'in the-t form of ' logs , Edata - sheets or other suitable forms. The required I

information may be contained in sin 6 1 e or multiple records, or a

' combination thereof.

f 10-9

SAR 5/91 4.

10 1 5.!- LiitLitiLEeGL1E 1if4 -ime records are records to be retained for the lifetime of

• Le teact , facility (Note: Applicable annual reports, if they contain all at the required information, may be used as records in this

<ection.)

a. Caeeous and liquid radioactive effluents released to the enViroMS,

b. Offsite e nv i r onme nt a l monitoring surveys required by Technical Specifications.

< Events that impact or effect decommissioning of the f.u:t1ity.

d. Radiation exposure for all personnel monitored.

e l'pdated drawings of the reactor facility 10.. 5.' l'! w Yea LEerind Records to be retained for a period of at least five years or for the life of the c o:r p o ne n t involved whichever is shorter.

a. Normal reactor facility operation (supporting documents such as checklists, log sheets, etc. shall be maintained for a period of at least one year).

b. Principal maintenance operations.

c hpottable occurrences.

d Surveillance activitiet required bv technical 3pecifications.

e Reactor facility radiation and contamination surveys where required by applicable regulations.

f. F.xpe r iment s nerformed with the reactor.

g. Fuel inventories, receipts, and shiptrent4

h. Approved changes in operating procedures _

i. Records of meeting and audit reports of the review and audit group.

10.1.5.3. One Irainipg. Cycle Training records to be retained for at least one license cycle are the requalification records of certified operations personnel. Records of the most recent complete cycle shall be maintained at all times the individual is employed.

10-10

_ - - - - ^ - -

SAR 5/91 5

4 i 10I7,5. Acceptance for Qncuth.D

-The' final startup program phase shall consist of the resolution of -

, all deviatiens from specifications. Deviations should be -resolvee .as l

.specified for quality . assurance : or other .- methods determined to be -

acceptaM e. -Three months after completion of requisite-initial startup

~

and power escalation testing ~ of. the reactor, or nine months after initial. criticality, a written report shall be submit ted to licensing authorities. The report shall include:

, Cha rac te ris ti c s of ihe reactor such as critical mass, _ excess reactivity, power calibration, control rod calibrations, shutdown margin E and experiment facility worths, describing the measured values of the -

operating conditions including:

a. Total control reactivity worth and reactivity of the red of highest reactivity worth,

+

b. Minimum shutdown margin of the w.ctor both at ambient and operating temperatures

-e. ' An evaluation of facility performance to date in comparison with design conditions and measured. operating characteristics, and a reassessment of the safety analysis

. when measurements indicate that there msy be substantial varicace from prior analysis submitted with the - license application.

Results of tLa startup program shall become a supplement to The-University of Texas TRICA Safety Analysis Report. Chapter 12 of the report will contain results of the startup program.

10-31

., . _ - _ _, . . . - . . _ - - - . , _ _ _ - . - .__ __