Text

Forwards Proposed Change 27 to Tech Specs,Enabling Insertion of Fuel Design Designated Reload G-1.Facility Rod Drop Analysis Encl
	ML20030A460
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	06/20/1974
From:	Lamley R, Sewell R; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Oleary J; US ATOMIC ENERGY COMMISSION (AEC)
References
	NUDOCS 8101090562
	Download: ML20030A460 (34)
	v • d • e

{{#Wiki_filter:. [ k :'.m ry Ra Cy. CODSumerS 'I a . 3 POYlCT 7 w C0mpany e e '3 n, 1 = p.- Zoenerai ornee.: 2i2 we.t u.ch6gan Avenue. Jach.on. Michigan 49201. Area Code 617788-0550 -9 Q Y r C' Q June 20, K j Q -d l. in lp. G Sy ' ; Ik J m tb) 'f.'- Mr. John F. O' Leary, Director Re: Docket 50-155 JUN211974 r-,~ Directorate of Licensing License DPR-6 ' P-[8 US Atomic Energy Commission Proposed Technical N h Washington, DC 205h5 Specifications Change. - . l * & /. \\

Dear Mr. O' Leary:

Transmitted herewith are three (3) executed and thirty-seven (37) ronformed copies of a request for a change to the Technical Specifications of License DPR-6, Docket 50-155 issued to consumers Power Company on May 1, 1964 for the Big Rock Point Plant. This proposed change vill enable Consumers Power Company to insert into the reactor at Bf3 Rock Point a fuel design designated Reload G-1. This fuel has been designed and vill be fabricated by Exxon Nuclear Company. It is similar to the already licensed Reload G fuel presently in use at the Big Rock Point Plant with the excep-tion that it will contain four (b) solid zirconium rods instead of one (1) as contained in the Reload G fuel. This modification enhances calculated fuel clad temperature performance under assumed loss-of-coolant conditions. It is our intention to manufacture and install Reload G-1 fuel type.n the reactor after approval of this Technical Specifications change has been obtained. By letter 6ated July 27, 1973, Consumers Power Company was requested to reassess and quantify rod drop accident safety margins using current analytical methods. This reassessment is included in Appendix A to the attached proposed Technical Specifications change. In addition, by letter dated January 5,1973, the Directorate of Licensing requested additional information with regard to Big Rock Point ECCS calculations previously submitted. Our letter dated July 2h, 1973 provided a response to all except three (3) of these areas. The three (3) areas not covered in our July 24, 1973 letter were additional verification of the rod wetting model; inside and outside metal to vater reaction; and the basis of the gam =a smearing allowance. Additional information concerning the verification of the rod vetting model is included in the proposed Technical Specifica-tions change. A report covering the basis for the gamma smearing UOh/ h h

- (- .1 ( l Mr. John F. O'Liary 2 Proposed Technical Specifications Change June 20, 1974 allowance _(NEDO 2021h, " GAMMA Heating Distribution in a BWR Lattice During Normal Operation and Accident Conditions") was submitted by General Electric Company during spring of 197h. Information on the - inside and outside' metal to water reaction still has not been cal-culated as models are not presently available. This information will be provided with our submittal con.erning the Final Acceptance Criteria. With the exception of the inside and outside metal to water reaction information, we feel we have fulfilled the requirements of ~ your. January 5, 1973 letter. Yours very truly, .,/ wp/c. RBS/mel Ralph B. Sewell Nuclear Licensing Administrator CC: JGKeppler, USAEC I 6

' ^ .-( b CONSUMERS POWER COMPANY Docket No 50-155 Request for Change to the Technical Specifications Change No 27 License No DPR-6 I. For. the reasons hereinafter set forth, the following changes to the Technical Specifications of License No DPR-6 issued to Consumers Power Company on May 1,1964 for the Big Rock Point Plant are requested: A. Delete the columns titled Reload B & C and Reload E and add the following column to-Table 5.1. General Reload G-1 Geometry, Fuel Rod Array 11 x 11 Rod Pitch, Inches 0.577 Mixed Oxide (Pu0 -UO ) Rods 25 2 2 UO Rods 8h 2 Cobalt - Bearing Corner Rods k Gadolinium - Bearing UO Rods b 2 Zircaloy Spacer Capture Rod 1 Zircaloy Rods 3 Spacers per Bundle 3 Fuel Rod Cladding Material Zr-2 Wall Thickness, Inches 0.034 Puel Rods Outside Rod Diameter, Inches 0.449 Fuel Stacked Density, Percent Theoretical (Including Dishing) 91.6 Active Puel Length, Inches 70 Till Gas Helium > 95% we d

k i i t. 2 1 P. Change Section 5 2.1(b) to read as follows: 5.2.1(b) Reactor Operation l 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, a-1 4 J-2 Reload G G-1 Minimum Core Burnout Ratio at Overpower 1 5* 1 5*" 1.5** Transient Minimum Burnout Ratio in Event of { Loss of Recirculation Pumps From Rcted i Power 15 1.5 1.5 Maximum Heat Flux at Overpower, Btu /h-ft 500,000 395,000 h23,000 ) Maximum Steady-State Heat Flux, Btu /h-ft 410,000 324,000 3h7,000 Maximum Fuel Rod Power at Overpower, kW/ft 21.6 13.7 lk.6 Maximum Steady-State Fuel Rod Power, kW/ft 17.7 11.2 12.0 Stability Criterion: Maximum Measured Zero-to-Peak Flux Amplitude, Percent of Average Operating Flux 20 20 20 M1ximumSteady-StatePowerLevel,My 240 2h0 240 Maximum Value of Average Core Power Density @ 240 MW, kW/L h6 46 h6 Maximum Reactor Pressure During Power Operation, P *: 1,485 1,485 3,485 Minimum Recirculation Flow Rate, Lb/h (Except During Pump Trip Tests or Natural 6 6 6 Circulation Tests as Gutlined 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 Wt Per minute when power is less than 120 W, less than 20 Wt Per minute when power is between 120 and t 200M(,and10MW perminutewhenpowerisbetween200and240M(. C. Delete Figures 5.2 and 5.3 and renumber existing Figures 5.4 through 5.8 as 5.2 through 5.6, respectively. Add new Figure 5.7 attached.

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 !!uelear Corporation Synthesized Hench Levy.

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.g ) ( (~ 4 II. Discussion - Relosd G-1 Fuel -Summary Reload G-1 fuel assemblies are identical to the Reload G assemblies currently loaded in the Big Rock Point ~ core, except that: 1. Each assembly contains four inert Zircaloy rods rather than one, ~ resulting in-2. A decrease in the number of fuel rods from 116 to 113, and 3. An increase in the number of plutonium-containing rods from 24 to 25 The most significant result of this design change is a reduction in the calculated peak clad temperature during a postulated Loss of Coolant Accident (LOCA). The maximum calculated cladding tempera-ture during a design basis LOCA, utilizing the AEC Interim Acceptance ~ Criteria assumptions, has been reduced to 2036 F from the value of 2286 F calculated for the Reload G assemblies. The differences between the Reload C assemblies and the Reload G-1 assemblies and their safety significance is discussed in detail in the following. sections. A. Fuel Description 1. Mechanical Design Reload G-1 fuel is identical to the Reload G assemblies in fuel rod and gelldc diameters, rod-to-rod pitch, cladding vall th*:kness, active fuel le%- h per rod, tie plate and spacer design, and overall as-sembly envelo; Figure 1 presents the layout of the various rod types within the assembly. The Reload G-1 fuel is designed to comply with the same mechanical design requirements as the Reload G assemblies. 'ine principal mechanical features of the tv iesigns are compared in Table I which illustrates the near identi' of the two designs. 2. Nuclear Design The neutronic parameters calculated for the Reload G-1 assemblies.are in most respects the same as those calculated for the Reload G assemblies. The net neutronic effects of the increase from one to four inert 71rcaloy rods and the increase from 2h to 25 mixed oxide rods are:

y FIGURE 1 REL0AD G-1 ASSEMBLY R0D MATRIX "DOD@@3@@DDCW OGOGOGOGGGQ O@@@DD@@@@Q D@D@DDD@D@@ D@@@@OOODCO D@DDDODODOO OGOOOOOGOGG OGOGOGOOQQO OG8GGGQQGQO OGOGOGGQOGO +OOOGOQQQOOOo Symbol Rod Type fio. per Assembly 5.45*wt.% Pu, Pu0 -UO2 25

Fissile Pu 2

4.33% = 35 Fissile Pu + Fissile U 4.6 wt.% U-235, U02 .6 w.$ 2 5.00% = 2 32 3.2 wt.% U-235. 002

- Zir alo soli slug 16 cladding.

2.3 wt.% U-235, UO2 g/ t W re Zircaloy Rod ** 4 Tie Rod 12 Spacer Capture Rod I

~ EN (_ 6 TABLE I Mechanical Design Features Reload G Reload G-1 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 93 5 Dish Volume, % 2.0

2. 0 '

Stacked Density, % Theoretical 91.6 91.6 UO -Gd 0 Pellets 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 .3715 Fuel'R'ods Puel uength, Inches 70 TO Pellet-Clad Diametral Gap, Inches .0095 .0095 Plenum Length, Inches 39 3.9 Clad OD, Inches .hh9 .4h9 Clad ID, Inches 381 .381 Acaamblies No of UO Rods 88 8h 2 No of Pu0 -UO R ds 2h 25 2 2 No of UO -Gd 0 08 2 23 No of Cobalt' Target Rods k 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 Pu0 + UO Kg 1hh.8 141.1 2 2 Bundle Wt, Lb hh5 hh0 Spacers - No 3 3 Frame Material Zr h Zr h Spring Material Inconel *(18 Inconel 718 uw- -rr* re t

1 .)- -[ 7

1) An increase of about 4% in the assembly maximum local peaking factor.

2) A decrease in Assembly k= of 17% k resulting from an-

' increased worth of gadolinia-bearing rods. Neither of these effects is significant since the appropriate asse"aly reactivity and peaking parameters will be used in evaluating the-overall core compliance with reactivity limits and power density limits, 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 of the neutronic parameters of the Reload G-1 e.ssem-blies and the Reload G assemblie's is presented in Table II illustrating the minor differences.between 'tIhe designs. Figure 2 presents a compari-son of the local peaking factors of the Reload G-1 assembly to those for the Reload G assembly illustrating the small increame in peaking factor obtained and the. shift in power distribution toward the center of the asserDly. 3 Thermal and Hydraulic Design The hydraulic design of the Reload G-1 assemblies is identical to that of the Reload G assemblies. The only difference in the thermal performance of the assemblies evolves from the reduction in the nusber 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 from 3410 F to 3600 F; and, the minimum critical heat flux ratio decreases from 1.68 to 1.$9

However, the resulting peak pellet te mperature is still far below the fuel melting point of about 5100 F, and the minimum critical heat flux ratio is com-fortably above the value of 1.00.

Other thermal hydraulic factors considered in the Reload G assembly analysis, such as the effcets of utilizing mixed oxide fuel and the potential effects of Pu0 particles, are not altered by the change p - to the Reload G-1 design. A comparison of the thermal hydraulic parameters for complete 1 cores of Reload G assembly fuel and Reload G-1 fuel is presented in Table III. 4

~.. ~ i.( ( 8 TABLE II Nuclear Parameters Reload G Reload G-1 Core Data Rated Power, MW 240 Operating Pressure,6Psia 1350 Total Core Flow, 10 Lb/h 12 3 Leakage Flow, % 19 5 Core Inlet Subcooling, Btu /Lb 22.8 Core Average Void, % 25 Number of Control Rods 32 Equivalent Core Radius, Cm 97 2 Assembly Average Enrichment W/O Fissile 3 98 3 97 Assembly Average Enrichment, W/0 U-235 3.05 3.01 Assembly Average Enrichment, W/0 Fissile Pu 90 96 Number of Enrichments per Assembly h h Number of Fuel Rods 116 113 Number of Gadolinia-Bearing Fuel Rods 4 h Concentration of Gadolinia in Urania, W/0 1.2 1.2 Water to Fuel Volume Ratic 2.62 2.69 BOL - Maximum Local Power Peaking (25% Void Content) Ho Gadolinia 1.15 1.20 With Gadolinia 1.18 1.23 E0L - Maximum Local Power Peaking (25% Void Content - 28,000 M'Jd/MTM Bundle Exposure) '1.10 1.lh Assembly k= (BOL - 25% Void Content no Xe or Sm) No Gadolinia 1.252 1.252 With Gadolinia 1.205 1.188 Ak= Moderator Temperature (20*C to 306 C) +.0123 +.0128 Ak= Void Content (0% to 50%) .0h27 .0426 ok= Fuel Temperature-Doppler (306 C to 515 C) .0049. .00h8 Fuel Weights by Assembly, Kg: Weight UO 1 3.1 139.h 2 Weight U 126.2 122.8 Weight U-235 3.93 3.Th Weight Pu0 1.635 1.648 2 Weight Pu 1.h4 1.51 Weight Fissile Pu 1.lh6 1.192 . Weight Gd 0 .0h44 .045h 4 23 4 w l

1 i 9 l 1.066 .976 1.167 1.111 1.094 1.046 .954 1.143 1.092 1.077 1.153 1.004 ' 178* 1.093 1.066 1.117 .955 1.130 1.069 1.044 1.086 .920 .812 .463 .551 .847 .802 1.233* G, 1.146 .946 .864 + Reload G-1 + 1.102 1.002 .945 .799 .785 6 .889 ~~ Octar t Syme try .845 FIGURE 2 LOCAL POWER DISTRIBUTION COMPARIS0N (Reload G Assemblier - G1 Reload) 25% Void Content - 0 MWD /MT - with Gadolinium

Maximum Local Power Factor

~ /. 10 TABLE III Thermal Hydraulic Parameters (Core Contains All of Each Type Fuel) Reload G Reloa i G-1 Core Conditions _ .B Reference Design Thermal Output, (MW )/(Btu /h) 2h0/8.191 t 6 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 c.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,h8 77,48 2 Flow Area, ft /in2 0.163/23.44 0.163/23.hh Design Power Peaking Factors Fraction Generated in Fuel, % 96.6 96.6 Fuel Assembly Power Factor 1.h5 1.45 Local Peaking Factor 1.20 1.25 Axial Peaking Factor 1.51 1.51 Engineering Heat Flux Factor 1.0h 1.0h Assembly Thermal Performance Maximum Heating Rate, kW/ft, at 22% Overpower 13 59 12 Sh Maximum Heating Rate, kW/ft, at Rated Power 11.14 11.92 Average Heating Rate, kW/ft 4.08 h.19 Maximum Heat Flux, Btu /h-ft at 22% Overpower 39h,300 422,400 Maximum Heat Flux, Btu /h-ft at Rated Power 323,500 3h6,200 2 Average Heat Flux, Btu /h-ft at Rated Power 118,500 121,600 2 Temperature, F, at 22% overpower 3h10 3600 Maximum UO

Maximum Clad Temperature, F, at Overpower 731 Th2 MCHFR at Overpower Conditions Axial Peak at X/L =.45 1.68

'1 59 Coolant Subcooling at Core Inlet, Btu /Lb 22.8 22.8 Assembly Hydraulic Performance Average Assembly Flow Inner Orifice Zone, Lb/h 132,900 132,900 outer 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 1 '3,000 c Assembly AP.at Average Design Power (Includes Orifice AP) 5.37 Psi 5 37 Psi Hot Assembly Engineering Enthalpy Rise Factor 1.10 1.10

Crud-Free Surface

I t. (.. i 11 l l l B. 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 outer row 2 2 of UO r ds next to the corner cobalt target rod. This would also be the 2 worst case for the Reload G-1 fuel design, although the resulting maximum rod power and fuel temperature in the Reload G-1 case would be slightly lower, due to the slightly lower initial power in this outer rod row location. 2. Reactivity Insertion Accident and Primary System Integrity The kinetics parameters of the Reload G-1 fuel design do not differ from those of the Reload G design; from this standpoint the conse-quences of a reactivity insertion accident will be the same for both de-signs. However, since the neutron kinetics model used to perform these calculations has been updated, the consequences of a rod drop accident have been recalculated for both a UO -fueled core (all Type J-1 or F) 2 mixed oxide-fueled core (al? Reload G or Reload G-1). The results of this updated analysis are presented in Table IV for an assumed 2.1% k rod drop accident and are compared to the previously reported results for this as-sumed accident. Inspection of the tabulated data indicates that the up-dated model leads to lower calculated enthalpy depositions for UO and 2 mixed oxide fuel types. This reduction is attributed to a more exact modeling of the dropped rod reactivity insertion curve plus inclusion of an axial Doppler weighting factor in the calculation. A more extensive discussion of the analytical methods used and calculated values for the enthalpy deposition as a function of the assumed dropped rod worth are presented in Appendix A to this submittal. Since the current calculations lead to lower values of enthalpy deposition, the conclusions with respect to fuel dispersal and primary system integrity reported in our Proposed Change No 31, dated June 16, 1972, conservatively apply to the revised f calculations. 3 Loss of Coolant Accident Analysis The consequences of the design basis Loss of Coolant Accident (LOCA)_have been calculated for the Reload G-1 assemblies with the same i b

g 4...

k. k.-

) 12 TABLE IV Enthalpy Deposition - 2.1% Ak Rod Drop Accident Reload G Core J-1 Core-(Mixed Oxide) (UO2 Fuel) ' 4-Current Previous Current Previous Analysis ' Analysis Analysis Analysis Peak Enthalpy Deposition (Cal /g) 340 370 350 410 4 s 9 .-r-

l

i

( 13 models and assumptions uaed for the Reload G assembly LOCA analysin reported in our Proposed Change No 31, dated June 16, 1972. The peak cladding temperature for the Reload G-1 design is 2036 F. The maximum ~1ocal zirconium oxide thickness is less than 0.002 inch, corresponding to to a maximum of less than 6% Of the 0.034-inch thick cladding. The core average metal water reaction is less than 0.3%. A plot of the clad temperature versus time for this analysis is given in Figure 3 a. Models and Assumptions The models and assumptions used in this analysis, including the heat transfer coefficient-time relationship (including spray coeffi-cients), the decay heat curve, the core neatup code, radiation emissivities, metal-water reaction, and channel vetting time, are as reported in our Proposed Change No 31. As in all previous Big Rock Point LOCA submittals, a constant value of 1000 Btu /h-ft -F* was used for the gap coefficient for all fuel rods. Using gap coefficient values based on the GAPEXX model described in XN-73-25 results in a calculated peak clad temperature of 21h6 F. The clad temperature versus time for this calculation is shown in Figure k.* b. Zircaloy Rods and Target Rods Wetting Time The centrally located Zircaloy rods were calculated to wet at 126 seconds by the method detailed in Appendix A of the May 18, 1972 submittal. This method yields a vetting time of 60 seconds for the four cobalt target rods. For the calculation using GAPEXX gap coefficients, these wetting times were 166 and 66 seconds, respectively. In order to refine the capability of predicting quench front velocities down the canister and passive rods, Exxon Nuclear has sponsored the development of a two-dimensional (r, Z) wetting model for the condition following LOCA in BWR subsequent to the initiation of emergency spray cooling. The wetting model takes into account direct heat transfer between the outer surface of the canister (or shroud or channel) and the boiling liquid film as well as the heat transfer between the nonwetted region of the surface and the vetted region of the surface through the "The peak clad temperature based on the AEC approved Exxon NucJ ear fuel densification model (letter dated December 13, 1973 from D. J. Skovholt to R. Nilson) is estimated to be less than 2150 F.

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t_

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5 ' (] (,)\\ (] (.)'l /~' s \\ 16 e wall of the shroud or passive rod. The aodel is sensitive to the following parameters: Wall Thickness Spray Flow Rate Density, Thermal Conductivity, and Heat Capacity Temi crature of the Spray Coolant Temperature of the Wall Surface Liedenfrost Temperature The subject model dubbed 2DQ (which stands for two-dimensional quench) has been checked out against the following principal sources of data and found to be in good agreement with the data: Yamanouchi, "Effect of Core Spray Cooling in Transient State After Loss of Coolant Accident," J Nucl Sci Tech 5 (1968). R. B. Duffey and D. T. C. Porthouse, " Experiments on the Cooling of High Temperature Surfaces by Water Jets and Drops," Crest Specialist Meeting on Emergency Core Cooling for Light Water Reactors, Munich, October 1972, Paper II 2, summarizing the Results of Crest Meeting in Europe. e J. D. Duncan and J. E. Leonard, GEAP-13086, June 1970 - BWR Flecht Data. Application of the 2 DQ model to the Reload G and Reload G-1 design using the AEC interim criteria results in peak clad tempera-i tures no higher than those calculated using the one-dimensional rod l vetting correlation displayed to the AEC during the licensing of the Reload G assemblies and adopted from the Yamanouchi differential correla-l tion model given in XN-73-34, "M0XXY - A Generalized Multirod Heatup Code," redubbed HUXY, December 31, 1973 c. Peaking Factors The reactor is assumed to operate at rated power (240 MW ) t and the radial and axial peaking factors were taken as the design values j of 1.k5 and 1 51, respectively. The local rod power distribution cor-l responding to an exposure of 11,000 mwd /MIM yielded the highest peak clad temperature over the exposure range from beginning of life to 11,000 mwd /MrM which represents the period of high radial power factor operation. p This distribution is given in Figur 'i. { !!l fi

l l i l i 17 i 1.055 .965 1.107 1.061 1.045 1.083 .951 1.102 1.031 1.004 1.020 .889 .804 1.100 1.079 .978 .918 .903 .854 Octalt Symn try FIGURE 5 - LOCAL POWER DISTRIBUTION USED IN LOCA ANALYSIS BIG ROCK POINT RELOAD G-1 Exposure 11,000 mwd /MN P00R ORIGINAL

~ ~ k l 18 k. Off-Site Radiological Effects Consideration The discussions of the off-site radiological effects presented in proposed Change No 31, June 16, 1973, addressed the question in terms t"*1 i" of the generic effects of substituting mixed oxide fuel for UO2 a typical light water reactor, or in terms of substituting plutonium isotopes for the U-235 isotope in assessing the effects of fission product yield differences. The results presented apply to the Beload G-1 design as well as the Reload G design. III. Conclusions Based'on the foregoing, both the Big Rock Point Plant Review Committee and the Safety Audit and Review Board have concluded that this proposed change does not involve a significant hazards consideration. i CONSUMERS POWER COMPANY By 4/ E -f-R. A. Lamley, Vice Presideht Date: June 20, 1974 Sworn and subscribed to before me this 20th day of June ICJ74. a_[ (SEAL) .) rte Sylvila B. Ball, Notary Public Jackson County, Michigan My commission expires May 18, 1976. 1 1 m w

( u i APPENDIX A BIG ROCK POINT R0D DROP ANALYSIS A.1 Results and Conclusions This analysis compares the peak energy deposition in the fuel resulting in the Big Rock Point BWR for two cores, each at the beginning of an equili-brium cycle. The cores are fueled with the XN urania fuel design (J-1) and XN mixed oxide design ' Reload G). The detailed core parameters assumed are shown in Table A-I. Figures A-1 and A-2 present the calculated control rod scram curve and dropped rod worth function used in the mixed oxide analysis. Similar results were obtained for the urania core. The dropped rod analysis was performed using the kinetics code WIGL2. The results of hne WIGL2 analysis are shown in Figure A-3 and Tables A-II and A-III, and the analysis procedure is summarized in Section A-2. As can be seen, the peak deposited enthalpy is approximately the same for mixed oxide fuel (Reload G) and for the urania fuel (J-1). The mair, difference between these results and those of the previous analyus presented in Proposed Technical Specification Change No. 31 is a lower enthalpy deposition in both the urania and mixed oxide cores. The reduction amounts to approximately 25% for the urania core and 10% for the mixed oxide core. This reduction is due to:

1) The use of different dropped rod worth functions in the WIC'.2 analyses (a linear approximation was used in the previous analyses)

2) The use of different scram bank worth functions in the WlGL2 analyses

3) The use of Doppler weighting factors

4) The use of calculated axial enthalpy peaking factors dependent on dropped rod worths.

(, 4; 3 A2 TABLE A-I ASSUMED CORE CONDITIONS 1. Equilibrium fuel cycle 2. B0C core average exposure 8.43 GWD/MTM 3. Beginning of transient fuel temperature 582*F 4. Moderator temperature 582*F, 0.0 Void -6 5. Beginning of transient power level 240 x 10 MW 6. Axial Exposure Distribution Exposure Exposure Zone (GWD/MTM) Top of-Core 1 4.40 2 7.44 3 9.59 4 10.94 5 11.45 6 11.15 7 9.61 8 7.06 r 9 4.21 Bottom of Core 7. Uniform radial exposure distribution A

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L ) A6 TABLE A-II R0D DROP ACCIDENT MIXED OXIDE CORE KINETICS PARAf1ETERS AND RESULTS Core: Mixed 0xide, Exxon Reload G Delayed Neutron Fraction (i): .00529 -5 Doppler Coefficient ~ (582*F to 1200*F): .960 X 10 ak/k/*F Scram Function: WIGL2 Calculation Scram Bank Velocity: 2.456 ft/sec. Scram Delay Time: 0.375 sec. 2 Rod Drop Acceleration: 29.3 ft/sec Rod Drop Insertion Function: WIGL2 Calculation Rod Worth (mk) 10.1 19.8 29.6 Radial Enthalpy Average (CAL /gm) 47.03 94.8 141.90 A. Radial Enthalpy Peak (CAL /gm) 86.78 203.22 337.49 B. Axial Enthalpy Peaking Factor 1.185 1.31 8 1.363 C. Local Peaking Factor 1.18 1.18 1.18 Peak Fuel Enthalpy (CAL /gm)* 121.4 315.1 542.8 Peak Fuel Enthalpy = A x B x C k.

( (' i ) A7 TABLE A 'II R0D DROP ACCIDENT URANIUM' CORE KINETICS - PARAf1ETERS AND RESilLTS Core: Urania, Exxon J-1 Delayed Neutron Fraction (i): .00591 -5 Doppler Coefficient (582*F) to 1200*F): .916 x 10 ak/k/'F Scram function: WIGL2 Calculation Scram Bank Velocity: 2.456 ft/sec. Scram Delay Time: 0.375 sec 2 Rod' Drop Acceleration: 29.3 ft/sec Rod Drop Insertion Function: WIGL2 Calculation Rod Worth (mk) 12.4 22.8 29.8 Radial Enthalpy Average (CAL /gm) 56.11 110.48 138.24 A. Radial Enthalpy Peak (CAL /gm) 103.04 237.36 336.16 B. Axial Enthalpy Peaking Factor 1.211 1.3 36 1.367 C. Local Peaking Factor 1.24 1.24 1.24 Peak Fuel Enthalpy (CAL /gm)* 154.7 393.2 569.3 Peak Fuel Enthalpy = A x B x C. 4 s e -r

~ ( s A8 i A.2 Analysis Procedure The following procedure was used to perform the Big Rock Point rod drop analysis using WIGL2. All curves are those which were used in the Reload G calculations. a) The core is divided into three radial zones: 1. Drop red zone; four fuel assemblies 2. Partial control tone; variable size 3. Uncontrolled zone; variable size b) With zone 1 controlled, the size of the controlled zone 3 (and likewise the size of zone 2) is varied and a search made for the amount of control necessary in zone 2 to achieve keff = 1.00. This yields a locus of points along with keff = 1.00 for a range of sizes of zont 3. With zone 1 uncontrolled, another series of calculations is made by 9anging the size c ' zone 3 and searching for the amount of control required in zone 2 to yield keff = 1.00 + akeff, where Akeff is the control rod worth desired. See Figure A-4. At the points of intersection of the curves thus generated are the necessary size of zone 3 and the fraction of control in zone 2 which will yield a keff = 1.00 when the center centrol rod is inserted and keff = 1.00 + okCR with the center control rod removed. c) Static calculations in the axial direction are performed to generate a scram worth function curve by calculating the eigenvalue at various scram bank positions. See Figure A-1. d) The axial case is then used in the same static manner to generate dropped rod insertion worth functions. See Figure A-2. e) Static calculations are made in the radial cases to generate dropped rod insertion worth functions. See Figure A-5 f) Figures A-2 and A-5 are used simultaneously to generate Table A-V, which provides tabulai input into WIGL2. This tabular input provides the code with the fractional change in crass sections between fully controlled and uncontrolled required to yield the axial worth curve as a function of , fraction of control rod removed.

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i t, g e a s All TABLE A-V ~ DROPPED R0D WORTH FUNCTION 20 mk WIGL2 Fraction Fractional Withdrawal - Withdrawn Worth Function .00 .00 .00 .10 .04 .10 .20 .15 .31 .30 .34 .56 .40 .49 .70 .50 .62 .80 .60 .73 .86 .70 .82 .91 .80 .90 .95 .90 .96 .98 1.00 1.00 1.00 4 9 I e 4 =

~ ( ) 5 f A12 g) The acceleration of the dropped rod was determined by assuming an initial velocity of 0 ft/sec and an exit velocity of 18.5 ft/sec. h) The scram bank velocity and scram delay time was detemined by the tech spec requirements of 1) 10% inserted at 0.6 seconds and 2) 90% inserted at 2.5 seconds. See Figure A-6.

1) An axial WIGL2 transient calculation was performed for each rod worth. This yields a Doppler Weighting Factor (DWF) and an enthalpy peaM ag factor in the axial direction.

The DWF is defined as follows: 2 Vc j DWF = Y' i i i [ -/T V where j (/(= 8) = Vj and Tj = Fuel temperature in region i (*R) T

re Reference fuel temperature (*R)

(; = Flux in region i Vj = Volume of region i The DWF is then a volume and flux-squared weighted a d.- j) The DWF's are then descrit ed is functions of the normalized average enthalpy deposited in the axial case for input in tabular form into the radial WIGL2 case. k) The axial case functions, rod worth and DWF's, are then used in the radial kinetics solution to yield the deposited enthalpy.

1) The peak enthalpy deposition is calculated as:

peak radial x peak ' average axial enthalpy x local neaking factor. m) All cross, sections are derived from standard Exxon design codes and no adjustments were required to obtain the range of rod worths.

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AEC DIS 1 g":' ION FOR PART 50 DOCKET MATMTAL L (TEMPORARY FOPl!) CONIROL NO:_ 5627 FILE: ~ FROMg DATE OF DOC DATE REC'D LTR TWX lRPT OTHER Consumers Power Company Jcckr;n, Mich. 49201 6-21-74 X Mr. R.B. Sewell 6-20-74 T0Z ORIG CC OHER SENT AEC PDR XXX SENT LOCAL PDR ^^^ J.F. O' Leary 3 signed CLASS UUCLASS PROP INFO INPUT NO CYS REC'D DOCKET NO: XXX XXX 40 50-155 DESCRIPTION: ENCLOSURES: Ltr requesting a change to the Tech Specs of Proposed changes to tech specs.(Change #27) Lic. No. DPR-6.....trans the following.... ACKNOWLEDGED (40 cys enc 1 rec'd) PLANT NAME: Big Rock Point DO NOT REMOVE FOR ACTICN/I?7CRPATIt'N 6-24-74 JB BUTLER (L) SCHWENCER (L) / ZIEMANN (L) REGAN (E) W/ CYS W/ CYS W/7CYS W/ CYS CLARK (L) STOLZ (L) DICKER (E) W/ CYS W/ CYS W/ CYS W/ CYS Pi=.(L) " us.aLLe (L) m?IcMTcM (E) W/ CfS W/ CYS W/ CYS W/ CYS KNIEL (L) PURPLE (L) YOUNGBLOOD (E) W/ CYS W/ 'CYS W/ CYS W/ CYS INTERNAL DISTRIb" TION REG FILp TECH REVIEW DENTON LIC ASST A/T IND AE C FITR HENDRIE GRIMLS

IGGS (L)

BRAIT!iAN D /0GC SCHROEDER GAMMILL GEARIN (L) SALTZMAN AiUNTZING/ STAFF MACCARY KASTNER GOULBOURNE (L) B. HURT CASE KNIGHT BALLARD KREUTZER (E) GIA?!3USSO PAWLICKI SPANGLER LEE (L) PLANS BOYD SHA0 MAIGRET (L) MCDONALD MOORE (L)(LWR-2) STELLO ENVIRO REED (E) CHAPMAN DEYOUNG (L)(LWR-1) HOUSTON MULLER SERVICE (L) /DUBE w/ input SKOVHOLT (L) NOVAK DICKER SHEPPARD (L) ./E. COUPE / GOLLER(L)dte) ROSS KNIGHTON SLATER (E) / Schemel P. COLLINS IPPOLITO YOUNGBLOOD SMITH (L) D. THOMPSON (2) DENISE TEDESCO REGAN TEETS (L) KLECKER vilEG OPR LONG PROJECT MGR WILLIAMS (E) EISENHUr FILE & REGION (3) LAINAS WILSON Q j MORRIS BENAROYA HARLESS J nD nDI'Ikl P I i STEELE VOLLMER IJ Il II n l 1513i iI EXTERNAL DISTRIBUTION ~ ' ' """""""M 'l - LOCAL PDR Charlevoix, Mich. 7

1 - TIC (ABERNATHY)

(1)(2)(10)-NATIONAL LABS 1-PDR-SAN /LA/NY

- NSIC (BUCHANAN) 1-ASLBP(E/W Bldg;, Rm f 29) 1-BROOKHAVEN NAT L\\E 1

1 - ASLB 1-W. PENNINGTON, Rm E-201 GT 1-G. ULRIKSON, ORNL 1 - P. R. DAVIS 1-B&M SWINEBROAD, Rm E-201 GT 1-AGMED (RUTH GUSSMAN) Vlo - ACRS 'nTnTT Sent to Diggs 1-CONSULTANTS Rm B-127 GT 6-24-74 NEWMARK/BLUME/AGEABIAN 1-RD..MUELLER, Em F-3M GT}}

ML20030A460

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