Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed

Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.

ME

LC

OR

New Modeling

SQA

Utilities

Severe Accident Modeling and Evolution Presented by R.O. Gauntt

Sandia National Laboratories

ERMSAR 2019 Severe Accident Research Conference

March 18-20, 2019 Prague, Czech Republic

WASH-740

Brookhaven Rept

NUREG 0771 &

NUREG 0772

NUREG 0773

Accident management assessment

Accident Tolerant Fuels/Cladding

Non-LWR Reactors

molten salts, HTGR, Na-cooled

WASH-740 (Brookhaven Report) 1957 (revisited 1964)

Motivated in part to support amending the liability limits in the Price Anderson Act and growing safety and risk awareness

Explored a “maximum credible accident” with conservative assumptions

Source Term Categories were defined lacking code based predictions

Uncontained release of 50% of core inventory to environment

Assumed pessimistic weather conditions

45,000 fatalities, 100,000 injuries and $17B economic damages (1964)

Estimated likelihood to be very low

WASH-1400 (Reactor Safety Study – Rasmussen Report) – 1975

Comprehensive study on reactor safety in 1972 prompted by growing public concern about safety

Used fault tree & event tree methods was first PRA – NUREG-1150 to follow

Identified specific accident sequences Showed importance of small break LOCAs

Identified Core-Concrete attack as a mode of containment failure

Detailed assessment of containment aerosol transport and deposition

Estimated probability of an accident producing greater than 1000 fatalities as ~1x10-6/yr

Compared nuclear risk to every day accepted risks

Severe accident analysis methods not much improved over WASH-740

Release Categories extended from WASH 740 were used informed by CORRAL code

NUREG 0772 – (1982)

Technical Bases for Estimating Fission Product Behavior During LWR Accidents Rolled out latest scientific understanding about mechanistic nature of source terms

Determined that CsI was dominant form of iodine (as opposed to elemental iodine)

Discussion of transport and attenuation processes from fuel, to RCS, to Containment

Description of computer codes for predicting source terms

Computer codes for various physical process were under development

Codes became the basis for the Source Term Code Package described in BMI-2104

Source Term Code Package (STCP) BMI-2104

STCP was envisioned as an “integrated” approach to quantitatively calculating severe accident progression and consequences

Separate codes had been developed for specific aspects of the problem

Individual code results were passed explicitly to subsequent codes in the “integrated” analysis

1982 Siting Study, NUREG/CR-2239

Last major consequence study performed using hypothesized source term categories

SST-1 was an energetic early containment failure (alpha mode failure) releasing significant fraction of inventory to environment

All US sites were addressed in the study

Alarming prompt and delayed consequences estimated for worst case weather assumptions

Deeply flawed and highly conservative calculation repeated for every operating plant in the USA

Cited by critics for decades

Superseded by the SOARCA Consequence Study

NUREG-1150 (1989) Extended probabilistic assessment pioneered for nuclear power plants in

WASH-1400

Full scope PRA – Level 1, 2 and 3

Uncertainty analysis performed

Used Source Term Code Package for estimating source terms and consequences

Used Expert Elicitation heavily for uncertainty propagation

Alpha mode failure contribution reduced to ~1E-7 (SERG insights)

Exceedance probability for 1000 prompt fatalities reduced below 1E-10

MELCOR Code STCP code had problems

Explicit and tedious package to package communication

Manual porting of code output to code input

Inconsistent database between codes

Phenomenological coupling often missing

Error prone

MELCOR was initiated in 1982 as successor to STCP Many STCP packages (eg. CORCON and Vanessa) brought into MELCOR

Unified database for all packages

Semi-implicit solution

Much improved phenomenological coupling and feedback

MELCOR maintained and upgraded New models for emerging issues – Spent Fuel Pools, etc

Extended for non-LWR applications – HTGR, Na, FLiBe, molten salts

MAAP, ASTEC, ATHLET, SOCRAT, SAMPSON similar strategy

Severe Accident Phenomena

Aircraft Vulnerability Study and Spent Fuel Pools

9/11 Attack on World Trade Center prompted Aircraft Vulnerability Studies

MELCOR used to evaluate large loss of area damage footprints

Study specifically focused realism without conservatisms Precursor to SOARCA it turns out

Utility of portable equipment was explored for mitigation Precursor to NRC B5b portable equipment orders and ultimately FLEX

Spent Fuel Pool Studies – part of aircraft vulnerability study Improved models for drained SFP accidents

Led to NRC orders to optimally arrange spent fuel

State of Art Reactor Consequence Analysis - SOARCA

Followed the post 9/11 Aircraft Vulnerability Study where best estimate analyses were desired

Best estimate without conservative assumptions

Peach Bottom, Surry and Sequoyah Select accident sequences – SBO, ISLOCA

Baseline analyses done first, followed by Uncertainty Propagation Studies

Uncertainty analysis revealed new insights into base case definition

Unmitigated and mitigated analyses were performed showing benefit of portable equipment (pre FLEX)

Peach Bottom Extended SBO were predictive of accidents at Fukushima

Uncertainty Propagation

Traditional deterministic analyses reveal only point values

Uncertainty propagation by sampling methods now feasible due to modern computing capacity

Embrace and “deal” with it

randomlysample

uncertainparameters

N-times

establishuncertaintydistributionsfor uncertainparameters

0

1

values

0

1

values

Input File 1

Input File 2

Input File 3

Input File N

MELCORInput Files

MELCORUncertainty

Software

MELCORExecutable

Output File 1

Output File 2

Output File 3

Output File N

MELCOROutput Files

MELCORBatch Execution

Software

StatisticalAnalysis

sample of distributionfor figure of merit

confidence intervalsusing non-parametricmethod

correlation analysis

0

1

values

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12

Time [hr]

Hy

dro

ge

n M

as

s [

kg

]

Evolution of Severe Accident Analyses Earliest analyses were unrealistically conservative WASH 740

Bounding accidents with pessimistic assumptions

Source Term Categories were developed

Unrealistic and alarming consequences such as 1982 siting study

Phenomenological understanding through experiments led to development of physical models used in WASH-1400

Computer codes were developed and linked to solve system behavior - STCP

STCP was used in NUREG-1150

Greatly lowered consequences resulted from more realistic modeling

SOARCA study undertaken to assess knowledge gained since the accident at TMI-2 and to rebaseline flawed 1982 Siting Study

SOARCA SBO analyses for Mark-I BWR design prescient of Fukushima accidents

New knowledge coming out of Fukushima accidents will further improve severe accident modeling in LWR’s

Emerging issues center around accident tolerant fuels and clads and non-LWR reactor designs

WASH-740

Brookhaven Rept

NUREG 0771 &

NUREG 0772

NUREG 0773

Accident management assessment

Accident Tolerant Fuels/Cladding

Non-LWR Reactors

molten salts, HTGR, Na-cooled