ML20010B554
| ML20010B554 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/17/1980 |
| From: | Ball S OAK RIDGE NATIONAL LABORATORY |
| To: | Tokar M Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19263D739 | List: |
| References | |
| FOIA-81-127 NUDOCS 8108170228 | |
| Download: ML20010B554 (45) | |
Text
.;
i OAK RIDGE NATIONAL LABORATORY openAtto av UNION CARBIDE CORPORATION NUCLEAR DIVl310N POST OFFICE sox Y OAK RIDGE, TENNESSEE 37830 Decerber 17, 1980 Dr. Michael Tokar Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
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Dear Mike:
Subject:
ORNL Calculations of Postulated Fort St. Vrain Reactor LOFC/WCD Accident' Response for Core Support Thermal Stress Evaluations
Background
The purpose of this letter is to describe the methods used in the subject analyses, summarize the results, and evaluate the accuracies of the calculations.
In May, 1978, audit calculations of several worst-case postulated Fort St. Vrain (FSV) loss-of-forced convection (LOFC) accidents following a design-basis earthquake were performed under ORNL's NRC/LSR-sponsored KIGR safety program.le2 Subsequently, Prof. Theophanous of Purdue not.ed that during the firewater cooldown (WCD) phase of the accident, the predicted temperature differences between certain lower reflector and core support block (CSB) nodes for adjacent refueling regions were very large (up to % 1500*F). There was thus some concern that such high thermal gradients could cause large thermal stresses in the support block regions. The task of calculating these stresces, given the output 3
of the ORNL ORECA code calculations, was assigned te LASL, and their results and conclusions are reported separately." Public Service Co. of Colorado (PSC) has also submitted an assessment of the problem.3 The large temperature differences predicted at the bottom of the core result from the uneven region (axial) temperature profiles that are generated during the earlier (LOFC) portion of the accident. Those regions with a high region power peaking factor (RPF) experience large reverse (upward) flows which transport the heat towards the top of the 8108170228 010529 PDR FOIA MULLEN 81-127 PDR 7
Dr. Michael Tokar Pagt 2 Deccmbsr 17, 1980
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Then after the cooldown beg' ins, the forward flow <, drive this heat
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core.
. downward, temporarily ra s ng t e temperatures at t e case of the core ii h
h to a much greater degree-than those low-RPF region that always had downflow. The uneven cooldown 1s accentuated by the fact that the Iow-t RPF region FWCD flows are larger (N 2:1 or 3:1) because of the lower effective flow resistance that re'sults from their lower temperature (and hence lower gas viscosity and larger. density). LASL also noted the possibility of high CSB stresses caused by sudden cooling from reverse flows that occur in some regions N2 hours into the FWCD.
i Some of the earlier LASL calculations indicated the possibility of relatively high stresses in the CSB for the reference 105% power case with che EQSB3 core,
7ggygertoevaluatethepotentialproblemsfor 2
present FSV operating conditions, a second set of initial operating i
parameters was generated by GA based on a maximum power of 72% and worst-case peaking factors and outlet temperature dirpersions for a cycle 2 E.O.L. core. The analyses described include Loth of these cases.
d 1
Description of ORECA Code Features ORECA simulates the 3-D thermal-flow transient behavior of the FSV core. The ORECA core node structure models the 37 refueling regions and an 18-channel approximation of the side reflector with 8 to 10 axial nodes each, for a total of 440 to 550 nodes. The model accounts for t
variable flow distributions between refueling regions (including reverse flow in individual regions), approximates the heat transfer and friction characteristic changes with flow regime, and includes expressions for gas and core material physical property changes with temperature. ORECA is alternatively run asid'sta6dI-alone code (as in the present case),
deriving time varying input values Of total core power, flow, pressure, and inlet temperature from other sources; or ORECA can be run as a p' art of'an ORNL overall system code ORTAP.'
n The results of ORECA calculations have been compared with data from 7
4 FSV scram tests and in numerous cases with output from the CA RECA 8
code, and the comparisons have generally been very good.
The detailed features of the ORECA code have evolved somcwhat since the original calculations were made. First, there have been numerous
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Decc=ber 17, 1930 Dr. Micheel Tokar" Page 3 improvement.s and corrections made to the code through " normal use" both in-house and by others, lesson learned from verification activities, and a comprehensive review by BNL. Second, the orig'inal intent of ORECA was more to predict outlet gas temperatures and maximum core temperatures rather than core support block temperatures. Hence a more detailed node structure was implemented. Where originally one axial node was used for each region's lower reflector and core support block, the present version uses two nodes for the lower reflector and a third for the CSB. The new version of ORECA also calculates and prints out the he-; flows into selected nodes via conduction and the heat flow out from convection. This Laformation is used as input to the stress analysis code (by LASL).
~
Regarding the difficult ouestion of interest in licensing matters, especially, as to how accurate the code's predictions are for postulated accidents, the following approaches have been taken:
j 1.
The code's models were developed from generally well-understood i
first principles to avoid problems with misuse of empirical models i
outside of their expected ranges. The problems of modeling distributed-parameter systems with lumped-parameter approximations were also addressed specifica11y'.
2.
Wherever possible, internal consistency checks are made.
3.
Peer reviews were conducted both within ORNL and by others.
l The code has been exercised for over 5 years on a variety of transients, j
4 f
large and small, and the results scrutinized by many people. A key means of checking and understanding the code behavior and l
accuracy is througheshe use of sensitivity studies, which are especially easy to run on ORECA because of its relatively simple input structure and low running times and costs (% $5/run typical).
This allows the investigator to alter the model and parameter assumptions and see directly how they affect t3,results.
7
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Dr. Micha21 Tckar P:go 4 Deconb:r 17, 1980
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5.
The results of ORECA have been compared on benchmark-type problems j
with other'similar codes, including RECA and FLODIS".and the
' agreement is generally good. The differences that do exist can usually be rationalized based on the known differences in the codes.
. 6.
ORECA output has been compared with FSV transient data, primarily 4 7
scram tests during cycle 1, and the agreement has generally been very goci. Our program is continuing in its efforts to verify ORECA (and other codes) as much as possible with existing' data and via proposed special tests.
ORECA Code Calculationh of the Postulated LOFC/WCD Acciden't The original FSAR scenarios for the postulated design-basis earthquake LOFC accident stipulated that the last-resort firewater cooling system for driving the pelton wheel turbines on the main circulators would be operational without any delay. NRC subsequently determined that t.p to a 90 min delay should be allowed in the accident reanalysis. The natural circulation flows that would occur in the core within this period ara sufficient to redistribute the heat significantly. Thus when forced circulation is restored, the distribution of the region cooling flows tends to aggravate the core temperature nonuniformities, since the higher flow resistance of the hotter regions (which need more cooling) restricts their flow more. Predictions of the maximum temperature differences in the lower core support regions, which occur N2 hr into the WCD, turn out to be quite sensitive to factors which affect the relative cooldown rates of the hotter and cooler regions, such as assumed total refueling region flow and the configuration of the region orifice positions. The total il~d ass'u'mption depends both on the estimate of l
the F D_ system output and the fraction of the total flow that bypasses the core. A low-resistance core orifice configuration assumption adversly affects the flow redistribution, because the wider-open the orifices, the less effect they have on the flow distribution relative to the region temperature conditions. Hence in the reference case ORECA runs, the most conservative condition was assumed, i.e. the widest-open orifice was fully open.
d e
me e
Dr. Michnol Tckar
. Pass 5 Decemb2r 17, 1980'
'i The ORECA code predictions of maximum support region AT's have changed'significantly since the original calculations for several reasons:
1.
Using the more detailed nodal structure, the CSB is now a separate node which has a ratio of heat transfer area to heat capacity that is much smaller than the average for'the lower reflector-CSB region; hence it doesn't cool down as fast in the FWCD phase of the accident as the node which originally represented both the lower reflector and the CSB, and the maximum AT's between CSB's are much smaller.
I 2.
The earlier version of the ORECA code had no printout which indicated the ragion orifice positions that would be required to give the individual region flows for a given overall core pressure drop. At one point in this analysis, some of the results were generated
.using unrealisi:ically small core pressure drops, i.e. the videst-a 9
l open orifices would have had to be open more than 100%. When more realistic values were used, the pre'dicted maximum AT's :ere smaller.
i 3.
Most recently, it was noted by LASL that the axial nodal region for which the maximum AT should be monitored is region 10 (the CSB),
rather than region 9 (the bottom half of the lower reflector).
Previously we had assumed that the AT's in the lower reflector, which were much larger, would have a more significant effect on the local stresses at the top of the CSB surfaces.
4.
At the November 7, 1980 meeting at NRC it was noted that there were major differences, between the CSB thermal conductivity expressions used by LASL and ORNL. LASL was using data for PGX graphite supplied recently by GAII, while ORNL was dsing a relationship given for the conductivity of therAowes reflector in GA-LTR-1". Figure 1 shows a comparison of the radial conductivity functiot.s used in the two cases.' Since the LTR-1 curves are for heavily irradiatad graphite, and the CSB's don't receive much radiation, it sas assumed that the PCK data was more appropriate. This means the assumed CSB conductivity.
is % 2-10 times greater in the never version, and since interregion r
conductivity is a relatively inportant 7 actor in the CSB heat transfer during the low-flow periods the change resulted in much.
1 smallur values of maximum AT.
V.
G-
4 December _17, 1980 Dr. Michael Tokar"
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-Page 6 Some resulta of the referencei runs for both the case of the 105%
. power EQSB3 core and the 72% power cycle 2 EOL core are shown in Figs.
2-5.
In the first case, the radial regions with the maximum CSB AT's are 19 and 36, and in the second, regions 35 and 36.
In the two cases, the maximum AT's between adjacent CSB's (axial region 10) are 657 and i
602*F.
For the nodes representirig the bottom halves of the icwer reflector (axial regh n 9), however, the maximum AT's are much larger: 1417*F in the 105% r uer case, and 1239'T in the 72% power case.
c Sensitivity Studies H
A number of sensitivity' studies have been done during the course of 2
this analysis, and it has been found that the maximum predicted AT s between adjacent re[r, ions vrt significantly when some parameters are varied within what may be cons a red error bands that represent ranges of uncertainty. This is particularly true of variations in AT's for the nodes representing the bottom halves of the lower reflectors, which, however, may not have a major effect on the* calculated thermal stresses of the core support structure.
Those parameters tested which did not have significantly large effects on the results were changes in overall core specific heat and thermal conductivity, laminar flow gas-to-core heat transfer coefficient, friction factors, and afterheat.
Those parameters which did turn out to be the sensitive ones were the assumed WCD flow that cools the refueling regions, CSB thermal conductivity (major in, crease), core flow resistance (i.e. orifice positions,,
and initial core power level. Reductions in the LOFC period also helped.
Because of the variety of mechanisms involved in core region transient behavior, it turns out that as a parameter such as the assumed region WCD flow is varied over vide ranges, different region pairs experience the greatest AT.
Furthermore, the predicted maximum AT for a given pair of regions peaks out at a given assumed value of WCD flow, while the peaks for other pairs occur at different flows.
For example, in the 105%
power EQSB3 case, WCD flow sensitivity studies show that a maximum CSB peak AT of 771*F would occur between regions 19 and 36 if the flow were I
0.8 of its reference value (vs 657*F for the reference flow). On the
Dr. Micha 1 Tokar" I
P g3 7 December 17, 1980 other hand, the peak AT between regions 20 and 21 increases from 602*F at the reference FWCD flow to 651*F at 1.2 times reference flow (Fig. 6).
The studies also showed that the maximum AT between lower reflector nodes was 1659'F and occured for 0.8 flow, (vs 1417*F at the reference flow), (Fig. 7).
In previous calculations where the C2B thermal conductivity expression was the same as that used for the lower reflector graphite, the maximum AT's for adjacent CSB's (105% power case) was N1020*F, compared to 657'F for the case of the new reference expression for PGX graphite conductivity.
An assumed reduction in the new CSB reference conductivity of 20% gave a maximum AT of 702*F, (7% increase).
A sensitivity run was-al'o made to show the effect of initial core s
AP on maximum CSB AT.
For a 20% higher AP (representing a relatively high resistance core, with maximum orifice openings of %50%), the maximum AT between regions 19 and 36 CSB's for the 105% power case was 566*F, or a reduction of 14% from the reference case. However, another pair of neighbors (regions 20 and 21) had the maximur AT, 608'F, which was larger than that of the other pair but still 7% smaller than the reference case AT.
The possibility of and severity of high CSB stresses caused by suddan cooling from region flow reversals during FWCD would depend on the magnis
'm of the reverse flows and the differences in CSB and outlet plenum temperatures at the time of reversal. The reverse flow phenomenon is relatively difficult to predict accurataly, however, since these flows are set up by small driving forces that depend on the axial temperature gradients in the core., The ORECA code predictions indicate that reverse flows during the FWCD period are much more prevalent in the 72% power case than in the 105% power case.
In the former, the first predicted reyersal occurs 2.5 hr after the start of the FWCD, and 4 hr after the start of FWCD there are 4 regions in reverse flow. The magnitude of the reverse flows are also quite sensitive to the assumed value of FWCD flow. In the 105% power case, the first reversal doesn't occur until 4.5 hr after the FWCD, and only that one region (region 20) ever reverses (at least in the first 6 hr).
e 4
9
D*, "tchael lokar**
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Prga 8 Decemb2r 17, 1980 Summary and Conclusions Calculations of core temperature transients during the postulated design-basis earthquake LOFC/FWCD accident sequunces were made for two sets of initial condition data: 105% power with an EQSB3 core, and 72%
l power with a cycle 2 E.O.L. core.' Summary results were presented in this l
report, and detailed outputs from ths ORECA code showing core block temperatures and pertinent heat flows were sent to LASL to be used as inputs to their stress analysis codes.
Sensitivity analyses were also run that showed how the predicted core support region temperatures were affacted by varidtions in the assumed
[
run conditions, models, and parameters.
I Estimations of the accuracy of these predictions should consider the following limitation of the code:
1.
No model verification tests have been done (and none are likely) which would confirm the redistribution of heat in the core, especially the extent of the shift to the top, during an extended LOFC. This first stage of the accident is crucial in that it sets up the conditions for the " race" between hotter and cooler region cooldowns during FWCD.
l 2.
Even though the ORECA model uses many nodes to approximate the core temperature distribution, the lumping is gross considering the need to calculate thermal gradients in specific block regions. The ORECA models also do not account for interr gion bypass flows, estimates of which are inco,rporated into the LASL stress calculations. The effect of thermal radiation heat transfer from each CSB to the lower plenum was also not included in the ORECA models, although its omission would probdkly make the calculcLed CSB temperature dispersion larger.
3.
Probably the greatest uncertaintics in the calculation are the starting l
times and the extent of the reverse flows developing during the FWCD phase. If these turn out to be crucial factors in determining potential is CSB damage, code verification tests such as those proposed by ORNL vocid be recommended.
O
Dr. Michael Tokar P;ga 9 Decemb:r 17, 1980 7
4.
On the other hand, there are a number of factors favorable to the credibility of the calculations that were enumerated previously in the section on code description. Considerin Tese factors, I e
believe that the calculations made for this investigation closely approach the best estimates that could reasonably be made in the absence of much more extensive testing and code verification efforts.
Under the circumstances, however, considering the uncertainties, there is no reasonable way to assign error bands and uncertainty estimates to the final results.
Please let me know if you have any further questions or comments.
Yours truly,
.y Af S. J. Ball, Manager HTGR Safety Studies for NRC/RSR SJB:rty cc:
C. A. Anderson-LASL P. R. Kasten G. C. Bramblett-GAC G. Kuzmycz-NRC J. C. Conklin F. R. Mynatt Ron Foulds-NRC Larry Phillips-NRC R. M. Harrington J. P. Saaders M. H. Holmes-PSC 7
M. Williams-NRC
- ~ ~g^h l
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Dr. Micha:1 T;kar P:go 10 December 17, 1980
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Referenceo t
1.
Letter from E. G. Case to S. Levine (NRC), Assistance in Evaluating and Confirming ECCS Analyses for Fort St. Vrain, Maj 1, 1978 2.
Letter from S. J. Ball to R.D. Schamberger (NRC), Evaluation and Confirmatio" ** ECCS Analyses for the Fort St. Vrain Reactor:
'Immediate..rks', May 25, 1978 3.
S. J. Ball, ORECA-r A Digital Computer Code for Simulating the Dynamics of HTGR Cores for Emergency Cooling Analysis, ORNL/TM-5159 April 1976 4.
LASA Report on Stress Calculations (in preparation) 5.
PSC Report on Stress Calculations (in preparation) 6..
J. C. Cleve3 eand et al., ORTAP: A Nuclear Steam Supply System Simulation for the Dynamic Analysis of High-Temperature Gas-Cooled Reactor Transients, DRNL/NUREG/TM-78, September 1977 7.
S. J. Ball, Dynamic Model verification Studies for the Thermal Response of the Fort St. Vrt.in HTGR Core, Proc. Fourth Power Plant Dynamics, Control and Testing Symposium, Gatlinburg, TN, Mar. 17-19, 1980, p. 15-1 to 15-12 8.
J. F. Petersen, RECA3: A Computer Code for Thermal Analysis of HTGR Emergency Cooling Transients, GA-A14520 (GA-LTR-22), August 1977.
9.
S. J. Ball, Approximate Models for Distributed-Parameter Heat-Transfer Systems, Trans. Instrument Soc. Amer., 3 (1), p. 38-47, January 1964.
10.
D. D. Paul, FI.DDIS: A Computer Model to Determine the Flow Distribution and Thermal Response of the Fort St. Vrain Reactor, ORNL/TM-5565, June 1976 11.
General Atomic Co., Graphite Design Material Properties, Document No. 904434/2, April 28, 1980, p. 62.
12.
General Atomic Co'., An Analysis of HTGR Core Cooling Capability, Gulf-GA-A12504, (GA-LTR-1), March 30, 1973, p.5-7 13.
Letter from S. J. Ball to 7. M. Williams (NRC), Status of Fort St. Vrain ECCS Confirmation in$1ys[s Planning, Dec. 22, 1977 9
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OAK RIDGE NATIONAL LABORATORY optmAfto av UNION C/.RBIDE CORPORATION NUutAR DIV1510N O
Po$r OFFICE box Y OAK RIDGE, TENNESSEE 37830 December 3, 1980 Dr. Charles A. Anderson M.S. 576, Group Q-13 Los Alamos National Laboratory Los Alamos, New Mexico'87544
Dear Chuck:
Subject:
ORECA Calculations for FSV Ther=al Stress Analyses Using New GA Thermal Conductivity Data for Core Support Block
~
Enclosed are the latest ORECA code calculations of the postulated 90-minute LOFC-FWCD accidents, one for the 72% power cycle 2 E.O.L. core and another for the 105% power-ECSB3 core.- The difference between these and previously submitted cases is the use of the new relationship for core support block (CSB) thermal ' conductivity obtained from reference.1 (enclosed). Other core component conductivity.relationshipe were taken from Fig. 5.2 of GA-LTR1 (also enclosed).
Previously, it was assumed that the CSB conductivity functions were the same as those shon for the top and bottom reflectors. %
Per cocversations with Tom' Burler, the latest runs were altered to include heat flow printouts for the entire FWCD per20d. Also, the axial e
region for which the maximum blo '-to-block AT shouit be monitored is region 10 (the CSB), rather than region 9 (the bottom half of the lower reflector).
Previously, I had assumed that the lower reflector's AT's, which were much larger, would have a more significant effect on th'a local str_ esses at the top surfaces of the CSB's.
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k For the 72% power case, the maximum AT between adjacent CSB regions is 600*F at t= 240 min, while for the 105% power case it is 602'F at t=
280 min.
These differences are much less than those of the previous-runs with the lower LTR-1 conductivity values for the CSB.
Please let me know if you have any questions or comments.
Yours truly, L
S.,*. Ball, Manager HTGR Safety Pregram.for'NRC/RSR Enclosure cc:
G. C. Bramblett, GA (w/ encl.)
M. H. Holmes, PSC (w/ encl.)
R. Foulds, NRC (w/o encl.)
G. Kuzmycz, NRC (w/o encl.) /
F. R. Mynatt (w/o/ encl.)
M. Tnkar, NRC (w/o e'ncl.)
-. ;g Ref.
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HTGR PROGRAM TECHNICAL NOTE HTGR SAFETY RESEARCH AT THE LOS ALAMOS NATIONAL LABORATORY
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K.R. STROH C.A. ANDERSON
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REFERENCES 1.
W. L. Kir', " Quarterly Report on HTGR Safety Research Program, October through December 1974," Los Alamos National Laboratory report IA-5870-PR (February 1975).
j
- 2.
W. L. Kirk, "HTGR Safety Research Program, Janua7y through March 1975,"
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3.
W. L. Kirk, "HTGR Safety Research Program, April-June 1975," Ios Alamos l
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l 4.
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g n
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LA-NUREG-6842-PR (June 1977).
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i 13.
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report NUREG/CR-0993, LA-7968-FR (August 1979).
1 20.
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21.
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22.
M. C. Stevenson, J. F. Jackson, and J. C. Vigil, " Nuclear Reactor Safety Quarterly Progress Report, January 1-March 31,1980," Los Alamos National Laboratory report NUREG/CR-1654, LA-8494-PR (August 1980).
23.
M. G. Stevenson and J. C. Vigil, " Nuclear Reactor Safety Quarterly Progress Report, April 1-June 30, 1930," Los Ala=os National Laboratory report NUREG/CR-1811, LA-8607-PR (November 1980).
24.
M. G. Stevenson and J. C. Vigil, " Nuclear Reactor Safety Quarterly Progress Report, July 1-September 30, 1980," Los Alamos National Laboratory report (to be published).
5.
J. E. Foley, "I-131 Release from an HTGR During the LOFC Accident," Los Alamos National Labor'atoty report LA-5893-MS (1975).
26.
G. E. Cort and J. H. Fu. " Core Heatup and Fission Product Release from an HTGR Core in an LOFC Accident," Los Alamos National Laboratory report LA-NUREG-6499-MS (1976).
27.
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2 8'.
C. E. Lee, C. E. Apperson, and J. E. Foley, " LEAF: A Computer Program to Calculate Fission Product Release from a Reactor Containment Building for Arbitrary Decay Chains," Los Alamos National Laboratory report LA-NUREG-6570-MS (1976).
C. E. Apperson, Jr., C. E. Lee, and L. M. Carruthers, " DASH:
A 29.
Multicomponent Time-Dependent Concentration Diffusion with Radioactive Decay Program," Los Alamos National Laboratory report NUREG/CR-0776, LA-7793-MS (1979).
C. E. Apperson, Jr., "SUVIUS: A Circulating and Plateout Activity 30.
Program for Gas-Cooled Reactors with Arbitrary Radioactive Chains," Los Alamos National Laboratory report NUREG/CR-0060, LA-7293-MS (1978).
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J" Experimental and Statistical Investigation of Thermally Induced Failure in Reactor Tuel Particles," Los Alamos National Laboratory report NUREG/CR-1787, LA-8547-MS (1980).
32.
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C63-Cd5 (1976).
33.
R. G. Behrens, H. O. Woodrow, and S. Aronson, " Vapor Pressure of Liquid Cesium by a Knudsen-Effusion Radiotracer Technique," J. Chem.
Thermodynamics 9, pp. 1035-1044 (1977) 34.
R. G. Behrens, " Thermodynamics of Transition Metal Carbonyls I, Fe(CO)5, Ru(CO)S, os(CO)S," J. of the Less-common Metals g, pp. 55-68 (1977).
35.
R. G. Behtens, "The Rcle of Transition and Lanthanide Metal Carbonyls it Verious HTGR Safety-Related Problems," Proc. ANS Topical Mtg. on Thermal Reactor Safety, Sun Valley, Idaho, July 31-August 4,1977 CONF-770708, Vol. 3, pp. 332-356.
36.
R. G. Behrens, " Metal-Carbon Bond Dissociation Energies and Enthalpies of Formation for Gaseous Metal Carbonyls," J. of the Less-Common Metals 3, pp. 47-54 (1978).
37.
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3'9.
R. G. Behrens and G. H. Rinehart, " Vaporization Thermodynandes and Kinetics of Hexagonal Silicon Carbide," NBS Special Publ. 561, Proc. of the 10th Materials Research Symp. on Characterization of High Temperature vapors and Gases, pp. 125-142 (1979).
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241-254 (1980).
41.
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42.
C. E. Apperson, Jr., L. M. Carruthers, and C. A. Anderson, " Fission Product Holdup in Graphite," Proc. US-Japan Seminar on HTGR Safety Tech., Tokyo and Fuji, Japan, Vol. 1, pp. 118-128 (1978).
43.
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44.
R. C. Feber, J. L. Lunsford, and W. A. Stark, Jr., " Application of the Complex Equilibrium Code QUIL to Cesium-Impurity Equilibria in the Primary Coolant of High Temperature Gas-Cooled Reactors," los Alamos National Laboratory report LA-NUREG-6373 (October 1976).
45.
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46.
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47.
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48.
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49.
J. L. Me r son and J. G. Benr e t t, "A Computer Methed for Analyzing HTGR Core Response to Seirmic Excitations," Los Alamos National Laboratory report LA-NUREG-6473-MS (September 1976).
50.
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l 51.
R. C. Dove, " Scaling Laws for HTGR Core Block Seismic Performance,"
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52.
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report LA-NUREG-6388-P (July 1976).
53.
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(Also presented at the Western Strain-Gage Spcing Meeting, March 3-4, 1981 in Las Cruces, NM).
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55.
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Symp. on Fundamental Research on Creep and Shrinkage of Concrete.
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56.
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57.
J. G. Bennett, F. D. Ju, and C. A. Ar.derson, "An Investigation of Failure Mechanisms for HTGR Core Supports," Los Alamos National Laboratory report LA-NUREG-6639-MS (December 1976).
58.
T. A. Butler, "Three-Dimensional Thermoelastic Analysis of a Fort St.
Vrain Core Support Block," Los Alamos National Laboratory report (to be published, May 1981).
59.
B. W. Washburn, "A Thermocouple Evt.luation Model and Evaluation of Chromel-Alumel Thermocouples for High-Temperature Gas-Cooled Reactor Applications," Los Alamos National Laboratory report LA-NUREG-6768-MS (April 1977).
60.
B. W. Washburn, " Accident Delineation and Evaluation of the High-Temperature Gas-Cooled Reactor System Concepts," Los Alamos National Laboratory report NUREG/CR-1200, LA-8170-MS (December 1979).
61.
J. C. Vigil, "Beginning-of-Life Nsutronic Anrn sis of a 3000 MV(t)
HTGR," Los Alamos National Laboratory report LA-6179-MS (January 1976).
62.
F. T. Adler and J. C. Vigil, " Integral Experiment Dcta for HTGR Safety Studies," Los Alamos National Laboratory report LA-5768-MS (November 1974).
~
, e3.
M. G. Stamatelatos, " Cross Section Space Shielding ia Doubly Heterogeneous HTGR Systems," los Alamos National Laboratory report LA-6157-MS (November 1975).
64.'
M. G. Stamatelatos, " Rational Approximations for Cross-Section Space-Shielding in Doubly Heterogeneous Systems," Nucl. Sci. and Eng.
61,, pp. 543-549 (1976).
4 65 P. G. Bailey, " Temperature Coefficient Analyses for the 3000 MW(t)
HTGR," Proc.' ANS Topical Mtg. on Thermal Reactor Safety, Sun Valley, Idaho, July 31-August 4, 1977, CONF-770708, Vol. 1, pp. 534-545.
I 66.
P. A. Secker and J. S. Gilbert, " Status of CHAP: Composite HTGR Analysis Program," Los Alamos National Laboratory report LA-6180-SR (January 1976).
67.
J. S. Gilbert, P. A. Secker, Jr., J. C. Vigil, M. J. Weeksung, and G.
J. E. Wi11 cutt, Jr., " User's Manual for the Composite HTGR Analysis Program (CHAP-1)," los Alamos National Laboratory report LA-NUREG-6576-M (March 1977).
68.
P. A. Secker, P. G. Bailey, J. S. Gilbert, G. J. E. Willeutt, Jr., and J. C. Vigil, " CHAP: A Composite Nuclear Plant Simulation Program for the 3000 MW(t) HTGR," Proc. ANS Topical Mtg. on Thermal Reactor Safety, Sun Valley, Idaho, July 31-August 4,1977, CONF-770708, Vol. 2, pp.
337-352.
69.
T. E. Springer and O. A. Farmer, "TAF: A Steady-State, Frequency-Response Simulation Program," Proc. AFIPS Fall Joint Computer Conf., San Francisco, CA, December 1968, Vol. 33, pp. 359-370.
70.
P. A. Secker, R. B. Lazarus, P. L. Rivera, and K. R. Stroh, " CHAP: A Gas Cooled Reactor Plant Simulation Program" Proc. 2nd US-Japan Seminar on HTGR Safety Tech., Tokyo and Fuji, Japan, November 22-25, 1978, Vol.
1, pp.88-108.
71.
K. R. Stroh, " Developmental Assessment of the Fort St. Vrain Version of the Composite HTGR Analysis Program (CHAP-2)," paper presented at the IAEA Specialists' Mtg. on Gas-Cooled Reactor Safety and Licensing Aspects, Lausanne, Switzerland, September 1-3, 1980.
+-
6
.1.
HISTORY REPORT DRAFT TITLE PAGE 3
March 1981 ORNL/NUREG/TM-
SUMMARY
OF ORNL WORK ON NRC-SPONSORED HIGR SAFETY RESEARCH, JULY 1974 - SEPTEMBER 1980 S. J. Ball, Manager Other Con'tributors:
J.'C. Cleveland J. C. Conklin J. G. Delene R. M. Harrington M. Hatta R. A. Hedrick L. G.-Johnson J. P. Sanders 1
l l
[
l
59 REFERENCES 1.
Oak Ridge National Laboratory, " Planning Guide for HTGR Safety and Safety-Related Research and Development", ORNL-4968, May 1974.
2.
Division of Raactor Safety Research, NRC, " Program Plan for Confirmatory HTGR Safety Research", Preliminary Draft, Feb. 28, 1975.
3.
Peter G. Kroeger, "0RECA code Assessment", BNL-NUREG-28707, July,1980.
4.
S. J. Ball et al., " Evaluation of the General Atomic Codes TAP and RECA for HTGR Accident Analyses", ORNL/NUREG/TM-178 (May 1978).
5.
A. Bardia and R. C. Potter, " TAP: A Program for Analysis of HTGR Nuclear Steam Supply System Performance Transients", GA-A-13248, January 30, 1976.
6.
J. F. Petersen, "RECA3: A Computer Code for Thermal Analysis of HTGR Emergency Cooling Transients", GA-A14520, August 1977.
7.
J. C. Cleveland, "CORTAP: A Coupled" Neutron Kinetics-lieat Transfer Digital Computer Program for the Dynamic simulation of the High Temperature Gas Cooled Reactor Core", ORNL/NUREG/TM-39, (January 1977).
8.
R. A. Hedrick and J. C. Cleveland, " BLAST: A Digital Computer Program for the Dynamic Simulation of the High Temperature Gas Cooled Reactor Reheater-Steam Generator Module", ORNL/NUREG/TM-38, (August 1976).
9.
J. C. Cleveland et al., "0RTAP: A Nuclear Steam Supply System Simulation for the Dynamic Analysis of High Temperature Gas Cooled Reactor Transients",
ORNL/NUREG/TM-78 (September 1977).
10.
S. J. Ball, "0RECA-I: A Digital Computer Code for Simulating the Dynamics of HTGR Cores for Emergency Cooling Analyses", ORNL/TM-5159 (April 1976).
11.
S. J. Ball et al., "High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research", Quarterly Progress Report, Jan. 1-March 31, 1977, ORNL/NUREG/TM-ll5, (June 1977).
12.
S. J. Ball and R. K. Adams, "MATEXP: A General Purpose Digital Computer Program for Solving Ordinary Differential Equations by the Matrix Exponential Method", ORNL-TM-1933 (August 1967).
13.
S. J. Ball et al., "High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research, Quarterly Progress Report, April 1-June 30, 1979", ORNL/NUREG/TM-356 (Nov. 1979).
14.
J. G. Delene, "A digital Computer Code for Simulating the Dynamics of Demonstration Size Dual-Purpose Desalting Plants Using a Pressurized Water Reactor as a Beat Source". ORNL-TM-4104 (Sept. 1973).
15.
J. C. Conklin, "0RTURB - A Digital Computer Code to Determine the Dynamic Response of the Fort St. Vrain Reactor Steam Turbines",
ORNL/NUREG/TM-399 (in publication).
16.
H. L. Bowers, "0RCENT - A Digital Computer Program for Saturated and Low Superheat Steam Turbine Cycle Analysis", ORNL-TM-2395 (Jan. 1969).
63 r
17.-. D. D. Paul, "FLODIS - A. Computer Model te Determine the Flow Distribution and Thermal Response of the Fort St. Vrain Reactor", OPJ.1/TM-5365 (June.1976).
-18.
.Public' Service of Colorado, Fort St. Vtain Reactor, Final Safety Analysis-Report,. Docket No. 50-267.
19.
J. C. Cleveland et al., " Simulation of the Response of the Fort St. Vrain High Temperature Gas Cooled Reactor System to a Postulated Rod Withdrawal Accident", Proceedings of ANS Topical Meeting on Thermal Reactor Safety, Sun Valley, DA., Aug. 1977,.V.2 p. 318-336, July 1-Aug. 4, 1977, Conf. 770708..
20.
S. J. Ball et al., High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research, Quarterly Progress Report, Oct. 1-Dec. 31 -1977", ONRL/NUREG/TM-164 (Jan.1978).
21.
Letter from S. J. Ball to R. D. Schamberger, Evaluation and Confirmation of ECCS Analyses for the Fort St. Vrain Reactor: "Immediate Tasks,"
May 25, 1978.
22.
S. J. Ball et al., " Investigations of Postulated Accident Sequences for the Fort St. Vrain HTGR," Proc. 2nd U.S.-Japan Seminar on HTGR Safety Technology, Nov. 24-25, 1978, Fuji, Japan, vol. I pp. 6-1 to 6-12.
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NOTE: This draft is issued for the purposes of review, editing and to elicit constructive questions and comments on its content and organization.
It may contain undetected errors, and should be-considered in this form as a working paper.
[
I tl
REFERE"CES Reactor Safety Research Program, Nuclear Regulatory Commission, " Program a.
Plan for Confirmatory HTGR Safety Research," preliminary draft, Feb. 28, 1975.
bl. Smith, C. L., " Fuel Particle Behavior Under Normal and Transient Conditions," Ga-A12971, General Atomic Co., Oct.1,1974 b2. Myers, B.
F., Baldwin, N. L., Bell, W. E. and Burnette, R. O., "The Behavior of Fission Product Bases in HTGR Fuel Material," Ga-A13723, General Atomic Co., Oct.1977.
b3. Myers, D. F. and Morrissey, R. E., "The Measurement and Modelling of Postirradiation Fission Product Release from HTGR Fuel Particles under Accident Conditions," Ga-A15018, General Atomic Co., Dec.1978, b4. Chandra, D. and Norman, J. H., " Diffusion of Cesium through Graphite,"
Ga-A13525, General Atomic Co., Feb. 2, 1976.
b5. Hanson, D.
L., "A Review of Fission Product Plateout Investigations at General Atomic," Ga-A14555, General Atomic Co., Dec.1977.
Mehta, S. and Zumwalt, L. R., "The Permeability and Retentivity of c.
Stressed Concrete Relative to Fission Product Iodine," Final reports Contract No. BNL 367132-5, ERSD 726(135), North Carolina State Univ.,
g Raleigh, March 8, 1979.
r d.
Zumwal t, L. R. and Phelps, J. S., Trans. Am. Nucl. Soc. 22, 212 (1975).
el. Hoinkis, E., "A review of the Adsorption of Iodine on Metal and Its Behavior in Loops," ORNL-TM-2916, fiay,1970.
e2. Compere, E. L., Osborne, M. F. and de Nordwall, H. J., " Iodine Behavior in an HTGR," ORNL-Tii-4744, February, 1975.
f.
Castleman, A. W., Jr. and Tang, I. N., Nucl. Sci. Eng. 29, 159 (1967).
~-
Castleman, A. W., Jr. and Tang, I. N., J. Inorg. Nucl. Chem. 32, g.
1057 (1970s, b.
Shiba, K., Hanada, M. and Yajima, S., N. Nucl. Sci. Tech. 6, 333 (1969).
1.
Growcock, F. B., Aronson, S., Friedlander, it., Skalyo, J., Jr.,
Hosseini, A. and Taylor, R. D. in Proc. 2nd U. S. - Japan Seminar HTGR Technology, Nov. 22-25, 1978, Tokyo and Fuji, Japan, vol. I, p. 140.
j.
Fi tts, R. B., Long, E. L. and Leitnaher, J. M., ORNL-TM-3385 (1971).
i k.
Got::mann. O. and Ohse, R. W. in " Behavior and Chemical Stata of Irradiated Ceramic Fuels," IAEA, Vienna,1974, pp 255-68..-
l REFERENCES (CONT'D.)
+
1.
Kleykarmp, H. Rept. KFK-2213, Karlsruhe, Germany (1975).
Aitken, E.
A., Evans, S. K. and Rubin, B. F. in " Behavior and Chemical m.
State of Irradiated Ceramic Fuels," IAEA, Vienna,1974, pp 269-85.
Tang, I. N., Aronson, S., Munkelwi tz, H. R., Dickey, J. H. and i
n.
Growcock, F. B. in Proc, Japan - U. S. Seminar on HTGR Safety Technology, Sep t. 15-16, 1977, Brookhaven National Laboratory, BNL-NUREG-50689 Vol. II, p. 280.
o.
Milstead, C. E. and Zumwalt, L. R., " Cesium Plate-out on Stainless and Carbon Steels," GAMD-7525,1966.
p.
Nic.olosi, S., Tang,1. and Munkelwi tz, H., "High Temperature Mass Spectrometry Study of Cs20 and R6 0," BNL-NUREG-51030, Brookhaven 2
National Laboratory, June,1979.
q.
Soo, P., Uheberg, G., Sabatini, R. L. and Schweitzer, D. G. in Proc.14th Carbon Conference, Penn State Univ., State College, Pa.,
June,1979, p. 467.
Soo, P., Shroy, R. E., Sastre, C. A., Schweitzer, D. G., Kraner, H. W.,
r.
and Jones, K. W. in Proc. 5th Conference Application of Small Accelerators, North Texas State Univ., Denton, Tex., Nov. 6-8, 1978.
Soo, P., Shroy, R. E., Rocks, B., Schweitzer, D. G. and Sastre, C.,
in s.
Proc.14th Carbon Conference, Penn State Univ., State College, Pa.,
June 1979, p. 109.
Chamberlain, A. C. and Wiffen, R. D., Geophys. Appl. 4j, 42 (1959).
t.
Tang, I. N., Munkelwi tz, H. and Nicolcsi, S.
L. in Proc. 2nd U. S. -
u.
'0" Janan Seminar HTGR Technology, Nov. 22-25, 1978, Tokyo and Fiji, Japan, Vol.
I, p. 208.
Nonnan, J. H., " Review of Yapor Pressures and Diffusion Coefficients v.
of Certain HTGR Core Materials and Fission Products for Use in Reactor Accident Calculations," GA-A12634, General Atomic Co.,1974.
Schwartz, M. H., Sedgeley, D. B. and Mendonca, M. M., " SORS: Computer w.
Programs for Analyzing Fission Product Release During Transient Temperature Excursions," GA-A12462, 1974.
Peterson, J.
F., " TAC-20. A General Purpose Two-Dimensional Heat Transfer x.
Computer Code," GA-8868, Sept.,1969.
140.
y.
Uneberg, G. D., Sastre, C. A. and Schweitzer D. G. in Carbon '80 Preprints. 3rd International Carbon Conference, Baden-Hader:, Gennany, June 30 - July 4,1980, p. 69.
Skalyo, J., Jr., " Fission Fragments and Activation Products in EOL Fuel Particles," Memorandum, Sept. 12, 1978, Brookhaven National Laboratory.. -
l ll,
O l
-References F
Reactor Safety Research Program, Nuclear Regulatory Commission, " Program a.
Plan for Confirmatory HTGR Safety Res~earch," preliminary draft Feb. 28, 1975.
b.
Romano, A. J. and Chow, J. G. Y. in Proc. Topical Meeting Thermal Reacto,r Safety, July 31 - August 4,1977, Sun Valley, Idaho, Eastern Idaho Sec-tion of the Power Division and Nuclear Reactor Safety Division of Ameri-j can Nuclea;* Cociety.
Isaacs, H. A. in Extended Abstracts,13th Biennial Conference on Carton, c.
July 18-22,1977, UCI, Irvine, Cali fornia..
d.
Growcock, F. B., Chandra, D., Heiser, J. and Skalyo, J., Jr. in Proc. 2nd U.S.-Japa? Seminar HTGR Safety Technology, Nov. 22-2E,1978, Tokyo and Fuji, Japan, Japan Atomic Energy Research Institute.
Velasquez, C., Hightower, G. and Burnette, R., "The Oxidation of H451 e.
Graphite by Steam, Part I: Reaction Kinetics," General Atomic Co.,
GA-A14951, August 1978.
f.
Growcock, F. B. and Heiser, J., in Extended Abstracts and Program,14th Carbon Conference, Pennsylvania State Univ., Park, PA, June 25-29, 1979,
- p. 459.
g.
Growcock, F. B., Eto, M., Heiser, J., III and Sastre, C. A., in Carbon
'80 Preprints, 3rd International Carbon Conference, Baden-Baden, Germany, June 30-July 4,1980, p. 238.
s.
h.
Grewcock, F. B. and Heiser, J. H., III, "High Temperature itass Transport 9
Experimental Program," Bill-HUREG-27502, Informal Report, 3rcokhaven flational Laboratory, liarch 1980.
1.
Caman, P. C., Flow of Gases Through Porous Media, Butterworths, London, 1956.
j.
Hutcheon, J. M., Longstaff, B. and Warner, R. K., Soc. Chem. Ind. Symp.
Ind. Carbon and Graphite, p. 259, 1958.
k.
Ash, R. and Grove, D. M., Trans. Faraday Soc. 56_, 1357 (1960).
1.
Yang, R. T., Liu, R-T and Steinberg, M., Fund. Ind. Eng. Chem.16, 486 (1977).
m.
Peroomian, M.
B., Barse11, A. W. and Saeger, J. C., "0xide-3:
A Compter Code for Analysis of HTGR Steam or Air Ingress Accidents," General Atomic Report, GA-A12493 (GA-LTR-7), January 1974.
n.
Wicke, E. and Kallenback, R., Ko11oid Z 97,138 (1947).
o.
Von der Decken, H. B., " Measurement of Gas Flow in Graphite," in Nuclear Graphite, 0FFC Dragon Project, November 1959.
p.
Dullien, F. A. L., A.I.Ch.E. J. 21, 299 (1975).
q.
Zeldovitch, J. B., Acta Physicochim. USSR _10_, 583 (1939).
r.
Th' ele, E. W., Ind. Eng. Chem. 3, 916 (1939).
s.
Hennig, G. R. in Chemistry and Physics of Carbon, Vol. 2, P. L. Walker, Jr., ed., Marcel Dekker, New York,1966, p.1.
l l
1 t
. i
__--_a--_--
Imai, H. and Sasaki, Y., Japan Atomic Energy Research Institute, Tckai-t.
a mura, Ibaraki-ken, Japan 319 '.1, private ccmmunication, December 10, 1978.
Eto, M. Grewcock, F. B. and Schweitzer, D. G., in Carbon '80 Preprints, u.
3rd International Carbon Conference, Baden-Baden, Germany, June 30 -
July 4, 1980, p. 255.
Krefeld, R., Linkenheil, G. and Karcher, W., lith Biennial Conf. on v.
Carbon, Gatlinburg, Tennessee, ORNL-CONF-730601,1973.
Mobaschewski, 0., private communication; Thrower, P. A., Progress Report w.
C00-2712-2 (1977).
Growcock, F. B. and Chow, J. G. Y., " Tensile Stress Corrorion of HTGR x.
Graphites," BNL-NUREG-24672, Informal Report, Brookhaven National g
Laboratory, July 1978.
Growcock, F. B., Heiser, J. H., III and Skalyo, J., Jr., " Kinetics of y.
the Oxidation of PGX Graphite in H 0(g)," Formal Report, Brookhaven 2
National Laboratory, in press.
Zimmer, J. E., in Proc. of Japar.-U.S. Seminar on HTGR Safety Technology, z.
Vol. II, p. 167.
k 34 -
-,e
~
Expe-iments will also be carried out to-assess how mechanical st-ength
'44 and ductility can be affected by simulated themal and helium chenistry tran-sients during a hypothetical accident scenario.
Fatigue and creep tests initiated under prototypic HTGR conditinos will be interrupted by higher-temperature thermal spikes it. sting for several days in order to quantify the losses in fatigue and creep failure times.
No work to date has been carried out in this area.
Similarly, coolant chemistry transients in which water is injected into the mechanical test system will be performed and the influence on failure rates will be detennined.
Additional mechanical tests on alternate structural alloys will also be initiated in FY 1981.
References i
1.
Soo, P. and Chow, J. G. Y., "A Correlatir.g High-and Low-Cycle Fatigue Data for Solution Annealed Type 304 Stainless Steel," Brookhaven National Laboratory Report No. BNL-NUREG-50601, December 1976.
2.
Soo, P. and Chow, J. G. Y., "The Effects of Mean Tensile Stresses on High-Cycle Fatigue Life and Strain Accumulation in Some Reactor Mate-rials," Brookhaven National Laboratory Report No. BNL-NUREG-50654, May 1977a.
3.
Soo, P. and Chow, J. G. Y., "Some Correlations of High-and Low-Cycle Fatigue Data for 2-1/4 Cr - 1 Mo Steel in the Annealed, and Normalized and Tempered Conditions," Brookhaven National Laboratory Report No.
BNL-NUREG-50610, January 1977b. -
I
l n =,p=.wwa:ml
"%l 4.
Soo, P. and Chow, J. G. Y., " Correlation of Low-Cycle and Hich-Cycle Fatigue Data for Solution-Annealed Incoloy 800" in Alloy 800, W. Betteridge, et al., Editors, Proc 3edings of the Petten International Conference on Alloys, Petten, The $letherlands, March 1978, North Holland Publishing Co., 1978, pp. 169-174.
5.
Soo, P. and Chow, J. G. Y., " Development of a Procedure for Estimating the High-Cycle Fatigue Strength of Some High-Temperature Structural Alloys" in Methods for Predicting Material Life in Fatigue, ASME,1979,
- p. 185.
6.
Soo, P. and Chow, J. G. Y., "High Cycle Fatigue of Solution-Annealed and Themally-Aged Type 304 Stainless Steel," J. Eng. Mat. Tech., Vol.102, 1980a, p. 141.
7.
['
8.
Soo, P., Sabatini, R. L., Epel, L. G. and Hara, J. R., "High Cycl e Fatigue Behavinr of Incoloy 800H in a Simulated High Temperature Gas Cooled Reactor Helium Environment," Brookhaven National Laboratory Report No. BNL-HUREG-51156, January 1980b.
9.
Lyman, T., (Edit.), Metals Handbook, 8th Edition, Vol. 7, ASM,1972, p.
163.
3.
'ihile the femation of some unmixed regions cannot be ruled cut, N
i especially in geometrically isolated areas, substantial mixing can be expected to occur under most conditions.
4.
With the extremely low probability of such events further work in this area does not appear warranted at this time, but could be resumed if required by a future shift in emphasis.
References (a)
D. I. Macnab, "The CONTEliT-G computer program and its application to HTGR containrent," General Atomic Company, Report No. GA-A-12692,1974.
(b)
I. Omata, "An Analysis of Gas Layering and Flammability in the Contain-ment Vessel of an HTGR Following Depressurization," Brookhaven National Laboratory, Report No. BNL-NUREG-50622,1977.
(c)
Shihi Pai, " Fluid Dynamics of Jets," Van Nostrand,1944.
(d)
W. Forstall, H. Shapiro, " Momentum and Mass transfer in Coaxial Gas Jets," Appl. Mech.,
7, 399-408, 1950.
(e)
J. L. Boccio, G. Weiderstein, S. Dash, "A Computational System for the Prediction of Low Altitude Rocket Plume Flow Fields," Vol. III, Mixing /
Af terburning Model, General Applied Science Laboratories, GASL-TR-239, Westbury, N. Y.,1976.
(f)
B. E. Launder, A. Morse, W. Rodi, D. B. Spalding, " Predictions of Free Shear Flows - A Comparison of the Perfomance of Six Turbulence Models,"
.I.
=
v in Free Turbulent Shear F1cws. Vol. I, Conference Proceedings,.1ASI JD Report No. SP-3' 1, P361-422,1973.
4 (g)
S. V. Pataukar, D. B. Spalding, " Heat and Mass Transfer in Soundary Layers," Intertext Books, London 1970, Chapter 2.
(h)
W. C. R!vard, 'O. A. Farmer, T. D. Butler, " RICE: A Computer Program for Multicomponent Chemically Reactive Flows at A'11 Speeds," Los Alamos Scientific Laboratory, Report LA-5812, March 1975.
(i)
J. L. Boccio, J. Colman, J. Skalyo, J. Beerman, "HTGR Depressurization Analysis," Proceedings of the Second U.S.-Japan Seminar on HTGR Safety Technology, Vol. I, p.172, JAERI,1978.
( Al so BNL-NUREG-25334).
(k)
H. B. Palmer, M. Sibulkin, R. A. Strehlow, C. H. Yang, " An Appraisal of Possible Combustion Hazards Associated with a High Temperature Gas g
Cooled Reactor," BNL-NUREG-50764,1978.
(1)
P. S. Bailey, "A Flammability Limit' Test Apparatus," MS Thesis, Univer-sity of Illinois,1978.
~-
is incurred as the hotter channels will tend to get less flow, thereby becom-s ing even hotter, and getting even l'ss coolant flow, etc.
e Figure _.1 shows the flow rates and the temperatures about midstream in a low flow vs. a high flow channel, for three computations, using different gas temperatures for evaluation o' average gas densities per node.
In the first case the solid block temperature of the node is being used as first approxima-tion for the gas temperature.
In the second case an arithmetic average gas temperature is used. A logarithmic average, representing an axial integration.
with the ORECA assumption of constant graphite block temperature per node is being used in the third case.
As expected, whichever density is used does not make any difference in the high flow channel. However, in the low flow channel average core tempera-ture differences of about 600'F are observed, as the low flow channel will receive less tiow, and heat up more over the time period from 100 to 250 minutes. This strong sensitivity of the resulting output temperatures 'o a relatively minor assumption in the analysis shows that under such low flow conditions the flow redistribution and ultimate core temperatures can be extremely sensitive to minor variations in design and operating parameters such as the flow crificing. An accurate analysis of such situations would require more refined modelling of the specific situation.
, References
[a]
P. G. Kroeger, "0RECA Code Assessment," Brookhaven National Laboratory, Upton, New York 11973, BNL-NUREG-28707,1980.
.Ch.
I l
i References
'l (a)
J. Skalyo, L. G. Epel, C. Sastre, "An Analysis of the Methods Utilized in Oxide-3", Brockhaven National Laboratorv. BHL-NUREG-50810, 1978.
(b)
M. B. Peroomian, A. W. Barsell, J. C. Saeger, " Oxide-3: A Computer Code for Analysis of HTGR Steam or Air Ingress Accidents",
Ganeral Atomic Co., GA-A12493, (GA-LTR-7), 1974.
(c)
R. C. Giberson, G. L. Tingey, " Reaction of Gaseous Impurities in a High Temperature Gas Cooled Reactor",
Battelle Northwest Laboratory, BHWL-974, 1968.
se===
1 S.te..a.-.s..W+ 5er March 10, 1981 PtjBLICSERVICECOMPANYOFCOLORADO P.O. Box 840
- Denver, Colorado 80201 Attention:
Mr. F. Swart Manager, Nuclear Projects
Reference:
Fort St. Vrain Nuclear Generating Station Unit No. 1 PSCC Document 111942 Stearns-Roger Project C-22240
Dear Mr. Swart:
Pertaining to our conversation today, I have had discussions with several individuals within our organization and cannot substantiate that any report on Fort St. Vrain has been prepared.
In addition, it is not our practice to number our reports, and so the Stea ns-Roger Report #308 requested by the DUC does not exist.
Please do not hesitate to call if we need to do anything more to eliminate this rumor.
Very truly yours, STEARNS-ROGER ENGINEERING CORPORATI0ft J. J. Donovan Project Manager JJ0:nc cc:
Bill Fitzmorris. " SCC FLWeigand/LFisher LmcBride, PSCC JJ0cnovan 450c CHER AY C9EEK CRIVE. P O. BOX 5888. DENvFA Cold A ADO W17. Dwmc rm ies.ne. mema aa. aara *** "'" a **
(..
.m
~
tM Dsate Steart s(W5?
t sy-
-apaer E NGINEE8 3 APCSA March 10,1981 PdBLIC~ SERVICE.COMPANYOFCOLORADO P.O. Box 840
' Denver,- Colorado 80201 Attention:
Mr. F. Swart Manager, Nuclear Projects
Reference:
Fort St. Vrain Nuclear Cenerating Station Unit No. 1 PSCC Document 111942
.Stearns-Roger Project C-22240
Dear Mr. Swart:
Pertaining to our conversation today, I have had discussions with several individuals within our organization and cannot substantiate that any report on Fort St. Vrain has been prepared.
In additien, it is not our practice to number our reports, and so the Stearns-Roger Report #308 requested by the PUC does not exist.
Please do not hesitate to call if we need to do anything more to eliminate this rumor.
Very truly yours, STEARNS-ROGER ENGINEERING CORPORATI0ft J. J. Donovan Project Manager JJD:nc cc:
Bill Fitzmorris, PSCC FLWeigand/LFisher LMMcBride, PSCC JJDonovan
._