ML20136F061
| ML20136F061 | |
| Person / Time | |
|---|---|
| Issue date: | 06/28/1984 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Royen J ORGANIZATION FOR ECONOMIC COOPERATION & DEVELOPMENT |
| Shared Package | |
| ML20136F066 | List: |
| References | |
| FOIA-85-485 NUDOCS 8407130407 | |
| Download: ML20136F061 (16) | |
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i J"*% UNITED STATES
,,.- I+. NUCLEAR REGULATORY COMMISSION j
jz WASHINGTON, D.C. 20555
(, f June 28, 1984 --
Dr. Jacques_Royen Nuclear Safety Division _
OECD Nuclear Energy. Agency .
38 Boulevard Suchet F-75016 Paris France
Dear Jacques:
Enclosed, as promised in my letter to you dated May 29, 1984, is a statement on our " current source term position" in the United States.
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Although I have no_t changed the title from that suggested in Dave Torgerson's circular' letter of January 9,1984, I would caution that it is primarily an NRC staff view and does not reflect any attempt at a concensus ' view to include U.S. private industry nor the U.S.
Department of Energy.
At the present time .it appears that I will be unable to attend the meeting of Principal Working Group 4 in September, but expect that NRC will be represented.
Sincerely, yy .'c~.-- . .
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R. ' Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration Office of Nuclear Reactor- Regulation
Enclosure:
As stated cc: D. Torgerson M. Silberberg -
.D. Muller J. LaFleur A. Millunzi F. Rahn yg}/30f0 YS
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Current National Source Term Position and Practices for the United States ,
s I. Introduction This sumary paper describes in brief terms the current status of ~
nuclear power plant accident source terms as they relate primarily to the regulatory and licensing process in the United States. The spectrum of accidents considered includes those postulated as design and siting basis accidents as well as those more severe that lead to core melt and potential containment failure. The latter class includes those formerly referred to as Class 9 accidents. Current usage in the U.S. favors the term " severe accidents," however. ,
There are a number of current and projected regula' tory policy develop-ment activi'ies t in the U.S. that are strongly related to the " source term position", question. These include Severe Accident Policy, Safety Goal Policy, and Revised Siting fulemaking (1, 2, 3). Current planning calls for adoption by the Nuclear Regulatory Commission of a formal Severe Accident Policy in 1984. Although the policy statement itself is not expected to hinge on the resolution of source term questions, its implementation would. . The proposed Safety Goal Policy is under-going a two-year evaluation period scheduled for completion in 1985.
In promulgating tWeiproposed safety goals, the Commission has en-visioned that they should be useful in evaluating the need for new regulatory requirements or for retaining existing ones, and for setting priorities on the allocation of resources for the resolution of safety issues. Using the proposed design objectives on individual and societal mortality risks and the benefit-cost guideline, clearly involves source term considerations. Possible revisions in nuclear power plant siting policy are currently expected to be actively re-considered in the 1985-1986 time frame following completion of the safety goal evaluation period and a current intensive effort to assess new technical information as it affects severe accident source terms.
This latter effort is discussed in Section III below.
II. Applications - Current Source Terms The nuclear power plant licensing process in the United States derives from two principal statutes, the Atomic Energy Act of 1954, as amended, and the National Environmental Policy Act of 1969 (NEPA). Safety issues are dealt with under the former, while environmental impact issues fal1 under the latter. Measures for the prevention and mitiga-tion of accident consequences are considered safety issues. The range of potential consequences of reactor accidents is discussed in formal Environmental Impact Statements from a risk perspective. Since late 1980,.these statements have derived from the probabilistic risk assess-ment methodology of the Reactor Safety Study (4, 5).
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. 2 The conventional and still current licensing process for nuclear power plants includes the safety analysis of postulated accidents, some of which have the potential for radiological releases to the biosphere. The guiding philosophy has been and is defense-in-depth.
. The implementation of this philosophy is articulated in NRC regula-tions and supported by Regulatory Guides and the NRC staff's Standard Review Plan (6). Chapter 15 of the la.tter document describes the treatment of postulated accidents and the criteria employed to assure that the offsite consequences of credible accidents are unlikely to exceed specified bounds. The source terms for most of these acci-
_ dents der i ve priinarily fron considerations of primary coolant ac-
_ tivity levels and criteria for fuel cladding failures.
The traditional and still current surrogate for the severe accident end of the spectrum that is used in this process is the accident postulated for siting purposes (7). The containment source term derives from TLD-14844, published, in 1962 (8). It is important to note, however, that a distinctive and important . feature of the ap-plication of the siting criteria involves the assumption that the integrity of the containment is maintained throughout the course of the accident. This considerction has led to considerable emphasis in .the safety evaluation process on the importance of measures taken to assure containment, integrity.
The evolution of severe reactor accident source terms over the period from 1957 to approximately 1981 has been fully described in a docu-ment published in November 1982 (9). Included are descriptions of the source terms (release categories) of the Reactor Safety Study as well as some later work to develop a set of " generic" source terms thought to be appropriate for the development of possible revisions to siting cri.teria (so called siting source terms, SST).
Over the past few years, the NRC staff has been engaged in risk as-sessment studies of a few specific nuclear power plants. The pro-posed studies were first set out in the TMI Action Plan (10) published in May 1980 as a result of the TMI-2 accident. Item II.B.6 of that plan called for studies of potential risk reduction for operating reactors at sites with high population densities (relative to all sites in the U.S.). The units specifically identified in the plan include those at Indian Point and Zion. Subsequently, these studies have been extended to units under construction at Limerick and Mill-stone-3. The NRC staff's studies in these cases are based primarily on Reactor Safety Study methodology with some corrections or modifi-cations but do 'not reflect anticipated improvements in our technical understanding of many of the important physical and chemical pnenom-ena associated with core melt accidents. References (11 ar? 12) list documents relative te these studies that contain the details of source term treatments, including those of the utilities that own the plants.
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l Another important application of severe reactor accident source term ,
methodology is in the area of qualification of essential mechanical and electrical equipment to assure operability in a post-accident ;
environment. Current practice in the U.S. continues to use the TID source term as a primary basis for NRC requirements for qualification to operate in post-accident radiation fields (13). i Finally, it may be of interest to mention that the NRC staff currently employs Reactor' Safety Study source terms in conjunction with value-impact assessments to prioritize the application of resources to the resolution of proposed generic safety issues (14).
It is necessary to point out that the NRC staff is following a de-- ~
. liberate policy decision with respect to regulatory applications of severe accident source terms. With one current exception involving a proposed standard plant design, the use of more recent research re-sults is awaiting completion of the thorough scientific peer review-process discussed in the next :ection. -
III. Research and Technical Issue Developments In order to provide more coherent technical support to deal with .
regulatory issues.jnvolving severe accident considerations, both the NRC and the U;E nuclear industry have developed special pro-grams. A " Nuclear Power Plant Severe Accident Research Plan" was formulated by the NRC and published as NUREG-0900 in January 1983.
Program element 9 of 13 total elements deals with Fission Product Release and Transport and is of primary relevance to source terms.
Appendix A of this summary is the description of that element copied from NUREG-0900. (The schedule and identification of specific tasks under this element is currently being revised and updated.)
Further. NRC focus on source term issues came about with the creation in January 1983 of an interim Accident Source Term Program Office.
The primary thrust of this Office has been to coordinate the work of NRC contractors and to assure thorough scientific and technical peer review of the research products reflecting improved methodol- i ogies (primarily models and codes) that can be used to define ac-cident source terms. The contractor reports BMI-2104 reflect the bulk of this effort to date. An NRC staff report (NUREG-0956) is-planned to be issued in 1985 following completion of the peer review process under the auspices of the American Physical Scoeity.
A considerable amount of HRC staff effort is also being devoted to two technical issues that currently seem to stand o'ut as having.
particular significance to source term questions and to the resolu-tion of regulet6ry policy matters. These are (a) the sequences and
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_4 phenomena that can lead to early gross failure of tfie containment, and (b) the potential for substantial increases in containment leakage under severe accident loads prior to (or even preventing) major struc- ,
tural failure (leak before break).
The- current rate of expenditure on direct fission product. source term research is about $12 million per year with a similar amount on other projects in the severe accident area that have a direct effect on the magnitude and character of source terms. .
The U.S. nuclear' industry in 1981-82 organized an Industry Degraded Core Rulemaking (IDCOR) program to provide the basis for industry participation in an anticipated generic rulemaking process that had '
been the announced intention of the NRC in 1980. Most of the tech-nical effort of this program has now been completed and is serving as a basis for the comparison of positions and possible resolution of the substantive technical issues in severe accident phenomenology between the NRC staff and industfy. Many of these issues are.directly or indirectly associated with source term characterization and quan-tification. '
To facilitate the resolution effort, the NRC staff is preparing a set of severe ' accident.. issue papers that focus on a comparison of tech-nical positjons. The subject headings of the 36 papers on severe accident phenomenology are given in Appendix B of this summary. As of this writing, this task has not yet been completed.
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References
- 1. "U.S. Nuclear Regulatory Commission Policy and Planning Guidance -
1984," NUREG-0885, Issue 3, January 1984
- 2. " Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation," 48 FR 16013, April 13,1983
- 3. " Safety Goals for Nuclear Power Plant .0peration," NUREG-0880, Revision 1, For Comment, May 1983
- 4. " Nuclear Power Plant Accident Considerations Under the National En-vironmental Policy Act of 1969," 45 FR 41738, June 20, 1980
- 5. " Reactor Safety Study," WASH-1400 (NUREG-75/014), October 1975 ,
- 6. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," NUREG-0800
- 7. Code of Federai Regulations, Titie 10, Part 100
- 8. " Calculation of Distance-Factors for Power and Test Reactor Sites,"
TID-14844, March 1962
- 9. "The Development gf_ Severe Reactor Accident Source Terms: 1957-1981,"
NUREG-0773, November 1982
- 10. "NRC Action Plan Developed as a Result of the TMI-2 Accident," NUREG-0660, May 1980 lla. " Preliminary Assessment of Core Melt Accidents at the Zion and Indian Point Nuclear Power Plants and Strategies for Mitigating Their Effects,"
(Preliminary Report), U.S. Nuclear Regulatory Commission, NUREG-0850, November 1931 lib. Ludewig, H., J. W. Yang, and W. T. Pratt, " Containment Failure Mode and Fission Product Release Analysis for the Limerick Generating Station: Base Case Assessment," Brookhaven National Laboratory,.
BNL-33835, April 1984 lic. Papazoglou, I. A., et. al., "A Review of the Limerick Generating Station Probabilistic Risk Assessment," Brookhaven National Laboratory, NUREG/CR-3028, BNL-NUREG-51600, February 1983 i
ild. Berry, D. L. , et. al. , " Review and Evaluation of the Zion Probabilistic.
Safety Study, Plant Analysis," Sandia National Laboratories, Jack R.
Benjamin & Associates, Inc. , NUREG/CR-3300, SAND 83-lll8, Vol .1, May 1984 ,
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lle. Pratt, W. T. , et. al., " Review and Evaluation of the Zion Probabilistic Safety Study, Volume 2; Containment and' Site Consequence Analysis,"
Brookhaven National Laboratory, NUREG/CR-3300, BNL-NUREG-51677, Vol . 2, To Be Published 12a. " Zion Probabilistic Safety Study, Commonwealth Edison Company of Chicago, 1981 -
12b. " Limerick Generating Station, Probabilistic Risk Assessment," Phila-delphia Electric Company, March 1981 12c. " Indian Point Probabilistic Safety Study," Power Authority of the State of New York, Consolidated Edison Company of New York, Inc., 1982 ,
- 13. "Environmen.tal Qualification of Certain Electrical Equipment Important to Safety in Nuclear Power Plants," Regulatory Guide 1.89, Revision 1, June 1984 ,
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.. g M'y 5.9 Fission Product Release and Transport * ,
. . . . i 5.9.1 Elem~ent Description Fission product release and transport research is directed at developing an experimental data base and models to predict .the- radiological source term for accident consequence assessment. This information is needed for emergency pre-paredness, risk assessment studies, siting rulemaking actions, and for equip-ment ciualification analysis. Although a signi,ficant amount is known about fission product release and transport under ccintrolled LOCA conditions, there are gaps in the data bas.e relative to fission product release and' transport behavior under severe core damage and core melt accident conditions.
. Nuclear power reactor sdfety studies consistently indicate that the uncertain-ties associated with estimating fission product release and transport behavior are among the jargest contributors to uncertainties in the risk to the public froni severe accidents at nuclear power plants. This result is not surprising
-for two reasons: (1) offsite consequences are directly affected by the magni-tude, timing, and makeup of the source term released from containment; and (2) t.1ere are 'large uncertainties regarding the actual potential source term.
The u'.timate objective of this research program is to improve the quality'of predictions of the potential fission product radiological source term released from containment under accident conditions.
5.9.2 Technical Issues Resolved by This Element
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. NUREG-0772 identified a number of key uncertainties related to estimating fission product source terms. The most important of these follow:
(1) Reactor coolant system (RCS) aerosol and fission product behavior (experi-mantal data for model verification),
(2) RCS thermal / hydraulic conditions under core melt accident conditions, (3) Containment failure time, mode, and location (experimental data and analysis),
(4) Fission product vapor phase and aqueous phase chemistry (experimental data),
(5) Less volatile fission. product, control material, and structural material aerosol formation rates (in-vessel and during interaction.with concrete)
(experimental data),
(6) Aerosol behavior in condensing steam containment atmospheres (experi-mental data),
(7) Re,cval of particulate fission products in water pools and ice beds (experimental data snd models), ,
(8) The effect of hydrogen combustion on fission product physical and chemical forms (experimental data), and 5-55
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- (9) Coupled models of containment fission product vapor transport, aerosol behavior, steam effects, and effects of ESFs.
,The objective of these research programs is to develop a data base for assess-
. ing fission. product release from the fuel and fission product transport behavior
- from the fuel to the environment. This research will focus on severe core ,
damage and core melt accident conditions.
3 The data base needs include informa-tion on the release _of fission products and nonradioactive aerosols from over-e heated and melting fuel, the chemistry of the released fission products, aerosol formation mechanisms, the transport behavior of fission products and aerosols I '; in the reactor coolant system and in 1.he containment, and the effectiveness of engineered systems ia mitigating fission product release under severe accident conditions.
- 5.9.3 Key Interfaces with Other Elements i ~
l Radiological source term and radiological source term analysis require defini-tion of accident sequence characteristics. This need is supplied by proba-i bilistic risk assessment studies (Elements'5.1, 5.2, 5:10, 5.11) that identify
- p. overall system performance and the dominant accident sequences.
O Fission product release and transport analysis also requires detailed informa-1;; tion on the physical processes that occur during severe accidents. Among the most important are l
1 i (1) i .i reactor coolant system (RCS) thermal-hydraulic behavior; v.; w L (2)
M 1 fuel heatup, telting, scvement, etc. (Element 5.4);
[ j (3) molten fuel / concrete interactions (Element 5.6);
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i (4) molten fuel / coolant interactions (Element 5.6);
' (5) containment response to severe accident loads (i.e., failure time and mode) (Element 5.8); and 1
(6) hydrogen combustion (Element 5.5).
L Major uses of the results of the fission product release and transport source Q
term research are equipment qualification, probabilistic risk assessment, O
definition of siting requirements, and emergency planning.
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h - An intensive program to evaluate realistic source terms for severe LWR acci-y dent sequences was conducted during the Reactor Safety Study (NUREG-75/014).
Ej - 'Because of the scarcity of applicable experimental data, large uncertainties
{ were associated with the fission product release and transport assumptions
, included in the study. In fact, in certain areas, so little information was available that'on'ly bounding assumptions could be made (for example, fission
' product attenuation within the primary coolant system).
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i Beginning about 1975, several studies were initiated by the NRC to investigate the release of fission products from irradiated LWR fuel rods under severe accident conditions and tc develop models for fission-product transport behavior within the re' actor coolant systems. These programs have provided (1) data on fission product escape from fuel rods under LOCA conditions in the temperature range of 500*C to 1600 C and (2) a mechanistic model (TRAP-MELT) for fission product behavior within LWR primary coolant systems under severe accident conditions up to and including fuel meltdown.
During the Reictor Safety Study, a relatively E mplei computer code (CORRAL) was developed to model the behavior of fission products in the containment atmos-phere. The original CORRAL code had relatively detailed models for spray wash-out of iodine vapor species; however, the spray removal of particulate fission products and surface deposition of aerosols and vapor species were crudely modeled. '
In the area of aerosol behavior within containment structures, significant progress that is broadly applicable to all aerosol studies has been made under the fast reactor ~ program. Experimental programs to characterize the genera-tion, agglomeration, and surface depositio,n rates of Na, UO ,2 and Na/UO2 aerosols have been conducted. The results of these experimental programs have formed the basis for a number of mechanistic aerosol behavior codes, including HAARM, ZONE, QUICK, fiAER05.
5.9.5 Plan of Work as a Function of Time The following three sectioE describe specific research projects and near-term results expected during FY83 and FY84. Figure 5.9 presents a detailed mile-stone schedule for these programs. .
5.9.5.1 Fission Product Release Research programs to investigate and quantify the release of fission products and aerosols from the fuel include the following:
(1) An experimental program to measure the release of fission products from commercially irradiated LWR fuel rod segments in a steam environment under elevated-temperature (1000 C to 2600 C) accident conditions.
Results at high temperature (2000 C) are scheduled for FY83 with the i higher-temperature tests (to 2600 C) to begin in early FY84.
(2) Experiments to investigate the release of fission products and structural material aerosols from larger bundles of fuel (0.5 to 10kg) using simulated irradiated fuel (fissium) and out-of pile heating technique (FY83-84). A numbe'r of 0.5 kg tests were completed in FY82.
(3) A program to investigate the release of aerosols from molten pools of core materials interacting with reactor cavity. concrete with core retention materials, and with residual coolant (encing in FY84).
(4) Examination and analysis of samples of the TMI-2 core (schedule depending on the TMI-2 cleanup schedule).
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(5) Developm2nt and improvement of mechanistic models (FASTGRASS AND START) to predict the release of fission products during interactions of the damaged and molten fuel with residual coolant and plant structures.
(6) Measurements of fission product release during Phase 1 severe fuel damage 3 testing in the PBF reactor (FY83 and FY84).
t y 5.9.5.2 Fission Product Transport i
[~0 Research programs in the areas of fission product vapor and aeroso1 transport p
f and deposition include the following: _
j (1) Continued improvement of the TRAP-MELT code-(models fi'ssion product 4 behavior with'in the primary reactor coolant system under severe accident conditions) and the coupling of the mechanistic, multicompartment TRAP-l.u-g MELT RCS code to models that predict containment fission product behavior and models for fission product (and aerosol) release from the core (on-g , . going, to be completed in FY84). Results from this program will be ~
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factored into the CONTAIN code. A thermal-hydraulic code MERGE for use 3 in fission product transport was developed in FY 1982.
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(2) An experimental and analytical
- program to provide model development data
!i for the TRAP-MELT code in the area of elevated temperature fission product vapor pressures; surface deposition rates and mechanisms; and fission 3 product chemical reactions with steam, prototypical surface materials,.
and other fission products (ongoing, to be completed in FY83, but may be extended).
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(3) Continuation of experimental and analytical programs to develop models for containment aerosol fission product behavior under severe accident conditions. The aerosol models will be incorporated into the TRAP-MELT, CORRAL, and the CONTAIN code to predict overall fission product transport behavior. These improved mechanistic codes will be used to benchmark-simpler models in MELCORR. (To be completed in FY83.)
(4) A series of small scale experiments will be initiated to provide data for interim verification of the TRAP-MELT Code. Thr e experiments will also' be directed toward investigating the potential tar resuspension of
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deposited aerosols from RC3 surfaces. This program was initiated in FY82 i and will be completed in FY84.
(5) Modification and operation of a facility to test and verify the primary syst'em fission product and aerosol transport codes. . Tests on volatile n
fission product (e.g., cesium, iodine, tellurium) transport will be initiated in FY83 and completed in FY84.
(6) An experimental program to investigate the chemistry of various fission product species (various forms of iodine and tellurium) in aqueous reac-tor solutions and their liquid /vapcr phase distribution under representa-tive accident conditions.
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5.9.5~.3 Fission Product Control Programs are. planned to ' investigate and quantify the effective ness of various engineered safety and mitigation features in reducing the potential fission product escape from containment. Within this area are progri.ms to function as follows:
(1) Investigate and quantify the radioiodine retention performance of impreg-nated activated charcoal adsorbers under accident conditions (completed in FY83).
(2) Conduct resear'ch on'the fission product mitigation performance of engi-neered safety features (e.g., containment spray systems, suppression pools, ice condenser beds) under the radiological and environmental.condi-tions predicted for severe core damage and core melt accidents. ,
3 (3) Study the effects of large aerosol sources (predicted for the most severe accidents) on engineered safety feature performance.
5.9.5.4 NUREG-0772Nollow-OnResearch
(1) Updated, severe accident, release-from plant, fission product source terms to supplement Reactor Safety Study estimates will be developed (completed in FY83).
(2) Quantitative estimatesr of.the uncertainties associated with these source
, term predictions and identification of the major sources of the uncer-tainty will also be provided by this study (completed in FY83).
(3) Analysis of past reactor accidents and core destructive tests for insights into fission product release and transport behavior and to compare current assumptions and models with measured releases are underway (completed in FY83).
. 5.9.5.5 Longer-Term Research Program Plan (FY84-88)-
(1) Fission Product Release.From Overheated Fuel--Beginning in FY84, tests will be initiated in the high-temperature fission product release program to investigate the release of fission products and aerosols from commercially irradiated fuel in the temperature range from 2000 C to approximately 2600 C. The test apparatus will include techniques (laser Raman spectros-copy) for direct in sit'u determination of fission product chemical form.
Two test series will be conducted in FY84, three in FY85, three in FY86, and two in FY87.
(2) Reactor Coolant System (RCS) Fission Product and Aerosol Transport Tests--
The tests on RCS fission product and aerosol transport will continue through FY85 and perhaos.into FYE6. In FYES, this expericental program will focus on determining the transport behavior of high-density aerosols within the RCS. Tentative plans call for tests with up to 800 kg of proto-typic core-melt aerosol materials.
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- i (3) Fission Product Transport Code (TRAP-MELT) Development--Pretest and post-test analyses of the RCS tests discussed above will be conducted with the TRAP-MELT code. Code predictions and experimental results will be com-pared and model improvements init.f ated (.if necessary) to correct defici-encies in the code. These analyses and model development activities should continue through FY86. At the end of FY86, the TRAP-MELT code will havetests.
been tested and validated by comparison with these large-scale ~
g integral Similar analysis will be performed using the extended TRAP-MELT code, the j CONTAIN ccde, and/or the MELCORR/ MATADOR code on large~-scale containment
! aerosol tests (to be conducted in-the Federal Republi.c of Germany). Again,
, these analyses should be completed by FY86.
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APPENDIX B SEVERE ACCIDENT ISSUES
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1.0 Severe Accident Phenomenology 1
1.1 Progression of core melt in the reactor coolant system 1.1.1 Reactor coolant system thermal and hydraulic behavior 1.1.2 Rate and magnitude of hydrogbn production in the. vessel and release from reactor coolant system 1.1.3 Fuel debris and in-vessel structure interaction 1.1.4 Fuel debris and vessel or vessel penetration interaction ,
1.1.5 L,ikelihood and magnitude of in-vessel steam explosions 1.1.6 Recovery potential pricy' to vessel failure 1.1.7 Primary system failure from overpressure 1.2 Loading of the containment
- 1. 2.1 Containment thermal and hydraulic behavior
- .: g 1.2.2 Rate and magnitude of combustible gas production, ex-vessel 1.2.3 Distribution of combustible gases and conditions leading to and resulting from detonations 1.2.4 Conditions leading to and resulting from deflagrations, Ldiffusion flames and flame acceleration 1.2.5 Likelihood and magnitude of ex-vessel steam explosions or
, steam spikes 1.2.6 Debris coolability in ex-vessel locations 1.2.7 Debris relocation following vessel failure 1.2.8 Fuel debris - containment shell, floor and internal structure interactions 1.2.9 Rate and magnitude of non-condensible gas production ex-vessel e
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1.3 Response of the containment and other essential equipment
- 1. 3.1 Characteristics and likelihood of containment leakage from shock loadings 1.3.E Characteristics and likelihood of containment leakage re-sulting from steam spikes and/or hydrogen burning .
1.3.3 Characterist.ics and likelihood of containment leakage re- '
sulting from slow pressurization l.3.4 Characteristics and likelihood of containment leakage re-sulting from external events .
1.3.5 Characteristics and likelihood of containment leakage re-sulting frcm thermal loading 1.3.6 Characteristics and likelihood of containment leakage re-sulting from internal missiles 1.3.7 Pctential for basemat penetration 1.3.8 Reliabi.[ity of early containment isolation 1.3.9 Equipment and instrumentation survivability 1.4 Fission product release and transport 1.4.1 Rate and magnitude of release of fission products from fuel (in-vessel) 1.4.2 Deposition of fission products during in-vessel transport-l.4.3 Rate and magnitude of release of radionuclides from fuel (ex-vessel)
'l.4.4 Deposition of fission products in containment due to natural processes 1.4.5 Effect of engineered safety features on fission product I retention 1.4.6 Deposition of fission products in other plant buildings I
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1.5 Ex-containment transport and consequences Environmental dispersion
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- 1. 5.1 1.5.2 Food chain transport 1.5.3 Dosimetry and health effects.-
1.5.4 Modelingofemergencyresponse .
1.5.5 Cost analysis D
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