safe operation of a triga reactor in the situation of leu ...

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The core refueling provided better core burnup homogeneity. ... computations of power and flux distribution, reactivity excess and fuel burnup was developed.
IAEA-SM-360/27P

SAFE OPERATION OF A TRIGA REACTOR IN THE SITUATION OF LEU–HEU CORE CONVERSION M. PREDA, G. NEGUT, C. IORGULIS, M. MLADIN Institute for Nuclear Research, Pitesti, Romania 1. INTRODUCTION Romanian TRIGA reactor was commissioned in 1980. The location of the research institute is Pitesti, 100 Km west of Bucharest. In fact there are two independent cores sharing the same pool. There are a 14 MW Steady State Reactor (SSR), high flux, and materials testing reactor and an Annular Core Pulsing Reactor (ACPR). The SSR reactor is a forced convection reactor cooled via a primary circuit with 4 pumps and 3 heat exchangers. The ACPR is natural convection cooled by the pool water. The characteristics of the two reactors are presented in the Tables 1 and 2. The reactor cores configuration is shown in the Figure 1. In Figure 2 is viewed the reactor building. In Figure 3 is viewed the Reactor Hall. In Figure 4 is viewed the reactor core and in Figure 6 the original startup core configuration.

2. TRIGA REACTOR FACILITIES The Romanian TRIGA reactor design (open pool, radial and tangential beam tubes) allowed along the years to be fitted with a lot of testing facilities. There were developed CANDU type fuel complete testing facilities, isotope production, and experimental physics application. In the Table 3 are presented the irradiation and testing facilities, their capabilities and utilization. In the Table 4 there are presented the other experimental capabilities with various applications. 3. SSR TRIGA HEU ORIGINAL CORE CONFIGURATION The reactor core is bolted to a mounting flange at the bottom of the aluminum-lined reactor pool. Sufficient space between the core and curved end of the pool allows for the in-core experimental loops and other equipment. The initial core consisted of 29 fuel bundles, 8 control rods and 44 beryllium reflectors, which is shown in the Figure 6. Of 44 reflectors 20 have a central hole for experiments. Since the total grid plate is a 12 by 11 array, grid plate plugs were installed in all the open holes, so that the coolant flow cannot bypass the fuel bundles. The original core configuration included three in-core experimental regions, two of one-bundle openings, each 8.89-cm square and one of four bundles opening. During the commissioning phase and reactor start-up these openings were filled with experimental devices, which contained CANDU type fuel elements. The reason was to reduce power peaking in adjacent fuel pins to a safe level. The grid plate plugs, fuel bundles and the reflector elements all use the same lower fitting, and therefore the components are interchangeable and can be placed in all grid positions. The 14 MW TRIGA fuel is a special fuel moderator alloy with U235 enrichment. The full behaved very well at fuel temperatures of 7000 C and 8000 C. U-ZrH TRIGA fuel has intrinsic

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properties designed to prevent a nuclear accident in the event of human error or mechanical malfunction. This is the prompt negative temperature coefficient that controls large, prompt reactivity insertions. Thus, any sudden reactivity addition causes an increase in power, which heats the fuelmoderator material instantaneously, causing the number of fission to immediately decrease due to neutron energy spectrum changes within the pins and due to the presence of Erbium in the fuel. 4. FOUR LEU BUNDLES IRRADIATION TEST In February 1992 it started a test consisting in introducing of 4 Low Enriched Uranium (LEU) fuel bundles into the active core. These fuel bundles replaced High Enriched Uranium (HEU) fuel bundles provided for the Romanian SSR TRIGA core. DOE of United States provided these fuel bundles on nonproliferation agreements and core conversion from HEU to LEU. The tests were included in the RERT program, managed by Argonne National Laboratory (ANL). The burnup results are presented in the Table 5. According to the measurements at the end of 1997 fuel burnup was between 20% and 44%, and behavior under irradiation was very good. The main part of the LEU irradiated fuel pins presented a burnup between 7.5 to 9.5 MWday/ fuel pin.

5. MIXED HEU-LEU CONFIGURATION The reason for the mixed HEU-LEU configuration was to mantain the core excess and the operability limit of 0.6$ (defined as minimum reserve to obtain suitable irradiation test flux distributions). At our disposition DOE provided 14 fuel bundles. Finally in the SSR TRIGA core were introduced new LEU bundles and replaced some of the HEU bundles. The final core configuration has now 35 fuel bundles in which 11 LEU bundles are included. The burnup of these fuel bundles is presented in the Table 6. The core refueling provided better core burnup homogeneity. The four HEU most burned fuel bundles were removed and there were replaced with LEU bundles. The latest core configuration is presented in the Figure 7. 6. TRIGA FUEL MANAGEMENT The management of the SSR core has the following goals: -

to obtain a homogenous burned core and a maximization of burnup at unloading; to meet with the requirements for the neutronic flux in irradiation devices; to maximize the fuel burnup at the discharge time; not to exceed the safety limits, especially concerning the hot spots;

The original core configuration at the first start-up was modified. The experimental fuel irradiation devices were: a bundle of equivalent size for the 100 kW loop and of 2 bundles of equivalent size for the capsules. It was important, also, to obtain suitable thermal flux for experimental needs. During the time there were made new experimental channels for isotope irradiation. The fuel behaved very well and there were not a single fuel to leak fission products. The original fuel pins had a longer burnup period that was predicted in the safety report. Anyway at the beginning of the nineties we ned to replace some of the original HEU fuel with the new LEU fuel provided by DOE to meet our needs.

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The new mixed HEU LEU core was computed to obtain the testing and experimental flux needs. 6.1. Core Models and Computer Codes Since 1992 an integrated computer codes system WIMS – DFA for three-dimensional computations of power and flux distribution, reactivity excess and fuel burnup was developed. WIMS computer code was modified to obtain microscopic cross sections on 7 energy groups for 42 nuclides and to obtain macroscopic cross sections for the structural materials. DFA computer code is a development from three-dimensional code 3DDT and it was added with a burnup routine and an identical burnup configuration with that of WIMS cell. Core model has a 73 × 63 × 27 mesh which describes the fuel Be reflectors and the water coolant. Fuel bundles have 25 pins in a 1.633-cm square lattice. Vertically there are 13 meshes on pin. Total number of modeled pins is 875. Burnup is surveyed at the bundle level, averaging the x-y flux in the 5 axial regions on each bundle. During burnup non- homogeneities caused by flux gradients are accumulated, so at two years interval burnup measurements are requested of each pin for future redistribution. From the first startup TRIGA core accumulated 20,000 MWday. Last configuration has 11 LEU bundles in a total of 35 bundles. This configuration is shown in Figure 7. This configuration has the characteristics presented in the Table 7. 6.2. Pin Power Factor Decrease and Fuel Economy Core configuration is not an optimum for the production of leakage fluxes. One of the most important parameters for the core safety is Combined Pin Factor (CPF) which is derived from the formula: CPF = PPF*APF Where: PPF = Power Pin Factor APF = Axial Pin Factor During refueling with LEU bundles CPF tends to overtake the prescribed safety limits, established for LEU fuel pins temperatures. The LEU bundles grouped in the core center cause this. Removing one pin from each cluster, more water will be available in the fuel cluster center, increasing the thermal flux, so the PPF for inner pins are increasing too. This will lead to a more uniform distribution of CPF. There were done computation for the normal fuel bundle and for the fuel bundle with the central pin removed. With the flux distribution obtained there were made thermal hydraulic computation for these two bundle configurations. The computation was done using COBRA code. The results are presented in the Table 8. 7.1. Safety Aspects for Romanian TRIGA 14 MW Research Reactor All research reactors have to fulfill specific safety requirements in operation. These requirements are summarized in AIEA Safety Series (e.g. 35-S2 and 35 G2). Although both research reactors and NPP’s have to meet severe safety criteria, the research reactors are to be considered

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separately, due to there own characteristics. Most research reactors have an obviously small potential for hazard to the public compared with power reactors but may pose a greater potential hazard to operators; also the need for greater flexibility in their use for individual experiments requires a different approach to achieving or managing safety. The main items which have to be considered in research reactor safety are: Safety objectives -Protection on individuals, public and environment against radiological hazards; -Prevention of potential nuclear accidents. Regulatory supervision National Regulatory Body, as an independent governmental organization is able to supervise all stages in the nuclear field activities, it has to: -Establish the safety principles, regulation and guides

-Approve and review the safety documents (operational limits and conditions, emergency plans, etc.). On the basis of documentation approval & review, this regulatory body licenses and associated device operation and the individual operation staff.

the

reactor

Responsibilities for safe operation It is divided at different levels, as follows: Operating organization -ensures the fulfillment of the operational limits and conditions issued in SAR and in the set of the documents for licensing; -prepares and keeps up to date the adequate SAR; -assures the continuous training of the staff; -maintains the associated facilities and services in normal operation; -provides the necessary authority to the reactor manager as to be able to fulfill his duties; -provides operational experience and information from other nuclear facilities, to detect precursor signs adverse to safety, so that corrective actions can be taken in real time. It is desirable that a safety committee to be established in order to advise on the safe operation of the reactor and associated experiments. Reactor Manager -supervises all activities in operation and maintenance of the nuclear devices, and also approves the work schedule and the modifications due to specific experiments; -has to establish (in write) the duties of the staff and communication lines; -is directly involved in training and selecting the operational staff, in health physics; -approves the detailed programme prepared in advance for all operational steps (startup, shutdown, core configuration, maintenance of classical facilities, etc.). Operating personnel -this includes shift supervisors, reactor operators, maintenance personnel and radiation protection personnel. -each person of the operational staff, out of the specific duties, has to be able to shut down the reactor to a safe state in an emergency situation. Safety analysis for operation It is based mainly on the SAR, which has to be updated at each modification stage, and provides the basic information required by the operator.

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An internal reactor advisory group for nuclear safety is formed, with the main role to analyze major aspects of operation or hazards, and to recommend the proper measures to be taken. Operational limits and conditions -Safety limits shall be generally defined in terms of maximum and minimum values. -Operational limits and conditions may include aspects related to administration and organization. Operating procedures They have to include: startup, operation, shutdown, experimental devices, loading/unloading the core fuel, preventive maintenance and periodic inspection, emergency action, security, radioactive waste handling. All procedures shall be consistent with the operational limits and conditions. Maintenance, periodic tests and inspection Shall be conducted to ensure: -compliance with limits and conditions; -adequacy of the safety status of the reactor; Reactor manager has the overall responsibility for these activities. The records are kept accordingly with the QA plan. Core management and fuel handling -it is the strategy used to produce safe operational cores consistent with the needs of experimental programms. -requires computation with validated codes the location of fuel, reflectors, safety devices, and experimental devices. -reactor manager is responsible for the core management and on-site fuel handling, after issuing adequate procedures to the National Regulatory Body. The management of the SSR core has the following goals: -to obtain a homogenous burned core and a maximization of burnup at unloading; -to meet with the requirements for the neutronic flux in irradiation devices; -to maximize the fuel burnup at the discharge time; -not to exceed the safety limits, especially concerning the hot spots. Records and Reports They consist in: logbooks, checklists, automatically recorded data. The operational organization should prepare summary reports. Reactor utilization The main activities in the Romanian TRIGA 14 MW are: -CANDU type and LEU-TRIGA type fuel testing; -Structural materials testing; -Isotope production for medical, industrial and scientific use; -Neutron transmutation doping of silicon ingots, neutronography, neutron diffraction.

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Technical assistance in different stages of CANDU NPP commissioning or for a TRIGA reactor abroad. Modifications Modifications having major safety significance shall be submitted for review and approval by the regulatory body. These modifications will be those which: -involve changes in the approved safety limits and condit ions; -affects items of major importance to safety; -reduce the existing margins of safety; -in the specific case of the Romanian TRIGA 14 MW, the major subject of modification is core conversion from HEU to LEU fuel. First four LEU fuel bundles were tested, at they had a perfect incore behaviour; -the whole strategy of core conversion is based on the computation done with MCNP-WIMS-DFA codes; -no one of the safety limits was exceeded; The National Regulatory Body separately licenses the conversion operation. Radioactive wastes -a special station is in operation especially for waste management at our site (low and medium activity). -spent nuclear fuel is integrally shipped to specialized facility in USA; -high activity waste can be stored for a long period in the hot-cells facilities, near the reactor. Radiation protection, health physics and emergency planning -all activities in this area are in line with regulatory requires; -these activities are periodically overchecked by the Health Department and Civil Defense; -no personnel overdose exposure or uncontrolled release in environment was recorded; -our institute health physicists performed the “zero level” of radioactivity in the CANDU NPP area and trained the personnel for this plant; -the emergency plan is periodically verified by practical exercises. Operators training According to Regulatory Body rules, the operator staff license must be confirmed every two years. In order to obtain this confirmation, all reactor operators should attend specific training course. Security -experts of IAEA performs periodical audit concerning the fissile inventory in the reactor; -the reactor was provided with modern security systems (electric fence, video monitoring systems, IR detectors, magnetic card access system, anti-bullet windows), body control for fire guns. Decommissioning Since the reactor was made critical in 1979, a full decommissioning programme, with adequate procedures, has to be elaborated and it will be included in the final form of SAR. The decommissioning programme will involve the operational organization, the National Regulatory Body, the IAEA, other governmental and local organizations. Quality Assurance All activities in the Romanian TRIGA 14 MW is covered by general provisions contained in QA Manual and specific procedures, in accordance with this Manual and approved by the National Regulatory Body.

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7.2. Conclusions Removal the central pin from the bundle leads to slightly temperature increase of aproximately 1% for the corner and edge pins, for the same pin power density. Also, the temperature slightly decreases for the 4 pins adjacent to the water hole. This is caused by the coolant flow redistribution. But, according to preliminary neutronic computations, PPF-s are decreasing, the edge and corner temperatures changes are no more detectable. DNB are decreasing, leading to a safer operation. Fuel management of TRIGA steady state core allows to obtain the requested fluxes for experimental reasons in the safer operation conditions. We can firmly state that the present operation of the reactor and the HEU-LEU core conversion fully respect the provisions of the National Regulatory Body and IAEA. On the other side, we have to mention the common fact that research reactors cannot sustain themselves in the financial domain. The lack of sufficient financial support leads to shortage of the maintenance programs and to reduce of activities and personnel member; this is a real danger in maintaining the actual standards of nuclear safety. During this transition period, the Romanian TRIGA reactor is used much its capability in the frame of international cooperation this facility can ensure support for various research programmes in the fields of interest.

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Table 1 Steady State Material Test Reactor

Technical specification Nominal Power Number of Fuel Bundles

Measurement units 14 MW Type Number Bundle fuel pin number Bundle dimensions Type

Fuel

Active length Clad diameter Pellet diameter Fuel composition HEU U235 enrichment LEU U235 enrichment Cladding material

Maximum Thermal Flux Maximum Fast Flux

1.0 × 1014 n/cm2 1.0 × 1013 n/cm2 Beryllium square cross section identical with fuel bundle 20 with 3.3 cm central hole Number 24 without hole Hot pressed compacts of boron carbide (B4C) Material with aluminum clad, natural B10 contents. Square cross section identical with fuel bundle Number 8 Control Rod Electrical motor with rack and pinion. Drive Electromagnetic connection with the control rod 10m depth, 5 m width, 9m length, aluminium tank, light water Material

Reflectors

Control Rods

Pool

Mixed HEU LEU 35 25 8.9 cm square cross section Solid, homogenous mixture of erbiumuranium-zirconium hydride alloy 56 cm 1.372 cm 1.27 cm Er-U-Zr1.6 93% 20% Incoloy 800

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Table 2 Annular Core Pulse Reactor

Technical specification

Measurement units

Steady State Power

500 kW Wide steady state power range from mW to MW. Suitable for tests and calibration measurements for very low power levels

Pulse Operation Maximum Pulse Power Minimum Period Pulse Width

Fuel

Control Rods

20000 MW 1.2msec 4.6 msec 1/2 pulse Type Enrichment Cladding material Diameter Cladding diameter Section length Rods number Type Poison material Number

12 wt% U-ZrH fuel 20 wt% 235 U stainless steel with dimples 3.56 cm 3.76 cm OD 38 cm 146+6 fuel followers fuel followered type natural B4 C 6 rack and pinion, electromagnetic connection with the control rod 2 fast transient rods and 1 adjustable transient rod air followered 92% enriched B4 C fast: pneumatic adjustable: rack and pinion drive

Rod drive Number Transients Rods

Type Poison material Rod drive

Maximum Thermal Flux Maximum Fast Flux Pool

1.0 × 1014 n/cm2 1.0 × 1013 n/cm2 10 m depth, 5 m width, 9m length

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Table 3 Irradiation Facilities A. 100 kW Irradiation Loop Technical specification Measurement units Forced Convection Heat Removal Operating Pressure 10.5 to 11.0 MPa Test Pressure 13.5 MPa Operating Temperature 2650 C Preheater Electric Power 100kW Pressurizer Temperature 3140 C Pressurizer Operating 10.5 to 11.0 MPa Pressure Normal Flow Rate 1.5 l /sec demineralized water pH 9.5 to 10.5 Working Fluid O2 < 50 ppb Conductivity < 100µS/cm 3 to 6 fuel samples and/or different structural materials overpower ramp Irradiation Capabilities power ramp B. Fission Gas Capsule Technical specification Measurement units Natural Convection Heat Removal Operating Pressure 8.0 to 10.7 MPa Test Pressure 13.5 MPa Operating Temperature 2650 C Fuel Clad Sample 3240 C Temperature Working Fluid demineralized water 1 fuel sample fission gas analysis Irradiation Capabilities fuel densification power ramp

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C. Fission Gas Pressure Capsule Technical specification Measurement units Natural Convection Heat Removal Operating Pressure 7.3 to 10.7 MPa Test Pressure 13.5 MPa Operating Temperature 2650 C Fuel Clad Sample 3240 C Temperature Working Fluid Demineralized water 1 fuel sample fuel elongation measurement Irradiation Capabilities fission gas pressure measurement power ramp

D. Structural Material Capsule Technical specification Measurement units Natural Convection Heat Removal Operating Pressure 0.1 to 0.5 MPa Maximum Pressure 0.6 MPa Operating Temperature 2650 C Maximum Cladding 2400 C Temperature Working Fluid and Coolant Helium V notch & tensile stainless steel samples zircaloy 4 Irradiation Capabilities cladding samples E. Structural Materials Capsule Technical specification Measurement units Natural Convection Heat Removal Operating Pressure 10.7 MPa Fuel Clad Sample Temperature 3100 C Working Fluid and Coolant Demineralized water zircaloy 2 and zircaloy 4 samples tensile samples Irradiation Capabilities corrosion samples water chemistry control

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F. Power Cycling Capsule Technical specification Natural Convection Heat Removal Operating Pressure Samples Cladding Temperature Working Fluid and Coolant Irradiation Capabilities G. ACPR Capsule Technical specification Hydrostatic capsule Natural Convection Heat Removal Working Fluid Operating Pressure Coolant TemperatureIrradiation Capabilities

Measurement units

10.7 MPa 3250 C Demineralized water 1 fuel sample in power cycling

Measurement units 120 mm I.D. and 12 mm thickness, total volume of 7.5l to fit in the ACPR dry cavity

demineralized water Atmospheric Aprox. 200 C 1 fuel sample in RIA type power pulse

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Table 4 Other facilities A. Focussing High Resolution Neutron Crystal Diffractometer Features: -

use of a disk shaped silicon crystal pneumatically bent in asymmetric reflection as monochromator - 60 mm diameter 400 mm length BF3 detector Capabilities - material structure studies B. ACPR Pneumatic Rabbit System ACPR is provided with a pneumatic rabbit system Samples Dimensions – diameter < 15 mm Length < 100 mm Capabilities – short & medium term irradiation - neutron activation investigation - location thermal flux 1.0x1012 n/cm2 C. SSR Pneumatic Transfer System Location

-thermal column 90 graphite blocks near reflector -Beryllium reflector - thermal flux 2.1x1013 n/cm2 -3 m distance from beryllium reflector air dry cavity with 110 mm diameter thermal flux 2.85x109 n/cm2 - fast flux 1.99x107 n/cm2 Samples Dimensions – diameter - 10 mm length - 19 mm Capabilities -short & medium term irradiation -neutron activation investigation

D. Neutron Radiography Facility Type – underwater in the pool situated near ACPR Technical features – L/D = 250 - Thermal flux in the probe plan – 1.0x107 n/cm2 - Transfer method - In, Dy foils Capabilities – nuclear fuel investigation

E. Gamma Irradiation Station Gamma Source – Cobalt 60 Total Activity –30000Ci

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Technical Features:

Dosimeters:

- cylindrical cavity 300mm diameter, 300-mm height with a maximum dose of 4.5kGy/h - Square cavity 600x600x900 mm with maximum dose rate of 1.4kGy/h - Dry channels system with two channels 150mm diameter, 5 channels 100-mm diameter with dose rate of 2.0 to 10.0 kGy/h

- Va-J-18 dosimeter measurement range 30 to 33300 R/min - Fe sulfate dosimeter measurement range 40 to 400 Gy - Ethanol Chloride Benzene measurement range 2-60 kGy

F. Radioisotope production facility Sealed sources for industrial use Isotopes for medical use and pharmaceutical use Ir192, Co60, Te204, I125, Mo99+Tc99, I131, Au198, Cr51 Re186, Ce141, Yb1609, W185, Cs134, S35, K42, Zn65

G. National thermal flux standard and sigma-sigma system -

Spherical graphite cavity of 0.50 m diameter (1 m by 1.2 m) Derived sigma-sigma system Sigma-sigma handling system Calibrated fission chambers NIM measurement chain for fission chambers signal processing

H. Neutron Transmutation Doping System -

Silicon ingots up to 3” diameter

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Table 5 LEU Fuel Bundles Irradiation Results, 1998 CRT. Number 1 2 3 4

Bundle Serial Number L-38 L-39 L-40 L-42

Reactor Operation Duration (Hours) 24,775 24,775 24,775 24,775

Burnup (MWday) 315.0 361.0 233.0 302.0

Table 6 Mixed Core LEU Fuel Bundles Irradiation Results, 1998 CRT. Number 1 2 3 4 5 6 7 8 9 10 11

Bundle Serial Number L-38 L-39 L-40 L-42 L-44 L-46 L-47 L-02 L-32 L-45 L-49

Reactor Operation Duration (Hours) 24,775 24,775 24,775 24,775 12,469 12,469 12,469 5,462 5,462 5,462 5,462

Burnup (MWday) 315.0 361.0 233.0 302.0 135.0 80.0 128.0 11.0 0.0 15.0 9.0

Table 7 Nuclear HEU LEU Mixed Core Characteristics Nuclear Core Characteristics Control Rods Bank Efficiency Cold, unpoisoned, with water in experimental channels excess Core without Be blocks excess Core without instrumented pins excess

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Value 12.7 $ 4.60$ 4.53$ 3.7$

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Table 8 PPF and Temperature Distribution for a Fuel Bundle in Both Configuration at 14 MW Power Level Table 8.1 PPF for the fuel in normal 5 × 5 bundle

2.25 1.85 1.74 1.72 1.81

1.79 1.58 1.53 1.53 1.61

1.65 1.48 1.45 1.44 1.53

1.62 1.47 1.43 1.42 1.51

1.99 1.86 1.80 1.80 1.87

2.17 1.80 1.72 1.67 1.75

Table 8.3 Normal Fuel Temperatures (ºC),

590.7 526.7 511.3 511.2 535.3

547.7 495.9 486.7 483.6 510.7

538.4 492.8 480.5 477.4 504.6

1.74 1.59 1.75 1.54 1.57

1.63 1.71 0.00 1.66 1.51

1.58 1.48 1.65 1.44 1.47

1.91 1.80 1.77 1.75 1.80

531.8 500.8 553.6 556.6 497.6

634.4 600.2 590.8 584.6 600.2

Table 8.4 Modified Fuel Temperatures(ºC), gap=1.27E-3 cm, 14 MW, Two pumps

gap=1.27E-3 cm, 14 MW, Two pumps 731.6 609.2 575.4 569.2 596.4

Table 8.2 PPF for the fuel in the 5× 5 modified bundle

644.4 612.2 593.8 593.8 614.9

715.2 600.2 575.3 559.8 584.7

Toutlet = 46.57ºC; ?p = 6.61E+4 Pa min. DNBR= 5.48

581.6 535.0 584.7 519.4 528.7

547.3 572.2 556.6 510.0

Toutlet =45.55 ºC; ?p=6.43E+4 Pa min. DNBR=5.72.

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14 MW STEADY-STATE REACTOR CONTROL ROD (8) ANNULAR CORE PULSING REACTOR

FUEL REGION EXPERIMENT REGION

CENTRAL DRY CAVITY

BERYLLIUM SOLID REFLECTOR PLUG OFFSET LOADING TUBE

REFLECTOR EXPERIMENT POSITIONS

GAMMA SHUTTER REACTOR TANK

GAMMA SHIELD

CAVITY REACTOR SHIELD RADIAL BEAM TUBE TANGENTIAL BEAM TUBE REMOVABLE TANGENTIAL BEAM TUBE

Figure 1 Dual core TRIGA ROMANIA 17

RADIAL BEAM TUBE

Figure 2 View of the TRIGA Reactor Building

Figure 3 Reactor Hall

Figure 4 View of the 14 MW TRIGA core

A B C D E

F G H

I

J

K

1 2

XT-1 XT-2 XT-3 XT-4 XT-5 XT-6

3

R-5

4

R-6

XS-1

XC-2

XS-2

XL-2

5

R-1

R-3 XS-3

6

XS-4 XL-1

7

XC-1

8

R-2

XS-5 R-4 XS-6

XL-3

9

XC-3

10

XS-7 R-7

11

R-8

XS-8

XB-1 XB-2 XB-3 XB-4 XB-5 XB-6

12 REFERENCE DESIGN FUEL BUNDLE LOCATIONS 1

2

3

4

5

6

7

8

9 10

REFERNCE DESIGN BERYILLIUM REFLECTOR BLOCK LOCATIONS

11 12 13 14 15 16 17 18 19 20 21 22 23 24 25

XC - IN-CORE EXPERIMENT XL - LARGE EXPERIMENT

XTXB XS -

BERYLLIUM REFLECTOR BLOCK WITH EXPERIMENTAL HOLE

R - CONTROL ROD LOCATION

PLUG Figure 6 Original startup 1980 14 MW TRIGA core configuration

A B C D E F G H

I

J K

1 2 3

H11 R-5 H3 H22 H14

4

H16

5

H13 H33 L45 R-3 L32 R-1

6

L39 L44 L46 L42

H37 H28 H30 H18

7

H4 H36 L38

L47 L40

8

H31 L2

R-2 L49 R-4

9

H27 H1 H34 H21 H25 H17

10

H15 R-7 H26 H12 H6

11 12 HEU FUEL

CONTROL ROD

LEU FUEL

REFLECTOR

EXPERIMENTAL LOCATION

Figure 7.

PLUG

Present 14 MW 35 bundles core configuration