DOE 2000. - Waste Isolation Pilot Plant - U.S. Department of Energy

WIPP RH PSAR DOE/WIPP-03-3174 CHAPTER 5
5.1.2.1.2 RH Radionuclide Inventory for Safety Analysis Calculations
Background
The establishment of a waste container radionuclide inventory (CI) for use in accident analysis
calculations must involve: (1) an evaluation of existing safety analysis orders and guidance documents to
establish the appropriate level of conservatism for the CI for safety analysis calculations; (2)
consideration of the projected waste inventory listed in Appendix A, and the desire to encompass as
much of the Pu-239 and Pu-238 waste as possible with the least design or operational impacts to both the
waste generator and the WIPP; and (3) evaluation of the existing RH WAC 10 transportation constraints on
nuclear criticality and Thermal Power criteria. The adequacy of the WIPP facility design, and operational
administrative controls is evaluated in detail in Section 5.2.4.
Each Pu-mix will be scaled to the RH WAC 10 nuclear criticality limit of 325 fissile gram equivalent
(FGE) for a RH 72-B waste canister and 200 FGE for a 55-gal drum in a 10-160B cask, using the
isotopic weight distributions and converted to PE-Ci (see Appendix B for a discussion of the PE-Ci
concept). Additionally, the maximum fissile loading will be no greater than 325 FGE per facility
canister.
The 10-160B cask certification requirements, limit the decay heat from all drums in the cask to 100 watts
per cask. The WIPP RH WAC Thermal Power transportation requirements, limit the decay heat from all
RH-TRU waste to 300 watts per 72-B waste canister. However, based on previous discussions, for the
predominant Pu-239 weapons grade operations waste, the most restrictive of the applicable WIPP RH
WAC criteria is the nuclear criticality criterion, which restricts a single canister to 325 FGE.
Past WIPP safety analyses have established a waste container radionuclide inventory (CI) for use in
accident analysis calculations based on inventory information from generator sites and on an average or
representative content of a RH waste container. Discussions between DOE and the generator sites
resulted in an agreement of plutonium-239 equivalent curies (PE-Ci) limits for the 72B canister. A 72B
canister that contains direct loaded waste has a PE-Ci limit of 80 PE-Ci while a 72B canister that
contains three 55-gal drums of waste (double confined waste) has a PE-Ci limit of 240 PE-Ci.
PE-Ci limits for the 10-160B cask of 20 PE-Ci are set in the NRC Certificate of Compliance. 13
drum (55 gal) in the 10-160B cask could contain up to 20 PE-Ci.
A single
Approach for Developing the Waste Canister Radionuclide Inventory for Safety Analysis Calculations
RH 72-B cask waste shipments to the WIPP are comprised of a DOT Type A (or equivalent) canister per
cask. Accident scenarios involve damage to the waste canister. Since the MAR for an accident scenario
is a function of the number of waste canisters assumed damaged in the scenario and their individual
radionuclide CI (MAR = CI * (number of containers damaged)), deriving a reasonable maximum for
MAR must also involve deriving a reasonable maximum for CI, as well as the distribution of PE-Ci
contents in the individual waste canisters assumed to be involved or damaged.
10-160B cask shipments to the WIPP meets the certification requirements for DOT Type B shipping
containers. The drums in the 10-160B cask (maximum of 10) meet the certification requirements for
DOT Type A (or equivalent) shipping containers. Accident scenarios involve damage to the 10-160B
cask, its load (55-gal drums), or facility canister containing a maximum of three 55-gal drums from a 10-
160B cask. Since the MAR for an accident scenario is a function of the number of 10-160B casks, 55-gal
drums, or facility canisters assumed damaged in the scenario and their individual radionuclide CI (MAR
= CI * (number of containers damaged)), deriving a reasonable maximum for MAR must also involve
5.1-5 January 22, 2003