ML19247A859

Discusses Bounding Estimates of Damage to Zircaloy Fuel Rod Cladding in TMI-2 Core 3-h After Start of Accident on 790328

ML19247A859
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Site:	Crane
Issue date:	06/20/1979
From:	Picklesimer M NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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ML19247A858	List: ML19247A857 ML19247A859
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{{#Wiki_filter:* f "%f, UNITED STATES f k NUCLEAR REGULATORY COMMisslON f e( g waseares, o. c. 2esss Wi .m1 2 0 575 s MEMORANDUM FOR: File ~ FROM: M. L. Picklesimer [ 2 Fuel Behavior Research Branch MN

SUBJECT:

SOUNDING ESTIMATES OF OMAGE TO ZIRCALOY FUEL ROD CLADDING IN THE TMI-2 CORE AT THREE HOURS AFTER THE START OF THE ACCIDENT, MARCH 28, 1979

SUMMARY

AND CONCLUSIONS A set of simp'ified bounding calculations have been made in an attempt to set upper and lowar limits on the damage produced in the TMI-2 core through the first three hours of the accident on March 28, 1979. The calculations use simplifying assumptions such as (1) the heat capacity of the fuel rod is constant with temperature, (2) the axial power profile in the assemblies is cosine,(3) the heat lost to the steam and surroundings is a constant fraction of the power developed in the rod by both decay heat and oxidation of the cladding, and (4) the core is uncovered at a constant rate. Also, it was assumed that the heat-up of the core was terminated at 50 minutes after the top vas first uncovered by two-phase and slug flow of coolant frem reficoding Radial and axial profiles of damage in the core were estimated frcm water. the calculations. It is concluded that the maximum damace likely in the core was: (a) All fuel rods burst with elevations ranging frcm about one foot frem the ~ top of the core in the center assembly to about three feet down in scme of the peripheral assenblies. (b) The total amount of Zircaloy reacted in cne first three hours to produce hydrogen was "guestimated" to be betveen 25t and 30% of all Zircaloy in the core. (c) Embrittlement of cladding by oxidation occurred to a depth of between 6 and 7 feet frcm the top of t.1e core in the center assembly, down to about 5 feet to 5 fee? in most of the assemblies, and did not occur on the lowest power corner assemblies on the periphery. (d) A liquid phase was famed between molten Zr + Zr02 eutectic r.id the outer part of the U02 fuel pellets next to the cladding at depthr fecm the tcp of the core dcwn to 5 to 6 feet in most assemblies and be'. ween 6 and 7 feet in the center assembly. It did not form in the corner assemblies. 7908030 Ab/ d

AV 20 m files (f) No significant oxidation of cladding occurred in the bottom five feet of the core at any radial position". It was concluded that the minimum damage that could have been produced was: (a) All fuel rods burst, with burst ranging from 1 foot from the top of the core in the center assembly to about 3 feet frem the top in peripheral corner assemblies. (b) Embrittlement of cladding due to oxidation occurred to a depth of 6 feet frem the top of the core in the center assembly, to 5 feet in most of the assemblies, and did not occur in the peripheral corner assemblies. Also, the top 1 to 2 feet of many of the outer assemblies was not oxidized to the embrittlement stage. (c) None of the assemblies was oxidized significantly at depths below 6 feet from the top of the core. (d) A licuid phase between molten Zr + Zr02 eutectic and the outer part of UO fuel pellets was famed only in the center assembly and down to a de th of about 4 feet. In view of the simplifications imposed to allow calculations to be made, and the assumptions required to establish a " scenario" of events frem the quite limited data available it is concluded that the more realistic estimate of damage lies near the upper bound described by the " minimum" and " maximum" plots of core damage. The " realistic" estimate of core dar. age allows expla-nation of most of the reported observations of fission product emission, hydrogen generation, high temperatures in the core, etc., with the exception of the survival of the themoccuples that passed through the hot core to the upper end fitting of many of the assemblies. The calculational procedures and many of the graphical plots can be used to improve the bounding estimates as new data are received and correlated to establish a better sequence of the events that actually occurred in the system. INTRODUCTION The data on the Three Mile Island, Plant No. 2 (TMI-2) accident on March 23, 1979 available to the Fuel Behavior Research Branch (FBRB), RES/NRC by about April 20, 1979 was examimed and discussed amongst the several members of the F3RS in an attempt to establish a " scenario" of events transpiring in the first three hours of the accident. Because the limited data available made difficult the establishment of a scenario that could provide a sound basis for calculations, the author decided to attempt a set of simplified, bounding calculations of damage to the core that could be modified relatively easily to produce better estimates and bounds for the extent of damage as new data or interpretations were developed.

"'40 53 Files A simple scenario was developed to hpve the following features: (a)although the coolant in the core was probably a two-phase mixture of flowing water and steam at the time, the core was adequately cooled and was not uncoverci at any position until after the second set of primary coolant pumps were virned off at 100 minutes after turbine trip. (b) the top of the core was expo ed (f.e., above the liquid coolant level) within a few minutes afterwards, (c) the water level in the core decreased at a constant rate of 12 feet per hour after the top was uncovered. (d) no position on a fuel rod could begir heating up above saturation temperature (Tsat) until it has oeen exposed by *he dropping water level in the core (e) the decay heat in a fuel red at any level was a fraction (to be. chosen as a parameter) of the operating power at that position in the specific assembly at the start of the accident (i.e., a function of the radial and axial power profiles in the core), (f) although known to be an approximation, such variables as heat capacities, steam ficw rates, heat generation rates by oxidation, etc., would be taken as constant with temperature or time, or were approximated by linear interpolation over small temperature or time increments, and (g) the minimum damage would be detemined by using only decay heat as the heat source while the maximum would be approximated by including an estimate of the heat generated by oxidation that would be made always on the high side of any interpolation. ASSUMPTIONS The folicwing assumptions have been made to allow simplified calculations to be made, even though several are known to be in error by as much as 10-15", and others are not adequately supported by data or other information from the plant. Improvements can be made wher justified by additional data or a need for better estimates of the damage to the core. 1. The voiding of the water in the irrediate volume of the core started at or "very shortly" after the shut-down of the sec::nd set of primary coolant pumps at 100 minutes into the accident. 2. The voiding of the water in the core ended during the rapid repressurization of the primary systen shortly after 2.3 hcurs into the accident, and refill-reflood of the core began. 3. The level of coolant in the core decreased linearly with time after the top was first uncovered, at a rate of 12 feet per hour. 4 The heat capacities of UO2 and Zircaloy are approximately constant with temperature (they increase by 10-15% with increasing temperature, and phasa changes are ignored). 5. Since the heat losses from the fuel rods to steam and to the neighboring control, poison, axial power shaping, and instrumantation rods / tubes can not be estimated with any precision, the losses can be epproximated by setting them equal to some arbitrary fraction of the heat developed in or by the fuel red after the water level has fallen below the specific elevation of the calculation. For the first estimation of damage. t.se fraction chosen is 25%. kb)

Files M 20:s;3 6. The total power peaking factor map of the core is the same as that reported for Rancco Seco (1/29/75) and shown in Figure 1. The axial power peaking facto.- is assumed to be 1.3 and the axial power profile to be cosine. 7. The heat generated by steam oxidation of the Zircaloy q; adding can be ~ calculated using the Cathcart-Pawel rate equations,tl'C the oxidation is one-sided, and the oxidation heat and amount can be calculated by linear ramps over 100'F increments, using linear heattag ramp calculations developed with the BUILD-5 code. 8. The decay heat in the middle of the period of calculations is it of the power developed at the start of the accident. 9. OxidationoftheZircaloyfuf.1rodlocallybygeamstopswhenthemolten c-Zr + Zr02 eutectic reacts with the'UO2 fuel to produce a ligt;f-phase at about 3480*F. The oxidation of the zirconium metal contained in the liquid phase contint E:.. but it no longer contributes heat to raise the temperature of the rod at Me spot where it had been before it melted.

10. Embrittlement of the Zircaloy cladding to themal shock (quenching) by oxidationcanbepredic*{gd{r ne-sided oxidation from the studies of Kassner and Chung (ANL)

• 5 for embrittlen. cst by two-sided oxidation.

They found that embrittlement was p-ent if the betc phase contained 0.9-1.0 weight precent oxygen or wa. hes thick. De code (BUILD-5) developed by Paw.l(e s than about 0.004 j 0 was used by Marino to calcu-lam the oxygen distribution and content in the beta phase for linear temperature ramps at rates encompassing the ones predicted for several positions in the TMI-2 core. The temperature detemined for the embrit-tiement conditions ranged between about 2500 and 260c de ending on the ramp rate. The lower temperature, 2500*F was chosen.

11. The burst temperature for the fuel rods can be 9gtimated from the pressure-burst temperature data of Chapman (MRBT, ORNL),M a terrperature ramp rate of about 1 F per second, and a knowledge of the cold fill pressure for the reds of 445 psi. Calculations show that if the pressure drop across the cladding is 40Cpsi, the fuel rod wP1 burst when it reaches a temperature between 1450 and 1500*F.

If it is 500 psi, the burst temper-ature is between 1400 and 1450*F. CALCULAT!CNS The details of the procedures used in the calculations are given in the attached appendix. The fomulations and. results are given be'ow. 1. Heat capacity of the fuel rod: aHUO2 = 30.42 x 10' Stu/ inch of red /*F 6.24 x 10-4Btt./ inch of rod /'F aHzr = aHrod = 36.66 x 10-43tu/ inch cf red /*F !} dl

Files -5 30 M 2. The axial power profile, P(z),.in the fuel red is calculated by the equation (9); P(z) = [B + A cos (w z/2L)] P

A, B = constants related to the pcwer B + 2A/w avg levels in the core Pmax = A + B, Pavg = 8 + 2A/w

Pmax = axial peaking factor Pavg tatal peakino factov_(rod) p

), [ axial peaking factor (rec)] p avg ( 3 ,) then for axial peaking factor = 1.3, and Pavg core = 6.0 kw/ft, P(z)=[0.5746+ces(vz/2L)]Payg = [0.4744 + 0.8256 cos (sz/2L)] P 0.5746 + 2/w avg 3. T(z) = Tsat + ai fordecayheatonly,ai=[gH.H.(at)(1-hif)] ~ rod for decay heat plus oxidstion heating, and T(z) s 1500*F, T(z) = Tsat + AT, aT = U 3 y and T(z) > 1500*F, T(z,1) = f(7,4.j ) + 100 t( z,1 ) = t ( z,1 -1 ) + a t ' ' ' . 100 (1-nif) [aHrod0.H.+0.H.) and the time increment, at', is calculated for the time recuired to heat 100*F above the last T(z) for all temperatures above 1600*F with the value of oxidation heat (0.d.) appropriate for T(z) where T(z) = temperature of red at elevation z T saturation temperature at time of uncovering by coolant sat 17 = tempe. 1ture increase durine time increment at at = time of calculation from start of core uncovering minus time of uncovering at elevattui. I at' = time required to heat frem T( g.) to T(z) with oxidation D.H. = decay heat hlf = heat loss fraction 0.H. = oxidation h. eat . l f. 1Hrod = heat capacity of fuel red kO!

M 2 0 ISTS Files 4. Axial power profiles are shown in Figure 2 for total peaking factors of 1.9.1.6,1.3,1.19, and 0.66, values representative of several areas in the core. RESULTS Time-temperature curves for the fuel rod cladding are shewn in Figures 3-7 for tntal peaking factors frem 1.9 to 0.66 and a heat loss factor of 25" to steam and non-fueled structures as functions of distance from the top of the core as the core was uncovered at a constant rate of 12 feet per hour. The straight lines are for heating by decay heat only, and the curved lines are for heating by both decay heat and by oxidation of the cladding. The latter allcws for the increase of oxide thickness with both time and temperature. The details of tr.e calculations are given in Table A-I of the Appendix. Radial and axial core maps of damage to fuel rod cladding are shown in Figures 8-11 for the maximum and asinimum amounts of damage estimated. The damage is graded into levels of (1) no significant oxidation, (2) significant oxidation but no emerittlement, (3) oxidation embrittlement of the cladding to thermal shock such as produced by reflooding coolant, (4) burst location, and (5) formation of liquid phase by reaction of the molten Zr + Zr02 eutectic with the cuter surfaces of fuel pellets. The curves of Pgures 3-7 are read by starting at the time of uncovery for a given level fre : the top of the core and folicwing it upwards. At 1600*F the choice must be made between inclusion or exclusion of the heat of oxidation. In Figure 3 (tetal peaking factor 1.9), the curves plotted show that the fuel rods in the center assembly at the one fact level reached the bu. st tempera-ture range of 1400-1500*F first, and the rods burst at that level at 18-20 minutes after the top of the core was first uncovered. If only decay heat is included in the calculation (straight lines), the fuel reds reached the Zr + Zr02 eutectic temperature of about 3480*F first at the 3 foot level, folicwed witnin 1-2 mnutes by the 2 and 4 foot levels. Mcwever, if oxidation heat is included (maxtum damage, curved lines), then this temperatura was first reached at the 1 and 2 foot levels essentially simultaneously and the 3 foot level did not reach that temperature until about two minutes later. The German core melt work of Hagant3) has shown that the eutectic liquid immedi-ately reacts with the UO2 fuel to fern a icwer melting liquid phase. The liquid forned ficws dcwn frem the region of formation, and oxidation of the cladding no icnger occurs at that specific elevation. The heatup at that point then continues only by decay heat. No allowance has been made in the plots for the decrease in decay heat generation caused by the loss of fuel at that position. The plots then show that by the maximum damage estimate for the assembly having a total peaking factor (tpf) of 1.9 the reds burst at 18-20 minutes after core uncovery started at the i foot level, and reached the eutectic reaction temperature after 34 minutes at both the 1 and 2 foot levels, foi-lowed by the 3 fcot level at 36 minutes the 4 foot at 39-40 minutes, the top of the core (0 foot) at 41 minutes, the 5 foot at 43 minutes, and the 6 fcot at 47-1/2 minutes. 48/ M

Files' M 0'!379 If the " turnover" occurred at 50 minutes after the top was uncovered, the rods at the 7 foot level reached a maximum temperature of about 2300*F and were not embrittled by oxidation. The embrf t'lement boundary was then just above the 7 foot level and below the 6 foot. Tne lowest level of the eutectic reaction was just above the 6-1/2 foot level. No significant oxidation occurred below the 7-1/2 foot level. For the minimum damage estimate for this center assembly (tpf = 1.9) the rods burst at the 1 foot level at 18-20 minutes, and only the 2, 3, and 4 foot levels reached the eutectic temperature. All rods were embrittled down to the 6-6-1/2 foot level, and no significant oxidation occurred below the 7-1/4 to 7-1/2 foot level. The same procedure can be followed for the remaining assemblies in the core, with interpolations being made proportionately to differences between the peaking factor values. These results are presented in Table I as depths of penetration from the top of the core in feet for the several radial peaking factor values used in the damage estimate. These results were then interpo-lated and plotted in Figures 8-12 as pPtographic core maps of the damage. ' DISCUSSION The detailed conclusions drawn frem the_ analysis and the scenario used are somewhat sensitive tc the details of the scenario used. However, the general conclusions as to the types of damage present in the core are not. Acceler-ation of the rate of change of level in the core does not drastically change the types of damage estimated, and preliminary calculations (to be reported in a subsequent, updating Memo to File) indicate that the depth of damage is changed only a little. Extension of the time that upper part of the core is uncoverad increases significantly the extent and depth of damage estimated and increased heat losses to steam and other heat sinks decreases or delays the damage estimated. The calculations are relatively easily repeated or the heat-up curves can simply be moved to different uncovery times as new data or interpretations are obtained to mooify the scenario of the events occurring during the time the core was uncovered during the first three hours. A faster rate of uncovery will cau;a the burst zone to move down the fuel rods, to a lower level, since the higher power regions will cvertake the upper, lower powered regions sooner in the ti.ie-temperature ramp, and they will reach the critical temcerature region of 1400 150G*F firsi.. Increased steam generation and flow will slow the rate of temperature rise by removing more heat from the fuel rod and delay the bursting. An axial power profile diffe.ent from the cosine shape assured may cause a faster rate of boil-off and uncovery of the first few feet frem the top, which will si.:ultaneously decrease the rate of temperature rise of an assembly and decrease the time to uncovery at any specific level. These a-e compensatint; effects, and whether the damage estimate changes will depend on the detailed calculations. I' kbI t

M 2 0 G73 Files The minimum damage estimated is shown.by the time-temperature plots using decay heat only (the straight lines). The maximum damage estimated is shewn by the curved lines departing from the decay heat lines at about 1600*F. The actual damage occurring for the scenario used for the plots should lie between the curve (oxidation included) and the straight line (decay heat only) for each core level, since the heat of oxidation has been approximated on the "high side" for the curved line and made zero for the straight line. An estimate of Zircaloy cladding converted to oxide can be obtained by using a modification of SUILD-5 to extend it to temperatures above its proven validity, making several assumptions as to the effects of phase changes in the oxide and comet f the specimens on the oxidation rate. This was done by R. E. Pawel ORNL) {07 at the writer's request and ';he results transmitted to him. The results cbtained for oxide thickness, alpha layer thickness, and total consumed for several different linear ramp rates and sets of assumptions lead to the estimate that the thickness of cladding converted to oxide is, in the first approximation, t. function of temperature reached in the ramp, and a function of time to reach that temperature only in the second order of approximation. Thus, the calculations indicate that between 1/4 and 1/5 of the original wall thickness has not been converted to oxide (but to oxygen-stablized alpha phase) at the time the Zr + Zr02 eutectic temperature has been reached. Further, only 1/6 of the wall tnickness has been converted to oxide when embrittlement is reached. From the core maps of damage and Table I, it can be seen that 20 assemblies did not 6xidize significantly (peripheral " corner" assemblies), none of the assemblies were oxidized significantly below 6 feet except for the center assembly and then only to 6 3/4 feet, and that significant contributions to the total amount of Zircaloy converted to oxide occurred only between the depths reached for embrittlement and eutectic forma-tion. Thus, only 4 feet of fuel rods in 40 assemblies, 51/4 feet in 56 assemblies, 6 feet in 60 assemblies, and 6 1/2 feet in one assembly were oxidized more than 1/6 of the wall thickness to contribute to hydrogen gener-ation. This then leads to the estimate that a maximum of 31", of the Zircaloy in the core assemblies could have been converted to oxide. A similar calcu-lation for the minimum damage estimate leads to the conclusion that at least 10% of the Zircaloy had been converted to oxide. Since it is probable that the "best estimate" of the level of damage is much closer to the maximum estimate, it is concluded that between 25 and 30% of the Zircaloy cladding was converted to oxide in the first three hours of the accident. This compares favorably to the 35-40% conversion estimated by others from the total amount of hydrogen present and burned immediately after the deflagration in the containment vessel at 9.9 hours into the accident. This latter estimate included any oxidation of cladding occurring after three hours. The time of the initial bursts was estimated to be at 18 to 20 minutes after the top of the core was first uncovered. If core uncovery started at 100 minutes into the accident event, then red bursting str.rted at 118-120 minutes. If the first two feet of the core were uncovered in tnree minutes, then rod bursting would have occurred at 115 minutes. The fi"st indication of fission product release at the fuel handling bridge just above the reactor vessel was found on the air sample monitor chart at 115 minutes into the accident. [} 8 [ 2

Files .g. M 2 01979 Estimates made by others(ll-14) from the fission product concentration measurements indicate that the fuel pellet temperatures of at least 30% of the fuel had to be greater than 4700*F or greater than 4000*F for significant times (several hours) if the release of fission products as to types and amounts were to be accounted for. The reaction between the aZr + Zr02 eutectic liquid and the UO, in the outer surface of the fuel pellets to produce another l'~ id phase containing equal or larger portions of UO2 could lead to release of a major portion or all of the volatile fission products present in the UO9 reacted. The Zr-0-UO2 ge diagram reported by the PNS core melt project at XfX, Karlsruhe, Gemany indicates that at 3630*F a liquid phase consistihg of approximately equal molar amounts of Zr and UO2 exists in equilibrium with a solid phase given as (U, Zr)0 =x. If the densities of U02 and Zr at 2000*C 2 can be assumed to be within 10% or so of their densities at roem temperature, then each volume of molten Zr will " melt" or dissolve approxjmately twice its volume of UO. The BUILD-5 calculations referred to earliert10) indicates that approxibately 0.013 inches of the wall remains as metallic Zircaloy when the fuel rod is oxidized in steam from one side and is ramped from 1500 to 3500*F in about 25 minutes. If this is assumed to be the case, then approxi-mately 0.052 inches of the diameter of the fuel pellet may be dissolved from the outer surface to form a liquid phase. This is about 26t of the volume of a fuel pellet. The maximum d aage estimate indicates that about 30% of the core reached temperatures allowing the liquid phase famation. This would then allow a rapid release ef about 8%,of the core inventory of volatile fission products. This would be in addition to the fission products in the gap and released by diffusional and cracking processes. If it can be assumed that the fuel pellets continued to heat up by decay heat only after the eutectic liquid formation occurred (locally), then additional release of fission product inventory could occur by two mechanisms: the increased temperature of the solid part of the pellet, and additional liquid phase femation as the temper-ature continued to rise. It is not possible from these calculations to esti-mate a caximum temperature for any of the fuel pellets. Since much of the oxidized cladding was susceptible to cracking fecm thermal shock, it would be expected that these parts of the fuel rods would crack and fragment when the primary pumo was -turned on, or when the core was being slowly reflooded from injected water. The discussion above indicates that many of the observations of the system and accident can be explained, predicted, or analyzed by the simple scenario used and the calculations made. These include most of the hydrogen generated, the first release of fission products at 115 minutes into the accident, part of the large fraction of core inventory fission products released, the flow blockage observed, and the high temperature readings of the assembly thermo-couples after the core was quenched. It fails, however, to exclain how and why the fuel assembly thermocouples survivid, when they pass through the length of the instrumentation tube from thin top of the core structure through the fuel assemblies to the bottom core strrcture, and then out of the reactor. It does not seem possible for the instrumentation tube and its contents to 00l

0m Files have survived when the temperatu"as of surrounding fuel rods must have been between 3000 and 4000*F. /. man detailed analysis and scenen is being developed to consider fuel rod behwior in different areas of each of the assemblies. Such an analysis has been examined cursorily and the resulting calculations indicate a possible mode for survival of the in-core thennoccuples. The analysis will be published in an updating " Memo to File" when completed. .. L. Picklesimer Fuel Behavior Research Branch Division of Reac+ar Safety Research

REFERENCES 1. R. E. Pawel, R. A. Perkins, R. A. McKee, J. V. Cathcart, G. J. Yurek, and R. E. Bruschel, " Diffusion of Oxygen in Beta-Zircaloy and the High Temper-ature Zircaloy-Steam Reaction," in Zirconium in the Nuclear Industry, ASTM STP 633 (1977), pp 119-149. 2. Cathcart, J. V., et.al., " Zirconium Metal-Water Oxidation Kinetics IV. Reaction Rate Studies," CRNL-NUREG-17, August,1977. 3. S. Hagan, et.al., Projekt 4241, "Experimentalle Undersuchung der Absc5mel phise van UO -Zircaloy - Brennelemeuten bei versagender Not Kuhlung," Projekt 2 Nucleare Sicherheit Halbjahresbericht, 1977/2 KfK 2600, P. 416-428. 4. T. F. Kassner, H. M. Chung, A. M. Garde, and S. Majumdear, "Zircaloy Caldding Embrittlement, Recorm: ended Criteria," presented at the Sixth Water Reactor Safety Research Information Meeting, NBS, Gaithersburg, MD, Ncvember 6-9, 1978, and published in " Memo to File" Corrents on the Symposium on Embrittlement of Zircaloy by 0xidation, 6th WRSR Information Meeting, NBS, Gaithersburg, MD, November 6, 1978, By M. L. Picklesimer, NRC, dated December 4,1978. Available at the NRC Public Document Room. 5. T. F. Kassner, et.al., LWR Safety Research Program Quarterly Progress Reports, Part III. Argonne National Laboratory. ANL-78-25, NUREG/CR-0089 October-December,1977, pp 31-44 (Published May,1978). ANL-78-77, NUREG/CR-0423, April-June, 1978, pp 19-42 (Published Cecember, 1978). 6. R. E. Pawel, Oak Ridge National Laboratory. Build-5, per se, is not reported in a formal document, but the basis is laid in.wo papers, R. E. Pawel, " Diffusion in a Finite Systen with a Moving Boundary," Jni. Nuc. Matls., 49, pp 231-290 (1974) and R. E. Pawel, "0xygen Diffusion in Beta-Zircaloy Turing Steam ']xidation," Jnl. Nuc. Matis., 50, pp. 247-258 (1974). 7. G. P. Marino, personal comunication to M. L. Picklesimer, BUILD-5 Calculations for Several Temperature Ramps, Acril 16, 1979. 8. R. H. Chapman, et.al., " Preliminary Multirod Burst Test Program Results and Implication of Interest to Reactor Safety Evalutions," Workshop on Multired Burst Tests, Sixth WRSR Informtion Meeting, Gaithersburg, MD, Novem er 6,1978, published in " Memo to File, Minutes of the Workshop on Multired Burst Testing, 6th WRSR Information Meeting, Gaithersburg, MD, November 6,1973," by M. L. Picklesimer, dated January 10, 1979, avail-able at the NRC Public Document Room. 9. L. S. Tong and J. Weisman, " Monograph on Therral Analysis of Pressuri:ed Water Reactors," ANS, July, 1970, p. 25, eq. 1.5.

10. Letter frem R. E. Pawel, ORNL, to M. L. Picklesimer, RSR/NRC, dated Acril 24,1979, transmitting ccmcuter printouts for oxidation data calculated by modified version of BUILD-5 for several tcperature ramps.

40/ 24

2-

11. " Core Damage Assessment for THI-2," Memorandum for R. J. Mattson, DSS /

NRR from R. O. Meyer, CPB/ DSS /NRR, dater. April 13, 1979.

12. Letter dated April 16, 1979, from J. Rest, Argonne National Laboratory to G. P. Marino, FBRB/RES/NRC, concerning fission product release calculations

~ using GRASS-SST.

13. Telephone conversation between A. P. Malinauskas, CRNL, and R. R. Sherry, FERS/RES/NRC, confirming estimates of temperatures required to release amounts and types of fission products observed.

14. Telex, dated April ll,1979 from R. A. Lorenz, CRNL to R. O. Meyer, CPB/

DSS /NRC, reporting calculations by Malinauskas of temperatures required toreleas the amounts of fission products observed at TMI-2. 15. P. Hofmann, et.al., Projekt 4244, "Konstitution und Reaktionsverbalten you LWR-Materialieubeim Corechmelzen," Projekt Nukleare Sicherheit Halbjahresbericht, 1974-2, KfX 2130, pp. 261-275, Kernforschungszentrum, Germany. 9 O 48~/ 2?'t(

TADLE I DAMAGE ESTIfMTE FOR Tile Till-2 CORE AT IllREE il00RSb Penetration Deptli frosi Top of Core (feet) liAXIMUM DAMAGE ESTlHATE HINIHUM DAMAGE ESTIfMTE lotal Peaking factor, (tpf) 1.9 1.6 1.3 1.19 0.66 1.9 1.6 1.3 1.19 0.66 Damage Type Burst 1 1 1 2 3 1 1 1 2 3 0xidized but not 1/2-embrittled 7-1/4 7 6-1/4 6 none 7-1/4 6-3/4 6 5-1/2a none only Embrittlement 6-3/4 6-1/4 5-l/2 1/2r none 6-1/4 S-3/4 1-1/2-2 4a none 5-1/4d Sa only only only Eutectic formation 6-l/4 5-3/4 5 1-none 1-1/2-none none none none ? 4-1/2 4-1/2d only only a. Occurs orily between these levels. b. Estimates as of S/1/79 50 (.n.D -a-C? Oc LT D C1 pt. x T": p{h: g. .;; 9 'h:r r

FIGURE 5 1-13 COMPARISON OF MEASURED AND CALCULATED TOTAL CORE PO'n'IR DISTRIBUTION RESU:.TS AT STEdDY STATE, EQUILIBRIUM XENON 92.6 % FP CONDITIONS Measured Calculated Control Rod Group Positions Gps l=4 100 100 t we S 100 100-t wd Gp, 6 Gp 89 87.5 5 we Gp 7 89 87.5 % wd Gp 8 16 18.3 6

• G Core Power Level 92.6 92.6

% FP Baron Concentration 1o95 1135 PPM Core Burnup 32 23.2 EFFD Axial Imbalance + o.55 + 0.40 % FP Max Quadrant Tilt - o.10 t Time 1o48 Cate 1/9o/75 8 9 10 11 12 13 14 15 ^ 1.88 1.75 1.47 L 66 1.45

4. 6a 1.82 1.32 H

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$/ l /fflfY/YY/b***"E - 4: \\ l ?/ W//Y////////fl/ / / / / l 2.2 I ?@^wn@ MW4MWAWW RWGM@VfAMMW _ 6' N M @ r#F2 mms.3 ' /@reeCei$AMMN @@fW4 M _ r 1@ WAM s FIGURE 5 No si:nificar.: oxdiation ESTIMATED MAXIMUM DAMAGE TO TMI-2 CORE AT TIME =3 HOURS Maxi =us da= age estimated to fuel red cladding. Decay F8#*1Y ** heat and oxidation heat included. Heat loss to steam not br M e ar.d " cold" rods set at 25P. of total of decay and oxidation heat appropriate for te=perature,esti=atec h Oxidi:ed to br:::;eness 0 ex:de thickness and power level at eawh position en .,,,, 3 3 7 fuel rod. Elevations in feet from tcp of core. gGa ud ' i

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== @N'i,D;t.iNLATQ"" fF1 PIN I i FIGURE O No siz.1 fica.;- oxdtatic.; ES!! MATED MAXIMUM DAMAGE TO TMI-2 CPI AT TIME =3 HOL'RS Ma.ximus da= age estimated to fuel red cladding. Decay ?*#~ ' Y #

• I*

heat and oxidation hea: included. Heat loss to steam U ~ and " cold" reds set a: :=r. of :otal of decay and oxidation heat acpropria:e for te=perature,estima ed , o,;

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" 3 3 ,, ; } cxide nick.ess and pewer level at eacn pos:::en on fuel red. Elevations in feet fres top of core. yQ /)f d y ~Q / 7 r W;' ~4]7*}j[

11 6.. g' r ..=.. ,5s 7 2 g y @@e 6 c@ @l@'@ Ce@ c ,i, op m. m - \\95'J W @ @ @ 1W @9 dbl_@le @@@ C ~ 1hs 9@@ @ R@ @ @'@ E i@Be@ @ @@@@@ 6' 4l @h@@ @@@M s' /~M' @iB @ @$39 ing@L @ @ @@' go RM 9 @ 4' FIG M 10 No si:nificant ESTIMATED MINIMUM DAMAGE TO TMI-2 CORE AT TIME =3 HCURS Partly oxidized but Minimu:s da= age estimated to fuel rod cladding. Decay heat only. Oxidation heat not included.Hea: loss to not bri::le steam and " cold" rods set at 05P. of decay heat at-0xidi:ed to bri :1 each position on fuel rod. Elevaritns in feet from g top of core. and/crover2500;eness F d '* *I '~' up:ured s n/,,. g e g =id p$ so

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• b ISTIMIED MAXIMLIM DAMAGE y y,j gj)l'y]en)

TO TMI-2 CCRE CLADDING AT 3 HRS. Decay ~ oxidation heat. j, f r p,,/u,'j f,fe, Heat loss at 25%. Decay 4cg heat at 1% full power REACTOR YESSEL & lHTERNALS-GENERAL ARRANGEMENT Elevatior ' feet from THREE MILE L5 LAND M'CI EAR STATICN 1; NIT 2 top of ccre. ,n np om n p1g* -----~ U* & DK nGu=E u-s k k k)di.idiiidlaa ~ P" d.o7 os c0 4 { APPENDIX A \\ CALCULATIONS: 1. Heat Capacity of the Fuel Rod 3 UO : density = 10.08 g/cc (92.5% theoretical density) = 0.3642 lb/in 2 O specific heat = 0.077 Stu/lb - F 3 vol/ inch of rod = { (0.370)2 x 1 inch = 0.1075 in / inch of rod mass UO / inch of rod = 0.03915 lbs/ inch of rod 2 -33tu/in 0F AH mass x specific heat = 0.03915.x 0.0777 3.042 x 10 UO 2 Zircaloy Cladding: 3 density = 6.54 g/cc = 0.2363 lb/in specific heat = 0.0830 Btu /lb OF 600-1100K 0.1315 Btu /lb OF 1100-1300K 0.0908 Stu/lb F 130CK assume = 0.0830 Stu/10-T 600-23C0K vol = - x (0. A10 + x 0.026 = 0.03308 in 3 2 -3 mass = 7.817 x 10 IN/ inch of rod ic heat = 7.817 x 10-3 x 0.0830 = 6.488 x 10-# 2H = mass x spec-Zr Stu/ inch rod-F Heat Capacity of Fuel Rod / Inch of Road 2H = 6.488 x 10'# Stu/ inch of rod-F Zr aH 30 42 x 10 Stu/ inch of md UF UO 2 aH = 25. 91 x 10 Stu/ inch of rod UF rod l h i e h ! '(b f . 2. Creep-Rupture of Zurcaloy Ciad Fuel Rods: Minimum internal pressure: from FSAR: total internal volume of fuel rod (free) with pellets = 1.75 in3 volume of the plena 3 = 0.87 in volume of gap + pellet dishes 3

0.88 in cold prepressurization = 445 psi He (300K) n = PV/RT

= 1.306a A, A = conversion facer .987 300 assra: 1. Top h of rod + upper plenum goes into film boiling. 2. The time is between 1 and 3 hours after turbine trip. 3. The hot rod peak clad temperature is 1470 F (800 C) U 4. The total number of moles of gas is constant n = n; + n2 = 1.2064 A 5. The fuel rod is separated into two parts, each at a different temperature, and connected by a smagl pressure capillary, the two parts having volumes of 1.09 in at 589K and 0.67inJ at 1073K then: P x 1.09 A P x 0.67 A " " "I + "2 = 1.3064 a. 1.987 x 587 * 'I.987 x 1073 = P = 1049 psi at 1470 F (800 C) . Maximum int mal pressure: from B&W estimate, internal pressure = 1200csi at 589K(operating conditions) 3 n = PV/RT = Axl.75 in x 1200 psi /1.987 x 589 = 1.7827 A n = n1 + n2 = ^1.987 x ce:09, A x P x 0.g = 1.7827 A

• 1 9

1.987 x 1Ces U P = 1436.5 psi at 1450 F (788 C) 48[

. Circumferential stress in the wa"11 of the cladding: a=2t

AP = internal pressure - external pressure D = 0.380 inches t = 0.0265 inches o = 7.1698 AP 1600 F for aP = 300 psi

, o = 2151 psi, T = rupture = 400 psi = 2868 psi, l= 1575 F 1540 gF = 500 psi = 3585 psi, = g 1500 F = 600 psi = 4302 psi, = 3. Axial Power Profile of Fuel Rods: Feminal axial peaking factor in Rancho Seco = 1.3 P g) = B + A cos (w z/2L) P**9

A,8 = constants reldted to the power 0*

levels in the core Pmax = A + B, P,yg = B + 2A/v ;g = axial peaking factor - 1.3 Pyx avg tnen at P - 1.3 P,yg, A = 1, and P = B + 1 = 1.3 3 + 2.6/w; 3 = 0.5746 max max 0.5746 + ces (v z/2L) then P(z) =L 0.5746 + 2/s avg p P(z) = 0.4744 + 0.8256 ces (x z/2L Pavg Table A-1 Calculation of Axial Power Pmfiles Total peaking Pavg{ core) Axial peaking Pavg/ft(md) factor Icwf. factor kw 0.66 6.C 1.3 3.046 1.19 6.0 1.3 5.49 1.30 6.0 1.3 6.0 1.50 6.0 1.3 7.38 1.90 6.0 1.3 8.77 48/ 26y

P(z) at 100% Power Pavg L_ff z/L 0.8256 ces (sz/2L) 3.046 5.49 6.0 7.38 8.77 2 6 0/6 90 0 1.445 2.604 2.85 3.50 4.16 6 1/6 75 0.2137 2.096 3.777 4.13 5.08 6.03 0 6 2/6 60 0.4i28 2.702 4.87 5.32 6.55 7.78 6 3/6 45" 0.584 3.223 5.808 6.35 7.81 9.28 0 6 4/6 30

0. 71 5 3.625 6.53 7.14 8.79 10.44 6

5/6. 15 0.7975 3.874 6.983 7.631 9.39 11.15 0 6 6/6 0 0.8185 3.956 7.14 7.80 9.63 11.40 4. For graphical solutions of temperature vs. time in rod for decay heat only: plot h ( F/sec) vs kw/ft inrod; rod = 3.69 x 10-3 Btu / inch OF aH f Pcwer in red = kw/ft x ft/ inch x 0.9846 Stu/kw-sec x 1/Stu/ inch OF = Ah rod 0.984(2 f 'raction 3 x Gw/ft at pcwer) = 22.4 (decay = fraction x pcwer) for 1% dacay heat, and 10 kw/f t power, f = 2.24 F/sec 0.75% decay heat and 10 kw/ft pcwer, h = 1.68 F/sec 0 0.St riecay heat, and 10 kw/ft ;cwer, k = 1.12 F/see then: T,t *Isat + ( t) where T = temperature of (:) at time of uncevery z sat t=t + at where t time of uncovery of (z), and, = g g a t, ime increment of calculation taphical solution of oxidation heat vs. temperature, use heat generation rate for time increments of 2 minutes as calculated by SUILD-5 for 0 temperature ramos frem 1500 to 3000 F in 20, 25 and 30 minutes plotted against temperature at the and of the time increment. See Figure A-1 [4($[ 2 0 For each 100 F step above 1500 F, add the oxidation heat at that temperature to the decay heat for that rod, enter Figure A-2 at the power sum, and draw the appropriate slope frem the icwer temper-ature of the increment to that temperature plus 1000F, or read incremental time required to heat that 100 F temperature incremer.t. 0 and extend the rod temperature vs. time plot at elevation z by that amount. Repeat in steps until the temperature for the fema-tion of the alpha-Zr + Zrt2 eutectic is reached. Mathematical description: T9 j=Tj + 100 F ti+1=ti+# T T T = aH. + aH j,y 9 g i AH is taken fmm plot of kw/ft vs. temperature for T J i ramp rate concerned and shown in Figure A-1. 4 i;g] L60

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D a n-sao Jul. 3 1979 ADDRESSEES - LETTER DATED M. L. Picklesimer, Chairman, NRC/RSR G. L. Bennett, NRC/RSR F. D. Coffman, Jr., NRC/ DOR G. P. Marino, NRC/RSR R. Van Houten, NRC/RSR D. A. Hoatson, NRC/RSR R. R. Sherry, NRC/RSR W. V. Johnston, NRC/RSR R. J. Budritz, NRC/RES R. O. Meyer, NRC/ DSS J. A. Norterg, NRC/OSD J. Voglewede, NRC/ DSS D. Powers, NRC/ DSS L. S. Rubenstein, NRC/NRR TIDC Accessions Unit /PDR (2) T. E. Murley, NRC/RES R. Scroggins, N",C/RES (50-320)/ P. O. Strom, NRC/ID L. S. Tong, NRC/RSR P. Boehnert, ACRS D. Bassett, ACRS R. Fraley, ACRS P. A. Shewman, ACRS Subcommitte on Fuels R. N. Oehlberg, EPRI J. T. A. Roberts, EPRI R. H. Chapman, ORNL D. 0. F:dion, ORNL C. S. H: 7tley, Univ. of Florida T. . Kassner, ANL S. Dagbjartsson, EG&G P. E. MacDonald, EG&G M. P. Bohn, EG&G L. G. Ybarrando, EG&G E. L. Tolman, EG&G R. R. Hobbins, EG&G H. M. Zeile, EG&G C. R. Hann, BNWL C. Mohr, BNWL A. Fiege, KfK M. Fischer, KfK J. S. Goedkoop, ECN-Petten S. E. Ritterbusch, CE P. G. Smerd, CE C. Crouthamel, Exxon .- so._ _@'I w~ r CM lh h)

ADDRESSEES - LETTER DATEp JUL 3 1979 A. L. Lowc, 3&W M. H. Montgomery, B&W L. D. Burman, W R. B. Adamson, GE, Vallecitos E. D. Hindle, UKAEA T. Healey, CEGB S. Kawasaki, JAERI M. Ishikawa, JAERI E. Zebroski, EPRI G. Folwer, NRC/0MPA A. DiPalo, NRC/0MPA R. E. English, Presidential Comission R. Bernero, Task Group #5 .}}