Startup range neutron monitor (SRNM) and power range monitor (PRM) systems are separately used from startup to the full power condition in nuclear measurement of BWRs. Since neutron flux level varies by about nine orders of magnitude from startup to full power, two different types of monitors are employed corresponding to each detectable neutron flux level. Both types of monitors are installed in the core as shown in Fig. 3.3. Fission ionization chamber type monitors are used in the SRNM system as a fixed arrangement type one. In a 1,300 MWe-class ABWR plant, the PRM system consists of 208 local power range monitors (LPRMs) and traversing in-core probes (TIPs) for LPRM calibration. LPRMs are inserted into the neutron flux instrumenta­tion tubes in the gap between channel boxes at the corner of the fuel assemblies and four LPRMs are axially installed in one instrumentation tube. LPRMs are components of the average power range monitor (APRM) system and they average the power signals as well as detect the usual neutron flux distribution in the core at normal operation [3].

One PRM or LPRM is usually used for every 16 fuel assemblies as shown in Fig. 3.3 and it corresponds to one for four fuel assemblies in considering a quarter-symmetric core. LPRM measurements are spatially discrete informa­tion which is not intended to directly measure the fuel assembly and fuel rod power. The assembly power distribution in the entire core is evaluated in combination with the relation between measured values and adjacent fuel assembly powers and the 3D computational model. Operation parameters such as maximum linear heat generation rate and MCPR are regularly moni­tored. Such power monitoring is implemented by a process computer referred to as core performance calculation system. It is necessary to prepare input constants of the process computer for the power distribution calculation using LPRM measurements before reactor startup; this is an important core management task.

Because of recent improvements in computer performance, a 3D nuclear and thermal-hydraulic coupled calculation model in a core design code is generally used in the core performance calculation of the process computer. Results of the core performance calculations are corrected by measurements from LPRMs and TIPs in each location to more accurately monitor and predict the core power distribution. The variation in burnup and fuel materials of each fuel assembly is also calculated in the process computer.

Alternatively, follow-up calculations of reactor operation are usually performed every month by off-line computers in which the same core design codes are installed. In the computers, calculations from the core design and measurements during operation, and results from the follow-up calculations and operation data are compared to evaluate and confirm the reproducibility of core design codes in the actual operation. The off-line computers are also used to calculate predicted operation in the next cycle and if necessary to re-investigate the control rod patterns. Thus, highly accurate and quick opera­tion management based on the real operation data is possible.