The nuclear design code system is validated by the experimental data of critical assemblies such as the VHTRC [42]. The VHTRC was constructed to study the nuclear characteristics of pin-in-block type HTGRs. The vertical cross section of

Steel frame

Graphite reflector

Heating wire

Fuel rod insertion hole

• Fuel rod A Safety rod ■ Control rod ® Neutron source

Fig. 4.32 Cross section of very high temperature reactor critical assembly (VHTRC)

the VHTRC is shown in Fig. 4.32. The VHTRC consists of two half assemblies. Each half assembly is formed by piling up the horizontal hexagonal graphite blocks and by fixing them into position by the steel frame. Each fuel block is equipped with the holes for loading fuel rods and inserting a control rod or a safety rod. The layout and loading number of the fuel rods can be easily changed. One of the half assemblies is fixed and the other is movable. Each fuel rod is formed by loading the fuel compacts into the graphite sleeve. Heating wires can be inserted into the graphite blocks so that experiments at desired temperatures can be done.

The major specifications of the VHTRC and the HTTR are compared in Table 4.8. The experimental data of the VHTRC, such as effective multiplica­tion factor, control rod worth, burnable poison worth, power distribution (reaction rate distribution of copper) and temperature coefficients, were used for validation the nuclear design code system [43, 44].

(1) Effective multiplication factor

The effective multiplication factor is measured by the critical approach experiment where the fuel rods are sequentially loaded into the core. The effective multiplication factor for the ideal condition is calculated by correcting the measured one considering the effects of insertions and temperature. This ideal value is used for validation. The reactivity worth of insertions such as control rods, detectors, etc. is measured by the period method or the pulsed neutron technique.

Since the error between the experimental data and the predicted values was within 1 %Ak, the calculation error used for the nuclear design of the HTTR was determined as 1 %Ak.

(2) Control rod worth

The control rod worth is measured by inserting the mock-up control rod into the critical core and then measuring the subcriticality by the pulsed neutron technique. The error between the measured and calculated values was evaluated as 2.6 %. The calculation error was conservatively deter­mined as 10 %.

The burnable poison worth was measured as well. Although the nuclear design code system accurately predicted the measured value, the calcula­tion error was conservatively determined as 10 %.

(3) Power distribution

Using the proportional relationship among reaction rate of copper, neutron flux and fission rate, the calculations of the neutron flux distribution and the power distribution were validated by the measured reaction rate distribu­tion of copper. In the measurement, copper foils were horizontally and axially inserted into the core and criticality was kept for a fixed period. The reaction rate of copper was obtained from the activation rates of the irradiated copper foils.

The calculated and measured axial reaction rate distributions of copper are shown in Fig. 4.33. The distribution is normalized so that the average reaction rate is 1.0. From that figure, the calculated distribution agrees well with the measured one in the fuel region. Since the errors between the calculation and the measurement were within 3 % for both radial and axial directions, the calculation error was determined as 3 %.

(4) Temperature reactivity coefficients [45]

The core was kept critical at room temperature and then the temperature was elevated using the heating wires. The subcriticality at the elevated temperature was measured by the pulsed neutron technique. From the relation between the temperature elevation and the decrease in the reactiv­ity, the temperature reactivity coefficients were obtained. The measure­ments were carried out for temperatures to about 200 °C. The nuclear design code system well reproduced the measured values. The error was 6 % at maximum. The calculation error of the moderator temperature coefficient and the Doppler coefficient for the reactor transient analyses were determined as 10 %.

Fig. 4.33 Comparison of measured and calculated axial copper reaction rate distributions

Fuel region Graphite reflector

Distance from core center [cm]

Based on those validations, the expected calculation errors in the nuclear design were determined. They are summarized in Table 4.9 with the errors used for core design [43].