Hong-Nian Jow

Conceptual structure of the 1996 performance assessment for the Waste Isolation Pilot Plant

The conceptual structure of the 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) is described. This structure involves three basic entities (EN1, EN2, EN3): (1) EN1, a probabilistic characterization of the likelihood of different futures occurring at the WIPP site over the next 10,000 yr, (2) EN2, a procedure for estimating the radionuclide releases to the accessible environment associated with each of the possible futures that could occur at the WIPP site over the next 10,000 yr, and (3) EN3, a probabilistic characterization of the uncertainty in the parameters used in the definition of EN1 and EN2. In the formal development of the 1996 WIPP PA, EN1 is characterized by a probability space (S{sub st}, P{sub st}, p{sub st}) for stochastic (i.e., aleatory) uncertainly; EN2 is characterized by a function {line_integral} that corresponds to the models and associated computer programs used to estimate radionuclide releases; and EN3 is characterized by a probability space (S{sub su}, P{sub su}, p{sub su}) for subjective (i.e., epistemic) uncertainty. A high-level overview of the 1996 WIPP PA and references to additional sources of information are given in the context of (S{sub st}, P{sub st}, p{sub st}), {line_integral} and (S{sub su}, P{sub su}, p{sub su}).

Performance Assessment in Support of the 1996 Compliance Certification Application for the Waste Isolation Pilot Plant

The conceptual and computational structure of a performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) is described. Important parts of this structure are (1) maintenance of a separation between stochastic (i.e., aleatory) and subjective (i.e., epistemic) uncertainty, with stochastic uncertainty arising from the many possible disruptions that could occur over the 10,000-year regulatory period that applies to the WIPP, and subjective uncertainty arising from the imprecision with which many of the quantities required in the analysis are known, (2) use of Latin hypercube sampling to incorporate the effects of subjective uncertainty, (3) use of Monte Carlo (i.e., random) sampling to incorporate the effects of stochastic uncertainty, and (4) efficient use of the necessarily limited number of mechanistic calculations that can be performed to support the analysis. The WIPP is under development by the U.S. Department of Energy (DOE) for the geologic (i.e., deep underground) disposal of transuranic (TRU) waste, with the indicated PA supporting a Compliance Certification Application (CCA) by the DOE to the U.S. Environmental Protection Agency (EPA) in October 1996 for the necessary certifications for the WIPP to begin operation. The EPA certified the WIPP for the disposal of TRU waste in May 1998, with the result that the WIPP will be the first operational facility in the United States for the geologic disposal of radioactive waste.

The NUREG-1150 probabilistic risk assessment for the Surry nuclear Power station

Abstract This paper summarizes the findings of the probabilistic risk assessment for Unit 1 of the Surry Power Station performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results show that the risk from internal initiators is well below the safety goals and is somewhat lower than estimated by Reactor Safety Study (RSS) over a decade ago. The risk from internal initiators is dominated by bypass accidents in the current analysis. The risk from seismic initiators is comparable to or greater than the risk from internal initiators, but still less than that estimated for internal initiators in the RSS. The uncertainty band for both types of initiators is considerably greater than that estimated in the RSS.

The NUREG-1150 probabilistic risk assessment for the Grand Gulf Nuclear Station☆

Abstract This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, consequence analyses, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term station blackout group and the anticipated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term.

Summary discussion of the 1996 performance assessment for the Waste Isolation Pilot Plant

Abstract The Waste Isolation Pilot Plant (WIPP) is under development by the US Department of Energy (DOE) for the geologic disposal of transuranic waste. The construction of complementary cumulative distribution functions (CCDFs) for total radionuclide release from the WIPP to the accessible environment is described. The resultant CCDFs (i) combine releases due to cuttings and cavings, spallings, direct brine release, and long-term transport in flowing groundwater; (ii) fall substantially to the left of the boundary line specified by the US Environmental Protection Agencys (EPAs) standard 40 CFR 191 for the geologic disposal of radioactive waste; and (iii) constitute an important component of the DOEs successful Compliance Certification Application to the EPA for the WIPP. Insights and perspectives gained in the performance assessment (PA) that led to these CCDFs are described, including the importance of: (i) an iterative approach to PA; (ii) uncertainty and sensitivity analysis; (iii) a clear conceptual model for the analysis; (iv) the separation of stochastic (i.e. aleatory) and subjective (i.e. epistemic) uncertainty; (v) quality assurance procedures; (vi) early involvement of peer reviewers, regulators, and stakeholders; (vii) avoidance of conservative assumptions; and (viii) adequate documentation.

Stochastic and Subjective Uncertainty in the Assessment of Radiation Exposure at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is under development by the U.S. Department of Energy as a geologic (i.e., deep underground) disposal facility for transuranic waste. An analysis is presented of possible radiation exposures associated with inadvertent drilling intrusions through the WIPP using future drilling rates obtained in accordance with requirements specified by the U.S. Environmental Protection Agency (EPA) in 40 CFR 194, Subpart B. The analysis attempts to maintain a separation between stochastic (i.e., aleatory) and subjective (i.e., epistemic) uncertainty as implied in the EPA regulations 40 CFR 191, Subpart B, and 40 CFR 194. The results of the analysis are presented as distributions of complementary cumulative distribution functions (CCDFs) for radiation exposure to oil field workers, where the individual CCDFs arise from stochastic uncertainty (i.e., many possible patterns of drill ing intrusions are possible over the 10,000 yr period specified in 40 CFR 191, Subpart B) and the distribu...

XSOR codes users manual

This report describes the source term estimation codes, XSORs. The codes are written for three pressurized water reactors (Surry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named ``XSOR``. The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. The codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena and their uncertainty which are not currently modeled by the mechanistic codes. The uncertainty distributions of input parameters may be used by an. XSOR code to estimate the uncertainty of source terms.