ML19345E436

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Forwards Request for Proposed Change 14 to Tech Specs of License DPR-6 to Insert New Fuel Design,Designated Reload E Fuel,Into Reactor
ML19345E436
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 02/06/1968
From: Allen R, Haueter R
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To: Morris P, Skovholt D
US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101190275
Download: ML19345E436 (28)
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Dear Dr. Morris:

Attention: Mr. D. J. Skovholt Transmitted herewith are three (3) executed and thirty-seven (37) coriforced copies of a request for a change to the Technical Specifications of License DPR-6, Docket I!o. 50-155, issued to Consumers Power Company on May 1, 1964, for the Big Rock Point :;uclear Plant. The proposed change (I!o. Ih) vill enable Censumers rower Company to insert into the reactor at Big Rock Point a new fuel design, designated Reload "E" fuel, which incorporates the General Electric Company's current commercial design features. This fuel design uses three enrichments and other features which will improve core peaking factors and take for more efficient uti-lisation of fuel in the Big Rock Point reactor. It is our intent to insert Reload "E" fuel into the bic Rock Point reacter during early June 1968. Because ve vill be out of fuel at that time, it is important that we receive license approval to receive the fuel by early May 1968. Yours very truly, ) GJU/d:b Robert L. Haueter Attach. Assistant Electric Production Superintendent - Iluelear ~. 1, i, 3 f.y W / T. d a M. w. - =

j $0ZL 4.s l:{ ..0 ~: 9 CONSUMERS POWER CO!FANY vi. /. x y, g, y _. hr[7 - Docket No. 50-155 LSS ' ' 4 adnd w/ur Dated Request for Change to the Technical Specifications License No. DPR-6 For the reasons hereinafter set forth, it is requested that the Technical Specifications of License DPR-6 issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Nuclear Plant be changed as follows: I. Section 5 A. In Section 5 1.5, change "(c )" to read as follows : "(c) Fuel Bundles The general dimensions and configuration of the six types of fuel bundles shall be as shown in Figures 5.2, 5.3, 5.4, 5.5, 5.6, 5 7 and 8.1 of these spec-ifications. Principal design features shall be es-sentially as follows:" B. In Section 5.1.5, add Figure 5 7 C. In Section 5.1.5, replace the present table of fuel bundle parameters with the following: 7'g.Ilf@ 4 9G O'c;*ipi-9 e e i e E g 2-v.)Q c s 7,.t g 4l 3 t

"5.1.5 (Contd) Fuel Bundles Fesearch and Development Centermelt Cente rmelt Original (A) Reload B & C Reload E "D" Puel Intermediate Aivanced Genersl Geometry, Fuel Rod Array 12 x 12 11 x 11 9x9 11 x 11 3x6 7x7 Rod Pitch, Inches 0.533 0.577 0.707 0.580 0.807 0.921 Standard Puel Rods per Bundle 132 109 74 109 36 29 Special Fuel Rods per Bundle 12* 12** 7*** 12 28# 20# Spacers per Bundle 3 5 3 7 5 5 Fuel Rod cladding Material 30h SS Zr-2 "r-2 30h SS, Zr-2 Zr-E Zr-2 Inconel 600 and/or Incoloy 800 Standard Rod Tube Wall, Inches 0.019 0.03h 0.0h0 0.010 to 0.030, Inclusive 0.035 0.040 Special Rod Tube Wall, Inches 0.031 0.031 0.0h0 0.010 to 0.030, Inclusive 0.035 0.0h0 Fu-l Rods Standard Rod Diamater, Inches 0.388 0.bh9 0.5625

0. l> 25 0.570 0.700 Special Rod Diameter, Inches 0.350 0.3hh

'0.5625 0.320 0.570 0.700 U0 Density, Percent Theoretical 9h 1 9h 1 Pellet 90 to 95, Inclusive 9h Pellet 94 Pellet 2 85 Powdered 87 Powdered 85 Powder 85 Powder Active P2el Length, Inches Stand 6rd Rod 70 70 69.5 68 to 70, Inclusive 66-674g 65-66JE Special Rod 59 (Corner) 6h.6 (Central) Fill Gas Helium Helium Helium Helium Helium Helium CFour Special Fuel Rods at Bundle Corners'Are Segmented s C? Reload B, C & F Fuel Bundles May Contain (in the Corner Regions of the Bundle) Four Zircaloy-2 Tubes Having Encapsulated Cobalt Targets Sealed Within C;5 Reload F Fuel Bundles Have a Epecial Central Fuel Rod To Which the Bundle Spacers Are Fixed

  1. In Additien, Two of the Interior Bundle Fuel Rods Are Removable Special Rods Have Depleted Uranium" ro

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s-a ( 3 17. Change.Section 5.2.1 (b) to read as follows: ~"(b) Reactor Operathn . The reactor operation shall be so' limited as to-be consistent - with the most conservative of the following: ~ Original. ("A"),"B" Reload and "C" Fuel ("E" Fuel

  • Minimum Core. Burnout Ratio at Overpower

'15 15 Transient Minimum Burnout Ratio-in Event _1. 5 ' l.5 of Loss of Recirculation Pumps From Rated Power Maximum Heat Flux at Overpower, Btu /Hr-Ft 530,000 500,000 Maximim Steady State Heat Flux, Btu /Er-Ft h34,000 h10,000-Maximum. Fuel Rod Power at Overpower, Kw/Ft 17 2-21.6 } Maximum Steady State' Fuel Rod-Power, Kw/Ft 14.2 17 7 -Stability Criterion: ' Maximum Measured 20 20 Zero-to-Peak Flux Amplitude, Percent of Average Operating Flux Maximum Steady State Power-Level, Mwt 2h0 2h0 Maximum Value of Average Core Power. LL6 h6 Density @ 2h0 Mwt, Kw/L-Maximum Reactor Pressure During. Power lh85 1h85 Operation, Psig Minimum Recirculation Flow Rate, Lb/Hr 6 x 10' 6 x 10 (Except During Pump Trip Tests or 4 -Natural Circulation Tests as Outlined in Section 8) Maximum Mwd /T of Contained Uranium for 23,500 23,500 an Individual Bundle l Rate of Change of. Reactor Power During Power Operation: Control rod withdrawal during power operation shall be such that t the average, rate of change of reactor power is less than 50 Mwt i per minute when power is less than 120 Mwt, less than 20 Mwt per minute when. power-is between 120 and 200 Mwt, and 10 Mwt per minute when power is between 200 and 2h0 Mwt. l

  • Based on correlation given in ' Design Basis for Critical Heat Flur Condition.

- in Boiling Water Reactors,' by J. M. Healzer, J. E. Hench, E. Janssen, and (~ S. Levy, SeptemberT1966~(APED 5286 and APED 5286, Part 2)." l i ...n, s

i 1 h II. Discussion - Reload "E"' Fuel-The' proposed changes-in Section 5.1 will enable Consumers Power Company to refuel,the Big Rock. Point-reactor vith a fuel design which very closely reflects current General Electric Company's. commercial r (nuclear fuel design concepts. - As in Reload "C". fuel, vibratory compacted-UO p wder vill be used. It is expected that all of the - fuel' designs - 2 in use in the. Big Rock Point reactor.will eventually be superseded by-the Reload "E" design. A. Fuel Description The Reload "E" fuel design is similar to the original ("A") fuel design in that machined castings are used for the upper and~ lower tie plates. The fuel bundle'is held together by eight tie. rods which are fuel rods with special threaded.end plugs and which mate with the~ upper and t ' lower tie plates. Rod clearances are maintained by spring clip spacers f which are held in place by a central spacer capture rod. Six rodt, one in each corner. and two in.the bundle interior, are provided with a-re-movable rod feature-to accommodate the cobalt _ irradiation program and pocsible future irradiation programs. The above design details are shown r in Figure 5 7. Additional fuel data are summarized on Table I. This basic bundle design has proven to be quite adequate for power reactor service over a number of years both at Big Rock Point and at other nuclear power plants. The re=ovable rod design has been in use at the Dresden Nuclear Plant for a number of years with no problems i

evident, i

B. Puel Thermal Data The. thermal conductivity data and the heat transfer coef-ficient between fuel and clad for Reload "E" powder fuel are the same as previously described for Reload "C" (see Change No. 10) and for the ten . pilot bundles included in the Type III-f reload fuel for Cycle h in the Dresden Nuclear Power Station. The previous thermal data were obtained from irradiations performed-under the High Performance UO Program spon-2 l sored by the Joint US - Euratom Research and Development' Program. The j UO2 Powder conductivity data, as submitted for the Dresden reload pilot i - bundles, yield an integral KdT from the fuel surface temperature (500 C j' by definition) to the melting temperature (2800 C) of L9 v/cm. (This was f I l, I

=, 5' reported as preliminary. data in the Dresden submittal.)~ In the su= mary ~ report o'f the High Performance Program'(GEAP-5100-1, "UO P}ovderand. 2 l Pellet Thermal Conductivity During Irradiation," by M. F. Lyons, et al, . Mr.rch 1966), the value of the integral as used for the Dresden. powder bundles was confirmed. Based on post-irradiation analysis, Lyons, et al, recommended an' integral value from surface to melt (as-defined above) 2 ~ l h9 w/cm'. The establishment of an' appropriate thermal conductivity i curve for powder UO is a deduction from this..The approach used to 2 establish the.vorking curve for.the Dresden and Big Rock Point powder 4 fuel is based on the following line of reason *ng. During the first few minutes,of operation, the-powder conductivity vill be uncertain und j probably change rapidly due-to sintering of powder above the UO sinter- ~ 2 ing temperature (s1600 C) and densification. However, as;these pro-cesses continue, the conductivity of the fuel above the temperature for the onset of grain growth vill become very close to that determined earlier-for pellets. The. remaining differences in conductivity and in f the conductivity integral values from surface to melting between powder and pellets are attributable to the peorer conductivity of the unsintered powder rim operating below the grain growth te=peratures. Working back-vards from the previously established pellet UO e nductivity curve, the i 2 powder t anductivity was assumed to be identical, at the melting point temperature, to that for pellets. Below this temperature, the powder curve was. assumed to gr-dually and smoothly fall below that for pellets. The deviation between the two curves was adjusted so that, upon reaching a reasonable value for the surface temperature (4500 C), the difference in aren under the two curves equals the difference in the integral values from surface to meltin.n. Below this temperature, the curve was simply extended smoothly to provide a reasonable mean through the out-of-pile data. (See Change No. 10.) Tables II and III show a comparison of the thermal per-formance characteristics and the thermal-hydraulic data, respectively, for the BRP reload fuel. It should be noted that at the 122 percent 7 overpower condition, from middle-of-cycle to end-of-cycle, the maxi-mum heat flux is expected to vary between 446,000 and L65,000 Btu /Hr-Ft. [

~ J j - 2 6 I N6,000 Btu /Hr-Ft corresponds to a linear heat generation rate of 19.2 Ew/Ft which is the point for onset of incipient center me5. ting for Beload 2 e "E" fuel. 'At h65,000 Btu /Hr-Ft, the linear. heat generatio'n. rate woald I be.20.1 Kw/Ft and approximately 0.003h volume fraction of the fuel rod would be molten. At a maximum fuel rod heat flux of.500,000 Btu /Er-Ft - the proposed license limit which' corresponds to a linear heat generation . rate of 21.6 Kw/Ft,~approximately 0.02h volume fraction of the fuel rod would be molten. As noted in the discussion of Change Request No. 13, the phase change volume-expansion of UO fuel is about 10 percent upon 2 melting..The Powder UO to be used in the Reload "E" fuel rods will have p a theoretical density-or 87: percent which will provide the necessary void to accommodate the expansion of the very small molten volume fraction } at the proposed license limit.of 500,000 Btu /Hr-Ft In addition,-the fuel rod damage limit based on plastic strain of the cladding has been ' defined as 'a molten volume fraction of about 0 50. Both of these factors assure that if the proposed license limit maximum heat flux at 122 per-cent overpower is ever attained for an extended period, the integrity of the fuel cladding will not be compromised. The proposed steady state rated power license limit of 410,000 Btu /Hr-Ft is comfortably below the onset of incipient center melting based on operating experience at Big Rock Point. In all of the. transients to date on the reactor, even though neutron flux traces.may - have indicated an overpower condition, nalysis showed that actual reactor power was always within a few percent of rated. This phenomenon is 4 ) attributable to the long thermal time constant of UO fuel which forces 2 i {_ the thermal events to lag well behind the nuclear events. The reactor safety system scrats on high neutron flux well i before the fuel reaches the thermal overpower condition. C. Fuel Physics Data The principal nuclear characteristics of Reload "E" fuel y have been calculated and are compared to Reload "C" fuel on Table IV. 1-4 - All of the differences between the two fuel types are small and do not 1 significantly~ change the nuclear operating characteristics of the reactor f core. Additionally, the control worth of Reload "E"' fuel has been found j to be equal to the Reload "C" fuel, which, in turn, has been found to be $i' -i f .,p' a 9 <. ~ .-r,.e, -,+,e-u y,.,.-. ,,y e -.[. y&, m ,,e

4-e. 4. J s l 7 . lower than Reload "B" fuel. Control rod worths are-then si,milar for given~ core. loadings of either "B," "C" or "E". fuel types'or combinations -thereof. D. Thermal-Hydraulic Data The principal thermal-hyd: 'ic characteristics of Re- ' load "E" fuel have been calculated and are compared to Reload "C" fue1~ on Table III. It vill be noted that the Minimum Critical Heat Flux- . Ratio (MCHFR) for the Reload "E" fuel is lover than that for the Reload "C" fuel,-but nevertheless, it is still comfortably above the licensedL limit of 1 5 at the 122 percent overpower condition. The-comparison on Table III is for a whole core of each type of fuel and as such does not, strictly speaking, represent actual cole configurations in the near future. Because of the difficulty of predicting core configurations:in the presence of the R&D fuel in Big Rock Point, specific core analyses must be p med during the refueling outages after fuel inspection l -and just p ' start-up. These analyses assure that all licensed limits are me.

  • he selected core configuration.

E. Operational Safety Aspects Change No. 10 included a discussion on the operational safety aspects of powder fuel. The information presented therein on the cause of cladding failures and propagation of. failures is still cur-rent and applicable to Reload "E" fuel. In the interim since filing for Change No. 10, additional experience has been gained in the irradiation of zirealey clad powder fuel. 1. Big Rock Point Sum =ary of Powder Puel Behavior, HPD Project i A complete program of fuel manufacturing techniques has been implemented in the Developmental Puel Assemblies Program, Task I .of the HPD Development Project at Big Rock Point. Members of each class of powder fuel bundle design have shown either verified or suspected fail-ures. The bundle types and exposures reached are shown in Table V. The D18-D20 assemblies, which are the most similar in design to Reload "E" fuel of the development assemblies, include verified and suspected fail-urcs but the Reload "E" fuel has been designed with important changes. i .~.

, Similarities between the designs; include the use of rocked Zircaloy-2T clad- ~ and vibratory compacted Dynapak UO2 p wder. Differences include inclusion

offa fission gas l plenum and deletion of the tight fitting zircaloy dice at the. top of the fuel column.

These modifications were made following observation of the - failure modes of the development assemblies. The Incoloy-800 clad bundles, 4 D10-D15,' demonstrated intergranular penetrations which appear to be similar to the stress assisted intergranular failure of 30h-SS observed in the VBWR-HPD program.' 'The Zircaloy-2 clad has been adopted for subsequent. . Big Rock Point fuel designs. The'zircaloy clad powder assembly failures (D18-D20) have been verified by visually locating one fai]ca fuel rod. This rod was circumferentially fractured in' a jagged manner near the bot-tom end -plug and was ~ partially cracked near the top end plug. Destructive examination of this rod is currently under way. This fuel design included a tight fitting zircaloy dise_at the top end of the fuel column. It'is theorized that the disc may have perturbed the clad by local stresses or local overheating resulting in failure near the upper end plug, and water was then able to enter the clad penetration and flow down. the rod. Such a_ path through the fuel rod was demonstrated experimentally bypassing argon through the rod. Water in the fuel rod could then be responsible for the clad hydriding observed. Further examination'of these fuel rods is continuing, but the zircaloy disc has been deleted from subsequent designs. As indicated on Table V, 33 Zirealoy-2 clad. powder assem-blies have been irradiated to levels above 5000 Mwd /T. These bundles have performed well to date with no indications of problems. 2. Dresden No. 1 Powder Fuel Experience Ten powder fuel assemblies were included with the Type III-f fuel loaded for Cycle 4. At the end of Cycle k, two bundles were identified as suspect failures from sipping signals. Fuel bundle examination re-vealed one failed rod in one bundle. No damage was found in the other assembly. The obcerved damage could be characterized by a local bulge in the clad with excessive hydriding at the bulge. Detailed examination of this individual fuel rod is now under way at the GE Radioactive Ma-terials Laboratory. 4 . = _.,,

a e 9 i 3 Powder Fuel Experience at Other Sites- _ The most extensive testing and application A,f compacted powder fuel'has been associated with the.PRTR program at Battelle.Horth-west Laboratories. Since start-up of the PRTR in.1961, a tota 1 Lor 361 elements i-rave been1 tested in the PRTR. Included-in the 361 elements have been J 75 aluminum plutonium spike enrichment elements,'68.UO elements and 2 216 UO -Pu0 elements. Both svaged and vibrationally compacted oxide fuels 2 2 T have been emphasized with most' attention being given recently te vibra- - tional compaction. More than TO variations in fabrication-techniques, fuel compositions and cladding have been tested in the PRTR. There are 4 21h' powder UO -Pu0 bundles with exposures ranging to 11,000 Mwd /T in 2 2 August 1967. The lead bundle of 68 UO p vder assemblies had then at-2 tained an exposure of 9,400 Mwd /T. The irradiation performance of UO2 packed-particle elements fabricated by swaging and vibrational ecmpaction has been excellent. In general, the irradiation performance of the me- ' chanically mixed Pu0 -UO bundles in the PRTR has also been excellent. Of - 2 2 nearly 5300 fuel rods irradiated in the PRTR prior to the start of HPD operation in January 1967, only 38 failed in service and these, but for one exception, were attributed to impurities in the fuel. No failures j have been attributed to inherent problems with the design of fuel rods utilizing powder fuel. F. Hazards Considerations The Reload "E" bundles described above will utilize hard-7 vare designed with current GE commercial techniques. Experience with this hardware has been excellent. The fuel rod spacers are of the spring type design now offered by General Electric for commercial fuel. The I - springs provide the necessary lateral pressure to minimize fretting wear of the zircaloy cladding. The three enrichments utilized in these bundles provide i a significant improvement in core peaking factors at a slight enrichment - penalty. The effect upon other nuclear and thermal-hydraulic factors is not significant. 1. Loss-of-Coolant Accident The ' loss-of-coolant accident for both a core load of Re-load "C" and Reload "E" has been evaluated. The calculation techniques ,m-s v w, ~e

m - g. ~ 1g a. -have been recently. detailed in Change Request No. 13.and the additional information submitted'with it. The results of the study hr}e shown on Fi ures 1 through 5 b a. Top Break Spectrum S D 1. Breaks Inside Enclosure - The analysis of possible- ' breaks in this'regionJudicates that the existing Big Rock Point (BRP) reactor could sustain, without melting any fue1~ cladding, a rupture as' ~ -large as that equivalent to the complete, instantaneous severance of a 17-inch downcomer, the largest credible break for this case. This is illustrated by Figure 1. Curves were drawn for the entire spectrum to-show the time at which the midplane of the reactor core would become uncovered, the time at which adequate core spray flow would be possible, and the time at which melting would commence-if. cooling were not pro-vided. f The assumed break area relates to_the complete sever-ance of two equivalent pipe sizes, dependent upon whether tU postulated rupture is single-or double-ended; i.e., if primary fluid reaches the-break via both parts of the severed pipe. For example, the complete . severance of the shutdown suction line would be a single-ended break while that-of a lb-inch riser would be double-ended. A. core spray flow of 250 gpm, which would occur at.a j reactor pressure of 85 psia, was assumed to provide' adequate cooling. This assumption is based on current design practice and distribution measurements which were performed with a mock-up of the BRP configura-tion. Figure 1 also shows the effect of rated and zero feed-water addition during the accident. While the times to uncover the core i_

midplane, i.e., in;tiation of heat up, are not significantly affected because.the blowdown is so fast, feed-water addition causes the reactor 2

pressure to decay faster and makes the core spray available sooner. The case-for zero as well as rated feed-water addition after the break was analyzed. For this region, however, loss of feedwater has only an in-- significant effect on those breaks for which the core'becomes uncoverea. m 1 1

( i t j \\ 11 Heat up calculations were performed for the fuel cur-rently in the BRP reactor which has an 11 x 11 array and also for the Type "E" reload fuel (9 x 9 array) which will be added later. The 9 x 9 fuel is shown on Figure 1 to begin melting slightly earlier than the current fuel if no cooling is provided. This is primarily because the Type "E" fuel has a greater diameter; consequently, there is more stored heat and power generation per fuel rod. Of further interest are the maximum cladding temperatures and percent of perforated rods with core spray cooling. Figure 3 shows these parameters for both types of fuel, and the 9 x 9 fuel is found to reach greater temperatures than the 11 x 11 fuel. However, for break areas less than 0.8 ft, fewer per-forated rods occur in the Type "E" fuel because of the number and lo'- cations of peaked rods. From a safety viewpoint, however, these dif-ferences cannot be considered to be significant. 11. Steam Line Breake Outside Enclosure - The analysis of possible breaks in this region indicates the existing system could sustain, without melting any fuel cladding, a rupture as large as that equivalent to the complete, instantaneous severance of the 12-inch steam line, the largest possible break for this case. The same break area versus time curves as discussed for breaks inside the enclosure are presented in Figure 2. The closure of the steam line isolation valve, assumed to occur 90 seconds after the break, is found to retard the pressure decay such that full operation of the core spray is slightly delayed. However, the time to uncover the core is unaffected by the valve closure since this would always occur before the valve was assumed to be actuated. The maximu= temperatures and percent of perforated rods are also shown on Figure 3 b. Bottom Break Spectrum The analysis of postulated breaks in this regicn indi-cates that the existing system could sustain, without melting any fuel cladding, a rupture as large as that equivalent to the complete, in-stantaneous severance of a 20-inch recirculation line, the largest credible break for this case. However, in order to sustain all breaks, feed-water addition is necessary. As seen on Figure 4, fuel clad melt-ing begins before the pressure is low enough to allow adequate core spray

l 12-flow for small breaks since the time-to-melt curve crosses over the core ~ spray curve. ;This.can~also be seen on Figure 5 where the cladding tem-peratures reach melting for both types of fuel for breaks between 0.02 2 and 0.03 PL. This'would' correspond to the complete severance of pipes -slightly greater than'two inches in diameter or, of course, partial breaks of-larger pipes'. A report on the Teig Rock Point emergency core cooling sys- - 1 tem'is currently being prepared for submission tc the AEC. -It:will be. addressed to the areas suggested by the lor: 4c-coolant accident analysis. 2. Reactivity Insertion Accidents -The Big Rock Point. reactor-operates with one specified rod withdrawal pattern. The rods are grouped in banks of two or more; all the rods in a bank are. withdrawn together, with a procedural limit of. one notch between any two-rods in a bank. This sequencing prevents large rod worths;'however, an' operator error or series of errors can-result in larger vorths. The possible rod drop situations and rod strengths when the core is critical and at hot standby are: Case 1: In-sequence potential of.008 Ak for drop from full-in position to drive position. Case 2: In-sequence potential of.021 ok for drop from-i full-in to full-out. Case 3: Out-of-sequence potential of less than.021 Ak for' drop from full-in to full-out. { Case h: Maximum theoretical worst case of about.0h5 Ak. Case 1 requires the following equipment malfunctions and operator error: a. Cod becomes uncoupled from drive. b. Drive is withdrawn (in-sequence), but blade hangs up temporarily. Operator does not notice that blade is j not following. Rod then unexpectedly releases and drops from full-in to c. position of the drive due to gravity. Case 2 requires an additional operttor error of with-drawing the drive completely rather than concurrent with the bank. Case 3 consequences are less than those for Case 2.

13 -Case k is considered hypothetical, as it r.equires still further compounding of unrelated errors beyond those enumerated above. The' analyses are performed for the hot standby (HSB) con- ~ dition; i.e., power at neutron' source level and normal eater level in the vessel. The hot standby rather than cold case is analyzed because: a. The rod strengths in the cold condition for the same potential rod drop situations are no greater than in the hot standby ~ condition. -b. According to our calculations, primary system integrity is more vulnerable when there is a free surface because of the possible water-hammer and subsequent vessel movement. This situation exists so long as the reactor is critical only when the vessel.is at a temperature safely above the nil ductility-temperature. Maintenance of this safety margin allows a strain to rupture of at least 13 percent. Big Rock Point operating procedures provide this margin. The zero power initial condition accidents are more severe than full power accidents because: a. The comparable rod worths are less at full power; e.g., the worth for Case 2 is.0095 vs.021 in the RSB situation. b. Peak enthalpy for equal values of reactivity insertion does not vary greatly, c. For equal peak enthalpy, the likelihood of fuel failure is greater in the HSB situation. d. The voided core in the full power condition is less susceptible to large pressure p61ses due to the moderator complJsnce, i.e., compressibility. Following the February 1968 refueling outage, the core I will contain six centermelt fuel bundles. Analysis is performed for a core of "E" fuel with the centermelt bundles included and with them withdrawn. To prevent a large amount of centermelt fuel from being in the peak neutron flux during a reactivity accident, the six centermelt bundles are to be loaded in the core in a dispersed array with a mini-mum center-to-center distance of h2 cm. This restriction means that the closest centermelt bundle spacing vill be no closer than two bundles in the x-direction and one in the y-direction. i ... ~. - - -

~ i Ik A

a. LKinetics Calculaticns

~ ) ~ LThe most important parameters in a nuclear excursion kin-etics calculation are: i. Quantity. of' Reactivity Insertion ii. Bate of Reactivity Insertion iii'.. Specific Power Distribution -iv. Doppler Coefficient .v. 1 Resonance Neutron Flux Distribution vi. Initial Power-The only significant difference getween the "C" core and "E" core is in the specific power distribution. The "E" fuel bundle lo-cal power. factor is about 15 percent less which would. reduce the peak j energy density as well as yield less fuel mass above any stated energy-level. ' Table VI indicates the relative fuel energy-density values re-sulting from a. reactivity excursion. This table shows that the energy-densities of fuel in the'"E" core are less than or equal to those in the "C" core. I' b. Primary System Integrity j As discussed at length in previous license applications (see Change Request No. 13) for this plant, the integrity of the primary l system depends upon the severity of any steam explosion. The severity of a steam explosion' depends-upon the following factors: f 1. Time of Fuel Failure 11. Mechanism of Fuel Failure f iii. Amount of Fuel Failed t. iv. Energy in the Failed Fuel l i v. Heat Transfer Rate to Coolant l' vi. System Geometry As has been shown in previous applications, a severe steam explosion vill result only if there is a significant quantity of promptly dispersed fuel in the moderator. For material to be promptly dispersed, [- it.must attain an energy density of 425 cal /g or more. Table VI shows I there is little, if any, promptly dispersed material in all the postu-lated accidents. It -is also seen that the "E" core is identical to the "C"' core in this 7espect. ~, - - - _. - - -

L.. -{ 15 i -III. Conclusions Based upon the above analyses and comparison ~s 'of a core of."C" fuel with a core of "E" fuel, with and without centermelt fuel, the following conclusions may be dravn: 'l. The mechanical design of the Reload "E" fuel is a -well proven concept and no problems should be expected, based on the good experience with the design to date.

2. 'The 9 x 9 lattice consisting of 77 fuel rods and 1

h' cobalt bearing rods, utilizes 3 zoned fuel enrichments which yield more uniform bundle power and fuel te=perature distributions and hence improve both steady ctate and transient performance of the. Reload "E"~ fuel. 3. The proposed license limit of a maximum fuel rod ] heat flux of -500,000 Btu /Hr-Ft at 122 percent overpower yields a 4 molten fuel rod volume fraction of 0.024 which can be accommodated by Reload "E" design withcut compromising the integrity of the cladding. h. The proposed license limit of a maximum steady state-heat flux of h10,000 Btu /Hr-Ft, in the light of BRP operating expe-rience, is unlikely to be exceeded even under transient conditions. 5 The principal nuclear operating characteristics of the Reload "E" fuel are basically the same as previously licensed BRP fuel designs which are well within the bounds dictated by good design practice. 6. The thermal-hydraulic calculations show that there is ample critical heat flux margin throughout a representative cycle. 7 Analyses show that the loss-of-coolant accident is not changed significantly from that presented in the FHSR and updated in Change Request No. 8. 8. A rod drop accident would not be expected to cause a breach of the primary system with either a core of "E" or "C" fuel. In fact, the reactivity insertion accident is calculated to be less severe in the."E" core.than in the "C" core. l

t. 16 1 Based ~on-the above consid'erations, we have-concluded that the use of the Reload "E" fuel in the Big Rock' Point reactor does not prese'nt a significant change in the hazards considerations' described or implicit in the_ Final Hazards Summary Report. CONSUMERS POWER COMPANY By \\ Vice President Date: February 6,1968 Sworn and subscribed to before me this 6th day of February 19C8. A_; GAn 3 A) Notary Public, Jackson County, Michigan My Con: mission Expires January 15, 1972 i- ) 1 w '{. 4. i ,i F 't +"*

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.. ~.- L- ~ ). s . l .L i TABLE I i . RELOAD "E" FUEL' DATA I n Fuel Rods Cobalt Rods-i j Fuel Diameter (Clad Inside -Diameter)- 0.4825L . Rod Pitch, Inches 0 707 .O.707 - C15dding Thickness, Inches 0.040 0.0h0' Clad Outside Diameter,' Inches 0 5625 .0 5625 . Active Fuel Length, Inches 69 53;-Central. Rod, 6h.62 t 4 i Fuel Material UO ~ 2 l Fuel Density, % of 'I'heoretical 87% ' i Cladding Material-Zr-2 Zr Number of Rods per Bundle 77 4 Enrichment (See Figure 5 7) Low - 2 35% 1 Middle - 2 93% 1 High - 3 55% Fill Ges Helium i l Fuel Bundle i Fuel Rod Array 9x9 { Weight UO per Bundlei Pounds 337 3 2 1 f Moderator-to-Fuel Volume Ratio 2 33 ? Number of Spacers 3 a T i j. 1 ,.m -x. -- e e

.?.. .4 ~' ' FABLE II . THERMAL PERFORMANCE CHARACTERISTICS COMPARISON OF REIDAD FUEL s

Reload "E" Reload "C" Fuel Diameter, Inches 0.4825 0 381

~ Cladding Thickness, Inches 'O.0h0 0.034 j Cladding Outside Diameter, Inches 0 5625 0.449 Fuel Density, % Theoretical Density 87 85 - Nucleate' Boiling Heat Transfer 2 Coefficient, Btu /Hr-Ft.op 10,000 10,000 Heat Transfer Coefficient Between 2 Fuel and ' Cladding, Btu /Hr-Ft - F 3,000 3,000 ' UO Conductivity Integral-2 T = 2800 C 1 q Kdt,W/cm 49 49 T'= 500 C Incipient Melting Temperature of j-Uo, F 5,080 5,080' 2 Heat Flux for Incipient Melting, Btu /Hr-Ft2 4k6,000 550,000 ~ FuelLineerHeatGeneration,Kw/Ft (For Incipient Melting) 19 2 19 i ~ .r., m -.y..

TABLE III THERMAL-HYDRAULIC DATA COMPARISON OF REIOAD FUEL Beginning of Cycle End of Cycle (Many Control Rods (No Control Rods Inserted) Middle of Cycle Inserted)~ "E" "C" "E" "C" "E" "C" Fuel Lattice 9x9 nxn 9x9 nxn '9x9 nxn Mwt (1.22 x Rated Power) 286 7 286.7 286.7 286.7 286 7 286.7 6 Flow, Lb/Hr 12 3 x lo 12 3 x lo 12 3 x lo 12 3 x lo 12 3 x lo 12 3 x 10 Assumed Ieakage Flov 20% 23 6% 20% 23.6% 20% 23.6% Teakage Power 3% 4% 3% 4% 3% 4% MaximumHeatFlux, Btu /Hr-Ft h22,ooo 353,000 446,000 374,000 465,000 392,000 Hot Channel Flow, Lb/Hr 120,000 117,000 123,600 n8,Too no,300 ni,400 MCHFR 1.80 2.16' 1 94 2 32 1 71 2.n. Quality at MCHFR 14.0% 14.4% 12.6% 13 3% 18.5% 18 3% Exit Quality 18 7% 19 2% 16.7% 17.6% 24.4% 24 3% ~ Radial

  • Power Factor 1.4 1.4 1 32 1 32 1.623 1.653 Axial Peaking Factor 1.45 1.45 1.625 1.625 1 385 1 385 Local Peaking Factor 1.21' 1.27 1.21' 1.27 1.21 1.27,

overpowei Factor 1.22 1.22 1.22 1.22 1.22 1.22'

[ k l. i TABLE IV PRINCIPAL CALCULATED NUCLEAR CHARACTERISTICS OF RELOAD FUEL, WITH COBALT Reactivity (koo) -Reload "E"- Reload "C" Temperatre; i '68 F, Zr Channels 1.262 1.244 572 F, Zr Channels, 0% Void, UO at 12000 F 1.280 1.272 2 572 F, Zr Channels, voided, UO at 1200 F 1.262 (25% 1.256(20% 2 Void) Void) 0 Moderator Temperature Coefficient (Akeer/kere per F in Zr Channel at 77 F) Reload "E" Reload "C" Start of Cycle +3 8 x 10 +2.6 x 10' ~ End of Cycle -1.8 x 10'5 +4.9 x 10' Void Coefficient ( Akerr/kere per Unit Void Within the Channel) 4 Reload "E" Reload "C" Cold (68 F) -0.07 -0.06 Hot (572 F) -0.11 -0.08 Doppler Coefficient (Akere/kert per F) Fuel Temperature Moderator Reload "E" Reload "C" -5 -5 l 68 F 68 F, O Voids -1 3 x 10 -1.2 x 10 -5 -5 1323 F 572 F, O Voids -1.0 x 10 -1.0 x 10 ~5 ~5 1323 F 572 F, Voided -1.2 x 10 -1.1 x 10 (25% Void) (20% Void) d

9 TABLE V GE POWER REACTOR DWONSTRATION (AS OF DECmBER 31, 1967) Clad Peak Avg Burnup Approx Clad Thick-UO2 Q/A 'of Leading Peak No. of Clad OD ness Density (Rated Bundle Burnup Reactor Bundles Material (In.) (In.) (%TD) Power) Mvd/T Mwd /T Remarks Dresden 10 Zr-2 0 5625 0.035 84.7 330,000 10,600 15,900 One bundle failed at 9,000 Mvd/T. One bundle failed at - 7,600 Mwd /T. , Type III-f. . Consumers 4 304L-SS 0.425 0.010 91 384,000 7,680 n,300 one suspect. One in core. BRP Bundles P01-P04. 6 Incoloy-0.425 0.on 91 h30,000 10,550 15,500 Four failed. Two in pool. 800 Bundles D10-D15 3 Zr-2 0.425 0.030 85 3h0,000 7,290 10,700 One failed.- Twb suspect bundles. D18-20. 40 Zr-2 0.449 0.034 85 h30,000 8,900 13,h00 Thirty-three ir. core. No suspected failure. Bundles C01-Ch0. )

--.~.. 1 l-f ~ TABLE VI REACI'IVITY INSERTION ACCIDENT COMPARISON OF ~ . RELOAD FUEL 6 C" Core "E" Core

  • W/Cm
    • W/O Cm
  • W/Cm
    • W/O Cm Reactivity Insertion, ok

.015.021.015.021 .015.021.015.021 Peak Enthalpy, cal /g 315 h50 225 320 315 450 200 280 f Mass of Fuel, Kg, Above: 4 425 e - O 1 O O O 1 O O i 280 cal /g. 3 35 0 '4 3 35 0 0 220 cal /g 100 100 <1 -110 100 100 0 100 a i

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l' E I ,,J 7 s-4 s. 4 A ~. 1 I FEJMs DATE OF DOCUMENT: DATE RfCEIVED NO.s kk8 g .communers Pasar Co. g g ) huon, Eid tra., ufuO. areOar. Otxta. Robert L Eauster g Os OtsG.e CC: OTHits Morris 1 35 cys reproduced ACTION NECESSARY O c ":Ut"c' O "^^"5*'' NO ACTION NECESSARY O c ""5"' O FLA151Fa eO51 OFF6C4 FILE CODES S REG. NO: Mhh DEscalPIlOhs War be Unclassif.ed) REFitRED TO DATI RECElYED BY DATE Ltr trans the followings g,,g r,,. Y ' "ecy Fl 2-1244 i" "" * ** '" Change in Tbch Spec's ENCtOsuats.( 3 33 D 0 i PROPOSED CHAT!GE # IL, notarised 2M ,, m,_,,,, Cil.1 enable CPL to insert into reactor g 1 !L, :,. En., O new fuel design, designated Reload I i . f,- cE' fuel, which incprporates 0.E. ~ ~ " ~ ' ' ' current ecomerical design,.. mrag .k..l.O ~ ' Ur' nu,h -w i DISTg 1. Formal .... v g. '- 1-suppl file Emmene 3.1: 1-ALC PIR 448 2-c=cli=ce 1-ooc men ram mim i U. S. ATOMIC ENERGY COMMIS$lON Mll CONTROL FORM FOna AEC-2245 e u. e. eovenus sur ensurine orricsi int-are4es e i Y PEismane s,>}}

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