ML20128F364
| ML20128F364 | |
| Person / Time | |
|---|---|
| Issue date: | 02/20/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20128F363 | List: |
| References | |
| FOIA-85-70, RTR-NUREG-0956, RTR-NUREG-956 NUDOCS 8507080217 | |
| Download: ML20128F364 (2) | |
Text
' ENCL 050RE M//D/f4L REVI!hD
.1).ao /is" Nu4EG-Or,gg, REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS Executive Summary CHAPTER 1.
REGULATING SEVERE ACCIDENTS^
2.
HISTORICAL PERSPECTIVE i
2.1 Maximum Credible Accident
- 2. 2 Design Basis Accident
- 2. 3 Reactor Safety Study 2.4 Review of Source Term Assumptions 2.5 Generic Siting Study i
- 2. 6 Current Source Term Analyses
- 2. 7 Ongoing Validation Programs 3.
CURRENT ANALYTICAL METHODS 3.1 Thermal and Hydraulic Behavior of the Reactor 3.1.1 Overall Assessment of Reactor Coolant System and Containment (MARCH)
I t
3.1.2 Detailed Flow Rates in the Reactor Coolant System (MERGE) 3.1.3 Detailed Core-Concrete Interactions in the Containment (CORCON)
- 3. 2 Fission Product Generation in the Fuel (ORIGEN)
- 3. 3 Fission Product Release from the Fuel (CORSOR) 3.4 Fission Product Retention in the Reactor Coolant System (TRAP-MELT)
- 3. 5 Fission Product Release from the Core-Concrete Melt (VANESA) 3.6 Fission Product Retention in the Containment 3.6.1 Natural Deposition of Aerosols (NAUA) 3.6.2 Retention in Water Pools (SPARC) 3.6.3 Retention in Ice Condensers (ICEDF) r 3.7 Code Validation Review 3.8 Uncertainty Study (QUEST) 3.9 Summary Evaluation of Current Analytical. Methods 3.9.1 Areas of Major Uncertainty 3.9.2 Fission Product Chemistry 3.9.3 Summary g ogo] 7 850329 j
SHDLLYSS-70 PDR iii
ENCLOSURE 4.
RESULTS FOR SELECTED ACCIDENT SEQUENCES 4.1 Summary Description of Accident Sequences Analyzed 4.2 Fission Product Inventory 4.3 Fuel Temperatures 4.4 Fission Product Release From Fuei 4.5 Upper Plenum Temperatures 4.6 Retention in the Reactor Coolant System 4.7 Fission P)(oduct Release from the Molten Core 4.8 Containment Leakage and Failure 4.9 Aerosol Retention in Containment 4.10 Summary of Sample Source Term Calculations 5.
REVIEW 0F SOURCE TERM WORK 5.1 Peer Review of NRC-Sponsored Source Term Work 5.1.1 Technical Expert Peer Review 5.1.2 American Physical Society Study Group 5.2' Discussion of Other Source Term Work 5.2.1 American Nuclear Society Study Report 5.2.2 Industry Degraded Core Rulemaking Program 6.
PROBABILITY AND CONSEQUENCES OF SELECTED ACCIDENT SEQUENCES 6.1 Core Melt Probabilities for Selected Sequences 6.2 Containment Failure Mode Analysis 6.3 Containment Event Trees 6.4 Offsite Consequences 7.
CONCLUSIONS AND REC 0mENDATIONS 7.1 Methodology 7.2 Selected Calculations 7.3 Applicability to Regulatory Process 7.4 Additional Observations and Recommendations APPENDICES A.
BASIC REACTOR SYSTEMS AND CONTAINMENTS B.
PROBABILITY ESTIMATES i
B.1 Accident Sequence Likelihood B.2 Containment Event Analysis and Estimation of Source Term Frequencies C.
CONTAINMENT ANALYSIS C.1 Summary of Containment Loads Working Group C.2 Summary of Containment Performance Working Grcup iv