ML12258A311

Attachment E - NRC Staff Answer to Motion for Summary Disposition of Contention 4

ML12258A311
Person / Time
Site:	Davis Besse
Issue date:	09/14/2012
From:	Atomic Safety and Licensing Board Panel
To:
SECY RAS
References
RAS 23469, 50-346-LR, ASLBP 11-907-01-LR-BD01
Download: ML12258A311 (3)

Text

FirstEnergy Nuclear Operating Company (Davis-Besse Nuclear Power Station, Unit 1)

License Renewal Proceeding NRC Staff Answer to Motion for Summary Disposition of Contention 4 ATTACHMENT (

ATTACHMENT E Transactions of the American Nuclear Society, Volume 102, 332, June 2010 Progression of Severe Accidents in the U. S. EPRTM 1 Zhe Yuan and Mohsen Khatib-Rahbar Energy Research, Inc: P.O. Box 2034, Rockville, Maryland 20847, zxy@eri-world.com, mkr1@eri-world.com Imtiaz K. Madni U.S. Nuclear Regulatory Commission: Office of Nuclear Regulatory Research, Washington, D.C 20555, Imtiaz.Madni@nrc.gov INTRODUCTION RESULTS U.S. EPRTM is a light water cooled and moderated Figure 1 shows comparisons of MELCOR- and plant designed by AREVA NP. This reactor is currently MAAP-predicted RCS pressure for a station blackout undergoing design certification review by the Nuclear accident scenario. It is noted that MELCOR and MAAP Regulatory Commission (NRC). The nuclear steam prediction of in-vessel accident progression are generally supply system is a four-loop pressurized water reactor consistent, except that MAAP-predicted event progression with four inverted U-tube steam generators. The is faster, resulting in earlier time of vessel breach as containment is a large-dry design and includes a number shown by the sharp drop in RCS pressure in Figure 1.

of unique severe accident mitigation features including an Generally, MAAP-predicted in-vessel hydrogen In-containment Refueling Water Storage Tank (IRWST); generation was found to be higher than the MELCOR a Severe Accident Heat Removal System; and a core prediction. This is due to a conservative enhancement of catcher with features designed to spread, flood, and cool the oxidation rate as modeled by AREVA. The the core debris ex-vessel on the containment floor. confirmatory analyses have shown that over the range of This paper focuses on assessment of the response of parametric values investigated to date, the induced Steam the U.S. EPRTM reactor, containment, and associated Generator Tube Rupture (SGTR) appears to be less likely systems to selected severe accident scenarios, including than other induced RCS failures (consistent with MAAP).

comparisons of MELCOR predictions of selected accident In addition, the impact of failure of instrument tubes on signatures and radiological releases to those of AREVA natural circulation is not as pronounced as for other using MAAP4. existing pressurized water reactors that have been studied

[1], nonetheless, the trends are similar.

DESCRIPTION OF WORK 2.0E+01 The analyses have been performed based on a 1.8E+01 MELCOR Results MAAP Results relatively detailed model using MELCOR 1.8.6 computer 1.6E+01 code. The MELCOR model consists of a detailed 1.4E+01 representation of the reactor pressure vessel internals, the Pressure [MPa]

1.2E+01 reactor coolant system (RCS) including the potential for in-vessel and hot-leg/steam generator tube counter-current 1.0E+01 natural circulation, the impact of failure of in-core 8.0E+00 instrumentation tubes on accident progression, lower head 6.0E+00 failure, melt behavior inside the reactor cavity, melt-plug 4.0E+00 failure, debris relocation onto the spreading floor, and 2.0E+00 coolability. Other modeling features include distribution 0.0E+00 of gases inside containment, passive autocatalytic 0 2 4 6 8 10 recombiners, and fission product release and evolution. time [hr]

Fig. 1. RCS pressure The largest differences exist in the behavior of core debris inside the reactor cavity following vessel breach, TM: Trademark of AREVA NP INC.

1 Work performed under the auspices of the U.S. Nuclear Regulatory Commission

where even though the MELCOR-calculated debris temperature is lower than that predicted by MAAP, the cavity melt plug failure (a special feature in U.S. EPRTM to enable melt stabilization before relocation to a region where core debris is expected to be cooled by another engineered cooling system) occurs later in MAAP as compared to MELCOR. There are several parameters that can potentially have a strong impact on the melt retention time. These include the total mass of corium in the reactor cavity, the amount of metal assumed to be in the pool layer in contact with the concrete, and the corium temperature. There are also differences in the modeling of molten core concrete interactions between MELCOR and MAAP that influence the calculated melt retention period.

Nonetheless, these differences are considered significant in so far as overall progression of the accident is concerned.

Provided uniform spreading of molten core debris material on the specially designed spreading floor occurs, passive flooding of IRWST water onto the containment spreading room results in melt cooling and stabilization for the accident scenarios that have been examined. The debris cool-down rate has been calculated to be faster in MELCOR as compared with the AREVA MAAP predictions.

Both MAAP and MELCOR results show that hydrogen concentration in containment remains below significant combustion limits under severe accident conditions (due to effective recombination by passive autocatalytic recombiners). The impact of reduced effectiveness of passive autocatalytic recombiners as a result of severe accident environment (i.e., poisoning or coking) on hydrogen behavior inside containment has been examined through a parametric approach.

In general, the calculated rate and magnitude of containment pressurization was found to be higher in the MAAP calculations. Finally, MAAP- and MELCOR-predicted fission product releases for scenarios involving intact or partially intact containment were found to be in reasonable agreement. However, significantly higher releases of volatile fission products were calculated by MELCOR for accidents involving containment bypass (e.g., due to steam generator tube rupture).

REFERENCES

1. A. KRALL, M. KHATIB-RAHBAR, Z. YUAN and R.

LEE, Analysis of the Impact of Instrumentation Tube Failure on Natural Circulation and Induced Failure of Reactor Coolant System During Severe Accidents, Cooperative Severe Accident Research Program Meeting, Bethesda, Maryland (September 2009).