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Q: Is there any source in the literature of nuclear science that poses the contention? | Q: Is there any source in the literature of nuclear science that poses the contention? | ||
A: Not exactly. However, the diagram called Attachment A here j essentially contains the contention. The diagram was taken l from the publication ORUL-NSIC-23, "Fotential Metal-Water Reactions in Light-Water-Cooled Power Reactors", August, 1968, by H. McLain. That publication took the Figure from an article, "A Review of Generalized Reactivity Accident for Water-Cooled l | A: Not exactly. However, the diagram called Attachment A here j essentially contains the contention. The diagram was taken l from the publication ORUL-NSIC-23, "Fotential Metal-Water Reactions in Light-Water-Cooled Power Reactors", August, 1968, by H. McLain. That publication took the Figure from an article, "A Review of Generalized Reactivity Accident for Water-Cooled l | ||
l | l | ||
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fetkn &c n G, f . ,f | fetkn &c n G, f . ,f | ||
=n | =n ATTACHIE"2 3 | ||
ATTACHIE"2 3 | |||
/ d . l d. /-lo L T i C t M ' ' | / d . l d. /-lo L T i C t M ' ' | ||
a' 8%t)$h.$ | a' 8%t)$h.$ | ||
| Line 342: | Line 339: | ||
h,M1 y m..=+= M d | h,M1 y m..=+= M d | ||
y NN m:;;;;w.E j; s 52 c | y NN m:;;;;w.E j; s 52 c | ||
;1 :re'''' W l';' ?! d.0,'c i 4' a | ;1 :re'''' W l';' ?! d.0,'c i 4' a | ||
' L , | ' L , | ||
| Line 424: | Line 420: | ||
. k RI A-ST-3 17 m 17 PWR 5.3 0 1.076 300 225 250 Did not fail O j; $ RIA ST-4 15 m 15 PWR 20 0 1.48 695 350'8 5308 Completely destroyed; pressure E O d i- pulse of 35 MPa measured 1! > | . k RI A-ST-3 17 m 17 PWR 5.3 0 1.076 300 225 250 Did not fail O j; $ RIA ST-4 15 m 15 PWR 20 0 1.48 695 350'8 5308 Completely destroyed; pressure E O d i- pulse of 35 MPa measured 1! > | ||
;, RIA I I Two Saston 5.7 4600 1.13 365 285 330 Complete shroud flow blockase j | ;, RIA I I Two Saston 5.7 4600 1.13 365 285 330 Complete shroud flow blockase j | ||
;I Two Santon 5.8 0 1.077 365 285 315 Severe failure; partial now e-l blockage $ | ;I Two Santon 5.8 0 1.077 365 285 315 Severe failure; partial now e-l blockage $ | ||
es | es | ||
| Line 562: | Line 557: | ||
. Jack R. Newman, Esq. | . Jack R. Newman, Esq. | ||
; The Several Intervening Parties e, es Docketing and Service A , | ; The Several Intervening Parties e, es Docketing and Service A , | ||
Res ectfully, 4 | Res ectfully, 4 | ||
/ | / | ||
Revision as of 04:47, 18 February 2020
| ML19345H112 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 04/20/1981 |
| From: | Jeffrey Scott TEXAS PUBLIC INTEREST RESEARCH GROUP |
| To: | |
| References | |
| NUDOCS 8104300511 | |
| Download: ML19345H112 (28) | |
Text
{{#Wiki_filter:. 7" ' 'I April 20, 1981
,h ,
bt li t UNITED STATES OF E'. ERICA i[.3 "*i'R tea *@"3 9%98> t I'UCLEAR REGITLATORY COFCIISLION
\'.A 4'
4 m - JFORE THE ATOMIC S/3ETY AIID LICESSING BOARD In the Matter of ) HOUSTOli LIGHTI'.!G & PC1:/ER COMPANY , Docket No. 50-466 (Allens Creek Huclear Generatin6 Station, Unit 1) 6 4 A cood*
# ge g INTERVENOR DOHERTY TESTI".CNY 4 9"'
- g 3 '5 iS B\ " '--
OF JMIES M. SCOTT - " secdL1 1 L 0 k W 638 ON DOHEETY CCliTEHTION ?-3 d
?
INADECUACY OF THE DESIGH SAFETY LIMIT o, OF APPLICANT'S FUEL RODS 1 Q: What is the purpose of your testimony? P00R ORIGINAL A: The purpose of my testinony is to show that, as alleged in Doherty Contention #3, the design proposed for the fuel system for the ACUGs will prsant an undue risk to the health and safety of the public and Mr. Doherty. Doherty Contention #3 states: The design safety linit of thermal enerr for each fuel rod is too high for fuel rois which will be in a cluster such as that on two General Electric 9/ proposed for ACliGS.
'6 in. outside Tests diam eter zircaloy rods, that had been irradiated aoproxi- .
cately one third (1/3) of the time a fuel rod troically is irradiated indicate that the cladding will rupture at between 147 calories / sram of uranium oxide fuel to 175 calories / cram of uran-ium oxide fuel if a power encursion (Heactivity Initiated Accident), occurs. The rupture of the rods i means there is danSer of:
- a. Fuel fragments escaping into the coolant from $O the rods due to the pressure of gases escaping hp through the rupture, L
6104300 W b. Pressure pulses froa fuel in contact with tue
/ water after it escapes from the fuel rod,
\
- c. Serious weakening of the cladding stren-th of the rods after ructure. Such rupturin;; means the rods vould not have sufficient resistance to with- .
stand normal pressure actions of the circulating coolant or disturbances due to the oower excursion (RIA) itself. The weakened rods would be bent out of allignment resulting in excess fissioning leading to further excess reactivity at the location where the rods are distorted and ruptured cladding, with results such as (a) and (b).- , I
- d. Fuel from damaged rods may prevent function of the control rods by jamming inteystices between the control rods and the reactor bottom.
3ecause ACUGs has more compact rods in each fuel bundle and a higher power core density than any operating 3';IR in the United States, petitioner requests the design
- safetylimitofthefuelrodsbeloweped,andvarious parameters in the reactivity control system be altered in accordance with this change.
Q: Do you agree with this contention? A: Yes, except for the statement in the;first part that the [ 'i ' fuel,rodsihadcbeen irradiated approximately one third (1/3) d d htE[btiEeb,,buel rod typically is irradiated. My testi-mony will show the irradiation approximately one tenth (1/10) to one fifteenth (1/15) of the amount a fuel rod is ty7ical17
~
irradiated. Q: Have you attached your personal qualifications? A: Yes. These are Attachment F to this filing. Q: Is there any source in the literature of nuclear science that poses the contention? A: Not exactly. However, the diagram called Attachment A here j essentially contains the contention. The diagram was taken l from the publication ORUL-NSIC-23, "Fotential Metal-Water Reactions in Light-Water-Cooled Power Reactors", August, 1968, by H. McLain. That publication took the Figure from an article, "A Review of Generalized Reactivity Accident for Water-Cooled l
l
. . 1
. l 1
l l and -Moderated UO 2 -Fueled Power Reactors, Nuclear Safety, 8 (2): 116-127 (Winter 1956-1967), by G. O. 3right. 1 Q: What is the " design safety limit" for fuel rods for the ACUGS7 A: The " design safety limit" for fuel rods for ACNGS and all BWR and P'JR reactors for production of electricity is set at 280 cal /gm. of UO 2 fuel as set in Regulatory Guide 1.77 of the NRC. This limit was arrived at largely throuSh re-view of the Special Power Excursion Reactor Test (SPERT) re-actor tests of 1968 through 1970. The requirement is that in the worst case, "Reactivit7 excursions will not result in
- a. radial avera7e fuel enthalpy greater than 230 cal /gs. UO 2
at any axial location in any fuel rod." (Regulatory Guide 1.77) Recently it has been recommended that a radial aver-age peak fuel enthalpy below approximately 240 cal /gm. would assure, "[N}either severe fuel-rod da= age nor loss of geometry would occur", (MacDonald, et al, " Assessment of Light-Water-Reactor Fuel Damage During a Reactivity-Initiated Accident", Nuclear Safety, 21(5) 582-602, Oct. 1980, p. 602) and that this is the appropriate desi 5n safety limit. That "The SPERT test rods, which were subjected to a total energy deposition of 280 cal /gm. UO 2 actually reached a peak radial ayerage fuel j enthalpy of approximately 230 cal /gs. UO2 . ,(Ibid. p. 583), explains this change in thinking about this limit for the re-searchers who authored the Nuclear Safetv article. The 280 cal /sm. linit is currently in use as the limit to the control rod drop accident under any conditions in the operation of Quad Cities Station, Unit 1, an 800 MWe BWR, (Amendment 61 to DPR-29, December 5, 1980, cperating License for Quad cities Unit 1) and Vermont Yankee, a 514 MWe BWR, (Amendment 55 to DPR-23, Oct. 26, 1979, Operating License for Vermont Yankee Nuclear Power Station), however. NUREG 75/087, the Standard Review Plan, (" Fuel System De-sign", Sec. 4.2, p. 4.2-4) states that, " ... (oj bserving the P00R ORIGINM ___.___,_,m
~
f limit... crevents widespread fragmentation and dispersal of the fuel and avoids generating pulses in the primary system during an RIA." The li=it thus defines the maximum heat energy that can be safely inserted in fuel rods when the power level is changed accidently. This is to assure the fuel rods do not rupturu and discerse fuel. Reactivity has been defined as the "... percentage excess of neutrons per fission that will cause fissions in the next generation" ( Webb, R. E. Accident Hazards of Nuc-j lear Power Plants, 1976, p. 13) or, "A measure of the de-parture of a nuclear reactor from critical (i.e. from a state where a self-sustaining chain reaction can take place), such that positive, values correspond to reactors above crit-i ical." (Sarbacher, R. 'A Dictionarv of Electronics and Nuclear Eng. 1957 Q: What are the tests described in the contention? A: These were tests performed at the National Reactor Test Cen-ter in 1969-1970 in the SPERT series, and reported in a limi-ted distribution oublication, (IN-ITR-113, "The Effects of Burnup on Fuel Failure, Part 1 - Power Burst Tests on Low Burnup UO Fuel Rods," 1970, p. 20). In test 568, a fuel 2 235 enri hment (called rod 5/16 in,outside diameter and 7% U a GEZ rod) ruptured at 147 cal /gm. causing a pressure pulse i of 165 p.s.i. Whether this 147 cal /gm was a measure of rad-i ial average peak fuel enthalpy or radial avera6e total energy ! deposition is not reported. The rod had been irradiated to 3,480 mega-watt days / metric tonne of UO2 fuel (mwd /t). Burnup of fuel reaching 45,000 mwd /t is now being attemp-ted at the Monticello plant, a 536 MWe BWR (Amendment-42 to DPR-22, Monticello Nuclear Generatine Plant, Table 3.11.1, Dec. 23, 1979) having passed a calculated 40,000 mwd /t. Burnup sreater than or equal to 45,000 mwd /t is planned for the ACNGS. (See Staff Reply to Question 12-20-10, Doherty Interrogatory Set #12, Feb. 19, 1980). The contention is incorrect in its reoresentation of the outside diameter of the rod. However, in all other respects 4 2- ()lll[illilil l>Ill)ll
. - _ . , . - . _ . _ . . - . , _.m..- ..m,-,, , _ - . -__..w ...,__m,,_.. ._..__...m,.,._.,,__..m ,,.,,..--,_,,,-.._....._m. _
A 5-but this and enrichsent, the rod in test 568 was a typical General Electric 3WR fuel rod. In the final analyis of this test, the internal crassure in the fuel rod, and not celt of fuel was considered the cause of the rod rupture. (IN-ITR-113,
- p. 49)
Q: ilhat were the results of other tests conducted in the SFERT . series? A: These were tests similar to the SPE29 tests in the conten-tion, but they used fuel rods with burnups of 13,000 and 32,000 mwd /t. At 7% enrichment, there were four tests per-formed, although the ori-inal plan was for 12 test excursions. (IN-ITR-118, "The Effects of Burnup on Fuel Failure, II. Power Burst Tests on Fuel Rods with 13,000 and 32,000 mwd /MTU Burnup", R. W. Miller, Dec. 1970, p. 1). One of the t;wo rods irradiated to 32,000 mwd /t,"had three fractures which to5 ether extended over most of the active length of the rod, and one ruptured blister."(Ibid. p. 16) This happened at 85 cal /gm. The second
. rod failed at 176 cal /gs, which disagrees with the result for lower burnup fuel given in IN ITR 113. These too were "GEX" rods, differing only in outside dianter and enrichment from ty7ical General Electric BilR fuel rods. In the 85 cal /g test there was no evidence the rod was waterlogged, because the measured pressure pulse from the transient was small compared to what '.tould be expected from a waterlogged fuel rod. (Ibid.
- p. 25) The energy deposition data was renorted later to have an absolute uncertainty of
- 12% and relative precision be-tween ceasurements was stated then to be about
- 25. (Report RE-8-76-187 (RAM 585-76) " Licht-Water Reactor Fuel 3ehavior Program Descriotion: RIA Fuel Behavior Experiment Require-cents", L. B. Thompson, D. L. Eagerman, P. E. MacDonald, Cet.
1976. l Other tests in the series attengted to determine the effect
, P00R ORIGINAL
of encursioning several rods together in an assembly such as in an actual reactor. The failure threshold for unirrad-iated fuel rods as measured at SPERT was 225 cal /gn. But, groups of five unirradiated rods in a cannister (which is I 1 similar to a BUR fuel channel box) showed a.5% to 10% decrease in threshold for fuel destruction under those coa:itions, that is 240 cal /cm to 225 cal /gs. (IR-ITR-116, I
"The Response of Fuel Rod Clusters to Power Bursts", L. J.
Siefken, May, 1970, p. 42). The rods used were "SPZM" rods four sixteenth of inch outside diameter,- with enrichment of 7% to 10 5%. The standard General Electric fuel assembly l consists of an enshrouded 62 fuel rod arransement. The use of more rods per assembly may result in rod destruction at lower reactivity insertion levels. The 5 rod asse=blies used in these SPERT tests ha1 the same ratio. of rod center to center distance and rod outside diameter,1 3, (Ibid. p.1) as those described in the GESSAR (Nure5-0152, Table 4-1,
- p. 4-3, March 1977) fuel assemblies.
l l Because ther,e are errors in two publications, NUREG/CR-0269 where the fact of fuel rod failore'(the publication's term) for the 85 cal /g test is omitted from Table C-IV, at p.144, and in Iluelear Safety, 21(5) Oct. 1980, p. 532-602, " Assessment of Light-Water-Reactor Fuel Damage During a Reactivity-Initiated , Accident" P. E. MacDonald, et al, at p. 585, Fig. 2, fote pages of IN-ITR-118 are presented as attachment B. Q: You mentioned the enrichment as higher than operating BJR fuel enrichment. What difference does enrichment cause in the results of these tests that might make their applicantion to AO:iGS erron-ecus? A: It is reported that initial fuel rod failure' threshold decreased with increased enrichment. Thresholds were determined to be: 265 to 277 cal /g UO 2 f r 5% enriched fuel, 254 to 264 cal /g UO2 for 10% enriched fuel, and 232 to 246 cal /g UO 2 f r 20% enriched fuel. ( NUREG/CR-0269, " Light Water Reactor Fuel Resconse During Reactivity Initiated Accident Experiments", Fu.jishiro, T., John-son , R. L. , MacDonald, P. E. and McCardell, R. K. , Aug.1978., p. 27. mm. x P00R ORIGLNAL i
Q: How ai-ht reactivity be inserted in an amount exceeding the thresholds discovered in these earlier tests, in the ACHGS? A: The reactivity excursion test results, as applied to Regul-atory Guide 1.77, have to viewed critically. The design safety limit was 280 cal /gs, but this gives full credit for heat enerry transfer to the fuel and does not include energy from the delayed neutrons. (Uuclear Safety 21(5) 563-603, Oct. 1980, p. 583 (Hereafter: Huclear Safety Article). The design safety limit for 3NR's under zero and low power conditions is 170 cal /sm " radial average energy density" (NUREG 75/087, " standard Review Plan," sec. 4.2, p. 4.2-4) and this document does not indicate the two factors which make the 280 cal /sm limit too great (heat transfer to coolant, and failure to include energy from the delayed neutrons) are factored into the~ term " radial average energy density". In Table 6 a of NEDO-20,626, " Studies of EWR designs for mitigation of ATWS,' Oct. 1974", the peak enthaloy for the main steam isolati n valve closure ATWS is 150 cal /En, which when considered a low power event and the failure to account for delayed neutrons means a rather small margin if not actually exceeding the low power criteria of the Standard Review Plan. An URC sponsored research group considers the most. severe Reactivity Insertion Accident to be a rod drop accident during l ;startfup.4 .("Ph(sics Analysis of Power Burst Pacility Test l
- RIA'1-3 (PNL 5ardware)", by E. H. White, Interin Report, 1
December, 1973. In the Nuclear Safet? Article, p. 601, MacDonald et al. suggest that a limit of 233 cal /gm should have been chosen
- instead of 230 cal /gs. It follows then the 170 cal /cs limit should be crocortionately reduced as well. On page 15 1-166a l of the Supp. Uo. 3 to the Montacue PSAR, a beginning of life l core enduring a rod drop of 5 ft/sec, with a worth of 0.01435 (the most severe conditions) yields a calculated peak enthal7y t
of 221.96 cal / cm. This sc.uares with results in Chaoter 4, l P00R OR G NAL
- p. 25, of NZD0 10,527 "2cd Drop Accident for Large 3?Rs",
where a 0.0123 worth rod dropped at a rate of 5 ft/sec inserts 2a5 cal /sm. In this same report, at 3 5 gigawatt . days /bonne core exposure a 0.0129 worth rod dropped at 5 ft/sec inserted 239 cal /ga. (Ibid.) NFD0 21,231 " Banked Position Rod ilithdrawal Sequence", Jan. 1977, p.2-1/2-2, lists a worst case for the highest rod worth (0.012 rod worth) would produce a " peak fuel enthalpy of 232 cal /cs in a desisn based control rod drop accident." The Staff in its review of ACNGS (Gafety Evaluation Report Supp No 2, p. 15-4) applied data from NEDO 10,527 Supp No. 1 which showed, " . 4a]value for the inserted reactivity worth which would produce a resultant peak fuel enthalpy greater than 280 calories per gras" wb$ch is "... greater than 0.013 ok/k." Hence it may be inferred that reactivity insertionh such as of 0.012 rod worth =ay well exceed the point at which fuel damage other than loss of hermeticity occurs because the reactivity insertion will be close to the 230 cal /gs, which MacDonald et al. (Nuclear Safety Ar-ticle sug est should be lowered. And, accordig to the Staff (Reolv to Doherty's 1ath Set of Interrogatories, question 9, June 9, 1980) if the i sost adverse en=bination of by-passed rods is assumed, 232 cal /ga is the energ7 yield of a rod drop accident. 0: 'fhct are the eroerimental results of the recene og a l the Power Burst Facilit and other recent tests? A: Most recently, URC s;onsored research has returned to the Reactivity Insertion Accident Et the Power 3urst Facility in Idaho. Unfortunately there have been no excurions of fuel rods irradiated core than 5,000 mud /t, so the SpIRT tests at 13,000 mud /t and 32,000 MWD /t, few as they are . 1 - remain neither confirmed or denied. According to Cdekirk (ANOR-1095 " Detailed Test Plan Report for Power Burst Facil-ity: The Behavior of Unirradiated FUR Fuel Rods under ?CMA Conditions", 4/74) the Power Burst Facility (P3F) tests will
,(e using fuel rods that differ from ACNGS rods in fuel en-richhent;and neutren s7ectrus during irradiaLion prior to " ' 'e~x' cur'sich' testing. ?uel enrichment causes differences as
_9_ shown above, which are not great. Neutron spectrus is a different matter. In the SPIRT tests, "... (a] 7t pical of SURs was the slightly lower fast-to-thermal flux ratio (between 0.15 and 0.20) which resulted in a slightly less than typical fast neutron dose to claddiag" during pre-irradiation prior to excursion testing.(NUREG/CR*0269,
" Light Water Reactor Fuel Response During Reactivity Initiated Experiments", Fujishiro, T. , MacDonald, P. E. , Johnson, R. L. ,
and McCardell, R., August, 1978, p. 35)'Taking this fact, one critic in arguing the current NRC design safety limit inadequate stated, " Fast neutrons dacare the claddin- by making it brittle, which suggest even more strongly the possibility of crumbling. Hence the cladding was not as damaged from this standpoint as 1, would be in an actual commercial reactor." ( Accident Hazards of Huclear Power Plants, Webb, R. E., 1976,'p. 54) The,recent test reports I have seen do not indicate if irradiation in the Sarton reactor irradiates uith a neutron spectrum m' ore like one fuel rods in the ACNGS will obtain. However, the quote- above by R. E. Webb has possibly become prediction, because data fron the PBF, indicate, "The consequences of failure at Boiling Water Reactor hot-startup s7 stem conditions appear to be more severe than previously observed in either the stagnant caosule SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed exten-sive variation in clad wall thickness (involving considerable plastic flow) and fuel shatterinr along rsin boundaries in both restructured and unrestructured fuel rer,-ions. Oxidation of the cladiing resulted in fracture at the location of clad- ! ding thinning and disintegration of the rods during quench."
- (Light Water Reactor Fuel Response During Reactivity Initi-ated Experiments", MacDonald, P. E., McCardell, R. K., Mar-tinson, Z. R. and Seiffert, S.L., from the International Collocuiun on Irradiation Tests for Reactor Safety Programs, June 25-23,1979, Petten, The Netherlands, CONF-790646-08)
P00R ORIGINAL
- - - , - . . . , . . , . , , , . , - . . . , . , -,,ww ,----,,.-,--o----,-
These saae authors state that "Che results of these test (RIA tests at 73F) indicate that, whereas the failure threshold for unirraf.iated and irradiated fuel rods of
<*22; and --140 cal /g UO2 (radial averace. peak fuel enthalpy) are generally consistent with previous S?ERT and USRR results, the consequencas of fuel rod failure.
at BWR hot startuo conditions are somewhat more severe than those observed in either SPERT or USRR. (Nuclear S2fetv Article, p. 591) . In test RIA 1-2, four fuwl rods were subjected to 5,000 HU4/t burnup in a cluster arran5ement and then were sub-decced to a power burst of 180 cal /gs. A sin 51e rod that had not been opened prior to excursioning failed at 135 cal /g
' UO fuel and had what appeared stress-corrosion cracks.
2 (Overview of Recent Power Burst facility (PBF) Test Results", , Paper Presented at the Seventh Water Reactor Safety Research Information Meeting, Nov. 5-9. 1979, By J. J. Zeile, Idaho l Hational ens i neering Laboratory) Test RIA 1-1, showed that the current 230 cal /gm design safety limit is inadequate. In that test, radial average peak fuel enthalpy of 285 cal /ga. produced complete flow l blockage in an encapsulated core within four seconds of the power burst. (Nuclear Safety Article, p. 594) A 256% expansion of fuel volume was required to block coolant in the P3F capsule. (Ibid.) See Attachment C for su= nary of the P3F tests.* Q: Have the results of any of these tests been. interprete,d to - rec.uire that the design safety limit be reduced? A: In the Nuclear Safetw Article mentioned previously, MacDonald, et al., finished by saying, " . . . [n] either severe fuel rod dam-age nor loss of normal geometry is expected at radial average i peak fuel enthalpy below 240 cal /g. Therefore, we conclude that an RIA in an LUR would pose no real safety concern, but l the URC licensing criteria sho,ul_i be r.9-el(aluated. (Ibid. p. 602) l This is Sacked further by a memorandun from D. A. Hoatson of 1
- Taken from the Nuclear Safety Article, p. 592.
J F'lll)ll [llllGlli;1L
, the Fuel Behavior Research Branch (NRC) to R. O. Meyer of the Reactor Fuels Section (NRC) title (RIA Energy Deposition',
dated Sect. "7,1979, It states, '"A reassessment of available data is needed to define a threshold in terns of ' fuel enthalcy'" because, "If a vendor calculated energy deoositions near the threshold using the ' fuel enthalp'y' definition the
' total energy' deposition could exceed 300 cal /cm." Mr.
Hostson further states, "It is our understanding CPB (Core Performance Branch) plans to ta'.:e the necessary actions to resolve the inconsistency between the Re ulatory Guide 1.77 position and the SPERT data on which it is based." Q: Were therc any tests between these two periods, and if n6t were any called for? A: The answer is no and yes. First, the various reviews of work on reactivity initiated accidents such as the Nuclear Safety Article, and NUREG/CR-0269 " Light Water Reactor Fuel Response During Raactivity Initiated Accident Erperiments", Aug. 1978, do not mention any tests between the SPERT tests and those in the Power Burst Facility with the exception of the work at the Nuclear Safety Research Reactor (NSRR) in Japan. That experimental work showed the significance of the pellet-clad gap variable in dternining the threshold energy deoosition for the onset of departure of nucleate boiling. (JAERI-M-8037 " Effects of Initial Gan Width on the Fuel Failure Behavior under RIA conditions", Saito, S. , et al. Jan.1979) But none of these l experiments attenpted to describe fuel rupture. By 1970, General Electric had plans to have reactors which produced sore than 1,000 MWe, but no- experi= ental basis for deter =inint the ph7sical correctness of its computer codes for calculating the effects of reactivity insertion in large l cores. These were called point kinetics programs, but in l WASE-1*46(" Water Reactor Safety Program Plan", prepared by the National Reactor Testing Station, Feb.1970) the Atonic Energy Concission ste.ted (p. III-93), "In general, a7 plication of tne point kinetics approach cannot be expected to be re-l - P00R ORIGINAL
12 - liable for large, loosely-coupled reactors." A proposed (Ibid. III-94) SPERT Large-Core Dynamics Program of research was never carried out- Today, with many large core reactors in use nationwide, the Power Burst Facility is doing small core ex,eriments. PTE-738(" Review of Generalized Reactivity Accident foi Water Cooled and Moderated Uranium Dioxide Reactors", G. O. Bright, 1965, p. 102) stated that a prob- - les area was excursion behavior as it related to core size, and spoke of the need for kinetics experiments, "..[a]dequate to determine the importance of the effects of s7 ace dependance under ootentially hazardous conditions." So, the need was expressed, but many reactors were built before the research work was begun at the Power Burst Facility. Q: 'clhat evidence indicates fuel will be extruded from the cladding in the event of a Reactivity Insertion Accident? A: The sequence ficare of a nuclear excursion accident (Attachment A) shows " Fuel Release" following " Thermal Stress Rupture" as an outcome when excess reactivity is available. (ORNL-NSIC-23
" Potential Metal-Water Reactions in Light Water Cooled Power Reactors, Aug. 1968, but ori-inally from Nuclear Safetv 8(2):
116-127, in an . article titled, "A Review of Generalize 4 Accidents for Water-Cooled and Moderated UO2 Fueled Power Reactors", by l G. O. Bright, Winter 1966-67.) Attachment C, (Nuclear Safety Article, p. 592) indicates rods receiving 250 cal /g UO2 and 260 cal /g UO2 1 st 10% and 15% respectively of their fuel. (At p. 598 of the Nuclear Safety Article, it states this loss was 13%) These were sin 518 rod tests. If 10% tc 15% of many fuel rods were to extrude - this percentage of fuel, then a considerable mass of fuel would be released. Q: What evidence indicates there will be pressure pulses from fuel contacting coolant under ' reactivity insertion conditions?
, P00R ORIGINM
_ 13 _ A: If molten fuel lea'res the cladding perimeter, it is thou-ht a steam enplosion will take place. M. L. Corradini, of the Li-ht t'ater Reactor Safety Studies Division of Sandia National Laboratories (SAND 80-1535c."A One Dimensional Transient Model for Analyzing Large-Scale Steam Explosion Experinents,"1980.) has divided this phenomena into three stages. "(a) mixins of the molten fuel and water; (b) triggering and spatial prop-oration of rapid fuel fragmentation through the fuel-coolant mixture; and (c) expansion of the stea= against the surround-inGs." Estimates of the quantity required to Produce the c:nsequences to the surroundings include a July 1971 evalu-etion that but 1% to 3% of core. mass melting would cause a s;eam explosion sufficient to destroy the (reactor) vessel. (B"I 1910 " Core Meltdown Evaluation", Morrison, Ap7endix C, 1971). > However, it has never been confirmed th'at a steam explosion has occured in a nuclear reactor. However, the explosion that accompanied the final Boiling Reactor Experiment (50RAZ) at the National Reactor Testing Station has never been explained as anything else tut a steam explosion. ( AHL 5323 "Excerimental Investigati*n of the Self-Limitation of Power During Reac-tivity Transients in a Subcooled, Water-Moderated Reactor," J. R. Dietrich, 1954, p. 39) A rough estimate of 6,000 p.s.i. to 10,000 p.s.i. was made (Ibid. p. 29) of the force of this ex71osion. Pressure pulses from the rupture of fuel rods have been shown to occur. In IN-1370 (' Annual Renort - SPERT Project , 1-3/33 - 9/69" R. W. Miller, p. 27, Attachment D) a 2,350 p.s.l.g. culse was generated on failure at 300 cal /gm of T' rece, tion following depostion of 3a0 cal /gm of UO2 to a rul of 3,000 mwd / ton burnup. The force of pressure pulses from groups of rods in one assembly will bend the channel box outward where it will block movement of the control rod. The BWR/6 design uses ff 2 P00R ORIGINM
. . - - - ,- ,.._,c _ . _ _ _ - . . _ , , . _ . , , _ __ _ , , - _ - . _ _ - . - - _ . _ ~ _ . _ _ - , . . _ _ _ _ _ , _ . . , .
a 0.260 inch thick control rod blade in a 0.482 inch water gap, leaving 0.111 inch tollerance between the control rod blade and any fuel rod channel, so only a small maount of deflection will interfere with blade progress. (Measurements from NEDO 10,566 "Testin6 of Cruciform Control Rods for BWR/6",1970, p.1) The ACHGS fuel channel wall thickness is 0.120 inches. ( ACNGS PSAR, Table 4.3-1, p. 4.3-17, Amendment 56, March 1981) Pressure pulses in addition to*dama5e caused in-directly by channel wall bulging, will propogate to other fuel assemblies through holes between the lower tie plate and channel, between the fuel support and lower tie plate (back cressure), and in th's lower tie plate through which it will act on the next adjacent fuel channel. (See Attach-ment E) Note: As shown on the attachment, the pressure nulse will travel downward from the fuel rods as indicated by the arrow drawn in the figure. Q: What evidence indicates fuel rod clad will be weakened by rupture, in particular that the affected cladding will not stand un to disturbances due to the reactivity insertion such as pressure pulses from the coolant contact as the contention describes? - A: The test RIA-ST-1, " Burst 2" at the Power Burst Facility (See Attachment C) produced 250 cal /g UO2 deposition into a fuel rod. In two publications, P. E. MacDonald, R. K. McCardell, Z. R. Martinson, and S. L. Seiffert of the NRC's contractor, EG&G Idaho, Inc. , showed that this deposition substantially weakened the fuel rods. I will cone to the weakening after giving their description of how the weakening occured. They suggested, "...($uring the oower burst the fuel enpands out against the claddin51
'ilm boilinc heat transfer is initiated and the cladding tenperatures reach values near the melting point; rapid heat transfer to the coolant results in vaporization and a j g i! & P00R ORlGlNAL
modest pressure pulse ( 6 2 MPa) which acts on the ductile cladding to defors it into thin and thick recions; the fuel rod claddinc remains in film boiling for about 20 seconds during which the cladling reacts with both fuel and coolant (steam) and becomes heavily oxidized; and, finally, rod fractureoccursuponcuench."!(CONF 790441-5, " LWR Fuel Response Durinc Reactivity Initiated Experiments, P. E. MacDonald,e_t,g.,EG&GIdaho)Inc.,1979,p.4) The weakening was extensive, such that 55% of the two unirradiated fuel rods, ahd 65% of the two irradiated fuel rods subjected to 285 cal /g UO 2 during experiment RIA 1-1, "... (@umbled into fine powder and larger chunks of fuel and cladding." (Nuclear Safety Article, p. 598) The irradiated rods were of 4',600 mwd /t burnup, well below
-f; projected 50,000 45,000! mwd /t for fuel at, ACNGS.
Cracking was observed in the single, unopened, irradiated to 5,000 mwd /t burnup, fuel rod in test RIA 1-2, which had absorbed but 185 cal /g UO 2 niexcursion testing. The same authors from EG&G Idaho, Inc. state in greater detail (Ibid.) the weakening is characterised by oxidation, a circaloy and water reaction found as zirconium oxide on the outside surface of the fu,el rod, and a uranium oxide-zirconium reaction whose result was found on the inside sur-face of the cladding. In thin re5 ions of the cladding the raaction fronts met, meaning the cladding was oxidized through. They state further that these oxide compounds are brittle com-l pc.ad to normal zirealoy tubing. In line with this discussion of departure from nucleate boiling and energy deposition, the USRR studies in Japan l concluded (JAERI-M-8087)that the gap width had significant influeace on the onset of departure from nucleate boiling as a result of reactivity insertion. The threshold energy depositions for the onset of deoarture from nucleate boiling f r a fuel rod with initial pellet-are about 180 cal /cm UO2 clad Eap width of 0.195 mm, about 140 cal /g U02 f r a fuel. rod with initial pellet-clad Sap width of 0.095 mm, and P00R ORIGINAL
, _ -..._._,..-.,_,_-_,-.-_y_ - _ , , _,
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about 110 cal /g UO 2 f r a fuel r d with initial pellet-clad gao width of 0.050 mm. (JAERI-M-8087 " Effects of Initial Gao Width on the Fuel Failure Behavior under RIA Conditions", Saito , S. , et al. , Jan. 1979, p.1) Q: What evidence indicatestbere would be an increase in fissioning if fuel rods were bent out of line and that this would result in more ructured cladding, fuel escaping into the coolant
'and pressure pulses from fuel-coolant contact?
A: Applicant's fuel rods will be 3 7 mm apart, according to NU2EG-3152, Table 4-1, p. 4-3 (Safety Evaluation Report to GESSAR-238", Nuclear Steam Supply Standard DesiEn, 1977). According to R. E. Webb, i m ard fuel rod bowing increases reactivity. ( Accident Ea::ards g Nuclear Pouer Plants,1976,
- p. 18) In the conte'ntion, the increased fissioning would lead to increased reactivity. The increased fissioning -
would be due to decreased neutron leakaSe, because this same author (Ibid. p. 17) states, " Moving fuel apart in a disassembly generally decreases the reactivity by enhancing l neutron leakage." In NEDO 10,173, " Current State of Know-l ledge of High Performance BWR Zircaloy-Clad UO p Puel", May, 1970, bowed fuel rods in the early overation of the Dresden-I olant had five rod failures from fuel rod-channel box contact. Where cressure culses distort the channel box, and force fuel rods inward (not outward as at Dresden-I) increased fissioning can be exoected between the rods which have been oushed more closely together. l When fissioninc fuel rods are closer to5 ether, the depar-ture from nucleate boiling ratio' decreases. Alt' lough recent research shows clearance can be reiuced to 0.76 mm (" Reduced Clearance Effect on Critical Power in BUR Fuel Bundles", Ikeda et al. Transactions of the American Nuclear Society, 33 (S98-9), 1980)ir rods are closer than this fission will increase as will denarture from nucleate boiling. Rods which hai undergone the transient would already be weakened as described in the results of test RIA-1, and the effect of departure from nucleateboilin3 on them would be more serious. This scenario is part of the "further core danage"
^- P00R ORIGINAL
~
_17_ which results fron " conversion to mechanical ener67" and lea $s to " Geometry hk" and " power rise" as shown in Attachment A. Q: Does that conclude your testimony on Doherty Coctention No. 37 A: I would point out that the hazard of fuel material clogging or obstructing control rod operation exists for many of the events where reactivity is inserted and the control rods do not insert, or are delayed in insertion, as in the
~
Browns Ferrv-III event of June 28, 1930, whero rods were in-serted after a fifteen minute delay. Reactivity insertion following an event is'almost instantaneous, but the effects as described in the contention, such as further collapse of fuel rods, and pressure pulses conceivably would not be instantaneous. I have mentioned bulging ef the fuel channels as oae mechanism of control rod movement inhibition here, but the second mechaniam, clogging with fuel bits might be important in an event where the consecuences of the reac-tivity insertion are not all instantaneous. Q: Thank you for your testimony, Mr. Scott. T P00R ORIGINAL
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=n ATTACHIE"2 3
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=2 ' s== Issued: December 1970 3
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# CIfice of the Secretary S
LService O' 1 THE EFFECTS OF BURNUP CN FUEL FAILURE II. Power Burst Tests on Fuel Rods with 13,000 and 32,000 mwd /MTU Burnup R. W. Miller - IDAHO NUCLEAR CORPORATION A Jointly Onned Subsidiary of AERGIET GENEPAL CORPORAfl0N , ALLIED CHEMICAL CORPORATION
, ~PMILLIPS PETROLEUM COMPANY I
l ....
\* . i N ./
'U. S. Atomic Energy Commission Scientific and Technical Report Issued Under Contract AT(10-1)-1230
./
k Idaho Operations Office
. N... M,._ P00R OR,GINAL
CONTENTS ABSTRACI . .
.........................11
- 1. INTRODUCTION . . . . . .. ..... ......... 1
- 2. EXPERIMENTAL RESULTS . ... .... ......... 4 2.1 Test 685; CEX Rod; 13,100 mwd /MTU; 186 cal /g-UO2**'
2.11 The Fuel Rod . . . .... ......... 4 2.12 The CDC Test ... .. ........... 5 2.13 Test Results . . ..... ......... 3 2.2 Test 684; GEX Rod; 12,900 mwd /MTU; 200 cal /g-UO2**5 2.21 The Fuel Rod . . . . . . . ......... 5 2.22 The CDC Test ... ...... ... .... 8 2.23 Test Results . . ....... ... .... 8 2.3 Test 756; GEX Rod; 32,700 mwd /MTU; 176 cal /g-UO28 2.31 The Fuel Rod . . . .... ..... .... 8 2.32 The CDC Test .. ... . . . . . . . . . . . 11 2.33 Test Results . . ... .. . . . . . . . . . 11 2.4 Test 859; GEX Rod; 31,800 mwd /MTU; 190 cal /g-UO2
- 16 2.41 The Fuel Rod . . . . . ....... . . . . 16 2.42 The CDC Test . .. .. .. .. . . . . . . . 16 2.43 Test Results . .. .... ..... . . . . 16
- 3. CONCLUSIONS . . . . . . ... .. ....... . . . . 24
- 4. REFERENCES . . . . . . .. . .... ..... . . . . 29 APPENDIX
- 1. The Burnup Process . . ... .. .... ... . . . . 31 1.1 -Method . . . . . . .. . ...... ... . . . . 31 1.2 The ETR . . . . . . . ... .... ... . . . . 31 1.3 The P-7 Loop . . . .. ... .... ... .... 31
, 1.4 Fuel Configuration and Hardware ..... . . . . 31 l 1.5 Operating Conditions . . . . . . . . . . . . . . . 32 l 1.6 -Operating History .. ... ....... .... 32 1 FIGURES
- 1. Schematic of GEX fuel rod showing construction details 3
- 2. Test 685 data on 10 maec/ division scale . ....... 6
, c m .
. 111 . .
P00R ORIGINAL.
i
,- 1
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l In summary, the rod initiated a fracture as a result of this test, but the failure was minor and fission products did not escape.
. _ ~
2.4 Test 859; GEX Rod; 31,800 mwd /MTU; 190 cal /g'
~ .-
2.41 The Fuel Rod. Fuel rod SE-33 received an average burnup of 31,800 mwd /MTU. The rod was under irradiation for 216 days during which its average linear power was 13.9 kW/ft and its maximum linear power was 18.1 kW/ft. A hot-spot factor of 1.04 raised the linear power to 18.8 kW/fe over a short section of the roa. All ' power and burnup figures are + 10%. Fast neutron (> 1 MeV) exposure to the cladding was 0 about 2.7 x 10 nyt. A neutrograph of the rod taken after the prairradiation indicated slight centerholing over the entire active length and several visible fractures (typically 3 to 4) per pellet. Appearance of the rod was identical to the preceding rod (SE-30) and indicated corrosion buildup over the active length. , ( As in all cases, this rod was carefully inspected before pre-irradiation and again after p reirradiation by magnifying periscope in i a hot cell prior to testing. No evidence in either inspection was I found to indicate a sub-standard condition of the rod, l 2.42 The CDC Test. The fuel rod was placed'in a. capsule con-taining room-temperature water under stagnant, atmospheric pressure conditions and subjected to a 3.94 millisecond period transient in the CDC. The maximum energy deposition during the test was 190 cal /g-UO 2
. Figures 13, 14, and 15, show the power burst and other recorded data taken during the test.
~
2.43 Test Results. The rod failed early in the transient _w .
\,
only about 85 cal /g-UO 2 had'been deposited.- The time of failure is indicated clearly in Figure 13 by the sudden rise of capsule pressure as well as by disturbane the~ i,ee 1 um che71 adding and fuel e, h transducers The rod is shown in Figures 16, 17, 18 and 19. Four independent failures were found: ' three fractures, which together extend over { uost of the active length of the rod, and one ruptured blister.
! r, t 2
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6".9 1 N [ Failure was accompanied by a multi-peak pressure pulse which reached about 70 psig me,ximum. Sufficient energy was released to cause the water column above the fuel rod to reach a maximum velocity of 5 ft/sec. It was determined that a nucleark o-mechanical t energy conversion of about 0.014% took place. Less than 1% of the active cladding was consumed in a metal-water reaction. Very little fuel was lost through the cracks into the capsula, l Metallurgical examination of the rod was not performed due to_. termination of the program. - I o e e 1 e e nr .
>, 23 P00R DRIGINAL 3m{}, t .a..
. m-S h*6 %
; =
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=
i n j L w , 4*, _m 3 s% 5) j i.7.w % i l 7.1N Table 2 RfA Test Results at the PBF t M< ' } *"" ?- Radial Radial average Radial average i DQ l'uel power antal energy peak fuel Peak local _ j e-4 Fuel-rod enrichment, Burnup, peaking deposition,* enthalpy,6 g,,g ,,gs,,ipy,c l ~ ~f= RIA test type %***U & lwd /t factor cal /g 00, cal /g U0, cal /g 00, Conurents i
, D 'n
- RI A ST-l 17 m 17 PWR 5.8 0 1.076 250 185 205 Did not fail;first test on ST l li
~~'I s.* Burst I rod I RI A-ST-1 17 m 17 PWR 5.8 0 1.076 330 250 275 10% of fuel washed out; second test t' e< f Burst 2 on ST-I rod
'[ i
! ",y RIA-ST 2 17 m 17 PWR 5.8 0 1.076 345 260 290 15% of fuel washed out g
. k RI A-ST-3 17 m 17 PWR 5.3 0 1.076 300 225 250 Did not fail O j; $ RIA ST-4 15 m 15 PWR 20 0 1.48 695 350'8 5308 Completely destroyed; pressure E O d i- pulse of 35 MPa measured 1! >
- , RIA I I Two Saston 5.7 4600 1.13 365 285 330 Complete shroud flow blockase j
- I Two Santon 5.8 0 1.077 365 285 315 Severe failure; partial now e-l blockage $
es .l . RIA l2 Four Saston 5.7 5000 1.13 240 185 285 One rod failed; three rods
, did not fait j! I
*l'ive methods were used to measure the test-rod radial avnage fission energy deposited du ing each transient.' A detailed independent review of the five
{l measurement methods confirmed that none were unreliable. The five measurement methods ha l estimated uncestainties ranging from all to sl45 These are j conservative estimates of the uncertainties and, based on previous PBI' results (where th average burnup measurement is within 3% of the average t l ,l T thermal-hydraulic power measurements), these results are considered to be accurate to within abo t s6%. 6 The FRAP-T5 computer code' was used to de?crmine the axial peak radial average fuel ent alpy from the measured totalenergy depositions.The Tsaction of energy generated by delayed neutrons after control-rod scram was calculated using she TWIGL c mputer code (Connguration Control Number 110099788).TWIGL lr. ' J-O solves the coupled time and spacedependent neutron diffusion and thermal-hydraulic equations or a reactor in two dimensions.
'This value will vary somewhat depending on the node stres in the analytical models used to convert total energy deposition to peak local fuel enthalpy.
' d The fuel enthalpy at the time of fa. lure was ~3 ans after the time of peak power. p O
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'A r 7 i TABLE VII !;
c) g l t' r M~\ SQ04ARY OF TESTS ON oE1 FUEL rod 5 WI"3 AND VTTHoUr BURNUP .p 'l' Ruclear-to-l
- I4 cE Total Test Marina = - Mechanical M Rod Energy Energy at Capsule Energy Metal-Water Deposition Burnup Test Failure Pressure Conversion Reaction h.a.
fjf ' (cal /g UO21 (NWd/MN) No. (cal /R UOo) (esfr) (T) ($) U
,4 o (No test - all quantities expected near zero -
no failure expected) 150 rs - 0.06 M i 571 (no failure) o o 3300 5 471 (no failure) o o 1.3 g - o
} @Tda!
M 200 FC gj :. 1.3 Si$ 3000 568 147 165 0.03 W . td 260 o o 8.8 I 260 O hT6 h@ 5 Y'h 0.05 4.9 , 567 225 350 3300 !? 0.36 19 0 o kT8 300 375
.i
_ . 3 3000 569 300 2350 0.21 -85 ; f] De relatively low burnup of about 3000 mwd /3dTM reauoed (1) the failure threshold from greater than 225 cal /g of UO2 j ya ' for unirradiated kol to less than 200 cal /g of UO2 for the i j ET I c3 II.i 1rradiated fuel. (A (2) Measurable capsula pressure and generation of mechaalcal i@%;i 83' . energy occurred at significantly lower energy depositions W for the irradiated fuel rods than for comparable unirradistel g fuel rods (200 cal /g compared with over 300 cal /g). w 1 l g ,h,h f (3) ne failure mode was apparwtly affected by irradiation. Little or no cladding melting was observed for any of the 1 b-19 L
.I Irradiated rods, whereas all unirradiated fuel rods showed '
cladding melting for energy depositions above about 250 cal /g ,s of UO2 . Near the failure threshold. unirradiated rods failed by ) cladding molting, whereas irradf ated rods failed by sudden J# N cladding rupture. '
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ATTACIEZiT E
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ACNGS-PSAR NOTE: PERLPHERAL FUEL SUPPORTE ARE WELDED INTO THE CORE
/ SUPPORT PLATE. FOR THESE SUNDLES. PATH R41 r qTg LOWER TIE PLATE 2.J AND 7 y %
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8- : s___ __ _ _ s ; APR S gB\> -r rgessurE PULSE DIAECTI 1, M & C' > I N, CORE SUPPORT PLATE FUEL SUPPOR7 l
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I CONTROL AOD Ib -t u - l P00R ORIGINAL
- 1. CONTROL RCD GUIDE TU8E FUEL SUPPORT f
lh 2. CONTROL ROD GUIDE TUBE CORE SUPPOhT PLATE
- 3. CORE SUPPORT PLATE INGRE GulDE TU8E
- 4. CORE SUPPORT PLATE SHROUD 5 j r CONTROL ROD 5. CONTROL ROD GUIDE TU8E DRIVE HOUSING
% DRIVE HOU3tNG 6. FUEL SUPPORT LOWER TIE PLATE 3'
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9. CHANNEL LOWER TIE PLATE ALTERNATE FLOW PATH HOLES HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Huclear Generating Station Unit 1 Schema?.ic of Reactor Assembly Showing the Bypass Flow Paths A ct. No. 56,' (3/81) Figure 4.4-1.
- _ _ . - . . - . . . . . . - . _ ~ - - -~ - . - . , _ . .? _? . __ 2 - - - . - . - . _ . - . . . . . . . . - . .
/ ; @ coe::,"_3 g. ATTAcEMER F
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h ME$fjtLMPI :nL UALIFICATIc"3 oF JA::23 ::c?GAn scoer. eso.
% v/ $,/ ,j I received a Bachelor of Science Degree fro: Arkansas 9
b Polytechnical University, Russelville, Ark., in 1963 My
=sjor subjects were Physics, Chemistry, arid IIathe=atics, meeting =ajor requirements in the three areas. The school was 13 milds from my home.
In the su=ser of 1963, I was a full time laboratory assistant in the hic-h energy physics laboratory at Iowa State University, Ames, Iowa, prior to a year as a teachin:; assistant in Physics. I chose,to enter the progra: at the University of Texas, in the following fall,.and worked full-ti=e the prior sus =er in the nuclear ohysics accelerator laboratory. I was an Undergraduate Laboratory Teaching Assistant, cart ti=e and at ended classes part time. In my second and third succers at the University of Texas, I did research in plasma physics, which was the subject of my caster's thesis, which was prepared instead of the Ph D. I took the M. S. , and a position with Ilestinghouse, because I was informed I no longer had a student deferrment from military service. Relevant to the Contention are the following I.'3sr ; , 3 undersraduate cou'rs,e: nuclear physics laboratory; and se c... ; . e , t graduate courses: nuclear physics, plasma physics, quantum mechanics, electro-magnetic theory, classical mechanics, statistical mechanics, heat and thermodynaaics, and solid state physics. My post-master's degree work experience includes two and a half years voriing as an assistant and later associate endneer with '.lestinghouse doing design, =anu- , facture,and testing of integrated circuits for various
solid state electronic comoonents for several government sponsored projects. This was in Baltimore, Maryland. From 1959 through 1971, I was employed by General Electic Company in Syracuse, H. Y. designing integrated circuits for G. E. products, such as computer components and radios. From 1971 to 1974, I was emplyed by Texas Instru-ments,in Stafford, Texas at their Semi-Conductor Division. There I designed integrated circuits for military commun-ications ecuiptnent. I was later a Product Engineer doing failure analysis of integrated circuits. From 1975.to 4977, I was a full time student at the University of Houston, Bates Colle6e of Law. The majority of my time in law practice has been on hydrology and air emission suits, in addition to the. Allens Creek proceedings. l ee +' t g* c. P00R BRIGINM
l
. ,: . .w, t
~= C9RTIFICATE OF SERVICE l
l Cooles of "INTERVENOR DOHERTY TESTIMONY TF JAMES M. SCOTT ON j )
- DOHERTY CONTENTION #3
- INADEQUACY OF THE DESIGN SAFETY LIMIT ;
! 0F APPLICANT'S FUEL RODS" were submitted to the Parties below via First Class U.S. Postal Service this 302% day of April, 1981 from Houston, Texas. sa e s i e Sheldon Wolfe, Administrative Judge /h 000lG7c3 ; iL Gustave A. Linenberger, Administrative Judge [., utgag : , Dr. E. Leonard Cheatum, Administrative Judge 7-' APg g 3 Richard A. 31ack, Esq., Staff t: o Atomic Safety Licensing & Appeal Board - 19873 O} J. Gregory Copeland, Esq. g p/
. Jack R. Newman, Esq.
; The Several Intervening Parties e, es Docketing and Service A ,
Res ectfully, 4
/
(34hnDoherty i l l l 1 l d e
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