ML19345E444
| ML19345E444 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/30/1968 |
| From: | Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19345E439 | List: |
| References | |
| NUDOCS 8101190310 | |
| Download: ML19345E444 (14) | |
Text
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SAFLTI EVALUATION BY THE DIVISION OF REACIOR LICENSING DOCKET NO. 50-155 CONSUtEPS POWER COMPA?N PROFOSED AMENDMENT NO. 1 INTRODUCTION By letter dated May 26, 1967, the Consumers Pbwer Company of Michigan has proposed Change No.13 to the Technical Specifications which we have redesig-nated Amendment No. 1 of License DPR-6 for the Big Rock Point Power Plant.
Supplemental information was submitted on August 15, 1967, Nove rber 10, 1967, and December 14, 1967.
1 Anendment No.1 would permit insertion of six high performance developmental fuel bundles into the Big Rock Point core as part of the normal core complement of 84 fuel bundles. The developmental fuel will be irradiated until the rest depleted fuel rods acquire 21,000 MdD/T U average exposure. This irradiation j
program, designed to investigate the performance characteristics of fuel rods with center melting, is an extension of the high performance UO pIngram sponsored jointly by the U. S. Atomic Energy Comission and E tom. The pre-posed change was reviewed by the Advisory Committee on Reactor Safeguards (ACPS) which concluded in its report dated December 20, 1967, that "the reactor can be operated with the high performance test assemblies without undue risk to the health and safety of the public". A copy of the ACRS report is attached.
4 DESCRIPTION The Table, which follows on pages 3 and 4 sets forth the important characteristics of the high perfermance fuel bundles and presents a comparison thereof with the Type "C" fuel bundles nest recently used. to refuel the Big Rock Point reactor.
The salient aspects of that table are further discussed below.
The eix high perforrance developr. ental fuel bundles, containing 324 fuel rods, 2
will oporate with peak heat fluxes of 450,000 +50,000 B7U/hr ft. 7b achicve the design objectives of perfomance and power output, it will be necessary that the high performance fuel bundles be repositioned into higher neutron fluxes during their appmximately 2 - 21/2 year in-core irradiation. 7he use of three i
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, different U enrichments, i.e., 4.3, 5.0, and 5.6%, to cont 1 individual fuel rod poh generation within a fuel bundle pemits nest of the fuel bundle power to be generated in about 60% of the fuel rods. 'Ihe remaining fuel rods contain depleted fuel, i.e., only 0.22% wt percent U and therefore generate only a smll fraction, about 3%, of the fue{3bu,ndle power. Two of the fuel bundles, designated as intermediate perforTrance fuel, contain 0.570 inch diameter fuel rods and are designed to produce incipient center melting. One of the intemediate perfornance bundles contains pellet type fuel and the other vibratory compacted powder fuel. The other four bundles, designated as advanced perfomance fuel, contain 0.70 inch diareter rods and are designed to produce substantial center melting. 'IWo of the advanced perfomance asse:rblies will contain UO, in the form of sintered cored pellets and the other two will contain UO in the fom of powder. Selected 2
rods will contain tungsten wafers to minimize axial novement of the nelten U0 2
during operation. All six high perfornance fuel bundles will be distributed within the core so that the minimum center-to-center distance of these fuel bundles is 42 cm (16.5 in.).
None of these bundles will be positioned in the outer row of the core. 'Ihe handles of the high perfomance fuel bundles will be notched to pemit visual verification of their core positions. Correct placement of the Inds within the fuel bundle geometry is assured by 3/8 inch identification letters sta.~ ped on the fue' rode.
'Ihe developrental fuel bundles are designed to pertrit disassembly when renoved from the core during refueling. This feature will facilitate exandnation and replacenent of all individual rods. Selected irradiated rods will be shipped to the Vallecitos Nuclear Center for destructive examination after a suitable period of radioactive decay in the spent fuel storage pit at Big Pock Point.
Pods selected for destructive examination will be replaced with new rods.
Lines 21 and 22 of the table show that the average power generation in the " hot" rods (high power producing rods) of both intemediate and advanced perfomance fuel is significantly larger than the peak power density in the Type "C" Big Pock Point fuel. Lines 24, 25, and 26 show that the total UO in each advanced performance fuel bundle is significantly greater than in eithdr Type C or the internediate perfomance fuel, but the total ancunt of U in either an advanced 25 or intermediate perfornance fuel bundle is less than for a Type C fuel bundle.
Further, the a-cunt of UO fuel in the power generating rods of each proposed 2
fuel bundle is noticeably less than the arount in the power producing rods of a Type C fuel bundle. 'Ihe cross-sectional area of the developmental fuel rods, howevar, is increased in accordance with lines 2 and 6 of Table I by a factor of 1.7 for the 0.570 inch dianeter rods and 2.65 for the 0.700 inch diareter rods.
i TABLE 1 COMPARISON OF PROPOSED FUEL TO BIG ROCK " TYPE C" FUEL' Proposed Fuel Big Rock Intermediate Advanced Type "C" Performance Performance Fuel 1
Geometry, Fuel Rod Array 11 x 11(1) 8x8 7x7 2
Standard Rod Diameter inches 0.449 0.570 0.700
[3 Number Standard Rods per bundle 121 36 29 4
Number Special Rods with depleted uranium 0
28 20
_5 Special Rod Diameter inches 0.344(4) 0.570 0.700 6
Standard Rod Tube Wall inches 0.034 i
0.035 0.040
[7 Special Rod Tube Wall inches
' O.031 0.035 0.040 8
Rod Pitch inches 0.577 0.807 0.921
,jl___jActive_ Fuel Length inches 70 66 - 67.3 65 - 66.3 10 UO2 Density, Percent Pellet Powder Pellet Powder Theoretical Pellet 85 powder 94 85 94 85 11 Fill gas He He He l_2 Spacers per bundle 5
5 5
13 i Clad material Zr-2 Zr-2 Zr-2 14 Wt Zircaloy clad per bundle Pounds 90 62.7 67.6 l5 DHcgagggfsinsidefuel inches 0.497 0.88 0.85
_16 Rod to Rod clearance inches 0.160 0.237 0.221
_17 Rod to channel clearance inches 0.128 0.160 0.155 18 Number of bundles UO2 Pellet N.A.
1 2
UO2 Powder 40 1
2 19 Steady state heat Tech Specs 500,000 500,000 500,000 flux limit BTU /hr ft2 550.000(3) 20 Power generation Pellet Powder Pellet Powder at limit KW/ft(6) ig(3) 21.8 21.8 26.9 26.9 21 Peak power at core rated conditions KW/ft 12.3 21.8 21.8 25.4 25.4 22 Average KW/ft core Hot Rods Hot Rods Hot Rods Hot Rods pcwer 4.1 16.4 16.4 19.4 19.4 23 Average Heat Flux BTU /hr f t' core Hot Rods Hot Rods Hot Rods Hot Rods 124,000 375,000
__375,000 360,000 360,000 00 / bundle kg.
133 136 129 159 152 24 W:
2 UO / bundle (U235) kg 4.83 Hot Rods Hot Rods Hot Rods Hot Rods 25 2
3.76 3.7 4.6 4.35 26 U02 in power producing Hot Rods Hot Rods Hot Rods Hot Rods rods / bundle kg 133 76.5 72.5 95 89.5 l
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- Table 1' Cont'd COMPARISON OF PROPOSED FUEL 19 BIG ROCK " TYPE C" FUEL
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Proposed Fuel 1
Big Rock Intermediate Advanced Type "C" Performance Performance Fuel 27 Heat transfer surface ft2 79 Hot Rods Hot Rods Hot Rods Hot Rads bundle 30.2 30.2 29.4 29.4 28'l Average bundle power MWt 2.86 3.25 3.25 3.25 3.25 29 Radial power factor in core 1.30 1.21 1.21 1.14 1.14 30 MCHFR at 122%_ Power
>1.5 1.53 1.53 1.54 1.54 31 Doppler coef. for (4) bundle at 10000K delta k/kOF
-1 x 10-5
-0.8 6 x 10-5
-0.9 5 x 10-5 32 Temp coef. 250C at (4)
BOL delta k/kOC v3.2 x 10-6
+1.04 x 10-4
+5.6 x 10-5 33 Void coef. 200C at BOL delta k/k0 unit void
-0.28
.24
.27 34 Cold pellet fuel to clad gap inches Powder 0.012 for pellets 0.013 for pellets 35 Central hole, pellet (5)
Pellet fuel inches N.A.
0 0.100 36 28000C Pellet fuel W/cm N.A.(5) 59 - 62 59 - 62 JIK de 5000C Powder fuel W/cm 49 49 49 37 Fuel enrichment wt*/. U-235 2.965.2 4.3;5.0; 4.3;5.0;
_ 5.6:0.22 5.6;0.22; 38 Fuel Rod lifetime MWD /T U 15,000 Ave, hot rods Ave. hot rod ave. rod 21,000 21,000 Symbols Minimum critical heat flux ratio.
MCHFR a
Hydraulic diameter of coolant channels.
D (1) 4 corner rods may be cobalt targets (2) 8 special corner rods and 4 Cobalt target rods for radial supports (3) Incipient melting. Tech spec limit is 500,000 (4) Proposed Change No. 13 dated May 16, 1967 (5) NA - not applicable (6) Melting starts in powder fuel at 19.4 KW/f t i
Melting starts in pellet fuel at 24.2 KW/ft i
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- Le applicant has indicated that pwer in each 7 x 7 fuel bundle is genemted in essentially 29 rods, because the remaining 20 fuel rods contain depleted fuel with little U,,b.
he arrangement of the hot (high power producing) rods and
" cold" (deplet fuel) rods in an annular geometry provides the maximum cold surfaces adjacent to the power producing rods and optimizes pwer distribution and coolant utilization within the four 7 x 7 advanced perfomance fuel bundles.
Le enrichmnts and placement of the 36 high power rods and 28 depleted rods with-in the two 8 x 8 array intemediate perfomance fuel bundles as with the advanced perfomance bundles were specified to provide the mximum number of fuel rodg operating at or near the desired heat flux, i.e., 450,000 + 50,000 B'IU/hr ft, at the axial peak.
It should be noted that there are only 29 high power rods in the 7 x 7 matrix in contrast to 36 in the 8 x 8 matrix; therefore, the average power generation in each of the 0.700 inch diameter rods in the 7 x 7 array is greater by the ratio of 36/29 than the power generated in the 0.570 inch diamter rods of the 8 x 8 fuel bundles.
Le power of the three types of enriched U rods (high power rods) decreases with inventory decreases. The power in the depleted rods (cold.
exposure as the 0,3b exposure because of plutonium production.
rods) increases wl To compensate for this reduction of pwer in the high power rods, the fuel bundles will be noved to positions of higher radial power factors as irradiation proceeds so that the design heat flux of the high power rods can be maintained throughout exposure life-tine. It has been estimated that the relative power generated by the depleted fuel rods increases over the fuel bundle lifetime by a factor of less than 3, and there-fore, would rot significantly affect the lifetine themal perfomance of the high pwer rods.
Since a linear power density of 19.4 W/ft results in incipient center melting in the vibratory compacted powder fuel rods and 24.2 W/ft correspondingly causes conterline melting in pellet fuel rods, it can be seen from lines 21 and 22 that nest of the rods will be near centerline melting at the hottest locations when the rcactor is at rated power.
In the case of the hottest, large dianeter, pwder fuel rods, there will be nelten fuel over a significant rod length, i.e., from approximately 1 to 1.5 feet on each side of the axial pwer peak, with the maximum nelten fuel cross-sectional area approximately 36% of the cross-sectional area of the fuel rod.
EVALUATION In consideration of the fuel characteristics identified above, our safety evaluation is concerned with:
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Normal plant operations, including consideration of the significantly higher average heat fluxes from the hot centermelt fuel rods than the peak heat flux for the normal Type "C" Big Rock Point Core, solid to liquid fuel phase changes, and-ave mge hot rod fuel exposure of 21,000 WD/T U, which is significantly greater than present fuel life-time exposures; 2.
Ioss of coolant conditions, including consideration of the greater fuel mass higher temperatures (fuel enthalpy or stored energy), and higher rod power levels in the proposed fuel rods in relation to the ability to dissipate the residual and decay heat; 3.
Reactivity excursions associated with the hypothetical control rod drop accident.
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- NOPMAL PLANT OPERATION I
Ebel Rod Pwer: The fuel rods have been sized and arranged within the bundle so Eto achieve the maximum cooling. Test results With hot rods adjacent to cold rods similar to the errangement in the six proposed developmental fuel bund'ee confirm that the multi-channel model for calculating the Minimum Critical.
Heat Flux Ratio (MWFR), currently used by General Electric to determine fuel rod heat-trensfer limits in boiling water reactors, is conservative for the proposed rod arrangement. Using this calculational model, it was determined that the MCHFR at 122% of rated power is greater than the minimum value of 1.5 permitted by the existing technical specifications. It has also been determined that the peak clad 0
temperature at rated power will be approxirately 900 F for those hot fuel rods wtiich remain in the core for the full irradiation lifetime of 21,000 ND/T.U.
'Ihe capability to operate at 122% of rated pwer without exceeding heat transfer limits provides a sufficient margin, in our opinion, to ensure fuel cladding integrity during normal operation.
'Ihe licensee has reported that failure of the main coolant circulation pumps or the main power generator with the high power density fuel in the core will not result in violation of the minimum critical heat flux limits. Although a nuclear power surge would occur due to pressurization and void collapse, the heat transfer limits would not be exceeded and fuel temperatures would not be excessive. The MWFR during the power and flow transient would be greater than the minimum value
- of 1.5 permitted by the technical specifications. We are in agreement with the calculational methods used in developing their conclusion. 'Iherefom, it is our opinion that such unanticipated transients would not cause a significant release of radioactivity from the fuel rods into the primary coolant system or diminish the integrity of the primary system.
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' he high lil. ear pwer density of the developnental fuel reds ses significant increases in the fuel temperature co p d to Type "C" Big Rock fuel, and results in possible nelting of the UO at the center of the hot fuel rods whenever the 2
p uer density is above 19.4 m/ft for powder fuel or 24.2 Kd/ft for pellet fuel.
Rese values correspond to Yd9 values of 50.5 W/cm and 63 W/cm. Only the 58 large diameter, vibratory mpacted, hot fuel rods will have significant molten fuel. The remaining 130 high power density fuel rods will have incipient fuel nelting conditions at the peak axial power locations. The effects of increased fuel temperature, as well as center melting, necessitated new design pmvisions which have been developed and tested during the AEC-EURA'IDM sponsored "UO High 2
Perfonrance Pmgrem".
Clad-Fuel Interaction: Mechanical interaction between the fuel and the clad re-sulting from differential expansion was considered as one of the largest potential contributors to high cladding strain. To minimize this potential cause of cladding failure, a large fuel-to-clad diametral gap, i.e., 0.012 inches for 0.570 inch diameter fuel rods and 0.013 inches for the larger rods, has been provided. At rated pcwer, the diametral clearance will not be less than 1 mil over the lifetime of the fuel.
Pod Fabrication: A depleted UO7 pellet is placed at the end of each active fuel column to minimize the effects of hot fuel at the end plug and plenum spring.
Dished and cored fuel pellets have been provided to allow sufficient volume to acco:modate the phase change volume expansion of UO on melting, based on 9.6%
2 neasured U3 volume changes during destructive examination of the irradiated fuel.
7 A 15% void space has been provided in each of the 94 hot fuel rods using vibratory compacted pwder by compacting to only 85% of the theoretical UO density. In 2
consideration of the test irradiation experience and the observed resultant sintering densification and accompanying center voids, we believe this void is sufficient to acconnodate the volume expansion caused by melting of pcwder fuel.
h e applicant has calculated that p wer levels of 136% and 200% of rated p wer would be required to cause cladding failure of the srcall and large diameter pellet fuel -ods, respectively. Le maximum nelt fraction for pellet fuel at 122%
of rated power is only 22% compared to the design capability of 71%. The maxirram nelt fraction of powder fuel at 122% of mted power is 45% compared with a theoretical limit of 100% nelting which represents a substantial margin above the MWFR power limit.
Fuel Micration: While operating with nolten fuel conditions, the possibility of fuel migration within the fuel rod exists. Studies performed by General Electric indicate that increased or decreased fuel concentration along the axial direction of the hot rods causes insignificant reactivity effects. Momover, as mentioned earlier, only the 58 hot, large diameter pwder fuel rods have rolten center fuel over a sufficient length to permit fuel novement by this neans. A proposed rodifi-cation to the technical specifications, which restricts the rate of initial power
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B increase between '.70 W t and 240 E t to 1/2 Mdt per minute, allow's time for fuel migration. This _s particularly significant for the powder fuel rods, when the fuel is sintered and densifies the first tine nelting occurc. This restriction also permits gradual axial expansion of the nolten fuel when the peak power has noved downward. We believe that this is a prudent restriction to avoid the renote possibility of fuel rod damage from these factors. Also, tungsten wafers are placed at 18 inch intervals in 4 hot, large diameter, pellet rods and 4 hot, large diameter, powder rods to minimize the axial novenent of nelten fuel for comparative perforrance evaluation.
Fuel Pod Lifetine: The capability of the hot fuel rods to achieve 21,000 W D/T U average accumulated exposure without failure will be confinred by renoving 4 representative fuel rods during each core refueling for destructive examination at the Vallecitos Nuclear Center. These examinations will reveal the a:rount of fuel burnup, extent of melting, fission product distribution, gas release, fuel neverent and dimensional changes. Deviations firm predicted conditions will be carefully evaluated to determine the effect on predicted fuel lifetine. The times required for fuel cooling, shipment, and examiration result in about a five nonth delay after rod renoval before results of the examinations are obtained. We believe that the confirmation of design performance by the destructive examinations is valuable in ensuring safety of continued irradiation for these bundles. There-fore, the nodified technical specifications require that after the first rod renovals, the four advanced perforrance bundles should not be reinserted until the initial destructive examination results are obtained and evaluated.
Fuel Rod Destructive and Non-Destructive Tests: A number of ron-destructive tests will be perfonrad on each of the developmental fuel bundles during each core re-fueling period. Each bundle will be leak tested by the " dry sipping nethod". The bundles will be examined visually using television and an underwater periscope.
The diameter of each of the 188 hot rods will be reasured with a go-no-go gauge to detect dimensional changes resulting from irradiation. In addition, several rods will be neasured with the profilometer for comparison with the pre-irradiation traces.
An unexpected diareter increase will be cause to discontinue irradiation of the affected rod. These examination procedures, first performed when the hot rods reach less than 15% of the expected lifetire, will provide added assurance that the fuel is performing in accordance with predictions.
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Continuous nordroring of the gases from the air ejector nonitor hill detect fuel element rupture. To provide early warning of a fuel failure, the nonitor alam s'et points' will be lowered to about twice the background level for the duration of these developnental fuel irradiations. 'Ihe few fuel rod failures experienced during the Euratom high power density fuel development program caused the release of only a small fraction of UO fuel into the coolant system.
2 Based on examination of the Euratom fuel rods and shrouds adjacent to the failed element, it has been concluded that a similar single rod clad failure would i
not cause adjacent rods in the six high performance fuel bundles to fail. 'Ihe l
188 high power fuel rods employ the same basic mechanical design and fabrication methods as the Type C Big Pock Point reload fuel which has performed with no known failures to date. 'Ihe developmental fuel is expected to be equally effective in maintaining fuel rod geometry and containing fission products.
Nevertheless, if a fuel rod failure should occur, the release of radioactivity to the coolant will be detected in the manner described and any fiscion products released to the atnesphere via the air ejector and stack will remain within the technical specification limits for continuous operation and will not create a hazard to the health and safety of the public.
4 i
IDSS OF COOIRTP As shown in Table i, lines 26 and 29, the two intermediate performance fuel bundles will generate the nost power in the least anount of fuel. Decay heat in these small diameter rods, following the design basis accident (double ended rupture of a recirculating pipe), will cause the fastest rise in average enthalpy at the l
various fuel rod cross sections, if film blanketing and insulated fuel rod con-ditions are assumed. However, the residual or stored energy in the large dia:reter 1
4 fuel rods, which normally operate with peak heat generation rates of 25.4 107/ft,
results in avemge fuel enthalpies which are initially the highest in the core, i
but which would rise nore slowly than those of the smaller diameter rods following the design basis accident. Redistribution of energy within the rods (after heat transfer from the red ceases) also influences the rute of clad temperature increase.
For loss-of-coolant caused by primary system breaks which are an order of magnitude smaller than those involved in the design basis accident (see Figure 7-1 of Applicant's supplenent dated Novcmber 10, 1967 for break analysis summary), nest of the residual heat would be renoved by the coolant before the coolant is expelled from the primary system to the containment. For such breaks, the present emergency core cooling system will avide adequate cooling for the core including the'6 high performance'Iael bundles.
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For the 'very smll breaks, where the system would not depressurize to permit the low-pressure core tpray system to function, the core mid-plahe could be.
exposed and fuel damge could occur. The slow loss of water, which keeps the system pressurized after the water level has decreased below the core mid-plane, prevents fuel cooling by the 400 gpm core spray header until the pressure is reduced below 140 psig. - Because of this behavior, much of the normal fuel md clad will_ be overheated _ and perforated before the spray system can-del ver water to cool the fuel.. The presence of the six high performance bundles does not significantly alter the consequences resulting from breaks of this size.
~
In the design basis accident (DBA) rapid depmssurization of the system occurs allowing the spray. system to perfom its cooling function. If the spray l system cools the core as ' effectively as expecteg, peak clad temperatures in the norml Big Rock Point fuel will not exceed 1500 F.
However, the peak clad tezrperatures for game of the 188 high power density rods will reach temperatures as high as 3500 F.
Thecapabilitytocoolthefuelbyspraywaterwithoutexcessivegamge and fuel configuration changes after' initial clad te.mperatures exceerd 2500 F has not.been demonstrated. 'lherefom, we believe that the pmsence of the high performance ~ fuel in the Big Rock Point core mkes the consequences of the design basis accident somewhat greater than with the nomal com, even assuring that 4
engineered safety features including emergency core spray water and containment isolation perfom as designed. Conservatively, we have assumed that the entire fission product inventory is released fmm the six developrental fuel bundles.
With this assumption, the maximum site boundary doses are approximately as previously reported in the Big Rock Point Final Hazards Summary Report for the 10% core melt case. The large break accident is less severe in tems of radio-dCtive release, and is Considered less probable than a small break accident.
Our conclusion, based on a review of all relevant infomation, is that in the nere probable range of primry coolant rupture accidents, i.e., the small breaks, the increase in hazard caused by the high power density rods is negligible. For the larger breaks, the negnitude of the radioactive release is larger than it would have been for a normal core, but is still less than the radioactive release pre-viously evaluated as acceptable. 'Iherefore, based on a relative evaluation of the Big Rock Point core with and without the high performance fuel, we believe that the consequences of loss of coolant do not represent an unacceptable risk to the health and safety of the public.
In addition, the installation of a new 44 kv power supply enhances the pmbability that high pressure feedwater pumps will continue to deliver water to the reactor vessel after a primary system rupture. 'Ihis l
inprovement in power reliability increases the probability of maintaining high A
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pmssure feedwater flow to the steam drum following primary system ruptures as large.as 11.5 square inches. With high pressure water delivery capability in excess of 16 minutes, the core will not be uncovered. Under these con-
'ditions the fuel, including the developmental fuel, would not be damaged.
REACTIVITY EXCURSIOIS We have reviewed the reactivity excursion which could occur if one of the con-trol rod blades inmediately adjacent to one of the two small diameter develop-mental fuel bundlesbecomes separated from its drive shaft, sticks in the fully inserted core position while the control rod drive shaft is fully withdrawn, and subsequentlydrops out of the core at a time when the maximum reactivity excursion would occur. For the ' purpose of our evaluation it is assumed that the center-to-center spacing between the developmental bundle and another s:rall diameter developmental bundle is at the minimum spacing limitation of 16.5 inches, set forth in the technical specifications. The small dia:eter rods were selected for the evaluation because they contain the smallest mass of active fuel to accumulate the energy released during the transient, and they will, therefore, experience the sharpest fuel enthalpy rise. If the fuel reaches te:rperatures in excess of boiling tempemturea, the internal pressure will increase rapidly.
The licensee has assumed that when 425 cal / gram fuel enthalpy is attained, as a result of very rapid power transients, prompt rupture of the clad will occur and the fuel, which has reached 425 cals/gm, will be instantly dispersed in the form of small spherical particles into the coolant.
The vaporized portion of the fuel will transfer its energy instantly to the water and the re: raining energy will be transferred on a time dependent basis, depending on particle size. The lowest and, therefore, most conservative time constant of 4 milliseconds for particles of 20 mil size was used to calculate the transfer of the remaining heat to the water. The applicant believes that unless the threshold of prompt failure (425 cals/gm) is exceeded, there is no mechanism for prompt fuel dispersal and rapid energy conversion; therefore, primary system integrity would not be affected. The results of reactivity excursions with peak energies in excess of 425 cals/gm, calculated in the manner described, indicate that peak enthalpies of 625 cals/gm would be required to cause the unrestrdined vessel to lift 0.5 ft.
Similarly, the applicant shows that, if the peak fuel enthalpy reached approximately 800 cals/gm, the energy release associated with this condition would exceed strain lirpits and cause vessel rupture.
The effect of lowering the prompt threshold failure was evaluated by the applicant for an energy release spectrum associated with a rod drop accident of 0.021 delta k/k. It was shown that excessive vessel strain or vessel novement would not occur for the sane power excursion with prompt failure thresholds as low as 230 cals/gm.
i All of the calculations assume reactor shutdwn by the Doppler 'effect alone, follmed by control rod scram within the nomal 290 milliseconds.
To produce the maximum reactivity-insertion assumed in the evaluation, the follwing compounded errors and failures nust occur while the reactor is in the. hot standby condition (HSB): (1) violation of normal operating procedures by fully withdrawing a single control rod of maximum worth rather than equally -
. withdrawing control rods. in banks; (2) this particular rod must stick in the core
-and simultaneously become separated from the drive, and (3) the stuck control ' rod.
then must free itself and drop-from the core at a precise time to cause the maximum reactivity excursion..If any of these conditions do not prevail, the excursion either would not occur or would be of a smaller magnitude.
We have concluded that' a reactivity excursion resulting from dropping a control rod worth 2.1% delta k/k, as described by the applicant,'is-extremely improbable but is a reasonable. upper limit for the comparative evaluation of reactivity accidents.with and without centerne.'t fuel. This conclusion is based on the follwing considerations:
1.
The irradiation period of the developmental bundles will be limited -
to approximately 24-30 ncnths.
2.
The two small diameter Id fuel bundles with the least fuel and highest potential fuel enthalpies can affect the accidents associated with only 2 of the 32 control rods in the Big Rock Point core.
-3.
The developmental fuel bundles will be located in the center region of the core, rather than the periphery where maximum control rod worths would occur.
4.
The probability of the rod drop accident is minimized by written pro-cedures which govern operator novement of control rods and assure that control rods are latched to their drive shafts prior to initiating control red withdrawals for approaches to criticality.
5.
There has been no evidence of the poison section of control rods sticking or binding within the cores of existing boiling water reactors.
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- Investigation ~ of the control rod drop accident, while at power cohidition and with operator errors equivalent to those assumed in the HSB analysis, indicates the maximum contml red worth to be only 0.95% delta k/k compared to 2.1% delta k/k for' the HSB rod worth, and the resultant excursion energy would be only 75 cals/gm.
'Ihis value, when added to the normal energy while at power, results in peak fuel enthalpy of approximately 280 cals/gm compared to 450 cals/gm for the HSB excursion.
Since the latter case is nore severe, our evaluation is based on the HSB condition.
t-Because of the tim dependence for prompt release of the fuel energy to the coolant during the reactivity excursion, as described by the applicant, and the lack of suitable te.3ts to confim the overall validity of the calculational nodel and assumptions,_ we cannot accept without reservation the conclusions based on _these calculational methods. The application indicates that, with a peak energy density of 450 cals/gm #or the control rod drop accident, the instantaneous release of the transient enargy in fuel above 320 cals/gm would correspond to about 32 Mi-sec energy burst. Similarly, 64 Mi-sec of energy would be released promptly if all of
- the energy above 230 cals/gm were considered. Since. fuel melting begins at enthalpies of about 220 cals/gm and is complete at about 280 cals/gm, we believe that the prompt energy release values derived in the above manner are reasonable boundaries. Our independent calculations, based on equivalent explosive energy releases, give vessel lift and stmin values higher than the applicant's estimation. We have detemined that vessel lift could be as much as 0.4 ft and maximum vessel strain will be below 0.7%.
We consider strain of 5% and lift below 0.5 ft to be acceptable as this amount of vessel novement would not damage major primary piping. We therefore believe the integrity of the primary system will be maintained throughout a reactivity excursion resulting from dropping a control rod worth as much as 2.1% delta k/k although there is a possibility that the 3-irch spray header pipe connection near the top of the reactor vessel might be damaged if all of the energy in the fuel above 230 cals/gm is released promptly. If the spray header should fail, the core mid-plane would not be uncovered for at least 18 minutes following a double-ended break of this 3-inch pipe. To provide a backup for the core spray system, we are requiring an additional means of introducing water from the fire main as an emergency core cooling system. This system will supply enough water to keep the core covered in event the spray system is damaged so that fuel temperatures will not rise excessively and core geometry will be preserved. The valves to activate this sytem will be located in an accessible location and will be opemted manually.
&st of the damage resulting from a control rod drop under the conditions described would be in the adjacent high performance fuel bundle with significantly less damage to other bundles in the immediate vicinity of the dropped control rod. We have concluded that sufficient control rods will scram within 0.290 seconds of scram signal activation to restore suberitical reactor conditions. Soluble poison nay be added during the 18 minute cool down and depressurization period prior to addition of core flooding water from the fire system.
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'IEONICAL SPECIFICATIONS We have reviewed the proposed nodifications of the technical specifications submitted by the applicant in conjunction with this high performance fuel irradiation prcgram. 'Ihe requested nodifications include ~ descriptions of the developmental fuel assemblies, and special operating and administrative restrictions appropriate for operation with the developmental fuel. We have made'some changes in the proposed modifications in the interest of clarification
-and two new specifications have been added.. The new specifications require:
(1) renoval of the advanced performance developnental bundles until the results of initial destructive examinations have been evaluated and (2) provisions for utilizing water from the fire main for core cooling as a supplementary core cooling system.
'Ihe changes made in the proposed modifications have been discussed with the applicant and he is in agreement with the revisions.
CONCLW ION We have concluded that there is reasonable assurance that the health and safety of the public will not be endangered by operation of the Big Rock Point Plant with the developmental fuel described in the applicant's Proposed Change No. 13.
We believe, therefore, that the Technical Specifications of License No. DPR-6 may be revised as indicated in Attachment A to Amendment No.1.
Donald J.
xovholt Assistant Director for Reacter Operations Division of Reactor !dcensing Date: JAN 3 01968 l
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