Structural and containment response to LMFBR accidents

The adequacy of the containment of fast reactors has been traditionally evaluated by analyzing the response of the containment to a spectrum of core disruptive accidents. The current approach in the U.S. is to consider fast reactor response to accidents in terms of four lines of assurance (LOAs). Thus, LOA-1 is to prevent accidents, LOA-2 is to limit core damage, LOA-3 is to control accident progression and LOA-4 is to attenuate radiological consequences. Thus, the programs on the adequacy of containment response fall into LOA-3. Significant programs to evaluate the response of the containment to core disruptive accidents and, thereby, to assure control of accident progression are in progress. These include evaluating the mechanical response of the primary system to core disruptive accidents and evaluating the thermal response of the reactor structures to core melting, including the effects this causes on the secondary containment. The analysis of structural response employs calculated pressure-volume-time loading functions. The results of the analyses establish the response of the containment to the prescribed loadings. The analysis of thermal response requires an assessment of the distribution and state of the fuel, fission products and activated materials from accident initiation to final disposition in a stable configuration

Review of effluent disposal practices in N Reactor Department

A survey has been conducted of current methods of disposal of radioactive, chemical, and sanitary wastes used both at the 100 Area and 300 Area sites of N Reactor Department Operations. In addition, liquid storage facilities have been surveyed for situations which might result in river water pollution. The survey and this report have been prepared in response to the request of the Manager, Richland Operations Office of the Atomic Energy Commission in accordance with Executive Order 11258. An audit of N Reactor Department waste disposal procedures and practices was recently made. The audit report provides detailed data on effluent streams, methods,d and sampling points. Therefore, this report does not include that information and instead provides a summary of experimental and analytical data which have become available since the audit. It also includes information developed in response to specific provisions set forth in the Executive Order

Safety analysis of experiments proposed for irradiation in FFTF

A safety analysis of all experiments to be irradiated in FFTF is required to demonstrate that the experiment does not constitute a hazard to the reactor or the public. This paper summarizes the review procedure, the scope of the analysis and safety acceptance criteria

Response of FFTF core to protected reactivity addition transients

The response of the FFTF core to protected reactivity insertion events was evaluated. Reactivity addition transients ranging from .05 cents/s to 3$/s have been considered. The evaluation method is based on a calculational model which predicts cladding strain from modified fuel-cladding differential thermal expansion. The results show that for all ramp rates considered, the Plant Protection System (PPS) controls consequences to required limits. Comparisons made between predicted fuel damage and results of TREAT transient tests support the conservatism of the results

FFTF containment of hypothetical accidents

The FFTF facility was evaluated for the consequences of an HCDA followed by failure of in-vessel post-accident heat removal, reactor vessel melt-through, and release of core debris and sodium coolant to the reactor cavity. Two cases are presented based on parameters considered to represent upper limits for rates of chemical and thermal attack of the reactor cavity concrete containment structure. The reactor containment building temperature, pressure, and leak rate histories were computed with the CACECO code which provided input into the HAA-3C code for prediction of aerosol behavior, and to the COMRADEX-H code for prediction of radioactivity dispersion. The resultant 30-day doses at the site boundary were judged to be acceptable considering the conservatism in the analysis and the low probability of the event

Chernobyl lessons learned review of N Reactor

A broad-base review of the N Reactor plant, design characteristics, administrative controls and responses unique to upset conditions has been completed. The review was keyed to Nuclear Regulatory Commission (NRC)-defined issues associated with the Chernobyl accident. Physical features of N Reactor that preclude an accident like Chernobyl include: lack of autocatalytic reactivity insertion (i.e., negative coolant void and power coefficents) and two separate, fast-acting scram systems. Administrative controls in place at N Reactor would effectively protect against the operator errors and safety violations that set up the Chernobyl accident. Several items were identified where further near-term action is appropriate to ensure effectiveness of existing safety features: Resolve a question concerning the exact point at which Emergency Core Cooling System (ECCS) activation by manual actions should be implemented or deferred if automatic ECCS trip fails. Ensure appropriate revision of the Emergency Response Guides and full communication of the correct procedure to all Operations, Safety and cognizant Technology staff. Train reactor operators in the currently recognized significance of the Graphite and Shield Cooling System (GSCS) in severe accident situations and cover this appropriately in the Emergency Response Guides. Complete reviews which establish an independent verification that pressure tube rupture will not propagate to other tubes. 15 refs., 3 tabs