Thermal-hydraulic transients of the primary system

This appendix details a simple calculation program that allows the rough evaluation of transients and accidents in the primary system of a PWR. It can however be adapted to other types of water reactors. 

As noted at the beginning of Appendix 2, here also, for historical reasons, some units of measurement are not those of the Standard International System. 

This program has been useful in preliminary sizing safety systems during the design phase and in the quick verification of them during safety reviews. 

This program was first developed (Petrangeli, 1983 Petrangeli et al., 1993) for the study of a new safety system (the CRS described in Appendix 10) based on the voluntary depressurization of the primary system and on the passive injection (by accumulators under pressure) of cooling water. This basic concept has been subsequently applied to various reactor designs. 

Calculation tools of this kind are very useful to the designer or to the overall system analyst (even if they leave the true specialists of the branch rather puzzled), as they allow the study of many cases and for transient times as long as are desired. It has been observed, with reference to the Three MUe Island accident, that if the time length of the calculated transients had been prolonged beyond the intervention time of the safety systems, the adopted thermal-hydraulic codes (RELAP and so on) could have shown the danger of getting to a situation where the pressurizer is substantially full of liquid while the reactor vessel is nearly empty. As it is known, this situation may cause the operators to erroneously think that all of the primary system is full and therefore make them shut off the safety injection systems. In fact, the calculations performed were stopped precisely at the moment of their intervention. 

---

Appendix 11 Thermal-hydraulic transients of the primary system 367... 

The SSC-K code [4] has been developed by KAERI for the analysis of system behaviour during transients. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. 

These design objectives were carried over to the work on the power reactor PIUS, basically a pressurized water reactor (PWR) in which the primary system has been rearranged in order to accomplish an efficient protection of the reactor core and the nuclear fuel by means of thermal-hydraulic characteristics, in combination with inherent and passive features, without reliance on operator intervention or proper functioning of any mechanical or electrical equipment. Together with wide operating margins, this should make the plant design and its function, in normal operation as well as in transient and accident situations, much more easily understood and with less requirements on the capabilities and qualifications of the operators. 

Passive core make-up water tank is used so as to eliminate high pressure safety injection pumps. Transient characteristic research of core make-up water tank in the case of small LOCA were performed. The break sizes are 2, 6, 12, 18, 30 mm respectively. Thermal hydraulic behaviors of pressurizer and draining flow rate measurement from core make-up water tank to primary coolant system following small LOCA were researched and carried out in the test. A total of 80 sensors were used to measure temperatures, liquid level and flow rate. 

It has to be remembered that boric acid may precipitate from the solution as various kinds of deposits (crud) which form on the inside primary system surfaces and especially on the hot surfaces of the fuel elements. Subsequently, in case of thermal or hydraulic transients, some of these deposits may peel off from the core giving rise to a reactivity transient. Over the years, no accidents due to this phenomenon have happened, notwithstanding the fact that the boron deposition on core surfaces has been observed and studied. The maximum reactivity which could be released can be evaluated of the order of 0.1 per cent in half a second (Petrangeli, 1967). 

---