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| document type = CORRESPONDENCE-LETTERS, EDUCATIONAL INSTITUTION TO NRC, INCOMING CORRESPONDENCE | | document type = CORRESPONDENCE-LETTERS, EDUCATIONAL INSTITUTION TO NRC, INCOMING CORRESPONDENCE | ||
| page count = 114 | | page count = 114 | ||
| project = | |||
| stage = Request | |||
}} | }} | ||
Latest revision as of 05:37, 24 September 2022
| ML20100C683 | |
| Person / Time | |
|---|---|
| Site: | University of Utah |
| Issue date: | 11/29/1984 |
| From: | Devries K, Sandquist G UTAH, UNIV. OF, SALT LAKE CITY, UT |
| To: | Thomas C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8412050551 | |
| Download: ML20100C683 (114) | |
Text
{{#Wiki_filter:-. . , t 1 - DEPARTMENT OF UNMEr&Y .~es=~eu.,~. 0 UTAH th,_,,, November 29, 1984
. Cecil 0. Thomas, Chelf
' Standardization & Special Projects Branch Division of Licensing U.S. ' Nuclear Regulatory Comission Washington, DC 20555.
Dear Sir:
The enclosed materials are submitted in support of the University of Utah's applicatior for renewal of its TRIGA Nuclear Reactor Operating License, R-126, Docket No. 50-407. The applicant (University of Utah) formally submits the following:
- 1. Responses to specific questions posed by the NRC on the facility, operations and assessments.
- 2. Re producible figsres required for preparation of the Safety Evaluation Report.
- 3. Corrections and amendments to the applicant's Safety Analysis Report.
The applicant has requested that the NRC grant a 20 yoar operating license for_the TRIGA Reactor Facility.
/Sincer(Ty.yo/ur d[ aeX C . n qui e Supe, L kg [ Eh4.u. -7 K. L.'DeVries Reactor Administrator GMS:ffc Enclosures j At 8412050551 841129 gf y/ DR ADOCK 05000 o e .
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a .... . UNIVERSIlY OF UTAH -TRIGA REACTOR (UUTR)
- 1. What are the principal uses of the UUTR? What is the . current use in
. megawatt-hours per year?
The University of. Utah's TRIGA Reactor (UUTR? is principally used in support of. its educational program in Nuclear Engineering and for research. Current use averages about 10 Mw hrs per year.
/
- 2. Provide a plan view of the current core configuration showing the number and . locations of the fuel elements (differentiate between stainless-steel and aluminum elements), the control rods, the D 2 0 reflector elements,
, the Dg0 reflectors tanks, the experimental tubes, the irradiation-facilities, the neutron detectors, and the startup source. Also indicate -
how many and where (if any) the instrumented fuel elements (aluminum or stainless-steel) are in the current core and what type of thermocouples are used. The Figure enclosed provides a plan view of the current core configura-tion. The single instrumented fuel element in position C3 is stainless steel clad. Three thermocouples (chromel-alumel) provide channel output
.for fuel / fuel temperature sensing.
- 3. What administrative limits or requirements are placed on fuel loadings?
The UUTR Technical Specifications limit fuel loading within the core by . reactivity and control requirements (Sec.3.1) and by cladding and fuel type (Sec.4.1). Fuel loading operations require the approval of the Reactor Supervisor and presence of a senior licensed reactor operator as specified by the Nuclear Engineering Laboratory Operations Manual for the UUTR.
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UNIVERSITY OF UTAH Shut down : $0.83 Excess React: $1.18
% Core Configuration #18 Safety: $1.76 Shim: $1.55
. April 25, 1984 ,
North Req: $0.46 i
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Wat is the current startup 4.-What is the total 2350 content in the core? source and what is its strength? - U-235 content. in core is 3.348 kg. Startup source is a 5.0 Ci Pu-Be source.
- 5. What are the measured excess reactivity and control rod worths in the current core configuration? <
Excess reactivity $1.18, Safety rod $1.76, Shim rod $1.55, Regulating rod $0.46, Shut down margin is $0.83..
- 6. What is 6-effective for the UUIR current core configuration?
Beta effective = 0.007.
- 7. How many spare fuel elements are there? How many have been irradiated?
Where are they stored? . Describe the fuel handling tools used and fuel storage facilities at the UUTR. Give details of the current inventory of fuels in these storage areas. There are 3 stainless clad and 45 aluminum clad fuel elements stored in the reactor tank outside the core or locked fuel storage pits. All have been irradiated (see fuel allocation table.) The fuel handling tool is the standard GA flexible cable with manual actuator which locks on the chamfered top of the element: Reactor Tank Fuel Storage Pits 43 A2 Clad 2 A2 Clad 3 SS Clad 0 SS Clad r , i l l l
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- 8. Provide a summary of the fuel element specifications for the stainless-steel-clad elements that .are used in the UUTR. Include a schematic drawing of a stainless-steel-clad fuel element assembly.
See Attached Drawings.
- 9. Describe the facility electrical power system and list all controls and instrumentation that are provided with emergency back-up power. Describe the emergency battery supply system.
The reactor console power is provided through an isolation transformer from building line power. The reactor console powers all reactor operations not including auxiliary functions such as lighting, overhead crane, recirculation pump, and facility security systems. The emergency power supply system powers facility security and lighting. The system (12 volts DC) consists of a series of large liquid acid batter-ies continually under recharge while line voltage is available. In the event of line power failure, the emergency power supply continues to provide power to the security system and lighting. If the reactor is operating during AC line power loss, the control rod magnet power is also lost, automatically scramming the reactor. However, the 12 volt DC emergency power supply continues operation, maintaining the radiation monitoring system and other emergency equipment.
- 10. Provide information on the dimensions of the neutron source holder and the manipulation of the neutron source during startup and power opera-tion.
See attached drawing. The neutron source is always located in the reac-tor core except when the power level of the core exceeds 1 watt of thermal power. Then the neutron source is removed from the core via a stainless steel cable attached to the source and stored during power operation in the tank wall storage rack.
- 11. What is the type and form of neutron poison contained in the control rods? What are the dimensions and what is the vertical travel length of the control rods?
See figure attached. Neutron poison in control rod is boron-carbide, vertical rod travel is 15 inches.
5
- 12. What is the withdrawal speed of the control rods and what is the accuracy of the rod position indication?
Control rod withdrawal speed is 3.75 inches per minute or 4 minutes to drive a 15 inch rod. Accuracy of. rod position ~ repeatability is about 2% or $0.03 for the shirs rod and i $0.01 for the regulating rod. Control Rod Worth Avg. Rate Max. Rate Safety $1.76 $0.44/ min $0.91/ min Shim $1.55 $0.39/ min $0.91/ min Regulating $0.46 $0.12/ min $0.29/ min
- 13. Provide information on the scrams and interlocks associated with the four neutron channels. Include set points and automatic actions (if any ) .
Scams and interlocks are specified in Section 3.3 of the Technical , Specifications. 1 s
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- 13. Provide information on the scrams and interlocks associated with the four neutron channels. Include set points and automatic actions (if any) .
Scams and interlocks are specified in Section 3.3 of the Technical Specifications. Specification The reactor shall not be operated unless the safety systens described in the-following table are operable. -
~
Safety System or Measurina Minimum Number Channel Operable Function Scram at or below limiting i Fuel element temperature 1(a) Safety System Setting Reactor power level 2 Scram at 120 per cent of full licensed potier Manual console scram button Manual scram ( 1 Magnet current key switch 1 Manual scram Console power supply 1 Scram on loss of electrical power Reactor tank water level 1 Scram at one foot below normal operating level Startup count rate interlock l Prevent control rod withdrawal when neutron count rate is less than 2/sec.
. Control rod withdrawal interlocks all Prevent manual withdrawal control of more than one control rods rod simultaneously.
(a) Fcr p'eriods of time for maintenance to the standard thermocouple fuel elerant, the reactor shall be in the shutdown condition with all control rods fully inserted, and, power to the control-rod magnets and actuating solenoid has been switched off and the key removed.
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- 14. Wat type of signal initiate scrams? Indicate if the meters open circuits, relays, or initiate other actions.
Signal: high power signals will trip set points on the percent power channel, high log power channel, and linear power channel. Action: relays cut magnet current and conne.ct scram indicator lights. Signal: high temperature signal will trip the set point on the fuel temperature circuit. Action: relay cuts magnet current and connects scrams indicator light. Signal: high radiation will trip set points on the area radiation monitor. Action: Relay cuts magnet current, connects scram indicator light, and trips security relay to warn campus polich. Signal: loss of high voltage power supply to core power detectors will trip relay. Action: relay cuts magnet current and connects scram indicator light.
- 15. Can the UUTR reactor scram on a loss of high voltage signal to the neutron detectors and/or on loss of power to other individual instru-ments?
Loss of high voltage to the neutron detectors trips a relay in the power supply unit which trips a relay in the scram chassis and thus scrams the reactor. Other individual instruments, such as the area radiation monitor and fuel temperature system have low level scrams, while the conductivity meter and ventilation have no power loss scrams.
- 16. Are there any rod withdrawal interlocks other than the source level and two "up" switches depressed at the same time? List those, if any, associated with the mode selector switch or the armature or lower rod limit switches. 7 The Shim / Safety and Regulating rod are interlocked with each other and the safety rod. Only when the safety rod is fully withdrawn and latched -
can the shim / safety or regulating rods be withdrawn. Upon a scram signal from any source to the scram buss, magnet power to all rods is inter-rupted, dropping the rods and the rod carriages are driven into their fully down position. Only one control can be withdrawn at a time.
- 17. -Are there any provisions in the scram logic circuitry and interlocks to exclude a loss of function on a single component failure?
Scram of the UUTR is realized by AC power interruption to the control rod magnets. The standard TRIGA circuitry and interlocks are designed so that no single component failure can prevent scram capability. This is accomplished through a common buss for all scram signal inputs.
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- 18. What provisions are used to prevent ganged rod withdrawal?
The solitary selector switch for rod withdrawal can only be set to one rod selection at a time.
- 19. Can the control rods be scrammed individually?
The control rods can be scrammed individually by interrupting magnet power to a given rod from the console. This process is used to determine individual control rod reactivity worths.
- 20. How is the pool bulk coolant temperature monitored and what is the frequency of temperature monitoring during reactor operation?
A thermocouple which can be rear at the console is located in the primary coolant circulation system. Also thermo couples which are located in the reactor tank and used for thermal power calibration, can be used to monitor tank water temperature. The primary coolant temperature is_ read during each reactor power operation as required by the checkout and operational logs. An alcohol thermometer situated in the reactor tank is used to verify the accuracy of thermocouples.
- 21. What is the maximum fuel temperature that has been attained for the UUTR? Indicate in what ring the instrumented element was located and whether the center irradiation facility replacing the B ring elements was installed.
The highest fuel temperature attained was 112 C while operating at 90% poner (90 kw(th)). The instrumented fuel element was located in the C-ring with the D 2 0 filled central irradiator filling fuel positions - in the A and B rings.
- 22. What restrictions are placed on the samples that are irradiated in the pneumatic transfer system? Describe the pneumatic transfer irradiation facility (materials of construction, dimensions, terminals, propulsion method, and cuntrols).
All samples irradiated in the pneumatic transfer system must satisfy technical specification 3.6 and are always encapsulated and sealed in a polyethelene sample vial and then inserted in a polyethelen pneumatic transport vial (see figure). A detailed figure of the pneumatic rabbit is enclosed. All construction materials are aluminum except the damper spring which is stainless steel. Air is used to propel the sample in and out of the core. The control system for insertion and removal is electronic employing a time and air- value relay system. Fissile samples are not permitted to be irradiated in the pneumatic system under the provisions of the Nuclear Engineering Laboratory Operations Manual for the UUTR.
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- 23. How ~ 1s the 41A r from the irradiation of air in the pneumatic system handled.
Exhaust air from the pneumatic system is vented to the reactor room exhaust system and is monitored by the ventilation system detector. See figure for ventilation system schematic..
- 24. Please address the following accidents: include all assumptions made, calculative methods used, accident scenarios, and references to any documents.
(1) Fuel element failure in air (2) Rapid insertion of the maximum allowable excess reactivity ($3.00) (step nuclear excursion) (3) Loss-of-coolant accident (4) Mechanical rearrangement of fuel (the consequences of dropping 1000 lb steel cask into core). The following accidents: (1) Fuel element failure in air (2) Rapid insertion of $3.00 reactivity into the core (3) Loss-of-coolant accident have all been carefully addressed in the SAR in Chapter 8. The analyses given were very conservative and experimental measurements made on Ar-41, N-16, and other radiation values confirm this conservency. An excerpt from the SAR covering these accidents is provided for review. The fourth accident (i.e.1000 lb cask dropped into core is addressed) here. (4) Mechanical rearrangement of core fuel to an dropping of the 1000 lb storage cask into the core tank. The dropping of the cask into the reactor tank is considered very unlikely due to the administrative and physical constraints imposed upon reactor crane operations. First, the crane is disabled by a lock-ed power supply box except when expressly approved for operation by the reactor supervisor and under his direct supervision. Second, the crane is not to be operated when the reactor is in operation. Third,
.an additional safety cable is attached (by, side body bolts) between the cask and the moving head of the crane whenever the cask is used over the reactor tank.
If the cask were to drop from crane, the heavy, structural beam bridge network that supports the control rod drives would at least deflect the cask so that it would not fall directly on the core. Tne control rods are bolted into the core assembly and any tilting or tipping of the core would result in the fuel moving from the core to the tank bottom into a subcritical geometry. It is possible that a fuel element could be damaged so that the cladding would release some fission products to the primary water. How-
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4 i 8.5 ' Reactivity Accident
.The reactor fuel loading under-the Techni cal Specifications can provide a maximum of 2.25% Ak ($3.00) _ excess- reactivity above a cold, critical condition. In this section it is -conservatively assumed that the core is loaded so that all of the innermost fuel positions contain fuel, starting.
from the . central fuel (or 'B) ring, which. consists of six elements. The
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maximum reactivity transient' which could possibly occur would be that pro-duced in such a densly packed core by the accidential insertion of the entire available amount of excess reactivity instaneously. Prior to the development of the pulsing type fuel used on present TRIGA' pulsing reactors, prototype TRIGA reactors (e.g. Torrey Pines, etc.) at Gulf Energy . and Environmental Systems, . Inc., using the same type of aluminum clad, low hydride fuel as is to be used in this reactor, were pulsed safely many times with up to 2.25% ak insertions. The - resulting power excursions attained a peak power of 1500 MW on a reactor period of 4.0 msec,
, .with a total energy release during the burst of approximately 20 MW-sec. The maximum measured fuel temperature of all these pulses was less than 500 C.
Curves of the typical transient power level and fuel temperature resulting from such power excursions are shown in Figure 8.5.1. These tests demonstra-l. ( ted that even a TRIGA reactor at zero -power using non-pulsing type ' fuel . F-l can withstand a $3.00 reactivity insertion without hazard to either operat-I ing personnel or the public. By deliberately violating the operating license and the several inter-
-locks and scrams a sudden insertion of reactivity is possible while the l reactor is at a power level up to 100 kW. From operating data reactivity loss due to temperature and xenon poisoning from operation at a steady i
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power of 100 kw is about 0.488% (or $0.65) and the average fuel element temperature is '70*C. Now to estimate the fuel temperatures resulting from a prompt reactivity insertion at power the Fuchs - Nordheim model will be used. Parameters used in the analysis here are as follows: a = negative temperature coefficient of reactivity
= 1.34 x 10-4 /*C C = heat capacity at equilibrium power kilowatts
= 0.718 x 105 + b (T(fuel temp.)- 25) watt-sec/*C
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Co + T Co = heat capacity at 25 C = 0.718 x 105 watt-sec/ C b = heat capacity rate of change with temperature
= 0.825 x 102 watt =sec/ C Po = 100 kw p = reactivity above prompt critical
= ($3 - $1) = $2 = 52 x 0.7 x 10-2/$ .1,4 x 10-2 This model which averages parameters, neglects heat transfer and delayed neutron effects yields the following set of coupled differential equations for reactor power P and fuel temperatures T:
O 1P = (p- aT)P O CT = P-Po These equations may be combined to give dP (p-aT) (Co + bT)P dT 1 (P-Po) and integration after separation of variables yields
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P~ T abT2~ l (P-Po) - Po in - =T bCo + (bp-aCo) - 2 3 _, Po. i
The maximum average fuel temperatures occur when dT
= 0 = P-Po dt-so P = Po and Tmax is found from the expression r .
7 - 16 A I 2p Tmax = -3/8(A-1) 1- )1 + 1/2 = 202.4*C 3 (A-1)2 _d ac 1.34 x 10-4 x 0.776 x 105 where A= ~ = 9.00 bP 0.825 x 102 x 1.4 x 10-2 The result for the core-averaged fuel temperature at the conclusion of the pulse is T = Tfuej + Tclad = 202.4 + 70 + 25 = 297.4 C The maximum fuel temperature in the hot test fuel element for a peak-to-average power ratio of 1.45 for the UUTR is found to be . Tmax = 431*C (in the hottest element) which is well within the 500*C range found from pulsing at zero power. Thus pulsing the reactor while at power up to 100 kw produces no significantly higher fuel temperatures than pulsing from zero power. This anlysis conservatively neglects any transfer of thermal energy during the step impulse and assumes that the system parameters remain constant. Experimental data has shown that some transfer of thermal energy does occur during the pulse and system parameters to change so as to reduce the power output from the pulse. Furthermore, without a specially design pulsing rod it is difficult to achieve any significant reactivity insert-ion within a few milliseconds. The first cladding failure in an aluminum clad, low hydride TRIGA fuel element was detected by general atomic in November 1959 after the element had been subjected to 709 pulses 410 of which were at or about $3. Sub-s
sequent investigation showed that failure was caused by ratcheting of cladd-ing along the fuel body. None of the UUTR aluminum fuel has ever been ex-posed to pulse operation and the probability of failure of aluminum clad in a single, accidental pulse is extremely low. 8.6 Loss of Reactor Pool Water 8.6.1 Total Loss of Reactor Pool Water Rapid and complete loss of the water in the reactor tank is con-sidered to be highly improbable. Such an event could only reasonably arise from an earthquake shock so intense that the reactor hazard would only be one of many serious hazards to persons in the Merrill Engineering Building. Furthermore, the UUTR primary coolant system employes two water tight tank enclosures, viz an 8 foot diameter aluminum tank which is contained within an outside 12 foot diameter steel tank. Probability of simultaneous failure of both tanks is very low. However, provision has been made for warning of the loss of pool water as an isolated event and calculations have been made of the fuel temperatures tnd radiation levels to be expected. A float switch alarm, operating on a continuous basis is provided in the tank to alert the operating staff and Campus Police Department it the water should reach about 6 inches below its normal level. In such event, the operating staff and/or the Campus Police Department will alert the Reac-tor Supervisor, and the Radiation Safety Officer immediately. If the l water lerel were to fall while the reactor is or is not operating, the i ceiling area radiation monitor over the reactor would also alert the operator of the high gamma radiation level and the reactor would scrammed automatically from either or both a low level water signal and a high radiation alarm and the area cleared. The operator would then attempt to i i . - - - - - . - , , - _ , _ . - , , _ _ _ _ _ _ _ - _ _ - _ _
maintain an adequate level of water over the core if possible by directing - an emergency water line (tap water) into the tank. The reactor cooling system and return lines contain anti-syphon holes approximately one foot below the normal water level to prevent the loss of significant water. The time it would take for the tank to 'crain may be conservatively estimated by considering a cataclysm that results in the complete opening of both containment tanks and the concrete pad to the ground soil at the bottom of the tanks. Under these conditions the flow of water through the soil is described by Darty 's law, where p is the pressure gradient dx
-= .Kp in meters of water head per meter and K is a constant depending on the nature of the soil, having the units m/ year. The time to drain the tank is then t = 1/K, years / meter (H2 0). The premeability K has a maximum value for sand of about 3 x 103,1 Thus to drain the tank to the bottom of the core (22 ft.) or 6.7 meters the time required is 0.0022 years or 19.3 hours.
This maximum drain rate allows ample time for emergency procedures to be enacted. Again the primary emergency response procedure is the continual flooding of the core tank from the tap water line in the laboratory until > radiation levels are acceptable, or the fuel can be removed to an acceptable storage facility. 8.6.2' Integrity of Fuel Element Cladding The following calculation shows that rupture of fuel element clad-ding is not likely following loss of cooling water from the tank. The calculation assumes that the reactor has been operating continuously at full power for one year prior to loss of the water which is considered to be lost instantaneously following reactor shut-down. The maximum tempera-
- 1. Handbook of Applied Hydraulics, McGraw Hill, New York (1952) p. 165.
- Note, in Section 5, internal pressures of up to 15 atmospheres at the time of 10% burnup were conservatively assumed. Even such a high pressure gives clad-ding stresses well below the yield.
ture reached by the fuel an consequently the aluminum cladding temperature is less than 61'C (152*C_ 9 250 kw). This temperature is such that .the pressure exerted by trapped air and fission product gases in the fuel is less than 30 psi. Such a pressure produces an internal stress of 264 psi (660 psi 9 250 kw). The yeild stress for the aluminum cladding is about 8,000 psi at 150*C.* Therefore, the fission products will be retained in the fuel' elements following operation at either 100 or 250 kW even with instaneous loss of coolant. It should also be noted that the calculated temperatures are conserva-tive (high) and assume that the only mechanism for heat removal from the fuel after the watger loss is through natural convection to ambient surround-ing the core. Actually some heat will be removed by conduction to the grid plates, evaporation of water film on the fuel and some will be removed by radiation. Method of Calculation use has been made of the two-dimensional, transient heat transport computer code, RAT, developed at Gulf Energy and Environmental Systems, Inc., for calculating the maximum temperature in the core after a water loss. It is assumed that at the time at which the water was lost, the tempera-ture distribution in the fuel element considered in this calculation was equal to the temperature distribution in the hottest (B ring) fuel element during steady-state operation of the reactor at 250 kW. The results of these calculations are obviously conservative for estimating the maximum core temperature at 100 kW operation.
~ _ - . _ -
- It _is also conservatively assumed that the reactor has been operating for a long time at 250 kW, and with a minimum fuel loading (only 62 elements) in the core. The rate of energy release in the B ring element was determined from consideration of the energy deposition of fission product gammas and betas only. The energy release -from delayed neutrons is relatively small (about 1700 watt-sec total in the B ring element) and has an average decay constant of abo'ut 0.08 sec-I.
The after-shutdown power density (in Btu /hr-ft 3 ) in the B ring fuel element is given by: W L q P
-=0.1pV v cos [0.78L -(x -)]
f 2
- 7
-0.2 -0.87 7 -0.2 -0.05 b (1) t+to + 10 t+to + 2 x 10 -
where p = Peak-to-average power density in the core = 2.0 P = Operating reactor power = 8.525 x 10 5 Btu /hr(250kW) Vf = Volume of the fuel in the core = 0.87 ft3 L = Length of the fuel = 1.25 ft x= Distance measured from the bottom of the fuel element, ft l. i t = Time after the core is exposed to the air, sec to = Time from shutdown to the time the core is exposed, sec. Equation (1) is a modification of the Untermeyer-Weill formula that matches the work of Stehn and Clancyl to about 5 x 104 sec after shutdown. It is also conservatively assumed that all the energy produced by fission product decay in the element is deposited in the element. While the decay gammas and betas are raising the fuel element tempera-ture, the flow of air between the fuel elements will be removing heat, tend-
- 1. J. R. Stehn and E. F. Clancy, " Fission Product Radioactivity and Heat Generation," Paper No.1071, Proc. Second United Nations Con on the Peaceful Uses of Atomic Energy, Geneva, 1958.
- - ' '
- a -~- --,,---.-.,,-..m,,.-,,.,,_y,, _ , . . , , , . _ , ,.
ing to lower fuel temperature. The air velocity through the channel can be determined by setting the frictional pressure loss equal to the buoyancy. Entrance and exit losses which are about 2 to 5% of the friction losses and have been ignored. 6P (buoyancy) = 6P (friction) * (2) Term on the left is given by L 6P (buoyancy) = (po - pj) - (3) where L is the length of the channel and po and p1 are the entrance and exit air densities, respectively. Since the frictional pressure drop calculations for laminar flow in non-circular channels are incorrect when expressed in terms of the hydraulic radius, the pressure drop for the TRIGA reactor must be predicted by other means. The method selected was to convert the free-flow area into an annulus around the fuel element. With an annular space of inner diameter D1 and outer diameter D , the frictional pressure drop becomes l 2 6P (Friction) = . - . 2 2 2 2 g Dv + Dj - (D2-D1) Ln (D2/D1 ) where u = Viscosity of air, lb/hr-ft v = Velocity of the air, ft/hr L = Length of the fuel element, ft Dj = Fuel element diameter, ft D2 = Dj + 2b, ft. The term b in D2 is the effective separation distance between the B
- 1. W. H. Adams, " Heat Transmission", McGraw-Hill, New York, N. Y.,
1954, p. 149.
i,
\
\
e
* \
> . ring fuel element and'those in the'C ring. The use of the minimum separation distance as b would yield too large a pressure drop and is considered too restrictive. .The use of the average separation distance, based on the free
_P flow between the B ring element and the C ring, would yield too low a pres- , sure drop since the pressure drop is not a linear function of the separat-ion distance.. As an ~ approximation, b is taken as the mean of'the two values, that is [0f=1/2(0.01130+.04930)=0.0303,ft. (5) For the channel between the B ' ring and ' the C ri ng, the equation for the pressure' balance (2) becomes - 9. (po_- pj) = 6.88 x l'0-5 uv, (6) where i is the average viscosity of the air (lb/hr-ft) in the channel and is a function of the entrance and exit temperatures, v is the average air velocity (ft/hr) in the chann.el, arid the entrance and exit densities are po and pj (lb/ft3 ), The mass flow rate of air in the channel is w = vpAc (7) where i is the . average density of air in the channel and Ac is the flow at;ed associated with the channel. Combining equations (6) and (7), one obtains the mass flow rate , w = 405.(po - p1)p/i . (8) Assuming that the average properties in equation (8) are the average of the properties at the entrance and exit, equation (8) beccmes 4- 1p"2 . pI 2) w = 405. (9)
; ( po + p1)
. \
x 3
,, i ,.
\
4 ,
. [h ,
Over the range of Mmperatures of interest the properties of air have been approximated by linear equations. Thus, 3 (10) P= lb/ft 2.5 x 10-2T and
- p =.(0.01135 + 0.6017 x 10-4 T ) lb/hr-ft where T is the temperature in ' Rankine.
Using these expressions .in equation (9) one obtains w= 3.24 x 105 (T + T )' C (TfT) O _ 0.01135 + 0.30085 x 10-4(T1 + To) The determination of the amount of heat removed by this air flowing past the element rests on the evaluation of a heat transfer coefficient. For a free-standing cylinder cooled by natural circulation, the heat trans-fer coefficient is given, conservatively, by kf h = 0.531 W GrPr]0.25 (12) L where kf is the thermal conductivity of air film at temperature Tf (Btu /hr-ft *F), Gr is the Grashof number, and Pr is the Prandtl number. Because the fuel element is surrounded by adjacent elements, the flow will probably not be laminar, even at low Reynolds numbers, and the heat transfer correla-tion should be larger than that assumed, perhaps by as much as a factor of 2. Again, over the temperature range of interest, one can write for the thermal conductivity and specific heat of the air ka T (13)
= (0.0009 + 0.26 x 10-4 ) Btu /hr-ft- F 1
- --- - ,, - . . , - . , _ , - , , , -.- ,,, - , . , , - ,- . -,,ag , , - - , - -----g,, , - , ,
- p. -. .m .
4 and Cp a = 0.240 Btu /lb *F with temperature, - T, in 'R. Using these values and those in equation
.c (10) .in equation (12) one obtains -
h = 50.16 0.0009. + 0.13 x 10-4 (Tw a +[1.25 T )fx310-2 (Tw + Ta l3" (14) [0.5(Tw + Ta )] [0.01135 + 0.30085 x 10-4 (T, + Ta )]) -I 0.25 (Tw-T where Tw is the wall temperature, Ta is the bulk air temperature, and Tf is the average of the two, all in *R. For the purpose of this anlaysis the heat transfer along the graphite end reflectors was neglected, and the fuel element was considered to be a 1.25-ft-long, 0.1225-ft-diameter cylinder of U-ZrH1 with specific heat and thermal conductivity given by: Cpf = (26.3 + 0.0245 T) Btu /ft3- F i and (15) kf = (10.7 - 6.42 x 10 -4T ) Btu /hr-ft 'F where T is the local temperature in *R. The temperature drop in the clad was ignored because, at the time of peak temperature, it has been calculated i. l to be insignificant (1 - 5'C). For the computer program, the fuel element was divided into five radial and five axial regions, and the temperature in each region was computed as a function of time after complete water loss. The maximum fuel temperature found from the program as a function of l t.ime after the loss of cooling water was found to be 152 C. The loss of water in the core was assumed to take place within seconds after shutdown, l and the maximum initial fuel temperature is 113 C. To determine the pre sure exerted on the cladding by released hydro-gen, fission products, and ' air trapped in the fuel can, the conservative L
assumption will be made that the entire system is at peak fuel temperature, i.e., about 152'C. The total number of fission product nuclei released to the gap between the fuel and clad was determined from Blomeke and Todd,l and the results of the experiment described in Section 8.7. The ' total quantity of Br, I, Kr, and Xe released to the gap in the B ring fuel element after continuous opera-tion at 250 kW for about one year will-be Ng = 0.028 x 6.87 x 1020 = 1.93 x 1019 atoms. (16) The number of gram-atoms.in the gap is 1.933 x 1019 a 3.22 x 10-5 gram-atoms (17) nfp = 6.02 x 10 23 The partial _ pressure exerted by the fission products gases is RT Pfp = nfp - (18) where, initially, the volume, V, is taken as the 1/8-inch space between ! the end of the fuel and the reflector end piece. This . is quite conser- [
- vative because the graphite reflector pieces have a porosity of 20% and the fission product gases can expand into the graphite. The initial volume, then, is-3 V O = wr h = w(1.80)2 0.317 cm3 = 3.23 cm ,
where the radius of the fuel material, r, is about 1.80 cm and the space width, h, is 1/8-inch (0.317 cm). Thus the initial pressure exerted by all the fission product gases l 1s 3.216 x 10-5 y = RT = 0.997 x 10-5RT (20) PfP 3.23
- 1. J. O. Blomeke and Mary F. Todd, "U-235 Fission Product Production as a Function of Thermal Neutron Flux, Irradiation Time and Decay Time",
ORNL-2127.
The partial pressure of. the air in the fuel element is RT Pair " 22.4 x 10 3" *
- and the total pressure exerted by the air and fission products is i PP f \ l 0.997 x 10-5 \
Pr = 1+ ~ Pair " l+ Pair = 1.22 Pair (22) Pair 4.46 x 10-3 T V0 Also Pair = 14.7 -- (23) 273 V where P,jp is in lb/in.2 and T is in K. The equilibrium hydrogen pressure over the U-ZrH1 ,0 fuel material is simply a function of the fuel temperature. At a temperature of 152 C, , the hydrogen pressure is negligible (see Figure 5.2.2) l The total gas pressure at the maximum fuel temperature of 152'C is 90.0 P=P H + Pr = 0 + 1.223 (14.7)273 ( 152 +V 273)V 3.23 = ' psi. (24) Ignoring the negligible increase in volume caused by the expansion of the clad the total pressure then is 90.0 90.0 90.0 P= = = = 27.8 psi. (25) V Vo 3.23 The tangential stress in the fuel element cladding when subjected to an internal pressure, P is S = Pr/t, (26)
, u --,-. . - - --,n, ., ----,,-.--n. w- ,,,_------.m, , - - - - - , -,,p.~.- . , , , . . , . . , , , - --- -,,,,,v..--- - , -- - , , , , . . . . - . -
where r is the radius of the fuel _ element can (1.80 cm) and t is the wall thickness (0.076 cm) or S.= 23.7P (27) The stress in the cladding is S = 23.7(27.8) = 660 psi. . (28) From the Aluminu.n. Company of America handbook,I the yield stress for type 6061 aluminum at 150'C is. greater than. 8,000 psi. Consequently it 'is concluded that, subsequent to the complete loss of cooling water, the release of hydrogen from the fuel and the expansion of air and - fission. product gases in the space between fuel and graphite end pieces will net result in the rupture of the fuel element cladding. 8.6.3 P.adiation Levels After Loss of Pool Water Dose rates have_ been calculated for two locations; 1. at the top of the pool, 2. at the floor level of the laboratory over the reactor. Again, the reactor was assumed to have been operating for a long time at power
' levels of 100 kW and 250 kW prior to loss of water.
The dose rates calculated for instantaneous loss of water at shutdown are given in Table 8.6.1 and are clearly hazardous. However, as shown in section 8.6.1, the instantaneous loss of all the water surrounding the core is not conceivable and it is more realistic to estimate the dose rates with gr6 dual water loss. These values are given in Table 8.6.2.
- 1. Aluminum Company of America, "Alcoa Aluminum Handbook," (1957), p. 26.
- - . .. . . - . - . - _ - - - - . _ - _ ~ - , . . - - . - - . - - - - . . , .
ge , ._ _ __ . . _ . . _ . . . . TABLE 8.6.1 CALCULATED RADIATION DOSE RATES AFTER TOTAL LOSS OF REACTOR POOL WATER Time after Direct Radiation Level (r/hr) shutdown (1) (2) Top of Pool Floor of Laboratory 100 kw 250 kw 100 kw 250 kw 0.1 hr 440 1100 4.1 10.2 1.0 hr 236 590 2.3 5.75 10 hr' 112 280 1.1 2.75 1 day 80 200 .69 1.73 1- week 36.4 91 .36 0.90 1 month 6.6 16.5 .065 0.164 TABLE 8.6.2 CALCULATED RADIATION DOSE RATES WITH LOSS OF P0OL WATER AT G.4 METERS PER HOUR Time after. Water Depth Direct Radiation at shutdown over core (m) Floor of Laboratory (mr/hr) 100 kw 250 kw_ 0.1 hr 6.1 10-10 10-10 1.0 hr 5.6 10-10 10-10 5.0 hr 4.1 2.1x10-10 5.2x10-10 10.0 hr 2.1 0.051 0.128 15.2 hr 0 880 2200** 24.0 hr 0 692 1730
**This value could be reduced to less than 1 mr/hr by maintaining the water level at about 6 feet in the tank from the main water line in the laboratory.
More than 5 hourb is provided by the alarm
- before the dose rate approaches the " permissible" 0.24'mr/hr level in the area above the reactor. Such time is ample to clear the area, investigate the damage and initiate remedial action. In addition a rough estimate of the radiation dose at ' the . side of the pool from scattered radiation indicates that an individual who did not expose himself to the direct radiation from the core could work there for at least 12 hours,1 day after the accident, without receiving a dose in excess of that permitted by 10 CFR 20.101 for a calendar quarter. This time would be sufficient to view the interior of the pool with a mirror' and to continue emergency repairs.
Method of Calculation The following assumptions were made:
- 1. The core had been operating at power for a very long time (1000 hr) prior to. shutdown.
- 2. The reactor is shutdown as soon as the leak develops.
- 3. The core can be approximated as a point source with effective strength determined by the photons escaping from an isotropic volume-equivalent spherical source.
- 4. The photon energy can be averaged to 1.0 Mev and only photons from fission product decay are important.
- 5. Attenuation by core components other than fuel elements is negligible.
! Assumptions 1. and 5. are conservative, and 4. is-optimistic. l l The net effect is conservative. i I l Effective Source Strength The fission product decay photon source strength is given by Perkins i and Kingl and the value of the effective source strength is assumed to be
~
t I all at 1.0 Mev. i l i
- 1. J. F. Perkins and R. W. King, Nucl. Sci. Engin., 3, 726 (1958).
- Activated by float switch
- ~, __ _ . _ _ _ _.. -__ _.. . _ _ . _ _ _ _ . _ _ _ - . _ _ _ _ . . _ _ _ . . _ _ _
The source strengths obtained are given in Table 8.6.3. Each. fuel element has a volume of 400 '.m3, thus a 67 element core has
- a. volume of 2.68 x'l04 cm3 which is equivalent to a solid sphere of 18.4 cm radius. -The fraction of photons which escape from the core is givenl as a function of Ucrc, and is 0.197, where pc is the attenuation coefficient of the core material for 1 Mev photons (= 0.207 cm-l; GA-5400, p 8-53) and rc is'the radius of the sphere, 18.4 cm.
The product of the source strength and the fraction escaped is the effective source strength of the core. These values are listed in Table 8.6.3. TABLE 8.6.3 EFFECTIVE SOURCE STRENGTH FOR SPHERICAL REACTOR CORE OPERATED FOR 1000 HOURS AT 100 KW Time after Source Strength Effective shutdown Mev/sec-100.kw Source Strength 0.1 hr 6.49 x 1015 1.28 x 1015 0.5 hr 4.52 x 1015 8.88 x 1014 1.0 hr 3.60 x 1015 7.08 x 1014 5.0 hr 2.20 x 1015 4.32 x 1014 10.0 hr 1.70 x 1015 3.36 x 1014 24.0 hr 1.20 x 1015 2.36 x 1014 1 week 5.48 x 1014 1.08 x 1014 1 month 1.00 x 1014 1.96 x 1013 1 ( i l l l
- 1. K. K. Aglintsev, Applied Dosimetry, (Eng. Ed.), Iliffe, London (1965).
l {
- - ~ ~ , - . . - - , . - -- .
6 Dose Rates-Direct Radiation The dose rate is computed on the basis that the core is a point source, 21 feet below the top of the pool: Seff - D= B(Ru)e ux 43R 2 g where D = the dose rate in rads /hr (= rem /hr for 1 Mev y) Serf = effective source strength (col. 3., Table 8.6.3) R = distance between core and location considered. K = conversion factor for Mev/cm2 to rad /hr ( = 5.77 x 105 for 1 Mev photons) B= build-up factor for medium involved. B is a function of Ru for any material, so will vary with R. Values for water were taken from ref. 1, those for concrete from ref. 2. u= linear attenuation coefficient, cm-l ref. I gives for 1 Mev photons u concrete = 0.149 cm u water = 0.067 cm x= thickness of material, cm. (10 'cm for concrete ceiling of reactor room). Attenuation by air was neglected. 8.7 Fuel Cladding Failure The effect of a fuel element rupture resulting in a release of fission product s from a B ring fuel element has been evaluated using a recently determined experiment value for the fission product release fraction. Also considered is the incredible situation in which cladding rupture occurs following the loss of pool water. This accident is considered incredible as it entails two essentially independent accidents occurring simultaneously. l l l l
- 1. H. O. Wyckoff, " Radiation Attenuation Data", in Radiation Hygine Handbook, H. Blatz, ed. McGraw-Hill, New York 1959. Chapter 8.
- 2. 'leactor Handbook, VoE III, Part B " Shielding", (1962).
l
8.7.1 Fission Product Inventory The inventory of radioactive fission product gases, halogens and noble gases, in the B ring fuel element in which the average power density is 1.6 times greater than the core average, was. determined ' by assuming long term operation at 100 kw yield with 70 elements in the core. Using the data developed by Blomeke and Toddl results in the compilation of the vola-tile fission products given in Table 8.7.1. For the purpose of these analysis, the gaseous fission products have been divided into two groups. The first comprises those isotopes that will remain in the tank water should the rupture occur when the pool is filled with water. -In this group are the bromines and iodines. In the incredible event that no water is present, these isotopes would be added to the radio-active cloud and add to the hazard. The second group comprises the insoluble volatiles, the krypton and i xenon isotopes. They are the major scarce of radioactivity in the room - (and outside) if the pool water is present. 8.7.2 Fission Product Release Fraction The steady state release of short-lived fission gases from specimens of TRIGA-type fuels has been studied at temperatures ranging from ambient to 400'C. The specimens (0.33 inches in diameter, 0.33 inches long) were cylindrical in shape and of the TRIGA fuel composition, U-ZrH l .7-
- 1. J. O. Blomeke and Mary F. Todd, "U-235 Fission Product
, - - - , , . . - - s -,, . _ , , . _ , , y, , - . ---m-. , _, - - ., --,,,,,-.,~-.,,,,ry. -
TABLE 8.7.1 GASEOUS FISSION PRODUCTS IN ' B-RING ' FUEL ELEMENT 100 - kw SATURATED ACTIVITY--70 ELEMENT CORE Nuclide DecayConstant(hr-ll . Inventory (curies /elem.) Group I Br 83 3.02 x 10-l 9.84 Br 84 1.31 x 100 22.6 Br 84m 6.95 x 100 .465 Br 85 1.39 x 101 30.6 Br 87 4.49 x 101 55.0 I'129 4.60 x 10-12 20.5 I 131 3.58 x 10-3 59.4 I 132 3.07 x 10-1 90.0 I 133 3.34 x 10-2 133.0 I 134 7.93 x 10-1 155.7 I 135 1.04 x 10-1 121.0 I 136' 2.90 x 101 63.6 Total Iodines 644. Total Group I 762 Group II Kr 83m 3.66 x 10-1 9.84 Kr 85m 1.59 x 10-1 30.6
. Kr 85 7.67 x 10-6 6.15 Kr 87 5.35 x 10-l 55.2 Kr 88 2.50 x 10-I 76.1 Kr 89 1.31 x 101 94.5 Kr 90 7.55 x 101 106.0 Kr 91 2.54 x 102 63.3 Xe 131m 2.41 x 10-3 0.594 Xe 133m 1.26 x 10-2 3.20 Xe 133 5.50 x 10-3 133.0 Xe 135m 2.67 x 100 36.4 Xe 135 7.60 x 10-2 98.1 Xe 137 1.07 x 101 12.1 Xe 138 2.45 x-100 112.4
' Xe 139 6.08 x 101 116.3 Xe 140 1.56 x 102 122.0 Total Group 11 '077 b.m_
The TRIGA King Furnace used for these studies is essentially a minature in-core-loop and is described in detail in several reports. (See the HTRDA Semi-annual report for July to December, 1964, pp 9-11, and Gulf General Atomic report GA-8597 by F. C. Foushee). The essential feature of the furnance is that it can be inserted into a fuel element position of the TRIGA reactor, a sample heated to any desired temperature from 100 to 1700*C (present license upper limit) and the gases released during a subsequent irradiation purged to a cold trap. Gases are then analyzed by y-spectrometry and their steady state release fractions determined. Aside from experimental errors and statistical scatter there is no temperature or large half-life effect from room temperature to 300*C. The mechanism of release through this temperature range is principally due to recoil of fission fragments. The runs were all performed on the same sample, and on the same day. As long as release was of a recoil nature the build-up of the longer lived fission gases from run to run did not affect the results. From these measurements it can be concluded that below 300*C the fractional release of fission products is about 1.5 x 10-5, Applying this release fraction to the inventory of gaseous fission products produced in the fuel , as given in Table 8.7.1 gives the total activity that would be released should the integrity of a fuel element cladding be compromised. Thus, the toital rate gas activity released would be 1.62 x 10-2 curies, and the total iodines 9.65 x 10-3 curies, and the total halogens 1.14 x 10-2 curies. , 8.7.3 Water Activity In the fission product release accident in which the pool water l t
i
~ remains in situ, the fraction of th'e total soluble fission products which are released from the element is distributed in the water. Thus 1.5 x 10-5 times the Group I fission products or 1.14 x 10-2 curies remains in the !
pool. Since the volume of water in the reactor pool is 3.57 x 107 cm3, the activity concentration is 3.20 x 10-4 pc/cm3 The activity remains moderately high because of the small decay constants for I-129,1-131, and 1-133. However, the demineralizer can be used to remove the soluble vola-tile fission products. 8.7.4 Exposure Inside the Reactor Room - The maximum exposure to a person in the reactor room will occur if the fission products from a ruptured element are distributed instantaneously within the room and no air change occurs. The total release of noble gases is 1.62 x 10-2 curies and the volume of the reactor room is 5.7 x 10 8cm3 so that the concentration is 0.284 x 10-4 microcuries/cm3, or 1.04 dis /sec-cm3 If the room is assumed to be equivalent to a hemisphere of radius R = 643 cm, the dose rate at the center is given by S D= 2Ku (1 - e -UR)
- S = source strength, dis /sec-cm3
. K = conversion factor for flux--dose rate
= 4.2 x 10 4y/cm2-sec per mr/ min for photons of 0.7 Mev (average energy of f.p. decay) p = attenuation coefficient for air
= 3.5 x 10-5 cm-l for 0.7 Mev y j .
i Thus the maximum dose rate if 0.8 mr/ min. If the halogens were to { be released as well, this would te increased to about 1.36 mr/ min. Thus l' b._
-- . -z . -- j an individual could remain in the room for about 75 min before exceeding the dose of 100 mr. The controlling factor, however will be the dose to the thyroid from halogen activity in the second case. In either event the dose rates are sufficient to set off the radiation alarms in the area so that the room will be cleared rapidly. In addition it is probable that the emergency purge exhause (see 4.6.2)
) 3 would be operating in such an accident, venting air at the rate of 150 cfm.
This will decrease the o'ose rate to a person in the reactor room. The dose to the thyroid can 'be calculated at any time (t) froml 2 (5.92 x 10 ) A f E ( 1 - e-A t) mx .. p where A = inhaled iodine, in curies f = fraction which is deposited in critical organ E = effective energy absorbed by thyroid per disintegration (Mev) A = effective decay constant, comprising both radioactive decay and biological elimination m = mass of the thyroid in grams in a long time, t = , and using T (=0.693/A ), the equation becomes: (8.54 x 102 A f E T) m
= constant x Aj Ej T j i
where the parameters are for the ith isotope.
- 1. J. J. DiNunno et al., " Calculation of Distance Factors for Power and Test Reactor Sites", TID-14844, 1962.
The inhaled iodine is given by fAjI Aj = R (V / Aj where-is the activity concentration, and R the breathing rate for a y time T. If x is made equal to 1 sec. then (8.5A ,10)fR 2 A1 Ej Tj . M i y For the standard manl, f = 0.23 m = 20 grams R = 10 m3/8 hr
= 3.47 x 10-4 m3 sec-I and the value of the constant is thus 3.41 x 10-3, The necessary activity concentrations are found from Table 8.7.1 using 1.5 x 10-5 for the release fraction and summation for I-131 through I-136 yields a value for D = 0.45 rads /sec.
Assuming that 300 rads to the thyroid is a limiting exposure,I a person will have approximately 37 minutes to evacuate from the reactor room. In fact the time will be longer than this since the iodine isotopes will plate out in the reactor room and the room exhaust will be in opera-tion and will reduce the dose. f
- 1. " Report of ICRP Committee 11 on Permissible Dose for Internal Radia-tion (1959)," Health Physics 3 June 1960.
'8.7.5- Exposure Outside the Building l' _
. On release of radioactive' fission products into the reactor room.
the continuous air monitor will close the normal exhaust and start the emergency purge. The i room air will then be exhausted through a filter at 150 cfm (0.071.m3/ sec) which will not remove the noble gases and'may not remove all of the iodines released in the event of pool water loss. At'the west penthouse roof level, on the Merrill . Engineering Building where the emergency purge exhausts, there are also many other exhaust out-lets, excluding those from the reactor area. Those in the insnediate vici-nity of the reactor exhaust have an exhaust rate totalling 18,500 cfm are designed to operate 24 hrs a day. Assuming this condition prevails, it is reasonable to assume a dilution of the exhaust effluent by a factor of 70 at the very' least. Using this, and assuming no further dilution or decay occurs outside the building the maximum dose to the thyroid may be calcu - lated. The calculation assumes: (1)' Complete mixing in the reactor room at all times (2) The person is immersed in the effluent from the building . in which the concentration of radioactivity X at any time t is equal-to I Ct /70 where Ct is the concentration in the exhaust and in the room. With (1) Ct is given by a simple first order rate law: L _
, Ct=C0 eXP --t I where a = exhaust rate in m3 /sec V = room volume in m3
- j. Co = initial concentration, at t = o
- 1. J. J. DiNunno et al., loc. cit.
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the differential inhaled activity is dAt= R X ~ dt an'd the total amount inhaled up to time t is RC . t a
'At" exp [ y't ] dt -
70 o
.RC V a
= 0 [ l - exp (
^ 70a V t) ] If the exposure is for an infinite time, ,
-A ._= RCoV/a(70) and, comparing 'withthe previous calculation, the total dose f
D.(rads /sec.) V(m3) int 0 " = 3 70 a (m /sec)
= 26 rads If the person were to remain in the effluent cloud for only one hour, .
the corresponding figure is about 15 rads. Thus we may conclude that for any location other than on the roof of the building, the maximum possible dose to the thyroid from a fission product release will be considerably less than 1/20th of a maximum permissible dose, even when no credit is taken for radioactive decay. Culf Energy and Environmental Systems, Inc. has developed computer programs which take allowance both for decay and for further meterological dilution.1 Calculations using these programs for a similar situation to that presented here,2 indicate that these factors will reduce the maximum-
- 1. GA-6511. " GAD 95E and GADOSET--Programs to Calculate Environmental Con-sequences of Radioactivity Release". GAMD-5939, "PHICHK and PHIBAR pro-grams to Calculate Atmospheric Dispersion".
- 2. Addendum to GA-6499, " Hazards Report for the Oregon State University 250 kw TRIGA Mark II reactor--dated June 23, 1965", April 20, 1966.
__ _ _ -._. _ _ _ .. _ _ . _ .._ ._ . _ _ . . _ _ . _ _ _ _ _ . _ . _ . _ _ _ , . . . . . ~ . . _
[. i_ . . _ _ . _ _ - . - . . _ _ dose to a person a short distance away from the building by several orders of magnitude.. Thus for a slow leak rate from the room (a = 1.69 x 10-2 m3 /sec) and very stable atmospheric conditions (see 8.4.4) they find: Whole Body y Thyroid Outside building Max dose (12 hours) 2.9 x 10-3 mr 2.6 mr Inside reactor area Max dose (1 hour) 6.0 x 10-1 mr 7.0 r These exposure values are negligible except for an individual in the reactor area. Even there the exposure is not unduely high for such a highly improbable accident. 9 l
w s. . , ever, the release of fission products to the tank water from the complete cladding failure of one B ring fuel element in water at 100kN satur-ated activity (note the reactor core should be shut down during crane operation. (See Table 8.7.1 and SAR, Section 8.7.3) is given by the relea:,e fraction of 1.5 x 10-5 times the concentrations of the respec-tive group of fission products. We find that the rate gas release is 1.62 x 10-2 curies, '9.65 x 10-3 curies for the iodines and 1.14 x 10-2 curies for halogens. The volume - of primary water is 3.6 x 10 7 cm3 so activities are respectively: 4.5 x 10-4 pCi/cm3 for rate gases, 2.7 x 10-4 uC1/cm3 for iodines and 3.2 x 10-4 pC1/cm3 for halo-gens. Although these values exceed 10 CFR 20 limits they pose no significant radiation exposure risk to laboratory workers and can be readily removed with the water purification system. The iodines and halogens would remain fixed in the primary water but it is possible that the rare gases could move into the reactor room air. Assuming an instaneous, complete loss of all rare gas fission products to the room air (5.7 x 108 cm3) results in an activity of 2.8 x 10-D pC1/cm. The dose due to this release is found to be (see SAR, pg. 8-44) 0.88 mr/ min neglecting the reduction that would occur from the ventillation system. Furthermore this level would trip the radiation alarm and the worker would have 75 minutes to exit the area before receiving a dose of 100 mr.
- 25. Confirm that you have no graphite loaded reflector elements.
There are no graphite loaded reflector elements in our present inventory.
- 26. Provide a schematic drawing of the primary coolant circulation loop and the primary coolant purification system. What are the coolant flow rates through the circulation loop and the purification loop?
Average primary coolant flow through the main circulation loop is 50 gal per minute and 4 gal per minute for the purification loop. See figure for schematic.
- 27. What is the concentration of tritium in the primary reactor coolant and the D2 0 in the reflector tanks and elements? What is the total j volume of D 2 0 now in the reflectors used in the reactor?
The tritium concentration in the D 2 0 reflector tanks in the reactor is 10.04 pCi/ml over the volume of about 60 gallons of D2 0 in the core. This represents a total inventory of 30 mci which if all released would
' result in an activity in the tank water of about 0.8 uC1/ml which is about 25% of the release limit (3.0uci/ml) in 10 CFR 20, appendix 3, Table 2, Column 2.
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- 28. Describe the current use of a cooling fan to assist evaporative cooling of the reactor pool water. Wat is the status of the proposed heat exchanger mentioned in your SAR?
A Cooling fan is employed to provide increased forced air connective cooling for the top of the reactor water tank in addition to the reactor room ventilation system. A heat exchanger system is under development to remove heat frcm core as optionally provided in the SAR. The freon compressor unit is being installed and the heat exchanger has been order. 3 i All activities on the heat exchanger syste.n are under constant review s by the Reactor Safety Committee.
- 29. What are the normal evaporation losses from the pool? Describe the makeup water system at the UUIR.
Normal evaporation losses (depending upon the mean tank water and ambient air temperature) are about 3 gal per day without the cooling fan and about 26 gal per day with the fan. Makeup water is added by opening makeup water lines and valving off any back-flow paths to force all makeup water through the demineralizer (seefigure), pre-filterandpost-filtersystems.
- 30. What is the volume of the ion exchangers used in the water purification system? How is the ion exchange material of this system regenerated?
What are the procedures for disposing the liquid and solid wastes from the regeneration and/or replacement of the ion exchange resins? The volume of the mixed resin used to purify the primary coolant water is 2.5 cubic feet. The ion resin is regenerated by a local firm under contract using a standard acid-base reverse wash. The resin is not released to the firm until activity levels when properly diluted are with 10 CFR20 restrictions for release. This surveilence is performed by the Radiation Safety Officer before release.
- 31. How often are the pH and conductivity of the pool water monitored?
What are the procedures used to maintain them within the Technical Specifications limits? The pH and conductivity of the primary water are monitored at the position shown in the enclosed figure. There are no Ifmits on pH or conductivity under the present Technical Specifications. However, the prop 0}ed, new tech specs will require a primary coolant pH between
< 5 x 10-0 mhs. This will be accomplished by resin bed regeneration andacid(HNO)bufferingasrequired.
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- 32. Describe the heating and cooling devices integrated with the ventilation system in the reactor area.
Reactor area heating and air conditioning is accomplished by the build-ing's general heating and cooling system. However, exhaust air from the reactor area is vented through a independent exhaust system which is continuously monitored by a radiation detector.
- 33. Provide a schematic drawing of the ventilation system for the reactor area, including the emergency filter arrangement and radiation monitors.
See attached figure.
- 34. Describe the system that supplies compressed air to the reacter area.
Compressed air is delivered to the reactor area from the building's air compressor and holding tank system. The air is filtered and dried before delivering to users. s
- 35. Describe the fire protection system at the reactor facility. Which fire department serves the fire-related emergencies at the reactor facility? 20 is reponsible for maintaining fire protection equipment at the reactor facility.
The building is fully outfitted with overhead sprinklers and near the reactor area is a fire-water line, chemical extinguishers and also in-stalled are smoke alarms in the reactor room. The University of Utah has a fully equipped fire station on campus which would respond to fire-related emergencies in the reactor area. The University fire marshall is responsible for surveying and maintaining all fire equipment on campus
- 36. Describe the central irradiation facility of the reactor and the different types of irradiators now in use at this facility. Provide the dimensions and schematic drawings of these facilities.
^ There are three different irradiators used in the TRIGA core central irradiator location. Attached drawings provide dimensioned drawings of these irradiators, Two are filled with 0 0 2 (one has a 7 hole con-centric row for neutron sample irradiation and the other has a single
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l central hole surrounded by2 D 0) and a third irradiator is air filled to provide a fast neutron spectrum for the sample. The reactivity difference between the D0 2 filled irradiator and the air filled irradiator has been measured at $0.62. For HO 2 in the irratiator this reactivity difference is less than $0.15.
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.37. Describe the wet and dry irradiation facilities at the trapezoidal
.D2 0 reflector tanks. Provide their disensions.
There are 6 trapezoidal. D2 0 filled reflector tanks each 6 inches-wide surrounding the-TRIGA core. Four of these tanks are completely D2 0 filled with no irradiation fixtures installed. The other two tanks do permit insertion of samples for neutron irradiation. One tank has a large cylindrical cavity which will accomodate any one of the irradia-tors ~ described in question. 1he other tank has three, 1 inch diameter openings which-can be voided of water and prepared to provide a dry irradiation space. Existing procedures and monitoring activities con-trol induced airborne activity that arises for such irradiations.
- 38. Wat are the volumes and Curie-contents of the solid and liquid wastes generated annually at the reactor?
The average annual volume and C1 content of solid and liquid waste is solids: 1 cubic foot at 10 9C1 liquids: 100 ml at 1 pCi typically these materials are held for decay, diluted and discharged or packaged and sent out for approved disposal.
- 39. Wat is the dose rate at the pool surface attributable to 16N and 14A r-when the reactor is operated continuously at the highest power level?
The radiation field at the top of the tank attributal to N-16 and Ar-41 is 0.1D mrad /hr. This value is in excess of the natural background value in the area of 0.012 mead /hr.
- 40. Wat is the estimated annual release of 41A r from the reactor facility to the environment.
The air monitoring system can detect down to a limit of 1/3 of Ar-41 limits found in 10 CFR20 APP. 8 TA8., !!. Since no Ar-41 has teen de-tected, the maximum Ar-41 release cannot exceed a value calculated from 1/3 of the table value. f if if if 1[ \ 4x10-8 901 1800 ft3 5.26x105 min 2.83x104 al 3.7x104 dis. 3 ml j min; yr j ft3 ;( pCi-sec j f 41; rams ar-41/ mole = /8.6n grams Ar-41' [6.59x103 soci X year ( . In z j (6.cz3x10" atoms /molej g j
- The maximum yearly release would then be 8.6x10-9 grams of Ar-41.
However, since the facility only produces Ar-41 while the reactor is running, a more realistic calculation follows. 8.6x10-9g Ar-41 I 6 hrs operation = 3.1 x10-10a-Ar-41 year 168 hrs year , k /( / Annual Release of Ar-41 would be less than 3.1x10-10 , g t
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- 41. Describe the administrative organization of the radiation protection program, including the authority and responsibility of each position identified.
The radiation protection program at the University of Utah is admin-istered by the Radiological Health Office with a Radiation Safety Offi-cer. The attached figure provides the administrative organization structure. The Institutional Council of the University of Utah is the licensee for the TRIGA reactor to the NRC. However, the Reactor Admini-strator is the primary point of contact with the NRC in behalf of the Institutional Council. The reactor Safety Committee and Radiological Health Committee serve as review and approval bodies for all activities which are under the primary control of the Reactor Supervisor. The Reactor Supervisor is responsible for all reactor operations and opera-tors in the Nuclear Engineering Laboratory.
- 42. Describe the responsibilities of the Radiation Safety Office staff at the reactor facility. Identify the radiation safety related tasks that are performed routinely by the reactor staff.
The Radiation Safety Office under direction and review of the Radiolog-ical Health Committee is responsible for all radiation safety activit-ies associated with the TRIGA reactor. The reactor staff provides routine monitoring functions and . personnel dosimetry coverage for personnel and activities at the reactor facility. However, all radiat-ion safety activities are subject to review by the Radiation Safety Officer.
- 43. Describe any radiation protection training for the non-Health Physics staff. If possible, provide a topic outline of the courses and indicate the normal duration of eacn course or lecture.
t The Radiation Safety Office provides periodic training in radiation I technology for all campus users. Reactor staff and students who need training and review of radiation related matters are requested by the Reactor Supervisor to attend lectures and laboratory work. These cour-ses include instruction on radiation physics, measurements, regulat-ions, procedures and practice; and generally require about 10 contact hours of instruction. I e
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- 44. Describe your program to ensure that personnel radiation exposure and
, releases of radioactive material are maintained at a level that is "as low as reasonably achievable" (ALARA). Identify steps taken to imple-ment the ALARA principle.
The Radiological Health Committee exercise's' review authority and seeks to insure. Compliance for all federal and state regulations. This comittee reviews and identify needs and actions appropriate to ALARA activities. See the Radiation Safety Manual supplied to the NRC review team for specifics.
- 45. De' scribe the gaseous effluent monitoring equipment with respect to
, 'ocation, stack flow rate, and probe geometry. For the fixed-position radiation and effluent monitors, specify the generic types of detectors
'3nad their efficiencies and operable ranges, and describe the methods and frequency of instrument calibrations and routine operational checks.
The gaseous effluent form the reactor is exhausted through a fixed volume, radiation monitoring chamber which is equipped with a calibrat-ed, G-M tube. A continuous strip chart recorder provides a permenant record of activity released through the dedicated exhaust to the roof
-of the building. The area monitoring ranges between 10-2 to 10-3 mr/hr.
The continuous air monitor. ranges from 100 to 100,000 counts per minute and monitors a control volume of 250 liters and operates at 0.1% ' I efficiency (i.e., the detector measures 0.1% of the total gamma activity arising from a uniformly distributed gamma source, Ar-41, in the control volume). Routine operational checks are perfor checkout procedures. Calibration is performed once'a month during the Monthly Checks Proce-dure.
- 46. For the radf ation monitors that are alarmed, specify the alarm set points and indicate the required staff response to each alarm.
.s The area radiation monitors consist of 1) the above core tank monitor,
- 2) the primary water monitor and 3) the exhaust air monitor. These-monitors have low level set points of approximately 0.1 mrem /hr and a high level set point of 1.0 mrem /hr. A low level alarm implies an elevated radiation field in the vicinity of the monitor and corrective r action is required. A high level alarm results is a scram of the reactor and further reactor operation is prohibited until the high radiation source is identified and correctea'.
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- 47. Identify the generic type, number, and operable range of each of the portable health physics instruments routinely.available at the reactor installation. Specify the methods and frequency of calibration.
Portable health physics equipment inventory, is
- Gamma Scintillation Detector range: 0 to 3 mR/hr sensativity: 1 micro R/hr
- Geiger Counter range: 0 - 2 R/hr sensativity: 0.01 mR/hr
- Neutron Bonner Sphere Texas Instruments range: 0 - 1 rem /hr sensativity: 0.1 mem/hr
- Self reading pocket dosimeters various manufac range: 0 - 200 mrem sensativity: 5 mrem all health physics instruments are calibrated semiannually..
- 48. Describe your personnel monitoring program.
The personnel monitoring program at the nuclear reactor facility includes the following respects: a) Familiarization with the' facility, the emergency procedures l and personnel monitoring program. b) Issurance of a TLA dosimeter which is changed monthly, creation of a personal radiation' history and record. c) Issurance of self reading dosimeters to personnel during reactor ' operations. d) Periodic urinanalysis of personnel involved with handling of isotopes that pose internal hazards, e) Monthly issuance of individual radiation exposure records and review of above minimum exposures for ALARA compliance. l l l
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- 49. Describe your environmental monitoring program and summarize the results for the past 5 years.
See enclosed environmental report provided as part of the applicants relicensing application. The applicant also has a complete environmental radiation monitoring
-station installed outside the reactor laboratory. This station includes:
a) Moble gas monitor (eg. Kr. Xe etc) b) Tritium sampler
- c) Air borne particulate monitor d)' Activated charcoal sampler (eg.1-13) e) Pressurized Ion Chamber for gross gamma survey f) TLD's for gamma field dose integration o
- 50. Coment on the ability of reactor components and systems to continue to operate safely and withstand prolonged use over the term of the re-quested license renewal. Include the potential effects of aging on fuel elements, instrumentation, and safety systems. L~ The reactor components and systems have proven reliable in general' l over eight years of operation. Those: systems and components that have experience undue wear or faculty performance have been modified ore replaced as necessary to provide safe, reliable operation. The radiation monitoring system has been upgraded and modernized as necess-ary. ;The present. reactor console is being upgraded with modern digital ! electronic subsystems as necessary and after review and approval of-the Reactor Safety Committec. The incore neutron detectors are eight years of age or less and have operated reliably and safely without any malfunctions. Rack and pinion control rod drives have been installed L and are being tested for acceptance and operation. With the installa-tion of an IBM series 1 Digital Computer,. more accurate, frequent and extended monitoring of operational and safety related parameters and variables will provide enhanced training and improved control of the reactor. The University of Utah TRIGA serves as a. major training resource for the nuc. lear program and it is the intention that students be subjected and trained in the most current and safety concious techn- ' ology we can provide. It is anticipated that over the renewal license period the reactor system will continue to be expanded and improved to meet both technical and education demands and NRC requirements. Fuel l resources and their integrity are deemed adequate to satisfy core fueling requirements for the foreseeable' future. l- . I L
- 51. What_ is the maximum rate capacity of the ventilation system airflow and what is the flowrate when the HEPA filter is in line?
l See the attached figure for the ventilation system under question 33 in , the reactor facility. The maximum airflow through the system is 1740 cfm with the HEPA filter off line and 150 cfm with the filter on line.
- 52. Provide a copy of the proposed Technical Specifications with a maximum operation of 100 kW. Provide a more applicable basis than you have for the maximum safe operating power of Al-clad, low-hydride UZrH fuel.
A copy of the proposed Technical Specifications is enclosed. The TRIGA Reactor at the University of Utah operates in a steady state mode only and pulsing operation is not permitted under the present license or proposed license renewal . The proposed Technical Specifications (3.2(4) and (5)) restrict excess reactivity worth under any assumed operating condition to $3.00 or less so this is assumed to be largest prompt re-activity insertion that could conceiveably occur. Furthermore, the maximum fuel temperature that has been monitored with the instrumented fuel element at 100 kW power is 112 C. The conservative assumption made for low hydride, aluminum clad fuel so as to prevent fuel distortion and possible cladding failure is that the fuel should not exceed the phase change temperature in zirconium hydride from a + 6 + y to a + 6 Phase. This phase change is known to not occur at a temperature lower than 530'C (reference GA-7882, Kinetic Behavior of TRIGA - Reactors, March 1967; K.E. Moore and W.A. Young, " Phase Relationships at High Hydrogen Content in SNAP Fuel System," USAEC Report NAA-SR-12587, Atomics International,1968; K.E. Moore and M.M. Nakata, " Phase Relat-ionships in Alpha-plus Delta Region of the Zr-H System", USAEC Report AI-AEC-12703, AI, Sept.1968.) Futhermore at achieving this tempera-ture in the fuel the f raction of release of fission products is given by f = 1.5x10-5 + 3.6x103 exp [-1.34x104 ] T l where T is mean fuel temp in K above 530 C or 803 K t Reference is Baldwin, N.C., Foushee and Greenwood, " Fission Product (' Release from TRIGA- Low Enriched Uranium Fuels" GA TOC-12, 7th Biennial U.S. TRIGA User's Conference, March, 1980. l The University of Utah's SAR details an analysis of the fuel temperature i reached in a LEU TRIGA fuel element- and shows the core averaged fuel i temperature at the end of a $3.00 pulse to be 297 C. From experimental l data obtained in the University of Utah TRIGA reactor, the hottest possi-l ble fuel element in the B-ring of the minimum fuel core (which leads l to the highest peaking of power in the core) is 1.45 so ( T(max fuel temp) = 1.45 x 297 C = 431 C l l T 1 w *e-r w w-W-*'e+r *Cy-> em4eut--4-p-T - wd-T---Fe1 -+h er 5-r ae r- as--- Ee e- v wme,rP-wpe-'9'---ev-'- eww-w---
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This anlaysis neglects any transfer of heat from the. fuel to the primary coolant during the pulse and assumes an instaneous step change in reactivity of $3.00. Futhermore, the heat capacity of the TRIGA fuel is not constant, but increases with temperature and has been shown to be at 400*C about 1.8 times it value at O'C. The assumption of a constant specific heat overestimates the core fuel temperatures by as much as 10%.
-(Reference GA-7882, Kinetic Behavior . of TRIGA Reactors, March .1976.)
Furthermore, experimental measurements from a thermocouple mounted near the center of the fuel for $3.00 step insertion yielded a fuel tempera-ture of 405*C (reference GA-7882). Furthermore, operational experience gained over many years with aluminum clad, low hydride TRIGA fuel ele-ments demonstrates the high degree of safety and integrity which such fuel exhibits particularly in a non pulsing facility such as the UUTR. In the course of experimental program carried out on the prototype TRIGA Mark I reactor at General Atomic, three cladding failures have occurred. The first took place in August,1959, in a special thermo-couple-equipped fuel element. It is believed that vaporization of volatile material in the insulation of the thermocouple wire created a local region of high pressure between the fuel rod and the cladding material and caused the failure. In November,1959, a cladding failure was detected in an aluminum clad fuel element. When the failure occurred, this element had been subject-ed to a total of 709 pulses, 410 of which were at or above 3 dollars. A subsequent investigation indicated that the failure was caused by ratcheting of the aluminum cladding on the fuel body. Further research > indicated that this ratcheting, as evidence by progressive lengthening of the element during a series of pulses. The second and third failure (May,1960) were both caused by ratcheting of the cladding material. In all three cases the release of radioactiv-ity to the surrounding atmosphere was confined to essentially insignifi-cant quantities of short-lived noble-gasses. It seems very conserva-tive to assume that for the small inventory of low hydride-aluminum clad fuel (17 elements in the present core all of which are in the F or G ring only) compared to the 70 stainless cladhigh hydride fuel (in the C-D-E and F rings) that a highly improbable single pulse of reactivity even up to $3.00 would lead to any major damage of the aluminum cladding that would result in a significant relase of radio-activity. Futhermore, the complete cladding failure of a single fuel element has been previously analyzed in question (24) and the conse-quences we deemed not to pose undue hazards to the staff or public.
- 53. What is the effect of radiation damage and aging on the electrical lines that lead'to the control rod magnets?
The electrical power lines that provide power to the electromagnets for control rod drive are at a minimum distance of 4 feet above the top of the core. At this position the neutron flux and gamma dose are estimated to be 6 x 103 nts)cm2 sec and 250 rad /sec respectively. The integrated neutron flux and gamma dose over 20 years of future operation
- m. -
,,?.. ..J. a .W .-n
- < - - e at anticipated operating history of 10 hrs per' week in 1.8 x 108 rads.
This dose is far below the~ damage threshold for metals such as copper
-and aluminum and less than 1% of the threshold damage- to synthetic
' materials used as insulation for. electrical cable. Futhermore, failure in any of the electrical power cables, would simply result in a reduction or loss of magnet force at the rod inteface and the rod would drop into the core, shutting down the reactor as a consequence. Thus, power line failure results in a fail safe, shut down of the reactor. Furthermore, a rack and pinion drive for the control rods is being installed and only the control rods and aluminum extensions will be exposed to radiation.
- 54. What is1the capacity of the overhead crane? What are the administrative and physical limitations on its use over the reactor?
- The overhead crane is rated at 4000 pounds and is employed in the facility for moving the fuel storage. Cask and other heavy objects. Because of .the potential risk posed by the risk of dropping a heavy-
. weight into the core with attendent fuel damage, the switch box which powers the crane is locked at all times except when authorized by the reactor supervisor for use. The crane is not to be operated during reactor operations. Futhermore, the heavy, structural beam members which support the control rod drives abovethe reactor core shield the core and are assessed to be capable of withstanding a vertical load dropped from the highest point of the overhead crane carraige. Final-ly, the control rods are securely attached by bolts to the core assemb-
! ly and any object dropped on the core would result in a subcritical core array as a result of water floding and fuel relocation.
- The restriction of no crane operations during reactor operation, and crane operations only under direct supervision of the. reactor super-visor and a normally. locked crane power supply box provide assurance of acceptable low risk potential for the core.
f
- 55. Where applicable, provide updated information and discussion on demo-graphy local industrilization, hydrology, meteorology, etc. Include student population information.
- 56. In your SAR and Technical Specifications, you have referenced relatively old reports. Please use more up to date references where applicable, and especially where significant differences exist.
The SAR and the proposed Technical Specifications reflect recent refere-nce and testing data. t 4 I
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- 1. Introduction I This report supports an application to the United States Atomic Energy Commission (AEC) by the University of. Utah (U/U) at Salt Lake City ,
Utah, for the construction and utilization of a TRIGA* Mark I nuclear re-actor. The reactor utilizes the fuel element the control . console, the I control rods and drives, and other minor components of the TRIGA Mark I 100 kw reactor previously located at the University of Arizona in Tucson, , Arizona. These components were given by the University of Arizona to the University of Utah, who with additional fuel supplied by the AEC proposes , to construct and utilize a modified TRIGA Mark-I 100 kw (thermal) reactor I which employs a water reflector, similar to that of the TRICA Mark III, rather than the standard graphite reflector. Because the reactor fuel is typical of that in most TRIGA's and the proposed system has a well es-tablished operating experience, no new significant research and devel-opment activity has been required in designing this s> ' tem. It has been - possible to use accepted, standard safety analysis techniques with regard to evaluating the characteristics of this facility. The reactor will be administered and operated by the University of Utah for education, research, and irradiation services. However, it is deemed that the proposed facility will benefit all interested parties within the regional area. The proposed reactor will be located in the Nuclear Engineering Laboratory (ground floor) of the existing Merrill Engineering Building where an AGN-201M 5 watt reactor is presently located and has been oper-ated safely for 13 years.
- Trademark renistered in U.S. Patent Office
. Additional fuel elemencs requAred to realize a critical system will be ] j obtained N = N1f Enerny and Environmental Systems Ine-
- 2. The U/U proposes to apply initially for operation at 100 kw and after acquiring operating experience and evaluating the reactor's performance at 100 kw to apply for an ansaandment if sarranted for operation at 250 kw.
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.and it is deemed that there is no danger from gansna radiation to persons in this second floor area.
4.3 Third Floor (Figure 4.1.3) i The third floor above the reactor area is comprised of office and classroom space. Maximum dose rate levels on this floor are deter- , E mined to be 0.004 ar/hr (0.010 ar/hr at 250 kw) which is well below natural i L background levels. c . L 4.4 Reactor Area (Figure 4.1.1 - Rooms 1001 A - G) The reactor area comprises seven rooms (Figure 4.4.1) a super-L visor's office, control room, laboratory classroom, counting laboratory, neutron generator office, radiochemistry laboratory, and the reactor room. {- Entry to the reactor room from inside the building is restricted to one , door in the control room. The reactor room has direct access through a 12 feet wide overhead door to the outside loading area.- This overhead L door and the doors between reactor and control room form the confinement [ enclosure and will be tape sealed for this reason. The reactor room L and the radiochemistry room will be treated as a single smit for vent-ilation and safety-confinement purposes. The walls in the area are of one hour fire resistance standard, - plaster and metal stud construction except the west wall which is an er.- terior reinforced concrete wall. ^ Visibility of the reactor room, laboratory classroom, the super- ~ visor's office, and the control room will be complete because of the large windows to these rooms. Nonporous enamel paint finishes are on all walls F L and the ceiling of the reactor area for ease of clean-up. ~ - 4-6 ['~ L r
. ,_- , - _ , . .. . _ , . -.._,..--__-_-__-,.-m4-_-..____-_-
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- 1. ._
f i l 4.4.1 Reactor Room (42 f t x 23 f t x 20 f t high- Room 1001 E)* c -- .~ ... - .. The reactor core will be situated near the bottom of a water-filled tank located as shown in Figure 5.3.1. The cylindrical tank has I aluminum walls, 5/16 inch thick on the sides and a 1/2 inch thick bottom. The tank and liner will be constructed similar to the design and specif-ications of Gulf Energy and Environmental Systems, Inc. for their Mark I I TRIGA installations. The design is considered to be mere than adequate t to endure earthquake shocks. A detailed examination of the aluminum tank r L will be made and hydrostatic pressure leak tests will be performed with
- [ the tank in place on its foundation pad. )
l ' Three existing fuel storage pits (Figure 4.4.1), each 10 feet deep, l . f with aluminum liners and surrounded by concrete have been approved and h l licensed by AEC and are serviced by an overhead crane. The crane travels on an overhead boom and is accessible to any location in the reactor room. I The crane has a 4000 pound capacity and is locally controlled from a pen- ~ t )
- dant box.
h r 4.4.2 Neutron Generator and Irradiator Facility (11 f t x 17 ft- Room 1001 B) South of the laboratory classroom is the neutron generator and w irradiator room. The valls of this room are 12 inch thick concrete. A 3 feet wide by 6 feet long by 10 feet deep pit is provided for a TMC-211 neutron 4 generator T(d .n)Re which produces 14 Mev neutrons (Figure 4.4.1) . The tritium target assembly is at the bottom of the pit and a pneumatic sample transfer system enables samples to be irradiated. A large movable concrete
- and hydrocarbon shield covers the pit and shields the room from neutrons.
The high voltage transformers and cooling units for the generator are
- An AGN-201M 5 watt Nuclear Reactor (License Number R-25) used for teaching and training purposes is 'also located in the reactor room. However no neu-tronic or hazard coupling is considered credible between the TRIGA and AGN.
All credible hazards associated with the AGN have been previously reviewed by the AEC.
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located in the room. The control console for the generator is located in the laboratory classroom. There is little danger from tritium leakage from the generator. In the event of an accident, the exhaust system is sufficient to prevent any overdose to operating staff. In addition to the f ast neutron generator, the facility will include a Cf-252 neutron irradiator. This Cf facility is fully licensed by the AEC and its potential hazards have been previously considered in a Safety Analysis Report submitted to the AEC. No nuclear coupling or interaction of this irradiator facility with the reactors is possible because of the separation distance (40 f t.) involved. 4.4.3 Supervisor's Office and control Room (16 ft x 14 ft- Room 1001 D) I, The control room will contain the reactor control console and i 1 will have good visual access to the reactor. Area monitor instrumenta-q tion will be located adjacent to the reactor control console. Portable b monitoring equipment will be retained here. The only nonemergency access to the reactor room is through the control room except under special cir-cumstances when the doors to the loading dock will be opened; these doors will not be opened while the reactor is operating. The door between the seals and will be pro-reactor room and the control room will have vided with a clo,ser so that,it will normally seal the confinement area. . 4.4.4 Radiochemistry (10 f t x 20 f t- Room 1001 F & G) Two stainless steel fume hoods are located in this laboratory. The venting for these two hoods has a rated capacity of 1800 cfm and is separate from all other ventilation from the building and exhausts above the penthouse level on the roof of the building. This independent venting system will serve as the exhaust system for both the radiochemistry hoods and the reactor ventilation system. _ _ _
- By-Product Material License No. 43-01864-15 49
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F u It is not intended that moderate or high levels of radioactivity shall be processed without special provision for shielding and waste disposal. The reactor administration will require that due provision be made before au-thorization o'f production of such quantities of activity. 4.4.5 Counting Laboratory (22 ft x 16 ft - Room 1001C) The pneumatic sample transfer system and counting facilities will be located in this room. The pneumatic terminus will be provided with an exhaust system connected to the reactor room exhaust. 4.5 Air Confinement As mentioned previously, the reactor room and radiochemistry laboratory are designed to act as confinement area for possible contaminated air. All penetrations of walls, ceilings and floors by pipes and ducts have been 1 specially sealed. The large doors to the loading dock and the door to the control room have weather seals. The ventilation system is designed to maintain a negative air pressure
- in this area both with respect to the outside and to the adjoining areas during reactor operation.
4.6 Ventilation System 4.6.1 Normal Operation The ventilation system proposed for this area is shown in Figure 4.6.1. A schematic of the control system is given in Figure 4.6.2. In normal operation fresh air is drawn into the building at the roof level cnd af ter appropriate processing (filtering, heating or cooling) is distributed around the building. A local thermostat controls dampers at the entrance to the area. Temperatures and static pressures in the main supply ducts are
*liinimum negative _ air pressure requirements are as follows:
1., 0.15.in. water with inlet vents closed and exhaust vent opened. ,
- 2. 0.10 in._ water with inlet vents closed and, exhaast vent dampered for HEPA ,
filter flow. f
- 3. 0.01inwaterwithinletventsandexhaustventopenel 4-10
I I High Level kj[li7el'ala . - Tri? . Monitor Relay i s r
, Reactor Room Air 2nput (1740 cfm*)
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N Input Air Emergency Damper (n.o.) roof - M exhaust O J Emergency Exhaust J Desper (n.o.) ,
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. \ B . *1800 cfm* Exhaust Reactor
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Room Au . g [ _b - Filters ei : I '
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*Without inlet vent filters in place Figure 4.6.2 ventilation control Schematic .
. 4-12
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I maintained by automatic controls. The complete system is under the control of the Physical Plant Center which is manned 24 hours a day. In the reactor area system, in addition to the control dampers, additional static pressure control and emergy isolation dampers are pro-vided. Th'e first is designed to regulate the supply air to the reactor room so that the negative pressure requirement is met; the second operates in emergency status as described below. Air discharge from the reactor room is accomplished through the laboratory exhaust vents in rooms 1001 F & G. .These exhaust vents will operate at all times during reactor operation and move sufficient air to insure at least four changes of air per hour throughout the reactor room (MEB 1001E). These vents are powered by a fan located on the roof of the building and the discharge is located 10 feet above the penthouse roof. 4.6.2 Emergency Operation In the unlikely event of an accidental release of particulate or gaseous radioactivity into the reactor room air, it is desirable that this be exhaused slowly to the atmosphere to allow dilution, decay, and filtration to reduce activity. Thus an emergency purge exhause, actuated by the air monitoring system, will be employed. Isolation dampers will close off the air supply to the reactor room. The outlet vent will be equipped with a by-pass HEPA filtration system which will be dampered into operation. This will provide a continuous filtered exhaust and maintain the reactor area (Rooms 1001 E, F & G) under negative pressure. The flow under filtering conditions will be such , that a negative pressure of at least 0.1 inches of water will be maintained. 4-13
L r 4.7 Radiation Monitorina System i 4.7.1 Continuous Air Monitor-Operati'on of Emergency System j Experience with the University of Illinois TRIGA reactorI indicates that in the case of the most probable accident, namely, leakage of fission pro-ducts from a fuel element, the release is more rapidly detected by a con'tinu-ous air monitor operating alongside the reactor than by a monitor in the ex-haust tank. Accordingly, for the U/U TRIGA, the emergency purge system will i be actuated by the high level radiation alarm with detectors in the air ex- I haust system and at the water tank. The unit provides audioMsual alert and alarm levels; the latter will be set to actuate the emergency purge system described previously. This unit must be operative at all times and a per-iodic testing procedure will be-a part of the written operating instruc-tions for th'e reactor and the periodic system inspection schedule. 4.7.2 Area Monitor } i An area monitoring system will be installed with control and readout adjacent to the reactor control console. Stations will be pro-vided in four locations.
- 1. Ceiling immediately over reactor core (also " sees" fuel storage pits)
- 2. Reactor tank
- 3. Radiochemistry Laboratory
- 4. Neutron Generator & Irradiator Facility The object of these locations is to maintain a check at the most likely positions of release of radiation to uncontrolled areas. Local warning
- 1. Nuclear Reactor Laboratory, University of Illinois, Report INRL-21, October 24, 1966.
4-14
r L F and readout will be provided in addition to central control. As with the L-continuous air monitor, a program of testing and calibration will be ini-I tiated. The instrument itself provides for detector failure alarm by the u.se of " live zero". [ The area monitoring system will be powered by an independent battery powered supply. 4-15 4
2.2. . 5.0 Reactor 5.1 Reactor Desian The TRIGA reactor to be constructed and installed at the University of Utah is similar to that which operated for 13 years at the University of Arizona.from 1958 to 1971. The core fuel elements are largely identical, having been transferred from Arizona to Utah,Therefore, much'of the reactor description given in the Safety Analysis Report for the original Arizona reactor (Hazard Report for a Training , Research, and Isotope-Production Nuclear Reactor at the University of Arizona, GA. 413 June 1958) is applicable to this reactor. There are, however, a few sig-nificant differences, as follows: . 1. Grid plate and lattice arrangement. The Arizona reactor had a grid plate providing a modified hexagonal-circular lattice characteristic of most of the early TRIGA's. However, the University of Utah has decided to adopt the uniform hexagonal lattice with 1.72 inch. pitch of the TRIGA at E the University of California at Santa Barbara (Figure 5.1).--The upper and lower grid plates have been fabricated in the machine shop at the University of Utah using Gulf Energy and Environmental Sys-tems, Inc. drawings and specifications. The grid plates are f virtually it--tical to those manufactured by Gulf Energy and Environmental Systems, Inc. and installed in the TRIGA at Santa Barbara, California.
- 2. Reflector.
The solid graphite reflector of the Arizona reactor will be replaced by either an all water or a combination ficavy water, gra-phite, water refiedtor. , Up to 47 dummy fuel elements containing gra-phite or heavy water in aluminum cans may be used_in the outer F-ring of _ 5-1---- - - - - - - - - - ~ ~ - ^ ~ '
W e 1ctrico, which ha3 fuel element positi:n3 62 to 91 er fer uno in the G-ring, position numbers 92 to 127, or in other positions in the n > l L core. These dununy elements will be aluminum tubes,1.47 inch 0. D., packed with powdered graphite or heavy water over the core length centered at the active core midplane. The use of the dummy elements l will slightly reduce the required fuel loading of the reactor and will thus enable the reactor core to accomodate various irradiation capsule experim,ents. Furthermore, several dummy sample irradiation elements fabricated from radiation resistant plastic (to minimized induced activity) will contain inserts that can be used for the inser-I tion of irradiation capsules. The d'=way elements so equipped can then be inserted in place of fnal alaments in different regions of the I reactor. The characteristics and operating parameters of this reactor have been calculated and extrapolated using experience and data obtained from existing TRIGA reactors as bench marks in evaluating the calculated data. There are several TRIGA systems with essentially the same core and reflec-tor relationship as the proposed U/U TRIGA so the values presented here are felt to be accurate to within 107.. The hexagonal fuel element array is identical to that being used by the University of California at Santa Barbara, and hence the characteristics of that reactor should be quite similar. Though the Santa Barbara reactor uses stainless steel clad ele-ments vs. the mixed stainless and aluminum clad elements of this reactor, the
~ ' '~
thermal utilizaiton of the two is nominal'ly equivalent due to a slightly different loading of fissile material. Table 5.1.1 gives nuclear parameters for the proposed U/U TRIGA core. I. E .g . University of California at Santa Barbara and Irvine, Michigan State University, University of Arizona, and Reed College. 5-3
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w mM1\.\ TABLE 5.1.1 V L- , TRIGA CORE NUCLEAR PARAMETERS - 1 u Cold, clean critical loading 59 elements 2.1 kg U-235 o Operational loading with used (partly 80 elements a burned) fuel elements both aluminum and stainless steel clad 2,8 kg U-235 4, Prompt neutron lifetime 39 ysec y / Effective delayed neutron fraction 0.0070 Prompt temperature coefficient -1.0 x 10" Akper.*C P Tg Average centerline fuel temperature 1000C L at.100 kw F T , Average weler temperature at 100 kw 27'c (with 160C pool temp-L- ature) w 1 Includes 3 control Ig!g The following are reactor physics parameters as described by L the 2-group equation. These are presented here merely for reference p- since calculations of reactivity effects are currently performed with mul-L tigroup computer codes. t n Neutrons released per absorption in U-235 2.0730 f Thermal utilization C,7241 s L p Resonance escape probability 0.877 C ( Fast fission factor 1.052 ~ 4, Infinite multiplication factor 1.3845 Y Fission neutrons to thermal energy 22.23 cm2 F L u 5-4
4 L [ TABLE 521.3 . L ESTIMATED CONTROL ROD NET WORIHS Worth (%Ak) C ring--regulating, 0 25 in. diam. 0.5 D ring--shim, g in. diam. 2.0 D ring--safety, 0.375 in, diam. 2.0 Total 4.5 The maximum reactivity insertion rate is that associated with a safety rod which can be fully removed from the core in 21/2 minutes I producing a reactivity insertion rate of 0.019% Ak/sec. at mid-travel. The total reactivity worth of the control system is about 4.5% Ak. With 'an estimated core excess reactivity of 2.25%, the shutdown margin with all rods down is approximately 2.2%dk and with the most reactive I . rod stuck out is 0.2% Ak.. All of these values are quoted from the cold, clean (no short-lived fission products) condition.
~ ~ ~
~the reactEvitiy worth of the Tuereliments 'is7ependent on their I position within the core. Table 5,1.4 indicates the average values that are expected in this reactor. (Data from University of California, Santa Bar-bara, Safety Analysis Report) Though these data are for stainless steel fuel elements, the worth of an aluminum clad element differs only slightly because of the lower fuel loading of the la,tter.
TABLE 5.1.4 ESTIMATED FUEL ELEMENT REACTIVITY WORTH COMPARED WITH I . WATER AS A FUNCTION OF POSITION IN CORE Core Position Worth (%Ak) B ring 0.98 C ring 0.87 5-6
,_ _._ __.._ . m
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---- m. . x w-I reactors with much higher operating limits than will apply to this reactor.
This TRIGA system will' not be operated in the pulsed mode, hence, there are only three major operatirit considerations and are major non-operating conditions used to define the reactor design bases:
- 1. Fuel temperature
- 2. Clad temperature
- 3. Reactor power
- 4. Retention of water in the reactor tank The following summarizes the conclusions obtained from the reactor design bases subsequently described:
a) Fuel Temperature The fuel temperature is a limit to steady state power oper-ation because of the need to avoid changes in the U-ZrHx fuel which could lead to sufficient pressure to rupture the aluminum cladding. Since most of the fuel to be used in this TRIGA is of the low hydride type (ZrHx where x<1.5), the phase transition that occurs at approximately 530 C0should avoided. 'Ihus, a limit of 530 C has been set for the fuel in this reactor, which will contain mostly U-ZrH i ,0, the older type fuel. A few of the elements may be of the never type, with x:Pl.5 in ZrH . xThese have much higher operating capability, approximately 1150 C for the stainless steel clad type. b) Clad Temperature Both the aluminum and stainless steel clad must maintain their cladding. Hence , they must not melt or reach a temperature sufficient to reduce their yield strength to the level that will allow the internal pressure inside the fuel rods to cause swelling and rupture. A conservative temperature limit of 400 C has been set for the more restrictive aluminum clad, at which the creep r ate from internal pressure is negligible.
3-. l . f u F c) Reactor Power . L Fuel and clad temperature limit the operating power of the reactor. F l u However, it is gconvenient to place an operating limit on core power which is . easily measured,- - The de-sign bases analysis indicates that operation at 2000 kw with natural con-vective flow will just reach DNB conditions (Departure from Nucleate Boiling) (~ u. and that below DNB, neither fuel nor clad temperatures will exceed their y j limits. The allowable K-excess limit is also an inherent limit on core c power, preventing long-time steady state operating powers above 600 kw, L and initial steady state power (before xenon buildup) above approximately f , 1000 kw. Thus, the requested licensed operating power limit of 100 kw with u 2.25% cold K-excess is most conservative. . u . d) Retention of Water by Reactor Tank Though not directly affected by operating power levels, the I L reactor tank must maintain its integrity so that the water level above p the core will not drop significantly enough to create high radiation L_ levels above the tbp of the tank'during' times that the reactor is shut-W down. The tank has been designed with double containment, and emergency tap water lines are available for adding makeup water should leaking occur. P Use of the double tank with sand between the liners makes the system more resistant to earthquake damage than would be a single rigid tank. 5.2.1 Reactor Fuel and Clad Temperature The reactor fuel temperature limit is dictated by the character-istics of the ZrH xceramic fuel matrix. These characteristics are shown in Figure 5.2.1, where it can be seen that the " low hydride" (x in ZrHx less than 1.5) undergoes a solid-solid phase transformation at approximately l 5300c. The hydride content in most of the ZrHx fuel elements to be used in this TRIGA is x=1.0. Therefore, 5300C has been established as the upper g 11mit for the fuei matrix. it this tem,erature, the equi 11brium hydrogen 5-12
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.a f pressure is less than 0.1 psia, as shown in Figure 5.2.2, and distortion L
of the matrix by phase transformations will not occur. Thus the cladding Y - L integrity is assured since the stress at 0.1 psig on the 0.030 inch thick aluminum cladding is only 2.4 psi and 3.6 psi for the 0.020 in stainless steel cladding; S = fr., i.e, tangential stress in a thin-walled vessel is t where r = 1.47 inch radius, t = vessel (clad) F 2 L thickness, and P = internal pressure The other possible causes of high internal pressure are: a) fission gas buildup N b) chemical reactions, such as metal-water reaction from water seeping through small breaks in the cladding and reaching the fuel matrix. P L In regard to fission gas release, the U-ZrH fuel has been shown u to retain a large fraction of even the gaseous fission products. However, assuming conservatively that all of the noble gases escape from the fuel matrix,107, burnup of the fuel loading in an element will create approxi-mately 0.003 moles of noble gas atoms. If all of these atoms leave the P L fuel matrix and accumulate in 10 cm3 of effectiv plenum area at a temper-y sture of 400 C, the total internal pressure in the fuel element due to these noble gases will be 15 atmospheres, or 220 psig. The use* of 10 cm3 ,, the characteristic volume is realistic since a large fraction of the noble gases will not even escape from the matrix. Using the above equation, o _ the stress _in the cladding is 5200 psi under these conditions. Typical stress vs. temperature data (for instance, see Nuclear L' Engineering Handbook, H. E. Thorington, Editor,1958) for aluminum and aluminum alloys show that the creep rate under these conditions is quite 1
- As shown in Section 8, actually about 3.3 cm3 of void is built into the
^ element as the clad-fuel gap. Other voids and interstices in the fuel probably account for much more than an effective 10 cm3 of void. re 5-14
l%=v===w =:, = x- . _ .- . . = _ _ __ __ -m_. . . tolerable, less than 10-3% per hour. Thus, 4000 C is an acceptable limit At for clad temperature with a fuel element burnup of as high as 10%. the 107, burnup level, the reactivity penalty will be intolerably high, and will generally preclude use of more than a very few fuel elements to this level of burnup. Fuel element logs will be kept in which approximate burnup will be periodically estimated.* v Among the chemical properties of U-ZrH,the reaction rate of the material with water that might leak in through the cladding is of particular inter est. Since the hydriding of zirconium is an exothermic reaction, water will react more readily with zirconium than with zirconium hydride, The reaction is Hence the water reaction is unlikily to occur with ZrH. f more likely to occur with the uranium. Experiments carried out of Gulf General Atomic Incorporated have studied the quenching with water of both powder and solid specimens of U-ZrH after heating to as high as 850*C. A relatively low chemical reactivity with both water and air was observed. Thus, a water leak in the cladding will not result in a rapid chemical s reaction, and gases produced in the slow chemical reaction would probably F In summary, limits of 530'C for the fuel element matrix and 400*C P for the aluminum fuel element cladding (conservative fnr th, menint.a eind P fus1L will assure sufficiently low internal pressures and sufficient strength L, for the cladding that the fuel element integrity will be maintained and the n ' cladding will not be breached. The design basis limit (safarv limie for reactor operation) is that the temperature of an aluminum clad low hydride l
- The estimated average burnup on the University of Arizona fuel is 0.6% This maximum, or less than 1% on the' hottest' (highest burnup) element.Similar op is following 13 years of operation.'
E cipated at the University of Utah. 5-16
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1 ( y full element shall not exceed,530*C under any conditions of operation. ' Futhermore, if the cladding should break for other reasons, these same temperature limits will assure that whatever chemical reaction occurs with { the fuel will be quite non-violent. L F s 5.2.2 Reactor Power ( bases limitsTheandpermissible fuel and clad temperature are the fundamental design a temperature instrumented fuel element will be employed to ' monitor full temperature in the core, as specified in the technical specific-L J ations for the reactor. r The anticipated peak full and cladding temperatures have been obtained by othera =4a= = m=arical methods, and the derivation - and results are reproduced in Appenatx u, as extractea careccty rrom cne F Safety Analysis Report for the TRIGA reactor at the University of California's u Santa Barbara Campus. It is sufficient to state here that the results of u that calculation show that, for convective cooling, departure from nucleate boiling (DNB) does not occur until above a total reactor power level of u 2000 kw. Furthermore, the maximum allowable K-excess of 2.257. will not allow long-time steady state operating powers above 600 kw. Prior to u reaching xcuou equtilbrium, this K-excess would allow operation to close j 'l000 kw. Thus, the 100 kw maximum allowable operating power is most conservative. Furthermore, the K-excess limit is a backup inherent power limit, which will limit operating powers such that fuel surface The following temperatures would remain far below melting conditions. ( table shows the approximate temperatures for the hotspot region of the core at a total operating power equal to 100 kw licensed power. 5-17
l Core inlet-water temperature 70 F Core average outlet temperature 110 F Core outlet-water temperature-hot channel 122 F Peak, cladding temperature (70C F coolant inlet) 297 F 0 Peak drop across 1-mil clad-fuel air gap 140 FAT l-l Temperature difference between center and edge 67 FAT of fuel (peak) E.. . It is apparent from the above table that the reactor at 100 kw, at which the initial operating license request is being made, is operating I at conditions that leave the hotspot highly subcooled and the fuel element meat and cladding temperatures well below the maximum allowable from the design basis. (AppendixIIIcontains details of the calculations of these l I temperatures.) The 100 kw power level _will also serve as a controlling _ parameter. The power will be continuously indicated to the reactor I operator by the readings of the nuclear ionization chambers. Calibration of these chambers will be undertaken at low power with the use of foil or wire type flux detectors. Both bare and cadmium covered data vill be taken, when feasible to determine the thermal flux activation component, from which the fission rate in the U-235 will be determined. Multigroup diffusion
- theory calculations could then be used to obtain details of the radial and longi-tudinal flux and power distributions and of the neutron energy spectrum.
i If possible, a U-235 fission rate using catcher foils will be measured at one location in the reactor. From these variety of data, it will be l possible to calibrate the flux chambers to within an uncertainty of less thani107.. 5.2.3 Retention of Water by Resetor Tank The reactor tank consists of an inner liner, 7 foot. 8 inch outer 5 diameter, of 5/16 inch thick aluminum and an outer liner of 3h61nch thick steel u 5-18 l
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l with a diameter of 12 f t. All soil surfaces of the latter tank are nainted with epoxy to inhibit corrosion. The two foot space between the liners is filled with tamped sand to the concrete pad base. The bottom of the aluminum tank is a welded sheet of aluminum material the same as the side walls. (6061-T6 aluminum) Both tanks are water tight, and welds on the inner aluminum tank have been made to meet ASME unfired pressure-vessel standards. (See Figure 4.4.1.1) Under a hydrostatic head of 24 feet of water, the hoop stress at the bottom of the inner tank is 3500 psig, while the hoop stress on the outer steel tank, if subject to the same head, would be 10,000 psig, both well below the yield stresses for these materials. Since the bottom of the aluminum tank rests on a r=M^---A "me rete end which forms the bottom seal for the outer steel liner, the maximum stress that the bottom of the aluminum tank is suh4--- :: i: 9 - hoop stress at the bottom weld. Therefore, the only major consideration is the ability of the tanks to resist the unexpected, such as damage from earthquake, penetrating objects impinging on the inner tank, or from corrosion. Corrosion problems appear to be of little concern. Normal ground 0 temperatures are approximately 60 F, and the reactor water will be main-tained within a few degrees of that temperatute when the reactor is shut-down for long periods (i.e. over night) . Through 13 years of operation at Arizona with tank water maintained in the 600F range, no buildup of corrosion has been observed on any exposed aluminum parts beyond that ini-tially formed as a protective layer. The outer wall of the aluminum tank and the inner wall of the steel tank will not be exposed to any moisture. The outer wall of the outer tank should be adequately protected by the epoxy paint, asphalt seal and polyethelene sheeting, particularly since the ground water level is w g below the bottom of this tank.
--________.____-19_ _ _ _ _5_ _ _ _ _ _ _ _ _
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- r. . . . i u,.w 57 Rupture of the inner tank will be detectable by a drop in water
' level .
Even a large breech in the tank wall, however, should not result - in a fast loss of water, which must percolate through the sand and concrete ~ pad
.:, below the aluminum tank. If the integrity of the outer tank has not been breeched, the inner tank water level cannot drop more than 6 feat, based on a sand porosity (packing density) of 607..
The possibility of an earthquake of sufficient magnitude to damage and breech both tanks, though quite remote, still exists. It is believed that under expected types of ground motion, the sand packing between the g - nular portion __of the two tanks adds ductility to the system such than it is better able to resist ground shear forces that might otherwise fracture a rigid fully concrete-filled structure. None of the above considerations relate to operating power levels of the reactor. In fact, operating power has no effect on the tank materials, W f; except possibly the concrete g . Aluminum is one of the most resistant materials known to fast neutron damage. Concrete will suffer some embrittle-ment, but is unlikely to be significantly affected over the projected 30 , year life of the system, for which peak integrated fast doses will be less than 10 19 nyt. The steel outer liner will be in such low flux that radiation damage will be negligible. In summary, the double containment _ system is designed to safely retain the water in the reactor tank, except perhaps under the most cata-strophic earthquake. Should the system loose some water, an emergency makeup i hose is readily available, connected to a tap water line. Furthermore, a low water level indicator on the tank immediately sounds alarms both in the reactor lab and at the Campus Security Office. Should uncontrollable 5-20
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. O F
l s . 5.3 Mechanical Desian and Evaluation l 5.3.1 General Description The reactor is located below ground level inside a shielded reactor tank. The general design of the tank and shield are shown in Figure 5.3.1. This design provides for shielding of the neutron flux from the earth by at least 2 feet of water
- in the tank and 2 feet of sand surrounding it. Approximately 20 feet of water and/or con-crete and/or sand and earth above and to the side of the core provides the
. necessary paisonal shielding. ,, ,, , , , , , , , _ , , , , ,, . , .
The reactor system will contain provision for 3 diagonally directed beam tubes between the reactor core and the reactor room 1 , floor. Each tube will be composed of two sections. aligned along a l l common axis. The top tube section will be a 1 foot diameter tube between the reactor floor and the wall of aluminum reactor tank. This tube will not penetrate the aluminum tank but will be sealed at the end. where it butts against the tank. Except when in use to extract a neu-tron beam for experimental purposes the upper beam tube will be adequately , shielded by being filled with sand and capped at the reactor floor level with a 1/16 inch thick locked steel cap for security. When the beam tube is to be used, the sand filler will be removed and a sealed aluminum beam tube (carefully weighted so as to have a net density gzester than water) will be installed between the inner tank wall and the reactor core shroud along the conunon axis of the upper tube. I i
- Operating procedures will require that at least 2 feet of water reflector-shield be between the core edge and the inner tank of aluminum.
5-22
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Special attaching' fEtures will b'e welTe'd'tolhe ' inner tank t/ wall to permit correct alignment. During beam tube use since both the lower and upper beam tube sections are sealed against air flow so that no radiation problems arising from argon-41' or nitrogen-16 buildup will occur. Furthermore, no penetration of the reactor tank is made for the beam tubes. Both inner tubes only butt against the reactor tank so no increased risk of inner tank water leakage is incurred. When the beam tube is being used appropriate shield at the reactor floor level will be made with dense concrete block and other shield ma.terials ava. ilable. E. s The core components are mounted between two grid plates, which are part of a fixed structure setting on a six-legged pedestal support. This entire pedestal structure can be moved as a unit by lifting it with a crane, thus permitting occasional changing of locations of the core within the tank in order to accomodate large experiments between the core and the tank wall (such as a thermal column or beam tube, if added in the fu ture) . Note, procedures will not allow shifting of the core without (
?firsc partially unloading the core to insure suberiticality.
Various irradiation facilities may be provided within the core of the TRIGA Mark I reactor. This particular grid plate structure will allow f n a central experimental flux trap the size of seven fuel elements and for possibly two off-center triangular shaped, three-fuel-elemer.t sized irrad-l intion facilities. These special facilities may be replaced with fuel elements, within the technical operating limit constraints, by inserting special inserts into the large holes in the upper grid plate. In addition f to each of these special irradiation facilities, and irradiation space 5-24 E
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I G can be provided in any one of the fuel element, spaces. One such special irradiation device is a high speed pneumatic " rabbit" transfer system, i operating.with nitrogen . gas as its propellant. 5.3.2 Grid Plates and Core Structure The upper and lower grid plates were constructed from drawings i and specifications supplied by Gulf Environmental Systems, Inc., except .. . that an effective lattice pitch of k,49 inches between centers has i L been adopted, identical to that of the TRIGA at the University of Cali-I fornia at Santa Barbara. l Upper Grid Plate - This plate is 3/4linch thick, type 6061-T6 , aluminum. It laterally positions the fuel elements, control rods, ir-radiation facilities, and neutron source. The plate carries no vertical load other than its own weight. It is mounted to six side plates, each 3/16 inch thick 6061-aluminum which are secured to the\o5 tom grid plato. ... The upper grid plate has 127 locations for fuel elements (or moderator rods, control rods, etc.). These 127 holes are in six hexagonal rings around the center hole and are 1.505 (+0.005. -0.000) inch in diameter. Cooling water passes through the top plate by means of the clearance provided ~. by the differential area between the triangular spacer blocks on the top , of each fuel element and the round holes in the grid plate. There are also several small diameter flux-wire insertion holes in the top grid plate in the interstices between the fuel element holes. A 7 central .! fuel element cutout is provided for A & B ring positions which can acco:2- f modate a 4.215 inch diameter experiment with a cross section area of 13.96 p r square i,nches. Y ..- E - - _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _
p -- - ~-- - . l
- L, u
Two triangular holes encompassing three fuel elements and with their spexes rounded to the shape of fuel element holes maybe provided in the upper grid plate at D and E ring locations. These cutouts may be removed to accomodate special experiments, or may be inserted into the upper grid, supported on an undercut step. Lower Grid Plate - The lower grid plate provides lateral posi-tioning and supports the entire weight of the core. It is also type 6061-
,T6 aluminum, and is 3/4 inches thick. It rests on six legs (pads),
six inches from the base of the aluminum reactor tank. Attached to the lower grid plate is the 3/16 inch thick hexagonal aluminum shroud plates for the core. l- The lower grid plate also contains 127 holes for fuel elements, but each hole is 0.25 inches in diameter with a 5/8 inch diameter counter sink. The tapered fuel element ends fit smugly into these holes. Cooling water cannot pass through the fuel element holes in the lower grid plates. Cooling water does, however, flow through 5/8 inch diameter holes in the interstices between each fuel element position in the lower grid plates. The fuel insertion holes in the lower grid plate of course line up with the similar holes in the top grid pl.it's. Since the control rods contain
'no followers, any fuel element position may also accomodate a control rod.
5.3.3 Moderated Fuel Elements The aluminum and stainless steel clad fuel elements and the optional , graphite or heavy water loaded reflector' elements all have similar outside dimensions. These are nominally 1.47 inches 0. D., except for the upper and lower and fixtures, and 28.44 inches long tip to tip. Figure 5.3.3 shows the typical fuel element. The fuel is a solid, homogeneous mixture of 5-26
-._c_,.---
- y. -- . - . . . . : .__. _ . _ ,
uranium-sitconium hydride alloy containing about 8 1/2% by weight of uranium enriched to 20% in U-235. The hydrogen to airconium ratio is approximately 1.0. Each aluminum clad element is clad with 0.030 inch (- thick sluminua and the stainless a eeel clad elements with a clad 0.020 inches thick. The active fuel portion is 1.40 inch nominal diameter by 14 inches long. On each end are 4 inch long sections of graphite. Aluminum and fixtures are on both ends of the fuel element, as shown in the Figure. The upper fixture contains a knob for grasping by the fuel handling tool. Table 5.3.3.1 is a summary of the fuel-element characteristics. ( I 4 5-27
L ..e i I . I TABLE 5.3.3.1 Summary of Fuel Element Specifications Item Nominal (Value) Fue oderator Material umin htainlessSteel l H/Zr atom ratio 1.0 1.0 Uranium content 8.5 wt7. 8.5 wt% Enrichment (U-235) 207, 20% U-235 mass (per element) 37 gm 37 gm lq Diameter of fuel meat 1.41 inches 1.41 inches Length of fuel meat 14 inches 15 inches Graphite End Reflectors Porosity 207. 20% Diameter 1.4 inches 1.4 inches Length (top and bottom) each 4.0 inches 4.0 inches Cladding Material Aluminum Stainless steel 304 g Wall thickness 0.030 inches 0.020 inches l Length 23 inches 23 inches End Fixtures Aluminum Stainless steel 304 Overall Element
- Outside Diameter _1.475 inches 1.475 inches Length -28.4/i inches 28.37 inches I Weight 7 pounds ;
8 pounds l I 5-2, I
(; ..
. . . . . - - . - . . . . . - - - - = - . - - - - .
k r L 5.3.3.2 Evaluation of Fuel Element Desian [ Gulf Energy and Environmental Systems, Inc. has acquired exten-( sive experience in the fabrication and operation of U-ZrH fuel elements. With these elements operating well below the design bases limits, no stress-associated failure of the elements is expected. The moderator elements if employed will be manufactured at the University of Utah shops. Since these elements experience no nuclearly induced stresses, failure of the elements would not be expected. If water should leak into the graphite f or heavy water, the effect would be a loss of reactivity. , , , , , 5.3.4 core The active core will have the approximate shape of a right _ hexagon. The fuel element spacing is such that 33% of the cross sectional core area in the lattice is water. The extra fuel element spaces on the outside of the active core allow for the insertion of reflector elements. The minimum critical loading using the elements that have been in the University of Arizona's reactor is anticipated to be 62 elements, with - control properly placed within the critical array and with the core surrounded by reflector elements. The operational loading with a cold clean k-excess of 2.256k is expected to be .pproximately 75 elements, if these are reflected with one row of reflector elements, or about 80 elements with no reflector elements._ 5.3.5 Reflector The reflector will generally consist of a single row of graphite or heavy ==*-- eeflector elements. These are such that the equivalent volume
, fraction of the first 1.6 inches of reflector is approximately 33% water, 62% graphite or heavy water and 5% aluminum. The remaining part of the 5-30
( - _ - _ _ - _ _ _ - _ - - _ _ _ _ _ _ _ - _ _ _- .-__ - _ _ . - _ _
l I reflector is water, plus the 1/4 inch thick aluminum core shroud. The effectiveness of this combination reflec, tor is expected to be approximately 1.5%AK 1ess than that of the mostly graphite reflector used on the University of 1. Arizona reactor. This means that the required number of fuel elements i to achieve a certain operational ~1oading will be between 5 and 15 more than the number used in the Arizona configuration. 5.3.5 Irradiation Facilities The irradiation facilities that will be installed are constrained by the si F of the.variours' available locatiors. The center seven elements will accomodate a facility up to 4.2 inches in diameter. However, since the central facility is intended to utilize a flux trapping technique,, much of this diameter will contain water or other moderator. There may also be provided two triangular openings in the top grid plate to accomodate a , round facility as large as 2.4 inches in diameter. All other facilities will need to fit through the' 1.5 inch-diameter fuel elenient positioning holes s in the upper grid plate, unless they are placed outside of the core shroud. I A pneumatic rabbit may be operated within one such 1.5 inch 0. D. tube. It will be driven by dry nitrogen gae.so as to minimize Ar-41 pro-
~
duction. As noted in Section 5.1, flooding of the central irradiation ; space from a sealed gas filled condition is the worst flooding accident. Its worth is estimated to be 0.6%Ak, less than one dollar. Thus the 5- worst flooding accident cannot make the reactor prompt critical if initiated from a steady power condition..
*0r acuivalent nas with a low radioactivation potential. - 5-31 F
L
-.x 2. u. . . . - . . . . _ . _ _ _
The pneumatic rabbit will have a slight curve in its tube such f that direct unco 111ded streaming of neutrons from the core to the sur ace Most dry irradiation tubes inserted into the of the pool cannot occur. reactor will not be dry to the top of the tank unless specifically de- I Under .s'uch' conditions, special pre-signed for use as beam experiments. cautions will be taken to adequately shield the floor above the reactor from irrad.iation by ,these b,eams. , , ,, l 5.3.7 Control ' These rods operate in The reactor contains three control rods. The tubes are f aluminum guidetubes in fuel-element positions in the core. The 3 rods are located (aocrog) '- f held in pla,ce by the top and bottom grid plates. l svarnetrically near the center of the core and include a safety rod, with a worth of approximately 2.0%; a shim-safety rod, with a worth of about 2.0%; Figure 5.3.7.1 and a regulating rod, with a small worth of about 0.57.Ak. 2 hows a control rod , , magnet ,, atnd_, guide tube,. , L .~. . These drives, of
'All rod drives will be essentially identical.
a vinchtype design shown in Figure 5'43.7.2 were designed at Gulf Energy and Environmental' Systems, Inc. and have been in use on the Arizona re-The components for the control-rod drive mechanisms actor for 13 years. The components consist of a motor, I are arranged in line on a base plate. . i L a brake, a cable drum, a 10-turn Helipot, and a limit-switch mechan sm, r The drives with suitable hangers and flexible couplings for mounting them. L are mounted below floor level on a concrete shelf alongside the reactor ( A housing will be provided for pit, underneath the reactor pit cover.
- each drive.
f The motor is a nonsynchronous, single-phase, instantly reversible - 1 I type. Contained in the motor housing is a four-stage spur-gear reducer 5-32 f I .
-_-_,=.-.-:=-'-
M S-i
- unit having a total speed reduction of 900 to 1 .
This motor gear reducer unit has an output speed of 1.9 rpm. shaft is locked by means of an electrical brakeWhen the . utsting rod, a special servonocor replac'es the con In the case of ventional motor. A stainless-steel aircraft cablezontally extends from the hori drum to an enclosed cylindrical sheave mounted di guide tube. rectly over the control-rod Attached to the lower and ofo the cable is a h l ding magnet. This magnet travela up and down in the guid on the absorber rod. e tube and ecgages the armature In the event of a po'wer failure or a scr am signal, all magnets de-energine, thus allowing the abs The pitch diameter of the cable drum is such aorber r s to give a linear rod speed approximately equivalent to 3 cents /second f , center of its travel. or the rod near the Overtravel is prevented by means of both r- elect i cat and mechanical stops on the limit switch anism. mech { The limit posi- . m tions are adjustable and can be set for an y desired travel. . Instrumentation and Control System \' - E The instrumentation and controls eactor foroperation normal r \ are contained principally in the main control co , nsole, shown in Figures 5.3.7.3and 5.3.7.4.' For the initial approach to criticality, additional instrumentation is provided in an auxiliary rack. Figure 5.3.7.5 shows a block diagram of the principal features of the i system. . nstrumentation and scram Detailsd Jpecificationsof the various neutron .
-measuring channels t
described below are given in Table ... 537 A -2 curie Fu-Be start-up source, the a initi l alb neutrons /sec, is loc aet d source strength of which is suspended in one of the top arid plate fueler h lin a spec - o es. (see Fig. 5.3.7.6). Once 5-35 W - - - - - - _ __
~
ition the reactor is critical the source can be moved from its in core pos f the secured to storage position by means of a cord attached to the top o An indicator light (and source interlock) (See Figure source holder cap. 5.3.7.4) visible to the operator will indicate if the source is in or out during startup. . The ion chambers and counters are located in the water approxi- i f the core-mately on the core axial midplane on the outer perimeter o I shroud assembly. A fission counter is used with a transistorized linear amplifier 1 s The log / and preamplifier to feed both linear and log count-rate circuits. Both linear and log count rates may be h ' count rate is read on a meter. te ( read on the linear and los recorders when the reactor is in the range of from about 10-3 w to about 2 w (sourch level). These source levels d 10,000 counts / r are < estimated to give count rates of about 5 counts /second an An interlock circuit is used to prevent rod with-second, respect ively. f l drawal unless the source count level is above the required minimum va ue The amplified pulses from the fission counter I of at least 2 counts /second. le. scaler ar'e also audible from a small speaker mounted within the conso l The log n channel uses a compensated ion chamber and, with the
' from about 1.0 w transistorized log amplifier, will give usable readings The log n output is read on a meter and on the log recorder.
g to 300 kw. The output of the log n amplifier circuit also feeds a transistorize d period [ cir cuit. This furnishes a period indication on a meter from -40 to co to
+7 seconds.
5-40
'l
?
i pf _- . . . . - . (- los
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3 a.5 5 5 I 10 5,' ,5 ._1, l t 5 5
. 10 2
I. t' o . 5 5 g N/C/SEC LOG g 3 , s 5" v( ,, . - 3adC/SEC LINEAR . a ", I .E l" a 'e I Icr3
.. SOURCE LEVEL IC/SEC COUNT-RATE LOGn POWER POWER (I-o o
CHANNEL CHANNEL NO.I CHANNEL NO.2 LINEAR COUNT-RATE ,. FULL-SCALE (RECORDER SCRAM) 30,lO0,300,1000,3000 C/SEC . DISPLAY RECORDER ONLY LOG COUNT-RATE 4 f FULL-SCALE (RECORDER SCRAM)1-10 C/SEC I
- DISPLAY 8aETER AND RECORDER f (QG_n I j
FULL-SCALE (RECORDER SCRAM)3.0 to 300 kw i WSDLAY NETER AND HECORDER POWE3t CHANNEL NOI FULL-SCALE (RECQRDER AND ELECTRONIC SCgAM)3.10.30.100,300 1000.3000.10000' 30000,100000,300000 w DISPIAY METER AND POWER CHANNEL NO.2 ~ FULL-SCALE FULL POWER (RECORDE'R'dND
, DISPLAY NETER ONLY ELECTRONIC SCRbt)0-N.
i
-Estimated range of reactor instrunent channels k
I ~ e i
.s Figure 5.3.7.7[ -
5-42 t 1 .. -
l B f TABLE 5.3.7 SPECIFICATIONS FOR NEUTRON-MEASURING CHANNELS 1
- 1. Count-rate Channel (one required):
- a. Counter--Westinghouse WL-6971 or equivalent fis on counter having a sensitivity of 0.14 count / neutron /cm
- b. Preamplifier--transistorized model.
- c. Linear amplifier--General Atomic transistorized model.
- d. Log count-rate meter--General Atomic transistorized model.
- e. Linear count-rate meter--General Atomic Model CR-100.
- 2. Log n and Period Channel (one required):
- a. Chamber--Westinghouse WL-6377 or equivaleg compensated ion chamber having a sensitivity of 4 x 10- amp /ny, electrical ensation, and compensated ( sensitivity of about 3 x comp 10- 3 amp /r/hr.
I
~
~j b . Log n amplifier--Bu'rr-Brown Model 3061/25 intearated circuit opera-
{ tionalamplifierfollowingamultiple-sigicon-diodeloginput
- * -eini t . Ranne: 5 decades from 4 x 10- to 4 x 10-9 amp.
- c. Period amplifier-General Atomic Model AP-130 transistorized period amplifier having a range of -40 to co to +7.
l d. Log n recorder--Varian G-11 (also used with log count-rate circuit).
- 3. Power-level Channels (two reguired):
- a. Ion chambers--Westinghouse WL-6937 or equivalent uncompen-sated ionization chamber using enriched U 30g and having I a sensitivity of 4.4 x 10-14 amp /nv and a ( sensitivity of 4.5 x 10-11 amo/r/hr.
is estimated to be 2 x 10 1(r /hr .) Full power T flux at chamber
- b. Power-level amplifiers--(1) Burr-Brown Model 3061/25 integrated) circuit operational amplifier, accuracy 0.02%, drift less than 0.03% of full scale for 8 hrs within temperature limits of 200 to 500C. (2) Chamber feeds meter and scram circuit directly through an attenuator. Range 0%-150% full power.
a 5-43
e L. . [ The layout of these controls and the various monitoring lights is shown in Figure 5.3.7.4. The reactor cooling-water temperature _is measured at the tank. The fission-product activity of the water is monitored con-tinuously, and a red pilot lamp lights, and an alarm bell sounds if the activity rises above a preset level.. A brief check-out procedure for the reactor console prior to start-up is as follows: The power switch is used to energize all in-strument and control circuits except the magnet-holding circuit. Filament circuits of the few tubes involved in the console are normally left on at all times (they may be turned off when necessary at the rear of the c onsole) . The key-operated switch will turn on the control rod magnet power supply. This arrangement allows technicians to check out most of the console without inadvertently pulling a control rod in the absence of a licensed operator. Only licensed operators have access to the console key. A minimum source count from the count-rate circuit is also used to interlock the rod-drive circuit to prevent raising of the rods unless the source count is adequate. The power-level amplifiers are first adjusted for zero and then a test signal is used which not only tests the amplifier but also calibrates it. A test signal may also be used to raise the signal level to check the scram circuit. Artificial signal sources for low and high calibration points are also available for checking the log n. A sig-
, nal is available for testing the period circuit and for setting the period scram. The operation of the count-rate channel can be checked by the source count and the channel can be calibrated at both high and low counting rates by means of test signals.
5-45 ,
_ jg_ -- L F l 5.3.8 coolina system Cooling of the reactor is accoinplished by natural convection of the pool water. A five ton rated capacity evaporative cooling tower _say also be uti11:ed as the heat sink. Heat will be rejected from the reactor tank to the cooling tower through a heat exchanger. Initially the reactor will be operated without the cooling tower heat sink by permitting heating of the reactor tank water up to a temperature of 95'F. The heat exchanger will receive warm water from the pool. and returns cooler water to the pool. The return line carries the water to a location several feet above the core where it is ejected into the pool ' I : water as a horimor.tal jet in order to diffuse the nitrogen-16 activity and pre-vent it from reaching the top of the pool as fast as it would otherwise. The discharge temperature will be about 50 F. iloth the inlet and return . lines to the pool have small =1/4 inch holes approximately one foot below the normal pool water level. These holes will stop siphoning action and i prevent draining of the pool below that level should a siphoning drain be accidentally initiated. The cooling tower water to the heat exchanger will be circulated at the rate of approximately 20 gallons per minute by a centrifugal pump. Since the cooling tower is subject to freezing, it will be provided with weather covers, and the water will either be drained from I it or be continued to circulate when temperatures are below freezing. Normally the reactor pool temperature will be maintained at about 60 F, which is approximately ,the ground temperature. For any opei ation above 25 kw, the heat exchanger-cooler has insufficient capacity to maintain the 60 F. The pool water heat capacity is app roxima tely 5-46 l
e- . ._ : ._ .__ . _ _ _ _ _ _ , l J L . .. _ ... . ._ . . 22 kw-hrs /*F. Thus, at 100 kw and with the heat exchanger operating the pool {. temperature will rise approximately _5'F per hour. The water temperature will not be allowed to exceed 95 F. However . af ter the reactor is shutdown, .. t the cooler can lower the pool temperature at approximately l'F'per' hour. - 5.3.9 Water Furification System A small ion-exchange water purification system will be installed inside the reactor building. Water from the pool will be circulated through [ this bed at a slow rate (1 tolo gallons per' minute) by a centrifusal pump. The water will be both taken fron'and discharmed'to tha' top:of tha'. pool. The ion exchanne system is of standard design, using readily availsbJe r [ mixed resins from commercial manufacturers.
. The water in the pool will be regularly sampled for ph and r i L conductivity measurements. In order to minimise aluminum corrosion, it F_ is desirable to maintain the water ph in the range of 5.5 to 6. Signifi-l L
. cantly lower ph than this value (more basic) will be buffered out by the I addition of some ni:ric acid. Also, the conductivity will be maintained L
within acceptable limits. Failure to be able to do so will then indicate filter bed repiscenent is necessary. ~ 5.3.10 Radioactive Monitorina and Disposal ~ Two types of automatic indicating alarms for high radiation will be continuously operating. These alarms will automatically indicate both ! in the reactor room and at the Campus Security headquarters, and thus will : ~ ~ _ _ _ _ , ireceive around-the-clock surveillance. The alarms when triggered w'111 provide a siaht and sound alarm in the reactor area and operate a red light alarm l L : ~ ^ 5-47 ' r .
. . I at the Campus Security Headquarters.
t One radiation alarm will actually be a low water level indicator, set to alarm if the, pool water level dropa_12 inches below the normal level. This alarm will be actuated bya standard microswitch levered to a plastic float. A 12 inch drop in water level activates the switch. The other alarm will be from a direct reading geiger tube mounted directly above the pool of the reactor. This will provide a high level radiation signal to a relay and will be set for levels higher than ex-pected in all normal operation. Few radioactive wastes will be generated, except those consisting of fission products in the fuel elements. The other main sources of wastes are from induced activity.in irradiation samples and from ion collection in the filter beds. The latter will routinely be monitored for radioactive buildup. Disposal of any of the wastes will follow standard university policy, as specified in the Radiological Safety Manual for the University of Utah. 'i. . .
.. .- i a 5-48 P
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l 6.0 Initial Reactor Startup and Calibration Tests E~ - Personnel All reactor operations during reactior startup and calibration i tests will be under the direct supervision of the Reactor Administrator. A Senior Reactor Operator has been trained on this TRIGA facility at the University of Arizona prior N theiime that the reactor was shipped to Utah. Chapter 9 gives a detailed account of the proposed operations and administrative structure for the Reactor Facility. Furt_hermore the University of Utah Radiation Safety Manual (enclosed with appli-cation) gives detailed emergency procedures for the reactor facility. The following is an outline of the tests to be performed during the initial startup and calibration of the TRIGA reactors 6.1 Pre-critical Test 6.1.1 Functional Tests of Mechanical Equipment These tests include verification of proper installation, and performing operational tests where applicable on the isotope production facility, pneumatic system, water system, fuel element handling tool, and other related equipment. 6.1.2 Checkout of Instrumentation and Control System A complete checkout of the control console is performed to insure that all circuits are calibrated and operating properly. All interlocks and scram initiations are shown to opercte satisfactorily. In addition,
, the rod drives are checked for proper. operation during rod withdrawal and scram situations. The fission and ion chambers are shown to be sen-sitive and responsive to neutrons.
A Keithley micromicroammeter and a scaler may be used as a temp-orary addition to the console instrumentation to provide more information during initial startup. l 6-1
k t concentration assuming air rather than nitrogen filled the specimen rack produces a discharge concentration about 17. of the RCG for Ar-41. To minimize Ar-41 activity the U/U pneumatic tube sys te a will be ' operated with bottled nitrogen gas *(less than 10 ppm argon) and will discharge into the reactor room exhaust system. Even under the most severe operational conditions and assuming that air at atmospheric pres-sure filled the in-core terminal during the irradiation and is completely removed and vented with each sample change, the concentration in the dis-charge would reach about 2 x 10-7 pc/cm3 . Even under adverse dispersion conditions, it is very unlikely that this discharge will not be diluted by a factor of 10 (to the RCG for uncontrolled areas) before the exhaust cloud reaches a populated area. Estimates of probable dispersion are given in Section 8.4.4 Calculation - g The concentration of Ar-41 produced in air irradiation of 100 kw j for time t is X= N(40)4o(40)") , ,-A '(41)t uc/cm3 3.7 x 10 <-
. where + = av. neutron flux = 1.76 x 10 at 100 kw 4.4 ex 10- at 250 kw N(40) = Number of Ar-40 atoms /
2.5x(0 e o(40)=CrosssectionforAr-41 production
= 0.53 x 10-24,2 A(41) = Decay constant for Ar-41 = 1.06 x 10~4 sec'I
- Or equivalent gas with a low radioactivation potential 8-6
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'9.0 Operations and Administration 9.1 ceneral Considerations The reactor will be operated by the Mechanical Engineering De-partment (Nuclear Engineering Program) of the College of Engineering of the University of Utah, Salt Lake City, Utah. The proposed administra-l l tive organization is shown in Figure 9.1. This organization is designed to promote efficient utilization of the reactor facility and yet to es-tablish adequate safeguards for the health and safety of the community.
Each unit in this -network is discussed further below. Also discussed i' are provisional criteria for the examination of proposed experiments and for ensuring safe. operation of equipment. The suggestions are not meant i to be limiting and will be subject to reconsideration by the appropriate administration in the light of operational experience. Biographical resume's of the Reactor Administrator, Reactor Supervisor, Reactor l Safety / Committee, and Reactor Operators are attached as Appendix IV. 9.2 Administration - 9.2.1 Resetor Administrator l The Reactor Administrator will be in complete charge of the facility. He and other members of the Reactor Safety Committee are appointed by the President of the University. The Administrator is Chairman of the Reactor Safety Committee. The Administrator is responsible for liason with the NRC regarding all mattera concerning the facility and for enforcement of all regulations. In addition he is responsible for recommending the appoint-ment of competent personnel as Reactor Supervisor, Operators and other staff of the facility. 9.2.2 Reactor Supervisor The Reactor Supervisor will be the operational administrator of the facility--an academic staff member of the Mechanical Engineering De-9-1
~ W f partment, holding a Senior Operator's License.* He will be directly respon-L - sible to the Reactor Administrator for the operation of the facility; and [ thus will enforce administrative rules and operating procedures, be re-sponsible for all planning and scheduling, and will review and classify l all proposed experiments and modifications to the reactor and submit them to the Reactor Safety Committee for approval. He will consult directly. with members of both the Reactor Safety Committee (RSC) and the Radio 163ical Realth Committee (RHC) concerning all safety aspects of reactor { operation. Within the reactor facility he will enforce compliance *with the requirements of the RHC and be responsible for monitoring operations. Other reactor operators and trainees will be under his control. . The Reactor Supervisor will be responsible for the operation, maintenance and scheduling of not only the TRIGA reactor but also of the . I i . AGN reactor, neutron-senerator. californium irradiator. and@ther , irradiation facilities in the Nuclear Laboratory. 9.2.3 Reactor safety Comittee (RSCl This Coasmittee,which is chaired by the Reactor Administrator, must approve new and unreviewed reactor' operations and till periodically ( I ,-
- review all operations of the entire nucle,ar , facility. Members 'are pirr-sons having experience and interest in the use of reactor facilities. The I Reactor Supervisor and, the Radiation Safety Officer are a ers._ , _
The Reactor Supervisor will report to this Consnittee on any problems concerned with the reactor. In particular, the approval of this Committee will be required for start-up of the reactor following major modification or repair. He must be able to satisfy the Committee at all ,
- Initially for the first critical startup of the reactor a single senior reactor operator will be licensed. The Reactor Supervisor will obtain a senior operator's license within one year of the initial startup.
9-3 L . _ _ _ . - - - - --- - -
R l times that the reactor and facility are operating properly. The Committee j may forbid any operation it does not believe is safe. This Committee must approve all documentation relating to reactor operation. All experiments, which may significantly alter safety consider-ations previously reviewed by the Committee, must be formally submitted
. to the Committee as test programs for the Committee's review and approval before such experiments may be conducted. Experiments which do not alter I
safety considerations but nevertheless are deviations from previously ap . i ., proved experiments must be documented in the reactor files and ' approved . _ _ _.. ..J by signature of the reactor supervisor. 9.2.4 Radioloaicai Hestth' Committee RHC) t This Committee, appointed by the President of the University advises the President concerning radiological safety on Campus. This Committee controls the movement and use of all radioisotopes and radiation producing machines on Campfa. All isotopes produced in the reactor for use outside the Department will be controlled and regulated by this com-mittee. The regulations of this committee will be enforced within the reactor area by the Reactor Supervisor who is a member of the EHC.
, \
The Conusittee and the Radiation Safety Officer will be consulted and kept fully informed concerning radiological problems in reactor operation. At present, this Committee com7?ises a Professor of Radiology (Chairman), a Professor of Internal Mtdicine, a Professor of Radiobiology, a Professor of Physics, a Professor of Medical Technology Science, a Pro-fessor of Mechanical Engineering (Reactor Supervisor), a Professor of Biology (Radiation Safety Officer), and a Professer of Chamf atry, \ 9.2.5 The Redistion Safety Officer An experienced Health Physicist serves on the Staff and functions as the Radiation Safety Officer for the University. He is responsible for 9-4 i
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l . the radiological health and safety of the University. Working through the Reactor Supervisor he ensures the sa'faty of operations in the re-actor area and is responsible for the transfer 'of isotopes outside the Department. He is available for consultation with the Reactor Super-l visor, and is available to any university official in the event of any emergency relating to radiological safety. He is a member of both the g RSC and the RHC, I p 9.2.6 Reactor Operators Re- \ All reactor operators will be licensed by the U. S. Nuclear ; l Initially a Senior Reactor , gulatory Commission to operate this reactor. 1 Operator has been trained to operate the TRIGA reactor at the Universip ,
- ~ ~- - - -- - ...-.. . . ... .
of Arizona where the reactor is presently located. Subsequent operators will be trained under his supervision. Graduate students may take ex-aminations for an operating license and be employed as reactor operators. Initially, the reactor supervisor will be trained on the console by the previously trained senior operator. Due to prior operating experience. the senior joperator can be licensed before the initial loading of t5e' o 'f. I I reactor. The Reactor Supervisor will be licensed (senior operator's license) t l within a year after the critical startup of the reactor. l I 9.3 Administrative controts and Procedures 9.3.1 Introduction This section describes the proposed administrative procedures pertinent to the operation of the facility and the performance of exper-1ments. The Reactor Supervisor will be responsible for ensuring comp 11-ance with the established controls, with the Reactor Administrator bearing ultimate responsibility for assuring enforcement of these regulations. 9-5
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F - i 9.3.2 Access to the Reactor Area l Normal access to the reactor will be through the control room. Persons entering the reactor room will be required to wear personal dos-imeters.* The overhead door to the loading area vill normally be ' closed and locked. Emergency exit may be made through the radiochemistry lab-E oratory or through the control room. When the reactor is not operating, I unsupervised access will be permitted only to persons holding authori-sation from the Reactor Supervisor. Such authorisation will require the holder to have a knowledge of safety and emergency proceduies~within
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- the reactor area. Emergency instructions will generally require that a
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I senior member of the reactor staff or the Radiation safety Officer shall 6 . be contacted prior to entry by Police or Firemens - I 9.3.3 Reactor Ooerstion , For each reactor operation, an Operator-in-Charge shall be desig-nated. He shall be respo'nsible for ensuring that: , (a) he is in a position to operate the controt rof the reactori at all times the reactor is critical.
/ i i
(b) the correct start-up check and shua.-down check procedures are followed and that tog sheets are filled out for all operations. e (c) any experiment is correctly guthorised and that any require-I ments noted by the ASC have been complied with. (d) the experimenter's proposed procedure conforms to RHC rec-onusended practice and that the experimenter has EHC approval _if the handling of radioisotopes is involved. I i
- Exceptions apply to tours of visitors at times when the reactor is shut d own. The authorised tour leader need be the only one wearing a personal l dosimeter, but the names of all visitors and the tour leader will be re-corded on the laboratory log.
9-6 _____
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- - - - - - - ~ - -- --
f ,I (e) all samples removed from the reactor are monitored, their activity levels recorded, and any necessary temporary access barriers or shielding are used. 1 (f) the experimenter and the Reactor Supervisor are informed in case of any unusual or unexpected incident, apparent equipment or in-strument failure or malfunction. (g) the Radiation Safety Officer has been notified if any experiment predicted to involve high radiation levels is to be perfermed., l The Operator will normally satisfy (c), (d), and (g) by ensur-ing that the Reactor Supervisor has correctly approved the proposed ex-periment and schedule. All members of the Reactor Staff will expected to be cognizant with basic radiation safety procedures, so that adequate safety is ensured even in the absence of the Radiation Safety Officer from the university campus. 9.3.4 Routine Monitorina and Test Procedures 9.3.4.1 Components The Reactor Supervisor will set up a program for regular testing of all safety equipment, procedures and certain reactor components. In particular, radiation monitors will be ;hecked frequently with calibrated sources, and the alarm set points on the continuous air and area monitors shall be checked . The operation of the emer-gency purge exhaust system and ventilation shut-down shall be tested semi-annually. Similar checks shall be made of reactor controls such as interlocks, control rod drop times, power level safety circuits and the water level alarm. Any malfunction of the above, or related systems shall be suf-ficient cause for suspension of reactor operation until the fault is c orr ec ted . t
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I The results of all tests will be recorded in a log book kept in the control room. _ 9.3.4.2 Radiation Levels l In consultation with the Radiation Safety Officer, the Reactor Supervisor will draw up a scheme for routine monitoring with portable . e., tipment to check radiation levels within and beyond the reactor area especially during reactor operation. Additional measurements will be made during installation start-up procedures so that radiation leakage - levels estimated in this report can be ascertained. The detection of any appreciable radiation above background at any point outside the general reactor area shall be cause for immediate investigation followed by corrective action in the form of procedural change or additiun of shielding. The results of such monitoring will be ', recorded and maintained in the reactor control room. , Personnel monitoring will be under the direction of the Radiologi-cal Health Office which will provide a radiation monitorina service. Theseviil be supplemented in the reactor area by pocket dosimeters which will also be used for occasional visitors. The Operator-in-Charge will ensure that persons entering the reactor room from the control room have a dosimeter. 9.3.5 Authortsstion of Experiments All experiments proposed for the reactor will be designated Class I (Routine or Hodified routine) or Class II (Special or New) by the Reactor Supervisor. , Class 11 comprises all experiments which must be submitted for i review and approval by the Basetor Safety Consittee. In addition.. Proce-dures involving radioisotope production for delivery outside the Noelear_ 9-8
Engineering Laboratory and/or high radiation levelo must b3 cypreved by the RHC. Experiment requests will normally be submitted to the RSC and IRS. (if *PPlicable) by the Reactor Supervisor af ter evaluation by him. Class I experiments include all repeats of successful experi-ments and those that are but minorm'odifications to a previous experiment. . The Reactor Supervisor has the authority to approve these experiments. 9-9 7 E R . R .
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List of Reproducible Figures Needed for SER l'. Facility Floor Layout
- 2. Reactor Tank Cutaway View
- 3. Core Configuration (See Question #2)'
- 4. Stainless-Steel-Clad Fuel Element Assembly
- 5. Aluminum-Clad Fuel Element Assembly
- 6. Primary Coolant Circulation and Purification Loops
- 7. Ventilation System ,
- 8. Control Rod Assembly
- 9. Control Rod Drive Mechanism
- 10. Block Diagram of Nuclear Instrumentation (Update)
- 11. Operating Ranges of Incore Nuclear Detector
- 12. Central Irradiation Facility, with multiple sample loading capability and rotation while being irradiated.
Some of these figures are provided in the proceeding question material
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Summary of Fuel Element Specifications l ?- l I Item Specification
- 1. Element Type Aluminum Stainless Steel 2.' Fuel-Moderator Material U-ZrH, U-ZrHn H/Zr atom ratio (n) 1.0 1.7 Uranium content 8.0 wt-5 8.5 wt-5 Enrichment (U-235) 20 at-5 20 at-5 U-235 mass (per element) 37 sm 39 sm Diameter of fuel meat ,
3 58 cm 3.63 cm Length of fuel meat 35.56 cm 38.10 cm i 3.- Graphite End Reflectors Porosity _205 205 Length (each) 10.16 cm 8.74 cm
- 4. Cladding i Material Aluminum S.S. 304 t-
! Wall thickness 0.762 mm 0.508 mm Length 58.42 cm 58.42 cm (
- 5. Overall Element i Outside Diameter 3.73 cm 3 73 cm l Length 72.24 cm 72.06 cm
! -Hass 3 17 kg 3.63 kg
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