ML091940021
| ML091940021 | |
| Person / Time | |
|---|---|
| Issue date: | 06/17/1988 |
| From: | Han J Office of Nuclear Regulatory Research |
| To: | Silberberg M Office of Nuclear Regulatory Research |
| Han, James T. | |
| References | |
| Download: ML091940021 (6) | |
Text
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2 MEMORANDUM TO:
Mel Silberberg, Chief Accident Evaluation Branch THROUGH:
Jocelyn A. Mitchell, Section Leader Source Term Analysis Section FROM:
James T. Han JUN 1 7 1988
SUBJECT:
DETAILED SEVERE ACCIDENT RESEARCH PLAN AND LOGIC DIAGRAMS -
NATURAL CIRCULATION IN THE RCS Enclosed please find my contribution to the Severe Accident Research Plan for the Natural Circulation in the RCS.
James T. Han distribution: circ; chron; AEB r/f; Han r/f; Han; 'wright; Meyer; Mitchell
[SILBERBERG/MEMO/HAN]
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3 NATURAL CIRCULATION IN THE RCS - A Research Plan
- 1.
Total Safety Issue Natural Circulation in the RCS involves multi-dimensional circulation of superheated steam and other gas including hydrogen in the uncovered region of the core and the rest of the RCS.
As a result, the RCS piping and steam generator tubes wi11 rise in temperature because of the continuous energy transfer from the uncovered core, and the RCS structural integrity may be challenged.
This activity addresses five technical questions: (1) Will Natural Circulation exist during a severe accident?
(2) How effective is Natural Circulation in heating up the RCS piping and steam generator tubes?
(3) Will Natural Circulation preclude high-pressure melt ejection (HPME) and direct containment heating (DCH) by causing RCS piping failure and system depressurization?
(4) Will Natural Circulation induce steam generator tube rupture (SGTR) and containment bypass of radionuclide fission products?
(5) What are the effects of Natural Circulation on the release of radionuclide fission products and hydrogen to the containment?
Natural Circulation in the RCS is one of the major areas of uncertainties identified in NUREG-0956.
- 2.
Major Parameters of Uncertainty There are three major parameters of uncertainty that may have significant impact on accident progression and the consequences.
Note that the uncertainty question on the condensation-induced hydrogen and steam separation (stratification) in the vessel upper plenum, as suggested in the Kouts' report (NUREG/CR-4883), is no longer considered to be a "major" one.
- 1.
Time-dependent Structure Temperatures The time-dependent temperatures of the surge line, hot leg junctions at the vessel, and steam generator tubes are most important in determining whether a RCS structure failure will occur early enough to prevent HPME and OCH.
The steam generator tube temperatures are also important in determining the induced SGTR which may lead to containment bypass.
These temperatures are calculated by either the MELPROG/TRAC or the SCDAP/RELAP5 code, in which thermal-hydraulics is coupled with core melt progression.
The uncertainty in predicting these temperatures is currently not known, but it will be estimated after the codes are validated against data from the Westinghouse 1/7-scale natural circulation I
- 2.
4 exper iments, the TMI-2 accident, and the core melt experiments of PBF, LOFT, etc.
Furthermore, an explanation for the absence of "high" upper plenum and hot leg piping temperatures in the TMI-2 accident will also be determined.
Break Size and Pre-Accident Conditions of Steam Generator Tubes The consequences of the SGTR induced by the natural circulation flows depend on several factors including not only the tube temperature but also the break size (at the failure temperature determined by the structural failure analysis), any defects or thinning of the tubes prior to the accident, and statistical nature of tube failure (because there are over 3000 tubes in a steam generator).
To address this issue requires both deterministic and statistical studies.
- 3.
Effects of the Elevated Structure Temperatures on Fission Product Retention in the RCS Temperatures of the vessel upper plenum structures, RCS piping, and steam generator tubes are likely to increase due to natural circulation flows.
The thermal and chemical effects of the elevated structure temperatures on fission product retention in the RCS are not well understood.
Additional analytical model development supported by experiments is needed in conjunction with the "Fission Product Behavior & Chemical Form" activity.
- 3.
Planned Research Activities To resolve the above major parameters of uncertainty requires interdisciplinary studies of thermal-hydraulics, core melt progression, and structural fai1ure analysis for RCS piping and steam generator tubes.
As a resu1t, the research activities cover a wide spectrum of phenomena.
The state-of-the-art computer codes such as MELPROG/TRAC and SCDAP/RELAP5 are being used to analyze severe accident progression in the RCS of a PWR in which theraml-hydraulics is dynamically coupled with core melt progression.
Structural failure analysis is handled separately using simplified models supported by test data.
Research activities completed include: (1) A state-of-the-art review of RCS Natural Circulation, which has been published in Chapter 2 of NUREG-1265 and a1so in Appendix J.3 of NUREG-1150, (2)
A COMMIX analysis for a Ht.LB ' accident in the Surry plant (a westinghouse reactor without vessel upper-head injection in a large dry containment) assuming intact geometry, which provides necessary information to SCDAP/RELAP5 and MELPROG/TRAC for modeling the counter-current flows in hot legs and steam generators, and (3)
Structural failure analyses for predicting the creep rupture temperatures as a function of RCS pressure and duration in time for the hot leg nozzle, hot leg piping, surge line, and steam generator tubes of the Surry plant.
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The planned research activities are listed be1ow:
- 1.
Two-dimensional MELPROG/TRAC and SCOAP/RELAPS analyses are being performed for the TMLB' accident with or without pump sea 1 LOCA in Surry to determine : (1) Whether an early RCS piping failure will occur so that the RCS is depressurized to preclude high-pressure melt ejection and direct containment heating, (2) Whether SGTR wi11 occur assuming no tube thinning or other defects and will result in containment bypass of radionuclide fission products, (3) The phys i cal and chemical states of the core me1t materials that enter the containment (as the initial conditions for core-concrete interactions study and/or direct containment heating study), and (4) The release of radionuclide fission products and hydrogen to the containment (as the initial conditions for assessing source term and hydrogen combustion in the containment).
- 2.
Similar analyses will be performed for the Sequoyah plant, which has a Westinghouse reactor with vessel upper-head injection and an ice condenser containment.
- 3.
Similar analyses will also be performed for a CE plant and a B&W plant.
- 4.
Two-dimensional and three-dimensional analyses using either the COMMIX or TRAC code are being performed to determine the reasons for the relatively low upper plenum temperatures in the TMI-2 vessel (compared with what is expected from natural circulation heating in the TMLB' accident).
- 5.
Additional tests will be performed in the Westinghouse 1/7-scale natura1 circulation facility in a cooperative program between EPRI and NRC.
Results will be used to validate and improve the codes for mode1ing natural circulation.
- 6.
Analyses for severe accidents in PWRs and BWRs will be performed as the n~eds arise.
- 4.
F1ow Charts Two flow charts are provided; the first chart shows the role of RCS Natural Circul ation in nuclear plant safety, and the second chart shows the research activities and resolution for RCS Natural Circulation.
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I ROLE OF RCS NATURAL CIRCULATION IN NUCLEAR PLANT SAFETY PWR HIGH-PRE SSURE RCS NATURAL CIRCULATION SEVERE ACCIDENTS RESEARCH AND RESOLUTION (TMLB',TMI-2,EfC.)
ACCIDENT MANAGEMENT
' i i
t NO HPME/DCH I HPME POSSIBLE I NO SGTR I I SGTR POSSIBLE I t
t NO CONTAINMET CONTAINMENT I,
BYPASS 8YPASS PROVIDING I.C PROVIDING I.C.
TO CORE-CONCRETE TO HPME/DCH INTERACTIONS STUDY STUDY r.c.
== INITIAL CONDITIONS I
RESEARCH ACTIVITIES ANO RESOLUTION FOR RCS NATURAL CIRCULATION (NC)
COMMlX ANALYSES FOR COUNTER-CURRENT FLOWS IN HOT LEGS
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ANO SGS OF THE PWR ANALYSES OF MELPROG/TRAC AND/OR SCOAP/RELAP5 FOR A "SPECIFIC" PWR (INCLUDING SENSITIVITY STUDJES)
STRUCTURAL FAILURE ANALYSES FOR RCS
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PIPING AND SG TUBES OF THE PWR VALIDATION AND IMPROVEMENT OF MELPROG/TRAC AND SCOAP/RELAP5 BASED ON
- WESTINGHOUSE 1/7-SCALE NC DATA
- TMI-2 PLANT DATA
- CORE MELT DATA FROM PBF,LOFT,NRU,ACRR,ETC.
NE\\</ MODELS RESOLUTION OF SAFETY ISSUES RELATED TO HPME/DCH, SGTR, AND CONTAINMENT BYPASSS I