ML20030A464
| ML20030A464 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/13/1975 |
| From: | Lamley R, Sewell R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8101090573 | |
| Download: ML20030A464 (19) | |
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CORSumBIS L
Power P-
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( rea we t utento. Avenue. Jackson, Echtgen 49201. Area Code S17 78R-OSSO
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October 13, 1975
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Director of Nuclear Reacter Regulation p kh h $:
US Nuclear Regulatory Commission d
Washington, DC 20555 g
4' DOCKET 50-155, LICENSE DPR-6 BIG ROCK POINT PLANT 4
Transmitted herewith are three (3) original and thirty-seven (37) conformed copies of a request for change to the Technical Specifications of License DPR-6, Docket No 50-155, issued to Consumers Power Company on May 1,1964 for the Big Rock Point Plant. This proposed change, when approved, vill allow the use of an initially all-uranium fuel type as reload fuel for the Big Rock Point Plant. This fuel type is similar in design to the 11 x 11 reload G fuel type that was inserted in the reactor over the past several refueling outages and is also similar to 11 x 11 all-uranium reload fuel that was uti~
lized as reload fuel during the mid-1960s.
This change was necessitated by the August 11, 1975 Memorandum and Order of the Commission. This Memorandum and Order made it clear that plutonium in excess of 50 kilograms could not be utilized until completion of a discrete environmental review and public hearing. As the 50 kilogram limit is being approached, it is necessary to obtairl approval of an all-uranium fuel design such that that fuel may be utilized as relosd fuel durin6 the next refueling outage.
The next refueling outage at the Big Rock Ioint Plant is scheduled to commence January 9, 1976. Because certain modifications vill be made to the plant during this refueling outage, it is enticipated that the plant will be shut down until early April 1976. Thus, it will be necessary to obtain approval of this Technical Specifications change prior to start-up at the end of this refueling outage.
The discussion contained in the attached proposed Technical Specifications change is presented in a manner that conforms as closely as possible to the
" Guidance for Proposed License Amendments Relating to Refueling" transmitted by Mr K. R. Goller's letter of June 23, 1975 We have experienced two diffi-cultiw in confoming completely to the " Guidance." The first difficulty is that much of the analysis was completed for the reload G-lU fuel prior to i
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issuance of the " Guidance." The second is that we cannot define a core loading map for the planned reload core until the condition of the existing fuel is ascertained during the refueling outage. Thus, the limits proposed in the M}
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Technical. Specifications are envelope limits upon which the core loading pat-
. terns and subsequent operating restrictions,. if any are necessary, will be
' developed. We belitve that setting a core loading scheme prior to ascertain-ing the. condition of the existing fuel is contrary to the principle of as low as. reasonably. achievable when applied to radioactive effluents.
In addition, and as stated in this Technical Specifications change, the analysis to. demonstrate that the new fuel design meets the requirements of Appendix K of 10 CFR, Part 50 has not yet been completed. This analysis is presently scheduled for submittal January 15, 1976. The delay in completion is caused by the existing workload associated with other Appendix K analyses that are required.
- %<&Y J
w Ralph B. Sewell
' Nuclear Licensing Administrator CC: JGKeppler, USNRC File f:-
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CONSUMERS POWER COMPANY Docket No 50-155 Request'for Change to the Technical Specifications l
License No DPR-6 I.
For the reasons hereinafter set forth, the following chanEes to the Technical Specifications of License No DPR-6 issued to Consumers Power Company on May 1, 196h for the Big Rock Point Plant are requested:
A.
1.
Delete the columns ertitled Reload B & C and Reload E and add the following column to Table 5.1.
General Reload G-lU Geometry, Fuel Rod Array 11 x 11 Rod Pitch, Inches 0.577 UO Rods 109 2
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Cobalt - Bearing Corner Rods k
Gadolinium - Bearing UO R ds h
2 Inert Spacer Capture Rod (Zr-2) 1 Zircaloy Rods 3
Spacers per Bundle 3
Fuel Rod Cladding Material Zr-2 Wall Thickness, Inches 0.034 Fuel Rods Outside Rod Diameter, Inches 0.hh9 Fuel Stacked Density, Percent Theoretical 91.6 Active Fuel Length, Inches Standard Rod TO Fill Gas Helium 3,95%
- 2. -Delete the references to Reloads B, C and E in Note 1 of Table 5.1.
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3.
Delete the reference to Reload E in Note 3 of Table 5.1.
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B.
Change Section 5 2.1(b) to read as follows:
5.2.l(b) Reactor Operation The reactor operation shall be so limited as to be consistent with the most conservative of the following:
Reload E-g and Modified E-G Reload F. J-1 & J-2 Reload G
_ %1U Minimum Core Burnout Ratio at Overpower 1.5*
1.5"*
1.5**
Transient Minimum Burnout Ratio in Event of Loss of Recirculation Pumps From Rated Power 1.5 1.5 1.5 Maximum Heat Flux at Overpower, Btu /h-ft 500,000 395,000 407,000 Maximum Steady State Heat Flux, Btu /h-ft 410,000 32h,000 333,600 Maximum Average Planar Linear Heat Generation Rate, Steady State, kW/Ft Stability Criterion: Maximum Measured Zero-to-Peak Flux Amplit 's, Percent of Average Operating Flux 20 20 20 Maximum Steady State Power Level, MW 2h0 2h0 240 t
Maximum Value of Average Core Power Density @ 2hD MW, kW/L h6 h6 h6 t
Maximum Reactor Pressure During Power Operation, Psig 1,h85 1,h85 1,h85 Minimum Recirculation Flow Rate, Lb/h (Except During Pump Trip Tests or Natural 6
6 6
Circulv ion Tests as Outlined in Section 8) 6 x 10 6 x 10 6 x 10 Maximum mwd /T of Contained Uranium for an Individual Bundle 23,500 23,500 23,500 Rate-of-Change-Of-Reactor Power During Power Operation:
Control rod withdrawal during power operation shall be such that the average rate-of-change-of-reactor power is less than 50 MW per minute when power is t
less than 120 MW, less than 20 MW per minute when power is between 120 and 200MW,and10kW Per minute when power is between 200 and 2h0 MW.
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t C.
Delete Figures 5.2 and 5.3 and renumber existing Figures 5.h through 5.8 as 5 2 through 5.6, respectively. Add new Figure 5.7 attached.
1
- Based on correlation given in " Design Basis for Critical Heat Flux Condition in Boiling Water Reactors," by J. M. Healzer, J. E. Hench, E. Janssen and S.' Levy, September 1966'(APED 5286 and APED 5286, Part 2).
- Based on Exxon Nuclear Corporation Synthesized Hench Levy.
j C**To be determined by linear extrapolation from the following points:
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i MAPLHGR (kW/Ft)
Planar Average Exposure
/Wd/STU)
Modified F E-G, F, J-1, J-2 Reload G Reload G-1U 0
6.38 See Note 1 200 9.5 9.h 2,h80 6.81 5,000 99 97 5,511 6.78 10,000 99 97 12,125 6.89 15,000 9.8 9.6 16,53h 7.03 20,000 87 8.6 23,1h8 7.02 25,000 8.7 8.3 30,86h 7.15 Note 1 - These values will be transmitted with the results of the additional analysis scheduled for submittal January 15, 1976.
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I II. Discussion - Reload G-1U Fuel A.
Introduction and Summary The purpose of this proposal change is to allow the use of en initially all-urani*
fuel with acceptable ECCS performance characteristics in the Big Rock Point reactor.
Present1/ licensed fuel types are 11 x 11 mixed-oxide fuel and 9 x 9 all-uranium fuel. The 9 x 9 all-uranium fuel was not utilized as a reload fuel at the last refueling. The 11 x 11 mixed-oxide, fuel was. This shift in bundle array was made to improve fuel operating eind ECCS performance.
The use of an initially all-uranium fuel for the next reload batch was necessitated by the Commission " Memorandum and Order" of August 11, 1975 which requires NEPA review prior exceeding 50 kilograms of re-cycled plutonium. This 50 kg limit will allow insertion of only-eight additional mixed-oxide assemblies in the reactor. The remaining as-semblies (approximately 20) will be of an initially all-uranium design as described in this proposed change.
This submittal contains information regarding mechanical design, nuclear design, thermal hydraulic parameters, and accident analysis including misplaced B2el rod analysis, reactivity insertion accident and primary system integrity and LOCA analysis. The LOCA analysis was done prior to the approval of models that conform to 10 CFR 50, Appendix K.
The heatup analysis will be performed utilizing approved models and sub-mitted to the NRC by January 15, 1976.
B.
Operating History The plant was started up for Cycle 13 following refueling in July 197h.
The core loading consisted of 60 9 x 9 fuel assemblies and 2h 11 x 11 fuel assemblies. The off-gas release rate stabilized following start-up at approximately 1350 pCi/s (corrected for specific gravity). The plant has operated continuously since that time, with the exception of one five-month outage which was not fuel related and several brief periods when the power level was reduced for a short period of time at a core
. output of approximately 200 MW. The off-gas release rate as of early October was approximately 2500 pC1/s (corrected for specific gravity).
a.,
1 The power level of approxinately 200 MW was chosen to conform to t
-MAPLHGR limits (which are ECCS related and the most limiting limits) tror 9 x 9 fuel..
It is anticipated that the' plant will continue to operate at approxi-mately 200 MW until the shutdown for refueling which is presently t
scheduled to occur on January 9, 1976.
A summary of Cycle 12 fuel performance and tests performed at the beginning of Cycle 13 is contained in Special Report No 19 dated October 24, 1975.
C.
General Description Projected-core loading patterns will not be made until later in the present cycle when the off-gas performance is known and the fuel per-fbrmance can therefore be better estimated.
Current plans call for loading 20 11 x 11 all-uranium assemblies and 8 11 x 11 mixed-oxide assemblies of the presently licensed Reload G design.
The final core loading vill be developed based on replacement of all
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fuel that is unsuitable for reuse in the reactor in order to make z
effluent releases as low as reasonably achievable and to maintin in-plant exposures as low as practical. The final core loading vill also be required to conform to licensed limits which are established for the G-lU fuel and those licensed limits which exist for the other fuel types. These limits include MAPLHGR, Minimum Core Burnout Ratios, Maximum Heat Flux and others.
D.
Reload G-1U Fuel Design 1.
Summary Reload G-lU fuel assemblies are identical to the Reload G assem-blies currently loaded in the Big Rock Point core, except that each assembly:
4 a.
Contains four inert Zircaloy rods rather than one.
b.
Contains 113 fuel rods rather than 116.
c.
Replaces the mixed-oxide rods with rods of high enriched (4.6 w/o U235) uranium, d.
Relocates the four gadolinia bearing fuel rods.
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The cobalt loading may vary between 0-70 grams per-foot.- The analysis was based on 23 grams per foot rather than 35 grams per foot utilized in the Reload G analysis.
The most significant results of this design change are the absence of mixed-oxide rods from each' fuel assembly and a reduction in the calculated peak clad temperature during a' postulated Loss of Coolant Accident'(LOCA). The maximum. calculated cladding temperature during-a design' basis LOCA, utilizing identical models generally in conform-ance with 10 CFR 50, Appendix K, has been reduced to 2116 F 1 rom.the value of 2286*F calculated for the Reload G assemblies. Heatup calcu-lations for Reload G fuel have been performed in full accordance with Appendix K and were submitted July 25, 1975 Heatup calculations for Reload G-1-U assemblies utilizing approved mndels will be submitted by January 15, 1976. The differences between the Reload G assemblies and the Reload G-lU assemblies and their safety significance are discussed in detail in the following sections.
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2.
Fuel' Description a.
Mechanical Design Reload G-lU fuel is identical to the Reload G assemblies in fuel rod and pellet diameters, rod-to-rod pitch, cladding wall thickness, active fuel length per rod, tie plate and spacer' design, and overall assembly envelope. Figure 1 presents the layout of the various rod types within the' assembly. The Reload G-lU fuel is designed to comply with the same mechanical design requirements as the Reload G assemblies. The principal mechanical features of the two designs are compared in Table I which illustrates the near identity of the two designs.
b.
Nuclear Design The neutronic parameters calculated for the Reload G..lU ascemblies are in most respects the sr.me as those calculated for the Reload G assemblies. The net neutrcnic effects of the increase from one to four inert Zircaloy rods and the change of 2h mixed-oxide rods to 25 highLenriched uranium rods are:
(1 (1) A decrease of about 2% in the assembly maximum local peaking factor.
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FIGURE 1 RELOAD G-1 ASSEMBLY R0D MATRIX "D@D@@@@@DD@
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OGOQQOOOOGQ O@DDDD@@@@@
OGOOOOOOOOG OQQOOOOOOOO OG8@@OQQGQO OGO@@QQQOOO
+OQQQ OQQQOOO+
Symbol Rod Type fio. per Assembly 36 4.6 wt,% U-235, U02 D 1:i"!:
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@3.z.t.,u-23s.u0 32
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2 g2.3.t.,u-23s.u0z ie cieaaie9 O w:n:~9" c
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J TABLE I
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Mechanical Design Features Reload G Reload G-1U Assemblies Assemblies Fuel Pellets UO Pellets 2
Material Density, % Theoretical 93 5 93.5 Dish Volume, %
2.0 2.0 Stacked Density, % Theoretical
_91.6 91.6 UO -Pu0 Pellets 2
2 Material Density, % Theoretical 93 5 Dish Volume, %
2.0 Stacked Density, % Theoretical 91.6 UO -Gd 0 2
23 Material Density, % Theoretical 91.5 93 5 Dish Volume, %
2.0 2.0 Stacked Density, % Theoretical 89 7 91.6 Pellet Diameter, Inches 3715 3TJ-Fuel Rods 3
Fuel Length, Inches 70 70 Pellet-Clad Diametral Gap, Inches
.0095
.0095 Plenum Length, Inches 39 39 Clad OD, Inches
.449
.449 Clad'ID, Inches 381 381 Assemb34es No of UO Rods 88 109 2
No of Pu0 -UO Rods 24 2
2 No of UO -Gd 0 ds b
k 2
23 No of Cobalt Target Rods b
4 No of Zircaloy Rods 0
3 No of Zircaloy Spacer Capture Rods 1
1 Rod Array 11 x 11 11 x 11 Rod Pitch, Inches 577 577 Fuel Wt pug 2 + UO, Kg 1h4.8 141.0 2
Bundle Wt, Lb hh5 hhD Spacers - No 3
3 Frame Material Zr-b Zr-4 Spring Material Inconel 718 Inconel 718 8
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'$l (2) An increase in Assembly k= of 0 5% in the not. 25% void, no Xenon or Samarium condition and 0 7% in tre cold condition.
Neither of these effects is significant since the appropriate assembly reactivity and peaking parameters will be used in evaluating the overall core compliance with' reactivity limits and power density 1Lnits, and those limits are more influenced by the overall fuel loading pattern in the core than by these small differences in assembly parameters. A summary comparison
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of the neutronic parameters of the Reload G-lU assemblies and the Reload G assemblies is presented in Table II illustrating the minor differences between the designs. Figure 2 presents a comparison of the local peaking factors of the Reload G-10 assembly to those for t'- Reload G assembly illustrating the small decrease in peaking factor obteined.
c.
Thermal and Hydraulic Design
. The hydraulic design of the Reload G-1U assemblies is iden-tical to that of the Reload G assemblies. The only difference in the thermal performance of the assemblies evolves frcm the reduction in the number of active fuel rods from 116 to 113 and the associated change in the local peaking factor. As a consequence of these changes, the peak fuel temperature (at 22% overpower) increases frc m 3410 F to 3900 F; and the minimum critical heat flux ratio decreases from 1.68 to 1.03.
- However, the resulting peak pellet temperature is still far below the fuel melting point of about 5100 F and the minimum critical heat flux ratio is comfortably above the value of 1.00.
A comparison of the thermal hydraulic parameters for complete cores of Reload G assembly fuel and Reload G-lU fuel is pre-sented in Table III.
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TABLE II Nuclear Parameters Core Data
-Rated Power (MW) 240 1TotalCoreFlow(10{ Psia)
Operating Pressure 1,350 Lb/h) 12.3 Leakage Flow (%)
19 5 Core Inlet Subcooling-(Btu /Lb) 22.8' f
Core Average Void Fraction (%)
25 Number' of Control Rods 32 Equivalent. Core Radius (cm) 97 2 R,eload G-lU Reload G Assembly Average Enrichment Total w/o Fissile 3.88 3 98 v/o U-235 3.88 3.08 w/o Fissile Pu 0 90 Number of Enrichments per Assemb.y 3
h Number of Fuel Bearing Rods 113 116 Number of Gadolinia-Bearing Fuel Rods 4
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Concentration' of Gadolinia in Urania (w/o) ~
1.2 1.2 Water-to-Fuel Volume Ratio 2.69
'2.62 BOL Maximum Local Power Peaking (25% Void) 1.154 1.177 Reactivity Parameters
- Assembly k. Cold 1.236 1.229 Assembly k. Hot 25% Void (No Xe or Sm) 1.229 1.22h Ak. Moderator Temperature (68 F to 583 F)
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+.0161-Ak. Void Content (0% to 50%)
.0411**
.0405 Ak. Fuel Temperature (583 F to 959 F)
.0048**
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- Calculated for 23 gm/ft cobalt loading, except as noted.
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- Calculated for 35 sm/ft cobalt loading. Similar values expected for loading
- j. A
.of 23 gm/ft.
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1.043 0 942 1.130 1.086 1.073
~ 1.097 0 986-1.169 1.106 1.087 1.112 0 954 1.154 1.118
~1.109 1.176' 1.011 1.177 1.087 1.058 o','559 o.947 0 955 0 95h 1.c87 0 916 0.805-0.h58 0 902 0 911 0 908 1.138 0 936-0.853 0 918 octant Symmetry 0.788 0 773 0 909
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X.XXX Reload G-1U Y.YYY Reload G Figure 2 - Local Power Distribution 25% Void No Xe or Sm o mwd /MT Both Assemblies Contain 23 Grams /Ft Cobalt
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- 1, TABLE III Thermal Hydraulic Paramew---
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(Core Contains All of Each Type Fut.1)
Reload G Reload G-lU
-Core Conditions 8
Reference Design Thermal Output,-(MW )/(Btu /h) 240/8.191g 10 t
Total' Flow Rate, Lb/h 12 3 x 10 Effective Flow Rate for Heat Transfer, Lb/h 9 9 x 106 System Pressure, Nominal in Steam Dome, Psia 1350 Assembly Description Rod Diameter, Inches-0.hh9 0.449 Rod Pitch, Inches 0.577 0.577 Number of Active Rods 116 113 Total Fuel Length per Assembly, Feet 676.7
_659 2 Heat Transfer Area, ft2 79,48 77.h8 2
Flow Area, ft /in2 0.163/23.h4 0.163/23.hk Design Power Peaking Factors Fraction Generated in Fuel, %
96.6 96.6 Fuel Assembly Power Factor 1.45 1.h5 Local Peaking Factor 1.20 1.20 Axial Peaking Factor 1 51 1 51 Engineering Heat Flux Factor 1.04 1.0h
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Assembly Thermal Performance Maximum Heating Rate, kW/ft, at 22% Overpower 13.59 1h.0 Maximum Heating Rate, kW/ft, at Rated Power 11.1h 11.hk Average Heating Rate, kW/ft h.08 4.20 Maximum Heat Flux, Btu /h-ft at 22% Overpower 394,300 h07,000 Ma.imum Heat Flux, Btu /h-ft at Rated Power 323,500 333,600 2
Average Heat Flux, Btu /h-ft at Rated Power 118,500 127,100 Maximum UO Temperature, F, at 22% Overpower 3h10 3900 2
- Maximum Clad Temperature, F, at overpower 731 754 MCHFR at Overpower Conditions Axial Peak at X/L =.45 1.68 1.63 Coolant Subcooling at Core Inlet, Btu /Lb 22.8 22.8 Assembly Hydraulic Performance 1
Average Assembly Flow
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Innte Orifice Zone, Lb/h 132,900 132,900 Oute r Orifice Zone, Lb/h 80,500 80,500 (2h Assemblies on Periphery of Core) 6 6
Active Core Flow at Design Power Lb/h 9 9 x 10 9 9 x 10 Hot Assembly Flow at 122% Design Power (Reference Design Flow) 123,000 123,000
- f Assembly AP at Average Design Power
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(Includes Orifice AP) 5.37 Psi 5.37 Psi Hot Assembly Engineering Enthalpy Rise Factor 1.10 1.10
- Crud-Free Surface 12
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!1eload G-1U Fuel Accident Analysis 1.
Misplaced Fuel Rod Analysis
-In the analysis of the consequences of a misplaced fuel rod in the Reload G assembly design, it was found that the worst instance of such an error would be placing one of the Pu0 -UO r ds in the 2
2 outer row of UO r ds next to the corner cobalt target rod. For 2
the G-lU design, placing a high enrichment (4.6 v/o) rod in this location is expected to be the worst instance. Because of the lower absorption and fission cross sections of the uranium rod, the loading error would be less severe than vltn a mixed-oxide rod.
2.
Reactivity Insertion Accident and Primary Systtm Integrity Previous submittals have addressed the reactivity insertion accident for both a UO -fueld core (Type J-1) and a mixed-oxide 2
fueled core (Type G and Type G-1).
The Type G-lU fuel has kinet-ics parameters similar to previous UO deoigns, and mechanical 2
b and hydr <.alie design similar to the previously analyzed Type G and T,e G-1 designs. It'is expected, therefore, that previous p_ayses satisfactorily address the reactivity insertion accident.
3.
LOCA Analysis A heatup analysis has been performed on the Type G-1U fuel for the design basis accident using blowdown calculations reported in our submittal of December 18, 1972.
The models used in this heatup analysis are consistent in most respects with the requirements of 10 CFR 50, Appendix K.
The peak clad temperature was found to be 2116 F, and the maximum local cladding oxidation was 5 5%. Both values occur at the be-ginning of life (0 INd/MI'M) and are below the 10 CFR 50 limits of 2200 F and 17%, respectively.
'A report of this analysis, XN-Th-Sh, " Design Basis Loss-of-Coolant Accident Heatup Analysis of ENC Reload G-2 All Uranium Fuel in the 1 - Rock Point Power Beactor," is attached as Appendix A.
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E..
Reload G-1U Fuel Accident Analysis 1.
Misplaced Fuel Rod Analysis In the analysis of the consequences of a nisplaced fuel rod in the Reload G assembly design, it was fount that the worst instance.
of such an error would be placing one of tae Pu0 -UO r ds in the 2
2 outer. row of UO r ds next to the corner ccbalt target rod. For 2
the G-1U design, placing a high enrichment (4.6 v/o) rod in this location is expected to be the worst inctsnce. Because of.the lower absorption and fission cross sections of the uranium rod, the' loading error would be.less setere than with a mixed-oxide rod.
2.
Reactivity Insertion Accident and Primary System Integrity Previous submittals have addressed the reactivity insertion accident for both a UO -fueld core (Type J-1) and a mixed-oxide 2
fueled core (Type G and Type G-1).
The Type G-lU fuel has kinet-ics parameters similar to previous UO designs, and mechanical 2
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and L.Iraulic design similar to the previously analyzed Type G and Type G-1 designs.
It is expected, t! erefore, that previous analyses satisfactorily address the reactivity insertion accident.
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3.
IhCA Analysis A heatup analysis has been performed on the Type G-lU fuel for the design basis accident using blowdown calculations reported in our submittal of December 18, 1972.
The models used in this heatup analysis are consistent in most respects with the requirements of 10 CFR 50, Appendix K.
a 2116 F, and the maximum v
The peak cla'l tenperature was found to local cl W **; exidation was 5.5%. Eoth values occur at the be-ginning of life (0 mwd /MIM) and are below the 10 CFR 50 limits of 2200 F and 17%, respectively.
A report of this analysis, XN-Th-Sh " Design Basis Loss-of-
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Coolant Accident Heatup Analysis of ENC Reload G-2 All Uranium Fuel in the Big Rock Point Power Reactor," is attached as Appendix A.
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This LOCA analysis will be supplemented by an analysis in full compliance with the requirements of 10 CFR 50, Appendix K, foi-loving. completion of analyses for those fuel designs currently residing'in the reactor core.
It is expected that this analysis will be provided by January 15, 1976.
G..
Start-Up Program The testing and start-up program planned for the next refueling outage and subsequent start-up will include:
1.
Control rod drive. testing as required by the Technical Specifi-
' cations including neasuring scram times.
2.
Core shutdown marg a verification with the most reactive rod withdrawn.
3.-
Critical con +rol rad pattern.
h.
Measurement of the moderator temperature coefficient.
5.
Measurement of flux shapes during power escalation and comparison to computer predications.
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The results of the tests described in 2 through 5 vill be reported
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in a special report following ctart-up.
III. Conclusion Based on the foregoing, both the Plant Review Co=mittee and Safety and Audit Review Board have concluded that this change does not involve a significant hazards consideration.
CONSUMERS POWER COMPANY 9
R. A. Lamley 0
Vice ' resident Sworn and subscribed to before me this 13th day of October 1975 60h 4
%2
/ Sylvia B. Ball
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. Notary Public, Jackson County, Michigan My commission expires May 18, 1976.
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L.,
NRC DISTRIBUTION FOR PART 50 DOCKET MATERI AL
- (TEMPORARY FORM)
CONTROL NO:
11966 e
FILE:
r Co.
ufe DATE OF DOC DATE REC'D LTR TWX RPT OTHER FROM:
Ralph B. Sewell 10-13-75 10-15-75 xxx TO:
ORIG CC OTHER SENT NRC PDR vv.
NRC q
3-sigped 37 SENT LOCAL PDR xxx CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
'40 50-155 xxxx DESCRIPTION:
ENCLOSURES:
Ltr trans the following:
Request for Change to the Tech-Specs DPR-6.... notarized 10-13-75.... in reference
[ [ ].l'~l to the une of an initially all-uranium fuel type as reload fuel for the Big Rock Point Plant
' s ':
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(
( 40 cys enc'1 rec'd)
PLANT N AME: Big Rock. Point FOR, ACTION /INFC RM ATION 10-16-75 JGB BUTLER (L)
SCHWENCER (L) ilEMANN (L)
REG AN (E)
REID(L)
W/ Copies W/ Copies W4 Copies W/ Copies W/ COPIES CLARK (L)
STOLZ (L)
DICKER (E)
LEAR (L)
W/ Copies W/ Copies W/ Copies W/ Copies PARR (L)
VASSALLO (L)
KNIGHTON (E)
SPIES W/ Copies W/ Copies W/ Copies.
W/ Copies KNIEL (L)
PURPLE (L)
YOUNGB LOOD (E)
LPH W/ Copies W/ Cupies W/ Copies W/ Copies INTERNAL DISTRIBUTION
[fGflLE.?
TECH REVIEW DENTON LIC ASST A/T IND.
[H. GE ARIN (L)
DIGGS (L)
BRAITMAN
..f RC PDR SCHROEDER GRIMES SALTZMAN
,OGC, ROOM P-506A MACCARY GAMMILL
/OSSICK/ STAFF KNIGHT KASTN ER E. GOULBOURNE (L)
MELTZ CASE PAWLICKI RALLARD P. KREUTZER (E)
GIAMBUSSO SHAO SPANGLER J. LEE (L)
PLANS BOYD STELLO M. RU3HBROOK(L)
MCDONALD MOORE (L)
HOUSTON ENVIRO S. REED (E)
CHAPMAN DEYOUNG (L)
NOVAK MULLER M. SERVICE (L)
DUBE (Ltr)
SKOVHOLT (L)
ROSS DICKER S. SHEPPARD (L)
E. COUPE GOLLER (L) (Ltr)
IPPOLITO KNIGHTON M. SLATER (E)
PETERSON P. CO LLINS TEDESCO YOUNGBLOOD H. SMITH (L)
HARTFIELD (2)
DENISE J. COLLINS EGAN S. TEETS (L)
KLECKER
[ FILE & REGION (2)
A-(K G OPR LAIN AS e
E T LDR G. WILLI AMS (E)
EISENHUT BENAROYA V. WILSON (L)
WIGGINTON MIPC VOLLMER H AW.ESS R. INGR "4 (L) 5 M. DUNCAN (a EXTEnNAL DISTRIBUTION V-LOCAL PDR Charlevoix,~Mich y-TIC (ABERNATHY) (1)(2)(10) - N ATIONAL LABS 1 - PDR SAN /LA/NY y-NSIC (BUCHAN AN) 1 - W. PENNINGTON, Rm E.201 GT 1 - BROOKHAVEN NAT LAB 1 - ASLB 1 - CONSULTANTS 1 - G. ULRIKSON ORNL 1 - Newton Anderson NEWMARK/BLUME/AGBABIAN q 4,- ACRS HOLDING 4WegT P00R ORIGINAL i
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