NSD-NRC-97-5067, Provides Westinghouse Responses to NRC Requests for Addl Info Pertaining to AP600 in-vessel Retention of Molten Core Debris Topic
| ML20138C146 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 04/15/1997 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20138C152 | List: |
| References | |
| NSD-NRC-97-5067, NUDOCS 9704300052 | |
| Download: ML20138C146 (52) | |
Text
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'O Box 355 Westinghouse Energy Systems Pittsburgh Pennsylvania 15230-0355 Electric Corporation NSD-NRC-97-5067 l DCP/NRC0814 )
Docket No.: STN-52-003 !
April 15,1997 ,.
I. aent Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY
SUBJECT:
AP600 RESPONSE TO RSQUESTS FOR ADDITIONAL INFORMATION
Dear Mr. Quay:
Enclosure 1 provides Westinghouse responses to NRC requests for additional information pertaining to the APJ00 in vessel retention of molten core debris topic. Specifically, responses are provided for RAls 480.946 and 480.956 through 480.966.
The NRC should review these responses and inform Westinghouse of the status to be hsignated in the "NRC Strtus" column of the OITS. The OITS numbers associated with these RAls are 501) and 5040 through x M.
Enclosure 2 provides the peer review commerns on the DOE /ARSAP report " Lower IIcad Integrity Under in-Vessel Steam Explosion Loads"(DOE /ID-10541). This enclosure is being provided based on a request stated in an NRC letter dated March 25,1997. ARSAP is working on responses to these peer review comments. Westinghouse anticipates the responses will be available and provided to the NRC by the end of May,1997.
Please contact Cynthia L. Ilaag on (412) 374-4277 if you have any questions concerning this tunsmittal.
6w Brian A. McIntyre, Manager Advanced Plant Safety and Licensing iI jml ,b
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I 250001 Enclosures ec: J. Sebrosky, NRC (enclosures) -
N. J. Liparulo, Westinghouse (w/o enclosures)
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9 Enclosure I to Westinghouse Letter NSD-NRC-97-5067 April 15,1997 t
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. NRC REQUEST FOR ADDITIONAL INFORMATION .
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l Question: 480.946 l l
Please provide estimates of the margins to failure, in a form similar to Figure 7.10 in DOE /ID-10460, for the following variations on the base case: (1) input parameters set individually to values specified in column 2 below (14 ceses), and (2) input parameters set simultaneously to values specified in column 3 below (1 case).
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NPC REQUEST FOR ADDITIONAL INFORMATION
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Value Parameter Individual Sensitivity Cases Combined Case Vessel Wall Melting Temperature (K) 1200 1600 1900 Pool Power Density, MW/m' --
1.6' Fission Product Decay IIcat in Metal Layer (%) 60 10 Metallic Layer Rickness (m) -
0.2 2 Convection Correlation for Pownward lleat Transfer Nu,=0.1453 Ra* 22" Nu,=0.1453Raa22n Convection Correlation for Upward lleat Transfer Nu,=5.884 Ra* "" N u,=3.825 R a" "
Ex-Vessel Heat Transfer to a Flooded Cavity --
0.9 x ULPU Lower Bound' Upward and Sideward Heat Transfer Within Metal Layer 0.167 x Globe-Dropkin Globe-Dropkin Melt Hermal Conductivity (W/m-K) I 1.5 7.713.8' 3.9 Crust Effective Hermal Conductivity (W/m-K) 8.6 5.812.8' 3.0 4803 6 2 W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION m:: --4
. .g Value Parameter Individual Sensitivity Cases -
Combined Case Melt Density (kg/m') 9000 8300 700' 7600 Metal Layer Emissivity --
0.35 Vessel Effective Thermal Conductivity (W/m-K) 42 32110' 22 1
2
- shift distribution in Figure 7.8 to the right by approximately 0.3 Mw/m'
- shift distribution in Figure 7.6 to the le.t by approximately 0.7 m 3 - alternatively, can use 0.7 x best fit of ULPU results with RPV insulation 4 - broader distribution with generally higher values 5 - not currently treated as an uncertain parameter. Alternatively, can ' Set to extreme value which minimizes margin to CHF 480.946-3 3 V'estinghouse -
. NRC REQUEST FOR ADDITIONAL INFORMATION mi "'u 3
Response
The NRC provided a table of sensitivity cases to be analyzed with the IVR heat transfer model and the results to be presented as a probabihty distribution of the ratio of the heat flux to the critical heat flux. Given the thousands of calculations that must be performed to generate the probability distributions, a bounding approach is presented in which all the parameters are set to their con.servative values and the parameter on which the sensitivity is being investigated is varied to the requested values. The results are presented in the attached figures. The cases and their parameters are defined as follows:
Individual Sensitivity Cases base: Vessel Wall Melting Temperature 1600*K Pool Power Density 1.4 MW/m' Metalic Layer nickness 0.8 m Melt Thermal Conductivity 5.3 W/(mK)
Crust Effective Conductivity 2.8 W/(mK)
Vessel Thermal Conductivity 38 W/(mK)
Melt Density 8150 kg/m' Metal Layer Emissivity 0.45 .
Metal Layer Decay Heat Fraction 0% )
Heat Transfer Correlations as from DOE /ID-10460 l l
case a) Vessel Wall Melting Temperature 1200*K Figure 480.946-1 caseb) Vessel Wall Melting Temperature 1900* K Figure 480.946-1 case c) Melt Thermal Conductivity 11.5 W/(mK) Figure 480.946-2 case d) Melt Thermal Conductivity 3.9 W/(mK) Figure 480.946-2 case c) Crust Effective Conductivity 8.6 W/(mK) Figure 480.946-3 case f) Crust Effective Conductivity 3.0 W/(mK) Figure 480.946-3 case g) Upward Heat Transfer Nu = 5.884 Ra"'"' Figure 480.946-4 case h) Downward Heat Transfer Nu = 0.1453 Ran 22" Figure 480.946-4 case n1) Melt Density 9000 kg/m Figure 480.946-5 case n2) Melt Density 7600 kg/m Figure 480.946-5 case n3) Vessel Thermal Conductivity 42 W/(mK) Figure 480.946-6 case n4) Vessel Thermal Conductivity 22 W/(mK) Figure 480.946-6 case n5) Metal Layer Heat Transfer 0.338* Globe-Dropkin Figure 480.946-7 case n6) Metal Layer Decay Heat Fraction 60 % Figure 480.946-8 Conclusions The individual sensitivity cases do not produce venel failure. There is sigmficant margin to-failure in each of the cases.
480.946-4 3 Westinghouse
. NRC REQUEST FOR ADDITIONAL INFORMATION taa HE Combined Sensitivities Case base Vessel Wall Melting Temperature 1600*K extreme Pool Power Density 1.6 MW/m' Metallic Layer Rickness 0.2 m Melt Thermal Conductivity 7.7 W/(mK)
Crust Effective Conductivity 5.8 W/(mK)
Melt Density 3300 kg/m' Metal Layer Emissivity 0.35 Vessel Thermal Conductivity 38 W/(mK)
. Metal Layer Decay Heat Fraction 10 %
gm, = 0.9*ULPU Heat Transfer Correlations as from DOE /ID-10460 case a) Melt Thermal Conductivity 11.7 W/(mK) Figure 480.946-9 case b) Melt Thermal Conductivity 3.7 W/(mK) Figure 480.94,6-9 case c) Crust Effective Conductivity 8.8 W/(mK) Figure 480.946-9 case d) Crust Effective Conductivity 2.8 W/(mK) Figure 480.946-9 case c) Melt Density 9000 kg/m' Figure 480.946-9 l case f) Melt Density 7600 kg/m' Figure 480.946-9 case x) Vessel Thermal Conductivity 42 W/(mK) Figure 480.946-9 case y) Vessel Thermal Conductivity 42 W/(mK) Figure 480.946-9 caseg) Upward Heat Transfer Nu = 3.825Ra*'*" Figure 480.946-10 l caseh) Downward Heat Transfer Nu = 0.1453Ran 22" Figure 480.946-10 l case i) Upward Heat Transfer Nu = 3.825Ra"" Figure 480.946-10 l Downward Heat Transfer Nu = 0.1453Ra*25 )
case j) Variable Pool Power Density with case i correlations Figure 480.946-11 j l
Conclusions 1
The combined sensitivities case is a more exterme condition than the " extreme parametric case" in the IVR report, l DOE /lD-l(M60, which results in a q/qc, of approximately 1.0. Therefore, it is not surprising that the combined I s.ensitivities case results in q/qm, values between 1.0 and 2.5 which would produce, by definition, vessel failure.
However, it is noted that the parameters selected are at the extreme bounds of their ranges of values and in some l cases beyond their range, and result in a " physically unreasonable" configuration.
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SSAR/PRA Revision: None.
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, NRC REQUEST FOR ADDmONAL INFORMATION i in m.;
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- NRC REQUEST FOR ADDITIONAL INFORMATION t['
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, NRC REQUEST FOR ADDITIONAL INFORMATION Y
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. NRC REQUEST FOR ADDITIONAL INFORMATION 1
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- NRC REQUEST FOR ADDmONAL INFORMATION
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. NRC REQUEST FOR ADDITIONAL INFORMATION illil 0.8 ...,...,. .
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NRC REQUEST FOR ADDITIONAL INFORMATION j!!! Ej r.
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. NRC REQUEST FOR ADDITIONAL INFORMATION p: m=
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. NRC REQUEST FOR ADDITIONAL INFORMATION
.c u Question: 480.956 For cases where the metallic layer is contained within the hemisphere, Equation 6.9 appears to be in error (the cosine terms become zero for cases where angle theta equals zero). A more appropriate form of the equation is:
2 sin 0,A u(Tu-T.)* = s.in'O,A,.,(T.-To.)* + (H/R)Ai ,,(sinO, + sinei)(T. T..)"
i Please clarify what equation is used in the full solution.
Response
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The area correction term in equation 6.9 should be sin2 0 instead of cos2 0. The last term in the equation is correct.
- The error exists only in the documentation. The code used to calculate the IVR heat transfer solved the problem correctly, so no error has been introduce <* in the analysis. j i
SSAR/PRA Revision: None. l I
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. NRC REQUEST FOR ADDITIONAL INFORMATION
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Question: 480.957 Figures 7.4 and 7.6 indicate that the oxide layer height will be 1.5 to 1.6m and the metal layer height will be 0.88 to 1.0m. 'Ihis would result in the top of the metallic layer being above the hemispherical portion of the vessel.
Please justify that the full solution adequately addresses this situation, and provide an estimate of the error introduced by the modelling assumption that metal layer is contained within the hemisphere.
Response
If the metal layer reaches elevations where Oi 2 90', the program uses the cross-sectional area of the reactor vessel.
. cylinder wall for the pool area. This is the proper area, and no error is introduced.
SSAR/PRA Revision: None.
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. NRC REQUEST FOR ADDITIONAL INFORMATION
--wi Question: 480.958 Please clarify.the nomenclature related to T, and Tu. In Equation 6.10 and in the discussion on page 6-4, it appears that T, should be replaced with Tu. If not, please justify why T,is used as a sink temperature in Equation 6.10 and T ois used as a sink temperature in Equation 6.12. Also, please show the equation used to relate T, to To and T,,.
Response
In Equstion 6.10, the temperature T, should be specincally T , and the typographical error in the report is corrected.
SSAR/PRA Revision: None.
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. NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.959 ;
Please clarify the nomenclature and assumptions in the third term of Equation 6.12. In this term, it appears that the molten metal layer emissivity, e, should be replaced with the radiative sink emissivity, e,. Please describe and justify the view factors and area relationships invoked to omit any te ms with the vessel emissivity. E, and surface area ratio (S/S,). It appears that the upper internal structure's heat was assumed to be entirely radiated to the vessel in'ner wall and that the upper internal structure surface area was assumed to be much smaller than the vessel inner surface, area (S, u S.). However, the latter assumption contradicts nomenclature shown in Figure 5.1 (S, pointing to both the upper internal structure inner surface and the vessel inner surface implies that these areas are equal).
Response
The emissivity in the third term of equation 6.12 should spe'cify the radiative heat sink emissivity, e,, and the 4
typographical error in the report is corrected. The inner wall surface area is taken as that of the core barrel and it is assumed to dominate in comparison to the upper internals. Figure 5.1 is only a schematic. As long as the effective area, S,, is la ge, which is the case, the upper internals area is negligible. See the sensitivity study presented in the response to 4T,0.966.
SSAR/PRA Re /ision: None.
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. NRC REQUEST FOR ADDITIONAL INFORMATION 3p:
Question: 480.960 It appears that the coefficient of the fourth term in Equation E.3 should be 1.35x10-2. Is this a typographical error?
Response
The coefficient in the fourth term of equation E.3 should be 1.35x10 2, and the typographical error in the report is corrected.
SSAR/PRA Revision: None.
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-. NRC REQUEST FOR ADDITIONAL INFORMATION
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Question: 400.961 Please clarify why the ratio (H/Rp2s s discussed on page 6-3. The equations used for predicting upward and downward heat transfer coefficients (Equations 5.12 and 5.28) dont have an (H/R) term.
Response
Chapter 5 is simplifed so the main effects can be visualized. For completeness in the calculation in Chapter 6, the H/R term is included.
. SSAR/PRA Revision: None.
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- NRC REQUEST FOR ADDITIONAL INFORMATION f
Question: 480.962 Please describe and justify the assumptions used to obtain Equation 5.41 from Equation 5.35.
Response
The only assumption used to obtain equation 5.41 from equation 5.35 is that the two boundary layers in the Globe.
Dropkin equation are of e, qual resistance. This is true, so it is not really an assumption.
Equation 5.35 is
- Nu= 0.059Ra o s given:
N Nu = G
- kAT l l
l and ya , 80 b TH' av so i
f *IS qN 800
=0.059 khT g av ,
The layer height. H, cancel out. If AT' m 0.5AT, then 2AT' gives the correct temperature difference. This is done to define the temperature difference from the bulk temperature instead of the surface temperatures.
_._f-- =0.059 gp2AT k2hT' < av ,
i from which 4= A AT' (2)(2)'8(0.059)AY r av , T' 480.9621 T westingtmuse
- NRC REQUEST FOR ADDITIONAL INFORMATION
+cc 1
substitute h= A A T' gives equation 5.41 i .
f 3in h=0.1Sk $ AT' t av ,
SSAR/PRA Revision: None.
4 480.962 2 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Eis Eg T II Question: 480.963 Please indicate the equation used to estimate A in the full model (Information on page 6-4 suggests that a Prandtl number ft.nction, rather than Equation 5-47, was used to evaluate the term, Ay).
Response
Chapter 5 is simplifed so the main effects can be visualized. Equation 5.35 is particular to the case where Pr = 0.13.
For the general modeling, equation 5.34 is used in which the Pr term is included.
SSAR/PRA Revision: None.
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W westinghouse
v NRC REQUEST FOR ADDITIONAL. INFORMATION F=
h Ouestion: 480.964 Please indicate what properties were varied with temperature in Equation 6.11 (the equation suggests that all the thermophysical properties were evaluated at the appropriate film temperature except the thermal conductivity, k).
Please specify the values used to obtain a single value for A (2764 W/m K) that is applicable to all directions (see captions for Figures 6.3 and 6.4).
Response
All properties, including thermal conductivity, were varied with temperature as shown in Table 7.1. The one value of A was used only for comparison with the simple model in Chapter 5 with very good results. The effect of properties is negligible in any case.
SSAR/PRA Revision: None.
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NRC REQUEST FOR ADDITIONAL INFORMATION
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..u Question: 480.965 For the following variables, please: (1) clarify if single values were used, (2) state the values or range of values assumed, and (3) provide the basis for these values:
- Upper internal structure surface area, S, Vessel internal surface area, S, (Figure 5.1 nomenclature)
- Upper internal structure thickness,6,
- Vessel thickness, 6 Upper internal structure emissivity, E,
- Upper internal structure thermal conductivity, k,
Response
The upper internal area is lumped with the core barrel sink which is modeled as the area of a 5 m tall cylinder,2 m radius with a 2 m radius disk. De total area is 75.36 m . 8The vessel internal surface area is modeled as equal to the core barrel area. The upper internal structure thickness is modeled as 5 cm. The vessel cylinder and lower head wall thickness is modeled as 15 cm. The upper internal structure emissivity is 0.8 which is the correct value for steel that is somewhat oxidized. The upper internal structure thermal conductivity is 30 W/m/K as opposed to the 32 W/m/K used in the lower regions. In the upper internals the steel is colder and the thermal conductivity increases with temperature.
SSAR/PRA Revision: None.
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480.965-1 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION ME SE Question: 480.966 In the response to RAI 480.459, it is noted that: (1) the initial core barrel thickness (2 inches) is assumed for the upper internal structure thickness, (2) the initial vessel thickness (8 inches) was assumed for the vessel thickness, and (3) the vessel side wall surface area above the melt (57.4 m') was assumed for the vessel internal surface area.
Please demonstrate that results areni sensitive to values assumed for these parameters. Also, clarify if the area of the upper internal structure surface area and the vessel internal surface area were assumed equal as suggested by Figure 5.1 nomenclature.,
Response:
- Fige:res 480.966-1 through 480.966-3 present the results of the requested sensitivity analyses. The results are not
%nsitive to the values use j in the analysis. The upper internal structure and the vessel internal surface area are assumed to be equal.
SSAR/PRA Revision: None.
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a v NRC REQUEST FOR ADDITIONAL INFORMATION jjH @{
i Base Case 00 . . .
base r O 3m cyl and no disk c ans ni and n disk. 2cni banel 0.5- _
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c NRC REQUEST FOR ADDITIONAL INFORMATION iRii Vessel Wall Thickness Effect in Base Case 0.6 ,
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- 20cm 0.5- .
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+ NRC REQUEST FOR ADDITIONAL INFORMATION i:y! in l
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Extreme Parametric Case
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, ,. n s n- e- n- -
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0.8-
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0 80 0 10 20 30 40 50 60 70 Angle (degrees) 1 i
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Enclosure 2 to Westinghouse Letter NSD-NRC-97-5067 April 15,1997 i
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01/10/97 FRI 11:22 FAI 1 630 252 4780 .0L-RE -. e S. SORRELL 4002
- 1- S-97 ; 15:18 CCLL. OF ENGR. ADMIN
- 1 830 252 4780;# 2 SENT SY QW-WAD 15CN l Y c.s $4 13~L i
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1 Nuclear Enalnearina and Enalnearina Physica MA i pepartment of magineering Physiee, univoremy se wiesoneen i
meaner seeny neeenmn censer 1ses Johneen othe Mediesn. wl a370s . .
Phens:(ess) ass 81ss4 80o1 Pas: (ses) ass 47ers44eo August 13,1996 l Prof. T.0/Iheofanous '
i Chemical and Machanle=1 Engr. Depts.
Director, Center for Risk Studies and Safety
! University of California, Santa Barbara i
Santa Barbera, California 93106-1070 .
l
Dear Theo,
[ I received your letter of July 9th,1996. and the enclosed reports. I am still waiting for the two
- verification reports (DOE /ID.10503 and 10504]. Sinos Ifirst neceived the initial sequest by Mr.
I Steve Sormll, I have boca thinkingabout this review process and have come to a couple of
- decisions that I wantedte inforrn you about. First, because of the depth of your IVB report and l thesupporting. documents it will take more than '24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' to adequately reviewyour work.
l Thus, I will only be able to begin the review in that time period,and plan to continue it to the i cxtent I feel is necessary to adequataly understand the analysis you and your Mhagues have i done. Second, because of my association with the Nucicar Regula!~y (%nmission in research and associated consulting, it is not aprvriate that I be paid for this review.
I will follow the tirnetable that has been set for this review n;. best as I canand will send copies of my cornments to the USNRC research staff.
Sincerely, .
l
- Michael Corradini Nuclear and Mechanical Engr. Depts.
Wisconsin Distinguished Professor l xc: Dr. Sud Basu l
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l i DETIAL COMMENTS sad QUESTIONS for DEO/ID 10 41
! 1] ns authors do a goodjob in giving a centsat for their wodc. However I smoot sors if tids analysis which is l provided is a failuss analysis for the AP600 reactor pressure vessel or a design analysis for the RFV. %s 4 former implies thadt would be a 'best. estimate' analysis, while the latter must account for factors of safety to I nasure survival Can the authors clarify what is the utdmate intent of the work?
)
l 2) In Chapter 3 the authors de6ne the failure critaria and the dragility curve for the roector pressure vessel. IfI j understand the approach a stralafeiture limit is used and the esenciated analysis suggests a RFV lower wall j failure probabiliry rangmg from 0 to 1 for a spectrum of loading patterns withimpulses between 200 - 400
) kPs-sec. Structural failure limits are a complex assant work and I have asked colleagues a help review this ,
j chapter. Currendy. I wonder how this failure envelope compares to that of previous LWR plaataeatysed for an j in vessel steam amplosion; e.g., the ZIP study in the early 1930s?
l 3] In Chapter 4 the authors' major point is that the case and veneel designis sufSeiendy diserent frera past LWR 4 such that the core malt behavior is quits olfferent. Two aspects are T 'M first, the lower coresupport plats and the non-active fuel length above it (30sm] is large enousbia elas to delay the core melt
> progression downward; escond, the core steel reflectorin the radial direstion is also thicker (over 15am), also delaying and changingthe details of radial core melt progressian. In essence, the ' race' to thebreach of the 4
corium raalt crucible, which is formed during the meltdown. downward or rechally outward is scv.a^y these h=adanies. %s authore use a specific 388 core melt =~1daar sequence taillustrate this behavior. If rew j
- accepts this prusniae about a radically difestent core geometrical design, a few questions artes
a] What is the sensitivity of maltdown timing to downward boil off of water? More esemples sze needed.
- b) is the casu melt event timing essendstly indarandaar of easidset asquence? No guidanos is given here.
c) The corium amit flowrats seems to be set by the 'rlp* in the redacturnlong the radial edge of the sors regica l
, at the very top of the pool. Is thias rea11stis =aeimata since it is not much moso than that one would calculaes
> from adiahade hastup and maltdown of the core; e.g., as evidenced atDG2 7 The ansbors soggest that 200to l 400 kg/ses " appears to be a reasonabis range physically to bound the behavior"; but I wonder if we really know that much about this core malt fal16@,. W in a radasally new ===a"is design that this flowress is a i
'reamaankla bound *7 Morejusdfication is needed for me to ' buy' the argumsat.
d) This last quescon really leads me to the key aWaa_ of this whole analysis; Le., the authors leave mo i'
with the impression that there is a good deal'of certainty in the melt progression and I have sigxdficant troubts acceptingthis premias. Speciftsally, the whole analysis hinges on the fact that the melt crucible which forms j during the malt progresalon has a structural lategntyo(enough certainty that it would release the melt radially a through a pour area no larger than 0.02 to 0.04 sq. meters. his estimaan also seems to berebast enough that it i would be a " bound" even with coolant reseed into thecore region and any possible disruptive events that may occur. I am vey dubious about this and need to ses more analysis toeccept this as a ' reasonable bound'. His melt eniture laceden and pour rate is the key deteradanat is lisaiting C ; " FG energades,
- 4) In Chapter 5, the authors dotad their mulddhnensional premixing analyses.As stated previously, the malt flowrats of 200400kg/s seems to predatarndne thabanign namre of the PCI energeties, but rrdning is also part
- cf the prosess. A few questions arias hers
a) Why has the effest of RFV pressure been noglected? Promisdng will occur steisvated preneures not 1 bar I
(2-5 bare) and this will aSeet tbs mining prosess. Also.will the des in presense lose 11y during mhdog wil!
canas the pool to subcool and this has been neglasted. Were calculations done to ' bound' theos essets?
" b) ne authore sessa as have only eensidered the premialog process as the maltfaus through tha lindtsd water pool from its surfeos to the curved RFV bomem.Would not anzing condnue as the melt sendonas to fall along -
the wall. Disasems to have been neglected. Is this promining pmcese of no importance or is the promising A
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1 analysis with PM.AL2HA not valid for thanalanger dmas?
! c) The biggest effect of % small pours in my mind is that they may casas localICIs which do not harm the RPV but totally change the malt peering charesteristles for subsequent malt pourst i.e.. thans small pours and
===ariatad PCIswill damage the core malt amsthis and markedly insmaas tbs melt flowress or changs its i larmeiaa '!he authors have gens to great pains to deterados the firagility of the RFVwaB. but talatly igners ths
- fragility of the malt omsible and the ef9sst of thses PCIs. I wouldsuggest that larger malt pours will bs induced from the boteosa of the emoiblaas well.as along its radially edges with larger holas, all caused by earty smal1 PCIs. How have thans events been considered or annaarvatively bounded for RPV survival?
! d) Phaally, the FM ALPHA modal has a parametric fbal breakup model that (amaa+iaamd briady, but has yet to be assessed against asperiments. For thanaamall peer ratas, abs medal strast is not of great inessent, but would be for larger pours in thans comples geometriesJs this madat dissuased in the support daanmaasa?
l 5) In Chapter 6, the authors use ESPROSE.m in a revised 3D vendan to simulaas tha==pta=I=a within ths
! RPV lower planum. Given the pnadaad mass of fuel we have a range of resulas given in Table 6.1. Only a couple of quesdans erlas:
a] Why is the trisgar thne so shest i.e., much less than l'asc7 Is it due to the tinw to the RPV wall?Why rannot thrther mixing along the RFV wall cause larger explosions? '
I b] Why is tbs impulas largest for the mid rangs value of ' beta'7 Is the impulseof 200 kPa ses asar the failure lignit? or am I reading this predictica corready?
i c) The detailed calculadonal results in Appendia C abrupdy step in manysaaes at 1 or 2 or 3 mil 11aamanda-
! Why? Is this an indirahaa of samathing numerically fintal in the ESPROsE.m ais==latiaa or whas is up?
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- Mussear seamy neeenan center j- 1s00Johneen Detwo M iame,wls870s j Phone (sos)asa aises.aeo1 Fes
- (sos)assesf/s.stoo '
January 3d,1997 l s
, Prof. T.G/Iheofanous i Chemical and Meehmakal Engr. Depts. l
} Director, Center for Risk Studies and Safety J
i University of California, Santa Barham Santa Barbara, California 93106-1070 l l
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DearTheo,
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- I am sorry for my tardiness a reviewing your W.s, DOE /ID-10503/04 After reasiving the
! PM-ALPHA report in latter Mahar, I had to finish my class and admlainwadve obligadoes and
) thus was unabic to give it the proper anaadna I have read them over the Christmas bensk and
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have some comments. On balance both reposts are well written and address the verificadon of
- tbc *~=~*~ models. If this were a model/ code review for the NRC, we would also nood a complete code trimand from each for review. Appendix A gives excerpts, but I wonder if we j can get the complete code theoryAlsers-manual? Attached are my comments and quesdans.
Hope to see you at the next CSNI meeting in sydng.
i Regards, i
I Michael Corradini Wisconsin Distinguinhad Professor l Nuclear and Mechanical Engineering
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! INITTAL COMMENTS and QUESTIONS for DEO/ID 10504 i
- The overall report is quite informative, but I do have specific comrnants/ questions that need to
! be addressed. -
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- 1) The analysis of the QUEOS exp.gi;; ts are very interesting. For any of the say # en
- [Q5,6, 8,10,11] the visual lmage is compared to the code, and the leading edge, level swell,
- steaming rata, staasn pr~4~~i and y..r irs is c=aami. My first question is what is the critarion to desarmins the leading edge? In the pictures for the tests, specifically Q10 and Q11 it
- seems to rue that FM-ALPHA is prWat the movement of the front to be faster than the data
- indicates. Yet in the plots the opposite is represented. Either there is a contradiction or I am l
observing numerical diffusion in the images and the researchers have a definition of ths leading l
edge that " corrects or compenastas" for this. I have seen the same behavior with IFCI and therefore, am sensitive to it. ' Ibis needs to be sosted out before I would say that the agrm-ent l
i in the idnamatics is Wale. 'Ibe MIXA results in Section 3 seem to indicate the same behavior to ma and thus I am worried about this numerical diffusion. There was also no study 4
cf the nodalization effects in Section 4 and this is surpnsing given the results in Section 2. This
- seerns to be a logical thing to do and really should be dans.
- 2) The second comment about QUEOS relates to the radiative heat transfer model. On pago l 2-16 the report states that the radiadve natal had to be changed from what is normal in PM-ALPHA to accomnwtata the experiments. Later on page 2-19, the report states that ths l
tests do not meet the ' fitness for purpose' critaria, and one reason is that the e,a-ature is too l
' low (2000C compared to 3000C]. I am troubled by this empirical "fix" to model the test and thus, am wondering about the " mixed" transport modelin PM ALPHA. This is imown to be a tough problem, clearly, but to arbitrarily change it seems too rough. Also, I disagree that the tests are not "6t for purpose".They are more fit than others and thus, am very misvant Thus, l
the proportion of the radiative transfer that goes into bulk heating versus steam production is important to consider and improve upon.
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! 3) I would also like to see a calculation of PM-ALPHA /3D for QUBOS if indeed benefit to a 3D calculation. It seems that the QUBOS tasts are the largest and highest i . temperature simulant bats to data with solid partic!ce; thus, it may be of use.
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SENT SY:UW-NAD180N 4]'Ibe report finally =raminaa the FARO-LWR test L 14 as a compedson with a large prototypical =Imntmar malt poured into water. This amama like a reasonable comparison test, but I am surprised about what data is compared. 'Ibere is an enormous ammmit of data avausble l
< over the that twenty seconds of the test [the first 54 seconds is reasonable before heat loss comes si.-alm ~atly into play) and yet the data c~a.=_dson is sparse at best. I would suggest the l
following vadables be displayed and -ug,g.sd over the first 54 seconds:
a] the total pressure and pressure rise rate (done now]
b] the steam and water temperature at a few locations since its 2-D l
c] the kiammaries of melt entry and arrival at the chamber base and Pling j d] the surface area sensrated by the breakup as a function of time e] the mean particle diameter as a function of time i f) the energy flow to vapor and cooicot liquid and loss by fhet g] the level swell of the pool (done but not for long enough timan]
i Also I am concerned about the arbitradness of the dynamic breakup model [$ value = 50], that i
is used and described in pages A-34/35.'Ihis whole yws,J.u. is a dia; exercise for some f value of beta unissa the results begin with ajet of ~10cm and break up to a size that is consistant with the post test debds data [as well as the amanar left as a ' cake']. It would seem advisable to compare the 'ftosen' model to other PARO tests to prove that results can be consistently predicted for LO6, LOS, L11, L19 and L20; all of which were high ptessure tests for quenching. Also the ' mixed' transport including radiative transport would have to be held constant in these compadsons to prove the match of L14 has some limited ' universality'.
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- INITIAL COnedENTS and QUESTIONS for DEO/ID 10503 l
'Ihe report is very well or-=ai-ad and describes in sufficient detail the ability of ESPROSBan to 3
pfws; shock propagation calculations for gas / water and vapoc/ water situations. I do not l completely understand the origin of the CHAT [or CHAT-QL] cods comparisons. Are these e7 I standard code models or a fbran!stion of the authors to do a code wz : ==r=i understood them to be the latter, and thus I wonder about the need to compare to acusal l
aspd= ^ 1 data on shock propagation in singis phase and multi-phase symamma This is a l
minor point, but I think for completaness a link to data is best., My main an====t= are about the l comparison to the KROTOS data.
l i 1] 'Iha inidal statement is made that the KRCyrOS tests are a challenge since they have h
imr fect characterization of the initial canditions. Ons question may be if thans are any ot w i
tests which give tbsm more insight? After my own search, however flawed these tests are, l l
these and other c-r- *===l~ -8 experiments are the best we have. My other mmment is about the initial conditions. 'Ibe camments on pass 4-26 indicate that the fuel mass and flowraw, but l
there is a problem as far as I can tall. 'Ihe mass is cartuct, but the initialjet size is not I can but 3 cm and I think the flowrate of 1 kg/sec is too low by at least 50%. Irinally, the fbal particle j
semperatures are different for each of the tests noted as is the location and timing of the trigl 1
I am not sure that the authors era aware of this. I can send them this information if needed, but in the case of KROTOS 38. the initial conditions are not correct; e.g., the jet size is 3 cm and l the trigger time is 1.12 see at or near K3 and not at the leading edge, with a pour time of abouc
. 0.75 sec.
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- 2) 'Ihe concept of using the parametric mixing model for a p value of 30 or 50, again raise
- _ question of what is approprises and why. 'Ihm kinamarina in Pisure 2 don't have any c iepe to the thermocouple data for position of the malt u a function of time and give no indicatina what 50 is "better" or more correct than 30 for a value. Also what is the time evoluti particles as ths jet breaks up froaa 3cm to what sise? None of this is discussed at all.
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' SENT BY:UW-NADISON i
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- 3) '!he final point is the uns of the parameter, xf = 0.5 to 1.0. Does & paremment mean that l
when the value is 1.0 all the fuel is quenched as it i- fragmented with some fracdon of water l
and steam? If that is the cormetini-a Mon then, tus pressure plots do not sneen to make l j sense to me. This is especially the case, since the predicted void in by Pigure 2 and Mgure 3 is i
s very small. 'Ihere is something missing in the description; since 1.5 kg of molten alumine has l the energy of 6-7 MJ and thus must be quenched by almost all the 35kg of coolant if there is to l '
be such a 'amall' pressurization with such little void. How much water is " assumed" to be l
intermixed with h fuel to give the pressure signature we ses? ' Ibis is never discussed and it is i
- the mest crucial post of the model 'Iha complete picture is missing and thus, I am not prone
- to agres this is a reasonable prediction until all the ' parameters' are specified and explained. i
' Also, wog sisons to more than one test is needed. ' Ibis has been done with other PCI models. !
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J',re10-1997 12:da FRCM LCSB ENGR. RESEfeCH CTRS. TO 65A u:620t35266249 P.01 If 5f4)
. December 30,1996 post-it" brand fex transmittal memo NTI ******9 L. W. Deitrich, Director w cj g eg = gMM Reactor Engineering Division 5 o Argonne National Laboratory . =, g g.g i 9700 South Cass Avenue l Argonne, IL 60439 **208 5 4 6249 ** 8CG 893 4927
Dear Dr. Deitrich,
j I have enclosed my review of
" Lower Head Integrity under In-Vessel Steam Explosion Loads,"
by T. G. Theofanous, et. al., DOE /ID-10541, June,1996.
A copy already was sent to Prof. Theofanous to help in preparation of his responses to be l included in a forthcoming document, which willinclude comments from all the reviewers.
i Thank you for an opportunity to review this work. I t}unk it is one of the most l
significant pieces of research I have ever reviewed. It is of both current and long-tenn l' importance to the nuclear industry.
Since I have spent my career in the nuclear energy business, I personally appreciate your long range viewpoint for energy needs, which is obvious from your support of this program.
l Yours truly, fAJ48 As '
Y 1
Frederick J. Moody 2265 Sunrise Drive ,
l San Jose, CA 95124 (408) 377-7900 (H)
(408) 925-6434 (W) copy: T. G. Theofanous (
)
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l)CSB ENGR. REh..-1 CTRS. TO 65333892085266249 P.02
- JPN-10-1997 12:40 FRD1 i
' " Lower Head Integ, sy under In-Vessei Steam Explosion Loads" -
by T. G. Theofamous, et. al., DOE /ID.10541, June,1996 1
Reviewed by: F. J. Moody The purpose for reviewing the subject report, with several other companion documents, was to assess whether "in-vessel retention" is damanatra'=d t o be an effective 3
severe accidant manneement concent for a remetar like the AP600.
- I have reviewed the work, and conclude that in-va==al r*=a+ inn han been shown to be an effective severe accidant mans-- ant ccEc=e for reactors with steommerv. fluid ouantities. event secuencine and thermochysical orcesties aimilar to those oer'ainine to
' the AP-600.
The documents provided for this review describe the steps taken to understand and predict the complex, multi faceted subject of steam explosions. Associated phenomena have been closely simulated by experiments, and predicted with determmistic theoretical formulations (causal relations) to a degree of accuracy that makes confident predictions possible for full size AP-600 systems. It appears that all controlling physical effects have l
been included, even without the need for a complete understanding of the exact timing and l
conditions necessary to trigger steam explosions. Already known or conservatively estimated ranges have been placed on parameter, timing, and scenario path uncertainties, and stillit has been shown that the expected range oflower head steam explosion pressure )
j i loads do not intersect the vessel fragility curve.
L
' I was asked specifically to review the material on steam explosion loads, as discussed in f
$ " Propagation of Steam Expk,3 ions: Esprose.m VeriScation Studies" i
by T. G. Theofanous, W. W. hm,10. Framman, & X. Chen, I DOE /ID-1050',, August 1996.
E The documents provihd for this review collectively lay an extensive foundation of i information, which testifies to the techm .al stature, competence, thoroughness, and integnty of the investigators. Indeed, the overall work is monumental in its scope and achievement, and it is communicated in a uriting style which is one of the most scholarly to be found in reactor safety studies. Bout the authors and sponsors should be commended for a carefbtly formulated investigative strategy (strong, in-depth, well-blended steps) resulting in the highest value obtained for the time and resources spent.
i Beyond steam explosions the progress and understanding achieved in this work are likely 1
to exert a major beneficial in8uence, both methodological and t+t#=! on other .
significant .nd complex thermal-hydraulic issues.
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< .Tteia-1997 12'41 FRCM UC5B EN3R. RESEARCH CTRS. TO 65333892085266249 P.03 l
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SUMMARY
- 1. The ROAAM has shown that vessel loads, resuhing from a comprehensive range of severe accident scenarios, melt conditions, relocation flow, timing of release from the core region, and thermal-hydraulic processes between the melt and surrounding water, 4 lead to the conclusion that vessel failure is " physically unreasonable" in an AP-600 type i
reactor. Parameters including pool geometry, melt release rate, shock explosive formation and propagation, and venting yield load distributions on the vessel wall which were i compared with the fragility curve in order to arrive at this conclusion. It is my opinion
- that even though all the mechanisms contributing to steam explosions are not fully
- understood, results embrace the extent of rahern=*
- which could eventually be made by l further experiments and theoretical model (causal relation) development.
- 2. I agree that it would be useful to obtain data from the QUEOS experiment 4 for a fully saturated water system, altbough it would not change the conclusion that vessel failure in AP-600 type reactor;i:" physically unreasonable." The value in such a test is to fill in a parameter range to give a more complete data base, and permit the technology to
- be extended to non-d-600 type systems.
- 3. One potential benefit of the ROAAM procedure is that it conceivably could be used in reverse. Suppose it was concluded that a system failure probability was larger l
than acceptable. The ROAAM could be employed to display which parameter (s) dominate the outcome, thus pointing the way for design or procedural changes to reduce i the failure probability.
I 4 How does the ROAAM accommodate different canut relations, such as
! PM ALPHA and ESPROSE.m, at different stages in the methodology if they might be strongly coupled through common variables? That is, the behavior of two systems alone
!, may be altogether different when they are coupled together (like two spring-mass l systems). The probability distributions of the parameters involved may combine l
differently when the reparate systems are strongly coupled, leading to different probability
- ranges on the variables which deternae success or failure of a system or process.
S. The source tr.rm fbr area production in Appendix A of DOE /ID-10503 is based on the assumption of particle number density remaining constant, while their size changes. A bit more explanation orjustification would help. Ulouldn't it make more sense to predict interfacial area growth by the formation of n ore particles as the melt decelerates in water? Taylor instability was employed to obten the Bond number criterion in interfacial area growth. Could that model be employed to obtain a fastest growing i wave length and droplet formation? ;
I
10504, flow regimes are identified by steady state correlations. Are these likely to be i nonrepresentative for such transient events u fragmentation, and not provide a conservative characterization of the actual heat transfer?
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N 1997 12841 MON UCSB BG. RESEMCH CTRS. 70 65 E 892085266249 P.04 ,
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- 7. Convective and radiative heat transfer from the fbal to the coolant is 1
estimated in much detail, drawing from various experimental studidowceo coolant and heated solid surfaces. Is there a backup analysis to show that for the ry!d heting 3
associated with steam explosions, the heat transfer is not limited to how fast it tm escape from the molten particles? Are there potential droplet sizes, relative velocities, nd fluid properties where internal conduction (or convection) might timit the heat excb se rate?
STRATEGY The severe accident management strategy addressed involves tM re% ion of core material in the reactor vessel following a postulated severe accident w a seactor like the AP-600 design. Inability to cool the core leads to malting ofcore material by decay heat,
- and relocating it in stages to the reactor pressure vessel (RPV) lower plenum Molten core debris, which may flow to the bottom of the lower plenum can melt through the RPV '
wall and undergo release to the containment. However, flooding the cavity to submerge the RPV bottom head is expected to be a means of arresting the downward relocation of i molten core debris.
Even if downward relocation of molten debris is arrested, there is the possibility 4
that some mass of debris could drop into water present in the lower head region, causing a i steam explosion and fbrther damage. Part of the overall study shows that failure of the bottom head by exceeding its structuralintegrity is" physically unreasonable".
4
- TIIE RISK ORIENTED ACCIDENT ANALYSIS METHODOLOGY (ROAAM)
A primitive method of handling uncertainties in power systems came in the early 1960's (Moody, F. J., " Probability Theory and Reactor Core Design," GE Report #
GEAP 3819, US AEC Contract AT(04-3)-361, January,1962). One of the greater l concerns for a nuclear core during normal operation was reaching the " burnout"
! condition, where a hot spot in the fbal could exceed design limits, and cause fuel damsge.
The fuel temperature could be expressed as a fbaction of several variables and parameters (causal relations), each with its own degree ofuncertainty. If one chose the most pessimistic limit of each variable and parameter, the " burnout" limit could be exceeded. !
- The most optimistic limits the " burnout" limit would not be exceeded. It was suggested that probability methods could be applied to give a reasonable assessment of the likelihood of exceeding the " burnout" limit. Data from power plant operating loss was gathered to obtain probability distributions for certain variables and parameters. Wherever data was not available, " expert opinion" was solicited. The results were then combined by the i
method proposed in an ASME paper (Kline, S. L, and McClintock, P. A.," Describing Uncertainties in Single Sample Experiments," Mechanical Engineering. January,1957),
which resulted in the expected mean and standad deviation for the hot spot temperature.
, Comparison with the established design limit showed that it was " physically unreasonable" to expect " burnout" in most cases.
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65333892065266249 P.05 t J4-10-1997 12842 FROM UC5B ENGR. RESEERCH CTRS. TO i
[ 4 The RO AAM is an extensive, opera:ional methodology which is more refined than any ofits primitive predecessors. It has the capacity for incorporatmg causal relations (describing equations relating the variables and parameters), based on weil-understood physics for the applicable phenomena, with specified parameter uncertainties, scenario
' bifurcations, and even a diversity of expert opinion. The process leads to a rationally-based prediction of those properties which deteradne the success or failure of a system or
- process.
j The structure of ROAAM embraces the current phenomenological state-of-the-art, i
built-in activation response of safety and control systems, man-machine interfaces, and 4 procedural understanding. As new information becomes available, the ROAAM can
- accommodate it. Where expert opinions may be diverse, the ROAAM provides a means
. of focusing further research to narrow the disagreements. That is, when experts strongly j disagree on the range of a parameter, the ROAAM can be employed as a tool to display '
the sensitivity, showing if the parameter dominates the outcome, or is only a minor l percentage effect on the overall result.
One question about use of the ROAAM involves the causal relations for various i
phenomena. If the parameters in a causal relation are independent, their probabilities can n
be combined in a certain way to obtain the expected mean and standard deviations of that i j function. If the parameters are not independent, the combination is more complicated.
l The question involves how the ROAAM accommodates the possibility that some 1
parameters appearing in more than one causal relation may not be independent. How would results from ROAAM compare with one deterministic mega-computation where all l I
- the parameters are treated by something like a monte-carlo process to obtain the i distribution of variables which determine success or failure of a system?
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- ROAAM APPLICATION
- I have seen the ROAAM work in two separate campaigns to close severe accident issues, namely the direct containment heating (DCH) issue for one series of PWR's, and the Mark I liner melt issue for one class of BWR containment. It is appropriate that this methodology should be applied to reach a conclusion on the in-vessel retention severe j accident management concept.
Application to in vessel retention embraces possible scenarios, melt conditions, coolant states, structural properties, debris moung with water, triggersng, explosion wave dynamics, and lower head fragility. Parameter ranges are associated with the amount of participating substances, the timmg of events, event patha, and state properties of various subsystems. Several analytical tools, based on physical modela, provide the causal relations employed, namely PM-ALPHA for enveloping the effect of melt breakup in water, ESPROSE.m for enveloping the effects of fragmentation and microinteractions on steam explosions, and AB AQUS v5.5 for enveloping the lower head failure criteria. The '
computer programs used for causal relations to envelope important variables have been y
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l i compared with other analyses and expenmental data to a level where their predictive capability of the tested parameters does not introduce uncertainties which are significant enough to consider.
The following comments are offered to help substantiate my conclusion that in-vessel retention has been shown to be an effective severe accident management concept l
1 for systems like the AP-600. l 4 MELT INTRODUCTION AND FRAGMENTATION 4
Early predictive models provide core melt scenarios and relocation rates with and r
- without reflood, which can arrest the melt progression. However, the melt state which may reach water in the RPV, and the subsequent breakup and penetration largely i
determine the rate of heat transfer, steam formation rate, and possible shock pressure loads. A quantity of melt arriving at the water can undergo Taylor unstable breakup or droplet formation at the leading edge and Helmholtz breakup or droplet stripping on those
' surfaces with parallel velocity components. The PM ALPHA model has been developed to incorporate the melt and coolant properties, and provide an envelope for the expecte range of momentum, heat transfer, and phase change interactions associated with br for premixing considerations.
l Single panicie acA particle cluster exp=iments have been employed to test predictive capabilities of particle motion and energy transfer dynamics in water (the MAGICO and QUEOS experiments). Particle cloud elongation, steaming, spreading, and mixing with surrounding water are captured by the PM-ALPHA code, which is employed as a causal relation in the ROAAM. Comparisons include particle cloud distortions associated with release door opening time, particle, and void volume fraction contours.
Of particular interest is the pinching of the vapor volume behind moving particles, cause by condensation for the particle introduction into subcooled water. Since the condensation acts to reduce mechanical energy transfer, I agree that it would be useful to
- conduct QUEOS experiments in fully saturated water.
One of the most important considerations in fragmentation is the formation of new i
melt heat transfer area. Appendix A in DOE /ID-10503 describes the " source term" for interfacial area production. Equation (3.69) is based on a change in size of particles for the same particle number density. It seems that before particles have reached a stable size, l they would undergo the formation of new particles. This assumption needs more explanation.
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STEAM EXPLOSION The mechanics of steam explosions are desenbed in DOE /ID-10503, detailing melt introduction to water, interfacial breakup and premixing of debris particles with water, the
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6 effect of voiding around the particles on heat transfer, the triggering of explosions, and j propagation of pressure waves with reflections from rigid mechanical and gas-liquid interfaces. It was earlier found that 1.0 OJ of energy could fhil the lower head. However,-
further understanding has led to a reexamination of the mechanics of steam explosion force generation to determine a more realistic criterion for lower head failure.
It was determined that the AP-600 could withstand 500 bars of pressure for
[ milliseconds without failure. Computations with the ESPROSE program displayed the 1
difficulty in generating such pressure impulses with attenuating phenomena like voiding, which resists triggering, and pressure venting from the water surface. Extensive i
development of ESPROSE have been performed with both data &om the SIGMA and y
KROTOS experunental facilities. Simpler analytical models have provided assurance that ESPROSE accommodates detonations, shock propagation, and reflection. 1 Significant effects embraced by ESPROSE result from the physics incorporated, which are consistent with experiments. Calculations show the strong attenuation of shock l
4 pressure loads with distance, and time by venting from the water free surface in the AP-600 systems. It is also realized that venting may not signi6cantly reduce loads if the water i depth is high in the lower head. Strong evidence is supplied that ESPROSE incorporates the appropriate physics, and can be used with confidence to provide the causal relation for
! enveloping the effect of trigger time on steam explosion severity.
The physical mechanisms considered by ESPROSE.m include shock pressure l
propagation from a trigger, which collapses voids, forces liquid onto the melt, producing l fragmentation and microinteractions, escalated heat transfer, further steam formation, and l
rapid expansion (explosion). A statement on page A-18 ofDOE/ID-10503 needs fiuther claafication. Where the pressure increases rapidly ahead of an explosion front, why does l the vapor become instantaneously subcooled? (If saturated steam is rapidly compressed, it i
would tend to follow an isentropic path off the vapor dome into the superheated region,
- not subcooled.) On the same page, it is stated that behind the explosion front where pressure is decreasms, the liquid can become superheated. (Ifyou decompress saturated water, the path drops into the steam dome.) It would make better sense to me (I can't speak for others) to note that the nonequilibrium states lag behind a steady state in the l
superheated or subecolod region.
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LOWER HEAD RESPONSE j Dynamic response of the lower head is based on well established physics of shells, modeled by the AB AQUS program. Mechanical failure of a shell depends not only on the magnitude of an applied load, but also on the Sequency content. It is stated that the shock pressure loads which lie in the steam explosion envelope have a short period relative to the structural response, so that the peak strain would be essentially Maaaadaat of the ,
pressure pulse time profile.
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The report has provided some " screening fragility" curves which would be used to determine if predicted steam explosion loads were of such a character that the failure ,
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' criteria envelope and fragility curve need to be further blended to provide a failure J t
likelihood. It was concluded from the range ofpressure loads and the lower head fragility curve, that for all relevant severe accident scenarios, melt conditions, and timing ofrelease from the core region, with ensuing mixing and explosion wave dynamics, steam explosion induced lower head failure in an AP600-like reactor is " physically unreasonable."
I REVIEW OF STEAM EXPLOSION LOADS ,
The verification of ESPROSE.m, based on stepwise experimental measurements and comparison with simplified theoretical methods shows that reasonably conservative assessments of steam explosions are possible in the present version.
l The discussions of DOE /ID-10503 provide foundational support of the physical modeling and numerical procedures to predict steam explosion properties for given melt addition rates and states. The basic physics involve wave dynamica, iWiag sound wave propagation and shock development and propagation in a water-filled region. Two-l
' dimensional calculations performed by ESPROSE.m are compared with simplified computations using the method ofimages and rolutions similar to classical waterhammer.
Some comparisons are included based on characteristic solutions. The results fonn a l
strong basis for concluding that the code is producing reasonable predictions for the expected range ofinput parameters. Pressure propagation speeds, attenuation from wave interaction at free surfaces, and wave amplification by reflection from rigid surfaces have all played a role in the verification. I I Numerous two-dimensional ESPROSE calculational surfaces are compared with solutions obtained from the method ofimages, and found to be sufficiently similar, leading to the conclusion that basic physics of explosions are included in the model. Several geometric parameters were varied, as was the source velocity function. Good comparisons were consistently achieved.
The SIGMA tests involved a melt droplet which was triggered at a specific
! l position, leading to local pressure traces. Comparison of the pressure traces with ESPROSE calculations showed reasonable tracking of pressure waves originating from l the droplet region to the rigid end of the test section, and re6ection back toward their j origin. Additional evaluation with the method of characteristics were provided. The wave i
dynamics, indeed, appear to be properly described in ESPROSE. 1 One piece ofinformation lack pointed out in the report is that the data base needs
- expansion for microinteractions with reactor materials.
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THE NEXT STEP i
j I understand that a number of experts are providing reviews of tlw documents provided. Some may believe (as I do) that even without a complete understanding of all i the phenomena, the remaining uncertainties, processed by the ROAAM, still permit a strong statement about failure likelihood being " physically unreasonable." Some experts i may feel that the uncertainty of a given parameter should be broader. This is a simpic
! exercise in ROA AM, which would then provide output with a range that accommodates
- the particular variable uncertainty. Other experts may wish to change the causal relations
! to reflect various " bottom up" or fine stmeture effects. This is always a possibility, but may be unnecessary, since the causal relations are based on macroscopic formulations of j i basic principles. Ifit were recommended that nonequilibrium models be employed for )
l causal relations, we would be farther behind than using ROAAM in its present structure, l because nonequilibrium models would have to be verified by experiments.
When strong disagreements have resulted in physical modeling, small working groups have been formed to reach agreement on acceptable formulation, with appropriate modifications in ROAAM.
Finally, it is possible that some would disagree with the ROAAM structure itself, suggesting that it skews results, or simply blurs our ignorance of phenomena. I would argue strongly that the ROAAM blends (not blurs) uncertainties (not ignorance) in a way that makes it possible to reach conclusions with a known level of confidence, i
OVERALL CONCLUSION 4
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! As a curious person who enjoys for=dadag better theoretical models, based on J i more complete experimental under=*arHae I r+x- --e=-i additional expenments (e.g., )
QUEOS experiments with fbily saturated water) to help close the few rernaining gaps in our understanding of steam explosion phenomena. l However, I believe that the studies provided for this review give substantial, in-depth evidence to help conclude that in-vessel retention is supportable as a severe accident management strategy in AP-600 type reactors without additional work to close the issue.
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10/13 96 FRI 13:18 FAS 1 630 232 4730 OL-RE *** S. SURRELL 2002 l g 5 6 UNIVERSITAT STUTTGARTINSTITUT FOR KERNENE Abteilung Reaktorsicherheit und Umwelt I
Stuttgan VaiNngen, den 02.10.1998 IKE Pfaffenwaldring 31 7055o Stuttgart Pf affenwaldring 31 Dr. L. W. Deitrich Telex 7 2ss 44s unN d Director, Reactor Eng. Division @ $jjjjs No* ,
Argonne National Laboratory l 9700 South Cass Avenue Unser zeichen: ts Argonne, Illinois 60439 Ihre NacMcht vom:
ihr Zeichen:
U. S. A.
i Review of the report ,,l.ower Head Integrity under in Vessel Steam
Dear Dr. Deitrich:
Mr. :
by order of my colleague Manfred Burger I send you encrosed l Burgt is on holidays until October 21. If you need Burger). further inform conta.:t me or my co!!cague Eberhard von Berg (same phone and I J
i With best regards E E C E }'/ ,i O l i
REACTOR ENGINEER:Ta D: ' "0N
-DIRECTCP/5 0.- :. ~ ,
MichaelBuck 0CT S 19 9 S l
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