ML20011A427
| ML20011A427 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 04/30/1973 |
| From: | GULF UNITED NUCLEAR FUELS CORP. |
| To: | NRC |
| Shared Package | |
| ML20011A420 | List: |
| References | |
| SS-1075, SS-1075-R01, SS-1075-R1, NUDOCS 8110130338 | |
| Download: ML20011A427 (34) | |
Text
.-. !
"fE SS-1075 (Rsv.1)
RESPONSE TO QUESTIONS BY AEC/DL WITH REGARD TO GULF UNITED REPORT SS-942, TECHNICAL EVALUATION, ADEQUACY OF LA CROSSE BOILING WATER EMERGENCY CORE COOLING SYSTEM April 30,1973 Revised November 15, 1973 Work Performed on Gulf United Project 2348 Contract AT(11-1)-1756 Chicago Oper,'tions Office of the United States Atomic Energy Commission GULF UNITED NUCLEAR FUELS CORPORATION s
Elmsford, New York 8110130338 G10929 DR ADOCK 05000
~
r' TABLE OF CONTENTS 1
Potential of LACBWR Fuel Cladding for Ballooning and Embrittlement and the Effect on a LOCA........
1 2.
Heat of Reaction for Steam-Cladding Reaction.......
5 3.
Stainless Steel Steam-Reaction Kinetics 5
4.
Verification of Radiatiu Interchange Factors.......
6 5.
Reanalysis of Fuel Assembly Heatup............
6
(
6.
Low Pressure Core Spray as Backup Only.........
7 7.
Leak Rates into Containment Building 8
8.
Explanaticns for Shape of RELAP Curves During Blowdown
......................8 REFERENCES 10 APPENDIX A - METAL-WATER REACTION CAL CULATIONS................. A -1 l
l
>1 11
f TABLES 1.
Results of Phase I and Phase II Burst Tests at 650 F 11 2.
Radiation Interchange Factors with Dry Shroud......... 12 3.
Radiation Interchange Factors with Wet Shroud.........
13 4.
Leak Rates, Recirculation Line, Double-Ended Break -
Discharge Side of Pump 14 5.
Leak Rates, Recirculation Line, Double-Ended Break -
15 Suction Side of Pump 2
6.
Leak Rates, Recirculation Line,1.0 ft Single-Ended Break..
16 2
7.
Leak Rates, Recirculation Line, 0.5 ft Single-Ended Break 17 2
8.
Leak Rates, Recirculation Line, 0.25 ft Single-Ended Break :
18 9.
Leak Rates, Recirculation Line, 0.1 ft: Single-Ended Break.. 19 2
10.
Leak Rates, Recirculation Line, 0.05 ft Single-Ended Break.. 20 11.
Leak Rates, Steam Line Break 21 A1.
Noninal Composition of Type 348H Ltainless Steel....... A-2 A2.
Composition and Atomic Weights Used in Calculating Energy Release
.........................A-2 A3.
Oxides Assumed in Calculating Energy Release......... A-3 A 4.
Calculated Molar Energy Release Accompanying Selected Metal-Steam Reactions...................... A -4 A S.
Energy Release Accompanying Stainless Steel-Steam Reactions Cal /g Stainless Steel
......................A-5 FIGURES 1.
Effect of Test Te.coerature on the Ductility of Irradirted Type 348 Stainless Steelin the Annealed CCadition -
Irradiation Temperature ~290* C 22 2
23 2.
Clad Temperatures, Intermediate Size Break - 0.2 to 0.75 ft 3.
Energy Release in Stainless Steel-Steam Reaction 24 25 4.
Reaction Rate of Type 304 Stainless Steel with Steam......
iii 9
f f
The following discussions are in response ce questions raised by AEC/DL at a meeting with DPC and Gulf Unite.i personnel ac Bethesda, Maryland, on February 8,1973 concernt".g Gulf United's Technical Evaluailon of the LACBWR Emergency Core Cooling System, submitted to you on May 31, 1973 as Report SS-942.1 1.
Potential of LACBWR Fuel Cladding for Ballooning and Embrittlement and the Effect of a LOCA
~
The behavior of the I.ACBWR clad material in the event of a LOCA is determined by its inherent ductility at the start of life in the reactor, '
by changes in its ductility due to radiation damage mechanisms during operation, and by further modifications accompanying the high temp-eratures during a LOCA.
A review of the pertinent properties of the LACBWR cladding materials and the changes in those properties during' operation follows.
From this review an assessment of the behavior of the cladding during a LOCA is made.
a.
Ductility During Normal Reactor Operation The LACBWR fuel rod cladding material is Type 34811 itain-less steel.which is a high carbon, stabilized austenitic stainless steel.
As such, annealed material is characterized by high ductility, excellent corrosion resistance, and good high temperature strength. The austenitic stainless steels cannot be hardened by heat treatment but they are strengthened by cold work. This strengthening is accompanied by a reduction in ductility.2,3 "he effects of co'.d work can be removed by recovery heat treatment during which strength levo are reduced over a narrow temperature range between 650* to 750* C (1200 to 1380 F) while 4
1
e O
ductility is recovered over the broader temoe;.uure rance of 300* to 3
760* C (570* to 1380 F).3 Fast neutron irradiation causes changes in I_
strength and ductility similar to those casued by cold work and these can be similarly removed by recovery heat treatment.8 ' Thermal neutrons also cause transmutations which yield izapurity atoms. Such impurity atoms may cause porosity or solid solution or precipitation hardening effects, but in BWR and PWR cladding they have not proven to reduce ductility to excessively low levels during normal burnup.
Actual experience in tests of irradiated cladding material from the Yankee reactor (Type 348 stabless stee') have shown that despite large increases in strength and decreases in ductility during irradiation "the material displayed generally ductile behavior."' It should be r.oted that this ductile behavior prevails in burst tests even when tensile ductility (as measured by ?o elongatio6) is low. This may be explained on the basis of the large values of reduction in area in tensile tests despite relatively low values of % elongation (see Fig.1, for example}.
Burst tests of VBWR cladding (Type 304 stainless steel) showed "a large degree of ductility stfil present in tubing that has been 21 nyt (>1 Mev). "5 This further substantines the irradiated to 2 x 10
~
experience with Yankee clad.
It is clear that irradiated Type 348 stainless steel retains 21 satisfactory ductility after neutron doses to 2 x 10 nyt (>l Mev). The chemistry (including carbon content) of the LACBWR cladding materini meets both Type 348 and Type 348H stainless steel specifications.8 It will therefore show similarly satisfactory ductility after exposure to 7
this neutron dose level. Since there is evidence that hardness and 21 burst strength both satunte at neutron dose levels beyond 3 x 10 nyt
(>l Met higher exposures will not have any further deleterious effect.
b.
Effect of 1OCA Cwditions on the Ductility of Type 348H' Stainless Steel At the start of life in the reactoi, the elevated temperature burst behavior of the LACBWR cladding will be that of unirradiated material. Tests of unirradiated Yankee cladding (Type 348 stainless steel) at 650* F resulted in a maximum diametral change of 2Wo k.
2
i-(Reference 4)(see Table 1*). LACBWR cladding should display similar behavior. This level of change will result in a reduction of flow area g-of less than 40% which will not restrict cooling.
4 Relatively little high temperature data exist.for Type 548H stainless steel and at temperatures from 1600* to 2400* F there are none.
4
{
In this temperature range, there are data for Type 304, Type 316, and i
Type 347 stainless steels.
4 Type 348H stainless steel is a high carbon modification of Type 347 stainless steel in which the stabilizing agent (columbium +
tantalum) has a limit set on the tantalum content at 0.1% Since these two elements are tiqually effective as stable carbide formers, the i
mechanical properties of the two grades are very similar. The carbon f
content of Type H material he.s a perynissible range of 0.04 o' to 0.10"o.
C Material certifications for the LACBWR Core I fuel cladding show that both Type 348 and Type 348H stainless steel.. specification requirements i
are met.8 Therefore, the Type 347 stainless steel data safely provide minimum mechanical property values. Figs. 7,1 and 7.2 of Reference 1 sumr'arize the available high temperature mechanical property data r
for Type 304, Type 347, Type 347H, and Type 348 stainless steels and indicate that appreciable strength and ductility remain (~3000 psi and
, (.
30c elongation, respectively) up to 2400* F.
o Of more direct interest is the behavior o.' the cladding material under the conditions prevailing in a LOCA. Under LOCA conditions, a rapid rise in temperature of the fuel rods will occur, with l
the four central rods of each element reaching the highest temperature.
j During this heatup period, temperatures in the range of 1700* to 2300* F l
will be reached. This temperature range covers both stress relief ~and solution anneal conditions tor Type 348H stainless steel. For thin sections, recommended time at annealing temperature is normally 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per inch of thickness. The LACBWR fuel rod cladding is 0.020 inch for which ne annealing time is between 1 and 2 minutes. Since l
the elapsed time at ur above 1600* F is 4 to 5 minutes (see Fig. 2), it can be assumed that during a LOCA much of the irradiation inducted i
property changes will anneal out. However, as shown in Item la above, s' fficient ductility remains even in highly irradiated Type 348H stainless u
I j
- All figures and tables appear at the end of this report.
1 k
l 3
.. _. _ ~, _ _.
steel to obviate brittle failure. Furthermore, circumferential elongation in burst tests of annealed material demonstrate that ballooning is not a
(..
problem.
The internal pressure resulting from tia combination of fission gas buildup and high fuel rod temperature prior to faitiation of the ECCS will subject the cladding to internal stresses which are best simulated in the closed end burst test, j
Recent tests at ORNLa on Type 316 stainless steel fuel e adding covered the following conditions:
Heating rate, *F/sec 26 to 412 Maximum pressure, psig 0 to 1810 Rupture temperature, *F 1880 to 2200
'In there tests, the ruptures were indicative of ductile behavior with a maximum circumferential expansion of 24% This behavbr is similar to that at lower temperatures as reported above.
Although these data are for Type 316 9tainless steel, the mechanical properties of this alby are sufficiently similar to those of Type 347 and Type 348 stainless steels, that they may be assumed to represent the behavior of the LACBWR cladding under similar conditions.
The anticipated conditions for LACBWR fuel rods during a LOCA are:
Heating rate, *F/sec up to 100 Maximum clad temperature, *F 2220*
Maximum internal pressure, psig (2 '7)
Therefore, the ORNL data will apply to LACBWR.
On the basis of the foregomg information, it can be concluded that at the start of life the ductility of the Type 348H stainless steel LACBWR cladding is at a level tla.t prc 'ludes appreciable ballooning should a LOCA occur. During s.bsequent accumulated radiation c
damage, there is a loss of ducti?.ity, but even at saturation there is still cufficient ductility remaining to insure ductile failures in the event of a LOCA. Furthermore; the high temperaturer., accompanying a LOCA will probably anneal cui much of the radiation damage, so that the material will exhibit adequate, but not excessive ductility.
4
2.
Heat of Reaction for Steam-Cladding Reaction 1
d' Considering the major elements in the LACBWR clad (Type 348 stainless steel) instead of only iron, results in higher values for the enthalpy change accompanying the reaction with steam (thnn the i
J29 cal /g quoted in Appendh D of Reference 1). Values as high as 264.8 cal /g stainless steel occur at 1832* F (1273* K), the lower temperature limit for significant reaction rates. Supporting calculations for these data appear in Appendix A. A curve of the energy release versus temperature, based on these calculations, is shown in Fig..l.
3.
Stainless Steel Steam-Reaction Kinetics The reaction rate actually used in the calculations in Appendix D of 1
SS-942 vas the linear rate obtained by takir.g the intersection of the 1200' C curve of Fig. 7.3 with the 6-minute time level to obtain the weight gain aftu 6 minutes. Division by 6 (the total time it takes to reach maximum clad temperah re) resulted in a reaction rate of 2.5 x 10-3 'C;)/
g 2
cm /mt1. However, the curves in Fig. 7.3 were obtained by measuring ve%ht gains experienced by samples held at fixed temperatures for fixed periods of tin c In the case of a LOCA, the clad temperature is diffelent J~ each rod of the fuel assembly and also changes with time.
For reanalysis of the LOCA, described in Item 5 below, the reaction rate was taken to be a function of clad temperature as shown in Fig. 4.
This curve was obtained from the steam corrosion data of Fig. 7.3 of Reference 1, by calculating the reaction rate for each temperature at 4 minutes, the time interval between the beginning of the reaction at a clad temperature of 1832 F and the time the maximum clad temperature i
l': reached. For example, for a temperature of 1150' C (2102* F), Fig. 7.3 2
of Reference 1 shows a weight gain of 1.2 mg/cm during the time interval between 3 and 4 minutes. The reaction rate for 2102* F at 4 minutes is then app:.oximately, 2
1.2 mg/cm 2
.0012 g(O )/cm / min
=
2 1 min x 1000 mg/g Since the reaction rate decreases with time, taking the linear rate, valid up to a time of ~20 minutes, is conservative. The decrease in reaction rate with time is due to the formativi. of an oxide skin which serves as a barrier to further oxidation.
'A 5
4.
Verification of Radiation Interchange Factors The radiation interchange factors,h*, used in the fuel assembly i
'f heatup analysis shown in Tables 5.12 and 5.13 of Reference I were
~
calculated by Gulf General Atomic (GGA) on the basis of an analysis described in Appendix C of Reference 1. In acco2 dance with DRL's request, Gulf United has now calculated these radiation interchange factors independently, and arrived at similar values, shown in Tables 2 and 3. These new factors were subsequently used as input for the reanalysis of the LCCA, desezibed in Item 5 below. These radiation interchange factors were obtained b,, first determining the view factors between individual fuel rods and between the fuel rods and the shroud frem the rod bundle geometry and then solving a series of c'.multaneous equations with the view factors and the emissivities as input, by means of a matrix invarsion routin9 on a digital computer to obtain the radiation interchange betors,Nj.
A recalculation of the LOCA with an intermediate size break (Table 6.2 of Reference 1) gave a peak clad temperature of 2204" F with the new radiation is erchange factors, as compared to 2186* F with the original factors.
5.
Reanalysis of Fuel Assembly Heatup The fuel assembly heatup calculations described in Sections 5 and 6 of Reference 1 were repeated for an intermediate size break (Table 6.2 of Reference 1) as this size break led to the highest clad temperature following a LOCA. All parameters were kept the same except for the following:
a.
The new radiation interchange factors calculated by Gulf United (see Item 4 above) were substituted in place of the original values.
b.
Heat generation in the cladding was added to account for the increase in estimated heat release during a reaction between cladding and steam, using the data of Items 2 and 3 above.
For each rod of the fuel assembly, the reaction rate in the clad was taken from Fig. 4 according to its temperature at the time when l
the center rod reached its peak cladding temperature. The reaction rate was then multiplied by the energy release from Fig. 3 to obtain i
6
the heat generation in the clad. For example, for rod No. 2 (Fig 5.24 of Reference 1) the clad temperature at 6.5 minutes is 2185* F. From j-2 Fig. 4, the reaction rate is 0.0015 g(0 )/cm / min and from Fig. 3 the 2
energy release is 264.8 cal /g.
The heat generation rate in the metal then is:
g@2) cal cuf -min g(0 ) x 264.8 x 96.36 "it q' = 1.5 x 10-3 x 2.606 3
g x 6.978 x 10-8
= 0.00675 kw/ft cal The heat of reaction adds about 4?o tc the decay heat of the center rod.
The results of the reanalysis are shown in Fig. 2. Up to the time the shroud wets, cladding temperatures are nearly the same as for the original analysis, leading to the same shroud wetting time. The maxi-mum cla:1 temperature, however, increases from 2186 '.o 2219 F.
6.
Low Pressure Core Spray as Backup Only
(
In the LOCA calculations, credit for the action of the Low Pressure Core Spray (LPCS) system was taken only in the analysis of a large break (Table 6.1 of Reference 1) where the heat transfer coefficient from shroud to spray before shroud wetting was taken to be 20 instead 2
of 10 Btu /hr-ft - F, on the assumption that the spray impinged on both inside and outside shrc,td surf aces due to the large flow rate of the LPCS.
This resulted in a peak clad temperature of 2170* F for the large break.
In the absence of LPCS action, the input conditions and peak clad temperature for the largt break will then be the same as for the inter-mediate size break (Table 6.2 and Fig. 6.2 of Reference 1), 2186* F which is governing as to maximum clad temperature for the entire spectrum of break sizes. Ignoring the effects of the LPCS, therciore, imposes no penalty in terms of maximum clad temperatures, and the Al?C criteria can still be met with the high pressure core spray system alone, leaving the LPCS system as a backup only.
9 7
s
7.
Leak Rates into Containment Building The leak rates from a break in the primary piping of the LACBWR plant to the containment building as a function of time after initiation of a LOCA are given in Tables 4 through 11 for a range of break sizes.
This information is taken from RELAP-3 computer output, generated in conjunction with calculations on the blowdown phase of the LOCA re-ported in SS-942.1 8.
Explanations for Shape of RELAP Curves During Blowdown An explanation is given here for the shape of some of the RELAP-3 curves presented in SS-942 showing core parameters during the blowdown 1
phase for singe-ended breaks. The characteristics of the double-ended break are covered in sufficient detailin SS-942 and are therefore not repeated here.
To understand the curves for ibe single-ended break, the following points should be considered:
The sequence of events following a break on the discharge a.
side of the recirculation pump has the following charac-teristics which can be 'oest seen from Fig. 5.15 of Refer-ence 1 giving core flow rates:
(1) As the break in all these cases occurs just below the core, the flov; reverses direction almcst immediately and reaches a peak in the negative (downward) direction within a short time.
l (2) Shortly after the peak negative flow has been reached, the core has emptied and the flow rate decays towards zero.
(3) At a later time, water from the downcomer and other parts of the system reaches the core on its way out, and a second downward flow peak is reached, l
I (4) When this water has emptied from the core,'the flow decays again to near zero.
b.
Another factor that should be pointed out is that the RELAP-3 2
input was modified for the calculations of the 0.25 ft and 0.5 ft break to give a more accurate value of the water 2
N, 8
Inval in the cot 2, as water 1svel was an important factor in determining the time of the HPCS signal and the mode of heat transfer in the core for the smaller breaks. As
(
a result, curves for water level and water mass in the 2
2 core for the 0.25 ft and 0.5 ft break sizes are not 2
directly comparable with those of the 1.0 and 0.1 ft breaks, although plotted on the same graph.
1 The individual f!gures of SS-942 applicable to the single-ended break are now discussed in detail:
a.
Fig. 5.11 - Core Pressura -No unusual features.
b.
Fig. 5.12 -Mass of Water in Core -The curve for the 2
0.5 ft break shcws the effects of the down-comer water entering the core at about 8 to 9 seconds. The curve for the 0.1 ft break 2
2 is below the curve for the 0.25 ft because of the difference in RELAP-3 tr.put used (see Item 8b above'.
c.
Fig. 5.13 -Heat Transfer Coefficients -The heat trans-fer coefficients drop rapidly as the water leaves the core and then recover somewhat
(
as the downcomer water enters the core.
d.
Fig. 5.14 -Mixture Level Above Core -This curve again is sensitive to the RELAP-3 input used so that no direct comparison should be made between the curves for the 2
0.1 and.25 ft break which are based on different steam separation models.
I e.
Fig. 5.15 -Core Flow - The phenomena discussed in Ite m 8a above can be seen in this figure, t
l f.
Fig. 5.16 -Core Pressure -The core pressure for a 2
0.25 ft break is shown here over an extended time period (44 seconds). No unusual features.
i 0
9
f REFERENCES 1.
Technical Evaluation, Adequacy of Lacrosse Boiling Water Reactor Emergency Core Cooung System, Gulf United SS-942 (May 31,1972).
2.
Chromium-Nickel Stainless Steel Data, INCO.
3.
Quarterly Progress Report, Metallurgy Research Operation, BNWL-20 (April - June 1965).
4.
Paxson, E. and Smellely, W. R.: Engineering and Metallurgical Evah;ation of the Yankee Core I Spent Fuel, WCAP-6084 (June 1967).
5.
Pashos, T. J., et al.: Performance Analysis of High Power Density Program Fuel Rods Irradiated in VBWR - Terminal Examination, GEAP 4926, General Electric Corp. (1968).
6.
Supplement No. I to LACBWR Refueling Plan for Fuel Cycle 2, DPC-851-58 (June 1,1973),
(
7.
Quarterly and Semi-Annual Progress Rcports on Yankee Core Evaluation PrograrA, WCAP-6056, 6064, 6070, 6078, 6080 and 6081 (1963 - 1966).
8.
Cottrell, W. B.: ORNL Nuclear Safety Research and Development Program Bimonthly Report for January - February 1972, pp. 42 - 44, ORNL-TM-3738.
9.
McAdams, W. H.: Heat Transmission, 3rd Edition, Eqs. 4-25,4-27, and 4-29, p. 74, McGraw-Hill Bouk Company, New York,1954.
10, Wilson, R. E., et al.: Isothermal, studies of the Stainless Steel -
Steam Reaction, Chemical Engineering Division Semiannual Report, July - December 1965, ANL-7125, pp.150-153.
11.
Elliot, J. F., and Gleisar, M.: Thermochemistry for Steelmaking, AISI, Addison-Wesley Publishing Co., Reading, Mass.,1960.
A 10
(
TABLE 1-RESULT 5 OF PHAS5: I AND II BURST TESTS AT ~ 650' F 8
Exposure, n/cm Burst, Pressure Equiv. Burst Max. Diam.
Fuel Rod No.
(> 1 Mev)
(Psi)
Strength (Psi)
Change (%)
9,000 69,000 27 Non-Irradiated Samples Phase I:
J3-C-f6 0.8 x 10 '
15,700 119,000 7
21 J3-NE-a6 0.7 x 10 12,100 92,000 8
21 H3-C-f6.
2.2 x 10 19,700 150,000 21 G4-C-al 3.3 x 10 17,400 13'4,000 13 21 G4-C-f 6 3.3 x 10 16,400 123,000 21 G4-SE-f5 3.2 x 10 14,800 113,000 9
21 E5-C-a6 3.1 x 10
,17,200 131,000 19 21 F4-NE-f1 3.2 x 10 16,000 122,000 9
(-
21 E5-NW-al 3.2 x 10 14,900 113,000 19
(
Phase II :
21 FS-C-al 6.0 x 10 15,900 121,000 21 21 FS-NW-al 5.9 x 10 15,800 1:;0,000 8
.j 21' F5-NW-b2 5.9 x 10 15,500 125,000 13
- Excessive tearing; no valid measuremeats possible.
11
f 1
<h TABLE 2 - RADIATION INTERCHANGE F*sCTORS,f/j, WITH DRY SHROUD ROD EMISSIVITY = 0.60, SIIROUD EMISSIVITY = 0.67 i j 2
,, 3 4
s 3
7 a
I 1
,156?
,,.?255
. c 71'3
. 542
.0443
. 112
- b
,,,n?4
,1177 1441
.:637
.13
.;658
. ?'?
2 3
. 173 713 1273
.o216 1 60 11'.3
. 37 77; 1133
.d534
.., 7 6 2
.J775
.i977
~
. 189
. 399 564
. 105 6 51 1776
.0578 5
? l
. 658
. 5F 5
. ^93
- 53-056.
267 31R5
.<??7 J533
.7 ?
. 624
. ;1M 7
.0 4%
. 223
.L'94
. 594
.2693
. 694
..?40 3W 9
. 074 1345
'J.j389
,.21' 16s
.9337
.. 7 7 t:
. 169 0
. 014
,4057
.n.36.
.:.1224
. ie 2 91
. 9 9, si 545
. iSB 1.-
n1'
. 047
.t>32
- H7 419-
. 541 5/2 145 It r?-
004
..143'
?
.p466
. !;136 6 '3 "
S'S
!?
. 01'
. 033
~ ~ ". 5 4 2
.. 55
.0291
. 074'
..i73
.t382
~
~
13 10 ii 8
.,029
.'..l' :. 2 5
. ;, 68
.0172
. 192
'424
..093
'a On5 117 0.17 32 0 71
..364 194 142 1;
. 00*
,,. 011
?4 42
... i S 9 86 1
"48 1*
. 012
.'1037 1
.. :! 2 4,
. a. 82
- 8) 93
,,255
.. ? 's 4
."368
.s.
g
.. 2
+.
[,/ +. 9
. 10
.5._
11 12-13 14 15' 16 097
.. ;. 0 2 2.......
- 3. 6. 5 1
.- 11
.u 16
.io 9
. ' ' 4
...i 5 5
- )
.07 7047
'..u084r
. J ' 19
'.0 29
. :, n ; 7
..u. e 6
- c 83 5
4
.r 072,
. ',r 0 6 3,,
r,2 7..,. '. '. 4 2
.0 51
. t: 0 3 3
.irld 3
9
' 11 1,,,,
, 224
. 187
. t. 2. <1
.. 2a
.o 63
,.632
. ' 12
.1a7
. 120
. 466
..,1 4 6
.0172 '
- .. 71
. '41 213
~~
'.201
' " '.): 5 41
' '.J13 67~
0.*37
'.0192
..i M 4
'.- 29
.!589 908
,. 5 7 2,.,,..i 6 81 ;
..J'a5
.0424
..194
. i i. 4 3. _..
..u6r1 7
. 545
.. '290
.1 > 5 0 2392
.n167
.s?n5 48
. 3.8 116 9
7GA 1636
.P2 8 49 3170
>144 94
.1943 1-
. 636
~
..;25 2 ' # "..i 445
.ul63
.0604
. 192
.'39
.1955 11
.r?59
.c445,,j,.'c559,
. 329
__.06H1
..49-155 2690 12 no?
326 1 >' 5 /
- R4 479' 1 f. 7 6
. 399
. i ' f; 7 t
3 to 11
.:141
..- 19 2
_. 690
- j. 305
.9276,
.. 671 563
.?191 179
.64
..i 7 ') d I'6
.h-Q1
. '14 8'
. 538
.-(; 7 2 8 15 19
. 108 (313
,.;309
..;271 11'27
. 216 3 15 I*
F53 858
. '. 2 9 8
.,221
.L878
. 962 652
. tj 33
/
l' O
TABLE 3 - RADIATION INTERCHANGE FACTORS,Dj, WITH WET SHROUD ROD EMISSIVITY = 0.60, SHROUD EMISSIVITY = 0.00 1
2 3
4 S
6 7
d i f r-1 1sa?
2255
.(713
.. 542
.J443
. 112 566
,.74, 2
117 1441
. 636 1*33
. r%57 256
'222
!??
7
}
1 751 1273
..'216
,1 68 1142
. 368 In7
.'D8 4
?7i 1133
. 534
,,761
.9775 925 SS2
.'104 5
2 ? '..
.. 647
_. 3 51
.-775
.0577 658-690
... 5 ? 9 '.
6 V.3 266
.i184
. 925
.9581
. f. 6 0 5 616
. 166 7
'43 222
.:.. 9 4 0552
.964*
.;616
'232
. 3:3 9
174
. 343 3 R r3
.d?.9 1359
. i333
- 766
. 168 1 t-ni.
. 194
.. _.. _ 3 4.
... a ? 1 /
.0192
. 48P4
.*521
. ' 15 2, _
o ni2 343 27
.J'79
.all2
..517
.:547
. 139 11 n?6
. 093
.r148
.a'97
.t463
.e177
. 671 593 12 n1
.iJ36
'.41 51.
.J287
.r062
'.,15 7'
'!.379
- i423..
..; 6-
.0163
,.nied
. r. 39 8
... ; : 7 7, ' _
.,, d, l
~
13 9 :. 7
.:026
.;i14
.s 14
, neJ
. a13 24
. a 61
.'037 166 135 15 0n2
.o0 6
.97
.; 12
.; Pg
..not
.47
.38 i
l 16
. 015
.i1046
.0031 ~
."L'4
.b119
..3?3
'..: 3 3 5 ~
,. ' ') P 6
' '" ~; O
_A...... _.
.... t j
\\j
. 15..
_'.1 1 12 13 14
.,. ! ?
1,
/
b 9
O 'i
. GPU
,4 Sc 11 0.14
.so 7 t2
. g70 T
1 I
2 054
'.'043
.0083 '
. u '.18 0 26 ~
.: 013
' '3'
.fi O 6 7'~~
3
.'069
... e 0 6 9.....c2 5
.u'41 0 45
.f>02e
.9ed?
6141 9
. 21 7
.'179
. 197
.o 26
.0 60
. (' n ? 4
.. tj u 6 "23A 4
132
.,.453.
. 143,,
. tit 63
. ; f' 61,
. ai14, 5
102..
27a
.'.9127
' '. 0 31
.0168
. 937
. i 911 ^.
. 0 7 36 ~
~
6
..884
.1517
~V 8
104 276
.1i4S
. 379
,s398
. _. t' 16.6
. ti': 2 4
.. 0763 ' :
.l547..
. :. 6 71:
.; 79 521 7 _
.0154
.n27.
. a.' 3 8 391 1
9 676
..556
.;180
.., ?8
~
.0 97
. g54 28
.2467
~
10
. 556-
. 171
.i 417
. t.14 2 0521
~ i.ii u
'.ac36 ~
'.2492'
~ " - ' "
11
.: 1 A
.:417,.... :5 9 3521
.0651 -
.i.449
.. 133.
.,19A3 12 o6 284
. i t' 4 3
. ; 1. 7 3
.u747 1599 367
.$279
.3 r
0 14 p r. 4
- 1. 0
~. 651
..s373
.0191
. 635 71 9541
~
13 q97 5?1
~77A2
.L449 2514 0635
..519 493 15 056
.t072 267
.01A7
.0143
..986
. :' 119 3878 I6
. I n /4 5 1389 379
.,/R2 1115 1221
. 8 '4 e 1 'd 3 2
if TABLE 4 - LEAK RATES, RECIRCULATION LINE, DOUBLE-ENDED BREAK-DISCHARGE SIDE OF PUMP
- Time, Leak Rate, Total Mass Sec lb/sec leaked, Ib 0.1 24462 2298
- 0. 5 18766 10212 1.04 18930 20595 2.12 13829 39302 3
5958 46614
- {
4 3883 51225 0
2000 54046 6
939 55438 7
583 56188 8
73 56589 e
14
. _ _.. _. _.. _ _ _.. _. _ _ _. _. _. _ _.. _ _.... _ _ _ _. _ _ _ _ _ ~ _. - _.. _. _. _ _ _ _,.
f TABLE 5 - LEAK RATES, RECIRCULATION LINE, DOUBLE-ENDED BREAK - SUCTION SIDE OF PUMP
- Time, Leak Rate, Total Mass Sec lb/sec leaked,Ib, 0.1 23535 2223 0.5 188h 10224 1
18610 19593 2.2 11745 39639 3.1 5035 45340 4
3717 49219 5.2 2028 52617 6
1480 54054 7
919 55233 8
378 55984 e
15
f TABLE 6 - LEAK RATES, RECIRCULATION LINE, 1.0 ft SINGLE-ENDED BREAK
- Time, Leak Rate, Total Mass Sec lb/sec leaked, Ib 0.1 8057 784
- 0. 5 8832 43 93 1.04 83 01 8037 1.94 7933 16184
(
3 7673 24453 4.02 7239 32055 5.04 4975 38030 6
3990
'42070 7
3920 45502 8
1949 -
48079 10 1681 52169 11 956 53387 12.4 699 54493 14.2 511 55517
- 16 438 56340 l
16
!f TABLE 7 - LEAK RATES, RECIRCULATION LINE, 2
0.5 ft SINGLE-ENDED BREAK
- Time, Leak Rate, Total Mass Sec lb/sec leaked, Ib 0.12 5610 533 0,4 4700 1872 1
4510 4641 1.5 4309 3844
{
2 4229 8979
- 2. 5 4095 11054 3
4078 13099 4.2 3970 17923 5
3895 21070 S,2 3.740 25660 7
3598 28596 8.6 3295 34106 10.2 3090 39211 11 2843 41594 t
l 17
I f
t TABLE 8 - LEAK RATES, RECIKOULATION LINE, 0.25 ft* SINGLE-ENDED BREAK Estimated
- Time, Leak Rate, Total Mass
'sec m
_ lb/sec leaked, Ib
.01 2129 17.8 1
234S 2586 2
2192 5016 4
2042 9584 5
1996 11770
(;
6 1952 13869 8
1851 17908 9
1812 19847 10 1782 21710 15.1 1680 30597 20.2 1382 38604 25 1027 44465 30.1 581 48537 35.2 269 50619 40 124 51512 45.1 58, 51953 50 28 52154 1
18
l.
f' TABLE 9 - LEAK RATE, RECIRCULATION LINE, 2
0.1 ft SINGLE-ENDED BREAK
- Time, Leah Rate, Total Mass Sec lb/see leaked, Ib 0.1 1016 96 0.5 985 491 1.04 984 1022 2.12 578 2081 3
972 2940
(
4.2 955 4096 5.4 937 5231 6
926
,5791 1
s i
19
---..,-,,..__-,+..,..w,.--..m.,-.--~--.,m-
.~,.,-...,,,,.--,,o,-,,~~,..e,..,.---,,,ee,,.-
,-,.-,r-,
,,,-...,,--,,-,,e, v..-nwr
,--.,-w-,
,wm-
i.
t
!f i
TABI.E 10 - LEAK RATES, RECIRCULATION LINE, 2
l 0,05 ft SINGLE-ENDED BREAK
- Time, Leak Rate, Total Mass Sec lb/seg_
leahed,Ib 1
486 487 2
485 971 4
484 1941 6
458 2898 8
427 3781 10 413 4611 12 404 5415 14 3 96
,3197 l
20 v -,
,e,_.
e...m.,
.c
.u.--.....+,.--,.,-.,-i,__,.-,-.---.e,-,.,-,...--,,w-.,
8 i
I I
1 1
k, #
4' k
TABLE 11 - LEAK RATE, STEAM LINE BREAK
- Time, Leak Rate, Total Ma.is Sec lb/sec leaked, Ib 0.1 1521 126 -
- 0. 5 1290 658 1
1186 1271 2
988 2350 2.4 965 2739 2.6 1470 2987
(
2.8 1855 3322 3
3137 3908 3.5 2906 5407 4
2622 6784 5
2452 9313 6
2178 11578 8
1886 15679 10
~1568 19136 12 1293 22013 14 1002 242/4 16 827-26103 18 694 27677 i
20 602 28941 22 531 30066 24 462 31065 26 394 31920 27 359 32296 4
l 21
. _ _. _... _... _, _. _ _ _ -. _ ~... -.... _ _ _...,.
il AlSt 348 - Transverse Specimens 60 l
q Unirradiated 50
{
40 N
g g
- 1. 0 x 1020 nyt
,s y
p, EO
~
K f
f
/
/ \\
%'~~
'/
-3 20
/
/
\\
20 -N
- 5. 0 x 10 s
/
V
~L II 21
%/
/
- 1. 0 x 10 nyt u
g m
/
I
' 'I' ~ ~ I ~ ~~"
I~.
I I-I 0
RT 100 200 300 400 506 600 713 800 900 Temperature, *C 80 AISI 348 - Transverse Specimens
{
70 Unirradiated
/
.0 h%$
2 o
~
1.0 x 10, nyt j
30
)
- 5. 0 x 1020 nvt 20 10
- 1. 0 x 1021 nyt I
I I
I I
I I
0 i4T 100 200 300 400 500 600 7uo 800 900 Temperature, *C Fig.1 - Effect of Test Temperature on the Ductility of Irradiated Type 348 Stainless Steel in the Annealed Condition -Irradiation 3
Temperature ~290* C l
s.
22
0 I
9 1
I I
i I
h I
i 5
g I
n it teW I
t i
f d
u 5
o 7
rh S
0 I
i o
t
%t 2
e 0
ru a
r d e 0
k up a
om a
1 e
r.
I eT r
S I
B i
I e
z l
I iS i
in I
e m
ta e
r e
i d
y I
m d
u ot a
cwi e
r Ra p
T r
m 1S 0
p r
2 d
e e
I e
r t
3 nm r
s 0
o p
e e e CT C
l t
a E
n I
n N
u I
R I
R e
s E
d r e
o u r
GI Ra t
u r
T r e t
e p a
n m r
roe e
/
CT 1
p 5
m 0
m
/ /
m I
e a
T I
I da I
l C
2
.g i
I F
I 1
0 0
0 0
0 0
0 O0 0
0 0
0 0
0 0
6 8
4 2
1 1
- gty O
ow t
<l1 Il
!j j
ij i * -
l i i l
iI l
i!l!i
,?
l l
(
)
oo N
I l
$ f,
]
oo N
eC 21 i
oc CJ M
5ao 8
C"Q
/
nN l
-oo u
GQ W
u.
02 o
N d
C o 5 8 To' YI N t CQ a
E b
O o
026?OZ 8
?
R roC IQ l
n
.Mk e
8,,
o E
R o
6
.s N
N
,o, e
N
$"IOht3 *astagay gggu3
\\
24
...,m.,w--n,.--,
.y,,
.,-y..--..-~
e
w
,.r--
/
0.01 l
l l
~
Time = 4.0 min I
4 I
)
.5 E
i r.!
O c.
o i
Ta 0.001 i
=
8 x
(
~
~
o o
0.0001 I
l 1800 1900 2000 2100 2200 2300 Temp 3rature, oF Fig. 4 - Reaction Rate of Type 304 Stainless Steel with Steam s
25
r APPENDIX A - METAL-WATER REACTION CALCULATIONS The calculations presented ir. Tables A-1 through A-5 assume that the oxides formed are Fe30 and complex spinels based on the Fe30 4
4 str0cture. Only the iron, nickel, and chromium are considered because these are the major components constituting approximately 96% of the composition of the clad, and the contribution of the other elements to the energy release is small. A further assumption is that in the formation of each of the complex spinel oxide structures from the constituent oxides in additional 6000 cal / mole is released as per Reference 10. The thermochemical data were interpolated from the
(
tabular data of Reference 21.
G O
A-1
. ~. -
i l
TABLE A-1 Nominal Composition of 348 Stainlocs Stcol f
Element weight %
'(
Carbon 0.08 max.
l Manganese 2.00 max.
Phosphorous 0.045 max.
Sulfur 0.03 max.
Silicon 1.00 max.
Chromium 17.00 - 19.00 Nicke) 9.00 - 12.00 Nb. + Ta,,
10 x Carbon min.
l Tantalum 0.10 max.
Iron Balance l
TABLE A-2 Composition and Atomic Weights Used in Calculating Energy Release Element Weight Fraction Atomic wt.
Iron 0.7033 55.85 Chromium 0.1874 52.00' Nickel d.1093 58.71
+
l
=
e e
A-2
I 2,
h TABLE A-3 0xides Assumed in Calculating Energy Release h
Wt. Fraction of Metal in Oxidized SS Y
s i.
Oxide M/0 Fe Ni Cr Total l
O.5457 Fe 0 166 0.5457 34 0.0529 0.0937 0.1466 Ni Cr 0 163 24 0.0937 0.1440 Fe Cr 0 160 0.0503 24 0.1637 NiFe 0 170 0.1073 0.0564 24 TOTALS 0.7033 0.1093 0.1874 1.0000
- M/0 = grams metal / mole oxide 9
9 6
4 9
e 0
I e
A-3
~ - - -..
M o
TABLE A-4 C:'Jculated Molar Energy Release Accompanying Seler:ted Me'. :11 - Steam Reactions f
~
6 H, cal / mole g
!\\
1273*K 1370*K 1470* K 1600*K 3Fe + 2 0 +Fe 0 ;
-260,750
-260,250
-259,750
-259,10d 2
34 b
l 4 (H 0+ H2 + I!I 3 )
+238,500
+238,900
+239,300
+239,760 2
2 2,250
- 21,3 M
- 20,450
- 19,340 3 Fe :- 4 H Fe 03 4 + 4H2 2
2)
Ni + 1/2 0 -+ Nio
- 55,855
- 55,850
- 55,870
- 55,000 2
2Cr + 3/2 0 ^ C# 0
-270,105
-270,100
-270,130
-270,200 2
23 4 (H 0+H2 + 1/2 0 )
+238,500
+238,900
+239,300
+239,760 2
2 O 86,700
-- su,340 N1 + zcr + 4n u.N10+crv +4H
- s7,4bu
- 87,udu 2
3 2
6,000 36,'000 6,000 6,000 Nio + Cr 0 + Nicr 0 23 24 92,340 Overall d H
- 93,460
- 93,050
- 92,700 f
3) 62,500
~ Fo + 1/2 0 + Fo0
- 66,160
- 62,960
- 62,760 2
2Cr + 3/2 0 -+Cr 0
-270,105
-270,100
-270,130
-270,200-2 2
3 4 (H 0 -+ H2 + 1/2 0 )
+238,500
+238,900
+239,300
+239,760 2
2 Fo+2Cr+4H 0-*Fe0+Cr 0
+4H
- 97,765
- 94,160
- 93,590
- 92,940 2
2 2
6,000 6,000 6,000 6,i100 FcO + Cr 0 *.
Cr 0
23 2
4 Oven 11 6 H
-103,705
-100,160
- 99,610
- 98,9T6~
f 4)
Ni + 1/2 0 + N10
- 55,855
- 55,850
- 56,350
- 55,900 2
2Fe + 3/2 0 -->Fe 0
-192,400
-192,050
-191,620
-191,100 2
23 4 (H 0+ H2 + 1/2 0 )
+238,500
+238,900
+239,300
+239,760 2
2 Y,M 8,670
,000 Ni+2Fe+4H O NiU+Fe O2 3 + 4H2 2
6,000 6,000 6,000 6,000 Nio + Fe 0 ->Ni Fe 0 23 4
Overall 6 H
- 16,755
- 15,000
- 14,670
- 13,240 f
~
NOTE: Thermochemical data ic from Reference 11.
A A-4
TABLE A5 - Energy Rolcaso Accompanying Stninlcra Stcal -
o Steam Reaction, cal /g Stainlesa Steel Temp., 'K 7
Reletion*
1273 1370 1470 1600 l
h).
- 1. 0,5457(-22,250/168)
?2.27
'I 0.5457(-21,350/168) 69.05 l
0.5457(-20,450/168) 66.43 l
0.5457(-19,340/168) 62.82
- 2. 0.1466(-93,460/163) 84.05 0.1466(493,050/163) 83.69 0.lar3( f2,700/163) 83.37 83.05 0.1466(-02,340/163)
- 3. 0.1440(-103,765/160) 93.39 90.14 0.1440(-100,160/160) 0.1440(- 99,610/1G0) 89.65
~
0.1440(- 98,940/160) 89.05 A. 0.1637(- 15,755/171) 15.08 0.1637(- 15,000/171) 14.36 O.1637(- 14,670/171) 14.04 0.1637(- 13,240/171) 12.68 TOTALS 264.77 257.54 253.49 2 7.60 l
1000 1097 1197 1327
'C 1832 2007
,2187 2421
- F
- Reaction numbers refer to reaction equations in Table A4.
e p
e e
S e
A-5
914-532-9000 ggg
- ruucu m ru.
.i.
GULF Urt!!TED SERVICES e
GRASSLANDS HOAD, ELMSFORD, NEW YORK 10523 In reply, please refer to: SS-1086 May 14, 1973 e
Mr. R. S. Shimshak Dairyland Power Cooperative Lacrosse Boiling Wa. tar Reactor P.O. Box 135 Genoa, Visconsin 54632 s
Subj ect:
Technical Evaluation of Fuel Densification, Lacrosse Boiling Water Reactor
Dear Mr. Shimshak:
Enclosed are ten (10) copics of the subject evaluation.
If you require any assistance in reproducing additional copies for distribution to AEC-i;L, please let me know.
Ver truly vours, kflCHM ^ *J TDH/ch Thomas D. Hawkins Enclosures GULF UNITED SERVICES l
t
.