**4. Summary**

The implementation of HELIOS as a design tool for core reload and safety analysis of the ATR is one of the first examples of using whole-core transport codes in such a capacity. Traditionally, codes such as HELIOS, are used for cross-section condensation for use with nodal diffusion codes. Since HELIOS is an arbitrary geometry code, this suits it well for creating cross-section datasets for much faster nodal diffusion codes that then analyze CANDU, RBMK, or VVER reactors. Generally, HELIOS would be an ideal code to support ATR fuel reloading analysis because the ATR was designed with minimum axial perturbation in mind; hence, the use of control drums over control rods. However, with the promise of higher solution fidelity has come more complex experiment designs. Since the KJRR-FAI cycles, more geometrically complex, high fissile worth, and/or high neutron absorber tests have been irradiated in ATR. The challenges of such experiments are as follows: finding fuel elements capable of providing the best irradiation conditions for all the customers of the ATR National Scientific User Facility (NSUF). In the case of high worth tests, this requires selecting fuel elements near the end of their life, but not so spent that they exceed their burnup limit by time the requested cycle-length has been reached.

Assuming this is possible, sufficient fuel must be loaded in the core to have sufficient excess reactivity to accomplish this cycle-length. However, this is limited by the amount of shut-down margin available in the OSCCs. If the loading is too rich, startup could occur in the non-linear range of the OSCC reactivity worth curve, risking a missed startup prediction. Not discussed here, but if a lobe is designed for much higher power operation, i.e., the core power is closer to the maximum rating of 250 MW, the margins to thermal-hydraulic safety limits can be challenged.

Finally, each time the chopped cosine assumption is challenged by such axial heterogeneity, the axial profile is calculated with MCNP, then measured in the ATRC, and ultimately triggers an update to the bounding thermal-hydraulic analysis using the new axial profile. Indeed, this was the case for the KJRR-FAI cycles. A new thermal-hydraulic limit for the ATR fuel element was derived such that the ATR's flow instability, departure from nucleate boiling ratio, and other safety limits under transient conditions would not be challenged by the KJRR-FAI's alternative axial profile.

The irradiation of the KJRR-FAI has essentially demonstrated that advanced codes can support advanced hardware. However, there is a tendency to believe that advanced codes can change the operating envelope of a nuclear reactor. The HELIOS and KJRR-FAI experience shows that the operating envelope is set by margins to the safety limits and that these margins are established by measurements. In the case of KJRR-FAI, these measurements were careful fission-wire measurements of the axial shape in the ATRC. The KJRRR-FAI test was a great success and required a great amount of teamwork among physicist and code developers that did the HELIOS code verification and validations, the reactor engineers who did the fuel reloading analysis, ATR plant operations who supported the fission wire measurements, and safety analysts who could understand the historical analysis with PDQ and connect those assumptions with modern application.
