Following a postulated design basis accident (DBA), the fission product release from the reactor core into containment is referred to as the "source term." Amendments to 10CFR50 Appendix A, GDC-19 and 10CFR50.34 now make possible the implementation of an alternate source term (AST) defined in RG 1.183/NUREG-1465 as opposed to the current basis described in TID-14844. The changes to the source term release parameters (release timing, iodine chemical form and release fractions) offered by the AST allow the justification of safe and cost beneficial plant changes in the areas of:

• allowable leak rate increase
• isolation valve delayed actuation
• filtration system’s simplification
• mitigation system’s delayed actuation

As a result, these changes can offer:

• reduced O&M costs,
• reduced occupational exposure,
• increased plant availability,

as well as improved plant safety in the form of:

• reduced core damage,
• improved performance of safety equipment and systems,
• elimination of unnecessary wear on equipment.

Potential AST applications include:

• elimination or reclassification of charcoal filtration systems (PWR & BWR)
• elimination of isolation valve’s automatic actuation or delayed actuation (PWR & BWR)
• improved EDG reliability through reduced sequenced loading (PWR & BWR)
• delayed actuation of control room pressurization systems (PWR & BWR)
• elimination of MSIV leakage control and/or improved MSIV allowable limits (BWR)
• increased secondary containment draw-down times (BWR)
• improved operating margin for containment pressure (PWR)

Guidance for the application of the AST is now defined in SRP 15.0.1, "Radiological Consequence Analyses Using Alternate Source Terms," and Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors."

The AAC staff has been responsible for the original radiological consequence analysis of ten (10) BWRs/PWRs in the USA, Asia, and Europe. In addition, the staff has experience in the radiological consequence reanalysis of over twenty (20) additional units. Our experience in this area includes all aspects of source term determinations, on-site and off-site radionuclide transport calculations, habitability analyses, and related calculations and safety analyses in support of areas such as EQ and RMS set-points.

AAC’s recent and current involvement in this emerging nuclear safety analysis technology includes:

• Development of a full implementation of the AST for Carolina Power & Light's Robinson Nuclear Plant (RNP) per RG 1.183 including RADTRAD modeling for all the plant's radiological consequence analyses and ARCON96 modeling for dispersion.
• Development of a full implementation of the AST for Carolina Power & Light's Brunswick Nuclear Plant (BNP) Units 1 and 2 per RG 1.183 including RADTRAD modeling of the plant's DBA's and ARCON96/PAVAN modeling for dispersion.
• Development of Fuel Handling Accident (FHA), Control Rod Drop Accident (CRDA), and Main Steam Line Break (MSLB) AST based analyses for Entergy's River Bend Station (RBS).  Review of RBS LOCA AST analyses.
• Development of scoping LOCA, FHA, CRDA, and MSLBA design basis radiological consequence analyses for PPL Services Inc.'s Susquehanna Steam Electric Station (SSES) using the AST and RADTRAD.
• Development of a full implementation of the AST for Alliant Energy's Duane Arnold Energy Center (DAEC) per DG-1081 including RADTRAD modeling of the plant's DBA and ARCON96/PAVAN modeling for dispersion.
• Development of Software Quality Assurance ( SQA) manuals for RADTRAD, ARCON96, and PAVAN-PC.
• Development of BWR Suppression Pool post-LOCA pH analyses fro Duane Arnold, Brunswick, and River Bend.
• Participation in NRC and NEI sponsored AST meetings.

For more information on how AST applications can lower your O&M costs and increase plant safety margins, contact AAC at AST@applied-analysis.com.