Cooling after shutdown

6 Cooling After Shutdown. After - an operating, reactor is shut, down, the total heat liberated due to.the-absorption of gamma and beta radiation from the fission products is given-by the empirical equation  

A = Reduction factor a reasonable, value is 0.08, which is an extrapolation to t = 0 from experimental measurements of energy, after shutdown (also, since at s)hutdown 14.7 Mev/fission of beta and gamma heat is released-comparedrto 187 Mev/fission.during operation, -the ratio of the two heftt values.is 0.08(i.e., 14.7/187).  

REACTION COSCEN- T TION (ppm) HALF- LIFE CROSS- SECTION RECOIL TYPE RANGE (cm) ACTIVITY (dis/cc sec)  

Mlnlmam Process water Flow After Shutdown 

The water temperatures given in the following table are based on this equation, a pile operating period of 15 days, and a 24,000-gal content of the circulating system. 

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The problem of cooling after shutdown has been extensively studied, and a full report has been written.After shutdown the power level is given by the formula developed by Way and Wigner ... 

Huffman, J. R., Cooling After Shutdown, ORNL CF-48-6-1S5, June 7, 1948. 

Steam. Steam connections to process resemble ordinary water connections. Pressure usually is of more concern than contamination, and connections to process systems without adequate design pressures must include some form of pressure relief. When steam is used to purge or humidify a process system, cooling after shutdown also can produce a vacuum. Protection is then required directly on the process system. 

Fuel cooling last at least 1 hr. after shutdown 4.7E-4... 

Moderate Water hose or piping Cooling char bed with primitive stream-water spray at supposedly safe interval after shutdown... 

Provision of emergency cooling systems for reactors where heat continues to be generated after shutdown, for instance, in some polymerization systems. 

Cooling Reqnirements at Reduced Power Levels and After Shutdown. 

The PDHR system is employed for both the hot stand-by and long-term core cooling modes. This system can operate at full reactor coolant system pressure and places the reactor in the long-term cooling mode immediately after shutdown. 

Consequently, the reactor must be cooled even after shutdown and even in case of an accident. Therefore several independent cooling systems are incorporated. The worst accident that is considered (maximum credible accident MCA ) is the complete shearing off of the primary circuit line. If this should occur, then a mixture of steam and air would be blown into the containment for a time span of 12 s (blowdown phase). When the pressure falls below 26 bar, water tanks maintained at this pressure automatically would empty into the reactor vessel from both sides (high and low temperature) of the circuit. If this water should evaporate or leak out again, at a pressure of 10 bar, a pump would provide water from other reservoirs. Water leaking out from the primary circuit will have to accumulate on the bottom of the... 

The residual heat removal system (RHRS) is used for controlling the water temperatures in both the RPV and CV for long term cooling after LOCA and is also used for long term decay heat removal during scheduled reactor shutdown... 

After a reactor shutdown harmful natural circulation of hot helium through the primary circuit is prevented, thanks to the thermohydraulic decoupling of heat source and heat sink. Consequently, there is no need to cool the steam generator after shutdown. Thus it can be shut off and left in a hot condition. 

Cool down time (hours after shutdown) 96 ... 

The cool, low-pressure moderator removes 4.5% of the fuel heat dining normal operation about the same as the amount of decay heat removed shortly after shutdown. It can therefore act as a long-term emergency heat sink for a LOCA plus LOECC the heat transfer is effective enough to prevent melhng of the UOjfuel and preserve channel integrity. 

HWRs have a shutdown cooling system that can remove decay heat after shutdown from full pressure and temperature conditions. It is not necessary to depressurize the HTS. 

Where a separate system is required for cooling the core after shutdown, an adequate and reliable system, in addition to the primary cooling system, shall be provided for the removal of residual heat. 

In order to go beyond such findings it is necessary to be able to describe the development of the fault up to cooling after the scram. This study must take into account the actual characteristics of the reactor plant, including systems for accident detection and shutdown. The Project code SURPASS [5.15] mentioned above provides for such a study and the demonstration of the enveloping character of the T.I.B. The parameters in such calculations are adjusted according to available test data and in particular to the Scarabee tests (not only the BE+ series but also the APL Tests [5.19]). 

As mentioned previously, the water content in the catalyst layer will be determined by the extent of drying of the fuel cell before freezing, so the various literature studies may have widely different starting water contents in the catalyst layer. Furthermore, because the saturated vapor pressure curve for water is a strong function of temperature, even cells that operate at undersaturated conditions can exhibit condensation and freezing as the temperature drops after shutdown. It is expected that liquid-water and ice contents will be highest at the periphery of the cell, as the cell should cool first at the outer perimeter, and water will be transported to the coldest regions, condense, and subsequently freeze. 

A reactor failure that results essentially in the confinement of air within a waterless reactor vessel will cause fuel elements to melt, even if the vessel walls are cooled and maintained at 60F by some means-This is because the shutdown heat production of the core cannot be transferred to the inner surface of the reactor vessel by air unless the air temperature is more than 2000F above the temperatvire of the vessel wall even after the first two hours after shutdown. Radiation cooling of the fuel elements by the vessel wall is limited to those elements at the outer radius of the core. The high flux elements near the center are hidden from radiation cooling except for a small section at their ends. 

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