ML20081A457

Radionuclide Release Under Specific LWR Accident Conditions, Vol V, PWR-Large,Dry Containment (Surry Recalculations)

ML20081A457
Person / Time
Site:	Surry
Issue date:	01/12/1984
From:	Cybulskis P, Denning R, Gieseke J Battelle Memorial Institute, COLUMBUS LABORATORIES
To:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
RTR-NUREG-1150-2-V2-B.42 NUDOCS 8403050230
Download: ML20081A457 (212)

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b'l Ct,/43-[(i74 RADIONUCLIOE RELEASE UNDER SPECIFIC LWR ACCIDENT CONDITIONS -- VOLUME V i PWR-LARGE, ORY CONTAINMENT (SURRYRECALCULATIONS) j Prepared for i

1

-! 0FFICE OF NUCLEAR REGULATORY RESEARCH U.S. NUCLEAR REGULATORY C0!HISSION Washington, D.C. 20555

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.j By i

1 BATTELLE'S COLUMBUS LABORATORIES Columbus, Ohio 43201 JA Gieseke, P Cybulskis, RS Denning,

, MR Kuhlman, H. Chen, and KW Lee January 12, 1984

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, 2. INTRODUCTION I

a l This is the fifth volume of a report on "Radionuclide Release f Under Specific LWR Acddent Conditions" and is concerned with analyses performed for a PWR-Large, Dry Containment. Specifically, the Surry

.l plant has been considered with cair;dations being repeated for most of l

the cases previously analyzed and reported in Volume I of this report, i-i but in the carrent analyses, improved analytical procedures were used. '

j The major changes in the analytical procedures have been the use of the MARCH 2 rather than the MARCH 1.1 code, improved representations of upper l plenum geometry, revised models in the COR5JR code treating the release

,l of tellurium and control rod materials,* expanded models in the NAUA-4 1 code to include homogeneous nucleation of water from the vapor phase, Il and consideration, for some cases, of multiple compartments ir the

.' containment. The analyses reported in this volume were performed with m the same versions of the various computer codec as were used in Volumes II and III with the exceptions of modifications noted above and described i in Chapter 5 of this volume. For continuity in reporting, a portion of

_ ] the Introduction from Volume I follows.

.l The radiological effects associated with fission product

! release from consnercial light water reactors during severe accident con-l ditions have been the subject of considerable concern and the impetus for much research. As research has progressed, the physical processes

controlling the magnitude of fission product releases have becone more j thoroughly understood and the ability to estimate fission product releases has been improved.

The design bases and siting criteria for most of the existin

$ population of U.S. reactors were based on the use of the TID-14844(2.1 assumptions regarding the release of fission products to the ccntainment

{ .in a severe accident. Although representative of the state of knowledge U at the time, a better understanding of the behavior of fission products in severe accidents has developed over the intervening years and many of the TID-14844 assumptions are now recognized as requiring re-evaluation.

A mere mechanistic treatment of fission product release was developed b for the Reactor Safety Study (WASH 1400)(2.2) and since that time the 1

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[j WASH 1400 source term to the environment for accident sequences has been j used extensively. Obtaining an improved characterization cf the scurce d of fission products to the environment in accidents is therefore an essen-Il tial stec in the comprehensive evaluation of current source term assump-tions and would serve as-a basis for formulating impacts on and changes j to licensing practice, emergency planning, safety goals and indemnifica-j tion policy. For this reason the NRC undertook a review of the state of-

,{! knowledge regarding procedures available for predicting fission product release and transport and in June of 1981 issued the report " Technical-f Bases for Estimating Fission Product Behavior During LWR Accidents".(2*U 1

As part of the " Technic 11 Bases Reportd, the assumptions, proce-h dures, and available data needed for predicting fission produc.t behavior

( were evaluated and calculations were made of fission product attenuation M) along the various flow pat'is from the fuel to the environment. Because

[ of the limitations of available computational tools at that time, release from the fuel and transport through various compartaents along the flow i path were treated separately and therefore oossible interactions were

' f.! r.ot considered. This procedure is the subject of the first major coment.

, .f on the " Technical Bases Report" (NUREG 0772, Appendix F) and was recog-

~ [. nized as a shortcoming of that report. The calculations and evaluations 2-being presented here are intended to overcome this shortcoming as well

? as to proviJe analyses performed with in' proved computational procedures.

h This report builds on previous computer modeling work performed 5 at Battelle-Columbus, Sandia, Batte11e's Pacific Northwest Laboratories,

and in the Federal Republic of Germany, and on experimental and model y evaltatien work performed at Oak Ridge, EG&G Idaho, Sandia, and Battelle's .

Pacific Northwest Laboratories. It is to be noted that in additioh to y the calculations performed at Battelle-Columbus, calculations concerned

'with the themal as well as the fission product release aspects of molten j4 core-concrete interactions were performed by Sandia. Research efforts

,] , specifically directed toward increasing oue understanding of fission 4 product release ano transport under severa accident conditions are '

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currently under way at the laboratories listed above as well as at other j,

.; research installations around the world. It is anticipated that over h

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-j the next few years considerable progress will be made in this area.

{ Therefore this report must be considered as an expression of current knowledge with the expectation that validation or modification of the calculated fission product releases will be forthcoming.

It is to be recognized that this report de~s cribes an analytical q approach for estimating radionuclide transport and deposition which d

incorporates physical and chemical processes on a mechariistic basis. ,

This approach is being evaluated for use in predicting fission product q source terms for release to the environment on a specific case-by-case j - basis (reactor, accident sequence) and.when verified would be expected to replace the generic tibular release fractions such as those in .

Table 6, Appendix V, WASH 1400 where release fractions are given for

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.i broad classes of accidents.

4 The purpose of this report is then to:

(1) Develop updated release-from-plant fission product source terms for four types of nuclear power plants and for acci-l '

i dent sequences giving a range of conditions. The estimated sour'ce terms are to be based on consistent step-by-step analyses using improved computational tools for predicting radionuclide release from the fuel, and transport and j deposition.

(2) Determine the ef fects on fission product releases associated

! with major differences in input parameters associated with plant Jesign and accident sequences.

(3) Provide irt-plant time and location dependent distributions of fission product mass for use irt equipment qualificaticn considerations.

i It is not necessarily the intent of this work to produce an all

encompassing definition of source terms but rather to make best est. mates ,

4 of source terms for a ra.1ce of typical plants c.nd severa! risk-significant sequences covering a wide range of conditions. These analyses were to

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be made with the best available techniques, in a consistent manner follow-ing along with release pathways for fission products, and at a level of l{

detail consistent with current knowledge of pertinent physical processes.

Based on state-of-the-art techniques, these best-estimate analyses shnuld

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provide an indication of the conservatisms inherent in current source ter1n assumptions and guidance for the development of new source terms.

It is important to note that the analytical methods and corresponding

predictions are based on currently available information and are subject

- ]1 to 7 vision and improvement as better analytical procedures are developed

}l and as a more extensive experimental base evolves. The preparation of j this' report, therefore, is an evolutionary process which will be carried; 4

'l out cver a period of time with verification and possibly revision of the procedures continuing over several years.

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References (2.1) DiNunno, J. J., et al, " Calculation of Distance Factors for Power and Test Reactor Sites", TID-l'4844 (March 21, 1962).

(2.2) " Reactor Safety Study -- An Assessment of Accident Risks in

{ U.S. Commercial Nuclerr Power Plants", tiASH-1400, NUREG-75/014 (October,1975).

'2.3) dTechnical Bases for Estimating Fission Product BeNvior During

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LWR Accidents", NUREG-0772 (June, 1981). +

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. 3. GENERAL APPROACH i

f The general philosophy behind this study is that mechanistic predictions of radionuclide release and transport are possible if proper j modeling is performed to represent the physical and chemical processes occurring during LWR accidents. The study, then, represents an attempt at describing in a reasonably complete but tractable fashion the proces-j ses influencing the radionuclide release to the environment for selected' 1 .

plants and accident conditions. The general approach taken in this study

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was specified by the cbjectives which called for a consistent analysis -

of radionuclide behavior by following their transport along flow paths

.< from their release into the core region up to their final release to the d environment. Nevertheless, numerous decisions and assumptions were d

required for the analyses. These decisions included selections of plants

< . and sequences *for consideration, choices of analytical tools to be used

? as available or upgraded, evaluations and incorporation of experimental data, and determinations of major physical effects which would be consi-dered on a parametric variation basis to illustrate the sensitivity of j calculations to such variations. Such decisions and assumptions are discussed throughout this repnrt as they arise in their technical con-text. However, some of the major considerations will be reviewed and J the steps comprising the overall approach will be discussed in this sec-

. tion.

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This study was based on selecting specific plants and accident sequences for consideration and then using consistent and improved analy-ses of fission product release from fuel, transport, and deposition to f predict fission product release to the environment for these specific I, cases. The approach is comprised of a series of steps performed in sequence such that in the combined analysis, the results are specific to

. an individual set of accident conditions and each successive transport st6p is based on results from analyses of the previous step.

h The first major step in the process was the selection of types  :

of nuclear power plant designs to be considered and a specific plant to }

represent each type. The types to be considered were: large, dry PWRs; i

Mark I BWRs; Mark III BWRs; and ice condenser type PWR designs. The I i

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3-2 specific plants chosen to represent each type are the Surry and Zion,

Peach Bottom, Grand Gulf, and Sequoyah plants. These selections were j made on a combined basis of typicality of design and availability of j design details needed for analyses.

, , Accident sequences are being chosen for each plant design based on risk and on a desire to have a range of physical conditions represen-ted by the analyses. The plants selected and the accident sequences .

j considered are listed below: -

.-ru j PWR Large Dry PWR Large Dry Containment Containment BWR Mark I (Surry) (Zion) (Peach Bottom)

AS TM.B ' TC 50 2 50 2 AE V TW

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Following the selection of plants and sequences the required

plant design data were collected and thermal hydraulic analyses performed for the accident sequences. Overall thermal hydraulic conditions on a time-dependent basis were estimated witn the MARCH 2 code,(3.1) and

. i detailed thermal hydraulic conditions for the primary system estimated

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with the MERGE code which was developed specifically for use in this

_.g program. The MERGE code was described in Volume I of this report.

{ The time-dependent core temperatures we're used as input to y.; ,

another code developed for this program, CORSOR, which predicts time and N temperature dependent mass releases of radionuclides from the fuel within

[ the pressure vessel. The latest version of CORSOR is described in Volumes

,..- I and II and Chapter 5 of this volume. Releases during core-concrete l

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interactions of radionuclides remaining with the melt were provided by Sandia National Laboratories using their computer code VANESA.

Using the MARCH / MERGE predicted thermal hydraulic conditions and the CORSOR predicted radionuclide release rates as input, a newly developed version of the TRAP-MELT code, described in Volume ~I, was used to predict vapor and particulate transport in the primary coolant circuit.

. Transport and deposition of radionuclides in the containment -

!' were calculated using the NAUA 4(3.2) code with modifications as noted in Volume I and in Chapter 5 of this volume.

The basic stepwise procedure described above is illust: ated in

.j Figure 3.1 which shows the relationships anong the computational models. .

The calculations were of a "best estimate" type using input derived, to the exte.1t possible, from experimental measurements. Types of data employed in the analyses include vapor deposition velocities, acrosol deposition rates, aerosol agglomeration rates, fission product release j]

rate's from fuel, particle sizes formed from vaporizing / condensing fuel I

materials, engineering correlations for heat and mass transfer, and physi-cal properties of various fuel, fission pr,oduct and structural materials.

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50.ECTION OF TYPES

,1 0F PLANTS I

'r SELECTION OF

$PECIFIC PLANTS v

l SELECTION OF .

4 ACCIDENT SEQUENCES

,e SPECIFI*ATION OF PLANT INVENTORY GEOMETRY AND ACCIDENT -___----

SEQUENCE PHENOMENA ORIGEN

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t i FIGURE 3.1. INFORMATION FOR RELEASE, TRANSF3RT AND DE':0SITION CALCtlLATI0d .

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References l

(3.1) Wooton, R. O., et al, " MARCH 2 Code Description and Users' I

Manual", Draft (Decenber,1982).

[i (3.2) Bunz, H., Koyro, M.,

and Schock, W. , "A Code for Calculating

.i j Aerosol Behavicr in LWR Core Melt Accidents, Code Description

'd Users' Manual".

.j (3.3) hinegardner, W. K., Postma, A. K., and Jankowski, M. W., '

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" Studies of Fission Product Scruboing Within Ice Compartments",

NUREG/CR-3248 (May, 1983).

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4. PLANT SELECTION AND ACCIDENT SEQUENCES 4.1 General Plant Description j A pressurized water reactor with a large high pressure contain-

{ ment design is the first reactor type to be considered M this study. In I

actuality, there are large variations in the design of large hi.}h-pressure -

containments in terms of the volume of the containment building and the design pressure of the containment. To some extent, generic accident i sequences can be defined which are independent of the variations that exist l in the nuclear steam supply system and balance d plant designs. For

.j example. the sequence A8-4 (-large LOCA, loss of all AC power and failure of l] containment by overpressurization) could occur in any PWR design. This.is

.j because the general safety philosophy and safety functions provided to pro- {

}l tect the plant are the same. Because the different vendors and architect-enginaers have chosen different approaches to sabisfy these safety functions, the likelihood of each sequence may vary greatly between plant design; simi-larly, the accident timing and sequence of events in a sequence may :.lso

vary depending on the design.

The specific plant design selected to characterize large high-

/l pressure designs is the Surry Unit 1 plant. In s ie respects, the selec-tion is rot ideal. The Surry plant is an older design. In comparison with

=l i ! average parameters for U.S. designs, the power output is low, the contain-t ment volume is small, and the containment design pressure is low. By the i

use of parametric variations, it has been posrible to examine some of the t important differences in containment design, however. An important reason for selecting the Curry plant was that this was the design analyzed in the j Reactor Safety Study. Thus, a direct comparison can be made betwean the magnitudes of the predicted source terms. f

. . The Surry unit is a pressurized water reactor (157-inch diameter I l' vessel with 157 fuel assenelies) designed by Westinghouse. A detaile.d  ;

f- description of plant data is provided in Table 4.1. The reactor is designed '

) to operate at a nominal power of 2441 MW(t) and a reactor coolant system

,l pressure of 2250 psia. The containment is a steel-lined reinforced t-i 4

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.l .1 TABLE Ll . PWR DATA

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9 Nominal power 2,441 Mwt 8,331 x 10 Btu /hr 4

)d Internal energy of water 246.9 x 10 Stu 2.66 x 1010 kg-m)

Sensible heat in the core 16.35 x 10 Stu 1.76 x 109 k

-; Total water in the system 423,200 lb 192.000. kg) g-m)

'l Aug. tanperature (Excl. pres.) 571.8 F ' 300 C) l Pressure . 2280 psig 15.7 MPa)

Reactor coolant system volume (387 ft3 237.5 m3 .

Pressurizer volume, total 1,336 ft3 (37.83m3) water 816 ft3 (23.1.m3 ) '

J/K

. steam 520 ft3 (14.7 m3)

Three eccumulators, total volume 4,350 ft3 (123.2 m3)

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:1un 2,775 ft3 78.58 m3)

. pressure 665 psig 4.585 MPa)

J' temperature 120 F 48.9 C)

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Containment recirculation spray 2 systems, flow each 3,500 gpm 5

y Containnent free volume 1.8 x 106 ft3 (5.1 x 10220.8 4 m) 1/s)

Initial temperature 100 F

Initial preswre 10 psia 3.778 6.89 x C) 104Pa) 2' Dewpoint -

80 F 26.7 C)

Primary system hot metal ) 686,285 lb (766,000 kg)

'% Temperature 572 F (300 C)

'5 , Containment tient St.ks Thickness Area

. Walls inside containment 1.0 ft (0.3048 m) 3,320 ft2 (308.4 m2) j: Walls inside containment 2.0 0.6096 27,600 2564)

. Walls inside containment 3.0 0.9144 19,400 1802)

..; Walls inside containment 4.0 1.219) 5,000. 464.5)

!' Walls inside containment 6.5 1.981 2,100 1 95.1 )

Contaf'$~snt wall 4.5 1.372 46,747 4343)

.'a Dome 2.5 0.762 25,000 (2323) {

'.k Floor above foundatio:: mat 2.0 0.6096) 11,250 (1045) 1

.1 Foundation mat 10.0 3.048) 11,250 (1045) i

,W Containment liner . j p.h Walls 0.38 in. (0.9652 cm) 46,747 4343) 1

,6; Dome 0.50 in. (1.27 cm) 25,000 2323)

'ey Floor 0.25 (0.635 cm) 11,250 1045) o Miscellaneous metal - 1,200,000 lb (544,308 kg)

M, Core

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Equivalent diameter 119.7 in. (3.04 m) 4p Active height 144.0 in. (3.658 m) e L/D 1.202

. T: Total cross sectional area 78.' ft2 (7.27 m2) y M

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TABLE 4.1. FWR DATA (CONTINUED)

Core (Continued)

No. of fuel assemblies 157 Rods per asscably 204 Pitch 0.563 in. (1,430 cm)

Assembly dimensions 8.426 in. square ,

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Fuel rod diameter 0.422 in. (1.072 cm) s Clad (Zr-4) thickness 0.0243 in. (0.0617 cm)

Total number of fuel rods 32,028 Core weight 226,200 lb (102,700 kg) 002 175 600 lb 79.820 kg)

Zircalcy 36,300 lb 16,500kg) e Misc.~ 14,300 lb

! Fuel pellet diameter, Region 1 0.3669 in. 6,500kg))

0.9319 cm 2 and 3 0.3659 in. 0.9294 cm)

.' Fuel pellet length 0.6 in. 1.524 cm)

, Diametral gap, Region. 0.0065 in. 0.01651 cm) 2 and 3 0.0075 in. 0.01905 cm)

l Fuel density, Region 1 94%

4 2 92 3 91 Fuel enrichment, Region 1 1.85 w/o 2 2.55 3 3.10 No. of grid spacers 7 i Neutron adsorber Ag-In-Cd

! iij Clad 304 ss Clad thickness 0.024 in. (0.06096 cm)

No. of control assemblies 53 Full length 48

.4 Part length 5

,d' Rods per assembly 20 Burnable poison rods 816 No. per assembly 12 No. of assemblies 68 1

Material Snros111cate glass 0.D. 0.4395 in. (1.116 cm)

I.D. 0.2365 in. (0.6007cm)

Clad 304 ss Soron (natural) loading 0.0429 g/cm U Reactor vessel l^ I.D. of shell 157 in. (3.99 m)

L'- 7.875 in.

8elt line thickness (w/o clad) 0.2000 m) j1; Head thickness 5.0 in. 0.127 m) ij Clad thickness 0.125 in. 0.3175 m) ld overall height 40 ft-5 in. (12.32 m)

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4-4 TABLE 4.1. PWR DATA (CONTINUED) l Reactorvessel(continued)

Inlet nozzles 27.5 in. (0.699 m) tapered to 35.4 ia. (0.899 m)

Outlet nozzles 29 in. (0.737 m)

I i Water volume with core and -

l internals in place 3,718 ft3 (1.053 x 102m3) -

l Core barrel I.D. 133.9 in, f O.D. ~

137.9 in. (,3.401 3.503 m) m)

L*t Thermal shield I.D. 3.622 m) 0.D. 142.6in.in.(

148.0 (3.759 m)

Safety Injection Charging Pumps

,.. Number 3

r. Design pressure, discharge 2750 psig 18.96 MPa)

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Design pressure, suction 250 psig 1.724 MPa)

Design temperature 250 F 121 C)

Design flow 150 gpm 9.461/s)

L/I Maximum flow Design head 600 gpm

. 5800 ft (37.8 1/s)

(1768 m)

,A Low Head Safety Injection Pumps Number 2 Design pressure, discharge 300 psig 2.07MPa) 4 Desigri temperature 300 F 148.9C)

Design flow .3000 gpm 189.2 1/s)

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. Design head 225 ft 68.58 m)

Maximum flow 4000 gpm 252.3 1/s)

Containment Spray Pumps

. Number 2 l1 Design flow 3,200 gpm 201.9 1/s) i Design head 225 ft 68.58 m)

4 Design pressure 250 psig 1.724 MPa)

Recirculation Spray Pumps Inside

, Containment

1. Number 2 1d Dasign flow 3500 gpm (220.81/s Design head 230 gpm (14.51/s) ) '

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Recirculation Spray Pumps Outside Containment l'[

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Number Design flow 2

3,500 gpm (220.81/s) i i

,t; Design head 249 ft (75.89m) l

..W Recirculation Spray Coolers

l " Number 4 Desig:. duty, each 55,534,520 Stu/Hr (16.3 MW) a Refueling ' dater Storage Tank 9 Volume 350,000 gal (1.32 x 106 1).

-C Boron con' antration 2,500 ppm i Design pressure Hydraulic head

,j Design temperature 150 F (65.6 C) j i; Water temperature 45 * (7.22 C)

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l 4-5 l i concrete structure with a free volume of 1.8 x 100 ft 3; it is op3 rated 1 subatmospheric with initial pressure and temperature of 10 psia and 100 F, l

} respectively. Figure 4.1 illustrates the layout of the containment design. '

.f The plant systems that perform critical safety functions are depicted conceptually in Figure 4.2.

4.2 Selection Basis and General Description j of Accident Sequenes j.

j The four accident sequences selected for the large dry PWR plant

! design analysis were AB, TMLB', S2 0, and V. Table 4.2 relates the letter

[ used in the accider.t identifier to the type of event and to the failure of

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the engineered safety systects. Two criteria warr used in selecting sequences.

First, it was desired to examine sequences that would be significant contri-butors to risk for a number of design variations within the large high-pressure containment category of PWRs. Secondly, it was considered important to cover a range of accident conditions and engineered safety system j

performance within the reactor coolant system and containment building. ,

Although the large, high pressure PWR containment design is often referred to as " dry", a great quantity of steam would be released to the containment building in each of the sequences analyzed, other than the V sequence in which the release is to the safeguards building. In these cases steam condensation on the walls and on aerosols can have a significant influence in enhancing the natural removal of radionuclides from the volume atmosphere. Even more effective are the containment spray systems in those sequences in which they are expected to operate. Sequences were selected to illustrate the performance of the containment system with and without

• J spray operation.

An aspect of the Three Mile Island 2 accident that played an important role in limiting the release of radionuclides to the containment atmosphere was the presence of a large quantity of water in the pathway of release to the environment. This is not characteristic of accident sequences j I

leading to complete core melting of the type selected for the present l

3 anslysis. In these sequences fuel heatup would not begin until the water level had dropped below the top of the core. Very hot steam and hydrogen

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.q TABLE 4.2. KEY TO PWR ACCIDENT SEQUENCE SYMBOLS

}1 lj A - Intermediate to large loss of coolant accident (LOCA).

I"' B -

Failure of electric power to engineered safety features (ESF).

4' 8' - Failure to recover either onsite or offsite electric power within about.

I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following an initiating transient which is a loss of ~

offsite AC power.*

k C -

Failure of the containment spray injection system.

q. D -

Failure of the emergency core cooling injection system.

n .

L F -

Failure of the containment spray recirculation system.

G -

Failure of the containment heat removal system.

~

H -

Failure of the emergency core cooling recirculation system.

Failure of the reactor protection system.

K -

L -

Failure of the secondary system steam relief valves and the auxiliary l feedwater system. ,

~~

M -

Failure of the secondary system steam relief valves and the power

~L conversion system.

1( Q - Failure of the primary system safety relief valves to reclose after 7 opening.

R - Massive rupture of the reactor vessel. 3

[ 51 - A small LOCA with an equivalent diameter of about 2 to 6 inches.

a ,

S2 - A small LOCA with an equivalent diameter of about 1/2 to 2 inches. .'

T - Transient event.

V - Low pressure injection system check value failure. [

4

'Nl Containnent Failure Modes:

r if a = steam explosion

[ s = containment isolation failure ,

j Y = overpressurization due to hydrogen combustion j 6, = early overpressure failure due to steam and noncondensible gases 1 :,

61 = delayed overpressure failure due to steam and noncondensible gases i

e = basemat melt-through t

.c -

---r---- - , . - - - -

7 _- ,-- ,p

4 4-9 1

leaving the melting core would be expected to superheat the structures in -

ls the pathway to the containment. Other sequences in which the exiting gases would contact water as in tna TMI 2 accident are possible, particularly those involving the partial performance of emergency core cuoling systems.

.6 Depending on the subsequent fate of the water, contact with water would be j expected to be effective ire retaining fission products within the liquid phase. Partial core damage sequences such as the TMI 2 accident would not be expected to be dominant from a public risk viewpoint. -

i

}

, j '

4.2.1 Sequence AB (Loss of Coolant Accident, j Loss of AC Power) j A large pipe b'reak accident was selected for analysis because it '

represents one end of the spectrum of reactor coolant system conditions t

during core meltdown. . Depressurization of the reactor coolant system would be expected to occur rapidly fo.llowing the break. In the case of loss of J

all AC power, the ' accumulators would discharge into the vessel to supply some emergency coolant but the pumped ECC infecticn systems would not operate.

f The water level in the reactor vessel after blowdown would be dependent on the location cf the large pipe break. For a break in *he hot leg piping,

, f the core would likely be fully covered; for a cold leg break, the core may be only partially covered as some of the accumulator water is swept out the l break. A break location in the hot leg rather than the cold leg was selected

~

l !

> in order to examine a case involving a minimum pathway and presumchly minimum i i fission product retention within the reactor coolant system. As the water level would decrease in the core due to continued generation of decay heat, heatup of the fuel and fission product release would occur at the same pres-sure as the containment building atmosphere. The flow path for fission products from the core, to the upper plenum, and to the hot leg break loca- ,

tion is illustrated in Figure 4.3. Flow through the other loops is assumed I

to be blocked in this sequence by hydrogen and possibly by water seals in the low points of the system. The flow path during the vaporization release ,

period .is shown in Figure 4.4.

lj In terms of reactor coolant system response, there would be little

.l ' difference between the cases A8 and AD (involving failure of pumped ECC l'

j

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I l 4-12 1

<l

injection rather than loss of all AC power). The containment conditions if for A8 would be very similar to those of another case analyzed, S2 3, in which the containment sprays are asst.med to be operating. It was felt

, that the A8 case provided an opportunity to examine a number of interesting

j '

containment failure modes in which nai: ural deposition was the retention mode for the released radioactivity.

t

< t The containment failure modes considered in connection with the .

j,f[ A8 sequence included containment isolation failure (8), early overpressure .

','i failure due to hydrogen burning (Y), and melt-through of the containment

,I

.i basemat (c). The latter would lead to the release of radioactivity through (i the ground; long-term overpressure failure (6) would involve similar quanti-ties of radioactivity release, but the path would likely be to the atmos-phere. Tne probability of containment isolation failure is believed to be quite small for.the Surry plant since it is normally operated with the containbrunt at subatmospheric pressure. This provides a means of continu-ously mor.itoring containment integrity. For other PWR containment designs containment ' solation failure may be a more likely and more significant mode of failure. Early containment failure du'e to hydrogan' burning would '

require a coherent burn of a large quantity of hydregen. If the hydrogen burning is spread o':t in time or takes place in a number of smaller combus- i tion events, the likelihood of cor.tainment failure would be very small.

Melt-through of the concrcte basemat and/or delayed overpressure failure would be more likely outcomes for the AB-sequence. The melt-through and je overpressurization would both require relatively long times and the occur-

., - rence of one may preclude the other. Because of the long times involved, i

[ there would be significant opportunity for fission prodact removal by various inherent processes and the magnitudes of potential radioactivity releases for these failure modes would be significantly reduced. i 14 4.2.2 Secuence TM.8' (Transient. Loss of

.f Primary System Heat Ren: oval)

• a b

t The TM.B' sequence was found to be a major risk contributor in

,1

> WASH 1400. The predicted release fractions for the containment overpressuri- l I

] zation failure mode for this sequence were used to characterize release '!

b,

- - - - 7:_ f r. .i a '"i "~TE~Ci ~

Z~"#E~ ~~

..=.-.; . .. .

~ ^

t 4-13 category PWR 2. The reactor coolant system behavior of TMLB' is in sharp contract to A8 because the reactor coolant system pressure remains elevated l (-2500) during core heatup and fission product release. The beginning of

i c re uncovery is also delayed for the time required to dry out the steam j generator secondary side and boil off a portion of the primary system coolant j inventory.

li I The flow path for fissio9 products through the reactor coolant .

j system is. illustrated in Figure 4.5. Prior to core uncovery the water in d the pressurizer is predicted to be carried from the pressurizer with the ,

I i steam flow or to fall back into the hot leg Neing periods when the relief valve is closed. During 'the period of core melting and fission product 1 release from the fuel, the path for fission product transport from the primary system would be filled with high temperature steam and hydrogen; there may, however, be water in the pressurizer quench tank through which i

the high temperature gases and radioactivity must pass. Cc!1 apse of the core into the vessel bottom head would lead to the evaporation of the '

residuaI water and heating of the vessel head. Since the primary system is expected to be at an elevated pressure, failure of the bottom head would

/

. take place due to the combination of pressure stresses and overheating.

i The foregoing describes the primary system scenario for the TMLB' sequence as it has been traditionally analyzed. An alternate scenario that has been suggested for this sequence involves failure of the primary coolant pump seals due to lack of cooling, changing the sequence to a small or intemediate break accident. Or, the pressurizer relief / safety valve could l , stick open after some period of operation, turning the seque1ce into a small break event. Such possibilities have not been explicitly evalcated in the .

present study.

As in the AB case, containment spray; are not available in this sequence due to the loss of electric power. One objective for selecting

, the TO' case was to investigate the effect of contair

nent failure time on l, fission product retention. Early and late failure of the c9ntainment by n overpressurization are considered. The conditions which woulo be required to laad to early failure of the containment involve energe. tic interaction 5 tween molten core material and accumulator water in the reactor cavity.

The magnitude of the ensuing steam pressure spike depends on mixing and I

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4 i heat transfer processes that are uncertain. Although early failure is

! considered unlikely, it was felt that this failure mode shoul.1 be evaluated 1

,j because of the potentially large consequences associated with it. Even at a relatively low containment failure probability, this could be a significant risk contributor.

In the absence of early containment failure there would be a

, decrease in the containment pressure after the time of head failure due to.

. condensation of steam on structures. Subsequently tne pressure would rise

]

again due to the continued evaporation of water from the cavity and the

.1 attack of the concrete by the core and structural debris. Eventually either I

overpressure or melt-through would be expected. Sinc'e substantial time -

l

) would be required for either of these failure modes, the consequences would d be expected to be reduced by fission product removal during that time.

In the absence of active containment safety features high steam partial pressures would be expected in the containment throughout the TM.8' sequence. These high steam partial pressures would inert the containment i atmospnere, thus precluding the possibility of hydrogen combustion.

a 1 4.2.3 Seauence S 2 0 (Small Pioe

'_; Break. Failure of ECC System) .

L ' i. ,

Because' of the availability of containment cooling and the con-tainment spray systems in this sequence, the expected release of fission products to the environment would be quite small. As a result, the contri-bution to the predicted public health risk would also be expected to be small. This type of event is believed to be comparatively likely, however, relative to other core melt sequences and is of interest for this reason. -

This is the only sequence analyzed in which the effectiveness of the contain-

) ment safety features is examined. The behavior of hydrogen combustion in this case is of particular interest because steam inerting will not be

present as in other sequences. Core meltdown occurs with elevated reactor coolant system pressure as in the TM.8' case but at somewhat lower pressures due to continuous leakage from the reactor coolant system. The rate of l coolant loss and the timing of core uncovery and melting will be dependent l on the size and location of the break in the primary system.

l i

'i

- .., .--.~.. -... -... ..--. ._ .:-.._

L 4,

~'" ~ : Gs

a 8

4-16 1

I The flow path of fission products in the primary system is illus-l trated in Figure 4.6. Other possible flow paths to the break through the intact loops were considered to be saaled by water in the low points of the primary system. If the flow path through the two intact loops is also

-. available, the residence time and retention of fission products in the j primary system would be greater than for the case analyzed.

]I Two specific containment failure modes were considered for the .

S2 0 sequence, early failure due to a hydrogen burn (Y) and melt-through of the ennerete basemat (c). For a hydrogen burn to present a significant threat to containment integrity, a large coherent burn would be required.

This would imply accumulation and distribution throughout the containment volume of a l'arge quant'ity of hydrogen, followed by ignition. If the burn-

' ,j ing were spread out in time or if a number of small burns were to take place, there would be no threat"to the containment. Melt-through of the containment basemat would be the most likely failure mode for this sequence. The reactor

( cavity in the Surry plant is understood to be separated from the containment

', sump, thus there is no direct access for the water on the containment floor to the reactor cavity. However,someofthecontainmentsprapwaterwould q be able to find its way into the reactor cavity. If sufficient water is able to continuously reach the debris, a coolable configuration may be .

achieved. In such a case the accident sequence may be terminated without a

J breach of containment.

.$ 4.2.4 Seouence V (Interfacino Systems 4

Loss of Coolant Accident) e The interfacing systems loss of coolant accident is initiated by f the failure of check valves separating the low pressure emergency core cool-

,}l ing system from the high pressure primary piping. Such a failure could not m only lead to the loss of reactor coolant, but also to the failure of the '

[ emergency core cooling system. The failure would be external to the reactor ,

.( containment and the released radioactivity would bypass the containment  !

safety features.

,{ This sequence was the largest individual contributor to risk for '

the PWR design identified in WASH 1400. Having recognized tne potential

,,I l d

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< l FIGURE 4.6. FLOW PATH FOR FISSION PRODUCTS IN SEQUENCES S2 D AND V y' .

- - - . . ~ . . . - . - .

%,.n-,._ _ . . . . . . -. .-

I t

1

{ 4-18 i

t

. i system weakness, it has been possible to reduce the likelihood of the t

j sequence substantially. The interfacing LOCA is of interest even at reduced probability, however, because the pathway for releast bypasses the protection normally provided by the containment building. The amount of retention in I

2, the reactor coolant system n thus particularly important in this sequence.

The flow path for release from the reactor coolant ' system is illus-f trated in Figures 4.6 and 4.7. ,

e

r. .

4 I

L s

A. '

l 3 h

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_...-.= =:~...,.-..-- , 9

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8

. w-1 FIGURE 4.7. RELEASE PATHWAY TO SAFEGUARDS BUILDING IN SEQUENCE V 4

,------------m__ _ ;,,_,j, v-- m w 4 w u- n- -<..+ . m- ,_7 . . .a v . . .

A

5. ANALYTICAL METHODS

The analytical st2thods and computer codes used to perfortn the

'j cdiculations reported in this volu.ne were identical to those described in Chapter 5 of Volume IV.

}

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=

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6. BASES FOR TRANSp0RT CALCULATIONS _

6.1 Plant Geometry and Thermal Hydraulic Conditions i

i MARCH 2 code calculations were perfomed for several variations

of each of the four accident sequences considered. The results of the .

MARCH analyses are used as input for three aspects of the fission product t I

release and transport calculations: '

(1) Thepredictedtime-dependnttemperatures

~ of the fuel are used by CORSOR to calculate fission product release. '

s

.d (2) The primary system pressure and flow of steam  ;

and hydrogan from the core are input to MERGE to {

q calculate primary system thermal-hydraulic input to TRAP-El.T.

(3) The thermal-hydraulic conditions in the contain-

.T ment building as well as leak rates out ara input to the containment transport codes.

.l 1

, J i;

4 i

A sumary of MARCH modeling options utilized in the analyses is presented  !

N in Table 6.1*. The design parameters for the Surry plant were sumarized I' J in Table 4.1.

l -i One of the many areas of modeling uncertainty for thermal-lj hydraulic analysis for this study has N en the behavior of the flow in the reactor coolant system in the pathway of release to the containment.

d In particular the conditions in the upper plenum and upper dome region .

are quite uncertain and could have a tignificant effect on the transport l' '

of radionuclides. The first problem in describing flow behavior in this region is in obtaining ut adequate enaracterization of the structures.

1 -

These structures are not described in detail in publicly available reports because of proprietary design features. This problem was alleviated by input from Westinghouse Electric Corpo.'ation, the reactor manufacturer for the Surry plant. In addition, even if the geometries of these strt:c-tures had been well known, there is significa:1t uncertainty as to the

' ~

• All tables in th's section of the report have been placed at the end of tha section.

t

. wp.w.e .;. 3 .-n 4

. .www a. c. . -c-7 7-S muN M N m *'- 7*~ ~ - ; ,_ . .,y,;2. ; p .

y7, _ . - - , . _ _ -

- _ - _ _ - - --- ~~~ - - - - - - -

t

_ ._- , _ _ _ _ - ..-- . - ., ._ _ - - . _ . _ _ _ _ _ _ _ . _ _ /t

a- -

d -

6-2

]1 nature of the flow patterns in the upper plenum and dome during accident j conditions, and as a result how much of the flow reaches the available

.] surtaces. The current version of TRAP only considers one structure with E

a given surface to mass ratio within a volume. While this limitation L can be circumvented by subdividing the volume of interest into a number of smaller subvolumes, each with a particular structure, such an approach

[] would require tr'iformation on the possible series / parallel flow splits .

among the subsolumes, i.e., how much of the flow is seen by each of the -

,3 structures. Such fine detail was not warranted by the existing level of understanding of flow patterns in the upper plenura. As a practical matter the gases leaving the core were assumed to flow in series through the

upper core support structures, past the control rod guide tubes and support columns, past the top support structure, along the upper core

\

ba:rel. and out the hot legs. Clearly alternate flow paths are possi-ble, e.g., after passing through the upper core plate some of the flow "E could go directly tio the hot legs without passing past the other

_e strmtures. The sensitivity of results to this approximation will be investigated later in the program.

In the following sections of the report, the results obtained with the MARCH and IERGE codes are described for each of the accident sequences. In Section 6.1.6, some of the uncertainties in the analyser and sensitivities to assuretions are discussed.

6.1.1 Secuence A8 (Hot Leo) ed A large pipe-break accident with failure of the active emergency .

core coolant injection system, a5 would result from total loss of AC r ;g power, would be er.pected to result in comparatively rapid core meltdown.

/ This is because core uncovery would occur essentially at the start of L: the accident with the decay heat level relatively high. The loss of

?1 electric rower will also preclude the operation of containment safety i y featares. Teble 6.2 indicates the times of kcy events as predicted by Ii the MARCH code for the input and modeling assumptions utilized.

[- Table 6.3 provides details of the core Ltd primary system conditions for

- {J .

this sequence. Core uncovery, hentup, and melting would occur at low a

g -

, .j l ,

i b._ - - - - - - - -...m ~cm - - - - - - - - - - ~" ~ ~ - ' ~ ~ " ' -

]

i 6-3 primary system pressure corresponding to the pressure of the containment.

The temperature of selected fuel regions is illustrated as a function of time in Figures 6.la and 6.lb. In this as well as rubsequent plots of core node temperatures the designation R00 (X,Y) denotes the core node

, in axial position X and radial region Y. For these analyses the core was divided into 24 axial nodes and 10 radial regions.

Prior to the accident, the pipirg and structures in the reactor.

l .

coolant system would be in the temperature range of 290-315 C. Because headp of the fuel and the release of fission products would occur at l about 172 kPa (25 psia), these surfaces would be expected to be considera j ably superheated. In addition, because of the high boiloff rate (high i

gj decay heat level) ad low density of gases fri the primary system (low

primary system pressure), the velocity of gases passing through the

.; reactor coolant system would be high in comparison to other accident

j iequences. At t.iie time of core uncovery, the velocity of steam in the l

j upper plenum is 1stimated to be approximately 1/2 meter /sec (2 feet /sec).

The to,tal residence time in the system fmm leaving the ccre to exiting 1 '

Ma break in the hot leg would be less than 1 minute. As the water level

.J in the core. drops, the production of steam decreases accordingly. Just i

prior to slumping into the lower plenum, most of the steam that is being generated. is predicted to be reduud to hydrogen.

1 Temperatures of the structure in the reactor coolant system i

, are illustrated in Figure 6.2. (In this figure time is measured from si %e start of core uncovery.) The gas temperatures leaving the core rise a

y rapidly as t4a core melts and begins to slump. Upon collapse of the core into the bottom head of the reactor vessel large quantities of steam f

H are generated, resulting in a sharp decrease in gas temperatures as well 1

! as cooldown of the structures in the upper plenum. The maximum tem- i

'i perature of gases leaving the hot leg is estimated to be in the range of j h 650 C (1200 F), but these persist only a short time. A schematic of the I gas flow path for the A8 sequence is illustrated in Figure 6.3. f k Several possible times and modSs of containment failure have

! been invasi.lgated for this sequence: failure to isolate (8), early 9

failure due to hydrogen burning (y), and basemat melt-through (c).

Table 6.4 presents the details of the containment response for the

/i i

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dance.

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< 20.0 15.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0

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1 TIME - (MINUTE) l i

FIGURE 6.2 UPPER PLENUM GAS AND STRUCTURE TEMPERATURES FOR SURRY AB-c SEQUENCE -

q U e

']

I r .

6

t 6-7 f

1

, CONTAINMENT 9

i r

A i

i

,$t

.f .

l UPPER PLENUM

'I (3 STRUCTURES) i1 1 4 1

- ~ . - .-

l ~1 l, CORE 1

1,

} FIGURE 6.3 SCHEMATIC'0F MERGE CONTROL VOLUMES FOR SURRY AB SEQUENCE

e

i' * .

4 j!

>f i

, . 4 ? ' larh'M 'E F N W r. Aa . . e s7

' er * , m Q* 1m I .-- *4- ) * * -7'IMv7G g 5

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.. _ _ . . . - ___- . ._ . ; ^!T _*~,, , T k ,

m - s, -- --,--e v

6-8 various cases considered. The containment isolation failure is assumed i to exist at the start of the accident. For this analysis it was assumed '

that the isolation failure wss in one of the penetrations from the containment to the safeguards or auxiliary buildings. Thus, both these volumes were considered in the analyses for the AB-s sequence. The containment pressure and temperature histories for this sequence are illustrated in Figures 6.4 and 6.5. In these figures Cu;upartments 1 and l 2 represent the containment and the safeguards building, respectively. ,

-l As a result of leakage through the isolation failure, the containment pressure decreases fairly rapidly after prir;:ary system blowdown. The large steam generation associated with the collapse of the core into the bottom head Teads to a temporary increase in *.he pressure. The pressure

in the safeguards building is seen to stay near atmuspheric except for a sharp increase at about 200 minutes when a hydrogen burn is predicted to take place. At other times the leakage from the safeguards building is able to keep the pressure low. It should t,e noted that a pressure rise

, of the type calculat'ed could lead to the failure of the safeguards build-Ing. The timing of the bura is controlled by'the combination of et increasing hydrogen concentration due to leakage from the containment v ! and decreasing s1!aam concentration due to condensation. The steam tends I

to inert the containment atmosphere early in the requence.' The early overpressure containment failure (y) considered was associated with a

hydrogen burn taking place during the concrete attack phase of the acci-( dent. The hydrogen was allowed to burn when flanrnable conditions were reached. The timing of the burn was determined by the increasing hydrogen h+' concentraticn from concrete attack and decreasing steam concentration due to condensation. High steam concentrations tends to inert the i.- containment atmosphere earlier in the sequence. The containment pres-

% sures and temperatures for the AB-y sequence are illustrated in Figures 6.6 and 6.7. Different containment pressura responses would be y[L -

predicted if the assumptions regarding hydrogen generation and burning were varied. The likelihood of containment failure due to a hydrogen burn or other event would of course depend on the failure pressure utilized j[; 'as well as the magnitude of the pressure. The quantification of the ,

[ probability of containment failure is not a part of this effort. If the jl '. 'i 4

1 L

$1,.g.7 n;:m=3r=33; ' ;- }_ _ T LDM M""""*"*b

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• 0 0 0 0

-, 0 5 0 5 0 5 3 2 2 1 1 MgmMNV EZNxk4N%oO s

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'i

] containment is able to withstand earlier challenges, loncj term overpres- ,

.j sure failure or basemat melt-through may be the eventual outcome for this secuence. The current MARCH calculations using as input a concrete composition representative of that actually used at Surry indicate that melt -through of the basemat would take place,before long-term overpres-

.; surization. This is the case that was analyzed as repre.3entative of

j delayed failure modes. The containment pressure and temperature histories

!j for this sequence are shown in Figures 6.8 and 6.9. Containment melt-i through is predicted to take place at about 1450 minutes into the accident.

Since upon melt-through the containmeat depressurizes througit the ground,

'( only a small change in the containment pressure is seen for the conditions Ij encountered here. It should be noted that there ':. substantial i . (' uncertainty regarding the progression of concrete attack and the timing i$ of the occurrence of melt-thrc:.9h could vary considerably. It ir also i

,. possible that overpressure failure could precede melt-through. In the J4 l Surry design the reactor cavity and the containment sump are not connected; e g thus sump water evaporation is not a source of containment pressurization.  ;-

1} After reactor vessel penetration the principal driving force for pressuri- ,

$ zation would be the release of gases from the decomposition of the

concrate. The time required to reach the assumed failure level of 0.69 Wa (100 psia) would be longer than the melt-through time predicted for y the particular case considered here. If, however, lateral attack of the concrete by the core debris leads ta the ingression of sump water into i'

i the reactor cavity, the rate of containment pressurization could be 1-increased. .

l.

Table 6.5 sunnarizes the containmer.t leakages for the various .

, cases corsidered that were derived from the MARCH results and used in tua evaluation of the fission product release from containment.  !

't 6.1.2 Seouence TM.8' l; InthetransientsequenceTM.5',abilitytoremoveheatfrom the reactor coolant system is lost and containment safety features are l, not available due to loss of all electric power. Decay heating following

!! reactor shutdown boils off the water in the secondary side of the steam m

i a h

(

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1 -

'r -m m _ , _ . _- .7- aw - ' ' 'ET .--* ' d.- d : ' ** " " - ' ' ' ' " ' ~ " " ' ' ' ~ ~ ' " ~ " " ' -

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6-16

-I generators. After steam generator dryout, the primary system pressure rises to the relief valve setpoint and reactor coolant is discharged l

l through the relief line to the discharge tank and ultimately to the l

! containment building. Table 6.2 indicates the times of key events as predicted by the MARCH code. Core and primary system conditions are  :

, given in Table 6.3. The temperature t'ransient of selected fuel regions is illustrated in Figures 6.10a and 6.10b. Core uncovery, heacup, and ,

melting occur with the primary system pressure at approximately 17.24 MPa (2500 psia). Because of the high density of steam at this pressure, the flow velocity in the primary system would be quite small, generally less than 1/2 cm/sec (1 foot / min) until the start of core slumping.

t' The Reynolds Number in the upper plenum is predicted to be in the laminar regime until core slumping begins, as for the A8 case. The 4

l Rayleigh Number, however, is substantially larger in the range of 1011 -

14

, 10 . Thus, significant mixing could occur in the upper alenum driven t I

by temperature gradients and the buoyancy of hyd;cgen. The temperatures of the gas and structures in the volumes of the primary system are q illustrated in Figures 6.11a and 6.11b. In these figures the time is

.j measured from the start of core uncovery. A sciiematic of the gas flow jl path for TM.3' is illustrated in Figure 6.12.

! Two specific containment failure mode possibilities were consi-dered for the TM.B' sequence, an early and a late failure. The early failure was assumed to be the result of the rapid steam generation from the interaction of the core debris with accumulator water in the reactor

cavity. The failure of the vessel bottom head releases the high pressure steam from the primary system to the containment as well as discharging

,h the core debris into the reactor cavity; the drop in the primary system

~

!- pressure allows the accumulators to discharge onto the top of the core debris. MARCH calculates the rate of steam and hydrogen production resulting from the debris-water interaction in the reactor cavity.

{ There are several user selected options in MARCH that can be used to C describe these interactions. These range from a simple quenching model 4 to several debris bed models; comoinations c1 models are also possible.

All of the models require user inputs for which there is little basis.

Perhans the most.imnortant input parameter is the assumed debris particle -

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25.0 i

30.0 i

35.0 i

40.0 i

45.0 i

50.0 i

55.0 60.0

TIME - (MINUTE)

y FIGURE 6.11a. PRIMRY SYSTEM GAS AND STRUCTURE TEMPERATURES p .. FOR SURRY TMLB' SEQUENCE fk l '. '+

. a.

e4== .

tv. .n :;~ 7. .y niy .: S b.s;v ::... - _ ,, v , . ..u; _ $ g v. y ;q ,,.,,. hay , 3 ,jr , 9:,y .y 9:.3,;,,=.;. 9 4;;, ;.. (;, y,_-

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x.

" ..p (J "

il 600.0 8 3

5 3 3 5 20.0 25.0 30.0 35.0 40.0 45.0 3

50.0 55.0 60.0

)

TIME - (MINUTE) .

FIGURE 6.11b. PRIMARY SYSTEM CAS AND H RUCTURE TEMPERATURES '

FOR SURRY THLB' SEQUENCE -

1 .

i i

.4 e e i

6-21 A

-1  ; CONTAINMENT f

~

e d k

. -> j i i t m. .. . __;

i .

l PIPING & l:

q~ . PRES 5U.tIZER F

' i

. (2 STRUCTURES) I e ,.

-a d .

I' w .. % .

i

! UPPER PLENUM

r. i

, I (3 STRUCTURES) l l

A 9

'i

! j.

t i CORE [

l

i FIGURE 6.12 SCHEMATIC OF MERGE CONTROL VOLUMES FOR SURRY TMLB' SEQUENCE  !

t i

I I

II 4,s 22 we. sn-y--... ~~-w  ; a w y v . . . a .., . . ,. . .

.. rr : t r - -, s: - -------*:-~.. --

. m r u ---- . ,,

. . _ - - _ . . _.n.. ,,,, _. _ ___,,__ _ ,_, _}

g

E-22 )

i l

)I .

size; additional inputs that may be required relate to debris bed

,o l porosity, bed area, criteria for transition from particulate to debris j bed heat transfer, etc. The rate and magnitude of the calculated

.i containment pressure rise can obviously be sensitive to the input and i

i modcling assumptions utilized. The use of the simple quench models in conjJnction with small particle sizes tends to maximize the amount as

,, well as rate of s15eam production from debris-water interaction; since in .

such casor the debris are quenched rapidly, only small amounts of hydro-j gen are predicted to be gejerated under these assumptions. The debris N

,. bed models can also predict large emounts and rates cf steam generation, Z depending on the input parameters, but generally lead to slower rates of

,9s containment loading. In some cases associated with marginally coolable

.h debris beds, large amounts of hydrogen genration can be predicted. The y containment pressures and temperatures for a representative TM B-c d sequence are givsn in Figures (i.13 and 6.14. The effect of the debris-J water interaction is seen in the sharp pressure increase at about iti0 minutes in Figure 6.13. In the part cular case illustrated the pressure

.Q 7 rise would probably not challenge containment integrity; higher pressures

could be predicted under alternate assumptions. If the containment 7 maintains its integrity through the above early pressure transiant, the

z. containment pressure will decline somewhat due to condensatic.: of steam on internal structures, but will later increase again due to the attack

.of the concrete basehiat by the hot core and structural debris. Since i.

1. the gas and vapor input rates from concrete decomposition are low, except j when %e debris is very hot, a long time would be required for the pres-l" sures to build up to levels at which the likelihood of failure is signi- ,

. 4 ficant. Since it may t'ake a long time to reach pressure levels 13ading to failure, it is pussible that basemat melt-through may precede and I. ] preclude such a long-term overpressure failure. Melt-through of the concrete basemat with venting of the gases through the ground.is reflected

]3 in the sharp pressure decrease at about 740 minutes in Figure 6.13. It should-be noted that the progression of concretc attack by the core debris

.s 9 and hence the possible timing of melt-through are highly uncertain. It I

may be noted that the high partial pressures of steam in the containment Jj atmosphere throughout most of this sequence are predicted to preclude

$ hydrogen burning.

U .

i

L. , .

~

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_- 7 pv - --~w y~., - .-

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ggy, g

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6-24

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^

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>=e gn 6

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I =

w

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- CE W i 1

v 8 -* x w

=

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O  ;

s- E .

. z

< C'3 2

>=

Z Q

U C

w

, - f's s 8" J

• A w

9 h

-8

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~

- . 9

. .. . C 6 6 s e rs o o C 9 9 9

^ ~^ ^

g B B S. 8 8 l4 e:i - -

HHn1VH2dM31 IN3M1HVdWOD 4

(

l '.

7 j

fj,_ _

7 E- - - , .y.f, - s-

-- y,m.. . .

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'N

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~

Figures 6.15 and 6.16 give the containmect pressure and temperature histories.for TM.B-6 in which the containment is assumed to

, fail early due to the pressure rise from the debris water interacting in

, l- the reactor cavity. It may be noted that some of the input parameters

for the TM.B-4 sequenco;were changed from those in TM.B-c leading to a .

l higher. containment pres:ure peak. In considering the occurrence of an  !

, early containment failure, ro representation is made as to the likelihood

, ofsuchfailure;th'eguaktificationoftheprobabilityofcontainment

failure due to such,interactionsJis beyond the scope of this study. It

'{

1 - '

l 1s suggested, however, that the magnitude of the pressures resulting i '

i from debris-water interactions may be sufficiently high that the possi-bility of failu v should be cens!dared. r Table 6.Esunturizes "the containment leak rate information ,

l derived from the i$6tCH results and used in the evaluation of the fission product release from the contairc9nt. .

4+ q 6.1.3 Secuence_V  % - . . _

3 ;, The V cequence or' interfacing systems LOCA is initiated by the L ll failure of the check valves separasing the low pressure emergency core '

cooling system f6m the primary coolant system. The release of the high

, pressure primary coolant inven Mry to the low pressure piping would not ,

only lead to the failure off the 'emergenc core cooling system but aixo  ;

4 provide a path for the release of radioactivity that bypasses the primary containment. It is also possible! that- the primary system blowdown would p.; , result in the failure of the safeguards or the auxiliary building. '

The interfacing systems LOCA sequence is of particular interest g because the containment building is bypassed for much of the sequence i

j l] and the primary system could represent the principal location for the J retention of fission products released from the core. The possibility Q of retention in the safeguards (or auxiliary) building also exists if it

} does not also fail, but would be quite design dependent. The thermal-Y

' hydraulic behavior of the system during the period of fuel melting is

)f similar to that of the Sequence AB. Following a period of 1/2 hour in

]p which blowdown and loss of reactor coolant inventory lead to the point 4

%. ,_ .= --- r mm

g. -.

mm._s;

,hg,_ _ ,. _ __.

_. - ~ _. - - - - - . - - - - -

a %-

6-26 I e i 8 i, =

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i.

c

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=

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=

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m, o 8 l

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l HHn1VH3dM31 ~1N3W1HVdWOD

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• ' ~ g

-. , _ , - - - - _ , - -. . - . , - . _ , , . . . . _ ~ . _ _ - - . _ _ . . . . - . _ _ . _ _ . -_ - ~______ - - -'

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4 t

6-28

- of core uncovery, melting of fuel would occur over ac interval of ancther 1/2 hour. The temperature histories of selected core nodes are given in Figures 6.17a and 6.17b. The timing of key events is presented in Table 6.2. The core and primary system characteristics at key times during the sequence are given in Table 6.3. In this case, the primary system pressure would be slightly more elevated (0.68 W a) and the velo-city in the upper plenum would be reduced to approximately 10 cm/s. The ,

velocity of the steam and hydrogen flowing back through the ECC injec -

tion line to the auxiliary building would be greater than 1 m/s. The residence time in the reactor coolant system from core exit to the

+

atmosphere of the auxiliary building would be on the order of 1 minute

, with the majority of the time spent in.the upper plenum. The predicted

< ~

temperatures of gases and structures in the flow path to the auxiliary building are illustrated in Figures 6.18a-d. In these figures Volume 2  !

represents the reactor vessel upper plenum, Volume 3 the hot leg piping, i'

Volume 4 the steam generator, and Volume 5 the emergency core cooling

'i.l system piping. Figures 6.19a-d show the results for an alternate break-down in which the emergency core i:ooling sy. stem piping was represented i as four connected volumes rather than a single volume. In both sets of

.; these figures the time is measured from the start of core uncovery. The  !

results are similar to those obtained for the AB sequence. A schematic of the gas flow path for V is illustrated in Figure 6.20. The auxiliary I1 building and containment characteristics at various times during the sequence are given in Table 6.4. Table 6.5 sunnarizes the c6.ntainant leakage flows derived fran the MARCH analyses and used to evaluate fission

[ p'roduct release to the entironment.

t 6.1.4 Secuence S 2,0 _

+

e The small pipe break accident sequence w'ith failure of ECC ~

3 injection involves conditions intermediate to the high pressure meltdown

[  :

sequence TM.8' and the low pressure sequences AB and V. In the S20 sequence the containment sprays and heat removal systems are operating; 3

'thus the containment pressure would be maintained at low levels, except i

, , perhaps f or brief periods when rapid transient loadings may be encountered.

y s tr I

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M OGAS INLET U AGAS OUTLET J,i +STRUCTURP. I c- g XSTRUCTURE 2 A-oSTRUCTURE 3

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.; 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 i TIME - (MINUTE) t

,.I

'. r j FIGURE 6.18a. PRIMARY SYSTEM GAS AND STRUCTURE TEMPERATURES FOR SURRY_V SEQUENCE cm .

+.

.E

n -

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i4 _

p.

r.. - . . e . ... , < ~, -r-:. c o . . t ,h h w r c um . - e -

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a GAS OUTLET l

+ STRUCTURE 1 -

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400.0 , , , , , , , ,

10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 TIME - (MINUTE)

FIGURE 6.18b. PRIMARY SYSTEM GAS AND STRUCTURE TEMPERATURES FOR SURRY Y SEQUENCE

~

l E s E1 j _... ...._..-...- .

7.= .

m. ,.. . . . .. .

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500.0-1 1

1 400.0 l 10.0 15.0.

20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 i

g TIME - (MINUTE)

J ll FIGURE 6.18c. PRIMARY SYSlEM GAS AND STR'JCTURE TEMPERATURES FOR SURRY V SEQUENCE

'h

• ?

1, <

c 1 R v_ . t - r ..- . . , ; n r_s. u_ a , z _ , ,y _z_ _. . : . ., _

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, 1' 4 f 580.0-N

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< 550.0-M N

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.., 530.0 . . , . . . , ,

! 10.0 '

15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 TIME - MINUTE) i

9 FIGURE 6.18d. PRIMARY SYSTEM GAS AND STRUCTURE TEMPERATURES FOR SURRY. V SEQUENCE ii

_ 6 L _.

mL* ,

2 0

.. 5

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i i i i i 10.0 15.0 20.0 25.0 30.0 35 0 40.0 45.0 50.0 550 TIME - (MINUTE)

FIGURE 6.19c EMERGENCY CORE COOLING SYSTEM PIPiliG TEMPERATURES FOR SURRY V SEQUENCE ,

!y yIIf t

ly. ,

~- I t

6-38 C

I c

o I.

30 1

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m

~

uca

~

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a -

D= D >

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-{ 03 D2 m

<< =

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w

( . C3

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=

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6-39 SAFEGUARDS BUILDING t . -

t A'

S

,.. . ECCS PIPING i

a

t l b

?

i. _4

(!

1 STEAM GENERATOR h

3

., PRIMARY PIPING

-t i

d' 1

, i ii UPPER PLENUM l '

(3 STRUCTURES)

'1.

8 D 1 CORE

'I 1

i FIGURE 6 20 SCHEMATIC OF MERGE CONTROL VOLUMES 4 FOR SURRY Y SFQUENCF i

d

. . , .c._ . = _ .w,,. . , em mw w- n = - n--,-~~n-..-.=-.. .

..t..'.=.b.....t'.' ..

'*" '_f.,-a._._.. _ r_. _ , _ _ _ _ - . . . - - _ . . . _ . . _ _ _ - - . ' ' ' .-.-

' " ' ~ ~..

~ . .~ n' s

3

!' 6-40 i Two containment failure modes were considered for this sequence: an early overpressure failure resulting from hydrogen combustion and basernt jl ,

melt-through failure with no direct atmospheric failure of the contain-l

.7 ment. The timing of significant events is given in Table 6.2. Core and primary system parameters are sunnarized in Table 6.3. Selected core node temperatures are illustrated in Figures 6.21a and 6.21b. The

] predicted temperatures of gases and structures in the primary system are ,

[' . illustrated in Figures 6.22a, 6.22b, anc 6.22c. In these figures the time is measured from the start of core uncovery. A schematic of the j gas flow path for 52 0 is illustrated in Figure 6.23.

,.j Table 6.4 semiarizes the containment response at key timec during the accident sequence. Since the rate of primary coolant loss is restricted by the break size and the containment sprays are operational,

, pressure in the containment would remain relatively low during most of this accident sequence. This is illustrated by the pressure history for the melt-through casa (520-c) in Figure 6.24; the corresponding tempera-ture history is given in Figure 6.25. Substantial containment pressure

i increases would require the accumulation and subsequent coherent burning of lar'ge amounts of hydrogen. For the case of failure due to a hydrogen

!l burn, the latter was assumed to take place following vessel failure when Ej the hot core debris entered the reactor cavity. The ejection of the hot j ,

core debris into the containment would provide the ignition source for the burn. The ccaiiainment pressure and temperature histories for such a hydrogen burn are given in Figures 6.26 and 6.27. This burn produced c

( peak pressure of about 0.62 W a (90 psia); the calculated peak pressure

can be sensitive to the timing of the assumed burn. In assuming contain-ment failurt due to such a burn, no representation is made as to the likelihood of failure. If the containment liaintains its integrity through

.;, challenges such as hydrogen burning, it is likely th7t the basemat will

] eventually be penetrated due to the attack of the concrete by the core and structural debris. The time required for melt-through would provide

,k ample opportunity for fission product removal by the sprays. If suffi-

,[ cient spray water can continuously reach the reactor cavity it may be

,[ possible for a coolable debris bed to form. In such a case the S 20 ,

sequence would be terminated without containment breach.

.s I

• f

& .. .-mc y- p _. m ww ~=- -

7

-.3 w b

j....._.

, ~ . . .. . _ . .____.

~

.9 i

6-41

- o o

! CI

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~&

O

_h u a

?

C .

I

_g m E

n i Z

ow

. , a N,, -

f 1 N ,' z m O ,

e2

=

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i ,

w Z

r=

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8

=

g N ,-

N , e =

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, b ;*.'i'. i'- E . . _ _ _ -r?,,

__ m e-w n.,-,.._.,__ _ . . _

. ., -e, .f hV .w. . , ,- l

_ - . _ . _ _ _ _ _ . i

I 1

I l

l 6-42 e

.1 e,

C2 e

e 1 -

g i

i e

-8N

\ e -

i

-8 eN 4

m  ;

e -

2

, - =

m i

y ~

^

D

,Z

.............y eM @

w I

e9 ,

I

,, sa y C-. ,

- z 2 Q '

oy f

C\2 i

-8 v - t CG #

> o i

g l

. Z .

-8_yg =

D a.,;. . - w g

_____- - - p g

_ CQ

• e e

.m.-

U w

a

'i e W m

-8 .

e W-

-y

~ .-. ___ ---

s..

i

~

e 2f252r ,' - - o w 888 ,' .

i i

s=

.; l -

l. i -

e '

i:

i -

-8

,, s s J

o e

s s a e s e L e e o e e e e o e e .

$, $ N N N N $ $

, , 8 M M M - -

ll f '3Hf11VH3dM31 'IVGON EH00 1

.y .

Y "9

.,,_,,.,,----a.

--r.~,

-.,-.-----.n- - - - - -

_w _ .-. . - _ , . _ , . . ,3 w -

u...-..

_a,,,

J. .. 4.A+.. ---=.A-,c * *

• 6 '

W;r ,

m. . .

$e

SURRY S2D EPSILON

}  :*000 0 ii OGAS INbW

+ U U  !

4 g ro S W 2

~ -

i C0 2500.0-i{;

• 2

.j

, p 2000.0-o

'!  ?

N 0 '

A 1500.0-CD * ..

. g a

1 p ,

1 P. 1000.0-4 <

}' @ 1 i

N ^

o-b y .t_ -

v c'~'

o

e r  ; _ -4

\

g 600.0-t) Y  :

'3

,s a- - e __.__.e _ _-- 0 o - 0 -co -

i I

0.0 - , , , , , , .

30.0 40.0 60.0 60.0 70.0 80.0 90.0 100.0 11 0.0 TIME - (MINUTE) ,

FIGURE 6.22a. PRIMARY SYSTEM GAS AND STRUCTURE TEMPERATURES 2 FOR SUPRY S D-c SEQUENCE ,

P

' {.

q r ir .

P

~- ~ - -

s

3, . . ;. -. . .

i o

' t ij 1

J1 J

SURRY S2D EPSILON

] 2000.0 J: O G AS IN!.ET

,;1 A GAS OUTLET

'y + STRUCTURE 1 g 1800.0-i ci '

i 1600.0-y' .

w

,_a -

O l p. 1400.0-1~ N e o

!' 1200.0-en N T 1

% i:

D 1000.0-

' E-4 \

Z )

800.0-32 -

N E-* '

~ -6 J " A A

O A A A 600.0- O 0 -

400.0 , , , , , , ,

30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 11 0.0 '

,]

TIME - (MINUTE) ,

FIGURE 6.22b.

1 PRIMARY SYSTEM GAS AND STRUCTURE TEMPERATURES FOR2 SURRY S D-c SEQUENCE 1

. ~ . . . - .. . . .

g, e - ,

..  : u. . _ . . . ,

m=. . . ...

t. I ti .

• IA n .

y .i .

2 1 SURRY S2D EPSILON -

• . 1600.0 ,

t OGAS INLET

>JI 'AGAS OUTLET

! ,'@E + STRUCTURE 1 rz.,

'! q 9 1400.0-

.$ N

0 X

.~2 di D *

.-l 1200.0-O .

> 1 l-i-'

Ex O

1000.0- '

i rr [

o, i Ci i p Z

Iy D

1. 1 Em p, 4 800.0-l M

,\ N

,~I

,- 4 [ O O O O O ^

y O O

[ N S00.0-b  ; 9::  : .;- J

~

a 1l - c, t i

?g ii etf!

j, 400.0- , , , , , , , ,

30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 110 0 TIME - (MINUTE)

.m .

l; { FIGURE 6.22c. PRIMARY SYSTEM GAS A.90 STRUCTURE TEMPERATURES FOR SURRY 2 S D-c SEQUENCE f

~"._7.__. .._,......_.-m - -

i l:

1 6 46

. i .

t l

J8 .

y h.

I i CONTAINMEhT 1

~

.4 j A i .

3 STEAM i

GENERATOR d

l l , PIPING A

A A

0'PER PLENUM

.1 (3 STRUCTURES) i

}k C

r.  ;

94

' 4 CORE I

l in

? l:: I

!' ; u FIGURE 6.23 SCHEMATIC OF TRGE CON.'ROL VOLUMES FOR SURRY S D2 SEQUENCE t .:

r._

' ~~ ' ' ' ' * ' ' ' " ' " ~ ~

""""'74'

^ '

-~~3._-

'f% -

_; +. 'Vef'=p v -, '[ _"*'~~,W['"~~~ "'~~['

l,

y3e i' : ,

b, l.

,  ?*"

0 0

0

0 1

0 0

_ ,0 9

0 0 E

_ ,0 C 8 N

_ . E

_ U Q

~ 0 0

E S

c

_ ,0 -

7 0

_ 2

- S Y

R .-

R

' =- 0)

E U -

0E S -

,0 T -

. D 6 U

R O

F 2 N E S 0 I S N -

.u Y . 0M i0 (

O P

5 S

1. c~ R -

E R

R E

.w R

E U -

U S

i00. M 0 S -

S 4 I T

E R

P T

. . - N E

-  %. 0 M

- ..L.

0 N -

- i0 I

~ ,

3 A *

. T N

O C

0 N i0 0 4 2

2 -

5 E ~.

. R T U -

0 G I

0 F i0

- 1

- ~ .

0 .,

0 .

n

- , 0 0 0 0 0 0 0 0 0

- 0 5 0 5 0 5 O 5 0 5 5 4 4 3 3 2 a 1 1 nn M%DrEM%A E>>yMEZ4AMoU aqEOE-

- - H

-7

- , < j; I - !3 1'

,i ii ' :, l:  ! - ,E

- y1 < yy? , t;j . '. - i: ,

1

- ' t"; ' . .

4!

r -

1 1

1 4

J 6-48

.I O i - C i 5 1

e 1

• O g .a 8 -

s

. 8 m

I m

$ g

-8> e~

m

- >=

cn =

=

a m

.c o ~

b o e m.

N rA o U bi

_g s m 5 e;

'. 4 g C i

w

=

M ew w _

C  ?

+

CG -82

'- 5

, S itw

-' / e

=

_g m t

' o E ,

l -

l'

=

C O

. . , g 1

-$ m m.

a

! w C o'

!~

_4 5  :

C

, i l

}j i, c #

la i

g .

e ls, e-o C C C

,y

@ M" s 8

- S

, i1 t.

2Hn1VHZdW21 INEN1HVdWOD e

l 6.

.IJ h

,I g,

'gN TT ' --- 'N e

6-49 q I C C

e3

.I c

. .: 8_

.]

.4 l

m 1

e g m

.i.I g

.J =

.t'

< c y 2 -@ ~

2 -

em

- a r

< -g r g

a

O- oj g

'a __

'l Q - bi C\1

-$"2  !

Co I c

=

o

. 4,

~

a m M - ~2

Z. p a y o E D ,_ -

-8,. =

1 4

CO .!

e .,

1

-@ 5 m

o le

-8a e -

_ m g.

o 2 l

1 -

f

C c

i Q Q o C Q Q c c c c l

8 8 8 8 8 8 2 i

3HnSSSHd IN3NIHVdM03 7V101 1

h 1

~

i w.v v.+..;. wu ~a . 4- #

l C. W......~..n..,,,,,-

a. ...,..n - . v e. '. . .:.~w.. . .

- .' m *~ ~

• T'~~ ~^~. ' . . - - '

- ~ - . ~ . - - - . . - -~'7WWWl m.---

L.

i 6-50

, e C

' C S

I e

C

-0

=

" C

.g _

m.

C g M E C-e C y 4 e g _5- m

~

% n OM z

- g :-- S a . .

c. D z m e

Q J 3~5 W CV E E CQ l d C "

-m

> J -52 w

Qi ,

r 2

~4

% i

~

p .

-8,. m CG %4 C

- .E

5 C

5 U

-8

~ l CJ N

\

g

( C W u -

-8 8 c

, o

.- ~ , , , , ,C o o o o A A f l $ $ $

i HHn1VHZdW31 INSW1HVdM03

's I

,r, o

- . g - ,s _ . ._ . Yuggr v , - , - - ---

. _--_ . . -L**

_7

_ . _ . . . _ _ _ ' _ . . _ . _ _ _ _ _ . . _ _ _--. _. .-i_ _ _ _

a.a --

E 6-51 i

Table 6.5 sumarizes the containment leakages derived from the '

uj MARCH calculations that were used in the evaluation of fission product f releases from the containment.

s N *'

, 6.1.5 Other Cases Considered ,

i t' i

The primary containment in the above cases was modeled as a .

single compartment with its inherent assumption of being well mixed.

For the A8-8 and V sequences the releases from the prima:y containment were input into the safeguards building before their release to the i atmosphere. Thus the potential for fission product deposition in the i( safeguards building was included in the analyses for these two cases.

y.j In the other sequences the release from the primary containment was f directly to the environment.

d !i a In order to investigate the possible effects of containment ,

compartmentalization on the release of radioactivity to the environment,  ;

the A8-8 sequence was reformulated to model the primary containment ts'c '

system of four interconrected compartments. While the Surry containment has many internal structurss and partitions, the flow paths between

~

, adjacent subcompartments are typically very large and multiple Vlow paths t

d are possible. Thus there is no obvious unique approach to containment j compartmentalization. The containment model in MARCil is restricted to j handling only compartments connected in series. The faar volume representation utilized was a practical cogromise between the actual cor tainment design and the code modeling capabilities. With the four crapah containment model the melt relaase took place into a different i-;

compartment than the vaporization release, f.e., the melt release was

• l assumed to enter one cf the steam generator cubicles and the vaporization L.1 release was into the reactor cwity. The location of the containment l

Q 1eak was in still another cogartment. -

d In the V sequence the rupture of the coolant system 'takas place

$ in the low pressure emergency core cooling system piping outside the

'1 primary containment. There is a long length of pipe between the core 4 where the fission products leave the fuel and the safeguards building j where they leave the piping. In the base case analysis this piping was 1

i i

&m%. , we . . << 7 4 Fj. s r..-..w.-.

7***= E# ' ' *** M F

-w, 7

~~-- --~~;

7 '

-" " " ~ ^*^*'

• u -

'd

,1 . .

t s.52 represented as a single volume. In order to gain insight on the possible

.f sensitivity of the predictions this analysis was repeated with the piping being divided into four equal volumes, t2

, . (Thermal hydraulic discussion for decay heat-up of primary system in preparation) 6.1.6 General Discussion J

i The release and transport of fission products are strongly h influenced by the thermal-hydraulic behavior of the accident. The y ~ cog uter codes MARCH 2 and MERGE that have been used to predict the

, thermal-hydraulic conditions treat various aspects of accident behavior I-g with different degrees of confidence. In the following paragraphs the i

, principal areas of uncertainty in the analyses, simplifying assumptions I and approximations, and the implications to fission product transport i

} b will be discussed.

In the MARCH 2 analyses of fuel hestup, the reactor core was  :

sundivided into 24 axi.al anc 10 radial nesh regions. The variation that i would occur in the timing of heatup and fission product release across the core is well characterized. Up to the point of cladding melting and

. fuel / cladding liquefaction, the theoretical treatments of the thermal

[

behavior. of the fuel and oxidation of the cladding are supported by i experimental data. Reasonable agreement has been obtained in the past between different coguter codes in analyzing this behavior. The MARCH '

g code- makes the simplifying approximation that the fuel would melt at a single characteristic temperature which is input. The selected input

[ melting tagerature of 2550 K (4130 F) has been chosen to be between the temperature at which the fuel would dissolve into molten zirconium and j

i\ -

the melting point of uranium dioxide. In the actual system, melting p would occur over the above range of temperatures up to the melting point *

, ) D of uranium dioxide, with the actual melting or liquids temperature being a function of Iccal composition and state of oxidation. As a result of the single melting point approximation, the peak fuel temperatures

+

B 4

l T.

}

• SA

~

E26 ~4 r ~ T. " *:

- * - * ~~~ ' ~

- - - . - _ .... m _ _ -. _ _ - _ .

3- t

, 6-53 '

i l predicted by MARCH may be underestimated for some quantity of fuel. The time for which fuel stays at elevated temperature is also very dependent l j on modeling uncertainties. These uncertainties will have little effect

! on the predicted release of volatile fission products but could affect the vaporization of involatile materials, most likely by underprediction.

The MARCH analyses for the present study utilized meltdown model "A" with no movement of fuel out of the core until the bottom node '

in any radial region was molten. At that point the molten nodes in that region were allowed to slump to the core support plate. As the bottom nodes in successive radial regions were molten these regions were allowed to slump. When 75 percent of the core was molten, the entire core was assumed to slump into the lower head of the reactor vessel. The assump- }

tions regarding fuel slumping and ultimate collapse into the vessel [

bottom head also affect the driving forces for fission product transport, the timing of reactor vessel dryout, and subsequent analyses of head

[

heatup and failure. MARCH does not attempt to describe mechanistically

,j fuel melting and movement. There is no actual redefinition of core geometry. The slumping models attempt to treat mathematically what would i happen if the fuel were to move in accordance with the selected scenario.

i These representations are highly intuitive, but with some support derived l from core meltdown simulation experiments.

[ The MERGE code was developed specifically for the aaalysis of lj '

reactor coolant system temperatures in this study. There is very little past experience in performing this type of analysis. The flow patterns i in the systes could be quite complex, particularly in the upper plenum

,- region, and are treated approximately. If there is substantial internal

?

l recirculation both the fission product residence time as well as the  ;

l heating of structures could be hipher than calculated here. The flow within the reactor coolant system is treated as one-dimensional with well-mixed volumes. Natural convection within the upper plenum is consi-dered in predicting the heat transfer to structures. Although convection l patterns are not examined explicitly, the mixing which is expected to result would be consistent with the well-mixed approximation. The one-dimensional treatment of the upper plenum does not take into account the radial temperature profile of gases leaving the core and transporting a

e

,8 *ge

• e- =

4 [ l c- -

I ,

)

6-54 through the upper plenum. The calculated temperatures are averages across the flow cross section and would be expected to be higher near the center and cooler near the periphery.

As demnonstrated later in this report, the timing of containment failure has a .sajor impact on the predicted release. of fission products to the environment. The pressure level at which the containment would be expected to fail is input into MARCH. To the extent that this failure '

1 pressure is uncertain (typically it ic quite uncertain), it would tend I

to compound any uncertainties associated- with MARCH code calculations.

The thermal hydraulic conditions within the containment can be predicted with relative confidence if the driving furces are well defined. The

[ principal early challenges to containment integrity are due to rapid 7 steam generation from core debris interaction with water and from the burning of hydrogen. The analysis of steam generation from debris quenching is particularly uncertain and sensitive to the input and model-ing assumptions utilized. This phenorenology is inherently uncertain t

and one in which unique answers, except in a bounding sense, cannot be

< expected.

l f

The prediction of the pressures due to hydrogen burning is

. fairly straightforward if the initial conditions and the timing of the

, burn are known, however, this is generally not the case and key assump-

, f tions must be made. The amount of hydrogen present in the containment at any point in time is subject'to the uncertainties in the prediction of core slumping, vessel failure, debris interactions in the cavity, etc. For- any set of conditions the composition of the atmosphere and

'; its pntential flann: ability can be tracked as a function of time. Except I in the presence of igniters, however, the occurrence of ignition cannot.

(

[

be predicted and must be assumed. Typically containment integrity would j

'. be challenged by large coherent burns, but would not be challenged by L

'E extended combustion; the timing of the ignition is the key difference  !

$ between the two predictions, j j Tables 6.6 and 6.7 provide information on which the calculations  !

g ,

of radionuclide transport and deposition in the containment were based. '

j Table 6.6 gives containment geometrical data and Table 6.7 provides f containment spray parameters. ,

?

.a. .: , (:- m ' ~~N ' ' ~" ~= " Lc .:1Q '-2 w w -~:W -n

7 . .. . ... _ . . . . . - . . . -

l 6-55 J

6.2 Radionuclide Sources 4

. 6.2.1 Source Within Pressure Vessel i Inventory I

s l The reactor fission product inventory which was used in all 4

four sequences considered in'this report is based upon ORIGEN calculations i

for the Surry plant with a three region model with the maximum burnup 1

corresponding to 33,000 MW days / ton. Table 6.8 contains the inventory ,

of the fission products and control rod structural materials. Since ,

l

,,; release rate information is not available for all of these species, rates j for members of the various groups considered in the Reactor Safety Study were taken to be equal when no other information was available.

The nonfission product materials constitute the bulk of the aerosol mass released during core melting. The value for Ag,in this table is based on a total of 1060 control rods composed of 81 mol percent Ag, 14 mol percent In, and 4.9 moi percent Cd. The inventory of fission I

.{'

products was distributed according to the power peaking factors in '

.; Table 6.9, and the control rod and structural materials were distributed

] according to the volume of each radial node. ,

Release From Fuel The rates for radionuclide re hase from the fuel were computed j using the CORSOR code for the core temperature profiles specific to each l

accident sequence. The mass of each species of interest released as a function of time is shown in Figures 6.28 through 6.32 for the AB, TM.B',

V, and 5 20 sequences. Two runs of CORSOR were performed for the S 0 2 l sequences,e and y. This was done because the conditions which give

[ rise to the y containment failure mode engender somewhat different j; temperatures in the melting fuel, and a significantly different time to ll bottom head failure. In these figures, aerosol materials were considered ll J

to be the sum of fission products Sb, Sr, Ba, Ru, Mo, Zr, Ag along with l,' . nonfission products, Fe, UO 2 , Zr (cladding), Ag, In, Cd, and Sn. After l4 -

l t

t;

'l .

. g.,- - -~ - ~ - -

-- a..

. a ra -:: . ;c .n. a. . .. _ _ _

c. - s -; . v . .:: + -

• p t:: r. a= -J s. . . . ..

a j

i i 30 7,

jA i ', ,

^

E m '

CSI ih.

!j

.x v

m 20-y W AEROSOL X .01 4

r m .

W 4 .

I s N

0 ,

CSOH X .1

b. 10 -  ?

8:

ld El /~y j s.

E

'_ / ' TE e

0 , , , , ,

? O 1000 2000 3000 4000 5000 6000 i- TIME (sec) ji 4

i FIGURE 6.28. CORSOR PREDICTIOki 0F SPECIES HASS RELEASED FROM THE CORE DURING THE AB SEQUENCE FOR THE SURRY PLANT f '. .

t

1 e

4 8 6-57 5

t i

o e e

.o a s [

• W

=

t x , ,. - -

-i o e 3

as - o s=

i

- e i

1 gl _o n 8 m e a v =c v W I x

, 1 e i

W il o ie

.: _o g o

o, e 1 >

n o e m

o a o d a

_oc &. ^

=

3 me -

Iz

! ._E =2 i

)\ i v >.

me s ,

\ o, Q

asm J

l O u.

CW I

i w mg

! sm

-o I - La.

.u m CU o mm i

s _o n 1

s o

~ om s

m l 2Q N a:

. N o  ; ,

o o o o m j m u - -

3 i

4 C aj C

(6>t) pesoela; sson

" D w $ , "= ,- _

Y ewa4 g

___,_'ar Ae$ _

[

dhI . .

_4 I

- =- ,..... . - _ , _ . .

_ ., , ,,. g

i i.

i l s-58 i,

4 i

i ,

o o

o w b i:

.i- s *

, e Z

} x s

,, - o a o x O w

\, o g

- e e to u a E ais w

=

o 5

\ _o

.1 =

1 c m o S i

9 m

a s or o 4 i o @O m

, *$ I -

um

_a o e um

, w

. g o m '?*

m, t

4

) oN g .=

OM t mw ia g sis

,i e , o pM C

i o .o u.

e o we a a s ms gs.

mm.

s s, oc w e 8-

_o

\;

N. 3 5

t W

Ii . -

i

,t

~ -o .=

a u 4 i

1 o o o o r

m N '

'W l1 (63) pasoela; sson

't i

i 1

., i m- pc~mmyrw rstscawawm..s r. es-- g _ . . m (.

.-x.- .

Pmme .

. : r.Pnrespm o

. =._ s .

,4 .

30 e

4

-t i

\

.< m .

CD / CSI '

H v

.x -

'?

20-y) y)

,,.______'2.-

TE 7 --

m4 y 4 - -.- 7 , # .- AEROSOL X .01

\' s' /

's (10-l' CSOH X 1

? .

LaJ

• II *

/'

i 0

~

l , , , , ,

0 1000 2000 3000 4000 5000 6000 7

TIME (sec)

~

FIGURE 6.31. CORSOR PREDICTIONS OF SPECIES MASS htLEASED FROM THE CORE s

', DURING THE S20-c SEQUENCE FOR THE SURRY PLANT e

h

,1) .

Y ,

i o

a

Wn d-1 , * 'n 1 t r ' m I "- ^J ' ' ' :' /~'?~ p c' ' '

n', >; * " ?' E. D ~ t' " ' r " ' * *.

i ___.

j _

i

, e'l 4

l

, .I 30 l i 8

l 4

i, Csl 1

l ...- -

^

l l 3

0) Y,-[ Te

) U 20- #

/

! C / Aer0 Sol x .01 w' f-U /

i

/

/ Cs0H x .1 .

I m 10 - ,' '

4 o 2

i/

i, -

l 3 1 L 0 500 1000 1500 2000 TIME (sec)

FIGURE 6.32. CORSOR PREDICTIONS OF SPECIES MASS RELEASED FROM THE CORE DURING THE S 0 y SEQUENCE FOR THE SURRY PLANT 2

S .

l.

i .

1 a

____. ___ . ... . . .~ ._ __ .

^

6-61

melt-through of the RPV release during the core-concrete interaction was taken as a release to the containment. The inventory available for

release during the core-concrete interaction is listed in Tables 6.10 l and 6.11.

6.2.2 Sources Within the Containment Radionuclides enter the containment as they are transported through the primary system and on melt-tnrough of the reactor pressure vessel, that material still suspended in the RCS is transported into the

! containment as the RPV and containment pressures are equalized. The final source considered is that material released during the core-concrete interaction. Because of a lack of release information or, even more generally, a lack of evidence that they are of potential importance, sources sometimes postulated as arising from stearc explosions (oxidation release) or from jet emission of ho't, molten corium at the time of RPV failure were not included in these analyses. F Release from Primary System i

The source to the containment of material penetrating the i

i primary system is defined in mass input rate by species of interest and ,

on a time-dependent basis by the output from the TpAP-MELT calculations.  :

l Also provided in the TRAP MELT output is the size distribution of the l l

l particulate material. This calculated information is included in the i

j subsequent report section on results. I'

, Release from Core-Concrete Interaction 1 The VANESA code (Sandia model described in Volume I) was used l,

!~ to make predictions of aerosol and gas release rates and compositions as

, a function of time. Corposition of the core materials contacting the l concrete was as determined with the CORSOR code to be the materials remaining in the melt at the time of head melt-through. These compositions l!.

1 for the various sequences are given in Tables 6.10 and 6.11. The concrete '

l

,,,.,yw -w- ~v e w~ - ' '

e ~ "

'Q

~~

g my ', ' Q'

...,.L_,- .E-~ - ' * * * ' " " ~

I i

i 6-62

l. was taken to be casaltic (CACO 3 = 0.05, CaOH = 0.09, 5102 0 = 0.60, l H2 O = 0.04, Al 02 3 = 0.22) and the initial temperature of the molten l1 material was as calculated with the MARCH code. The total release rates and composition of the release are given in Tables 6.12 through 6.16.

These ratas and compositions define the sourcs to the drywell after

~

, vessel failure.

I =

r .

l .

I t

.  ?

I l

t 1.

l

'1

=. ;

~

P-p 4

's '

~

l i + .

dj

's i

]

, I,. . .

I .

'.4

j. -

TE Tj ll<)

' f,d l}

fi a,

a

.e i T

3

.k_ ry-----yp y:<x~~w+s.m y L9=+-f3 p++ ~~3_,;w.n - 2 m n,+myg g w}

~.

~

i L ,

6-63

TABLE 6.1. REACTOR CHARACTERISTICS, CONTAINMENT PARAMETERS, AND MARCH

. OPTIONS FOR LWGE ORY PWR CONTAINMENT i

i

! ECC storage and iniection tanks Accumulator RWST l Weight of water 171,300 lb 77,700 kg 2.92 x 10 lb6 6 1.3245 x,10 h; i Initial pressure 665 psia 45.9 MPa 14.7 psia j

i Temperature 120 F 0.1 hPa g 48.89 C 45 F 7.22 C Fractional value of RWST to start ECC recirculation: 0.01

,' Fractional value of RWST to start spray recirculation: 0.143 Laroe LOCA blowdown l

l Tine, rain Enthalpy Blowdown Rate Stu/lb J/kg 1b/ min ko/s 0 602.7 288,400 2.115 x 100 1.599 x 104

.20 602.7 288,400 2.115 x 10 6 1.599 x 10 4

!l '

.201 89.73 42,930 2.770 x 10 5 2.094 x 10 3

401 89.73 42,930 2.770 x 10 5 2.094 x 10 3 Calculated model input

.i l

Core heatup section:

Number of radial iones: 10 Number of axial zones: 24 ,

Meltdown model: BOII. model A Core melting temperature: 4130 F (2277 C) l .. Core slumping: Starts when lowest node in a zone has melted

'O Core collapse: Occ n when 75 parcant of core has meltM Zircalloy - water reaction: Urbanic-Heidrick reaction rate data, hydrogen j,. blanketing, steam limited, continues for melted ,

nodes, reaction of molten Zircaloy in the bottom  ;

s .

head calculated. ,,

1 I i

)

m_

u . . . _ . .g . - _. + . . . . , .:. .

~ - - - - - -

_4 .. g

j

s .

l .

6-64 i

! TABLE 6.1. (Continued) i Q

.4 End of blowdown conditions for large LOCA:

.a Water in vessel: core covere<1 Peak core temperature: 1700 F (927 C)

.; Accumulators: empty ,

1 Bottomheadfailuresociion:

,. Head melting temperature: 2800 F (~.538 C)

Debris melting temp,arature: 4130 F (2277 C)

., Heat loss frcm top'of debris: Radiation to core barrel Debris thermal condsctivity: 8 Stu/hr ft F (0.1384 w/cm/C)

) Tensile strength of vessel: e = min (80.000, 1.49 x 1016 TEMP-3.9105),

lb/in2 9 Reactor cavity processes, debris fragmentatfor.:

Particle diameter: 0.5 inch (1.27 cm) i Particle thermal conductivity: 2.0 Stu/hr ft F (0.0346 w/cm/C)

,i Reactor cavity processes, concrete decomposition:

j -

Metal-concrete interface heat transfer coefficient: HIM = 0.01 w/cm2 K 0xide-concrete interface heat transfer coefficient: HIO = 0.01 w/cm2 x h- Top surface emissivity: E = 0.5 Heat to cover water: surface boiling plus 50 percent of area radiating '

at internal temperature of top layer.

d Containment Section:

'k Atmosphere-wall heat transfer coefficient:

[ h = he (TSAT-TWALL) + 0.19 (T-TWAL!.)4/3 /(T-TWALL) h f he = 0 if TSAT < M1  ;

,y 2.0 < hc= Uchida data < 280 Stu/hr ft2 F 7 Containment break area: 7.0 ft2 overpressure failure (0.65 m2 ) ,

.,; 2 0.349 ft2 isolation failure (0.0324 m ) l o -

L  !

r: . .

9 .

i

. , , __ _ _ _ P4 WMeruuk' s- - . v tWii..= ~ m'.4 M9y- E." .

a. - T *T Tr . T AW.; .,*'e =4 < pgime. - ~ .-

" * * ~ . v. 4

l 6-65 TABLE 6.1. (Continued) i 1

i.

t

.'. j  ; Failure of safety systems:

.. ] (1) Containment failure fails the containment spnys J.1  ; (2) Containment failure fails ECR if sump is saturated.

l.

t s:

d

.i .

l

't

1 t

d i

i i

t j

9 J i

, \

I e

b l

1 a

= , - " *E ' '- "* " ' ' ' '

W" . .l h e ' ~ hfMN- --w p N N Wmg .. - . , Mm_

_"A ._.._r

-w +.w e.y.-ll

. _ _ .- ~ . . _

1 6-66 TABLE 6.2 ACCIDENT EVENT TIMES y- , Event Time, minutes

p Surry AB-c

• (

5 Core Uncover 9.4 -

~ '

Start Melt 24.8 t

,p Core Slump 42.1

~r Core Collapse 43.3 j Bottom Head Dry 64.4 i

' Bottom Head Fail 110.1  !

Start Concrete Attack 110.1 f Containment Leak '

1450.6

{

End Calculation 1639.6  !

'i l

.; Surry AB-Y q

.I; Core Uncover 9.4 o Start Melt 24.8

s. Core Slump .

42.1 1 -

e.

Coie Collapse

{ 43.5  !

% Bottom Head Dry 64.4

~ Bottom Head Fail 110.1 Start- Concrete Attack

{ 110.1 p Hydrogen Burn 268.9 0 Containment Fail 268.9 -

l[ End Calculation 710.7 6

a u .

'j -

f -

E.

i d

4 b

b .

s re; ' ~ "

" ' " ~ '

~

T ' l3.7 ~ '~.2M K= J h.l? .i ' _L lL ., ,.1

~'

~' ' " ,l' ,.

q 6-67 i -

TABLE 6.2 (Continued)

. Event Time, minuter

.i Surry AB-8 n

4 Volumes 2 Volumes a .

j Core Uncover 7.1 9.4 g Start Melt 24.6

~ 27.2 q Core Slenp 41.5 44.6

. j

-4 Core Collapse 42.5 45.6 Bottom Head Dry 62.5 65.8 i

Bottom Head Fail 97.9 104.3

I Start Concrete Attack ~ 97.95 104.4 j Hydrogen Burn 108.1 198.2 ll Hydrogen Burn 129.1

Hydrogen Burn 145.4 I

• j Hydrogen Burn

. , , 226.9

.7 ,; End Calculation 608.3 704.6 1:

.t Surry TMLB'-6,_

j! Steam Generator Dry 67.5

. I iL Core Uncover 95.5 Start Melt 118.3

!, Core Slump 146.3 Core Collapse 148.0 .

Bottom Head Fail -

152.8 Containment Fail 152.9 Reactor Cavity Dry 177.2 l] Start Concrete Attack 254.2 -

End Calculation 1073.4 I

l'  ;

j g i

u .

ri f

1 (gpygyb#ff'.7CTL _ _ . . _ - . l_ --

NIIS-Lhj

.'r j

.9 1

-1 . .

'l .

6-68 il -

I'ABLE 6.2 (Continued)

r. .

Event Time, minutes

.a Surry TMLB'-6) _

{ Steam Eanerator Dry 67.5 Core Uncover 95.5 ,

t Start Melt 118.3

[ Core Slump 146.3 Core Collapse q -

147.3

{ Bottom Head Fail 157.3 1 -

Reactor Cavity Dry 214.9 Start Concrete Attack 289.9 s Containment Fail 738.2

.J End Calculatior 1100.0 t

Surry S2D-c b Containment Spray Injection On

j. P.O.0

( ]. Containment Spray Recirculation On 25.0 Core Uncover 27.8

.', Accumulators' Empty -

91.5 -

g

-( Containment Spray Injection Off 114.7

.J. Start Melt 134.6 ,

' c:;~ Core Slump 147.3 i l'$

i Core Collapse 148.8 '

.) Botton Head Fail 22,'. 5 .  !

j. Reactor Cavity Dry 325.7

$ Start Concrete Attack 407.7

' I' End Calculation 2210.4

y -

i$ '

~

a.

c

' l'

! I.

y

,- i.

n _. ,

.,- __ ----,..t,- ,

i 6-69 TABLE 6.2 (Continued)

Event Time, minutes i

Surry S20-Y

)

il Containment Spray Infection On 20.0 Containment Spray Recirculation On 25.0 4 Core Uncover 27.8

[f Accumulators Empty 91.4

.! Containment Spray Injection Off 114.7

_.j Start Melt 134.0 Core Slump 146.6

~

)3 Core Collapse 150.8 Bottom Mead Fail 163.6

7. ,, Containment Fail .

163.7

>' Reactor Cavity Dry 264.9 Start Concrete Attack 336.9 g

End Calculation 1114.6

-{ Surry V

..t

. Containment Fails 0.0

,j Core Uncover 3.9 y Start Melt 39.7 Core Slump 56.6 Core Collapse 60.5 Bottom Head Fail 149.9 s

Start Concrete Attack 149.9 .

End Calculation 750.2 e

a a -

a s

'l

~ ' ~ ~ ' ~ ~ ' ~ ~ ' ~

-,7 g : __ ;c ';;;_ . . _ _ .'" ...m S W5+ r"' ' ' v' ' ~'" ' . - _ _ " ~ .

E

4 '- -isEE i-l i

Q y;- : = : r;; 1 = ~ -- A- -~

.q

_ 33 _. _ __ ._

i:

1 1

TABLE 6.3 CORE AND PRIMARY SYSTEM RESPONSE l

di Primary

. , Primary System jJ System Water Average Core Peak Core Fraction Fraction i Accident Time, Pressure, Inventory. Temperature. Temperature, Core Clad Event minutes psia Ibn F F ' Melted Reacted 4

Surry TMLB-6) _

Core Uncover 95.5 8.58 x 104 2369 669 675 0. O.

Start Melt 118.3 2366 5.65 x 10 4 1990 4130 ' O.00 0.06 Start Slump 146.3 2362 5.37 x 10 4 3709 4147 0.55 0.33 Core Collapse 147.3 2364 4.79 x 10 4 4130 ---

0.82 0.58 Bottor Head Fall 157.3 2368 1.95 x 10 4 3820 --- --- 0.59 I i'

~

6,_

, Surry Tit.B o Core Uncover 95.5 2369 8.68 x 10 4 669 675 0. O.

Start Melt -

118.3 2366 5.65 x 104 1990 4130 0.00 0.06 i Start Slump 146.3 2362 5.37 x 104 3'658 4150 0.55 0.33 Core Collapse 148.0 2367 4.52 x 10 4 4130 ---

0.79 0.89 i

Bottom Head Fall 152.8 2369 9.33 x 10 4 4130 --- ---

0.93 Surry S20-Y l Core Uncover 27.8 1164 1.01 x 10 5 577 585 0. O.

Start Helt 134.0 292 7.03 x 10 4 2067 4130 0.00 0.08 Start Slump 146.6 131 6.37 x 10 4 3688 4139 0.58 0.48 l Core Collapse 150.8 337 5.88 x 10 4 4207 ---

0.77 0.89 ,

Bottom Head Fall 163.6 617 2.46 x 10 5 3925 --- ---

0.89 s

._p g ,

., r i _ m.. r. _ . & ,m_ . _.

_ -....x v. -..g 3 ,; ro a1 ,

r.g e in ca n__, _ _,. _ _ , ;

. l,Il a

/y, i

' p j' s

l: TA8LE 6.3 CORE AND PRIMARY SYSTEM RESPONSE

,..f

>h s

~

,W "rimary i lf Primary System '  ;

' 4.i System Water Average Core Peak Core Fraction Fraction i1g Accident Time. Pressure. Inventory. Temperature. Temperature. Core clad iW

. !t Event minutes psia Ibn F F Melted Reacted .

} Surry S20-c l,

~

Core Uncover 27.8 5 1164 1.01 x 10 577 585 0.

jf Start Melt 134.6 293 7.04 x 10 4 2046 4130 0.00 O.

0.07

' I#I Start Slump 147.3 129 6.38 x 10 4 3658 i

4147 0.53 0.48

! ,' Cora Collapse 148.8 213 6.20 x 104 4130 ---

0.76 0.59 i

I

'lf s Bottom Head Fall. 227.5 18 2.% x 10 5 4130 0.60 T

~

l )h Surry V

~

i Core Uncover 3.9 617 2.20 x 10 5 567 597 0. O.

j Start Melt 39.7 24 7.97 x 10 4 2094 4130 0.00 0.05 i Start Slump 56.6 17 7.74 x 10 4 3597 4135 0.49 0.22 1 i Core Collapse 60.5 122 7.31 x 10 4 4130 1 ,

0.75 0.41

-{ Bottan Head Fail'

! 149.9 15 0. 4130 --- ---

0.41 Surry AB-8 (2 Volumes)

Core Uncover 9.4 44.2 8.32 x 104 288 l

296 0. O.

Start Melt 27.2 36.3 5.64 x 10 4 1952 4130 0.00 0.05 Start Slump - 44.6 30.2 5.34 x 10 4 3629 4132 0.54 0.24 Core Collapse 45.6 31.7 5.05 x 10 4

.! 3729 ---

1.0 0.40 ff.4 Bottom Head Ory 65.8 36.4 0. 3234 --- ---

0.40

/,0 Bottom Head Fall 104.3 24.9 0. 4130 A

• 0.40

4. :'

s -----

gym t; ~ n . . .u :;>:. y } -: :- ..:~G. - : m'~-_=

t, -

I 1

j TABLE 6.3 CORE AND PRIMARY SYSTEM RESPONSE

i Primary '

Primary

,;, Systes

System Water Average Core Peak Core Fraction Fraction

j. . Accident Time. Pressure. Inventory. Temperature. Temperature. Core Clad Event minutes psia lha F F Melted Reacted

, Surry AB-8 (4 Volumes)

Core Uncover 7.1 41.3 8.40 x 10 4 ,.

286 295 0. O.

i Start Melt' 24.6 29.8 5.62 x 105- 1962 4130 0.00 0.05 Start Slump 41.5 24.0 5.34 x 10 4 3623 4131 0.54 0.24 Core Collapse 42.5 25.2 5.08 x 10" 4207 ---

0.76 0.39 Bottom Head Dry 62.5 29.4 0. 3236 --- ---

0.39 I. Bottom Head Fail 97.9 18.7 0. 4130 ' --- ---

0.39  ?*

M

, Surry A8-c/Y Core Unct,ver 9.4 40.2 7.50 x 10 4 2% 357 0. O.

Start Melt 24.8 35.6 5.70 x 10 4 1960 4130 0.00 0.05 Start Slump 42.1 31.8 5.33 x 10 4 3644 4139 0.54 0.24

,, Core Collapse 43.5 33.5 5.02 x 10 4 3732 ---

1.0 0.40

,l i, Bottoia Head Dry 64.4 41.2 0. 3299 --- ---

0.40 I Bottom Head Fall 110.1 33.9 0.

i 4130 ---

g --- 0.40 e

e I

• w a ==

._ % . ,.  : - -] ,-j _. L ' - b . : i*. ' _ _ _ - h _ lN , ~ s: ~

~

Q[ - ; _ ..; _ w ...a...... _ ..l

{ . .i

+ 'If.

q

.l t

O  : TABLE 6.4 CONTAllMENT RESPONSE L

?( .

1 '

Reactor W Sump Reactor Cavity Steam l Compartment Compartment RWST or CST Sump Water Water Cavfty Water Cond.

p; Accident. Time. Pressure. Temperature. Water Mass, Mass. Temp., Water Mass. Temp., on W ils

i. > ?, Event minutes psia F Ibn Ibn F 1ha F Ib_njain '

Surry TMLB-aj _ i, S'ceam Generator Dry 67.5 13.0 136 3.0 x 106 3.39 x 104 138 0. --- 1,051 P Core Uncover 95.5 ?8.8 219 3.0 x 106 2.07 x 10 5 197 0. --- 2,498 (

, b. Start Melt 118.3 26.7 209 3.0 x 100 2.47 x 10 5

200 0. ---

994 b hl Start Slump , 146.3 22.5 197 3.6 x 10 6 2.69 x 10 5

198 0. --

584 6

Core Collapse 147.3 24.9 205 3.0 x 10 2.70 x 10 5 197 0. --- 1,378

!q , Botton Head Fall 157.3 45.9 253 3.0 x 10 6

2.82 x 10 5 5 198 1.71 :t 10 120 11,810 p I -

Start Debris / U

'3.0 x 106 1 Water Interaction 157.3 46.0 253 2.82 x 10 5 198 1.71 x 105 120 12,400 .

I 6

Cavity Ory 214.9 58.6 272 3.0 x 10 4.06'x 105 221 0. ---

1,405 i

Start Concrete Attack 289.9 46.0 253 3.0 x 106 4.54 x 10 5

725 0. ---

653 Containment Fall 738.2 53.7 249 3.0 x 10 6 4.85 x 10 5 227 0. ---

0 End Calculation 1100.0 14.8 263 3.0 a 106 4.66 x 10 5 180 0. ---

0 Surry TMLB-6,_

Steam Generator Dry 67.5 13.0 136 3.0 x 106 3.39 x 10 4 138 0. ---

1051 I

Core Uncover 95.5 28.7 219 3.0 x 106 2.07 x 10 5 197 0. ---

2498

!-i Start Melt 118.3 25.8 209 3.0 x 10 6

2.47 x 10 5

200 0. ---

990

! 4 Start Slump 146.3 2?.3 196 3.0 x 10 0 2.68 x 10 5 196 . O. ---

668

[ Core Collapse 148.0 26.5 210 3.0 x 106 2.68 x 10 5 196 0. ---

0

< i. -

1 y .

vu.m_Lwm..w.;~ n. . _ . .-- -

g. .

U:

i

,] .

TABLE 6.4 CONTAINMENT RESPONSE 1:

! ,1 3- Reactor i Sump Reactor Steam i

Compartment Compartment RWST or CST Sump Water hter Cavity Cavity hter . Cond.

. Accident Time. Pressure. Temperature hter Mass. Mass, Temp., h ter Mass. Temp.. on Walls p Event minutes psia F 1ha Ibn F Ibn F Ibm / min j' (Surry TM.B-8, Continued)

'! Botton Head Fall 152.8 46.5 252 3.0 x 10 6

2.75 x 10 5

197 0. --- 12,300 Start Debris /

Water Interaction 152.8 46.6 252 3.0 x 10 6 2.75 x 10 5 197 1.71 x 10 5 120 12,510 6 5 Containment Fall 152.9 06.0 336 3.0 x 10 2.76 x 10 197 8.82 'x 10 3 109 0 Cavity Ery 177.2 14.7 202 3.0 x 106 3.16 x 105 202 0. ---

238 Start Concr.ete

• m Attack 254.2 14.6 187 3.0 x 10 6 3.30 x 10 5

187 0. ---

278 4 End Calc,ulation 1073.4 14.8 207 3.0 x 10 6 3.33 x 10 5

159 0. ---

0 '

. Surry S2D-c lI' Ccntainment Spray 1 Injection On 20.0 23.4 201 .2.90 x 10 6 2.14 x.10 5

194 3.96 x 10 2 196 3.184 C:ntainment Spray

.;, R: circulation On 25.0 20.9 191 2.77 x 10 6 3.58 x:10 5 191 4.14 x 10 4 189 1,301 l ,

Cere Uncover 27.8 19.4 183 2.70 x 106 4.00 x 10 5 184 6.53 x 10 4 185 823 Accumulators Empty 1.04 x 106 2.08 x 10 6' 3

91.5 12.6 126 131 2.46 x 100 128 0.

Start Melt 134.6 13.5 136 4.12 x 10 5 2.75 x 10 6 131 2.46 x 10 5 128 113 5

Ccre Slump 147.3 14.2 13? 4.12 x 10 --- ---

2.46 x 105 128 27 Ccre Collapse 5 148.3 14.4 132 4.12 x 10 --- ---

2.46 x 10 5 128 45

[ B:ttom Head Fall 227.5 13.6 144 4.12 x 105 2.81 x 10 6 141 2.46 x 10 5 128 355 ,

I

_ w. _ _: ..z. ;T ~ - .E . J . :- - ~ - - - - - -ex--

. . :. :2

. ;7.. -

~!

.i

( TABLE 6.4 CONTAllMENT RESPONSE 3

i -]f.

. Ktsctor

.. Sump Reactor Cavity Steam

f Compartment Compartment RWST or CST Sump hter hier Cavity hter Cond.

, ! p; Accident Time. Pressure. Temperature, h ter Mass, Mass. Temp., Water Mass. Temp., on Walls

! Event minutes psia F 1ha Ita F 1ha F lbm/ min f; (Surry S2D-c Continued)

!l t

Start Debris /

<d Water Interaction 227.5 13.6 144 4.12 x 10 5 --- ---

2.46 x 10 5 128 .i75 Reactor Cavity Ory 325.7 16.1 162 4.12 x 105 3.06 x 10 0 164 0.

lI 15 Start Concrete Attack 407.7 13.7 141 4.12 x 10 5 3.06 x 10 6 147 0. --- 0 h End Calculation 5 2210.4 13.8 131 4.12 x 10 2.88 x 106 132- 2.39 x 10 5 212 12 l-i' ili Surry S2D-Y n -

Containment Spray Injection On 2.90 x 106 20.0 23.4 201 2.14 x 105 4.76 x 102 t[1 8 Containment Spray 194 196 3,185 Recirculation On 25.0 20,9 2.77 x 106 3.58 x 10 5 190 191 4.iS x 104 189 1,299 2.70 x 106 Core Uncover 27.8 19.4 WJ 3.99 x 10 5 4 184 6.52 x 10 185 824 d Accumulators Empty 91.4 12.6 126 1.04 x 106 2.08 x 10 5 131 2.46 x 10 5

131 0 Start Melt 134.0 13.5 4.15 x 106 2.75 x 106 136 2.46 x 106 3

' 130 128 114 Core Slump 146.6 4.15 x 105 2.76 x 106 14.2 132 132 2.46 x 106 128 24

! Core Collapse 150,8 4.15 x 10 5 2.76 x 106 15.4 136 132 2.46 x 10 5 128 188 Bottom Head Fall 163.6 23.1 4.15 x 105 2.77 x 10 6 I

186 134 2.46 x 10 5 128 6,254 Start Debris /

Water Interaction I 163.6 24.0 4.15 x 105 191 2.77 x 10 6 134 2.46 x 10 5 128 6,561 x

-l f

y ii-y-gy,gyy ,sQg ;3;g . ;. . .;.f. Q pish 5.y .: m .{3) c, ,

'e Y %,Q .;, w.g3 , _ ,, ;_

Q 4' ; .

+q TABLE 6.4 CONTAI MENT RESPONSE

,3l ?

m Reactor Reactor Cavity

-:p ((I'Accident Sump Steam

[ Compartment Compartment RWST or CST Sump hter hter Cavity Water cond.

Time. Pressure, Temperature. Water Mass. Mass. Temp., Water Mass. Temp., on Walls p; Event minutes psia F 1hm lha F

.4 lte F Ibm / min ft (SurryS20-YContinued)

Containment Fall 163.7 88.6 1696 4.15 x 10 0 2.78 x 10 0 134 0. --- 0 f, Reactor Cavity Dry 264.9- 14.6 203 4.15 x 10 5 2.88 x 106 137 0. ---

491

'd Start Concrete

[ Attack 336.9 14,7 184 4.15 x 105 2.89 x 106 137 0. ---

0 l[,1 End Calculation 1114.6 14.8 2E8 4.15 x 10 5

2.90 x 10 6 137 0. --- 0 k -

cn e

' j, Surry AB-c 4

[

Core Uncovers 9.4 40.2 246 3.0 x 106 3.45 x 105 247 0. ---

3.878 Start Melt 24.8 35.5 236 3.0 x 106 3.78 x 105

[; 237 0. ---

1,797 Start Slump 42.1 31.8 225 3.0 x 10 6 --- ---

O. ---

0 Core Collapse 43.3 33.6 249 3.0 x 106 --- ---

0. ---

0 Botton Head Dry 54.4 41.1 245 3.0 x 10 6 --- ---

0. ---

1,793 Bottom Head fall 110.1 34.0 POS 3.0 x 106 4.42 x 10 5 227 0. ---

777 Start Concrete Attack 34.0 3.0 x 106 l 110.1 205 4.42 x 10 5 227 0. ---

0 Centainment Leak 1450.6 45.8 2.0 3.0 x 106 4.93 x 10 0 227 0. ---

0 End Calculation

, 1639.6 43.0 288 3.0 x 106 4.93 x 10 0 227 0. ---

0 r -

Surry AB-Y*

N Containment fail 268.9 87.4 1152 3.0 x 10 6 5 4.52 x 10 227 0.

N End Calculation 710.7 14.9 271 3.0 x 10 0 4.45 x 10 5 213 0

-r3

0. ---

0 -

G

• Containment response same as Surry AB-c out to Start of Concrete Attack.

. r alEW----_____,._m._ ,

y- , .

qq .

~

)?

?

I il TABLE 6.4 CONTAINMENT RESPONSE  ?

1 hi -

b Compartment Reactor j Compartment Pressure. Tempe ature. Sump Reactor Cavity Steam

) RWST or CFT Sump Water hter Cavity h ter Cond.

Accident Time.

D

h ter Mass. Mass. Temp. h ter h ss. Temp., on W ils

Event minutes 1 2 1 2 1ha Ibn F 1ha F Ibm / min a Surry V

-6 i{ Containment fails 0.0 10.0 15.7 100 ICI 3.0 x 10 6 0. O. O. ---

0/1163*

,l Core Uncover 6 3.9 10.0 16.5 100 290 3.0 x 10 --- ---

0. ---

0/0

Start Melt 6 4 39.7 10.1 14.8 100 210 3.0 x 10 3.38 x 10 45 0. ---

0/225 Start Slump 56.6 10.1 14.9 100 428 3.0 x 10 6 3.38 x 10 4 45 0. ---

0/0

, Core Collapse 60.5 10.1 15.0 100 739 3.0 x 106 , 3.38 x 10 4 45 0. ---

0/0 jj Bottom Head Fall 149.9 12.3 12.3 114 141 3.0 x 106 3.70 x 104 50 0. ---

1084/0 y

Start Concrete

[:1 Attack 149.9 12.3 12.8 116 131 3.0 x 106 3.89 x 10 4 54 0. ---

1413/0 If:. End Calculation 750.2 14.9 14.9 135 150 3.0 x 10 6 6.62 x 10 4 81 0. ---

0/0 5

Volume 1/ Volume 2 1

e 1

h 1 1 t

,I u

{riidrdrhtCd;;70d!at.WgpT W Kl%;pn.F T ";f4""- WNc!wn : " ' .pg2. ,._

t- r J

_a,_....,. . __ fi_. ,

ll

[l l

fJi '

(% TABLE 6.4 CONTAINMENT RESPONSE

-l.i  !

il

.i i[ Compartaent Compartment Reactor j Pressure. Temperature, Sump Reactor Cavity Steam Cavity

- psia RWST or CST Sump Water hter Water Cond.

T Accident Time.

F Water Mass, ~ Mass. Temp., h ter Mass. Temp., on Walls Event minutes 1 2 1 2 lbe its , F Ibn F Ibs/ min

>h Surry AB-s (2 Volumes) >

.1 6

i Containment. Fati 0.6 54.2 15.7 260 99 3.0 x 10 2.24 x 10 5 258 0. ---

3026/0*

(( Core Uncover 9.4 44.2 15.0 244 155 3.0 x 10 6 3.40 x 10" 245 0. ---

3815/319

. Start Melt 27.2 36.3 15.0 230 170 3.0 x 106 3.76 x 10 5

231 0. ---

1525/250 i Start Slump 44.6 31.7 15.0 241 171 3.0 x 100 3.89 x 105 218 0. ---

0/0 Core Collapse 6 45.6 51.9 15.0 244 171 3.0 x 10 3.89 x 10 5 218 0. ---

0/0 Bottom Head Fall 164.3 24.9 3.0 x 10 6 4.18 x 10 5 15.0 211 181 212 0. ---

521/0"[

ij, Start Concrete 6

. Attack 104.4 24.9 14.8 211 181 3.0 x 10 4.18 x 10 0 212 0.

534/0 0

End Calculation 704.6 14.9 14.8 184 186 3.0 x 10 4.03 x 10 5 176 0. ---

0/0 2

• Volume 1/ Volume 2 .

f.

i

-1 tb p>t .

,g-.

• ,*#w

• r . '- '**- "b ( e "[ *= ", a ..

3

'4

- :\

• r

, ,4

?

TA8tE 6.4 CONTAIIMENT RESPONSE 2

l[ Cospartment Compartment Tempe ature $,,, g,,,g,, Ey,'g g

I l Pres 85 [. Sid5T or CST - Susp Water Water Cavity hier <a W1'Is i4 Accident Tlas. idater Mass. Mass. Temp.. Idater Mass. Temp.. Ite/ min t ! Event minutes 1 2 3 4 1 2 3 4 lba Ibe* F Ibn F 2 3 4_

li 1

A Surry AB-s (4 Volumes)

,o

$ Containment Fall 0.0 15.7 15.7 15.7 15.7 110 109 90 174 3.0 x 100 --- ---

S. ---

0/0* 0/6155*

Core Uncover 7.1 41.3 41.3 41.3 41.3 126 193 255 268 3.0 x 10' 9.80 x 10 2 120 0. ---

18/1530 2082/1694

. Start IIelt 24.6 29.9 29.8 29.8 29.8 103 163 235 246 3.0 x 10 0

1.008x10 3 104 0. ---

0/310 697/0 j Start $1use 3.0 x 10 6 41.5 23.9 23.9 23.9 23.9 98 3 154 220 377 1.006 x 10 98 0. ---

0/262 453/0 .

3.0 x 100

'l Core Col' 2 42.5 25.2 25.2 25.2 25.2 109 163 239 614 1.008 x 10 3

98 0. ---

0/584 0/0 Sottoe * .d Fall 6 f 97.9 18.8 18.8 18.8 18.8 124 187 221 215 3.0 x 10 1.139x10 3 99 0. ---

50/0 0/172 1 Start .,acrete

'i Atti 98.0 18.8 18.8 18.8 18.8 127 187 221 215 3.0 x 106 1.143 x 10 3

0.

100 ---

66/0 0/156 j Iv- alculation 698.3 14.8 14.8 14.8 14.8 468 148 279 173 3.0 x 106 2.141 x 10 118 0. ---

0/42 0/0 m iI -

' to

{ Volume 1/ Volume 2 Volume ?/Volur.4 4.

5

't I

. 's 1

i4 84 t A, f.

L

+.

.+

g i,

'd:s

^ ~ ' " ~ ~ ~ ~

t

- E,.__..__._. _

' ~ ~ ~ * * - ~ " ~~ ' - * ~ -- -

r vr w_

n

. v- cdu :, ' ?:m _

e ~ nr: - " ;_*; _.y7,; ~'q:&2 y::::- 3, .. . ._ 3 .

TA8LE 6.5 SumARY OF CONTAINMENT LEAK RATES

.4 .

< I 1

I- CSIS CSRS l**I"8' Time Leak

. ;! Pnssure Tm J Subsequence Start, sin 'minEnd, min Start. min End.ein Interval,v/hr Rate.(a) Wa psia *F *C Remarks i TMLB-6 95.5 4.2E-4 0.20 i

29 219 104 Core uncovers 1

95.5-118 4.2E-4 0.19 28 217 103 Core heats

,;., 118-146 4.2E-4 0.16 24 201 94 Core melts lr , 146-148 4.2E-4 0.17 25 204 96 Core slumps and collapses

148-152.8 4.2E-4 0.20 29 216 102 Reactor vessel heatup 152.8 4.2E-4 0.31 46 251 122 Reactor vessel fails 152.8-152.9 4.2E-4 0.51 75 327 164 Bolloff of H 2O 152.9 4.2E-4 0.58 86 337 169 Containment falls T*

152,9-164.4 7.5 0.26 38 245 118 Bolloff of H 2O E 164.4-166.4 3.8 0.11 16 204 95 Bolloff of H 2O l

166.4-168.4 1.9 0.10 15 203 95 Bolloff of H 2O 168.4-173.2 0.9 0.10 15 203 95 8011off of H 2O f' 173.2-254.2 0.01 0.10 15 194 90 8011off of H 2O l 254.2-272 4.2E-4 0.10 15 186 86 Concrete decomposition 272-325 0.02 0.10 15 184 84 Concrete decomposition 325-567 0.09 0.10 15 180 82 Concrete decomposition l , 567-627 0.19 0.10 15. 204 96 Concrete decomposition lC 627-1073 0.08 0.10 15 222 106 Concrete decomposition s,

4

'TMLB-c 95.5 4.2E-4 0.20 f 29 219 104 Core uncovers

.l 95.5-118.3 4.2E-4 0.19 28 217 102 ' Core heats .-

4 7 , 118.3-146.5 4.2E-4 0.16 24 203- 95 Core melts i( (a) Nermalized to a containment free volume of 1.8 x 106 fg3. Units are volume fractions per hour. -

4.y - , _

w. . __ s . - -

., - - -a - ~.

2__ .

fpa

• r, u i

d I

,3 TABLE 6.5 SUMARY OF CONTAINMENI' LEAK RATES .

53

.i -

4 Leakage CSIS CSRS Time Leak ressure h Start. End. Start. End. Interval. Rate.(a)

Subsequence sin min min sin min v/hr Wa 9sta *F *C Remarks TML8-c(Continued) 146.3-147.3 4.2E-4 0.15 22 197 92 Core slumps and co11 apses

. 147.3-157.3 4.2E-4 0.20 29 216 102 Reactor vessel heatup 157.3 4.2E-4 0.31 45 252 122 Reactor vessel falls

(, 157.3-290 4.2E-4 0.37 55 267 136 Boiloff of H 2O 290-620 4.2E-4 0.32 48 248 120 Conchete decomposition

]li f

, 620-738.2 4.2E-4 0.35 52 248 120 Concrete decomposition 738.2 4.2E-4 0.36 54 249 120 Containment fails

. 738.2-742.8 7.7 0.26 38 231 111 Concrete decomposttion p 4 a

>j 742.8-770 1.97 0.11 16 236 114 Concrete decumposition E q

1 770-1100 0.10 0.10 15 275 135 concrete decomposition jg .

l S2D-c 20 2210 20 2210 27.8 4.2E-4 0.13 19 183 84 Core uncovers 1 27.8-134.6 4.2E-4 0.09 14 140 60 Core heats

! -j 134.6-147.3 4.2F-4 0.09 14 133 56 Core melts

!] 147.3-148.3 4.2E-4 0.10 14 132 56 Core slumps and collapses 4

} 148.3-148.8 4.2E-4 0.17 25 580 305 Hydrogen but ns l 148.8-148.82 4.2E-4 0.31 46 1424 773 Hydrogen burns l , }lj 148.82-148.83 4.2E-4 0.37 '55 1789 976 Hydrogen burns

!: f 148.83-149 4.2E-4 0.32 48 1337 725 Hydrogen burns l' [' 149-149.2 4.2E-4 0.23 34 659 348 Hydro 9en burns 149.2-227.5 4.2E-4 0.10 15 153 67 i Reactor vessel heatup .

b 6 (a) Normalized to a cor.tainment free volume of 1.8 x 10 ft 3 .

Units are volume fractions pe,r hour.

tg..

J- ._ .. _. ,

f. ,6Er0175,ir .1MfUhr. 4/tI f"TfC,7l ,""" ~ . - M(. C'"; ,17 .t '  : ,

4

.- . _ - - , _. _. m_..__. o

{ ,*l

.f,

J t .

TABLE 6.5 SUMARY OF CONTAINMENT LEAK RATES i-.

L'*"***

} CSIS CSRS Time Start. End, Start. End. Interval. Rate,(a)

Leak Pressure Tm i Subsequence sin ein ein sin sin v/hr Wa psia *F *C Remarks s

i S20-c (Continued) 227.5 4.2E-4 0.10 14 142 '61 Reactor vessel falls

f 227.5-227.6 4.2E-4 0.13 19 153 67 Initial bolloff of H O 2 j! 227.6-233 4.2E-4 0.21 32 226 108 Bolloff of H p0 l 233-407.7 4.2E-4 0.11 17 164 73 Bolloff of H 2O

': 407.7-1248 4.2E-4 0.10 14 145 63 Concrete decomposition 1 -

s 1248-2210 4.2E-4 0.10 14 134 57 Concrete decomposition i

• i S20-Y 20 91.4 20 163.7 27.8 4.2E-4 27.8-134.0 4.2E-4 0.13 0.09 19 14 183 140 84 60 Core uncovers Core heats

{

li 134.0-150.8 4.2E-4 0.09 14 133 56 Core slu.aps and collapses 150.8-163.6 4.2E-4 0.11 16 142 61 Reactor vessel heatup l,; 163.6 4.2E-4 0.16 23 186 85 Reactor vessel falls j, 163.6-163.7 4.2E-4 0.34 50 825 441 Hydrogen burns

! 163.7 4.2E-4 0.60 87 1795 980 Hydro 9en burns 163.7 4.2E-4 0.60 88 1726 941 Containment fails 163.7-164.5 11.1 0.50 73 1308 709 Hydro 9en burns 164.5-166.7 8.7 0.31 46 '588 309 Hydro 9en burns j 166.7-201.5 1.7 0.16 17 207 97 Bolloff of H 2O j 201.5-407.7 0.2 0.10 15 190 88 Bo11off of H 2O 407.7-477.6 4.6 0.14 20_ 478 248 Hydro 9en burns 477.6-479.1 11.8 0.24 36 1599 870 ~ Hydrogen burns -

6 (a) Normalized to a containment free volume of 1.8 x 10 ft3 . Units are volume fractions per hour.

U .

j

- .- - J, -,- - .-

,, y J. -

' e. ,

lt n -

1, ,..

t1 -

TA8LE 6.5

SUMMARY

OF CONTAINMENT LEAK RATES i}

}1

( ! ,' ** 88' CSIS CSRS Start. End, Sta rt.- End. In al, Ra e a) Pressure Tm Subsequence min min min ein v/hr

} min ,tfa psia 'F 'C Remarts

}';((,

S20-Y (Continued) 479.1-480.7 S.) 0.11 16 . 694 368 Hydrogen burns ,

L! 480.7-591.3 0.'l 0.10 15 256 124 Concrete decomposition j 591.3-591.5 9.2 0.23 34 1260 682 Hydrogen burns

,. 591.5-592.3 12.3 0.26 38 1785 974 Hydrogen burns

{i 592.3-593.3 9.6 0.14 20 973 523 Hydrogen burns

.] s 593.3-750.7 0.1 0.10 15 250 121 Concrete decomposition j 750.7-750.8 9.0 0.23 34 1249 676 Hydrogen burns

!j i

3 750.8-751 13.2 0.34 50 2252 1233 Hydrogen burns  ?

751-752.2 11.3 0.22 32 1470 799 Hydrogen burns O ':

! 752.2-758 1.2 0.10 14 425 218 Concrete decomposition 758-1115 0.07 0.10 15 238 114 Concrete decomposition 3

AB-c -- -- -- --

9.4 4.2E-4 0.28 40 246 119 Core uncovers

'J 9.4-24.8 4.2E-4 0.27 39 244 118 Core heats 24.8-42.1 4.2E-4 0.23 33 230 110 Core melts 42.1-43.3 4.2E-4 .23 33 238 114 Core slumps and collapses n 43.3-110.1 4.2E-4 0.25 37 238 114 Reactor vessel heats

}j 110.1 4.2E-4 0.23 34 229 110 Reactor vessel fails 110.1-1451 4.2E-4 0.29 42 232 111 Concrete decomposition

.! 1451 4.2E-4 0.31 46 236 113 Containinent leaks a

j 1451-1452.4 5.8 0.28 41 230 110 Concrete decomposition

1452.4-1639.8 4.2E-4 0.29 42 295 146 Concrete decomposition 3

h (a) Normalized to a centainment free volume of 1.8 x 106 ft . Units are volume fractions per hour.

L '

.l j '

yes _ . . . - -. . - -.- - -- --

g?y-- g-b-l:y:_ .;; & l i T* ] .I-yf II ,-;,-

, + -:

4 1

2

.l: TABLE 6.5 Sit 1 MARY OF CONTAINMENT LEAK RATES

.t; d CSIS CSRS l**k

Time Leak Start, End, Start. End, Interval, Rate, a Pressure Temp.

~

i j Subsequence sin sin sin sin min v/hr Wa psia *F *C Remarks l ', AB-Y .

-- -- -- -- 9.4 4.2E-4 0.32 46 246 119 Core uncovers j 9.4-24.8 4.2E-4 0.27 40 245 118 Core heats 24.8-42.1 4.2E-4 0.23 34 230 110 Core melts

] j 42.1-43.5 4.2E-4 43.5-110.1 4.2E-4 0 23 0.25 33 37 243 117 238 114 Core slumps and collapses Reactor vessel melts 110.1 4.2E-4 0.23 34 230 110 Reactor vessel fails

' l( 110.1-268.9 4.2E-4 0.25 37 230 110 Concrete decomposition 268.9 4.2E-4 0.46 68 770 410 Containment falls /H burns m

' ][ i 268.9-269.6 14.7 0.93 137 2567 1408 Hydrogen burns I

, 269.6-271 12.9 0.56 82 1809 987 Hydrogen burns j 271-271.9 11.1 0.33 48 1211 655 Hydrogen burns

! 271.9-274.2 9.5 0.19 28 769 '409 Hydrogen burns

.y j<

.r 274.2-277.4 4.4 0.11 16 443 228 Hydrogen burns

!i 277.4-299 4.2E-4 0.11 15 345 174 Concrete decomposition 299-710.7 0.07 0.11 15 281 138 Conct ete decomposition f

, (a) Normalized to a containment free volume of 1.8 x 10 ft. Units are volume fractions per hour.

l i .

l 9

~ . . ..

p ..- .

- m ..

m _

n .

I N. .,

9

,3 '

)

TABLE 6.5 SIMtARY OF CONTAllMENT LEAK 8ATES a ,,,, ,,,, te...

in Co .,t . o te. . . in Co ,t.ent 2

~

'i St., , E. >r - . 206) ar --

. .. i.u..t i. 2 ".v,,,r> > ..

St.,.. Ew 1->_

1 5 _. ... . .. .. .si. y C vm . .s,. y C -,t s

,) '

AB-s(2V) -- -- -- -- 9.4 0.32 0.30 44 244 118 4.'E-4 t 0.10 15 86 30 Core uncovers 5 .i 9.4-27.2 0.70 0.29 241 116 2.3 42 0.10 15 156 69 Core heats 27.2-44.6 0.50 0.22 33 223 106 2.2 0.10 15 170 77 Core .elts 44.6-46.0 0.59 0.22 32 244 118 2.3 0.10 15 170 77 Core slumps and collapses

~ j{

46.0-104.3 0.57 0.21 31 226 108 2.3 0.10 15 179 82 Reactor vessel heats W

104.3 0.45 0.17 25 212 100 2.3 0.10 15 181 83 Reactor vessel falls l '

1G4.3-104.4 1.96 0.17 25 212 100 2.3 0.10 15 181 83 Solloff of H O2

(' 104.4-198.2 0.42 0.15 22' 215 102 2.3 0.10 15 181 83 Concrete decomposttion m 198.2-198.3 94.1 0.13 19 219 104 0.44 0.14 20 1345 730 Hydrogen burns /n

,j 198.3-201.5 4.2E-4 0.13 219 104 2.3 0.10

• 19 15 540 282 Hydro 9en burns li

'i 201.5-257.4- 0.29 257.4-442.1 0.13 0.12 0.11 18 16 228 109 238 115 1.95 1.0

• 0.10 0.10 15 15 210 99 Concrete decomposition 192 89 Concrete decomposition 442.1-704.6 0.05 0.10 15 125 87 0.43 0.10 15 125 87 Concrete dectmposition

}

V (2V) -- -- --' -- 0.04-4.0 4.2E-4 0.07 10 100 38 169.0 0.12 18 232 111 Core uncovers 4.0-40.0 4.2E-4 0.07 10 100 38 14.7 0.10 15 214 101 Core heats

, 40.0-57.0 4.2E-4 0.07 10 100 38 1.6 0.10 15 253 123 Core melts

,j 57.0-61.0 4.2E-4 0.07 10 100 38 13.0 0.10 15 583 306 Core slumps ar.d collapses l 61.0-149.9 4.2E-4 0.07 10 100 38 3.3 0.;0 15 248 120 Reactor vessel melts j 0 ,,_

(a) Normalized to . containment free volume of 1.8 x 10 ft. Units are volume fraction /hr.

! (b) Norma 11:ed to a compartment free volume of 2.746 x 10 f t . Units are volume fraction /hr.

b i

?

o e

l'

.[

~ . - - - - - . . . . _ . . ,

f

3 gjr.p L--

eri

• 3rr -ey--j--- " - ^ -

1 i

• 1 .

s t

ii

'J 4

, ,5

'5 TABLE 6.6 $UHHARY OF CONTAINt1ENT LEAK RATES

-2 i --

! CSIS C$as Leakage in C apartment I teakaoe in Connartment 2 f Start. End. Start. End, la 41 R4Ie *I I'**' I**"

R (b) Pressure Teen

., subsceuence ein als ela als ein v/hr MPa esta *F _C w/hr MPa esta *F *C R,eearts i  ;

J* V (2V)*(Continued) 149.9 4.2E-4 0.08 113 45 0.08 12 4.2E-4 12 142 61 Reactor vessel falls h 149.9-159.4 4.2E4 0.10 15 118 48 4.2E-4 0.10 15 127 53 Concrete decomposition

? 159.8-371.5 0.11 0.10 15 126 52 0.76 0.10 15 165 74 Concreta decomposition

. 371.5-750.2 0.09 0.10 15 133 56 0.64 0.10 15 158 70 Concrete decomposition li A8-p(4V) -- -- --

0.9 0.2b 7,1 24.6 38 121 49 0.01/0.18/ 0.26 38 185 85 Core uncovers and heats (CcNartments 1 & 2 1.03 j are given first.

24.6-41.5 0.7 0.18 100 38 26 0.01/ 0.18 26 156 69 Core melts fumpartments 3 & 4 1.6E-3/  ?

areonpage9) 1.00 g l 41.5-42.5 0.1 0.17 25 104 40 02/ 0.17 25 159 71 Core slumps and collapses

, 4.2E-4/

1.01 42.5-97.9 0.0 0.16 24 107 42 0.04/0.53/ 0.16 24 185 85 Reactor vessel melts I,

0.69 I

i 97.9 52.3 0.13 19 122 50 4.2E-4/ 0.13 19 181 86 Reactor vessel falls

'; 2.8/0.64

? 97.9-97.95 4.2E-4 0.13 l') 124 51 0.11/ 0.13 19 197 86 Bolloff of 18 0 2

4.2E-4

0 0.64

'l 5 3

) (a) Norreltzed to a cos.'*Innent free volume of 4.199 x 10 ft . Leaka9e, in volume fraction /hr. Is from Compartment I to Compartment 2.

l (b) Normallied to a compas' ment free volume of 5.380 x 10 gg3. Leakages are respectively. Compartment 2 to Compartment 1. Compartment 2 to 5

i Compartment 3. and Comp **tment 2 to the environment.

f

• See footnotes on Page 6 fo.' V(2W).

J i

2 m --

I.

j ,. _ . . g . . -- . . ( - . ., . . .

.  ; p .,, ,

.S tt 3

.!j 1( ,e G

.$ h ,

I-TA8tE 6.5 SIMtAAY OF CONTAINNENT LEAK AATEs n.t l *******'"C'"*""***1 Csis Csas 1 ***** 8" C8"*nment 2

St.n. Co.. St.n. es.. i.!L"ai. 2"e> > *~~-- '- 2"e> > *~~- '-

i* Subseeuence ela sla_.E in ela ela v/hr wa asia *F *( v/hr we asia *F *C Samarts I

AS-6(4V)(Continued) 97.% 79.6 0.13 19 129 54 4.2E-4/ 0.13 19 187 86 Initial ccacrete attack

• 3.8/0.61 ,

1 97.95-108.1 6.9 0.12 18 174 79 0.32/0.17/ 0.12 19 178 81 Concrete decomposition 0.90

! 108.1 173.4 0.12 18 225 107 4.2E-4/ 0.12 18 175 79 Hydro 9en burns 9.1/0.88 108.1-129.1 9.6 0.11 16 409 209 0.96/0.09 0.11 16 *175 79 Concrete decomposition

) 0.74 cn 129.1 89.5 0.11- ,16 , 493 256 5.88/151.1/ 0... 16 189 87 Hydrogen burns b 0.62 N 129.1-145.4 10.5 0.11 16 508 264 4.2E-4/ 0.11 16 219 104 Concrete decomposition 0.14/0.56  :

• 145.4 76.9 0.10 15 518 270 6.07/155.4/ 0.10 15 197 92 Hydro 9en burns

0.25

.} 145.4-226.9 8.7 0.10 15 528 276 2.4E-3/ 0.10 15 188 87 Concrete decomposition l ,-( s -

0.43/0.13 ps 226.9 56.5 0.10 530 277 15 1.61/ 0.10 172 78 Hydro 9en burns jj 4.2E-4/

15 O.26 4-226.9-698.3 5.2 0.10 15 502 261 0.05/0.16/ 0.10 15 161 72 ~Cor. crete decomposition 0.17 l

4 j (a) Normalized to a containment free volume of 4.199 x 104 ft.3Leekaga, in volur.e fraction /hr. is from Compartment I to Compartment 2.

-j (b) Normalized to a compartment free volume of 5.380 x 105ft 3 . Leakages are respectively. Compartment 2 to Compartment 1. Coms artment 2 to g Compartment 3. and Compartment 2 to the environment.

~

O

'fd -

1'

,d .

I g.) *

~3  ;. g .,z.., .

% a

• a. -.--.--------*2..... -- .hC- -

• I ~= -

' 'I l

fd, g .

4w

, .h i

!b i[

~

fi TA8tE 6.5 StaclARY OF CONTAINMENT LEAK RATES

L fi _-- .

2 ,,,, ,,,, ie.t..e in Co r m .t 3 ie.t. in C n.nment 4 -

4 St.n. ela E . St.a. Eu. i.2::.i. 2"e> +- '- ap -- '-

3 g Subseeuence ela sin ela ela v/hr we osia *F *C w/hr wa esta *F *C Renarts t

j A8-s(4V) (Contiwed) 7.1-24.6 0.89/0.11 0.26 38 250 121 5.8 0.26 38 263 128 Core uncovers and heats y 24 5-41.5 0.43/0.46 0.18 26 227 108 2.2 0.18 26 253 123 Core melts

.f$ 41.5-42.5 2.67/ 0.17 25 230 110 30.3 0.17 25 521 272 *Core slumps and collapses 4

4.2E-4 ld 42.5-97.9 0.86/0.40 0.16 24 232 111 4.6 0.15 24 253 123 Reactor vessel melts

) 97.9 4.2E-4/ 0.13 19 221 105 4.2E-4 0.13 19 .215 102 Reactor vessel falls

'Y :

\.

0.46 QL 97.9-97.95 0.51/0.20 0.13 19 - 221 105 0.35 0.13 19 215 102 Bolloff of H 2O 97.95 4.2E-4/ 0.13 19 221 105 4.2E-4 0.13 19 215 102 Initial concrete attack m

• s'

[ 0.54 4

97.95-108.1 0.46/0.13 0.12 18 219 104 0.23 0.12 18 212 100 Concrete decomposition CD i 108.1 2.19/1.2 0.12 18 217 103 1.09 0.12 18 210 99 Hydrogen burns

]1 108.1-129.1 0.39/0.17 0.11 16 213 101 0.32 0.11 16 .205 96 Concrete decomposition 129.1 4 ?E-4/ 0.11 16 213 101 4.2E-4 0.11 16 202 95 Hydrogen burns

.p 15.0

[ 129.1-145.4 1.71/0.13 0.11 '16 299 149 1.18 0.11 16 199 93 Concrete decomposition

'{ 145.4 2.21/15.04 0.10 15 288 142 1.67 0.10 15 196 91 Hydrogen burns

. 145.4-226.9 0.611/0.18 0.10 302 150 15 0.78 0.10 15 189 87 Concrete de m osition

' 'I

• 226.9 12.9/ 0.10 15 298 148 155.4 0.10 15 236 113 Hydrogen hier s 4.2E-4

[' 226.9-698.3 0.15/0.13 0.10 15 289 143 0.37 0.10 15 (c) Normaltred to a compartment free volume of 9.136 x 105ft 3. Leakages are respectively. Comparta.ent 3 to Compartment 2 and Compartment 3 to 182 83 Concrete decomposition

( Compartment 4.

f 5 3 (d) Normalized to a compartment free volume of 1.731 x 10 gg . Leakage is from Compartment 4 to Compartment 3.

}m

j. . -

~

N

y -

. _.a .

- _ a ._ _.c. -

,_; .- ~ _

.=_ a_ _ _ . . . .. .. .

' (.kf '

'111

c

• I l ,

i. a

l TABLE 6.6. OIENSIONS OF PWt USED FOR CALCULATIONS g.

j.. _

l 9 Containment Volume Wall Area Floor Area i

1 Design Compartaent 7t3 m3 ~~Tf2 ,2 - Tf2 m2-

.$ Large, high pressure Containment 1.80x10 6 5.097x10 4

2.36x10 5

2.19x10 4 4 3 1

1.374x10 1.277x10 General, for V sequence 8 4 5 4

] containment 1.80x10 5.097x10 2.36x10 2.19x10 1.347x10 4

1.277x10 3

[d Aux Building 1.50x10 5

4.248x10 3

5.25x10 3

4.88x10 2

1.875x10 3

1.742x10 2

-i

{d 1.1 1 ::

\

• En TABLE 6.7. PWR CONTAINMENT SPRAY PARAMETERS *

{i 1

Flow Rate Height Temperature Oroplet Diameter.

ij Pumps lb/ min kg/s ft a F C pm t 1

! 4 2

. t Injection 2.60x10 1.966x10 90 27.4 120 48.89 490 j Recirculation 5.80x10 4

4.385x10 2

90 27.4 120 48.89 400

,p -

)

y .

J i

.) -

W '

$1 h

ai

?!

$ ~ ~ ~ ~ ~ -_ _

I e e t

i

~;

i 6-90 t

,i

'j TABLE 6.8. INVENTURIES OF MDIONUCLIDES AND STRUCTURAL MATERIklS FOR SURRY l j Fission Products Actinides / Structural *

.. Element Mass (kg) Element Mass (kg)

>]

,l Kr 13.4 U 70,210 Rb 14.7 Pu 469 Sr 47.6

Y 22.9 Cr 8,130 Zr 179 Mn 159

,l Mo 155 Fe 51,880 ,

I Tc 37.1 Ni 4,517 i-

. Ru 104 Zr 16,460 ,

[ .~ Rh 20.9 Sn 262

.n-t Pd 52.5 Ag 2,750 Te 25.4 In 505 t s

!' I 12.4 Cd 173 l

Xe '

260 '

., . Cs 1 31 l.

i-,

Ba 61.2 l1 La 62.3 l L ,1,..- Ce 1 31' -

'i j% Pr 50.7 .

l 1 ~

i Nd 171

1. Sa 34.0

,l '

>y..t e

=.

u I

.p i k

l l4-P l

) * '

i.l.y g.,_ . .

p .. s._a +vu-9 m._ m w w u n u::.u + a> w ++ . - - . : = =~c *u

~ -G--

hH+?MiYM 1 1

r 6 - 91 i

TABLE 6.9.

{ ' GEOMETRIC AND POWER PEAKING FACT 0kS FOR CORE CONFIGURATION OF THE S'JRRY PLANT Radial Power Axial Power Peaking Factor Fraction of Core Peaking Factor in Radial Zone 1.03 0.09 0.047 1.055 0.26 0.062 1.04 0.43 0.083 1.04 0.59 0.062 1.055 0.74

. O.062 1.05 0.89 0.062 1.03 .

1.02

~ 0.083 1.04 1.14 0.124 0.95 1.24 0.166 0.71 1.32

. 0.249 1.39 1.43 1.44 r-1.45 l- .

1.43 1.41 1.35 1.27 1.18 .

1.07 0.94 0.80 l 0.63 0.49 L

,';g3,"-  ; y - = 2 7 '. - ~~~ M : 7

~

' :.h:P.'A,

~

1 1

4

6-92

. i l TABLE 6.10. MATERIAL PRESENT IN MELT AT TIME OF RPV FAILURE FOR THE SURRY. PLANT a

AB-6 TM.B-6 S2 0-c S2 0-y V ,

Melt Melt Malt Melt Melt Content Content Content Content Content i Species (kg) (kO) (kg) (kg) (kg)  !

I.

Cs 0.0 0.7 0.0 0.0 0.0 j

. I 0.0 0.1 0.0 0.0 0.0 l Xe 0.0 1.5 0.0 0.0 0.0 Kr 0.0 0.1 0.0 0.0 0.0 Te 16.4 16.4 5.9 0.61 14.0  :

Ba 38.6 49.1 38.2 49.1. 31 .0

.,; , Sn 38.0 152 44.0 123 20.0

-} Ru 102 103 102 .103 101 U0 2

79594 79630 79591 79624 75559 Zr 9981 6690 6367 1783 9846 Zr0 2 8760 13210 13640 19830 8912 Fe 40520 34140 42110 34050 51880 l

Mo 125 140 125 142 117 l

Sr 39.8 43.7 ~ 39.7 44.1 36.5 Ag 1410 1460 1410 1400 1400 Cd 36.0 38.0 36.0 36.d 36.0 In 431 433 431 430 430

.[ Sb 0.11 0.31 0.11 0.60 0.05

.x.: -

'-~

'0 1 :. .

Yj

e I'

l.i l-o f

/ .

E - ,

$ r .. ., e i ,. .:.. , e ' ' ~ ': J-~ L~ ~ ~ ~ ~ '~ ~ ' '

I i

5-93 e

t i

j TABLE 6.11. MELT CONTENT FOR SPECIES NOT SPECIFICALLY INCLUDED 1

.: j IN CORSOR CALCULATIONS FOR THE SURRY PLANT a

l AB-6 TMLB-6 S2 0-c 5 0-y V Melt Melt Melt Melt Melt Content Content Content Content Content Reference Species (kg) (kg) (kg) (kg) (kg) Species Rb 0.0 0.1 0.0 0.0 0.0 Cs j Y 22.9 22.9 22.9 22.9 22.9 UO

'f Tc 36.4 36.7 36.4 2 -

36.7 36.0 Ru

. Rh 20.5 20.7 20.5 20.7 20.3 Ru Pd 51 5 52.0 , 51.5 52.0

• 51.0 Ru La 62.3 62.3 62.3 62.3 62.2 UO 2

Ca 131 1 31 1 31 1 31 1 31 U0 2

Pr 50.7 50.7 50.7 50.7 50.6 U0 2

) Nd 1 71 1 71 1 71 1 71 171 UO 2

Sm 34.0 34.0 34.0 34.0 34.0 UO 2

Pu 469 469 469 469 468 00 2

1 Cr 6351 5411 6734 5411 8130 (a)

Mn 153 ~ *157 153 157 151 fe Ni 3528 3006 3741 3006 4517 (a)

(a) These values are taken from the MARCH code predictions.

O bi W

O

'L ;f?-f, l.. .'

OU tu

.-,'_ D g I  ! O" N %4g '

b N 'Y'

~ ' ' ' l; :\ f

^

%:WhhwMW!ff:; '" ~ +' " ~

il TABLE 6.12. AEROSOL COMPOSITION AND TOTAL RELEASE RATE FOR p SURRY A8 EARLY, CORE-CONCRETE INTERACTION I 5,.cl , u . s.c -

q s o neO 2400 3600 4800 6000 7200 8400 9600 10800 i2000 iu00 1t Fe0 --

12.1 9.93 7.56 9.67 14.0 18.9 1.01 0.90 0.90 0.82 0.72 '

y

• erp3 -- -- -- -- - -- --

4 x 10-4 1 x 10 -3 1 x 10-3 2 x 10-3 2 x 10 -3 31 4 2.43 0.44 1.89 4.76 2.46 0.26 0.10 0.25 0.20 0.21 0.18 0.16

-6 f flo 3 x 10-5 2 x 10'I 2 x 10 1 x 10-5 4 x 10 8 x 10 2 x 104 4 x 10

-5 3 x 10 -5 4 x 10 -5 4 x 10-5 5 x 10 -5

'I Su 3 a 10-5 2 x 10-6 2 x 10 -5. I x 10'4 3 x 10-5 7 x 10'I 1 x 10'I 2 x 10*I 1 x 10'I 1 x 10'I 1 x 10'I 9 x 10 0 I so 0.10 0.06 0.14 0.24 0.16 0.04 . 0.02 0.18 0.17 0.18 0.17 0.17 56 3 x 10-5 1 x 10-5 2x10 -5 3 x 10-5 2 x 10-5 8 x 10 '6 x 10-6 2 x 10-5 2 x 10-5 2 x 10-5 2 x 10 -5 2 x 10 -5 Ta 0.95 0.49 0.62 0.66 0.60 0.35 0.30 1.04 1.05 1.05 1.07 1.09 At 24.0 21.3 19.1 14.4 18.5 14.8 8.61 25.0 22.8 23.4 22.2 20.9

iti 8.89 2.94 5.96 7.32 6.14 1.78 1.11 3.37 3.14 3.20 3.09 2.96

$ Ca0 --

12.3 10.6 8.87 10.7 14.3 12.4 0.55 0.52 0.46 0.44 0.49 f A3 0 23 6 a 10-4 4.50 6.79 5.46 1 x 14-3 2 x 10'3 8 a 10 ~3 9 x 10'3 'O.01 0.01 0.01

! Ita 0 --

7.22 6.05 4.63 5.90 8.33 11.0 5.34 5.46 5.48 2 5.49 5.41 m

K0 --

12.6 10.6 8.04 10.3 14.8 19.1 55.5 59.9 2 59.3 61.7 64.3 'g

, 510 -- 26.0 21.4 16.5 20.87 30.1 27.9 6.98 2 5.26 5.24 4.22 3.23

!8. UO 2 3.24 0.35 1.43 3.95 1.67 0.13 0.06 0.63 0.53 0.54 0.51 0.48 II 2r0 0.04 6 x 10'3 0.02 0.08 0.03 6 x 10'3 8 x 10 -3 0.03 0.03 2 0.03 0.03 0.03 i Cs 0 2

Sa0 3.46 1.30 1.03 0.90 0.79 0.44 0.20 0.03 0.02 0.02

0.02 0.02 j, Sr0 8.63 2.05 2.14 2.20 1.59 0.52 2 x 10'3 0'.19 2 x 10'3 2 x 10'3 2 x 10-3 2 x 10'3

) La 0 23 3.81 2 x 10 -4 2.04 6.62 2.52 1 x 10 1 x 10 -4 4 x 10-4 5 x 10~4 4 x 10 5 x 10'4 5 x 10'4 Ce0 7.38 0.70 2.55 6.45 2.60 0.15 0.02 6 x 10'4 9 x 10'4 6 x 10'4 7 x 10 -4

'E 2 7 x 10'4 iib25 0 4 x 10 2 x 10-5 2 x 10'4 1 x 10'3 4 x 10'4 1 x 10 -5 2 x 10 -6 7 x 10 -6 7 , yg -6 7 x 10 -6 7 x 10 -6 7 x 10 -6 2

i_ Csl -- -- -- -- -- -- -- -- -- -- - --

4 Cd 36.9 -- -- -- -- -- -- -- -- -- -- --

i Source

, , Rate 170 97.1 33.1 123 230 183 98.3 17.5 10.6 11.2 12.8 10.3 j 9/sec

I i.t

!h U -

3e4 a

.'l .

s- .=

~

,..- T-- . , . :.L .J.

- [} p.._ _- .

~,,\

II -

g TABLE 6.12. (Continued)

(J h

r Species.

.j

~

s 14400 15600 16800 18000 Time, sec 19200 20400 21600 22800

.3

!5 Fe0 0.64 0.54 0.40 0.32 0.29 0.30 0.31 0.32 Cr23 0 3 x 10'3 4 x 10'3 5 x 10~3 6 x 10'3 7 x 10'3 7 x 10~3 7 x 10-3 7 x 10~3 Ni 0.14 0.12 0.09 0.07 0.07 0.07 0.07 0.07

a I h 7 x 10-5 9 x 10-5 1 x 10'4 2 x 10'4 6 a 10 -4 1 x 10'3 1 x 10'3 -1 x 10'3

] Au -- -- -- -- -- -- -- -- '

l Sa 0.18 0.18 - 0.17 0.18 0.21 0.20 0.25 0.26 k Sb 2 x 10-5 2 x 10-5 2 x 10

-5 2 x 10-5 2 x 10-5 2 x IL -5 3 x 10-5 3 x 10-5 Te 1.10 1.12 1.14

] Ag 20.0 18.5 1.16 1.19 1.20 1.20 1.20 15.7 14.4 14.1 14.4 14.5 14.4 2 2.88 2.72 2.43 2.32 2.36 2.46 2.50 2.50

! ,* Ca0 0.50 0.49 0.45 0.46 0.51 0.56 0.58 0.58

.j ni,03 0.ci 0.02 0.02 0.02 0.02 0.02 0.02 0.02 Na 0 5.33 5.22 5.00 f 2 4.71 4.44 4.31 4.26 4.24 j K02 66.0 68.7 73.0 75.1 75.6 75.2 75.0 75.1

.ien

$10 2.58 1.88 1.12 0.75 0.59 2 0.55 0.54 0.53  ;

.j UO 2 0.48 0.46 0.42 0.44 0.51 0.58 0.60 0.60 Zr0 0.03 0.04 0.04 0.04 0.05 0.05 2 0.04 0.04 Cs 0 -- -- -- -- -- -- --

2 --

Ba0 0.02 0.03 0.03 0.03 0.04 0.04 0.04 0.04 SiO 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10~3 '2 x 10'3 2 x 10'3 2 x 10'3 2 x 10'3 "i

sa2 0, 5 x 10-4 5 x 10 6 x 10'4 6 x 70'4 7x,10-4 7 x 10 ~4 7 x 10'4 6 x 10'4 i Cc0 2 7 x 104 8 x 10'4 9 x 10'4 9 x 10'4 1 x 10 ~3 1 x 10 ~3 I x 10'3 9 x 10'4 Nb 25 a x 10 4 8 x 10 9 x 10 4 9 x 10 1 x 100 1 x 10 1 x 10 4 I x 10

CsI -- -- -- -- -- -- -- --

A Cd --

.f Source e Rate 9.1 8.5 7.1 5.5 3.8 3.0 3.0 3.0 i

, 9/sec a

C .-

l t i

Lh .  !:

. .\ .

l .I . ,

l if

.I C _ __ _ _. .- 'r-

~ ~ - ~ --~ --'

--~ ~ " - ' ' - ~ ' ~ ~ ~

j_;;; h . m , '

. & .t-? ?

' - '*Y -_ ~ .._ . .... -,.. ~ .. ....... _ - . -

4

i i TABLE 6.13. AEROSOL COWOSITION AND TOTAL RELEASE RATE FOR

.o

.) SURRY AB LATE, CORE-CONCRETE INTERACTION h

. f, Species Time. sec

.; I 0 1200 2400 3600 4800 6000 7200 8400 9600 10000 12000 13200 J fe0 --

13.6 12.0 )0.7 9.20 10.5 12.2 13.2 14.2 15.5 22.1 1.31

-y ,

Crg3D -- -- -- -- -- -- -- --

0.12 n Ni 2.43 0.22 0.47 1.20 2.60 1.55 0.52 0.38 0.26 0.20 0.21 0.39 i Me 3 x 10'0 6 x 10 2 x 10'I I x 10-6 5 x 10 2 x 10-6 3 x 10'I 2 x 10'I 8 x 10 5 x 10 -8 4 x 10 8 x 10-5 au 3 x 10-5 5 x 10'I 2 x 10-0 1 x if f 4 x 10-5 1 x 10-5 2 x 10 ) x 10-6 7 x 10'I 5 x 10'I 4 x 10'I 6 x 10'I f Sn 0.19 0.04 0.06 0.11 . 0.17 0.12 0.06 0.05 0.04 0.03 0.04 0.23

) Sb 3 x 10'S 1 x 10-5 1 x 10-5 2 x 10-5 3 x 10-5 2 x 10-5 1 x 10-5 1 x 10 -5 8 x 10 8 x 10-6 9 x 10-6 2 x 10-5 d Te 0.95 0.44 0.48 0.58 0.65 0 58 0.43 0.39 G.35 0.34 0.42 0.98

's Ag 24.0 14.2 22.3 20.6 17.7 20.2 22.8 18.4 14.4 12.6 14.2 30.0 k Mn 8.89 2.14 3.04 4.85 6.90 5.21 2.81 2.36 1.89 1.68 1.93 4.17 '

i0 Ca0 --

13.80 . 12.J 11.2 10.1 11.2 12.6 13.6 14.5 15.3 3.83 0.70

', 6xM 4 7rM ,

Al230 1 x M 1 x d 1 x r.

Na 0 8.09 3.04 5.92 3.73 IxM 2 xM 6 m d i 2

7.20 6.50 5.62 6.38 7.32 7.89 8.38 9.09 12.5 4.80

,,) K0 --

14.2 12.8 11.4 9.79 11.20 13.0 13.9 14.7 15.6 cn 2 20.2 44.1 e fj $10 2

29.4 25.8 23.1 19.9 22.6 26.2 28.4 30.5 29.0 24.3 12.2 00 3.24 0.20 0.34 0.87 1.98 2 1.02 0.28 0.18 0.11 0.09 0.09 0.97 i Zr$

2 0.04 0.007 0.006 0.015 0.G35 0.018 0.006- 0.006 0.006 0.006 0.009 0.02

'lp Cs 0 2

0 0 -- -- -- -- -- -- -- -- -- --

1; 840 3.46 1.36 1.03 1.01 0.97 0.84 0.57 0.40 0.29 0.19 0.05 0.03 i Sr0 8.63 1.86 1.65 1.96 2.14 1.62 ' O.85 0.55 0.36 0.23 0.05 0.03 La 0 3.8! 2 x 10'4 2 x 10'4 L15 i

23 3.01 1.44 1 x 10'4 1 x M'4 l x 10'4 1 x 10 ~4 1 x M'4 3 x 10-4 Ce0 7.38 0.36 0.63 1.59 3.39 1.71 0.40 0.20 0.10 f 0.05 0.009 5 x 10'4 Nb 0 25 4 x 10'4 3 x 10-6 2 x 10-5 1 x 10-4 4 x 10'4 2 x 10-4 3 x 10 -5 2 x 10-5 1 x 10 -5 2 x 10 2 x 10

' 5 x 10

Csl -- -- -- -- -- -- -- -- -- -- -- --

Cd 36.9 -- -- -- -- -- -- -- -- -- -- --

[ Source Rate 158 80.5 205 214 73 136 119 i 84 73.1 62.2 45.3 17.6 (above

{ Pool)

{ 9/sec k

i '

!j  ! .

j .-

~

L -

7 .c- g .

. . , . , , . -- ~-

n.. . ._ .

i'k il .

I

' i .i, !,

{

q y .

TABLE 6.13. (Continued)

~

'q i j' i Species Time, see PI .',

5 14400 15600 16800 18000 19200 20400 21600 22800 24000

'y .

m 1.22 1.07 0.80 Fe0 0.91 0.69 0.59 9.43 0.36 0.33 ,

Cr23 0 6xM ixM d 2xd 2xM 3xM d 4xM O 5xM 5xM 6xM

D Ni 0.33 0.27 0.22 0.18 0.16 0.13 0.09 0.08 0.07 l

f

-5 ho 7 a 10-5 5 x 10 6 x 10-5 7 x 10-5 8 x 10-5 1 x 10'4 1 x 10-4 2 x 10'4 3 x 10'4 S

• Ru -

4 x 10'I 2 x 10'I 2 x 10'I 1 x 10'I 8 x.10-8 ,, __ .. ,,

jf Sn 0.'22 0.20 0.20 0.19 0.19 0.19 0.17 0.18 0.20 y Sb 2 x 10-5 2 x 10'0 2 x 10-5 2 x 10-5 2 x 10-5 2 x 10-5 2 x 10 -5 2 x 10-6 2 x 10 -5

,J To 1.01 1.06 1.09 1.11 1.13 1.15 1.17 1.19 1.21 [

28.3 26.0 Ag 23.8 22.2 20.7 M.3 15.9 14.9 14.7 jl ['d Mn 4.01 3.78 3.55 3.38 3.22 3.06 2.66 2.58 2.60

? Ca0 0.60 0.52 0.52 0.54 0.53 0.52 0.46 0.47 0.50

] A1 0 23 7xM 8xM 0.01 0.01 0.M O.M 0.M8 0.02 0.02 Na 0 2

5.06 5. 4 5.38 5.36 5.33 5.24 5.05 4.82 4.64 m i K0 2 48.6 54.1 58.8 61.8 64.5 67.2 72.4 73.9 74.4 to

$10 9.71 7.03 4.98 3.78 2.89 2.14 1.21 0.89 0.74 2

U0 2

0.82 0.67 0.58 0.54 0.50 0.48 0.42 0.44 0.48 Zr'l 2 0.02

.. l 0.03 0.03 0.03 0.03 0.03 0.04 0.04 0.04 Cs 0 -- -- -- -- -- -- -- -- --

! 840 2 x 10-3 0.02 0.02 0.025 0.026 d.027 0.03 0.03 0.03

. Sr0 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10-3 2 x 10'3 2 x 1G'3 0.002 0.002 2 x 10 -3 j La23 0 4 x 10-4 4 x 10'4 4 x 10'4 5 x 10'4 5 u 10-4 5 x 10-4 6 x 10-4 6 x 10'4 7 x 10 -4 i Ce0 2

5 x 10 6 x 10'4 6 x 10 -4 7 x 10-4 7 x 10'4 7 x l0'4 9 x 10 -4 9 x 10 -4 9 x 10-4

Nb 0 2$ 5 x 10'0 6 x 10-6 7 x 10-6 7 x 10 7 x 10-6 8 x 10 -6 9 x 10 -4 9 x 10 1 x 10

-5 Csl -- -- -- -- -- -- -- -- --

'Cd 95.7 37.8 29.1 27.9 -- -- -- -- --

Source i Rate 18.4 17.8 13.2 10.3 S.9 8.5 6.3 4.3 3.6

! g/sec

^

s. .

<M a .

.] .

a.. . .

f 1 6-98 i

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u o= = o u o M N .

= = - .e. = o o : oo = o =

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=a

3. * * * ~*

• t e = = = = =

~

4 en e u en --

tt.

T'? 7 7e 7TT?? -

I g g g e o.g e ~ a

. me = =- = -

e o

o. 4 .g g

.o .

n

2. *4 4* 3.

• ol 3.

==~-=a = =* =* =* =* *- l en,--

4 TT  ?  ? 7TTT?

g ~N gee~e I, . me = .- =-,4**-

- -

• g o ,~ o *

• 4"4 2. * ! 3. * ~* ~* ~* * ! 4*

.e =n --o o = = = = = -

em o n .nw en T. . w.

• *** n. n. e. .ww . . .

I g "N

. ee 2. =*.-* S-nm

=-

• 4,5.e

• n

==w

2. 4 4.* 3. o

o3.= *= *= *= *=

• one~ u 4

e e m m m e

= e= m.

g

~asamasara3salo- s ~ c 3 .g - N M-

)M o sasab

'4-[$ - , 4,tB '

1. • .'5 g . . .y, '

h M

_---g

{

e W F g --Be ,g -

.. _ . , _ a ..,, ,

h

-._.._m.

,ji'*

. . = .

1: _j

._.,-)..,L' us

a. - _. . .. .

_h(, .

j6 v6 .g., , ,j; t

q .

4

@'i -

TABLE 6.15. AEROSOL COMPOSITION AND TOTAL RELEASE RATE 'FOR SURRY S D, CORE-CONCRETE INTERACTION lt l

I~

Species, Tlee, sec

? 5 0 1200 2400 3600 4800 6000 7200 8400 9600 10000 12000 13200 i

. Fe0 -- 13.7 12.1 11.3 16.6 19.9 20.0 20.1 20.7 22.1 0.96 .

1.10 h1 Cr 0 23 0.02 '0.01 -- -- -- -- -- -- 2 x10 4 4 x 10 48 8 x 10'4 0.04 0.11 0.12 0.09 0.04 0.04 0.05 0.06 0.00 0.31 0.39

-8

.i Me -- --

2 x 10'I 2 x 10 1 a 10'O -- -- -- -- --

4 x 10 6 x 10 4

h Su -- 3x10-0 2 x 10'I 2 x 10'I 1 x 10'I 3 x 10'8 3 x 10'O 4 x 10'I 5 x 10'8 7 x 10 ~8 3 x 10'I 5 x 10'I En 3 x 10'3 0.01 0.02 0.02 0.02 0.01 0.01 0.02 0.02 0.02 0.22 0.26 I Sh 6 x 10'I 2 x 10~0 ( x 10 4 x 10-6 4 x 10 3 x 10 3 x 10 3 x 10 3 x 10 4 3 x 10 4 2 x 10-5 2 x 10-5 y Te 0.05 0.11 0.13 0.13 0.15 *0.13 0.13 0.14 0.14 0.16 0.54 0.50

,g Ag 0.46 4.66 3.98 8.99 8.41 5.69 5.68 6.41 7.17 4.23 31.3 34.0

'2 Its 0.12 0.87 1.54 1.52 1.49 1.07 1.06 1.18 l.30 1.48 '5.54 5.83 l Ca0 --

0.47 2.56 5.30 8.58 9.?2 9.96 10.2 9.35 6.72 0.36 0.45 5 Al23 0 -

3xM 1xM d 2xM SxM 4 9xd IxId IaM 4 1xM 2xM 4 6xM 5xM 4 7

} Na 0 2

-- 7.70 .6.91 6.50 9.64 11.6 11.6 11.7 12.0 12.8 4.86 4.56 g j; K0 2

-- 12.1 11.2 10.8 16.5 20.3 20.6 20.4 20.? 21.6 47.4 42.1 0 510 --

15.4 17.7 21.8 29.8 30.1 29.5 28.7 27.6 26.2 7.54 9.63 2

U0 0.03 0.08 0.14 0.12 0.11 0.08 0.06 0.06 0.06 0.06 0.91 1.12 9 x 10 -3 g; 2r0 0.02 0.01 8 x 10~3 0.01 0.01 0.01 0.01 0.01 0.01 0.04 0.03 2

Cs 0 -- -- -- -- -- -- -- -- -- -- --

2

<: Ba0 0.22 1.16 1.36 1.10 1.22 0.73 0.52 0.42 0.33 0.21 0.03 0.03 Sr0 0.69 1.60 1.95 1.61 1.56 1.03 0.72 0.57 0.43 0.26 4 x 10'3 4 x 10'3 2 x 10 -4 8 x 10 4

~4

,[ La 0 23 4 x 10 2 x 10'4 2 x 10'4 2 x 10'4 2 x 10'4 2 x 10 -4 2 x 10'4 2 x 10'4 2 x 10~4 8 x 10'4 I Ce0 6 x 10'4 0.07 0.19 0.18 0.11 0.04 0.03 0.03 0.02 0.02 8 x 10'4 8 x 10'4 Hb 0 25 6 x 10-6 4 x 10-6 3 x 10 3 x 10-6 4 x 10 3 x 10 3 x 10 3 x 10 3 x 10 -6 3 x 10 8 x 10 -6 8 x 10'4 Csl -- -- -- -- -- -- -- -- -- -- -- --

Cd 98.3 42.0 35.0 30.4 5.68 -- -- -- -- -- -- --

• Source .

Sate 0.8 8.1 20.0 32.2 51.4 59.6 45.0 40.1 -38.2 38.4 8 .11 5.3

.e (after li Pool)

.. 9/sec .

,q .-

h 9 A - - - _ _ . . . - _ _ - _ . . _ . _ _ . _ __

~

-m . _ _

- . , . v. . . . .. ,. . , ,

i .

p%n t TABLE 6.15. (Continued)

.,y

.8

<I* Species, Time, sec

, 1 14400 15600 16800 18000 19200 20400 21600 22800 24000 25200 26400 27600

?

u .r .

i , .. Fe0 1.24 1.37 1.47 1.55 1.62 1.67 1,71 1.72 1.71 1.65 1.56 1.44

'5 Cr23 0 , 0.20 0.26 0.31 0.36 0.39 0.42 0.43 0.0 0.38 0.33 0.28 0.23

. Ni 0.48 0.57 0.66 0.72 0.77 0.80 0.80 0.76 . 0.70 0.63 0.55 0.47 i ?,

^

Me 1 x 10'4 2 x 10'4 2 x 10'4 3 x 10'4 4 x 10-4 4 x 10'4 5 x 10'4 4 x 10~4 4 x M'4 3 x 10-4 3 x 10 ~4 3 x 10 lha 8 x 10*I 1 x 10'4 2 x 10-6 2 x10 -6 3, jp -6 3 x 10'4 3 x 10-6 3 x 10-6 2 x 10-6 2 x 10'4 1 x 10-6 8 x 10'I i Sn 0.29 0.32 0.35 0.37 0.38 0.39 0.40 0.40 0.40 0.39 0.38 0.37 lI Sb 2 x 10 5 2 x 10 -5 2 a 10-5 2 x 10-5 2 x 10 -5 2 x10 -5 2 x 10-5 2 x 10-5 2 x 10-5 2 x 10'4 2 x 10-5 2 x 10-5 Te 0.46 0.41 0.38 0.34 0.32 0.31 0.30 0.32 0.34 0.36 0.39 0.42 Ag 36.1 37.8 38.6 39.0 39.0 38.9 38.8 38.7 38.4 37.7 36.6 35.2 h 6.00 6.07 6.06 5.98 5.80 5.70 5.66 5.80 5.82 5.82 5.77 5.65

. Ca0 0.59 0.67 0.71 0.74 0.77 5.79 0.81 0.82 0.82 0.82 0.81 0.19

~3 A1 0 23 4 x 10'3 4 x 10'3 3 x 10-3 3 x 10 2 x 10'3 2 x 10'3 2 x 10'3 3 x 10'3 3 x 10~3 4 x 10'3 5 x 10'3 6 x 10~3 sia 0 4.17 3.80 3.48 3.23 3.06 2.96 2.96 3.13 3.40 3.72 4.03 4.31 2

K02 36.5 31.4 27.5 24.6 22.7 21.6 21.6 23.4 26.3 30.3 34.5 38.9 5

510 2

12.4 15.5 18.4 20.9 22.8 23.9 24.0 22.2 19.6 16.4 13.4 10.7 p UD 1.40 1.70 - 1.96 2.18 2.33 2.42 2.42 2.27 2.06 1.81 1.58 1.38 2 o" 0.02 0.0i 0.0i d zr0, 0.02 0.02 0.0i 0.0i 0.0i 0.0i 0.Di 0.02 0.e f Csg0 -- -- -- -- -- -- -- -- -- -- -- --

Ba0 0.04 0.04 0.04 0.04 0.04 'O.03 0.03 0.03 0.03 0.03 0.A4 0.04 .

. Sr0 5 x 10-3 6 x 10'3 6 x 10~3 6 x 10'3 6 x 10'3 6 x 10-3 5 x 10'3 5 x 10-3 5 x 10'3 5 x 10 0 5 x 10'3 5 x10 -3

, i.. La 0 23 4xd 3xM 2xM 2xM 2xd 2xM 2xd 2xM 2xd 2xd 2xM 3xM

'U Ce0 6 x 10'4 5 x 10'4 5 x 10'4 5 x 10'4 5 x 10~4 5 x 53'4 4 x 10'4 4 x 10'4 4 x 10'4 4 x 10'4 4 x 10'4 5 x 10'4 5xM 4 Nb 0 2$ 4xM 4 9 xM 1xM d 1xM ixM ixM 1xM 1mM 8 x 10 4 x 10 4xM

} Csl -- -- -- -- -- -- -- -- -- -- -- --

I U 1

Source Rate. 4.9 6.2 8.4 11.2 14.2 17.0 21.9 24.6 23.8 21.0 18.0 15.2

f'  :

(above IJ 9/sec P

11 1 .

G

j. .

3

.t.

m -

, .  ?

'p-6-M; w cA-1.rured * ~- # ,'%yrp.u,. <

. - s uwi. cam.:s.u.1. cw ::tc.

W.iuf _,.- ._.

,d _

e TABLE 6.15. (Continued) 4

-, Species, u se, sec

'j 1 28800 30000 31200 32400 33600 34800 . 36000 Li g r0 1.31 1.19 1.oe 0.98 0.89 0.78 0.69 4 Cr 0 23 2 x10' 2 x 10'3 3 x 10'3 3 x 10'3 3 x 10'3 4 x 10'3 4 x 10 ~3

NI 0.41 0.36 0.32 0.28 0.26 0.22 0.19 2 3 x 10'4 Me 3 x 10'S 3 x 10'4 3 x 10'4 4 x 10" 4 x 10'4 6 x 10'4

,h 6 x 10'I 4 x 10'I 3 x 10'I 2 x 10'I 1

Ru 2 x 10'I 1 x '10'I 1 x 10'I j- Sn 0.35 0.35 0.34 0.34 0.35 0.35' O.35 Sb 2 x 10-5 2 x 10-5 2 x 10 -5 2 x 10 -5 2 x 10 -5 2 x 10 -0 2 x 10 -6 j Te 0.44 0.46 0.47 0.48 0.49 0.50 0.52

1 Ag 33.6 32.2 30.7 29.5 28.4 26.9 25.4 0 its 5.51 5.37 5.21 5.08 4.97 4.79 4.62 k) Ca0 0.77 0.75 0.74 -

0.73 0.72 0.71 0.70 Al23 0 7 x 10'3 8 x 10-3 9 x 10~3 0.01 0.01 0.01 0.01 Na 0 4.55 4.69 4.78 4.82 4.83 4.82 4.77 2

jq K02 43.3 46.6 49.8 52.2 54.3 57.1 59.6 510 8.47 6.84 5.48 4.I9 3.7I

~

U0 2

1.21 1.11 2.86 2.19 7

~

2 1.02 0.97 0.94 0.89 0.86 o

.] Zr0 0.02 0.02 0.02 0.03 0.03 0.03 0.03 2

j Cs 0 . -- -- -- -- -- --

2 --

2I R40 0.04 0.04 0.04 0.04 0.04 0.04 0.04

'. Sr0 5 x 10-3 4 x 10~3 ~ 4 a 10'3 4 x 10'3 4 x 10'3 4 x 10'3 4 x 10 -3

.( La 0 23 4 x 10-4 4 x 10'4 4 x 10-4 4 x 10'4 4 x 10'4 5 x 10 5 x 10'4

'! Ce0 5 x 10'4 5 x 10'4 6 x 10'4 5 x 10-4 6 x 10'4 7 x 10'4 7 x 13'4 hh 0 2$ 5 x 10 5 x'10 5 x 10 6 x 10 6 x 10 6 x 10 4 7 x 10 j Csl -- -- -- -- -- -- --

Cd Source

. I ;'-

Rate 12.6 10.6 9.1 7.9 7.5 7.0 6.2 (above pool) g/sec l .

l -

e

• i SE am .-

.. .m , _ . . , m .s . . . .. w. ..

g-- .

3. . .

.Y TAllLE 6.16. AEROSOL COMPOSITION AND TOTAL RELEASE RATE FOR 3& SURRY V, CORE-CONCRETE INTERACTION 4

y

, 59ecles. Time. see t 1 0 1200 2400 3600 4800 6000 7200 8400 9600 10000 12000 13200 Fe0 --

14.2 12.7 11.0 9.58 13.6 15.3 18.4 23.7 1.11 1.02 0.u6 Cr 8 -- -- -- -- -- -- -- -- --

4x1 i I x 10' 2 x 10'3

' 23 Ni 2.57 0.24 0.51 . 27 2.78 0.40 0.16 0.12 0.11 G.28 3.24 0.18 I m 3 x 10 5 x 10~8 2 x 10'I 9 x 10'I 4 x 10 1 x 10'I 2 x 10 I x 10-8 1 x 10

-8 3 x 10-5 3 x 10-5 2 x 10 -5 Ru 2 x 10-5 4 x 10'I 2 x 104 8 x 10

-0 3 i 10-5 g ,yg-6 2 x 10'I 1 x 10'I. I x 10'I 2 x 10'I 1 x 10'I 9 x 10'8 .

Sn 0.08 0.02 0.03 0.04 0.07 0.02 0.01 0.01 0.01 C.08 0.08 0.07 Sb 1 x 10-5 3 x 10 4 x 10 6 x 10 8 x 10 3 x 10-6 2 x 10 2 x 10-6 2, a 10 8 x 10 7 x 10-6 7 x 10-6

% Te 0.70 0.32 0.36 0.43 0.48 0.32 n.25 0.25 0.28 0.88 0.89 0.91 f Ag 25.5 12.1 19.3 21.2 18.4 16.2 8.96 7.89 8.04 22.6 21.0 18.u

• 7.55 1.84 2.66 4.14 5.95 2.19 . 1.28 1.16 1.21 3.49 3.31 3.03 .

Ca0 -- 14.4 13.0 11.6 10.6 13.9 15.5 13.4 4.53 O.59 0.48 0.42

{f Al230 ,

6 x 10 4 8 x 10 d 2.53 4.97 1 x 10 -3 1 x 10'3 2 x 10'3 2 x 10'3 8 x 10' 9 x 10 -3 0.01 Na 0 8.40 7.60 os 6.65 5.81 8.10 9.01 10.8 13.5 5.55 5.70 5.77 K0 2 L 14.7 13.4 11.6 10.0 14.3 15.9 18.6 22.1 2

510 30.6 gu.7 56.7 60.2 64.9 8 2

27.3 23.7 29.2 32.3 28.8 26.3 7.94 6.36 4.48 00 3.43 0.20 0.35 0.89 2.06 0.22 0.08 0.06 1}g 2 0.06 0.68 0.58 0.47 Zr0 0.04 2 x 10~3 7 x 10~3 0.01 0.03 6 x 10~3 7 x 10'3 8 x 10'3 9 x 10'3 2 0.03 0.03 0.03 J Cs 0 -- -- -- -- -- -- -- --

2 -- -- -- --

I 840 2.64 1.00 0.99 c. 74 0.71 0.49 0.28 0.18 0.06 0.02 0.02 0.02 Sr0 7.48 1,64 1,81 1.59 4.42 0.81 0.37 . 0.23 0.07 3 x 10~3 2 x 10~3 2 x 10'3 i u,0, 3.82 2 x iO -4 2 x iO 4 i.i2 2.37 i a iO 4 i a iO -4 i x iO 4 i x iO4 5 x iO4 5 m iO4 5 x io 4

$ Ce0 6.96 0.33 0.59 1.44 3.10 0.30 0.07 0.03 8 x 10'3 6 x 10'* 7 x 10'4 7 x 10'4

~0 Nb 0 2$ 4 x 10'4 1 x 10-5 3 x 10 I x 10-4 5 x 10'4 2 x 10 -5 2 x 10 2 x 10 -6 2 x 10 -6 7 x 10-6 7 x 10

-6 8 x 10 -6 Csl -- -- -- -- -- -- -- -- -- -- -- --

2 Cd 39.2 -- -- -- --

}I Source .

Rate Ib6.0 81.4 20.7 24.6 147.0 157.0 100.0 66.0 49.9

' -]g g/sec 14.4 16.0 13.3

,  ?,

a .'

ei .

• i l j -

4 '. .

^""'

- - ~ . - - . ~ . --- . , . . . . . .

y--, & & t. M . 5 P " T m ' f z; -

p . x -4 ' ~. it;, .t. ..r. .r -l ~ x '. ~ ;.i.' 1:

t .

.n 1f a

'J TABLE 6.16. (Continued) ty .

is Species, Time. sec g 5 14400 :5600 16800 10000 1D200 70400 21600 22 66 24000 25200 t

4" Foo 0.79 ' O.72 0.66 0.66 0.56 0.43 0.38 0.36~ 0.32 0.32

'i Cr 0 23 2 x 10'3 3 x 10'3 3 x 10'3 4 x 10-3 '4 x 10'3 5 x 10~3 6 x 10'3 6 x 10-3 7 x 10'3 7 x 10'3 NI 0.17 0.15 0.14 0.14 0.12 0.08 0.08 0.07 0.07 0.07 h 3 x 10 -5 3 x 10-5 3410-5 5 x 10-5 5 x 10-5 5 x 10-5 y 'a 10-5 1 x 10~4 i x 10 d S

2 x 10 Ru 7 x 10 -- -- -- -- -- -- -- -- . --

f.' Se 0.07 0.07 0.07 0.07 0.07 0.06 0.06 0.07 0.07 0.08 4 Sb 7 x 10 7 x 10 7 x 10 7 x 10 7 x 10-6 7 x 10 7 x 10 ' 7 x 10-0 7 a 10-0 H x 10 -6 Te 0.91 0.92 0.92 0.93 0.93 4

0.93 0.95 0.96 0.97 0.99

.Ag 17.9 17.0 16.3 16.9 15.3 12.9 12.4 12.2 11.6 12.4 i[p u 2.92 2.82 2.73 2.82 2.61 2.30 2.25 2.25 2.19 2.34 C. a0 0.48 0.48 0.47 0.51 0.50 C 42 0.42

.. 0.44 0.44 0.50 6 Al23 0 0.01 0. 01 0.01 0.01 0.02 0.02 0.02 0.02 0.02 0.02 J n.20 5.70 5.66 5.60 5.56 5.4e 5.28 5.n s.00 4.83 4.76 y K0 66.9 68.5 70.0 69.3 .72.1 75.9 76.9 l

2 510 2

3.69 3.11 2.60 2.60 1.95 1.24 1.01 77.3 0.88 78.3 0.72 77.2 0.72 g

f U0 2 0.45 0.43 0.41 0.44 0.40 0.35 0.36 0.37 0.38 0.44 Zr0 0.03 0.03 0.04 0.03 2 0.04 0.04 0.04 0.04 0.04 0.04 Cs20 -- -- -- -- -- -- -- -- -- --

... 840 0.02 0.02 0.02 0.02 0.02 0.02 0.02' O.02 0.02 0.0J E Sr0 2 x 10'3 2 x 10-3 2 x 10'3 2 x 10'3 2 x 10'3 2 x 10~3 2 x 10-3 2 x 10'3 2 x 10'3 2 x 10~3

![ Le23 0 5 x 10-4 5 x10 5 x 10'4 5 x 10-4 5 x 10 4 6 m lo d 6 x 10 4 6 x 10 4 7 x 10 6 x 10 4 7 x 104

~

Ce0 7 x 10'4 8 x 10 7 x 10'4

• 8 x 10'4 9 x 10 -4 9 x 10-4 2 9 x 10-4 9 x 10-4 9 x 10'4 8 x 10 -6 9 x 10 -6 Mb 0 2$ 8 x 10-6 8 x 10 7 x 10 8 x 10-6 9 x 10 -6 9 x 10-6 1 x 10-5 9 x 10-6

Cs1 -- -- -- --

i.- -- -- -- -- -- --

Cd -- -- -- -- - -- -- -- -- --

Source Rate 9.4 8.6 6.4 5.7 7.2

ll 5.8 4.1 3.6 3.2 3.3 l 9/sec j _ _ _

9

g __ ... _- . . _ . . _ __ .:. _

.]

c

} 7. RESULTS AND DISCUSSION ii 1 7.1 Intreduction

.J l

• l .Results of calculations for the transport and deposition of radionuclides are presented and discussed in this section. The plants and sequences selected for consideration were discussed in Chapter 4, l) the analytical and calculational methods were described in Chapter 5,

'. and the assumptions and bases for the calculations were described in Chapter 6. Results presented in this chapter include the' deposition and J

release from the reactor coolant system of radionuclides leaving the

l core region. These results are based on TRAP-MELT code calculations.

Also included as results are the masses of radionuclides airborne and t

deposited in the containment as well as the airborne materials leaked to E the environment, based on NAUA-4 calculations.

Four basic system sequences were considered in the ana. lyses:

, A8, TMLB', 5 0, and V. Combined with these sequences are several 2 ,-

variations in the analyses allowing for diffetcat assumptions of contain-i ment failu're,' compartmentalization of the containment, compartmentaliza-l tion of the ECC injection pipine for the V sequence, and heating of

,;, pr,imary system structures by decay of deposited radionuclides.

l, 7.2 Transport and Deposition in Primary System i

7.2.1 RCS Transport rW Deposition for Secuence AB i-lt

, The A8 sequences are characterized as low pressure, generally g high flow rate. scenarios with respect to RCS behavior. These sequences, ;l when considered as hot leg breaks, as they are in this document, offer  ;

jh the minimum volume of the RCS to the fission products being . transported 6 d to the containment atoosphere. This, of course, usually minimizes the opportunity for retention in the primary system. A variety of contain-

.L ment failure rmdes have been examined for this~ sequence in the Surry

( plant and are reported in this volume.

L i 1v . t 3

N. .~ e.. _ m._

Tz{%QG: . ' . , .-.w-

. ~- , . - .-,--, _ _ . .- _ :. _ 4:_~=~ Q

- .. . . - - ~ ~ ~~

.g -

P}

l

.; 7-2 i

j The TRAP-MELT predictions of fission product retention for tr.e 8

AB sequence are found in Tables 7.1 and 7.2. The extent of retention is seen to be different for the various species, and the cause of the differences is apparently in the R',,5 gas flow rates as a function of time. Examination of the gas flow rates leaving the core shows that for this sequence there is low flow for approximately the first 1000 s after ,

the start of core melting. Th*s low flow rate of gas through the upper plenum permits very efficient retention of Te vapor due to its rapid -

reaction with the structural surfaces. This behavior is illustrated in i Figure 7.1. At about t = 1000 s, the RCS gas flow rates increase dramati-cally as portions of the core begin to slumo into the water remaining in l 3 the lower plenum. Thus, while Te continues to be emitted by the melting l Y fuel, practically no additional Te vapor is retained due to its rapid I

,~

transit through the upper plenum control volumes. After about 2500 s, . j if the gas flow rate decreases to very low levels again becaus'e the water has been completely exhausted, and the additional release of Te is f

S i effectively scavanged.

The influence of the RCS thermal hydraulics is manifest in '

Figures 7.2 and 7.3 also. For thers specie:4, the surface reaction is not nearly so rapid, and the mechanism of their retention is condensation on the particles followed by particle settling in the lower portion of the upper plenum. This is seen to occur until the gas flow rate increases h; greatly, at which time the temperature of the upper grid plate increases .

l . from 950 F to 1340 F, resulting in the re-evaporation of the Csl and i

'?

[ Cs0H into the swiftly flowing gas stream. The surfaces in the upper plenum, and the gas temperatures quickly decrease to values which would permit condensation, but the inventory of Cs and I has been nearly .

expended by this time so that almost no retention of these vapors occurs

) in the RCS for this sequence. -

,9 The aerosol retention factor for the AB sequence is seen in

3 i Table 7.2 to be approximately 50 percent. Figure 7.4 illustrates that  ;

h about one-half of the ultimately retained aerosol mass accumulates in ,

l[

i1 the core and uppar plenum during the first 1000 s, which is characterized by low flow and high aerosol generation. As one would expect, the high I

l$ flow rate period of the in-vessel phase of this accident sequence permits I

1a .

h j

W .

I .I

, . . . . ~

' 9 Q # Y % 3 f/5/41P P R'&-

$-~4* '

" + "_ C - W W 9 .ceyvr M 4 M , , _ ,,g

,. - ,;Jj

73 .

TABLE 7.1. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE AB SEQUENCE FOR THE SURRY

-! PLANT

'1-(Times Measured from Start of Core Melting)

Cs! Cs0H Te Aerosol j '

Time Ret Total Ret Totaf Ret Total Ret Total' j (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg)

..i -

-1 300 2.9 12.4 15.9 67.7 0.3 0.4 31 9 663 t

.j 600 5.3 18.9 28.3 100 0.8 1.0 523 974

.1 900 6.9 22.7 37.0 120 1.5 1.7 682 1235 t.;

'; 1200 0.5 24.6 3.9 130 2.4 2.9 71 3 1810

.I i 1510 0.5 25.3 3.9 133 2.4 3.6 71 4 1875 t 2110 05- 25.6 3.9 135 2.4 4.4 1l i 714 1900

', . 2710 0.5 25.8 3.9 135 2.4 4.9 721 1920

) 3310 0.6 25.9 4.6 136 2.9 5.7 785 1990

,! 3910 0.7 25.9 4.9 136 4.1 6.9 911 2120 1 '

4520 0.7 26.0 5.0 136 5.4 8.2 1030 2235 i 0.7 j 5120 26.0 5.1 136 6.5 9.3 1135 2340 l .

e d

I d

t.

Q3 48=

'.4 & 4, -34*m"'

- d # - '

M*

,"*M r K --'m._..x:_ _. ,gwegh

.~

a_ . ~ -

"a

_ _ _ _ ,_, _ _ _ _ _ _ ._'_ "7 r.: 7N ,

fr. :iF3RMT .W 6 5 #S."*if 1: ~ > ' " ~ . ; 7

• TM W h ^ " ' t- . .~- _

t:

. l; j,,

T y

TJ TABLE 7.2. TRAP-MELT PREDICTIONS OF PRIMARY SYSTEM RETENTIdN FACTORS (RF) AND VOLUE

' SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIME FOR THE A8 SEQUENCE FOR THE SURRY PLANT

(

t

,i 0 Cs! Cs0H Te Aerosol 1: Time Core Core Core Core (s) RF Plate RF Plate RF Core Plate RF Core Plate 300 .23 .22 .23 .23 .82 0 .78 .48 .32 .16

60) .28 .27 .28 .27 .86 0 .83 .54 .36 .17

1 900 .30 .29 .31 .30 .93 0 .91 .55 .37 .18

! 1200 .02 --

.03 .01 .85 --

.77 .39 .25 .13 s 1510 .02 .03 i.

2 --

.01 .67 --

.61 .38 .25 .13

{ 2110 .02 -- ,03 .01 .55 --

.51 .38 .24 .12 i;

2710 .02 --

.03 .01 .49 --

.45 .38 .24 .12 j- 3310 .02 --

.03 .01 .51 .09 .39 .39 .27 .12

}i

.1 3910 .03 --

.04 .01 .60 .25 .32-

.43 .31 .11 i

j 4520 .03 --

.04 .01 .66 .36 .27 .46 '.35 .11 ilM 5120 .03 --

.04 .01 .70 .44 .24 .48 .38 .10 i

1 1

'i .

m 1

y .

m

- L ..

_  : . .. .. = -

.-_ v i} =-

y

'-I 1

sI m .

5;ii( 10 li L.s.nd

',d, WOL 8 8- v a ' + vo_' 2.

g);

9 vats i._8._

f d5 --

il v) 6 s- p

{ ,

, l 4

4- <

l e / r I

g 2- '

_.- //

i

) I I I I i

O 1000

~

2000 3000 4000 5000 -

6000 i

TIME'(sec)

I I f I FIGURE 7.1. MASSES OF Te EMITTED FROM C0kE AND RETAINED IN THE RCS CONTROL YOLUMES AS FUNCTIONS OF

=;

TIME FOR THE AB SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 = Guide Tubes, J Vol 4 = Top Plate, Vol 5 = Upper Plenum Annulus). Times measured from start of core

2 melting.

.J .

, y t .:

u

w.:; . -

- =

m . .

s._z: ~. ,. w . .+. . ~ . .

s11 -* m - , . . _ .. . . - , , -

,yy r ,. *

,s _} 4 v ..

1 a

1 t

15 0 M<

lI -- -

j p i.

4 m

/

l x", G x

,.I 10 0 -

.y 4 u)

Ti @

11 2

' r.a c

a

'a z

<( 50- ~

s -

a, 9 .d

. (E Legend i VOL 2 *'

4 EMITTED ,

/

,i 0 i , , , ,

it 0 1000 2000 3000 4000 5000 6000

!f -

TIME (sec) s

, FIGURE 7.2. MASSES OF Cs0H EMITTED FROM CORE AND RETAIEDIN THE RCS CONTROL VOLUMES AS FUNCTIONS OF

. TIME FOR THE AB SEQUENCE (Vol 1 = Core. Vol 2 = Upper Grid Plate, Vol 3 = Guide Tubes, a Vol 4 = Top Plate, Vol 5 = Upper Plenum Annulus). Times measured from start of core . . ,

! melting.

t .

t -

4 y

N . - - - . ..

.- ., m . . . . .m . m ..

m

. s. . mm. _ . .

u't .

tl 43 l

?! -

,y 30 ,

n

\A

• n

/,

ii3

a] .

a

.x /

/

lA -

20- /

(A /

, r]a

.i i

g

,n 2

e

. w

.: z j R

10 - / ~

)

g Legend j

llp

.VOL 2 j EWitTED ,

i i'ii i 0 , i i i ,

] O 1000 2000 3000 4000 I 5000 6000

.! TIME (sec)

.I f

l -

1 FIGURE 7.3. MASSES OF Cs! EMITTED FROM CORE AND RETAINED IN Tile RCS CONTROL VOLUMES AS FUNCTIONS OF

, TIME FOR Tile AB SEQUENCE (Vol 1 = Core, Vol 2 =. Upper Grid Platp, Vol 3 = Guide Tubes,

, Vol 4 = Top Plate, Vol 5 = Upper Plenum Annulus). Times measured from start of core melting.

4

_ . _ . =

l r-

_ Ss Ne

_ Ob I u TTe

_ C Neo l

idc r

0 Fi uf 0 SGo 0 A

= t

= 6 S E3a r

L M Ul s t

y Lo OVm V o w 0 Lef

,r Ot i 0 Rad Tl e w

y ...

0 5

NPr O

Cds u-n y

/ Sre cgm R

Eee ia rs s

e - i 0

0 Hpm T pi N

UT .

/

q 0 I =

- 4 D2)

- E s x Nl u

/ )ce I ol AVu T n

/

,n

E ReA 0 r 0 (s Dom i NCu

- 0 A n .

A. -

u 3 E M

E R1P O

= le I Cl r -

T oe .

Ae MVp O( p R U FE 0 C=

+

_ 0 DN i EE5 w-_, _ 0 TU TQl 2 IEo y HSV E

O B ,

LAe O

sea t

Oil l l 2 o-t --

o,, i 0

0 0

RTP E

ARp Oo A d v

+ o-1 FFT O

E= g 0

. n e:

ga i

t o w u- ,

, / SM EI4i ST S ll n

t L

ev vr

,/ /

l AFoe MOVm S

I - - - -

0 4 A- 0 0 0 0 0 0 7 0 0 0 0 0* E A

R -

5 0 5 0 5 U -

A

. 1 2 2 1 G I

N F A,-

mmvu x) s o i.L z 4 a

E c.

xa .

S _

O J Jd, alI< ,aw! tam 3,,i ,.s i '

i i

?' .

il  :!i ,

(ia'Ij g (-

i

'o .

7-9 -

little added retention because of the low aerosol concentrations and l residence times. After vessel dryout occurs, nearly all the emitted ,

! aerosol is retained via gravitational settling.

l There are ma,jor differences in the input to the TRAP-MELT code for this analysis from that used in the analyses performed in Volume I  ;

'of this report. The upper pier.un! has been subdivided in this case into  :

{ four ctatrol volwnes, and the geometries and masses of the structures in  !

the upper plenum have been described more accurately. Also different is' the inclusion of control rod material in the aerosol inventory, which- -

e leads to higher masses of aerosol being released from the core. The most significant difference manifest in these results, however, is the j

lack of retention.of fission products which is a direct consequence of the gas flow rates predicted by the MARCH code, and the timing of fission

?t product release relative to the gas flow rates. The use of a different

') description of the core melt progression to loss of geometry appears to h se a dominant influence on the predicted retention. '

3 q

D' 4

N .

4 t

T 1,

k y=<u = -.n ..m~.a.a.a

= = . . . .,nu, c.ww=~..

. . . ~ . . -  :. - - - _

. ~ - . ~ . - = :--- =

n -

n-.

~~c : - 9

, , a 4q 1

.a j 7-10

1) 7.2.2 RCS Transport and Deposition i for Sequence TMLS' This sequence is a high pressure one with generally low flow rate, which yields long residence times of the emitted materials in the RCS. These long residence times lead to quite high retention factors for all the species considered, as seen in Tables 7.2 and 7.4. There is a significant difference in the flow rates through the RCS used in this -

/ ,

analysis and those used in Volume I of this report. In the previous case the flow rate was so low as to prevent all but a few percent of the I 1 -

emitted materials from entering the containment until the time of bottom t

1 head failure, with nearly 75 percent of even the early emitted materials

[ still residing in the core region, where gas and surface temperatures prohibit any condensation, at the time of vessel failure. This fact may

(

b well have led to large underestimates of the ex Kat of the vapor species' retention.

h In the present analysis, the low flow rates retain the emitted T materials in the core region until just after t = 1500 s. In Figure 7.5 one can see that no significant retention of Te occurs until after this

[

.( time. This is because the TRAP-MELT code, as used here, does not permit

.' , the surface reaction of Te to occur in the core region. As the gas flow

s. ~

rate from the core increases at 1500 s, the suspended vapors and particles j f are swept into the upper plenum control volumes, where the Te reaction L{ leads to nearly complete scavenging by the structure surfaces. This 7 period of high flow rates due to core slumping also leads to increased l '4 .

Zr cladding oxidation, which, in turn, increases the Te release rate Lc ig from the melting fuel. As the gas flow rate continues to increase, .

eventually the material is transp'rted o through the RCS too quickly for l4 the surface reaction to be completely effective, and the retention factor I} decre,ases, although it remains at a fairly high value.

3 The Cs0H and CsI behavior indicated in Figures 7.6 and 7.7 is ' '

f not dissimilar to.that of Te. The cause of the small early retention of lg the Cs species is condensation of the vapors on particles which subse-7 quently settle out in the upper pienum volumes. The absence of any

'f .

1 retention of these species in the core volume is due to the high gas i

d w *

.r ap t.Q-.

b. . :fwm W -

"-w+~- ~

.: 7- ~we r"n)

7-11 TABLE 7.3. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PPEDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE TMLB SEQUENCE FOR THE SURRY PLANT (Times Measured from Start of Core Melting) i i

Cs! Cs0H Te Aerosol .

Time Ret Total Ret Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg) i 102 --

3.1 0.4 20.8 --

0.1 5.5 262 f 203 0.2 6.2 1.5 36.3 --

0.1 20.4 372 4

305 0.4 8.9 2.7 49.9 --

0.2 50.2 473 l

407 0.6 11.1 3.6 61 .1 --

0.4 92.2 567 I-61 0 0.9 14.4 5.2 77.8 -- 0.7 181 726 J.l

,. ; 81 4 1.2 17.1 6.7 90.9 --

1.0 273 854 1017 1.4 19.1 7.8 101 0.1 1.4 365 967

!j 1221 1.4 20.7 8.2 109 0.1 1.8 449 1065 oi 1424 ~ 1. 6 22.0 9.1 116 0.1 2.3 535 1160 1526 2.3 22.6 12.6 119 0.2 2.8 560 1225 1628 7.0 23.3 37.9 123 1.3 3.6 736 1315 1730 17.7 24.0 93.9 126 4.5 5.0 1155 1435 1831 22.1 24.6 117 129 6.3 6.3 1455 1635 ,

2035 22.9 25.3 121 132 6.9 7.7 1565 1740 L. 2340 23.2 25.7 122 134 7.5 8.9 l'600 1765

\

l It l

1 ,

I l .

• y m, ,_ y,.,x zuFfT%""'W'

L ..

.~

. .w.a . . y ,

.;-&'&A : _tp;,*-3u_,z y"" ,

~

- _ u --

• _.c._._._......m , __.__._. _ __

]

1 I

l i TABLE 7.4. TRAP-ELT PREDICTIONS OF PRIMARY SYSTEM RETEEION FACTORS (RF) ANO VOLUME .

1 SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIME FOR THE TMLB SEQUENCE FOR THE SURRY PLANT Csl Cs0H Te Aerosol i Time Core Pres- Core Pres- Core Core

) (s) RF Plate Piping surlaer RF Plate Piptog surizer RF Core Plate RF Core Plate 102 .01 .01 -- --

.02 .02 -- --

.04 0 .04 .02 --

.02

! 203 .04 .04 -- --

.04 .04 -- -- .05 0 .05 .05 -- .05 305 .05 .06 -- --

.05 .05 -- -- .05 0 .05 .11 .04 .06 407 .05 .05 -- --

.06 .06 -- --

.05 0 .05 .16 .30 .07 610 .06 .06 -- --

.07 .06 -- -- .05 0 .05 .25 .18 .07 814 .07 .07 -- --

.07 .07 -- -- .05 0 .05 .32 .25 .07

{

1017 .07 .07 .08 .07 .04 0 .04 .38 N

.31 .07 1221 .07 .07 - -- . 07 . .07 -- --

.03 --

.03 . 42. .35 .07 1424 .07 .07 -- --

.08 .08 -- -- .04 0 .04 .46 .39 .07 1526 .10 .30 -- --

.11 .10 -- --

.07 0 .07 .40 .38 .08

, 1628 .30 .30 -- - - -

.31 .31 -- -- .37 0 .37 .56 .35 .20 1730 .74 .59 .09 .03 .75 .60 .09 .03 .90 0 .89 .78 .32 .37 1811 .50 .54 .18 .12 .91 .55 '.17 .11 .99 .18 .79 .89 .30 .34 l

l 2035 .91 .52 .18 .14 .91 .54 .18 .13 .89 .17 .67 .90 .29 .32 2340 .90 .51 .18 .14 .91 .53 .18 .13 .84 .16 .58 .90 .29 .32 l

}

^)

.4 -

a s . .

-- . . - ~ . - - _.

.. .-.. . 2 a' -- -

di

.m .

l

.i]1 -

l

.l*

s.

N s:;

10-t Lee A

(?.r **' .

/ .

n 8- vot a + veJL

/

/ ,

t a voi2t_3._ .

j'1 3 vots :-s

~~~l O N .v.%. !.!-1 8:..  : =.::: .

g .6 - EWTTED 1< 1 i

{[ 8 z 4-

~

-: R ~

> 4

/ b x ,'/

4 < 2-

/ 4 i

1

/,  :;

/

0 . , , ,

0 500 1000 1500 2000 2500 '

TIME (sec)

FIGURE 7.5. I g MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUtiSTIONS OF TIME FOR Tile TMLB' SEQUENCE (Vol 1 = Core Vol 2 = Upper Grid Piste Vol 3 = Guide i4 Tubes Vol 4 = Upper Plenum Annulus, Vol 5 = Ilot Leg Vol 6 = Pressurizer). Times

(

t' measured from start of core melting. ' ,

h, , -

3 i 4

a _. _ .

. w_ .+ . . w- , +_,, w_ . .,.

. w.c- - -

+. -

~ - - - -

F m'

a,

15 0 -

n L*e="8 u va i h .

va i + va._3

'7 g vots t-3__ /y ,',,____________

g vots :-s.__

/,/ ,/

10 0 - N E A---- /

v) amo / /

t

- /

/ >

.l /

, y / -- ---

/

8 z /

/

4H 50- /

/ ~

4 ..

m

/

/

m 0 -

i , ,

i 0 500 1000 1500 2000 2500 TIME (sec)

.s

,f FIGURE 7.6.

i MASSES OF Cs0H EMITTED FROM CORE ANO RETAINED IN Tile RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR Tile TMLB' SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 = Guide Tubes, Vol 4 = Upper Plenum Annulus, Vol 5 = Ilot Leg, Vol 6 = Pressurized . Times measured

from start of core melting.

y -

.~

f -

a 1

i _ __ _ . _ - . .

w .::.?t-_. .- ' . xza :_ _ . v. . -.- .

a__ .

k,p:~~

0.;h .\

11 y .

M 30 4

w u s.nd

4 . vot

,j vet I + vo_t 2

• i n vots i,,s, _

'g t 4

I y v

CD vas s-s Y9!2 ti

/

/'y ,. . ------------

1 20- /

1 a m N

o.m.

/,/ . /

'/./

8,

, E /

/-

.y

'9 /, .--_.. -

V

^

z / -

W 3 10 -

/ 7 i / G

,i E j

'4 i

4 ii 0 -

3 3 I l 0 500 1000 1500 2000 2500 TIME (sec) l ,

i . t FIGURE 7.7.

' MASSES OF Cs! EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF i TIME FOR THE TMLB' SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate Vol 3 = Guide Tubes,

}q Vol 4 = Upper Plenum Annulus. Vol 5 = Hot Leg. Vol 6 = Pressurizer). Times measured from

start of core melting. - ,

t' . .

4 .

11 5._ . i

vy , e - -- _

(

i 7-16

,[ temperatures there which prevent condensation of the vapors on the

~

l particles. As soon as the aerosol and vapors are transported into tt.e f relatively cool upper plenum, condensation and retention of these speciss occurs :.t a rapid rate. The process does not, however, go "to completion" due to the large velocities which are achieved, and due to the low emis-sion rate of Cs and I after about 1400 s.

i e

The aerosol retention factor reaches a value of 90 percent for I this transient sequence, which is similar to values obtained for such sequences in other plants. The explanation is found in the long RCS

! residence time the low flow rate affords the aerosol. This .esults in j the particle mass median diameter being greater than 3 um when the aerosol

}

. reaches the upper plenum, where there is a greater alaount of horizontal surface area available for settling of the particles. Thic effect is

.;, clear in Figure 7.8. Significantly, the aerosol generation rate decreases  ;

),w .

greatly during the high flow rate period. This results in a lower amount of aerosol mass being injected into the containment than would occur for the case where aerosol generation proceeded at a high rate along with 4

the high gas flows.

While the aerosol retention factors for this sequence as analyzed here and in Volume I are similar, the location of the retention 1

is different because of the different flow rates used in the two TRAP-i MELT analyses, and the total mass infected into the containment is nearly 4

l i

. three times as great here (165 kg) because of the higher aerosol genera-l -j tion rates used in the current analyses.

L i, i 7.2.3 RCS Transport and Deposition i for the Sequence V .

1.1

The V sequence, from the perspective of the primary system, is p characterized much the.same as an intermediate size break in the cold L leg of a steam generator. The diffetence, which is very significant Ji y from a risk ftandpoint, is that the primary system comunicates directly '

l(ij with the auxiliary building (s) via a long pipe. The gas flow through

~3 the primary circuit can be classified as being fairly low during the first 1000 s after the start of core melting, and then reaching very J l

J .,  ;

!O .- l

'h s=

- . . . - - .- . = .s .

1

'i{ .

V.) ,

, ..? s

,? -

w . ;-

lLS 2000 i

, .; legend r l , .s v, ,, ,__________ .

[e '

vet I + q ,',s n  !

" '-* ._ ~ ,

i tn 1500- --- -

5 m '-t - / '/

7 F u m e ...

l ---~

• ,o

, in , .

,i

l. Q ,,,..', --

y 1000-

,,____ ~ - j

] g ,,,

o

• z eme ',*.

\

4 ','

y s ~ l

,jh

" 500- ,

l

,- /  ;

T , /. .

t 3

l 0

,/ s/ i i i i O 500 1000 1500 2000 2500 '

I

[

TIME (sec) .

.l t

FIGURE 7.8. MASSES OF AEROSOL EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE TMLB' SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 = Guide Tubes, Vol 4 = Upper Plenum Annulus Vol 5 = Hot Leg, Vol 6 = P.ressurizer). Times l*

o measured from start of core melting.

hl ,

i is t .

) _ , _ -. . . . . ..-

- I

. . l

. \

f I

7-18

, high values until approximately 2400 s, at which time vessel dryout l occurs. Bottom head failure is not preoicted by MARCH to occur until about 4000 s after vessel dryout -- resulting in a long period during l which the melting core is emitting into a stagnant primary system. The total masses of the species of interest emitted from the core and retained throughout the primary system are given in Table 7.5 and the retention i factors for the RF and specific control volumes are given in Table 7.6.

The behavior of the Te illustrated in Figure 7.9 can be explained entirely in terms of the gas flow through the RCS. Durino the initial period of low, but non-zero, flows the Te is scavenged by the upper plenua surfaces very efficiently due to the long residence time in

.q that region and Te's high rate of reaction. Since this reacted material

.;. $ is not liberated, except at very high temperatures, it is retained 1 throughout the sequence. The lack of additional retention of the Te

,l emitted 5avond t = 1200 s is caused by two factors. Until 2400 s, the flow rate thra gh the system is too high for the reaction with the sur-4 faces to occur at a significant rate. Beyond this time, i.e., after

[ 1 vessel dryout, the lack of flow out of the core region results in the vapor remaining suspended in the core region until the time of bottoc head failure, at which time it is injected into the containment atmosphere.

The primary system retention of Cs0H and Cs!.is depicted in Figures 7.10 and 7.11. In this sequence the materials which are retained in- the first 1000 s, principally via condensation on particles which are

[.i subsequently deposited, are partially liberated by the surface heatup l

that occurs at about 1200 s when the gas flow increase dramatically.

i ,

After this time, the surfaces cool sufficiently to permit condensation,

, but the inventory of Cs and I has been nearly exhausted so that no further

• I retention is possible.

Approximately two-thirds of the aerosol mass emitted from the g -

core in this sequence is retained within the RCS. 'It is clear in

,f Figure 7.12 that from t = 1200 s to t = 2400 s, the high flow rates (cr

,n$.

l

?. very short RCS residence times) lead to no aerosol retention on the

] surfaces. After vessel dryout, however, almost all further emission of v $.,

7

. aerosol is retained. As a result, there is but little aerosol injected H into the containment at the time of vessel failure. Figure 7.12 also l A ,,

W 3 "

}

&=~- g g i. ic ': m:; - } - W yyv-y ff e f-' *f~Mei,

. l 7-19 i

l l

! TABLE 7.5. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE I

~! I CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE V SEQUENCE FOR THE SURRY j PLANT Csl Cs0H Te Aerosol i Time Ret Total Ret Total Ret Total Ret Total -

i (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg) l i 300 5.4 12.0 30.1 65.4 0.3 0.4 340 687

. 600 10.3 18.2 54.8 97.0 0.9 0.9 588 999 900 13.6 21.7 71 . 9 115 1.5 1.6 805 1240

! 1200 14.0 23.5 73.6 124 2.3 2.3 941 1520 1500 13.0 24.4 69.4 128 2.6 3.0 948 1680 2100 13.0 25.3 69.4 133 2.6 3.7 948 1690 2690 13.0 25.5 69.4 134 2.6 4.0 955 1710 h I

25,6 f 3290 13.0 G.5 135 2.6 4.8 1005 1760' i

l t i j 4190 13.0 25.8 69.6 136 2.7 6.5 1150 1910

'l 5090 13.0 25.9 69.6 136 2.9 8.5 1330 2090 i 6585 13.1 25.9 69.7 136 3.2 11.4 1565 2320 l l . . . - . --- _ _ ..

t.

1. . .

l-

)

i l .

f.

t

.h [,www~.E. e ',L = k

• sK . ~L V Q d' 7 '~ .. @:*:- ~ i% " V ' %s ' -

~ '

~ UY~ ~ ' > *W * ' ~ ** A5 4 %,.M* ' T V'N TO*C-- - - - - - - '

.A . -e-e*- w-i.i ; . . .w ~ . , _ ~ . ~ . O

, - , -- c - , , ;w  ;- ------------:------ . ,;- -

(Edi. cP .s3 :i3-'hMRrhiif; ; TT+ 7 " r" ~ '~.~

~ j' - 5 ~~~~~~ 7^' ' 'T "

4'i 2"

,,,1,_,, ____ _ _

,f,.J .

.\

i l TABLE 7.6. TRAP-MELT PREDICTIONS OF PRIMRY SYSTEM RETENTIdN FACTORS (RF) AND VOLUME j- SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIE FOR THE V SEQUENCE FOR THE SURRY PLANT

's

)

~

Csl Cs0H _

Te Aerosol Tina Hot Hot Core Hot (s) RF Leg Piping RF Leg 81 ping RF Plate RF Core Leg Piping a .

U 300 .45 .18 .20 .46 .18 .21 .90 .83 .50 .03 .18 .21

~

'i 600 .56 .22 .23 .57 .21 .23 .91 .86 .59 .7 .20 .22 j -

900 .63

.24 .24 .63 .23 .24 .93 .89 .65 .10 .20 .21 1 1200 .60 .24 .26 .60 .21 .25 .98 .90 .62 .09 .19 .21 2 1

! 1500 .53 .23 .25 .54 .23 .25 .86 .78 .56 .08. .17 .19 i y 3

<n 2100 .51 .22 .24 .52 .22 .24 .71 .64 ,56 .08 .17 .19 p

!i 2690 .51 .22 .24 .52 .22 .24 .64 .58 .56 .08 .17 .19 ,

6 3290 .51 .22 .24 .52 .22 .23 .55 .49 .57 .11 .16 .18 4190 .50 .22 .23 .51 .22 .23 .42 .36 .60 .18 .15 .17 i

5090 .50 .22 .23 .51 .22 .23 .34 .29 .64 .25 .14 .16 I

,j 6585 .50 .22 .23 .51 .22 .23 .28 .23 .67 .32 .12 .14

,?j i

l

}y

.~

w - . . . . . . - _ _ _ _ . . _ . _ . .._ _ .

.s .

q -

4

, fj

4 15 d

';f! Legend vot i vot i tyot 2

.][h n vots t-s i 7 CD r.amo -

3 ---

. g 10 - / "

! (n /

k1 .

/ i:

'it /

Q ,

L.1

, Z /

,r 5 5' / l

, N x -

h l

j / -=.=

-= -

al 0

/ e n , , , ,

5 0 1000 2000 3000 4000 5000 6000 7000 i ) TIME (sec) l i~, FIGURE 7.9. MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF

! .. TIME FCR THE V SEQUENCE (Vol 1 = Core, Vol 2 = Uprir Grid Plate, Vol 3 = Guide Tubes, Vol 4 = Upper Plenum Annulus, Vol 5 = Hot Leg, Vol 6 = Steam Generator, Vol 7 = Piping).

l: t Times measured from start of core melting.

')

i

,.m fl

,( .

d. ,

u.

t 7-22 L .

. O ^

O

.l m .c i

\

l 80~

-n-l f

, U34 H

4 o go a 1 - o wem k W35 >

l i, 0"-

l i 8

, gmg

.I8.

. al

.,. , I .

o a--

eo. ,

3 al a+ "

e s

. o >>L 1 w,l  : o a 8 l'

&W 8 E 8 > e

.i l .3>>>>2,.5! .

.2

- 3. e z-

. .1, Ii o u

& .e o e

u . o-o stm I

_oo

=wn W ue a

o =.

i- a-CL O I . ii ==>

l- ,

i. ~,, .

C I

o W N G3 Un o a i

z a

. oO

, e l2 H>

h ch l U seB er v(n ogu-

[ W (d*

I l

o s o_ m.se ac >o l '

I o }-- C" .v g

. I n EO3O

-m

.' ' O>-

g g, s

,' cc w 3 M 6 C g W C .L 4

lg ,' C U aC w

' Wz ./I

' o CWEs o - cr c o

\ '

b x W as s.

WM-l II Z> T

/ Q LG r \ #

! m W 43 L

[ UZ43

\ ' O 4 3 l

' o c8,y s (- -

i g

o mm W em o D -

1 2., *C = 0 -

% Z I- > H

=\'Q.g

'b,  %

" ' ~

g J i i ,

o -

(l[ o o o o '

.- O O ar) W i

~ - g e

t .. m l: (Sn) SSVW GNIVEM l{

i L

ly'

! ' N, - .

g - ,--%.wege.. # u .#w = w xr .

e - ._ 3,y

3. ,__ . _. _..- .: , ,

u .-:. ._ 1, ..m_; -

.lj '

4 .

.N ,<

il -

' 9 30 I ig gg vas s-.

.jq _ , , ,- vas s-r was[5 F*

I.lj H vas a-s -

.% m /

Lt.: .x 01

/ (

Il

• 20- / '-

(A l'

$, Q /

,a 1 2 /

e O .--

w l ,

7 4- 10 - I

',/ y I

$ I'/ %q A

m I .j i

f./ i 0

- - - ~

j 1,' I I I I I I j O 1000 2000 3000 4000 5000 6000 700C TIME (sec) -

i FIGURE 7.11. HASSES OF Csl EHITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE V SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 = Guide Tuoes.

Vol.4 = Upper Plenum Annulus, Vol 5 = Hot Leg. Vol 6 = Steam Generator, Vol 7 = Piping). i i{< Times measured from start of core melting. '

>d .

- w..= -- . = = = . .

7-24

$2*

• EF C

OBE o 5.E

u. =

o wq" b >3 a -

\ \. t

• 8 3 b.

,' \ \\ '

g s

' o e25 8'*

m3%

\ \.

\ _o o shB*

., \

o wu.

\

s gg

\ ' \ =&*

\ ~

ng i \ s o Sm> '

o a z .

,1 \

\\ o @O E5F w>a m ,

\'

\ $$

'\ u ma.: '

2 \

\

\

)

\.

\\

\

i s a.;

o&

m w.-u 8 ->n 5E

• E8.*

, i e st i o

-Q S$at Cs&B

'i l C l

i E wmg E s *g

( l

' a>c6 mE* ,

\ Q khLE l"Ul...

I N

N

\ .% k

_o o

Wa: ah

u. 2 & M .

i --

+. 4 A A =

wi N '\ - 's

- o me.3 i 8WI I >

. i i  % \ '

3>>>>8.5 I s

3 M

m p * "E f ,>g

h. .

' s-A o o i-j bt o o o o o o "

o o o o o -

!.. e o o o e

• l,n N N - -

g g

} 3' (6a) SSVW G3NIV138 2 t

E -

L N'

.t.

. R ., ~. I ."._~ A~ ' ~ ' ~' ~

- --- ~~--- ^ ~

^

7  ;

illustrates that the ECC piping -- Volume 7 in the figure -- is an

. important, though not dominant, contributor to aerosol retention in this sequence. The total mass retained in this piping is predicted by TRAP-l MELT to be 323 kg at the time of bottom head failure. This is equivalent

l to a deposition of just over 50 g of aerosol per cm of pipe, which leads j to questions concerning this being physically possible. As the aerosol

{ deposit iti the pipe increases, one would anticipate that the constriction ,

in the flow path could lead to higher gas flow velocities. These higher velocities could, in turn, result in resuspension of deposited aerosol, 4

reducing retention. Formation of a plug in the pipe composed of the deposited material may be possible, but the existence of system pressures in excess of 100 psi make this open to question also.

^

7.2.4 RCS Transport and Deposition

l for the 5 0-c 2 (Cold Leo) Secuence '.i This sequence exhibits low flow and intermediate pressure in

, the RCS. The principal contributors to fission product retention in the

, primary system are the core region, the upper plenum annulus, and the

< steam generators.

Tables 7.7 and 7.8 illustrate the release of the species of

inte est from the core, and the retention and distribution of retained masses throughout the RCS during the in-vessel phase of the sequence.

The high overall retention factors for each species is characteristic of a low flow, high residence time sequence such as this one.

The retention of tellurium in this, sequence is nearly complete; the largest portion retained on the first post-core surface, which is -

l the' upper grid plate. This retention is due exclusively to the chemical i reaction of tellurium with system surfaces. The high retention decreases ,

slightly starting at 4100 seconds because of vessel dryout and the coupled loss of flow of emitted vapors from the cor%. Figure 7.13 111ustrates lq -

the behavior of tellurium in the RCS.

CsI and Cs0H are retained throughout the system in a comon j; i pattern, as shown in Figures 7.14 and 7.15. Each displays, at 840 seconds, !l a drop in retention in the core plate and guide tubes. The core has ll t

l, cmaac2%w, '~mL m:X X ' ~ - ~ ~ ' ' - ' -

_Y""=":_": m .w L -- .- .-. ....- . .- -.. - lL.

• a\ I

.=. .-. n

.n. .a . . -. . y - . . . . . .-

,11 . s o .1 .*

.. A ' s . L '.

• L* 't - .- n .

., _1 ~

i q

e

.A My TABLE 7.7. TRAP-ELT PREDICTIONS OF PRIMARY SYSTEM RETENTION FACTORS (RF) AND VOLUME SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIE FOR THE S 2 0-s (COLD LEG)

f SEQUENCE FOR THE SURRY PLANT l' '

a~ .

il,9 Cs! Cs0H Te Aerosol

!! Upper Upper Upper Upper Upper Time Plenum Steam Plenum Steam Grid Grid Plenum Steaa

i, (s) RF ' Annulus Gen RF Annulus Gen RF Plate RF Core Plate Annulus Gen

280 .48 .17 .06 .54 .17 .07 .74 .72 .71 .24 .18 .16 .07

.s 1 560 .73 .22 .11 .75 .21 .11 .80 .78 .82 .30 .19 .18 .08

.s

! 840 .81 .21 .35 .83 .20 .30 .99 .93 .85 .24 .16 .19 n .19 f 1400 .77 .19 .34 .79 .19 .29 .97 .91 .78 .22 .14 .16 .20 ?

.76 M

1960 .19 .33 .78 .18 .29 .97 .90 .77 .22 .14 .16 .20 I' i

! 2800 .75 .19 .33 .77 .18 .28 .96 .90 .76 .21 .13 .16 .19 2

3630 .74 .18 .33 .77 .18 .28 .96 .89 .73 .20 .13 .15 .19

. 5591 .74 .18 .32 .76 .18 .28 .90 .8) .73 .29 .11 .13 .16 l'

't 6 k

2 -

I J e

e i .-

f - _ _ . . . - - ~ . .

, . o 7-27 i

TABLE 7.8. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED IN THE RCS (RET) DURING THE S D-c E (COLD LEG) SEQUENCE FOR j THE SURRY PLANT I

! Cs! Cs0H Te Aerosol i Time Ret Total Ret Total Ret Total Re: Total (kg) (kg) (kg) (kg) (kg)

(s) (kg) (kg) (kg) 280 3.9 8.0 24.3 45.3 1.0 1.4 351 492 560 12.5 17.2 68.6 91 .8 5.0 6.2 771 935 840 18.2 22.3 97.9 117 12.9 13.1 1150 1360 1400 18.6 24.1 1 00 126 14.4 14.8 1240 1590

-j 1960 18.6 24.5 100 128 14.8 15.3 1240 1600 2800 18.7 24.9 101 1 31 15.8 16.5 1250 1650 9

j 3630 18.8 25.3 101 132 16.9 17.7 1280 1760 5591 18.9 25.6 102 134 17.4 19.3 1570 2140 i .

1 t

e 9

m 4 4

?

ib,ysWY." +;n F M-T 7 T'"ZE'"~"___._ * ~"I

• 9 a.

1.

y: A

• s y Tu'c: '.r s :T: ';fx ' ' n;;" ':~^~~~-

. ~ ^ - ':; * ~

r., _

J

')

20 0 -

!! Legend ' ~

I VOL 2

'7-

' ~

VOL 3 -

F#

~

$ n suinto - #'

CD 15 -

3 r lI s.

E y) il. 4 h 2 10 -

I @

Z

)

(

. 4-

~

i .AJ 4 i E 5- co I

I

\.

! 0 , , , , ,

n i 0 1000 2000 3000 4000 5000 6000 a

ei TIME (sec)

. FIGURE 7.13. MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS C0tiTROL VOLUMES AS FUNCTIONS OF TIME FOR THE2S D-c (COLD LEG) SEQUENCE (Vol 1 = Core. Vol 2 = Upper Grid Plate, Vol 3 =

1 Guide Tubes. Vol 4 = Upper Plenum Annulus, Vol 5 = Hot Leg Vol 6 = Steam Generator).

l' Times measured from start of core melting. -

k .

Uh

s-

_v u. . . . . ~ .. . .

2  :.. _ _ _ .. ___ _ , . - .. ...

r

• I:

' )2 s g

?! -

15 0 ,,_,

vot a

j.; vot a + vo_t ,3 vots -s

[yTgo __ _

3

? m y.ot u .i ..

,/.---~______--------- .

2' cn /

v

_x 10 0 - / - -- - -- - - - -- - -

(n k

1

-l J

l C

k l s- -

i -

50- / /, --------------------------------------------------------------

s.

li W

4 // l 0

l ,

/ s__ ___ _ . _ ___ ___ ___

l 0 , , , ,

I' O 1000 2000 3000 4000 5000 6000 TIME (sec)

FIGURE 7.14. MASSES OF Cs0H EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCT OF TIME FOR Tile 2S D-c (COLD LEG) SEQUENCE (Vol 1 = Core Vol 2 = Upper Grid Plate, '

1' Vol 3 = Guide Tubes, Vol 4 = Upper Plenum Annulus, Vol 5 = Ilot i.eg. Vol 6 = Steam

<; Generator). Times measured from start of core melting. '

.I '

F 1, '

?

,, m ,. . 1- - .c.. . -m m. .: ; . ,. ___

d e

4 h

D il 30

Q Legend vots 2-s ji vota vots

-s y vot a + vo.u twnto 1

y s u .t.... - _

m /_ -

G V f i d

. m 20- / . . . -

W

h. < l l 2 /

I' C u u /

z l

4m 10 - / s

-- ~

a O

I' :: A p

3 y lyl'~,.............................................................

l I :,

ka (j // - - _ _ _ _

T s i 0 '

e i e I

) 1000 2000 i i i

'O 3000 4000 5000' 6000 i

.)

r.1

. TIME (sec)

.k FIGURE 7.15. HASSES OF Cs! EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF ij TIME FOR THE2S D-c (COLD LEG) SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 =

Guide Tubes, Vol 4 = Upper Plerium Annulus, Vol 5 = Hot Leg. Vol 6 = Steam Generator).

3 Times measured from start of core melting. '

1 -

F t

t

~

}. .

4 .

l 7-31 slumped at this time and the resultant increase in surface temperature

,- to 1900 F in these volumes causes re-evaporation of the two Cs species l; and their transport to downstream volumes. Emission of Csl and Cs0H I -! from the core is essentially complete at 1200 s and the retention of the

! i l j , . two species undergoes little change after that time. The steam generators

! are seen to 'be the site of the bulk of the retention of the Cs species. b

} Aerosol retention, as illustrated in the tables and  :

Figure 7.16, shows a less pronounced response to the core slump. '

g Retention continues to increase, but at a slightly reduced rate. The increased flow rates accompanying steam generation when the one slumps serve to reduce agglomeration. This is evident in the aerosol mass I

median diameters calculated wy the TRAP-MELT code. In the core region, ldj the particle diameter falls from 3.6 um at 560 s to 1.0'um at 840 s.

I Gravitational settling in the core region continues at a reduced rate, and transport to cooler volumes allows continued retention. After this  ;:

time retention shows little change as aerosol emission from the core l, slows considerably, until 4100 s. At this time flow has stopped, the f

g freshly emitted aerosol staying in the core region, where gravitational settling effects an increase in overall retention.

t' 7.2.5 RCS Transport and Deposition for the 5 0 (Hot Leo) Sequence 2

l This sequence is characterized by low flow and intermediate pressure in the RCS. Retention occurs primarily in the core region and the upper plenum annulus. Use of the hot leg as the site of the small pipe break removes the steam generators from analysis. The resulting .

decrease in overall retention, except for tellurium, is the standout i

, feature when this sequence is compared to the previous one. Tables 7.9 and 7.10, along with Figures 7.17 through 7.20, illustrate the release,

, retention, and distribution of species of interest.

Tellurium shows increased retention, when compared to the cold-leg case because the flow in this sequence does not subside after core

, slump. This is due to shorter duration of the in-vessel portion of the Li . accident, 1750 : versus 5600 s.

i!

ah 1.tpg.p. A. @;NQT27'@7N N b -

3EAEb

n% ,

. c:v :- rg+:, = . --:.- - -

~

. . ~ - -

' =

~'

___a_____,

llw,!,

J _.l

i ,

'! 1 i1 l, 2500 l, L.o.nd l 1911-i, EtLis v.n a -

Ij 2000- YtM tri--- -

,, A vets t-s 2 rau.1-s '

v. Leu __ -

!i u) f--

_~ .

!- y) 4 1500- p'.-  !

/ '

i' 1 j - -.. . .-----***~~~../ ,.

c y 1000- / - [,I. ,.' ' ' . .  !

4 l ,- i, L

/ ' ,

t tr 500-

,/'f _ _

/ /

A y'.-

h

/ if 0- , , , , ,

3 0 1000 2000 3000 4000 5000 6000

, TIME (sec) i FIGURE 7.16. MASSES OF AEROSOL EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF TIME FOR THE SpD-c (COLD LEG) SEQUENCE (Vol 1 = Core, Vol 2 = Upper Grid Plate, Vol 3 =

Guide Tubes, Vol 4 = Upper Plenum Annulus Vol 5 = Hot Leg. Vol 6 = Steam Generator). Times j measured from start of core melting. -

1

[- .

g-___... _. ._ . . . _ _ _ _ . . _ . . _ _

. o 7-33 i

i TABLE 7.9. CORSOR PREDICTIONS OF MASSES OF SPECIES RELEASED FROM THE '

CORE (TOTAL) AND TRAP-MELT PREDICTIONS OF MASSES RETAINED I

IN THE RCS (RET) DURING THE S 0 2 (HOT LEG) SEQUENCE FOR THE SURRY PLANT ,

l

.l Cs! Cs0H Te Aerosol Time Ret Total Pit Total Ret Total Ret Total (s) (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg)

. . . - -s 103 1.0 3.8 7.5 24.4 0.1 0.3 224 336 206 3.1 7.6 19.9 43.5 0.8 1.2 348 494 1

309 5.8 - 11.5 34.1 63.2 1.7 2.3 481 652 41 2 8.5 14.9 47.8 80.4 3.0 4.2 629 81 5 H 51 4 11.2 17.9 61 .9 95.3 5.3 7.0 776 974 61 7 13.6 20.3 75.9 108 8.4 9.9 907 1130 720 11.7 22.0 76.9 116 11.3 12.1 1040 1290 l 1029 12.6 24.5 78.1 129 17.0 18.1 1190 1610 l i337 12.3 25.5 76.7 134 22.6 23.0 1240 1790 i

1749 12.3 25.7 76.5 135 23.8 24.3 1240 1870 e

i

- 1, e

t O

IA S 1, pH %el

{ *S'"

._{ . " ??N s.e.. - ..f.-

,'& . % " "wf b?h' i ' ' .* 5b e ' '. ' " S 'N

' %' ? -_-, ~_,

, p; &W_c ,^ t ; r.;_v.11b nh ,. a t.-a

. . ._ \ > i

.".s',. .___. m.a a__._.

t . . . . - . .

' !I .

.!ir.

t .

I b

'l .

TABLE 7.10. TRAP-MELT' PREDICTIONS OF PRIMARY SYSTEM RETENTION FACTORS (RF) AND VOLUME N SPECIFIC RETENTION FACTORS AS FUNCTIONS OF TIME FOR THE S 2 D (HOT LEG)

SEQUENCE FOR THE SURRY PLANT

1. .. ,

if $ upper CsI er upper Cs0H Te upper Aerosol upper per 4 'iles .

Grid Guide P enus Hot Grfd Guide esaa Hot Grid Grid Penus ij (s) RF Plate Tdus Annulus Leg RF Plate Tubes Annulus Leg AF Plate AF Core Plate Annulus

? 103 .25 .18 -- .06 .02 .31 .21 --

.07 .02 .46 .45 .64 .40 .16 .06 206 .41 .22 --

.15 .05 .46 .25 .01 .15 .05 .66 .64 .70 .37 .18 .11 .

'I 309 .51 .25 --

.20 .06 .54 .27 .01 .20 .06 .74 .72 .74 .37 .19 .14 1

.f 412 .57 .27 .01 .22 .d7 .59 .30 .01 .22 .07 .71 .70 .77 .39 .19 .14 7 u

514 .63 .30 .02 .23 .08 .65 .33

, .02 .22 .08 .76 .74 .80 .40 .20 .15 I

', 617 .67 .32 .05 .22 .09 .71 .36 .04 .22 .09 .85 .83 .80 .40 .20 .16 720 .52 .12 .09 .21 .11 .66 .28 .08 .21 .10 .93 .91 .81 .38 .20 .l?

., 1029 .51 0 .15 .21 .15 .61 .14 .13 .21 .13 .94 .9) .74 .32 .17 .38

'i g 1337 .48 --

.13 .20 .15 .57 .14 .12 .19 .12 .98 .93 .69 .30 .16 .l?

1749 .48 --

.13 .20 .15 .57 .14 .12 .19 .12 .98 .92 .67 .28 .15 .16

.i 1

h, 1

i -

.1

i __ . _ _ . . ~

__1 ,

] _ __ _ , _

t,

!!j 30 lj; j' t  :

p3.

n Legend '

• • CD VOL 2 vx d ,

j VOL 2 + VOL_3

[

20- -

M EWITTED~~~ e ..

J; ii B

C ,

f i:

1 <(

10 -

e y

a m m

i W

/

r, 0 , , ,

0 500 1000 1500 2000 i TIME (tec)

)

~

Ila a

{ FIGURE 7.17. MASSES OF Te EMITTED FROM CORE AND RETAINED IN THE RCS CONTROL VOLUMES AS FUNCTIONS OF j TIME FOR THE2 S 0-y(HOT LEG) SEQUENCE (Vol 1 = Core. Vol 2 = Upper Grid Plate, Vol 3 =

Guide Tubes Vol 4 = Upper Plenum Annulus, Vol 5 = Hot Leg). Times measured from start 1 of core melting. ,.

1 1

l.

j . .

4 7-36

%N as o C. QJ M C. J

• <3 a

I o M is e "m:

o Q II

-1 N is am a -o

., Ce>

,, J mL

• Ho *

• l

• 8 Z U ut 8

C 3 I U as -

8 O s*

. M-c e

I i u e a

a 8

g m-< o 4

Q W>E e =w3

.4 e

I o H W@

c g N

" 2O-it w Z Q.

g I

• C3L t'*' 3 l W C* Go g I Z W C.

I - m c. .

". I i 4 3

g 8 I

He m W C 15 c l4 I mW w

',' s

• I O

O c

zw-M

' il 1 8 I @ # E S .g' e i. W-

.. o a .u A

.>' l t

l aQ C ). ut o U e 43 u 4

i n w C4 4 M

', g *N ~o s%

• • 4 =  % ( 2 m

m WWDa 1  %

3  % =7 6 C & **= @

L. .  %

/ W 3a

. t / H e c vi c.'1  % F

/ Wo

,i l *

[ w 6 is E

• g

8. E o

, g (

W L*J P3 6

<,, y,

% E  %

= ls \ w-

',  % MWo%

\  % C > 43 g

% Q ut L. 6

, -o uoJg supe e s *

%, c u. m u -

ozg q;

, .J j l e8 ** ,  % m .C c, E r O I  :* *,  %% Wp y,

> gg e e *

% muve

7 T_ +, 3 ye l Of 8

• .M2w E OLw CM N g

g , ,'s, WCH .

- ea .a a .4 c. ,

.y 3O m > > > > 3e O O O esse Q

CO

' - g 8.*

1,

%,- g f W h l )

m o a

.; o o o o 8 S -

o e o C O

j 9

a (6>t).SSYW 03NIY138 n

)

a

.d d

1 N I N Y '

[ Y *", f 5 -

. l= & Y ** Y* ** * 'O ~' ' * * ' ' ' ~

• } WA 5 VM((! , , , )

e e 7-37 "

mM 6

- et3 MOM Z>M Q

e . in HW Q uaL Z tts Lb-D-

O Q mm'

<- 3

, O MG*

b W i .

N

?

t 8

yGJbI

! IaM l o c. W

>D E w.

' ' s 3 If H

( 8 O .

i 8 %N g l H

• g E-m 3 l I CCQ j i e i u>c J -

g Q m a  !

V JM I, *

' O w '- o a:

' c

-  !$8 ' 3

'u Gn "

e, ,

m l

,? I l

l l E-i

, - - .f

1. 4 -O .

! i ce>

w>

'I -

i g Z% . i

! 3 - s9 4W3 l~

.l l t t I A HU-

%F WZ3 l rl g I zwr g

l g DC l' t

'g l

l $

1 O$

Q

• b' z

<mg

] e

• Oy w-c 20W

'q l ,

g g )

t 1

"2 _

ow-V -.3 A t

t f

g

, ,I I h EHb co s

S MZ1 l 1 i'

g # wwa

( s D g

• C >=

t  %

W e u l HQ l

', s  % H Nw l (

-m .'

i i ,

3" - .

l

' 's- \  % \

O w W O On z>c l

-O ';;

• 3 8

I sept l

s l

e l

a 1

g

@ U QC M -

C c; m

• s J 8  % WW4R o g I ' 's  %% C 2 l'

O *

% m r y + N 1

N I as *.

• w-eo mHVU CN N w e ,

• , m -

e

, ea a 5 5 ..e4 ,, , 18a%

ao

. o o o  ;

eJ > > > > w' N -

'- , m

[

', 7 i

s O w e

Q Q O O D M N "

~

u.

1 (6 M) SSVW G3NIY138 e

~

4..M 7.a

^

1  % en g * = , ', . , 4 e

,T7 g ,y*.ygygg p=*** s-------y ,.

- - - - > , .1,

~ .

s.m +_ .. .~

pn y- 2 ' . :. :, . . . . . - .. - - .

. ._.....-...-.: =- -- .-- -- - - - - - - - - --- -

~ fl a:i N

lii '

, ' 5,3

• Xf 69 2000 1 i ljj Legend 3 VOL 1

.l2 VOL 1 + VOL---2 ,'

!e n

t a a 1500- v0Ls 1-4 ,,',,

H v 2 ---- -

s VOLS 1-5 -- ,e'

,1 M EMITTED , '/

q l Q y

y~ -

-3 1000-- ,'

( ,'*

, s s' - - - - - ~ -

Z_._ , ' -- s s

4 ,' L,

/ w b

.t ,' I' /

}

W 500-

,',' '/ /

l1 l, /

, i! /

!!j 0 , , ,

, 0 500 1000 1500 2000

}

i4 FIGURE 7.20.

TIME (sec) ll MASSES OF AEROSOL EMITTED FROM- CORE AND RETAINED IN THE RCS CONTROL VOL

1 0F TIME FOR THE S2D y (HOT LEG) SEQUENCE (Vol 1 = Core. Vol 2 = Upper Grid Plate, Vol 3 =

Guide Tubes, Vol 4 = Upper Plenum Annulus, Vol 5 = Hot Leg). Tings measured from start

, .i i of. core melting. ~

Al

-l

• J 0

. lI

,n '

6 e

t 7-39 The Cs species again show the decrease in retention attributable

') to core slump at 720 s, with its resultant tcaperature increase from 1530 F to 1980 F. Cs1 has a negligible sur/ ace reaction rate compared 5

to that of Cs0H, and it shows the more abrupt decrease in retention at

! this time. Retention for both species has recovared by 1100 s, at which

] time Cs emission from the core has slowed considerably. Retention is

.! e:.sentially unchanged from this time to the end of the sequence.

1 The aerosol retention pattern is similar to that of the previou's

sequence with one exception as illustrated in Figure 7.20 at 1100 s.

Lj While emission from the core continues, retention has stopped. This is

] due to the increased flow beginning at 1070 s, which results from the j core slump and subsequent steam generation and pressure increase. The j mass median diameter drops from 3.27 um to 1.79 um in the core region, Fj and from 5.1 um to 2.5 um in the hot leg, from 1030 s to 1130 s. The

'I shorter residence times and reduced agglomeration serve to halt gravita-

{l tional settling in all volumes.

. 7.2.6 5 0s with Fission Product 2Heatup l _

-t (In preparation) 1

'j  :

-1 M

J n

i '

{. ~,We

=

4'Jw3 -~ t t t =*

cJ... m _ vg,

.- r-c -. -- -- - --

9

e I

j 7-40 I

'/ .3 Transoort. Deposition and Leakage in Centainment Calculated results are presented in this section for analyses performed to examine the transport and retention of various fission products in the containment. The NAUA code that was described previously jl was utilized in the analyses.

I In general, the NAUA code used here needs information on the thermal hydraulic conditions of an accident of interest. The conditions provided by the MARCH computer calculation were used. The typical required thermal hydraulic conditions are time-dependent containment

, temperature, pressure, and wall temperature, and the rates at which steam b

enters the containment, condenses on the containment structure, and leaks from the containment.

Perhaps the most important and critical input that containment g codes also need is the fission product source term for particulates.

.y The source rates calculated as release from the primary system (TRAP-MELT code) and the VANESA calculations for release during the core- '

concrete interaction were taken for the melt and vaporization ' releases,

,' respectively. For the NAUA calculations, CsI, Cs0H, and.Te were

, distinguished. All these species were assumed to be in the particulate

form in the containment atmosphere because the temperature and pressure

{ under the containment conditions indicate that these species will remata i i as particulates for all practical circumstances. Although it is assumed '

ih J, in the calculatinn that individual species are distributed evenly over all sizes of particulates, differential amounts of these species at a l

[jU given time due to different source timings were taken into consideration *

{

l I'

Lh in the calculations.

p The aerosol behavior. mechanisms considered in the present 3 calculations include agglomeration due to Brownian diffusion and due to M sequential gravitational settling, particle size change due to steam condensation / evaporation, sedimentation, Brownian diffusion, and diffusio-

• t phoresis. Additionally, homogeneous nucleation of water vapor was '

l[ included to model the water droplet formation in the containment. It p should be noted that diffusiophoresis and homogeneous nucleation of water 7

b '

,k w .

-_s e:

--[=&.*Q.s **T2M*"2: - MOD' M.F *2& ~"

?L ' ~ 7 ; '

. . - . . - - = -

j ,

7-41

vapor were not considered in the previous calculations (BMI-2104,

'] Volume I).

y Four different accident sequences, AB, TM.B', S'20, and V, were considered in the present calculations. The timing of the fission product source for the containment calculations was permitted to coincide f with the prescribed accident sequence that was listed in Chapter 6. Thus, jj the melt release of aerosol mass occurs as the core starts melting and i

the vaporization release takes place as the core-concrete interaction

'il begins. The time-dependent source rates in mass per unit time were pro -

j vided as input to the NAUA calculations. In certain cases effe:ts of I

• multiple compartment modeling, notably for the A8 4 and Y sequences, were further examined. The timing of accident events of each accident i sequence considered in the present calculations is sumarized in Chapter 6.

1 \;

7.3.1 AB Sequence _ _ __

I  :

This accident represents the sequence in which a minimum reten- -

I

}

tion of fission products is expected to occur in the primary system due to the presence of relatively rapid steam flow conditions combinea with f

i a short transport pathway. For analysis of the A8 sequence, three l different containment failu*e modes were applied and these are identified using WASH 1400 nomenclature as A8-6, A8-y , and A8-c. The containment failure times corresponding to these sequances are shown in Chapter 6.  !

~

For analyzing the A8-8 sequence the safeguard building was included in l the analysis. As mentioned previously, calculations based on a multi-  !

j compartment model were additionally made to compare with that based on a '

reduced volume assumption for the A8-6 sequence.

A8-6  !

This containment isolation failure mode was analyzed using two f compartment models. The NAUA code was run sequentially to calculate the L aerosol behavior in the containment and in the safeguard building. t Figures 7.21 and 7.22 are the airborne mass in the containment and  !

j i

i i

6. f 5 ( 4 "'7 F* "" '" " T * * ~ ~'Y' '*

. Q,[ N * "Y."

• A t th7 w.v.

.d T ~

• ^ .. s

, ,,;, ;g,

. - - - - ~ - -

(

i g 7-42

~

AIRBORNE LEAKED -

~

f j

b:

.f.

e .

-1 .

l '

/- -

.V co .

. (n .

. (n '

4 Cb:

r- -

j d:

s -

/

. o . ./

, s .

i -

P l

' b_

i I

1 ,  ;

., i-

' L e b. 10 '* . . . . . . . . , . .

't . 10~ 10 10' y

TIME, MIN r

3 FIGURE 7.21. AIRBORNE AND LEAKED MASSES, AB-s (Safeguard Bufidincj).

O r

d j i

.v ji

'V '

. k. ,

+gni : ,, . , 'L? -

' ~ ~ -

K}', C ' ' ^ ~ J -.D.~'~~

' '~

= :: [

-}--- . . _ _ _ _ - - . - _

. . . . . . . . . . . .---;- .= . .. _

=i 7-43 l AIRBORNE O j .' LEAKED / -

\ ._

j p .-

1

[ '

r b_ -

• i

/.  !

. i

. . t i

cs . .

' m m

To.

~

d-w -

cs . -

H . .

2 i

)  !

4 -

.-j .

f f

i J

.b- . . . . . . . . , ' . . ....,

ggn g ' . . . . . . . .

TIME, MIN l

.i I

FIGURE 7.22. AIRBORNE AND LEAKED MASSES, AB-s (Safeguard Building)

- ,,_-(.c ;giu.u m. y v.., n ;. ,, ,. .

- - A- - - . _ -

m .- - - -. - -- _ _,... _ - .

a

' q l}{

- ~^

q

~

7 44

. safeguard building and Figure 7.23 is the accumulated mass released to the environment from the safeguard builifing. Figures 7.24 and 7.25 are f the particle size distribution in the containment and safeguard building.

,j' The location of each species after the accident is completed is shown in Table 7.12. Table 7.12 shows that a core inventory fraction less than 5 h percent for the CsI and Cs0H species is shown to be retained in the '

) ' primary system due to relatively high flow velocity. As a result, nearly; 80 percent of the Cs! and Cs0H inventories are retained in the contain-i

[' ment and the safeguard building. Distribution of Te is seen to be I M substar,tially different from those for CsI and Cs0H in the table. As

'[ already discussed in Section 7.2, the timing of the Te release during t the melt release is much delayed and a substantial portion of the Te' .

core ;nventory is released directly into the containment during the core-3

{s concrete interaction causing the distribution of Te to differ from those for Cs! and Cs0H.

In order to examine the effects of the multi-compartment model-ing on the release fraction, the containment was divided into four

{ separate volumes and similar calculations were performed. The calculation results showing the locational distribution of the CsI, Cs0H, and Te species are shown in Table 7.13. It is interesting to note that the 9 fraction of core inventory for each species that is released to the e environment is smaller than the corresponding fraction shown in

'5 Table 7.12. While it is rather difficult to assess precisely the effect h of the multi-compartment calculation on the vadionuclide behavior, the j observed increase in retention appears to be primarily due to altered y flow patterns and thermo-hydraulic conditions, such as recirculated and

[ ,

periodic reverse flows, as modeled by the NRCH calculation. "

( ).i

,{ A8-y

') .

,} This containment failure mode represents an early overpressure

,yj failure due to hydrogen burning. The containment failura takes place at f a time of 269 minutes which is about 159 minutes after the vessel failure.

7 The VANESA calculation resdits shown in Table 6.13 were used as the source

.[l during the core-concrete interaction for this sequence and for the AB-c -

l . L.

lD -

3 l 2

g ., . = g,33;;;  ;;3== = =- -

H

t I

~1 7-45 1

b:

CSI ---------

1

' ' CSOH - -

il TE -- '

1 D1: OTHERS --------- :

1

.s -

1 l

.i  % - ~

n -1 l -- -

./ .

1.  !

,. / ./

b_

i

-e ,/ .

g - '

/

m ot

-e / _

(1 @ "

f E- '

/

db s -1 C  :

e  : -

. I 3 b, -

's i 4

l I

r o..

I 7 .

. o.,:

e 4 o '

g

' ' ' ' 'tO' ' , , ,

,,,,,,8 , , ...

10 10 t,

TIME, MIN '

s 6

FIGURE 7.23. ACCUMULATED LEAKED NUCLIDES, AB B

-g 1

4

'd' I

. _ . ,_., m- _

. ~ c;'- -

ys,3, .- _ ~ .

_- _ , g 7 ~ .: ,; g ._ ,s.,jj,i-

.l)!

l  ; 7-46 t o 1

j  : .6 HR.

4.2 HR. - -

.i 6.8 HR. --

['

14. 4 HR . ----------- ,

20. 0 HR . -------

.y g*:w

- N 1 .

~ .

4.j

,e s, \

\ -

s .4 o

c.,  ;.

s

\.

/ ll 's

% a. -

.-. s 3-~ a: '. ' '

z

\ 'ss \

h ll .

1 ,C3

,: . i a e / ,

\ i' x

~

ll \ ,

,,  : i .

. r ,  : i M-.

Z a- '

', ,/ :[ I. t g d ..

i i .

8-o .

• ll

• i

\\

\ ' ,

,  : i, w l l: \ i z .

i

.i a a:  : i i

\

• i \*

x:

l li

\ i

i c '

ll i

\:: 'i. -

i.

. i

. . i .

.. .f i t.b I .  ? I o

F

'o . -

\

e, t

.y .

3

?

=.o i . . i ..i i , , , , , , , ,

......i*

U 10* 10 10 10 10*

0 RROIUS, MICRONS r

i .

lh -

FIGURE 7.24 PARTICLE SIZE DISTRIBUTION. AB-s (Containment)  !

s

?0 t< a 4

.5

-^-,-nm , ,.- R =9e6 W , =v ' w * ~ y 4 x, ll Q_

_ _ _ _ . - _ _ _ _ . _ - _ - _ . . _ _ _____._______i

q .

. 7-47 f

a

!  : .7 HR.

4.3 HR. - -

. 6.7 HR. --

.i e 13. 8 HR . -----------

-i

18. 9 HR. ---

'o l.

.A

./N

• 8.3 -

/ ~

\.

N  :

, cs .

. i g'

-, o_

f .'s \

w -: ,f .. . .

c : .

s i x s -

i l!

if

.\

i

- a .

i,

\

,o .:

/

-1 f ll \. '

' a- -

\: \,

//

,i:t i

!M / / '

g:

m /

ll/ .

\ ,','

t l

e

-e

• i: \

.  : i C -: o- l l l: \ '

. ,; . i.

,: s

,: n t ,

. ,: s I. .

,'l

?o.

l

. ,, t T . ,

\

l \

l

, 'o 10* 10' 10 10 10' -

RADIUS, MICRONS l i

FIGURE 7.25 PARTICLE SIZE DISTRIBUTION, AB-s (Safeguard Building) l w .. , .,, . .-

...~~..n-.-__... _ _ _ .

I. . : . : . :. L L

~ - ' "~F'[." t

.4i .

'i

-1 7-48 i

TABLE 7.12. LOCATIONAL DISTRIBUTION OF SPECIES AFTER

'- l. ACCIDENT IS COMPLETED, AP-s 2

~

Fraction of Core Inventory

, Safeguard

-! Species RCS Containment Building Environment -

t i

F-i Cs! 2.7 x 10-2 0.59 0.29 8.7 x 10-2 I:

e.

Cs0H 3.8 x 10-2 0.59 ^ 29 8.5 x 10-2 Te 0.26 0.19 c. 7.0 x 10-2 j,

7

% .u l

.- TABLE 7.13. LOCATIONAL DISTRIBUTION OF SPECIES AFTER

,1 ACCIDENT IS COMPLETED, A8-8 (4 Volume)

I

!' , Fraction of Core Inventory Species Vol.1 Vol. 2 RCS Vol. 3 Vol. 4 Environment lfu CsI 2.7 x 10-2 0.33 0.36 0.23 1.8 x 10-3 5.0 x 10-2

f. Cs0H 3.8 x 10-2 0.33 0.36 0.23 2.2 x 10-3 4,9 x 10-2 ,

1. Te 0.26 4.3 x 10-2 5.3 x 10-2 0.19 4.9 x 10-2 4.2 x 10-2

4 --

( ' ',"j

, - S; l !_1 iU Lb,s -

[ f .0 .

I

,il l i '

p e, - -. -

= .

==- -

.I -

i  :

}

7-49

, l 4

sequence. Figures 7.26 through 7.28 show the airborne mass, the accumu-

-lated mass leaked, and the particle size distribution for this case.

3 Table 7.14 is the locational distribution of each species after a sufficiently long period of time elapsed. Table 7.14'shows that compared with the calculated CsI and Cs0H fractions that are released into the environment as shown in Table 7.12 for the AB-S sequence, slightly lower fractions are found to escape the containment. However, the Te release fraction is substantially higher because an increased amount of,Te is available in this sequence, compared with that for the A8-8 sequence.

, A8-c

~

J 3

Figures 7.29 through 7.31 show the airborne mass, the accumu-lated mass leaked to the environment, and the airborne particulate size distribution at various times for the A8-c sequence. As shown in Chapter 6, the cor.tainment does not fail until a time of 1450 minutes in

} this~sequcAce. The first peak in the airborne mass shown in Figure 7.29

} .

represents the source from the RCS prior to the bottom nead failure; the e

second peak represents the source during the core-concrete interaction.

The particle size distribution shown in Figure 7.31 attains a bimodal distributional apparently due to the condensation of steam vapor onto Ij particulates. However, when the thermal hydraulic conditions in the j containment become such that there is no further condensation, the size i

. distribution becomes a unimodal distribution which seems to fit approxi-

[ mately the lognormal function. -Table 7.15 shows the results of the fraction of core inventory for each species located in various compart-ments. *

- 7.3.2 TM.8.' Sequence ,

l~.

Two containment failure modes were examined for the subject

.[ accident sequence. In the first failure mode designated TM.B'-de the ,

q containment fails 152 minutes after the core uncovers and this time i 1

coincides with the event the bottom head fails. At this point the  :

containment pressure reaches 89 psi. In the second mode designated i i:

l. ,

i <

ti s

N

bS i 7-50 1

b M FIIRBORNE

! . LEflKED - -

f t

b_ .

<G. . -

. ,. c3 - .. / .

m m -

E.

s C ,

H .

a .

H

b. -

1 l

• l!

I:-

~

i i

a ,

4 l; ,

1

?,.

k

.i -

I l.c

.y l' i *S . . . . . .... . - - i - - ' -

10' H. TIME, MIN t , .;

t f, FIGURE 7.26. AIRBORNE AfD LEAKED !%SSES, AB y.

.I s.-> _

%, i ly" \ , e ., .

, M..e x2 .L - , we . - ~ -- ~

. . . - - ^.~-V~-

. -~

' ' ' ' ^

'" ~ ' N" ' ' ~M -

• Y???"Y

a m .A i ,

t l

7-51 CSI e  : CSOH f - ~.' ' ' ~ - - - - - - -- -

_ TE ,' - -

l _

OTHERS --

I

,i

- I t

b_

i i . .

i i j _ i i

CD

. I m <

)

m ,.

Eo. -

1 g- p._. - . .-

. s

e. -

r H _

i . I l

l' i

b. .

l 1

.i i I

I l

,,,,' , , , , , , , , .i 10' 10 10' TIME, MIN  !]

t 11

! FIGURE 7.27. ACCUMULATED LEAKED NUCLIDES, AB-y i i

t s l

4i

.

• N ,YN' "W *'PWT.* g _gg4gpp g 96g_ - l

.__.~#

1. s

.; x,;,gacy

+

$.T

• t .

i:

f- 7-52 t

o

.6 HR.

4.2 HR.

/.m' l - - l i

6.4 HR. --

I -

14. 0 HR . ----------- -

y

19. 9 HR . ------

[ \- i

. /

. \ -

i  !

s. r

{I .

Q

,1 e

Ora i l .

N: -

,/ .e e  : /

, s . ,

,. . ,: . i r - '  :

l l'

(..

z o

,i

\

\:

i

. - ,e s . i C

l l! \ \ t l j5tc

/ \\

z l;/  ; i i

\,

• 5, .

l ll

,. \.

i i

o .

,  : i

\

,i \ ','

t LJ ,:  : i Z

i "r l! i  ;

ll \: \

S o_

, ,i i x ~: ,; \ \

,! ! ii ,

C .  ; :i i

, ,e :i

.i l

e . ,: :i

,: 1 i

, ,i ,  : i

5.o _ -

-: l j s  : ,

l 1.

o -

l

. r.. ,.

~

.P, l

=.a .

. . . . . . . . . ..rT

.......i' ' ......' i

., 10* 10' 10 10 10' .

H RADIUS, MICRONS

~1

y. l

! FIGURE 7.28. PARTICLE SIZE DISTRIBUTION AB y  !

I-N r

f- .e . - WA,

. . > m.-

.n, Tw" . . f-

~.--w,

.. +^.-

.~. J, - . . _ ' .

, _ . -_ . 4 .; w ... n p. ., .

7. ,,, , es y_ w_. .. .., r ..

,l

. ... j 7-53 J

~1 TABLE 7.14 LOCATIONAL DISTRIBUTI' 0F SPECIES.AFTER

! ACCIDENT IS COMPLETED, AB y 1

7 Fraction of Core Inventory

)l 1

Species RCS Containment Environment l Cs! 2.7 x 10-2 0.92 5.7.x105

.a

-: Cs0H 3.8 x 10-2 0.90 5.9 x 10-2 .

2 Te 0.26 0.31 0.14 a

d

, . TABLE 7.15. LOCATIONAL DISTRIBUTION OF SPECIES AFTER

' -i. ACCIDENT IS COMPLETED, AB-c

, .{

-:l Fraction of Core Inventory 4 Species RCS Containment Environment Cs! 2.7 x 10-2 0.97 4.8 x 10-5 5

Cs0H 3.8 x 10-2 0.96 4.7 x 10-5

', Te 0.26 O.45 4.0 x 10-5

-4 -

l l

'I e

,1 4

'{

r !!

4 5

LL% Lw.t u- o n=;w;.~ ..  : ,wpy;.. _ x. M..<_yg,,~37wo% wvmw:wx***w* ~v-:- . - - .

a - .~ .._

_ _ . _ _ _ _ _' ~~~~ ' ~ !~

\

'"^;_..___ _- ... _ . _ _ . - , . , _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . - . . _ . _ _ . . _ _ _ _ _ _

r i

t

. 7-54 t

b-F '! 2 RIRBORNE l -

LERKED -- -

t b_-:

b_

f -

4 -

c_ .

cs -

u,~ -

i. (/)

~

.. _ .~ei

- ,r  :

.c- .

s .

o .-

s -

b_ /.

i ~! -

/.

/.

/.

o I

.' b_

.. i j .

p .

'o

?.1

~ , , , , , , , ,, ,

. . . . . . i i i 1

J 10' 10' 10' 10,  :

f TIME, MIN  !

2r '. I

.L FIGURE 7.29. AIRBORNE AND LEAKED MASSES, AB-c -

n

.y

a. gi "O hi g fY 4- ~L. (; >

. s m '. *', s '/6- . r- ,'N Z a * .-w.

• , _ , _ . . + + m% e. -' _*%'.c ,+' =. .

,% >g < < r= : ** .4 e -  : v~ .w m 2I I

______-__.___-_______'s-

~ ___.

n.i

.-n..- . . , . - . - -

~-4.*= g

I I

i i

7-55 "O

- E. -

i CSI CSCH - -

b., TE --

~5 OTHERS ----------- -

~o 1 ,a'

,# ,p.

t

%1 l y*

I, i l

.i -

b - I

/

ao C.9 *-',: .

m -

/

m -

E o, =

g :

6 I m;= , .

.: J

. I

%.1  : I

~

b-l 1

- i Y

g i

-1 s

,0 . . . . . . . . . . .

- i i i . .

.is' iiiii *

. 10* 10 10 10' TIME, MIN FIGURE 7.30. ACCUMUt.ATED LEAKED NUCLIDES, AB-c 1- ; n , yy, 24, -.w,x; --r.--- y. : - m . _7---=---- -- -

a$

1 i

7-56.

l 'o

.7 HR. -

h',5h

~ ~

l -

14. 3 HR . -------- --

,Dw h.

y 20. 0 HR. -- -- - ,

N g-4=

, s a

i..*, -

Ch //.\ '

. i i.

N e S::

/

,/ / \. \,

\ , .

l :/  : i

', t z . ,i :f i '

O - ,:  : ,

m i ,i s

s ,/ !/ ',

c -

i s ,i s i z: l :i i i

w  :

,i i

i, o

\. ,'.

8-o -

f-ll

-i i

,:  : i, w - * \ i 1 '

+

l'a.

x~;

l ,

ll!

\\

1

\

- - 1.I ,i c  : ll

\: i i l i ,

- .  : i ,

. l I .i s.o:

- \

3- -

i

)f .n g i.

i.- . .o. . . . . ...., .

.......* 10' 10 10" 10 10 i s RADIUS, MICRONS

. .I 41 FIGURE 7.31. PARTICLE ST.ZE DISTRIBUTION, AB-c l.

g I'd Q3' ' H J'- = <C * == *s - *s'aQ3 ~ ~ *~; f f '= i % - * * ~r~ek ;% ,-zn W t!

< mM *w?pyk~ W An

• 6 '[QQ Wy: }y At'<m3.',

S

~ "

t 7-57 TM.B'-s, the containment fails at a time of 738 minutes with a contain-j ment pressure of 55 psi. Unlike in the AB accident sequences where fission products have a minimum retention in the primary system due to a j!

relatively short pathway, 90 percent of the Cs! and Cs0H core inventories are retained in the TM.B' seguence as discussed already in Section 7.2.

.I TM.8-4 ,

l ,

j TM.B'-4 , represents a scquence in which the containment fails shortly after the reacter vessel is ruptured. Connared with the TM.B'-c sequence, the sourca particulate released from the primary system and

[ from the core-concrete interaction tend to escape the containment before being subjected to significant natural retention. No engineered safety features are assumed available in this sequence.

The calculated results for the airborne mass and accumulated leaked mass for the containment for tha TMLB'-6 , sequence are shown in Figures 7.32 and 7.33. The locational distribution of species after a sufficiently long period of time elapsed is shown in Table 7.16. The calculated core inventory fractions shown in the table are comparable to

, -i those obtained for the A8 y sequence shown in Table 7.14 but somewhat

, lower than those for the AB-6 sequence shown in Table 7.12.

l  ; . TM.B ' -s l .

l:i The calculated results for the airborne mass and the accumulated

,q 1eaked mass in the containment for the TM.B'-c sequence are shown in l . Figures 7.34 and 7.35. In th's accident sequence the containment melt- '

Jl through occurs at a time of 738 minutes. The containment pressure reaches I k' about 55 psi at this time As traditionally assumed, no engineered safety l,'

features such as sprays are operating due to the loss of the off-site I; power source. Details on the accident event timetable and the time-dependent pressure and temperature conditions can be referred to in ;_

J Section 6.1. -

, t-ll3 Figure 7.34 shows the mass of the total airborne particulates l'

' ?

as a function of time. Again, the first peak of the curve represents I r.

(

I

's 1

i _ sa a = =2x = n . m m

• f W ~: T ^' r W T ~ "'*"" ' " = ~= - c+ d1 o

,1 7-58 i b-

. _' AIRBORNE

.i . LERKED - -

e

- i,

.j l

4 1 .

r' u, .

I i CI . f t,

% i

<n

<n .

Eb_

d.

s

. f*-

l

, o . .

H l - .

i d

1 1

! 4

. . n,

-:.a4 rr

, a. - b-- , ,

.=. , , ,

t 10* 10 10' TIME, MIN.

e.

r .

. 1.< .

( .

FIGURE 7.32. AIRBORNE AND LEAKED MASSES, THLB'-6 e

. L, 1

11 o,

4-t._r,_--.,..,...n..-.n. ww - ,w ..,_.,wn -

,,ms-- , g.y; ww- -.g---sa..

r-_- .I; s

l l

1

~

7-59 i

=

C CSI - i

'  : CSOH ._

l TE OTHERS -----------

1 -

,i

.i-s' . '

i

' t p r

t

. b._.

,' i

.; l

- p i -

1 I.

1 , i l

.l I .

cs .

v m

~

m Eb_ - .

a g :- .

n ,. .__ -._ . ._ _

a _

w _

4 4

c i

~: _

C-

_ .I b- -

a 10'

g TIME,' MIN FIGURE 7.33. ACCUNULATED LEAKED MUCLIDES, TMLB'-oe 4

P

' * *e 96 s )^ p M4 'A & To i ~ - _ ,,

"'lW~'v3= rsm.n-~m 1,.m,m f'

• '}-  : a'.'. . .-.? vmL-_ __ - -- w.

_ _,cu ;J

- - ~ - - - - - . _ _ . ____-___

. 2 i,

i 7-60 TABLE 7.16 LOCATIONAL DISTRIBUTION OF SPECIES AFTER i

ACCIDENT IS COMPLETED, TNLB'd, t

4 i~

a. l Fraction of Core Inventory Species e

RCS Containment Environment

Csl 0.90 5.6 x 10-2 4,4 x 10-2 ,

- Cs0H 0.91 5.2 x 10-2 t 4.3 x 10-2 Te 0.30 0.16 0.-11 ,

I i1 i i

d l

TABLE 7.17. LOCATIONAL DISTRIBUTION OF SPECIES AFTER  !

a ACCIDENT IS COMPLETED, TMLB'-c 1

Fraction of Core Inventory Species RCS Containiaent Environment l

l Cs! 0.90 9.8 x 10-2 2.6 x 10-3 ,

Cs0H 0.91 9.5 x 10-2 3 x 10-4 Te 0.30 0.19 7.9 x 10-2 b -

E ll

, h s"

11  !

2 a . - x +mn ~~ u -w -~~w _ - - - _ _ _ _ - _ - _ _ - - - - _ _

+~ ~ ~ ~ ~ - - - ~ < = ~ ~ - - mw '

1 . .

i 7-61 -

s b

RIRBORNE

. LERKED - -

~ ._

. r a

b_ .

o cm .

u,- -

u, q E. .

j -

d s .

\

C3 -

s

.: I y_ .

e

. b- . . . . . . . . . . . . . .

....' 10,

, 10' 10

< TIME,. MIN FIGURE 7.34. AIRBORNE AND LEAKED MASSES, TNLB'-c

g " w { ', Y. N ' "'

e"" * * * ' " -' - -

A

. -, , ng,z;,,_,_,g g .

" _'*- - - - - _~- ~

. - . . . . _ . - . . . . . _ _ _ __ au

i i

t

. 7-62 i b i - :

CSI l -

CSCH . .-

TE -.

OTHERS -----------

(_____

b.: I 1

s. .

. e a ,

+ ,

. . i 1

o.

.o - i

.. i j: = =

n 4

cs .

u- -

- i u)  :

i Eo':

. 1 g _-

e s  : I

-,.. o

. n -

1 b-:  ;

4 - r c.

t

b. -

r,  :<  :

c .

t y, ,

y,a s .

?-

j, -

.is

b. '

't 10 10 10'

.t..

TIME, MIN cs1 b

3. ,

L f.; FIGURE 7.35. ACCUMULATED LEAKED NUCLIDES, TMLB'-c h j r ,

1 lG7 i+4 r -

., =.

.=_- -.

= - -

,._ a.

- .-  :. = .. - . .- . . - - --- - - - - -

- -, 3 7 63 '

i i

the melt release prior to the bottom head failure and the subsequent increase is due to the release from the VANESA calculations during the core-concrete interaction. It should be noted that in this accident sequence, a water or mass of 1.117 x 105 kg at 326 K is assumed present at the time 'of head failure (t = 157 min) and it is further assumed that l the mass ovaporates over a time period of approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The j presence of water, of course, delays the vaporization release substan- .

I tially. Figure 7.35 shows the accumulated leaked mass for CsI, Cs0H, and Te as a function of time.

' 1

7.3.3 V Sequence -

1 l

This sequence represents one of the accidents that permits a j . maximum amount of fission products to escape to the atmosphere. In the i event check valves that provide a barrier between the RCS piping and the low pressure ECC system should fail, the fission products can bypass the -

l containment safety feature and be released to the auxiliary building.

If the auxiliary building is treated as a containment for modeling purposes, calculational methods identical to those used for the other accident sequences can be applied. Results calculated with the TRAP-El.T code were, used for the melt release source for NAUA containment calculations. In treating the vaporization release, it was assumed that a-no attenuation of particulates occurs in their passage from the reactor cavity through the primary system. This simplification is expected to overestimate the amount of fission products released to the environment.

To simulate this sequence more rigorously, it may be necessary to use a multi-volume approach to assess the effects of retention of particulate material in the primary system for the vaporization release term. Two separate TRAP-MEl.T calculatiens will be necessary in that approach.

The fractions of core inventory of various species released to

. the atmosphere as' predicted by the NAUA cede are listed in Table 7.18. -

As expected, the calculated results show that this accident sequence still remains as one 'of the most ig ortant sequences in terms of total 1

release with the fractions being about 40 percent of the Cs and I inven- 1 tories. The airborne mass, accumulated leaked mass of each species, and  !

. i big.g;_ZaF[ min * *% N.7IitUEI. E ~'" 2-

~ 52 * -I (

l

~!

l 7-64 ei TABLE 7.18 LOCATIONAL DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED. Y Fraction of Core Inventory Species RCS Containment Safeguard Building Environment

\

i Cs! 0.50 1.8 x 10-2 6.9 x 10-2 o,41 l i

Cs0H 0.51 1.7 x 10-2 7,1 x 10-2 o,40 j i

Te 0.13 0.71 4.4 x 10-2 0.12

,. 'l -

i. ' .

I

' f

.i, 96 4.

-t g

'e z

.y.

5

{%

l4 h' I,

[,

i: .

e d e ,

~ ~ - - ' - =' = - - -

L., . . _ . _ _ - . ..--w. vm ~nse:- ~xt -

2~--~~m m)

7-65 I

the size distribution in the sa'1! guard building are shown in Figures 7.36 through 3.38, respectively.

7.3.4 S20 Secuence l

The S2 0 sequence is a small pipe break accident as already discussed. The containment failure mode examined was y and c. Unlike the A8 and TM.B' accident sequences, containment spray systems operate j during this sequence in order to condense steam and to reduce the contain-

,?] ment pressure. Since the original MAUA c e did not have any provision for engineered safeguard, calculations c:e made by adding the removal -

mechanism of aerosol particles due to spraying. Details on modification of the NAUA code have already been discussed in Chapter 5 of Volume I.

j In general, spraying systems are highly effectivo in removing aerosol particles, and it is expected that the spraying mechanism will dominate all- other natural retention mechanisms in the case of 520 sequence. -

50v2 The major accident events in the, containment for this sequence j were already discussed in Chapter 6. The spraying systems were assumed to cease operating at the time the containment failure takes place.

,g Figure 7.39 shows the time-dependent airborne mass for the y j containment failure mode. In Figure 7.39, it is seen as expected that

,, the airborne mass remains very low during the melt release period due to the high removal effectiveness of spraying coupled with the relatively '

l high retention of particulates in the primary system. It should be noted 4

that in the 5 02 calculation shown in Figure 7.39, a drop size of 400 um f was used. Figure 7.40 is the leaked mass. Table 7.19 shows the fraction -

Q .,

of core inventory leaked to the environment. '

's  !

-5 l N . i p

r 4

• l

" ~ ~ ~

.N zg-n.-

e g s ,n u . Q--

r

= 9 W Y~ ?'.d - - . ' W N " h* ?:" N ***

. ~ - - - - - - - - - - -

~ ~ N XE* +N

_ _ _ _ _ _ _ _ - _ - - - - . __ f

.I i

i 7-66

!-l. 4 l

i ..I i

D .

RIRBORNE LERKED - -

? - ._. . . . .__

t  ;

t .

7

./ l 3

t 1 -

xi

,- f.

") b_ 1 - .

I cs -

)w -

i i

A w

- f

.d e V

g- .i t,

+ e u

c d-t -+

1 i

o .

i E-* t i

O b.

d.

.I i' r i n -

j

' l

o

l

.e- g- , , ,

lj 4*10 i

10 g.10' p TIME, MIN I IJ i

• . l

! i.

FIEJRE 7.36. AIRBORNE AND LEAKED MASSES. V (Safeguard Building)

I V,

i t '_ . . . . -* - >. - "- n..-- s %.M,%i F'? T*I

\ c :. w p g E*r s 7^,Ep5M) * , ye. **e ' ' ' . '

_._._~.---Vr --- ,

- - " ~

- s .

I

. 7-67 .

1 CSI i . .-

b_ .CS..OH -

~! 5 -

4 OTH E R S , _;----------- ----__._________________-------------------

I

. o o

l b.1: ,

,s .------. . . . . . . .-

, t .

i

,1 r ,

l . o ,

O_

• _

"i .

s'

. ,r .

o r . . -

l '

b. ,. ,.- - - . . . _ . _ . _ .

9 s . * .

~

tn .

tn .

~i .

s .

• H -

O s -

O.

b._:

. s 7

C. :

- I T

~

C. : .

4 1o' ia' 9 10'.

TIME, MIN FIGURE 7.37. ACCUMULATED LEAKED NUCLIDES, V (Safeguard Building) t I e h* O

~

T

• f *. * -' '

'I d'. N ~ ?s

, . . r.v'we . . _ t .- ,,,.3 Iy..-...-m..

.~%. -

w r..- u -- &..**Q

_ . . . . .*. .l1_ W*%

, . .. . Ma5f*

- - - *s . . _ . - -

a AW"*R'WR . ..

~

,..e.,

4 . -

l.

t, i

i, .

. 7-68

>o

& ~

1.0 HR.

HR. . .-

3. . 5.

l 9 . 2 HR . ------- ----

Y l o.

14. 0 HR . ---.----

~

f l

o I

a g

t CD et '

Ea.

w .

Ns s

c ..

M ,

g s -

f '

3 ". m, ' ,

\

s,

, g

, s a . . ,

.q . a -

o ,

,/  :

. . ~ . . .,

i, s

. t

.  : . s

,,t . .

g i g - s o l .

c .

s

~ ,  : .

\

'c-

~4 a- ,,,

l

/

l 1

o  : .

l.

l l \: s

~

o  :  :

o  :

,,o  :  :.

- . rl l \:

' ,o  :

t .

I o ,  :

t  :

. o

.c._ . , "' , . . . ....

10 10 10 2=10' RADIUS, MICRONS l "n I e .

~

FIGURE 7.38. PARTICI.E SIZE DISTRIBUTION, V (Safeguard Building)

-n .

a

?

p x; , r3wem n.-._.cw , v- g c.2.3 z 2., m.2..._ . . r ...___ _ _ m.,a...,,._,,,,_,.. .- _

,. . ., . . l

~

.\

t e ~ ,

i 7-69 i

i l l

l l TABLE 7.19 LOCATIONAL DISTRIBUTION AFTER THE

] CALCULATION IS COMPLETED, S 2 0-y 4

~1 Fraction of Core Inventory _.

< Species RCS Containment Environment i

, Cs! 0.48 0.52 4.2 x 10-5

> Cs0H 0.57 0.43 S.4 x 10-5

! Te 0.91 5.8 x 10-2 3.3 x 10-2 s.-

9

~

.j .

y l

l

'l b

I

t. < .

i 4 '

l f .

• 4 I i ' 4 i

1 *

' gy ,. 2,, g' _

~

+ '

N ,I'

}

I -

J i

7-70

?

1 1

• g l E. RIRBORNE

1 LERKED - -

F,,j

.)

. b._.

.1 -

1 . -

n e.a

~.-

cs i m'o_

.. m ..

E~ . . . ./

t 3

a w c s

i

\ o I w

M

.i

/, .

"o._.

t I:

i 9

s l' -

l .:.i. .

)

. r.

et b - ' . . . . . . .

u 1(f 10' 2*10' I *i .

TIME, MIN l'!

l .4 FIGURE 7.39. AIRBORNE ANG LEAKED MASSES, S 2 0-y

,i

.u ju .

~

g- ,

• ~ T_11l,. ' ' ' :. Zi:

~ " * ~ ~ ' ' '

7- 71 l 1

l 1

CSI . l CSCH - -

7 _ _ _

i JTHERS -----------  ! ~ ,,,)

t_

I - ,

1 . ,

,I

s. ,

I hf.

/

l

\

lt

. l 1

cm )

w _

E ,-e -: .

~

~

al w -

r o

H -

l b_: 1 l

l b_n. -. . . . . . . .

1, $g 2,;g TIME, MIN FIGURE 7.40. ACCUMULATED LEAKED NUCLIDES, 2S 0 y 4

{,

O Wb -C -

--m -' * *

% T m, A # O Mh .* #W F N ' F ". -

7"$ '

J* 4' - - - - , . . , >

.___-,. . _ _= _ _ . .

l .

.i i

q 7-72 i

! S D-c 2

1 This accident sequence is similar to the S 0-Y 2 sequence except containment failure does not occur for an extended period of time and that containment sprays contira e to operate throughout all the accident

. events. Due to availability of water in the reactor cavity during the core-concrete interaction, the VANESA calculation resul's taking account into the presence of water were utilized as the source term during the vaporization scheme. Figures 7.41 and 7.42 are the airborne mass and accumulated leaked mass of species for the S 20-c sequence and Table 7.20

, presents the locational distribution of the CsI, Cs0H, and Te species.

? .

w 7.3.5 General Observations The results of'nine NAUA calculations have been presented.

i #

Based on results of the present calculations, the following observations can be made on transport and retention of fission products in.the contain-ment. The amount of particulates escaping the containment is largely dependent upon timings of the two releases from fuel (melt and vaporiza-tion) and upon containment failure time. As demonstrated in calculations for the sequences involving an early containment failure (AB-8, AB-Y,

.( TM.8-6,, and V), these accidents lead to the highest release fractions.

It is worthwhile to note that the retention of fission products *

, that takes place in the flow path of the primary system for sequences

'[.

involving early containment failure reduces the source term to the con-

, tainment and will directly reduce the amount of fission products that escape the containment. However, the retention, effects of the primary

, system on the release to the environment would not be so apparent for 4

c accidents involving a late containment failure mode. In such accident g sequences, the fission products that are not retained in the primary

? system will be retained in the containment due to prolonged residence $

a);

times. One can then conclude that alteration of the melt release rate

'g would not significantly change the final release fraction in sequences V of late containment failures.

!_n -

I -

.$is .

t. p .. -_ _

~_ p _ } ~ ; _ _- ==- -[-  ;- + -- _ _ : = _. __ . . :2.-__::-

.. s 7-73 RIRBORNE M -

LERKED - -

1, -

l- .

i e

l e -

t 6

11 -

's' C

s - .

i c.s -

1 w

/

k. - ./.

H )

o- - .\

!l M 'o _

-: p i \ . . _ _/

i t

.d' '!

I 7

o_

1

~ ,

- f j ,

,O._

I

- 1 t .

. j i a- i 11 if 2a 10' 1 l TIME, MIN t'

tt .

is i

FIGURE 7.41. AIRBORNE AND LEAKED MASSES. S2 0-c

1 s

l 1

4 1

.w1

.g

,g. m y . . ,..r-~ .

,,~k' s ,

~#

eM 4'

1. ~*

_.-m

~'

,7* aM. wwrqa f & *d-J E AN==NEEY lI .

4 4 7-74 1

CSI ..

i '____

CSCH - -

'o , ;-; . _ . . - - l

~i . OTHERS ----------- ,'

'o ,.. l

= ,

i .

.a ,

= ,

.f-- . _ . .

r, ,i

• a ~

~1 t

1 cn (n -*,=

c  :

_.t n  : '

)

.a - '

co H1

. o 5

.i .

H  :

i ,o ,- e

-r '

-1

I

. f r'a '

-1 .

i

. . i 1

_=

m w . -

-:- 10' 10' 2*10' 2 TIME, MIN 1;

c. .

'1

', FIGURE 7.42. ACCUMULATED LEAKED NUCLIDES, 25 0-c 4

f ,

i a.-v :y_ .

__  ;]l-

.g , Q ..

7-75 a

TABLE 7.20 LOCATIONAL DISTRIBUTION AFTER THE CALCULATION

.. IS COMPLETED, S2 0-c 1

)

Fraction of Core Inventory Species

}" RCS Containment Environment

.' l l.

! CsI 0.74 0.26 1.5 x 10-8,  ;

Cs0H 0.76 0.24 1.4 x 10-8 l

4

. Te 0.69 0.15 7.7 x 10-8 i

4 j .

V e

g 9 m

i e

1 t'

4 i "

~} .

'l

.g .

4

, l-4 ,!

i'

)

, ,w .. .. , _ ~ c... . v . . . - ., -

- m ,-- - ~ ~ .. - - - - - ' -. - ~ n, . c .x.

. };"j: ;_

'l _ _ _ . _

. . . s -'.

.~,:. _4

~, -.

W#6 v

i

." 4 6

h i

9 f

D 4

i.

I e

\ .

4 i

i 1

.e  ;

4 . <

M

)

~

+

. .d

-4 e

9 I

i e

0 6 J

4 i  !

.j .

'. s .

t B

I

-6 O

e l .

t'

• 4 1

4 9

..=.--w.e--=~- ,==---.-% - ~ ~ . - ,-. -- ,__,w 2, ..gunag. .w #:,e

_ 5ii

)* .Q _4 . . . ,% w _.. .%, .c._,_ ,_ _ . . ,. 3,. . ---------2_

_s. - r

_.-- m

'j