Fuel cladding integrity

The safety barrier function is expected to be lost if thermomechanical analysis has demonstrated a loss of cladding integrity. 

Fuel pellet Temperatnre 2800°C Volume averaged fnel enthalpy 10 kJ/kg 

Fuel cladding Maximnm temperatnre 1200°C for DBAs Maximnm local cladding oxidation 8% Maximnm core-wide hydrogen generation 1% 

Fuel channel tube Nominal wall temperatnre 50°C at 13.4 MPa Maximnm wall temperatnre 650°C at 4.0-8.0 MPa 

Reactor cavity Excessive pressure 210 kPa for a DBA 300 kPa for a BDBA with MPTR 

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Experiments were carried out on power to measure the fuel SA outlet temperature with CCPM stuck at 80 mm position and a temperature attenuation of 7% (average) was found in Mark I SA. However, this attenuation is large for Mark II SA where sodium flow is less. Studies were conducted to find out the probability of plugging during reactor operation and found to be acceptable. 3D analysis of outlet plenum thermal hydraulic was carried out to establish the level of plugging that can be detected viz., allowable plugging for fuel clad integrity. 

Any fuel assembly with a leaking rod is promptly detected by the system for fuel cladding integrity monitoring and is immediately removed from... 

The safe operation of a reactor necessitates that the fuel be kept adequately cool at all times to prevent loss of fuel cladding integrity and the consequent dispersion of radioactive species into the coolant. The safety systems that prevent or mitigate fuel damage are described below. 

Water chemistry shall be controlled in order to guarantee the spent fuel cladding integrity throughout the planned periods of wet storage. 

Limits for the permissible specific activity of the reactor coolant should be stated in order to ensure the protection of persormel and the environment as well as to provide a measure of fuel integrity, as discussed in the safety analysis report. If online measurement of coolant activity is used to monitor the fuel cladding integrity in operation, the minimum provisions for the detection and, where appropriate, identification of failed or suspect fuel should be stated. 

SCHWARTZ, M.W., WITTE M.C., Spent Fuel Cladding Integrity Durii Dry Storage, Rep. UCID-21181, Lawrence Livermore Laboratory, Berkeley, CA (Sep. 1987). 

The fuel clad integrity requirement would be met if WSRC can demonstrate that no fuel damage precursor conditions occur. Both Action Plans indicate that thermal hydraulic limits (FI and ECS), which preclude reaching fuel damage precursors, will be applied to all DBEs. Maintenance of fuel clad integrity is an open item. 

The core should remain subcriticaT and/oir amenable to cooling considering calculated changes in 0re geometry.. The/.stiff considers that if the fuel-clad integrity requirement is met 7this Requirement would also be met. The loss of fuel accident postulated by. WSRC will, require special consideration. 

The cladding temperature that was obtained by the three-dimensional coupled core calculation is the average temperature over the assembly. The peak cladding temperature of a fuel rod is necessary for the evaluation of the fuel cladding integrity. The subchannel analysis code of the Super LWR is coupled with the fuel assembly bum-up calculation code for this purpose [25]. Fuel pin-wise power distributions are produced for various bum-ups, coolant densities, and control rod positions. The pin-wise power distributions are combined with the homogenized fuel assembly power distribution to reconstmct the pin-wise power distribution of the core fuel assembly. The power distribution over the fuel assembly is taken into account as shown in Fig. 1.11. The reconstracted pin-wise power distribution is used in the evaluation of peak cladding temperature with the subchaimel analysis. 

Surveillance measures that should be taken to verify the integrity of fuel cladding include, but are not necessarily limited to ... 

For the integrity of fuel claddings to be confirmed, it is therefore essential to ensure that the following maximum values of fuel rod parameters are not exceeded ... 

The preserved integrity of fuel claddings and comphance with the dose criteria are indicative of an acceptable response of the plant to fuel handling accidents. 

Defence in depth requires the provision of multiple barriers to the release of hazardous fission by-products from the fuel. In general, these barrios include the fuel matrix itself (UO2), the fuel cladding, the primary heat transport circuit boundary and a surrounding containment structure. Autonomous reactor operation for a specified time interval requires that ail contairunent barriers are initially intact and that no events are foreseen that could compromise barrier integrity within this duration. 

The fuel cladding, in conjunction with other plant systems, is designed to retain integrity so that the consequences of any failures are within acceptable limits throughout the range of normal operational conditions and abnormal operational transients for the design life of the fuel. 

Besmann, Th.M., Lindemer, T. B. Chemical thermodynamics of the system Cs-U-Zr-H-l-O in the light-water reactor fuel-cladding gap. Nucl. Technology 40, 297-305 (1978) Burman, D. L. Methods for estimating numbers of failed rods from coolant activity analysis. Proc. Workshop Fuel Integrity Monitoring by Coolant Activity Analysis, Charlotte, N. C USA, 1986... 

Reactor core cooling (typically protects integrity of the fuel cladding or... 

The remaining concepts not yet discussed often employ significant innovation and, therefore, they may require multiple years of R D for technology development prior to a prototype integral demonstration. In general, if new fuel/cladding/coolant combinations are employed, the deployment times can be no sooner than 15-25 years. 

The fuel lifetime and the period between refuellings are both about 1800 effective full power days (EFPD), or more than 5 years. From the viewpoint of the neutronics, a 10-year (about 3650 EFPD) refuelling interval is possible but the integrity of the fuel cladding is not confirmed for more than 5 years of operation in the conditions similar to those of a PWR core. 

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