ORNL-5388 - the Molten Salt Energy Technologies Web Site

5-10
The fuel performance program under LWR data-base development would consist of the
establ ishment of safety-related fuel performance information such as transient fuel damage
limits, thermal performance both for normal operation and with respect to LOCA* margins on
stored heat, dimensional stability (densification and swelling), gas absorption and release
behavior, and fuel cladding interaction. The initial phase of this program should consist
of in-reactor properties experiments, power ramp tests, transient fuel damage tests, and
fuel rod irradiations. The in-reactor properties experiments would be similar to the
program currently underway in Norway's Halden HWR and would be designed to provide information
on such parameters as center-line temperature, swelling and densification, and fissiongas
release during operation. The power ramp experiments would consist of preirradiation
of the fuel rod segments in existing LWRs and the subsequent power ramping of these segments
in special test reactors to establish anticipated fuel performance during power changes
typically encountered in the operation of LWRs. Examples of such programs are the international
inter-ramp and over-ramp programs currently being undertaken at Studsvi k. The
transient fuel damage experiments would be designed to provide information on the performance
of the denatured fuels under the more rapid transients possible during operation and in
postulated accidents. Lastly, the fuel rod irradiation experiments would provide information
on the irradiation performance of prototypical thorium-based fuel rods, and, with
subsequent post-irradiation isotopic analyses, would also provide information on burnup
and plutonium production. (As noted previously, the fuel performance program costs are
included, though not specifically delineated, under the fuel cycle R,D&D discussed in
Section 5.2.)
In addition to the data base development, some as yet unidentified reactor components
development could be expected. To cover this aspect of the program, an estimated cost of
$5 - $25 million is included in Table 5.1-1.
The remaining fuel-cycle-related R&D for LWRs would be devoted to developing core
design changes and safety analysis information in preparation for a reactor/fuel cycle
demonstration. In this phase of the program, safety-related behavior of alternate fuel
would be determined using the specific design attributes of the demonstration reactor.
The effects of alternate fuel cycles on plant safety and licensing would require examination
of safety criteria and the dynamic analyses of design basis events. Appropriate
safety criteria, such as acceptable fuel design limits and limits on maximum energy deposition
in the fuel, would have to be determined. Changes in core physics parameters that
result from alternate fuel loadings and the implication of these changes on reactor design
and safety would also have to be identified and accommodated within the design. For
example, changes in fuel and moderator temperature reactivity coefficients, boron worth,
control-rod worth, prompt-neutron lifetime and delayed-neutron fraction must be addressed
since they can have a large impact on the performance and safety of the system. The effects
of alternate fuel cycles on the dynamic system responses should be determined for all
transients required by Regulatory Guide 1.70, Revision
determine the implications of denatured fuel cycles on
performance to determine whether the response of plant
*LOCA = Loss-of-Coolant Accident.
2. It would also be necessary to
plant operation and load change
control and protection systems is