ML20002D105
| ML20002D105 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/26/1967 |
| From: | Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20002D102 | List: |
| References | |
| NUDOCS 8101190444 | |
| Download: ML20002D105 (9) | |
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T.2LE 1 COMPARISON OF MI!!IlLE CRITICAL HEAT FWX PATIO (MCHFR) WITH NEW kND 01D CO Power Shape C Type Fuel at 122% Power Beginning of Life Radial Axial Old Critical Heat Flux New CHF Fatio Many Factor Factor Cormlation Correlation New/Old Control Rods Inserted 1.4 1.45(6) MCHFR 1.99(8) 2.23(7) 1.12 Steam Quality Xc 0.151 0.1265 2
Q/Ac BIU/hr ft 273,000 325,000 2 342,000 342,000 Q/Ap BIU/hr ft Middle of Life 1.32 1.652(5) MCHFR 1.92(5) 2.32(7) 1.21 Xc 0.0415 0.116 Q/Ac 374,000 326,000 Q/Ap 374,000 374,000 End of Life 1.623 1.335(2) MCHFR 1.42(8) 2.11(6) 1.48 All Control Rods Removed Xc 0.2205 0.167 Q/Ac 184,000 292,000 Q/Ap 392,000 392,000 Xc = Steam quality at the locat on of MCHFR Q/Ac = Heat Flux at the locati/m of MCHFR Q/Ap = Peak Heat Flux Axial length of fuel divide ( into nine nodes. Numbers in parentheses irdicate nede number whem node number 1 is at the bottom of the fuel.
14 cal power factor of 1.3 used for all cases pasented ('IWX - Feb. 6,1967).
122% overpower - basis 120% scram set point - refennce p. 26, Amendment 14, Nov. 15, 1963.
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Vibmtory ccr:pacted ziroaloy clad fuel - 109-3.449 inches diameter rods --
12-0.34 inch diameter corner rods,11 x 31 fuel rod array - 0.577 inch rod pitch.
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Frtrn the gbove table, it can be seen that the peak local heat flux, 392,000 BTU /hr ft' at 122% power level, occurs at the erd of core life ard is 5%
greater than the middle of core life peak heat flux. Also, it is evident, since the minimum critical heat flux ratio at the erd of core life is 1.42 for 122% po,er, using the old correlation, and below the tech spec MCHFR of 1.5, that rated power at the erd of life would not be pemitted. Towazxi the erd of an operating cycle when the central control rods are finally witMrawn frm the botte of the core, the flux peak moves toward the bottom as the axial factcr column above shows, e.g., at the beginning of life, the axial peak occurred at node 6 in contrast to nada 2 at the end of life. The greatest gain in MCHFR according to the tal>1e above is in the high quality region where the ratio of the new-to-old MC5fFR is 1.48.
Although the peak heat flux is 5% gmater than the mid-life value, the heat flux at the MCHFR location is 10% less at the end of life than at the beginning or middle of core life as the table shows when the new CHF correlation is used.
EVAIDATION Conramers Power Company has proposed that the new General Electric CHF correlation presented in APED 5286 be used to update the Big Rock reactor thermal hydraulic parameters to the most recent and representative numbers.
The present CHF correlation was based on approximately 1000 data points, essentially all of which were from tests of single, internally heated rods with annular coolant flow. Since the corner red was clearly design limiting before it was economically feasible to ire.uycrate enrichment variation within each fuel element, the geometry was selected to be representative of this rod.
The few multi-rod points available at the time fell significantly and con-sistently above the single rod design line.
The new CHF correlation is based on approxinately 700 multi-rod data points taken on four-red and nine-rod test sections. Since the comer rod Eecmetry, red spacing, pressuzw and flow correspond to actual boiling water core con-ditions according to General Electric, the correlation (as can be observo:1 in the caparison below) is valid for General Electric boiling water reactors including Big Rock Point.
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i TABII 2 Multi Rod Geometries to Develope
- Big Rock Po et.
The CHF Correlation Fuel Bundles Heated length inches 36, 45, 4b.and 6r 70.
Hydraulic Diameter inches 0.324 - DJ485 0,508 (1)
Rod to Rod Spacing inches 0.095 - 0 J87 0.129 Rod to Channel Spacing inches 0.060 - 0.335 0.157 (2)
Pressure psia 600 - 14SD 1350 6
2 6
.46
.7 x 10 Flow Rate lbs/hr ft 0.2 - 1.E x 10 Steam Quality 0 - 0.6
.23 maximum Heating Distribution Uniform ad Incrmsed non-unifom Heat in Cbrner lb:t or Centrul. Rod
- Information received by telephone from Conramer's Fower Company March 3,1967 (1) The effecte of hydraulic diameter appear to in ruegligible over the range from 0.? to 0.5 inches.
(Ref APED-5285)
(2) The rod-to-channel spacing is important in dernemining the critical heat flux. Its effects become noticeable fur very :small rod-to-channel distances, i.e., lower than the range examined.
(Ref APED-5286)
Great care was exercised, according to G.I. to mah sure that the test geometries were representative of boiling water reacter fuel m:rangements in order to match the anticipated local flow and steam quality conditisns, since the application of four and nine rxxl data points is justified nnly if the flow and enthalpy re-distribution takes place in the same manner in the test sections and reactor fuel assemblies. 'Ihe test rxxis consisted of Inconel-X tubes through which electrical current passed to simulate the nuclear heat generation.
A multichannel model was developed to predir:t the ccaolant behavior in complex l
gecmetries, including reactor fuel asaamb W a.
E r model subdivides the geometry into several parallel and adjoining channels which run over the entire flow lergth. Each channel has its own characteristic hydraulic diameter and heat input and the effects of boiling in some or all of the channels are included within the accuracy of engineering correlations of two phase flow effects. The mass flow rate into the channels is assumed to be uniform, and local flow in
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- each channel is obtained by allowing the flow to redistribute itself between the various channels until they all sense the same pressum at each axial position. Tne best fit QF correlation for an internally heated annulus was employed for the corner rods, and circular pipe correlations were used for the spaces between the rods. 'Ihe proportionality constant, i.e. t he mixing constant was adjusted to obtain acceptable agree ent.between rer:e1 and test critical heat values. The uncertainty associated at.this time with these mixing constants is w of the reasons for developing nulti-rod correlations based on average alal enthalpy or steam quality, and average flow for the fuel bund: s rather than local QF mter pwWties. 'Ihis analytical model was applied to the four and nine rod test sections for which critical heat flux data are available. In this manner the ua2 of.
four and nine red test results to predict the critical heat flux in reactor fuel assemblies containing up to 121 rods has been justifieid.
The QF limit lines described by the proposed new nulti-rnt J.E. correlation were derived with minimal red spacers in the test assembly. 'Ihe red-to-rod and red-to-channel spacing of the test assemblies was maintained by slender spool type spacers desiptwi to have negligible influeme on the flow. Other tests were perforred to show that fuel rod spacers used in boiling water reactors could have a beneficial effect by promoting mixine, and thereby ircreasing the critical heat flux, although in this particular Big Rock application the benefits of irrproved mixing have not been considered in calculating the HChTR.
G.E. har stated hat the new QF corralation based upon nulti-rtxi test data cannot be applied indiscriminately to any gearretry and in particular, the correlation is valid only for lattices typically found in boiling water reactors. We have listed the important physical characteristics of the Big Rock Pbint fuel geometry in Table 2.
Further, we have been assured by G.E.
that the new correlation also accounts for 3cc.1 peaking factors of the magnitude to be encountered in Big Rock cores, i.e.1.24 For the present core configuration:
1.
All of the original stainless steal clad fuel bundles yet remaining in the care are in the outer low power region where the flow is also reduced. (Most of these fuel assemblies will be removed during the June i
19G7 refueling.)
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2.
'Ihe central, high power region of the core, is occuplied by type B arxi C zircaloy clad fuel assemblies which utilize two zone enrichment with the outer two rows containing the lower enrichment fuel rods. Three I
small diameter rods as in the original fuel assemblies (in scue cases l
l cobalt targets vwplace one of the thme srall diameter corner rods) are provided at the corners of the fuel bundle to improve heat transfer conditions in the regions which are nomally limiting.
3.
The corner (or adjacent rod where coabit targets are substituted for the corner rods) red although amaller in diameter and of lower enrichment than the inner bundle fuel rod enrichment is the thermally limiting red, i.e. the location of the minimum critical heat flux ratio (MCHFR). Consumers Power Cmpany reported (telecon April 21, 1967) that although the highest local peaking factors with the improvenents described above is 1.2, they will continue to use 1.3 for additional conservatism in their calculations.
(Refer to Section 5.1.5 of the Technica'. Specifications for Big Rock Point Nuclear Power Plant for fuel bundle description.)
Earlier fuel failum following the maximum credible accident of a loss of core coolant, is an undesirable characteristic inherited with the increase in local power peaking resulting fran the new General Electric CHF correlation.
In response to this concern Consumers Power Cm pany reported by telephone on April 26, 1967, that fuel cladding perforations would occur several seconds earlier than most mcently reported in the evaluation which acccrnpanied Proposed Change No. 8 December 23, 1965. Further, the curve representing percentage of fuel rod perforations as a function of time after MCA will cross the old curve (see attached figure) when 20 to 30% of the fuel clad has failed.
The volume percent of UO2 fuel over the 30000F clad melting temperature is not noticeably affected. The significance of these fuel rod failum changes caused by increased core power peaking in relation to previously reported post accidents conditions sunrarized in the attached curve is in our opinion negligible.
In sumary, our <nnelusion that the new General Electric critical heat flux (CHF) correlation for the Big Rock Point operatioral themal calculations may be safely used is based on the premise that the test assembly heat fluxes and geometries suitably ratch the Big Rock Point boiling water reactor conditions and is further supported by the following factors:
1.
Rated power level of 240 MWt will rerrain unchanged.
2.
Average power density at rated condition 46 MW/ liter will remain unchanged.
3.
The minimum critical heat flux ratio, MCHFR, of 1.5 at 122% overpower continues in effect.
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4.
The Technical Specifications limiting heat flux and fuel md power remain unchanged at 530,000 B7U/hr ft2 and 17.2 KW/ft respectively. (392,000 B7U/hr ft2 heat flux which corresponds to 122% overpower at end of core life in the mpresentative results presented in Table 1 is equivalent to 10.3 KW/ft for the small 0.344 inch diameter corner rods.)
5.
The average fuel bundle exit steam quality will rerrain unchanged at 8.1%.
Peak power generation in the corner rods prior to MCA is, consistent with 6.
the infomation presented above, 8.45 KW/ft, and although this value is approximately 5% greater than the midlife peak value for the same core con-figuration we believe this is an insignificant change.
CONCUJSION We have concluded, for the reasons stated above, that Proposed Change No.12 does not present significant hazards considerations not described or implicit in the hazards su m ary report and there is reasonable assurance that the health and safety of the public will not be endangered.
Accordingly, we believe that the Technical Specifications of License No. DPR-6 may be revised as indicated in Attachment A.
A nald J.
holt Assistant Director for Reactor Operations Division of Reactor Licensing Date:
May 26, 1967 l
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l ATIACRENT A
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CONS m EaS s ta C m PANy CHANGES 'ID 'ECHNICAL SPECIFICATIONS LICENSE NO. DPR-6 GANGE NO. 12 1.
Delete the footnote in section 5.2.1(b) and replace with:
"* Based on correlation given in " Design Basis for Critical Heat Flux Condition in Boiling Water Reactors," by J. H. Healzer, J. E. Hench, E. Janssen, and S. Invy, September 1966 (APED 5286 and APED 5286, Part 2)."
2.
Delete the footnote in section 8.3(a) and replace with:
"* Based on correlation given in " Design Basis for Critical Heat' Flux Condition in Boiling Water Reactors," by J. H. Healzer, J. E. Hench, E. Janssen, and S. Imvy, Septerrber 1966 (APED 5286 and APED 5286, Part 2)."
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BIO ROCK POINT Tran?,isnts cf Rrzeter System Rupture For Maxianan Credible Accident Evaluation With Metal Water Reaction Peak =35.6 PSIA 35 300 r
I Dost Incident Spray System in Operation
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Enclosure Pressure /
' Core Spray System 100 30 250 in Operation 1
7 Peak =223 F g
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Enclosure Pressure, FHSR (No Metal Water Reaction) is#
N Enclosure 80 p,
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Temperature (ek i
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k' b 20 6o so s
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y Per Cent of Fuel Il 3
Rods Whose lad
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Exceeds 15 at Y
Bottest Point, Perforation 15 $00 Voltane Per Cent M
I hMelting 20 10 50 y.
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10 10 10 ABSTRACT FROM BIG ROCK TECil SPEC Time After Accident % conds PROPOSAL #8 (DECEMBER 23, 1965) s
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