ML20031C428
| ML20031C428 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/20/1967 |
| From: | Palladino N Advisory Committee on Reactor Safeguards |
| To: | Seaborg G US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20031C429 | List: |
| References | |
| NUDOCS 8110070137 | |
| Download: ML20031C428 (4) | |
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COPY ADVISORY COMTIEE ON FI/COR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D. C.
20545
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Decerber 20, 1967 HonoraDie Glenn T. Seaborg Chaiman U. S. Atxic Energy Comission Washington, D. C.
Subject:
REPORT ON BIG ROCK POINT NUCLEAR PL M
Dear Dr. Seaborg:
At its eighty-ninth meeting, Septerber 7-9, 1967 and ninety-second meeting, 7-9, 1967, the Advisory Cctrmittee on Reactor Safeguards considered Dece:rber the application by Consumers Power Co:rpany for authorization to install six high perfomance develognental fuel assmblies in the Big Rock Point Reactor.
During its review, the Comittee had the benefit of discussions with repm-sentatives of Consumers Power Company, General Electric Company, and the AEC i
Fegulatory Staff and of the documents refemnced. Subcc:nnittee meetings were held in Washington, D. C. on September 6, 1967 and Decerber 5, 1967.
Tne six developmental assemblies are designed to obtain infomation concern-ing fuel mds operating with central melting. Two of the six assemblies will centain fuel rods with a 0.57 inch outside diareter designed for incipient center melting during full power operation of the reactor. The fuel rods in the other four assenblies have an outside diareter of 0.70 inches and are to operate with considerable center melting. Censumrs Power Company plans to install the test fuel asserblies during the next refueling period, currently scheduled for January 1968. After a peak exposure of 3000 L'D/T has been reached in the test fuel-the reactor will be shut down and the test fuel recoved for exarination. These fuel rods will be subjected to nondestructive exarination. After a three-renth decay period, four fuel rods will be shipped to Vallecitos Nuclear Center for destructive tests. During continuation of the irradiation program, the reactor will be shut down for periodic fuel examination.
8110070137 671220 PDR ADOCK 05000155 0
COPY Honorable Glenn T. Seaborg Lrember 20, 1967 Because experience with high perfonunce fuels, peticularly those in which some of the fuel is in the molten state, is limited, the Comittee recoatnends that the applicant use caution in prosecution of the test program. The Com-mittee recocinends that extra caution be given to operations if signs of failure of the experimental fuel appear. To this end, the Comittee urges that wrk be done to develop improved methods for prompt detection of gross failure of an experimental fuel. element.
As a further precaution, the Comittee recomends a more conservative approach to returning of the 0.70-inch center melt rods to the reactor after the first irradiation period. Specifically, it recommends that, after the first inspection period (at approximately 3000 %f/T burnup),
the 0.70-inch dianuter rods not be Inturned to the reactor until the re-sults on the four rods selectea for destructive exarination have been obulned and show to the satisfaction of the AEC Regulatory Staff that the red themal conditions and perferrance during irradiation am essential-ly as pmdicted and that additional irradiation ray be undertaken safely.
The applicant reported that, as additional protection against core meltdown in the unlikely event of rupture of the core spray line, means will be pro-vided for pnrpt delivery of water from the fire protection system to the reactor pmssure vessel. The Committee recommends ccepletion of this system before the experimental assemblies are operated.
In recognition of the safety aspects of the operation, Consumers Power Campany has several programs in progmss to improve the safety character-istics of the Big Pock Point Reactor. These include installation of another external power line (nearly completed); examination of the safety instru-mentation system for possible negation of safety action through single faults; installation of a nossle support system to protect against rod ejection in the unlikely event of control red nossle failure; and nodification of the emergency core cooling system to improve its protective capacity (within a year or two). Tne Comittee approves of these steps, and urges expedition of their ccrnpletion.
'Ihe applicant is also planning a detailed inspection program for the primary system piping and ccrnponents.
It was stated that the inspection is hampered by the radioactivity in the prirary system, so that special arrange; rents and fixtures will be required. The Comittee recoamends that the in-service inspection program be expedited, and that consideration be given to a hydro-l static test of the primary system, Lefore the developmental assemblies are installed, at a pressure only sli;htly lower than the pressure-relief valve l
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Honorable Glenn T. Seaborg December 20, 1967 The Conrnittee telieves Jut, in view of the favorable characteristics of the Big Rock Point site,.a M subject to the e ndations stated above, the reactor can be operatea with the high performance test assemblies with-out undue risk to the health and safety of the public.
Sincerely yours, 4-ORIGINAL SIGNED BY N. J. PALLADINO N. J. PallaMno Chairman i
References Attached.
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COPY Hencrable Glenn T. Seaborg Decereber 20, 1967
' References - Big Rock Point Reactor 1.
Censu:rers Power Company letter, dated May 26, 1967, with Proposed Change No.13 to the Technical Specificatiorus for the Big Rock Point Reactor.
2.
Consumers Pbwer Canpany letter, dated May 29, 1967, with Errata Sheet and Figure 25 for ".oposed Change No.13.
3.
Censumers Power Ccrrpany letter, dated August 15, 1967, with Answers to Additional Infomation Requested Regarding Proposed Change No.13.
4 General Electric 'IWX, dated August 24, 1967, to Division of Reactor Licensing re Pasults of 2-Pump Trip frcen Pated Pcuer.
5.
Consumers Pcuer Ccupany letter, dated Noverrber 10, 1967, with Additional Infomation Required in Support of Proposed Change No.13.
6.
Peferences Associated with Answer to Question 2 of Answers to Additional Infonration Requested Reganiing Proposed Change No.13.
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. LICENSE AUTHORITY FILE C0FY
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Not Rep @
SAFE 1Y EVALUATION BY THE DIVISION.0F REACTOR LICENSING I
DOCKET NO. 50-155 CONSUMERS POWER COMPAW PROPOSED AMDiDMENT NO.1 INTRODUCITON By letter dated May 26, 1967, the Consumrs 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.
Supplenental infomation was submitted on August 15, 1967, Noves er 10, 1967, and December 14, 1967.
A~endment No. I would pennit insertion of six high perfonnance developmental fuel bundles into the Big Rock Point core as part of the nomal core complement of 84 fuel bundles. The developmental fuel will be irradiated until the rest depleted fuel rods acquim 21,000 MID/T U average exposure..This irradiation program, designed'to investigate the performnce characteristics of fuel rods with center melting, is an extension of the high perfomance UO9 program sponsored jointly by the U. S. Atomic Energy Comission and Euratom. The pro-posed change was reviewed by the Advisory Comittee on Reactor Safeguards (ACRS) witich concluded in its report dated December 20, 1967, that "the reactor can be operated with the high perfomance test assemblies without undue risk to the health and safety of the public". A copy of the ACRS report is attached.
DESCRIPI' ION l.
Tne Table, which follows on pages 3 and 4 sets forth the important characteristics I
of the high perfonnance fuel bundles and presents a comparison thereof with the i
Type "C" fuel bundles nost recently used to refuel the Big Rock Point reactor.
l Tne salient aspects of that table are further discussed below.
Tne six high perfomance developmental fuel bundles, containing 324 fuel rods, 2
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will operate with peak heat fluxes of 450,000 +50,000 BIU/hr ft. To achieve the design objectives of perfomance and power output, it will be necessary that the high perfomance fuel bundles be repositioned into higher neutron fluxes I
during their approximately 2 - 21/2 year in-core irradiation. The use of three 1
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different U en-ichments, i.e., 4.3, 5.0, and 5.6%, to control individual 35 fuel rod poder generation within a fuel bundle pemits rest of the fuel bundle power to be generated in about 60% of the fuel rods. The reraining fuel rods contain depleted fuel, i.e., only 0.22% wt percent U and thereforegenerateonlyasmallfraction,about3%,ofthefue$bu,ndle power. Tao of the fuel bundles, designated as intemediate perfomance fuel, contain 0.570 inch diameter fuel rods and are designed to produce incipient center melting.
One of the interrediate perfomance bundles contains pellet type fuel and the other vibratory compacted powder fuel. Tne other four bundles, designated as advanced perfomance fuel, contain 0.70 inch diameter rods and are designed to produce substantial center melting. Two cf the advanced perfomance assemblies will contain U3,3 in the fom of sintered cored pellets and the other two will contain U3 in the fom of powder. Selected rods will contain tungsten wafers to mini $ize axi.al nove ent of the rolten UO 2
du-ing operation. All six high perfomance fuel bundles will be distributed within the core so that de minint:m 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. The handles of the high perfemance fuel bundles will be notched to perit visual verification of their core pcsitions. Correct placement of the rods within the fuel bundle gecretry is assured by 3/8 inch identification letters sta ged on the fuel rods.
Tne developrental fuel bundles are designed to pemit disasse-bly when renoved from de core during refueling. Tnis feature will facilitate examination and replacerent of all individual rods. Selected irradiated rods will be shipped to the Vallecitos Nuclear Center for destructive examinaticn after a suitable period of radioactive decay in the spent fuel storage pit at Big Rock Point.
Rods 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 interrediate and advanced perfomance fuel in significantly larger than the peak power density in the Type "C" Big Rock Point fuel.
Lines 24, 25, and 26 show that the total UO in each advanced 2
perfo mance fuel bundle is significantly greater than in either Type C cr the interediate perfomance fuel, bu: the total arount of U in either an advanced or interrediate performance fuel bundle is less than for d5 Type C fuel bundle.
2 Funher, the arount of UO fuel in the pcraer generating rods of each proposed i
fuel bundle is noticeably less than the arount in the power producing rods of a Type C fuel bundle. Tne cross-sectional area of the developmental fuel rods, I
however, is increased in accondance with lines 2 and 6 of Table I by a factor of 17 for the 0.570 inch diameter rods and 2.65 for the 0.700 inch dia.neter rods.
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TABLE 1
'COXPARISON OF PROPOSED FUE'L
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TO --
BIG ROCK ~" TYPE C" FUEL i
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1 Proposed Fuel Big Roc k Intermediate Advanced Type "C" Performance Performance Fuel-1 Geometry, Fuel Rod Array 11 x 11(1) 8x8 7x7 21 Standard Rod Dia=eter inches C.449 0.570 0.700
_3 i Number Standard Rods per bundle 121' i
36 29 4
Nu=her Special Rods with depleted uranium 0
28 20 5i Special Rod Diameter inches 1 0.344(Z) 0.570 0.700 6i Standard Rod ';'ube Wall inches 4
0.034 1
0.035_
0.040 0.040 7
I Special Rod Tube Wall inches 0.031
! 0.035 8
Rod Pitch inches 0.577 0.807 0.921 l
9
-Active Fuel Length inches I' 70 66 - 67.3 65 - 66.3 10 UO2 D4.nsity, Percent Pellet Powder Pellet Powder j.
Theoretical Pellet-85 powder 94 85 94 85' 11 J Fill' gas i
He He
'He 12 i Spacers per bundle 1
5 5
u 5
f Clad material f Zr-2 I Zr-2 i Zr-2 1
Wt Zircaloy clad per bundle j
-l Pounds 90 62.7 67.6 15 l DHcggggsinsidefuel inches l 0.497 0.88 i 0.85
_16 i Rod to Rod clearance inches 1 0.160 0.237 i0.221 17 i Rod to channel clearance inches 0.128 L 0.160 l0.155 18 Number of bundles UO2 Pellet N.A.
1 2
UO2 Powder 40 1
2
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19 Steady state heat Tech Specs 1 500,000 500,000 500,000 flux limit BTU /hr ft2 1 550.003) 20 Power generation Pallet Powder Pellet Powder at limit KW/ft(6) 19(3) 21.8 21.8 26.9 26.9 21 Peak powcr 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 power 4.1 16.4 16.4 19.4 19.4 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 We UO / bundle kg.l 133 136 129 159 152 24 2
UO / bundle (U235) kg 4.83 Hot Rods Hot Rods Hot Rods I Hot Rods 25 2
3.76 3.7 4.6 4.35 26 UO2 in power producing Hot Rods Hot Rods Hot Rods Hot Rods rods / bundle hg 133 76.5 72.5 95 89.5
'. Table 1 Cont'd COMPARISON OF PROPOSED FUEL BIG ROCK " 'PE C" FUEL Proposed Fuel i
Big Rock Intermediate Advanced Type "C" Performance Performance Fuel 27 Heat transfer surface ft2 79 Hot Rods Hot Rods Hot Rods Hot Rods bundle 30.2 30.2 29.4 29.4
_2_8 i Average bundle power MWt 2.86 1
3.25 3.25 3.25 3.25 29 Radial power factor in core 1.30 1.21 1.21 1.14 1.14 30t MCRFR at 122% Power
>1.5 l
1.53 1
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 6 x 10-5
.32 Temp coef. 250C at (4)
BOL delta k/kOC
+3.2 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 l
.24
.27
~~t Cold pellet fuel to clad gao inches Powder 0.012 for pellets 0.013 for nellets 35l Central hole, pellet (5)
Pellet I
fuel inches N.A.
0 0.100 W/cm N.A.(5) 59 - 62 59 - 62
[28000CPelletfuel 36 K dt i
5000C Powder fuel W/cm 49 1 49 49 4
37 Fuel enrich =ent vt7. U-235 2.965.2 4.3;5.0; 4.3;5.0; 5.6;0.22 5.6;0.22; 3-Fuel Rod lifetime MWD /T U 15,000 Ave. hot rods Ave hot rod ave. rod 21,000 21,000 Sv=bols i
MCEFR Minimum critical heat flux ratio.
g Hydraulic diameter of coolant channels.
D (1) 4 corner rods may be cobalt targets l
(2) 8 special corner rods and 4 Cobalt target rods for radial supports j
(3) Incipient melting.
Tech spec limit is 500,000 l
(4) Proposed Change No. 13 dated May 16, 1967 (5) NA - not applicable (f i Melting starts in powder fuel m: 19.4 KW/f t I
Melting starts in pellet fuel at 24.2 KW/f t
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.~. The applicant has indicated that pcreer in each 7 x 7 fuel bundle is genemted in essentially 29 rods, because the reraining 20 fuel rods contain depleted fuel
" cold" (deple$. Tne arrangement of the hot (high power pmducing) rods and with little U fucl) rods in an annular geo etry provides the raximum cold surfaces adjacent to the power producing rods and optimizes power distribution and coolant utilization within the four 7 x 7 advanced perfomance fuel bundles.
The 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 raxmra number of fuel rodg operating at or near the desired heat flux, i.e., 450,000 + 50,000 BTU /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 ratrix; therefore, the average pcreer generation in each of the 0.700 inch dia.eter rods in the 7 x 7 array is greater by the ratio of 36/29 than the power generated in the 0.570 inch diareter rods of the 8 x 8 fuel bundles.
Tne pcwer of the 9.ree types of enriched U rods (high pcraer rods) decreases with exposure as the U inventory decreases. Tne power in the depleted rods (cold.
5 rods) increases wdn exposure because of plutonium production. To compensate for this reduction of pcreer in the high pcreer. rods, the fuel bundles will be noved to positions of higher radial power factors as irradiation pmceeds so that the design heat flux of the high power rods can be raintained throughout exposure life-time.
It has been estimated that the relative pcuer generated by the depleted fuel rods increases over the fuel bundle lifetime by a factor of less than 3, and there-fore, would not significantly aff act the Jifetime themal perfomance of the high.
power rods.
Since a linear power density of 19.4 W/ft results in. incipient center relting in the vibratory compacted powder fuel rods and 24.2 W/ft correspondingly causes centcrline melting in pellet fuel rods, it can be seen fmm lines 21 and 22 that nest of the rods will be near centerline melting at the hottest locations when the reactor is at rated power.
In the case of the hottest, large diameter, powder fuel rods, there will be colten fuel over a significant rod length, i.e., from appm xirately 1 to 1.5 feet on each side of the axial power peak, with the mavir.um molten fuel cross-sectional area approxirately 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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Nomal plant operaticns, including ccnsideration of the significantly higher average heat fluxes fmm the hot centerelt fuel rods than the peak heat flux for the norral Type "C" Big Rock Point Core, solid to liquid fuel phase changes, and average hot rod fuel exposure of 21,000 MG/T U, which is significantly greater than pmsent fuel life-tim expocures; 2.
loss of coolant conditions, including consideration of the greater fuel rass, higher temperatures (fuel enthalpy or stored energy), and higher rod pwer levels in the proposed fuel rods in relation to the ability to dissipate the residual and decay heat; 3.
Reactivity excursions associated with the hypotnetical control rod drop accioent.
NOFF.AL PIANT OPEPMION Fuel Pod Power: Tne fuel rods have been sized erd arranged within the bundle so as to achieve the maximum cooling. Test results with hot rods adjacent to cold rods s4-4'ar to the arran p.ent in the six proposed developrental fuel e
bundles confirm that the multi-channel nodel for calculating the Minimum Critical heat Flux Ratio (FNR), currently used by General Electric to detemine fuel rod heat-transfer limits in boiling water reactors, is conservative for the proposed rod arrangerent. Using this calculational rodel, it was deterrdned that the MChTR at 122% of rated pcwer is greater than the minimum value of 1.5 perritted by the existing technical specifications.
It has also been determined that the peak clad g
terperature at rated power will be approxirately 900 F for those hot fuel rods which rarain in the core for the full irradiation lifetime of 21,000 MG/T U.
Tne capability to operate at 122% of rated power without exceeding heat transfer li-its provides a sufficient margin, in our opinion, to ensure fuel cladding integrity during norral operation.
The licensee has reported that failure of the main coolant circulation pumps cr the rain p wer generator with the high pcwer density fuel in the core will not result in v,*.olation of the rini. um critical heat flux lirits. 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 rot be excessive. Tne MCG during the pcwer and flow transient would be greater than the minimum value of 1.5 perritted by the technical specifications. We are in agreenent with the calculational methods used in developing their conclusion. Therefore, it is our opirlon that such unanticipated transients would not cause a significant release of radicactivity from the fue' mds into the prira"v coolant system or divinish the integrity of the primary 3 stem.
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'Ihe high linear power density of the developmental fuel mds causes significant increases in the fuel terperature co w d to Type "C" Big Fock fuel, and results in possible melting of the UO, at the center of the hot fuel mds whenever the power density is above 19.4 w/ft for powder fuel or 24.2 IM/ft for pellet fuel.
Tnese values correspond to KdG valt.as of 50.5 W/cm and 63 W/cm. Only the 58 large diameter, vibratory mpacted, hot fuel mds will' have significant rolten fuel. Tne remaining 130 high power density fuel rods will have incipient fuel nelting conditions at the peak axial power locations. Tne effects of increased fuel tenperatum, as well as center melting, necessitated new design provisions which have been developed and tested during the AEC-EUR.IOM sponsored "UO High 2
Perfomance Prrgram".
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 minirize this potential cause of cladding failure, a large fuel-to-clad diametral gap, i.e., 0.012 inches for 0.570 inch diameter fuel mds and 0.013 inches for the larger rods, has been provided. At rated power, the diaretral clearance will not be less than 1 mil over the lifetime of the fuel.
Rod Fabrication: A depleted UO Pellet is placed at the end of each active fuel 2
coltrm 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 accoradate the phase change volume expansion of UO n mlting, based on 9.6%
2 measured U3 volume changes during destructive exannation of the irradiated fuel.
A 15% void $ pace has been provided in e.rh of the 94 hot fuel rods using vibratory compacted pwder by conpacting to only 85% of the theoretit.1 U0 density. In 9
consideration of the test irradiation experience and th; observe 3 resultant sintering densification and accompanying center voids, ve believe this void is su"ficient to acconrodate the volume expansion caused by melting of pwder fuel.
The applicant has calculated that pwer levels of 136% and 200% of rated power would be recuired to cause cladding failure of the small and large diameter pellet fuel rods, respectively. 'Ihe ma>.imum mit fraction for pellet fuel at 122%
of rated power is only 22% compared to the design capability of 71%.
Ihe raximum melt fraction of powder fuel at 122% of rated power is 45% compared with a l
theoretical limit of 100% melting which represents a substantial margin above j
the MCEIFR power limit.
Fuel Micration: Unile operating with nolten fuel conditions, the possibility of fuel migration within the fuel rod exists.
Studies perforned by General Electric indicate that increased or decreased fuel concentration along the axial direction of the hot rods causes insignificant reactivity effects. Moreover, as mentioned earlier, only the 58 hot, large diameter powder fuel rods have ac. ten center fuel i
over a sufficient length to pemit fuel novement by this neans. A proposed nodifi-cation to the technical specifications, which restricts the rate of initial power i
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increase between 170 MYt and 240 Mdt to 1/2 Wt per minute, allows tim for fuel migration. This _ is particularly significant for the powder fuel mds, when the fuel is sintered and densifies the first tim melting occurs. This restriction 4
also permits gradual axial expansion of the molten fuel when the peak power has -
roved downward. We believe that this is a prudent mstriction to avoid the remote possibility of fuel red damage from these factors. Also, tungsten wafers are placed at 18 inch intervals in 4 hot, large diameter, pellet rods and 4 hot, large diamter, powder rods to minimize the axial novement of molten fuel for t-comparative perforrance evaluation.
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Fuel Rod Lifetime: The capability of the hot "uel rods to achieve 21,000 WD/T U _-
average accumulated exposure without failure wm a confirmed by removing 4 representative fuel rods during each core refueling for destructive examination.
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at the Vallecitos Nuclear Center. These examinations will reveal the anount of fuel burnup, extent of melting, fission product distribution, gas release, fuel novenent and dimensional changes. Deviations from predicted conditions will be carefully evaluated to determine the effect on predicted fuel lifetine. The times; required for fuel cooling, shipment, and examination result in about a five nonth
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delay after rod re:: oval before results of the examinations are'obtained. We believe that the confimation of design performance by the destructive examinations is valuable in ensuring safety of continued irradiation for these bundles. There-
-fore, the rodified technical specifications require that after the first rod anovals,' the four advanced perforrance bundles should not be reinserted until the initial destructive examination results are cbtained and evaluated.
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Fuel Rod restructive and Non-Destructive Tests: A number of non-destructive tests i
will be performed on each of the developantal fuel bundles during each core re-fueling period. Each bundle will be leak tested by the " dry. sipping nothod". Tne bundles will be examined visually using television and an underwater periscope.
L Tne diameter of each of the 188 hot rods will be measured with a go-no-go gauge to t
detect dimensional changes resulting from irradiation.
In addition, several rods will be measured with the profiloneter for comparison with the pre-irradiation traces.
An unexpected dianeter increase will be cause to discontinue irradiation of the affected rod.
Tnese examination procedures, first performed when the hot rods l
reach less than 15% of the expected lifetime, will provide added assurance that l
the fuel is perfcming in accordance with predictions.
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Continuous wnitoring of the gases from the air ejector ronitor will detect fuel element rupture. To provide early warning of a fuel failure, the monitor alam set points will be lowered to about twice the backgcaund level for the duration of these developmntal fuel irradiations. The few fuel red failurec experienced during the Euratom high power density fuel de.velopmnt program caused the release of only a smil fraction of UOg fuel into the coolant system.
Based on exara. nation of the Euratom fuel rods and : ouds adjacent to the failed elemnt, it has been concluded that a similar single rod clad failure would not cause adjacent rods in the six high perforrance fuel bundles to fail. The 189 high pcwer fuel rods employ the sam basic mchanical design and fabrication mthods as the Type C Big Rock Point reload fuel which has perforred with no known failures to date. The developrental fuel is expected to be eqmily effective in naintaining fael rod gemtry and containing fission products.
Nevertheless, if a fuel rod frA should occur, the release of radioactivity to the coolant will-be det.
in the ranner described and any fission products released to the atnesphere u the air ejector and stack will remain within the technical specification linuts for continuous operation and will not create a hazard to the health and saSty of the public.
I LOSS OF C)OLET As shwn in Table I, lines 26 and 29, the two internediate perforrance fuel bundles wi'l generate the rest power in the least a cunt of fuel. Decay heat in these srall diareter rods, follcwing the design basis accident (double ended rupture of a recirculating ripe), will cause the fastest rise in average enthalpy at the varios fuel rod cross sections, if film blanketing and insulated fuel rod con-ditions are assured. However, the residual or stored energy in the large diameter fuel reds, which norrally operate with peak heat generation rates of 25.4 Kd/ft, results in average fuel enthalpies which are initially che highest in the core, but which would rise core slowly than those of the smaller diameter rods follcuing the design basis accident. Redistribution of energy within the rods (after heat transfer from the rod ceases) also influences the rate of clad terperature increase.
l For loss-of-coolant caused by primar" system breaks which are an orxler of magnitude i
smaller than those involved in the d_ sign basis accident (see Figure 7-1 of l
Applicant's supplement dated Noverber 10, 1967 for break analysis strrary), rest l
of the residual heat would be re oved by the coolant before the coolant is xpelled l
from the primary system to the containmnt. For such breaks, the present l
emergency core cooling system will provide adequate cooling for the core l
including the' 6 high perforrance fuel bundles.
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4 For the very srall breaks, where the system would not depressurire to pemit the 1cw-pressure core spray system to function, the core mid-plane could be ex;csed and fuel darage could occua. Tne 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 pmssum is reduced below 140 ps;g. Because of this behavior, much of the nomal fuel red clad wiu be ovemeated and perforated before the spray system can deliver water in cool the fuel.
Le presence of the six high perfca ance bundles does not sigm.ficantly alter the consequences resulting from breaL cf this size.
In the design basis accident (DBA) rapid depressuaization of the system occurs allowing the spray system to perfonn its cooling function.
If the spray system cools the core as effectively as expecteg, peak clad temperatures in the norral Big Rock Point fuel will not exceed 1500 F.
However, the peak clad temperatures for game of the 188 high power density rods will reach temperatures as high as 3500 F.
The capability to cool the fuel by spray water without excessive gamge and ftal configuration changes after initial clad terperatures exceed 2500 F has not been deronstrated. Therefore, we believe that the presence of the high perferrance fuel in the Big Rock Point core mkes the consequences of the design basis accident somewhat greater than with the norral core, even assuming that engineered safety features including emergency core spray water and centainment isolation perfonn as designed.
Conservatively, we have assumed that the entim fission product i:wentory is released from the six develcpmental fuel bundles.
With this assumption, the raxi,mrn site boundary doses are approxi:ately at previously reported in the Big Rock Point Final Hazards Simrary Report for the 10% core melt case. The large break accident is less severe in terrs of radio-active release, and is considered less probable than a srall break accident.
Our conclusion, based on a review of all relevant inforration, is that in the nore probable range of prinary coolant rupture accidents, i.e., -he srall breaks, the increase in hazard caused by the high power density rods is negligible.
For the larger b"aaks, the magnitude of the radioactive release is larger than it would have bee.1 for a norral cene, but is still less than the radicactive release pre-viously evaluated as acceptable. Therefore, based cn a relative evaluation of the Big Rock Point core with and without the high perferrance fuel, we believe that the consequences of ic_s of coolant do not represent an.1 acceptable risk to the heal.h and safety of the public.
In addition, the installation of a new 44 kv power supply enhances the probability that high pressure feedwater pumps will continue to deliver water to the reactor vessel after a primev system rupture. Tnis irgrovenent in pcreer reliability increases the probability of raintaining hig;n
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.- pressure feedwater flow to the steam drum following prirary system ruptures as large as 11.5 square inches. With high pmssure water delivery capability in excess of 16 minutes, the core will not be uncovemd. Under these con-ditions the fuel, including the developm ntal fuel, would not be damaged.
REACTIVITl EXCURSIONS We have reviewed the reactivity excursion which could occur if one of the con-trol rod blades innediately adjacent to one of the two srall diamter develop-mntal fuel bundlesbecoms 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 tim when the raximum reactivity i
excursion would occur.
For the purpose of our evaluation it is assumed that the center-to-center spacing between the developmental bundle and another small diameter deve]opmental bundle is at the rinimum spacing limitation of 16.5 inches, set forth in the technical specifications. The imall diameter rods were selected for the evaluation because they contain the srallest rass 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 temperatums in excess of boiling temperatuies, the internal pressure will increase rapidly.
Tne licensee has assumed that when 425 cal / gram fuel enthalpy is attained, as a result of very rapid power transients, prcrgt rupture of the clad will occur and the fuel, which has reached 425 cals/gm, will be instantly dispersed in the form of srall spherical particles into the coolant. Tne vaporized portion of the fuel will transfer its energy instantly to the water and the remaining energy will be transferred on a time dependent basis, depending on particle size. The lowest and, therefore, nest conservative tire constant of 4 milliseconds for par-icles of 20 mil size was used to calculate the transfer of the remaining heat to the water. Tne applicant believes that unless the threshold of pru::pt failure (425 cals/gm) is exceeded, there is no mechanism for pmmpt fuel dispersal and rapid energy conversion; therefore, prirary system integ*ity would not be affected. Tne results of reactivity excursions with peak energies in excess of 425 cals/gm, calculated in the manner described, indicate tP at peak enthalpies of 625 cals/gm would be required to cause the unrestrained vessel to lift 0.5 ft.
Sirilarly, the applicant shows that, if the peak fuel enthalpy reached approxirately 800 cals/gm, the energy release aesociated with tnis condition would exceed strain lirits and cause vessel rupture.
The effect of lcwering the prompt threshold failure was evaluated by the applicant for an energy release spectrum associated with a md drop accident of 0.021 delta k/k.
It was shown that excessive vessel strain or vessel novemen v mld not occur for the same power excursion with prompt failure thresholds as 1s as 23G cals/gn.
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-- w All of the calculations assume mactor shutdown by the Doppler effect alone, follcwed by control rod scram within the norral 290 rilliseconds.
To produce the raximum reactivity insertion assumed in the evaluation, the following compounded errors and failums must occur while the reactor is in the hot standby condition (HSB): (1) violation of norral operating procedures by fully withdrawing a single control rod of naximum worth rather than equally withdrawing control rods in barks; (2) this parti-clar rod must stick in the com and si::altaneously becoas separated from the drive, and (3) the stuck control rod then must free itself and drop from the core at a precise tire to cause the raximum reactivity excursion. If any of these conditions do not prevail, the excursion either would not occur or would be of a smaller ragnitude.
We have concluded that a reactivity excursion resulting from dropping a control rod worth 2.1% delta k/k, as described by the applicant, ?s extrerely improbable br is a reasonable upper lirit for the comparative evaluauion of reactivity accidents with and without centerrelt fuel. Tnis conclusion is based on the follcaing considerations:
1.
Tne irradiation period of de developrental bundles will be lirited to approxicately 24-30 conths.
2.
The two srall dia.~eter rod fuel bundles with the least fuel and highest potential fuel enthalpies can affect the &ccidents associated with_only 2 of the 32 control rods in the Big Rock Point cora.
3.
Tne develop ental fuel bundles will be located in the center region er the core, rather than the periphery where maxi:rs cen rol rod worths would occur.
4 The probability of the rod drop accident is rinirized by w*itten pro-cedures 'Jnich govern operator rovement of control rods and assure that control rods are latched to their d*ive shafts prior to initiating control rod withdrawals for approaches to critica]irj.
5.
There has been no evidence of the poison s'cticn or control rods sticking or binding within -he cores of ensring boiling water reac c"s.
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Investigation of the control rod drop accident, while at power condition and with operator errers equivalent to those assumed in the HSB analysis, indicates the maxi:mnt control md worth to be only 0.95% delta k/k compared tc 2.1% delta k/k fer the HSB rod worth, and the resultant excursion energy would be only 75 cals/p.
Tnis value, when added to the nomal energy while at power, results in peak fuel en-halpy of approxirately 280 cals/p compared tc 450 cals/gm for the HSB excursion.
Since the latter case is core severe, our evaluation is based en the HSB condition.
Because of the ti:ne dependence for prorpt release of the fuel eners to the coolant during; the reactivity excursion, as described by the applicant, and the lack of suita:le tests to confirm the overall validity of the calculational rodel and assu ptiens, we cannot accept without reservation the conclusions based on these calculational mthods. Tne application indicates that, with a peak energy density of 450 cals/p #cr the control rod drop accident, the instantaneous release of the transient enargy in fuel above 320 cals/p would correspond to about 32 W-sec energy burst.
Simila-ly, 64 W-sec of energy would be released promptly if all of the energy above 23:. ls/gm were considered.
Since fuel melting begins at entha' pies of about 220 cals/gm and is c:.plete at about 220 cals/gm, we believe that 9e prenpt energy release values derived in the above manne" are reasonable bounteries. Our independent calculations, based en equivalent explosive energy releases, give vessel lift and strain values higher than the applicant's estimation. We have deterrined that vessel lift could be as much as 0.4 ft and maxi:m:n vessel strain. '_11 be below 0.7%.
We consider strain of 5% and lift below 0.5 ft to be acceptable as this arount of vessel rove ent would not camage raior primry piping. We aerefore believe the integrity of the prirary system will be raintained throughN: a reactivity excursion resulting fm. drcpping a control rod worth as much as 2.1% delta k/k although there is a pos'ibility that the 3-inch spray header pipe connection near the top of the reactor vessel night be da aged if all of the energy in the fuel above 230 cals/gm is released promptly.
If tne spray header should fail, the core mid-plane would not be uncovered for at least 18 ri~ttes following a double-ended break of this 3-inch pipe. To provide a backup
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- ccre spray system, we are requiring an additicnal means of introducing water frca tne rire rz2n as an emergency core coc.,.ing system.
nis system will supply i
enough water to keep the core covered in ev?nt the spruy system is damaged so -hat fuel te peratures will not rise excessively and core 3ecme*ry will be p-esened. Tne valves to activate this sytem will be located in an accessible
- cation and will be operated manually.
Fcs: of the darage resulting frcm a control rod drop under the conditions desc-ibed would be in tne adjacent high perfomance fuel bundle with significantly less crage to other bundles in the immediate vicinity of *he dropped control rod. We have concluded that sufficient ecntrol rods will scram within 0.290 seconas of scram signal activation to restore suberiti " ' w +cr conditions.
Soluble p ison ray be added du-ing the le minute cool down and depressuricati:n period prior to addition of core floodiiig water frcm the fire system.
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[l TECW. CAL SPECIFICATIONS
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We havt reviewed the pmposed nodifications of the technical specifications submitted by the applicant in conjunction with this high perfonrence fuel
-irradiation program. The requested nodifications include descriptions of the -
l developmental fuel assemblies, and special operating and administrative restrictions appmpriate for operation with the developental fuel. We have made some changes in the proposed nodifications in the interest of (larification and two new specifications have been added. The new specifications requim:
(1) renoval of the advanced performance developnental bund'es until the results of initial destructive examinations have been evaluated and (2) provisions ' for :
I utilizing water from the fire main for core cooling as a supplementary core cooling system.
The changes made in the proposed nodifications have been discussed with the applicant and he is 'in agreemnt with the revisions.
4 CONCLLMON We have ecnoluded that there is reasonable assurance that the health and safety j
of the public will not be endangered.by operation of the Big R6ck Point Plant with the developnental fuel described in the applicant's Proposed Change No.13.
We believe, therefore, that the Technical Specifications of License No. DDR-6 may be revised as indicated in Attachment A to Amendment No.1.
1 Donald J. xovholt Assistant Director for Reactor Operations.
Division of Reactor Licensing I
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