ML19269D198

Refueling Plan for Cycle 6

ML19269D198
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/28/1979
From:	Buck P, Raffety S DAIRYLAND POWER COOPERATIVE
To:
Shared Package
ML19269D197	List: ML19269D196 ML19269D198
References
LAC-TR-067, LAC-TR-67, NUDOCS 7903070254
Download: ML19269D198 (24)

Text

{{#Wiki_filter:LAC-TR-067 REFUELING PLAN FOR CYCLE 6 OF LACBUR S. J. Raffety Dairyland Power Cooperative and P. Buck Nuclear Energy Services, Inc. February, 1979 Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601 79030704Sf

LAC-TR-067 TABLE OF CONTENTS Page 1.0

SUMMARY

1 2.0 CYCLE 5 ANALYSIS..................................... 7 3.0 LACBWR CYCLE 6 RELOAD PLAN........................... 3 3.1 Location of Fuel Assemblies for Cycle 6......... 3 3.1.1 Shroud Cans.............................. 3 3.1.2 Limitations on Rate of Power Increases... 3 3.2 Control Rod Program............................. 4 3.2.1 Bias for Calculation of Criticality with TRILUX.............................. 4 3.3 L e ng th o f Cy c l e 6............................... 5 3.4 Reactor Physics Calculations for Verification of Technical Specifications..................... 5 3.4.1 Cold Shutdown Margin..................... 5 3.4.2 Ejected Control Rod Worths............... 5 3.4.3 Linear Heat Generation Rate and Average Planar Linear Heat Generation Rate....... 5 3.4.4 Minimum Critical Power Ratio............. 6 3.4.5 Misplaced Assembly....................... 6 3.4.6 Uncontrolled Rod Withdrawal.............. 9 4.0 APPENDIX............................................. 11 Analytical Methods.........................$... 11 4.1 4.2 Design Concepts................................. 11 4.3 Operating History of Current Cycle.............. 11 4.4 Nuclear Design................................ . 11 4.4.1 Moderator Coefficients................... 11 4.4.2 Doppler Coefficient...................... 11 4.4.3 Maximum Radial and Axial (or Total) Peaking Factors.......................... 11 4.4.4 Rod Drop Analysis........................ 12 4.4.5 Changes in Nuclear Design................ 12 ii

LAC-TR-067 TABLES Table No. Page I LACBWR Fuel Assemblies to be Discharged at End of Cycle 5.............. 13 II Calculated Shutdown Margin With One Rod Out............................... 14 III Worths of Ejected Control Rods............ 15 iii

LAC-TR-067 FIGURES Figure No. Page 1 Proposed LACBWR Reload Configuration for Cycle 6............................... 16 2 Control Rod Patterns and Associated Nodal Power Peaks and Exposures Ex-pccted During Cycle 6..................... 17 3 Lt.;BWR Core Configuration and Expected Pcel Assembly Exposure at End-of-Cycle 5.. 19 4 Rod Withdrawal Transient, DOC-6........... 19 Relative Core Power vs. Control Rod Position 5 Rod Withdrawal Transient, BOC-6........... 20 MCPR vs. Control Rod Position IV

LAC-TR-007 1.0

SUMMARY

LACBWR fuel cycle 5 is expected to end about March 15, 1979 when the exposure of the lead assembly reaches 15,000 MWD /MTU. The average core exposure at EOC-5 will be approximately 9930 MWD /MTU and the length of Cycle 5 will be approximately 4200 MWD /MTU. The Cycle 5 core consists of 40 Type II assemblies and 32 Type III assemblies. Twelve stainless steel shrouds are used for reactivity control. It is expected that at EOC-5 the 28 highest exposed Type II assemblies will be discharged and replaced with 26 fresh Type III assemblies and 2 fresh Type I assemblies. The proposed reload pattern for Cycle 6 is shown in Figure 1. The core average exposure at BOC-6 is estimated to be 4325 MWD /MTU, with 58 Type III, 2 Type I, and 12 Type II fuel assemblics. Eight stainless steel shrouds will be used for reactivity control. Calculations with the TRILUX fuel manage-ment code show that with the reload configuration of Figure 1, the length of Cycle 6 will be about 4275 MWD /MTU. At this time, although additional reactivity will still be available for more burnup, the lead assembly will have reached 15,000 MWD /MTU, The control rod program used in the TRILUX study is shown in Figure 2. This program produces acceptable power distributions throughout Cycle 6 with Technical Specification requirements on Minimum Critical Power Ratio (MCPR) and maximum average planar linear heat generation rate (MAPLHGR) being conservatively met. The shutdown margin with a stuck rod is at least 1.5% ak/k for the worst rod compared with the Technical Specification require-ment of at least 0.5% Ak/h. The ejected rod worth for the worst rod is less than 2.5% t.k/k as required. No control rod bank switches are made during Cycle 6. LAC-TR-067 2.0 CYCLE 5 ANALYSIS Cycle 5 of LACBWR will end about March 15, 1979 with an average exposure of approximately 9930 MWD /MTU at which time the ex-posure of fuel assembly 2-48 will be 15,000 MWD /MTU. Approxi-mately 2000 MWD /MTU additional burnup would be available before the cycle would end due to reactivity limitation. The expected exposure distribution at EOC-5 is shown i., Figure 3. Cycle 5 has operated so far well within allowed limits on off-gas activity. No indications of broken fuel pins such as occurred in Cycle 4 have been noted. At the end of Cycle 5, an examination of all fuel assemblies will be performed as is customary at the end of each cycle. The 28 highest exposed Type II assemblies will then be discharged with estimated ex-posures as shown in Table 1. The average discharge exposure is expected to Le 14,183 MWD /MTU. Minor adjustments may be required to this discharge scheme if it is found that some of the lower exposure Type II assemblies have defective pins. In that event, some additional assenblies may be discharged and replaced with additional Type III fuel. o LAC-TR-067 3.0 LACBUR CYCLE 6 RELOAD PLAN 3.1 Location of Fuel Assemblies for Cycle 6 After the 28 highest exposed assemblies from Cycle 5 are dis-charged, the remaining 12 Type II assemblies and the 32 exposed Type III assemblies will be relocated as shown in Figure 1. Twenty-six fresh Type III assemblics and'two fresh Type I assenblies will be located as shown in Figure 1. The exposed fuel is intermixed with the fresh fuel to control radial power peaking and to limit power density increases com-pared to those experienced during Cycle 5. Highly exposed assemblies with burnups of approximately 12,000 MUD /MTU are located on the outer adge of the core to prevent them from being exposed to high power at advanced burnup and also to del.ay their reaching the burnup limit of 15,000 MWD /MTU as long as possible. In addition, in these locations, the high exposure assemblies are remote from control rod movement at power. 3.1.1 Shroud Cans The Cycle 6 core will use eight stainless steel shroud cans for reactivity control. These are necessary because with the 15,000 MWD /MTU burnup limit at EOC-5, the BOC-6 exposure is rather low and the excess reactivity high. These stainless steel shrouds are located as shown in Figure 1. 3.1.2 Limitations on Rate of Power Increases During Cycle 5, the limiting conditions on rate of power in-crease were: 1) During the initial power escalation and after all cold shutdowns, increase power to an initial level of 40% at an average rate of approximately 1% per hour, and from 40% to 603 at a rate not to exceed 10% of rated power per day. From 601 to 1001, increase power at a rate not to exceed 5% of rated power per day. These restrictions are because of pc.'.let-clad interference and allow the pellet time to icform and relieve the stress on the stainless steel clad. Other power escalations above SOS of full power were limited to a rate of approximately 1% of rated power per hour. 2) Additionally, during any power escalation above 25% of rated power, limit movement of any rod to 4 inches per hour (s 1" per 15 minutes). LAC-TR-067 3) During power operation after the initial escalation, limit the withdrawal rate of any rod to one inch per day whenever the rod is withdrawn farther than it has been previously withdrawn during the current fuel cycle. The restrictions on power escalation rate and on control rod movement which were used in Cycle 5 will be used in Cycle 6. 3.2 Control Rod Progran A complete burnup history for the Cycle 6 reload has been per-formed using the TRILUX code. Several BOC studies were per-formed before an initial rod pattern was deternined which allows satisfactory peaking factors. The BOC pattern selected has rods 8 and 12 withdra.n 60", rods 7 and 11 at 35", and rods 1, 9, and 13 at 16". Burnup steps of 600 PUD /MTU were used for the study and new control rod patterns were determined for each step. The basic rod pattern withdraws the B bank rods (6, 10 and 8,

12) and the A bank rods (7, 11 and 9,

13) in pairs with a 25-inch separation between pairs.

When rods 9 and 13 are 25" belou rod 1, rod 1 is then withdrawn, maintaining a separation of 25" also. This rod program has been found to give good axial power distribution throughout the cycle. Figure 2 shows the patterns used at each 600 FWD /MTU step along with the ex-posure interval and the nodal peak. Rod locations shown in Figure 2 are expressed as nodes inserted, 10.0 representing fully inserted, 0.0 reyresenting fully withdrawn. 3.2.1 Bias for calculation of criticality with TRILUX_ Records from past cycles show that TRILUX calculated keff for the hot, full power critical condition at the beginning of Cycle 2 to be 1.021 instead of 1.000. For beginning of Cycle 3, TRILUX calculated 1.009. The Cycle 2 reactor core had no stain-less steel shrouds and the Cycle 3 reactor core had 32 stainless steel shrouds. Since the Cycle 4 reactor core had 16 stainless steel shrouds, an interpolation was performed that indicated that TRILUX would calculate a value of 1.015 for the'beginning This actually turned out to be the value for of Cycle 4 keff. the BOC-4 keff. On this basis the TRILUX model was predicted to calculate a BOC-5 k gf of 1.017 for the hot, full power, equi-e librium xenon condition since 12 stainless steel shrouds were present in the core. The actual value calculated for Cycle 5 was very clcse to this, 1.015, with the slight difference being attributed to a very small discrepancy in calculating k rg for e Type III fuel. Based on the variations in predicted keff as a function of the number of steel shrouds and of the number of Type III assemblies, the prediction for calculated kegg at the beginning of Cyclc 6 with 8 steel chrouds and 58 Type III assemblies is 1.0166. _ 4 _

LAC-TR-067 3.3 Length of Cycle 6 The detailed studies of the burnup of Cycle 6 with the TRILUX code show that the lead exposure assembly, number 2-67, will reach a burnup of 15,000 MWD /MTU at a core average exposure of 8600 MWD /MTU and that the length of Cycle 6 will be approxinctely 4275 MWD /MTU. 3.4 Reactor Physics Calculations for verification of Technical Specifications Calculations have been made with TRILUX and XENOLUX to verify the cold shutdown margin, the ejected rod worth, the linear heat generation rate and the minimum critical power ratio. The maxi-num average planar linear heat generation rate (MAPLHGR) was also calculated and the consequences of a misplaced assembly and an uncontrolled rod withdrawal frca power are discussed. 3.4.1 Cold shutdown Margin The cold shutdown margin with 28 rods inserted and one rod stuck out must be greater than 0.5% according to the Technical Specifi-cations. TRILUX calculations were performed for each rod with-drawn from the core (excluding symmetric duplicates) with the Cycle 6 reload shown in Figure 1. The results are given in Table II along with the results for five previous startups. For each withdrawn rod, the shutdown margin is greater than 0.5 ok/k as required by Technicel Specifications. 3.4.2 Ejected Control Rod Uorths The Technical Specification linit on the worth of any ejected rod is 2.5% Ak/h. The XENOLUX code was used to calculate ejected rod worths because this code allows the void and xenon distributions in the core to be held conservatively constant during the ejection. All calcule'_ ions were peformed at full power-full flow conditions at begin.ng-of-cycle since previous calcuintions have shown this condition yields the highect ejected rod worths. Calculated rod worths are shown in Table III along with initial control rod patterns and xenon distributions. All ejected rods were calcul-ated to be worth less than 2.5% ak/k. 3.4.3 Linear Heat Generation Rate and Average Planar Linear Heat Generation Ratr, The Technical Specification limit on peak linear heat generation rate during steady-state operation is 11.52 kw/ft for the Type III (Exxon) fuel and 11.94 kw/ft for Type I* and Type II (Allis-Chalmers) fuel. The overall peak power will be maintained below

• See proposed Technical Specifications for Type I fuel.

- 5 _

LAC-TR-067 the above values throughout Cycle 6, as determined by TRILUX with estimated maximum values of 9.1 kw/ft, 5.5 kw/ft and 10.7 kw/ft respectively for Type I, Type II and Type III fuel at steady-state full power conditions. The Technical Specification limits on average planar linear heat generation rates are 8.11 kw/ft for Type III fuel and 7.78 kw/ft for Type I* and Type II fuel for average planar exposures up to 24.0 GWD/MTU. TRILUX calculations indicate the maxinum expected APLHGR values during Cycle 6 steady-state operation will be 6.57 kw/ft, 2.61 kw/ft and 5.43 kw/ft respectively for Type III, Type II and Type I fuel assemblies. Average planar exposures will not exceed 24.0 GND/MTU. Therefore, the Tech-nical Specification will be satisfied. 3.4.4 Minimum Critical Power Ratio The proposed Technical Specification for Mininum Critical rower Ratio (MCPR) using the XN-2 correlation from Exxon is 1.73 for Type I fuel for normal full power operation and 1.59 for Type II and Type III fuel (unchanged from Cycle 5). These limits were established from the results of the rod withdrawal transient which, as in Cycle 5, has been concluded to be the most limiting transient at full power operatinn (See Section 3. 4. 6 ). At core flows less than rated recirculation flow, the nost limit-ing transient is the recirculation two-pump speed up caused by a failure of the motor-speed control system. The MCPR Limits for steady state operations at less than rated recirculation flows are conservatively maintained at values established for Cycle 5, as Cycle 5 fuel maximum peaking factors were higher than those proposed for Cycle 6. 3.4.5 Misplaced Assembly TRILUX calculations were performed to ascertain the ef fect of an assembly loading error. When assembly 3-35 (core location C-3) was interchanged with assembly 3-15 (core location G,7) so that a cluster of 6 fresh Type III fuel assemblies was formed near the G-7 location (see Figure 1) and full power-full flow equi-librium beginning-of-cycle conditions were assum d, the follow-ing results were calculated.

• See proposed Technical Specifications for Type I fuel.

LAC-TR-067 Misplaced Assembly Reference Case Values Results Without Loading Error (kw/f t 0 Location) (kw/ft 0 Location) CPR, Minimum Type I 1.92 0 F,9 1.93 0 E,9 Type II 2.03 0 B,9 2.03 0 B,9 Type III 1.69 0 H,8 1.73 0 H,8 APLHGR, Maximum Type I 5.5 0 F,9 5.4 0 F,9 Type II 2.7 0 K,9 2.6 @ B,9 Type III 6.8 0 G,8 6.6 0 G,8 LHGR, Maximum Type I 9.2 @ F,9 9.1 0 F,9 Type II 5.6 @ K,9 5.5 0 B,9 Type III 11.0 0 H,8 10.7 0 H,8 When the fresh Type I assembly in core location F,9 was inter-changed with the fresh Type III assembly in core location H,8, the following results were calculated for equilibrium BOC-6 full power-full flow conditions. LAC-TR-067 Misplaced Assembly Reference Case Values Results Without Loading Error (kw/f t 0 Location) (kw/f t 9 Location) CPR, Minimum Type I 1.78 @ H,8 1.93 0 E,9 Type II 2.03 0 B,9 2.03 0 B,9 Type III 1.74 0 C,3 1,73 0 H,8 APLHGR, Maxinum Type I 6.5 0 H,8 5.4 0 P,9 Type II 2.6 0 K,9 2.6 0 B,9 Type III 6.6 0 H,7 6.6 0 G,8 LHGR, Maximum Type I 10.5 0 H,8 9.1 0 F,9 Type II 5.5 0 K,9 5.5 0 B,9 Type III 10.6 0 C,3 10.7 0 H,8 As can be seen from the data in these tables, the limits on CPR, APLHGR, and LHGR are conservatively met during Cycle 6 even if a fuel assembly is misplaced during the core loading. The effect of a misplaced assembly on the shutdosa margin in a cold, no xenon, zero power, BOC-6 condition was also investigated. When one of the fresh Type I assemblies was placed in core location B,9 and control rod number 23 was withdrawn, a kerf of 1.00116 was calculated. When a fresh Type III assembly was mis-placed in core location B,9, a shutdown margin of 0.586% ak/k was calculated and with the lowest exposure previously irradiated fuel assembly misplaced in core location B,9 the shutdown margin was found to be 1.139% ok/k. These results indicate that if one of the two fresh Type I assemblies was misplaced in,one of the four corner positions, the result could be significant. No other loading error causes the shutdown margin to be reduced below the Technical Specification requirement. The potential for mis-placing a frech assembly in a corner position is negligible. All the previously irradiated assemblies are placed in their desig-nated core positions before any fresh fuel is loaded and the core loading is checked to assure that only the locations designated for new fuel arc still empty. The fresh fuel is easily dis-tinguished from previously irradiated fuel at all states of hand-ling and af ter installation in the core because it is brigh* and shiny whereas the irradiated fuel is a dull brownish color. It g_

LAC-TR-067 is also possible to readily distinguish Type I fuel from Type III fuel by the difference in design of the upper fittings. After the refueling is completed, the core loading will again be doubJe checked to assure that the fresh fuel is located in the proper positions and that the two fresh Type I assemblies are located properly. Therefore the results of the nisplaced assembly calculation with respect to shutdown margin are deemed acceptable. 3.4.6 Uncontrolled Rod Withdrawal The effects of an uncontrolled rod withdrawal were studied with the TRILUX code using the following initial conditions: Full Power, 17.5 Mwth No xenon No samarium in fresh fuel Average core exposure at EOC, 4325 MWD /MTU 18.50 Btu /lb Ah = g Recirculation flow = 10.8 x 106 lb/hr Control Rod Pattern: Pod Nodes Inserted 1 8.27 6 3.18 7 8.27 8 6.18 9 8.27 10 3.18 11 8.27 12 6.18 13 8.27 The consequences of withdrawal of various control rods from the BOC-6 control rod pattern were determined with the result that the withdrawal of Control Rod 11 leads to the worst transient. Figure 4 shows the increase in core power as a function of rod 11 withdrrwal. Figure 5 shows the change in the MCPR for Type I, Type II, and Type III fuel respectively. As can be seen from these figures, the MCPR for all fuel types remains satisfactorily above the minimum allowed value of 1.32 throughout the transient. This transient study indicates that the minimum steady-state MCPR's at rated thermal power and flow conditions which assure that MCPR will not go below 1.32 during a control rod withdrawal transient are 1.73, 1.49 and 1.50 for Type I, Type II and Type III fuel respectively. Accordingly, as discussed in Section 3.4.4, _9 -

LAC-TR-067 the proposed Technical t'pecification limit for MCPR for Type I fuel at normal full power operation has been established at 1.73. For Types II and III, the limiting MCPR value of 1.59 from Cycle 5 has been retained for conservatism and continuity. LAC-TR-067 4.0 APPENDIX 4.1 Analytical Methods No changes in analytical methods have been made for Cycle 6. TRILUX is used for power distribution and burnup studies while XENOLUX is used for rod ejection studies. 4.2 Design Concepts No new fuel types will be used in Cycle 6. Fuel types will be Type I (Allis-Chalmers), Type II (Allis-Chalmers) and Type III (Exxon). 4.3 Operating History of Current Cycle No operating anomalies have yet (February 1979) been noted in Cycle 5. 4.4 Nuclear Design 4.4.1 Moderator Coefficients There will be basically no significant change in moderator co-efficients from Cycle 5 because similar fuel will be used in Cycle 6. Although the moderator ccefficient for Type III fuel is different from that for Type I or II, the difference is very slight and TRILUX calculations show satisfactory operation with all Type I, all Type II or all Type III fuel. Therefore, Cycle 6 is bounded by these calculations. 4.4.2 Doppler Coefficient The Doppler Coefficient is about 8% less for Type III fuel than for Type II fuel. Since Cycle 6 will have 58 instead of 32 Type III fuel assemblies, the Doppler Coefficient for Cycle 6 will be about 3% less than for Cycle 5. Transient studies for all Type III fuel have shown satisfactory operation and adher-ence to Technical Specifications; therefore Cycle ( operation is bounded. 4.4.3 Maximum Radial and Axial (or Total) Peaking Factors The maximum radial peaking factor will be about 1.36; the maximun nodal peaking factor will be about 1.98. Refer to Figure 2 for results of the TRILUX burnup study. LAC-TR-067 4.4.4 Rod Drop Analysis No significant change in delayed neutron fraction will occur because of the close similarity in enrichment between Type I, Type II and Type III fuel. The scram function used in rod drop analyses for LACBUR was very conservatively chosen (fast linear ramp). Therefore Cycle 6 behavior is bounded. The control rod patterns used in Cycle 6 are shown in Figure 2. No rod interchanges are planned; the rod patterns are very similar to those of Cycle 5. 4.4.5 Changes in Nuclear Design None. LAC-TR-067 TABLE I LACBWR Fuel Assemblies to be Discharced at End of Cycle 5 Fuel Assembly No. Core Position Exposure, MWD /MTU 2-48 K-8 14,999 2-25 B-3 14,947 2-23 F-1 14,872 2-41 B-8 14,821 2-2 L-14,809 2-32 K-3 14,796 2-1 F-10 14,784 2-40 H-9 14,725 2-13 A-5 14,704 2-33 C-2 14,645 2-29 E-1 14,634 2-24 L-5 14,577 2-36 H-2 14,556 2-9 A-6 14,535 2-49 C-9 14,490 2-47 E-10 14,434 2-70 F-8 13,956 2-51 E-3 13,919 2-69 E-8 13,644 2-52 F-3 13,608 2-62 D-4 13,551 2-59 G-7 13,547 2-68 H-6 13,411 2-63 G-4 13,362 2-53 C-5 13,358 2-58 D-7 13,348 2-65 C-6 13,056 2-56 H-5 13,047 LAC-TR-067 TABLE II CALCULATED S!!UTDOWN MARGIN WITII ONE ROD OUT SIIUTDOWN MARGIN, % t2k/k Control Rod At BOC At BOC At DOC At BOC At BOC Out* At BOL 2 3 4 5 6 No. 1 7.0 7.3 5.0 6.0 8.5 7.0 s No, 4 5.8 5.4 5.1 5.9 5.5 6.7 No. 11 4.8 5.3 4.0 5.9 4.8 6.4 No. 12 3.5 4.0 4.6 3.1 3.3 2.6 No. 23 0.53 1.2 3.4 3.1 4.1 1.5 No. 22 2.2 3.7 1.4 2.0 2.9 3.4 No. 28 1.5 4.0 5.7 2.4 4.6 4.5

• Control rod as listed or syrametric duplicate.

LAC-TR-067 TABLE III WORTHS OF EJECTED CONTROL RODS Rod Rod Worth l Rod Configuration

• Xenon Ejected

% Ak/k 1 and A at 16" No 11 1.90 6 and 10 at 58.3" 3 and 12 at 33.3" 1 and A at 16" t:o 12 2.21 6 and 10 at 58.3" 8 and 12 at 33.3" 1 and A at 16" No 1 1.44 6 and 10 at 58.3" 8 and 12 at 33.3" 1, 9 and 13 at 16" Yes 9 1.66 7 and 11 at 34.8" 8 and 12 at 59.8"

• A Bank - Rods 7, 9,

11, 13. Rod position in dial inches withdrawn. LAC-TR-067 A B C D E F G H K L 2.-37 240 f /2.05 //.721 I d6"~ 1 1 47 V 3f 3-2 3-33 3-3Y 3-3 3-1 2-40 II.74 l r.17 Lov 0 o Cao Y.83 11.6 9 IA It: l( li ~~ 3-1 3-3.Y 30b 3-4 3-7 3 37 3-31 3-Y b 2.bb 0 0 5 75 5'W D o 2.3? ___ G 7 @F8 3-13 3-39 3-11 3 - Yo 3 -41 3 -11 '3-V2 3-/& 4 V.7 Y o Y.7y go 4.97 0 gja @yo l '3 1 ,/P I P 'i i P 2,,s 3 d. 3-7 3-vy 3 - ll 3-10 3-Mr 3-12 3-yh 2 4y 5 12, o s-a ga r.w ny, sc x13 o i =.8, L_ _ PC9 r f2 / 9 @f 2-F 2-67 3-v7 1-21 3 -97 3-23 3-21 3-v9 3-2y 3,50 2-6G b /2.90 0

5. s v L o.

3 47 I vy Lo r G6 o /2. 9 7j @L.S 4 to N 3-17 36/ 3-1y 3a 3, 73 3-/r 3-59 3-20 y %30 o y.77 d d 1.?Y 0 Y.33 12 ll (?) ID 3-Z1 3-55 1 44, 3-26 3-27 3,T7 3-37 3-32 8 z.SG 0 0

5. 'if S.77 0

0 2 66 @ _ _ ?/ i M 7/ 9 4 1 41 3-1r 3-30 l-3 l-2 y 3-JJ 3-17 2-bY 9

11. b3 4.s Y C ol 0

0 LcY C. I 7 ll.79 pg 1 /0 2-7/ 2 72 /I.97 /2.c7 IN CORE FLUX MONITDRS Q PLAhiT NORTH rUEL ASSEMBLY uUMnER xxx AVERAGE EXPOSURE (GWD/MTU) ,YYY DENOTES STAINLESS STEEL SIIROUD CAN---*^ FIGURE 1 - PROPOSED LACBWR RELOAD CONFIGURATION FOR CYCLE 6. Tile BOC CORE AVERAGE EXPOSURE IS ESTIMATED TO BE 4325 MWD /MTU (TRIAL RELOAD F) LAC-TR-067 FIGURE 2 - CONTROL ROD PATTERNS

• AND ASSOCIATED NODAL POWER PEAKS AND EXPOSURES EXPECTED DURING CYCLE 6 PATTERN 1 PATTERN 2 PATTERN 3 0

5.50 2.50 0 4.80 1.80 0 4.15 1.15 8.27 8.27 8.27 7.80 8.27 7.80 7.15 8.27 7.15 2.50 5.50 0 1.80 4.80 0 1.15 4.15 0 Nodal Peak = 1.98 Nodal Peak = 1.93 Nodal Peak = 1.87 4325-4925 4925-5525 5525-6125 PATTERN 4 PATTERN 5 PATTERN 6 0 3.40 0.40 0 2.40 0 0 1.40 0 6.40 8.27 6.40 5.40 8.27 5.40 4.40 7.40 4.40 0.40 3.40 0 0 2.40 0 0 1.40 0 Nodal Peak = 1.82 Nodal Peak = 1.83 Nodal Peak = 1.87 6125-6725 6725-7325 7325-7925 PATTERN 7 0 0.90 0 B A B 3.90 6.90 3.90 0 0.90 0 A 1 A Nodal Peak = 1.82 7925-8525 B A B xxx-xxx +- Exposure, MWD /MTU

• The nwmbers represent the locations of the "A" and "B" banks and Rod 1 in terms of nodes.

10.0 represents a rod fully inserted; O represents a rod fully withdrawn. "A" bank consists of control rods 7, 9, 11, and 13 and "B" bank consists of rods 6, 8, 10, and

12.

LAC-TR-067 A B C D E F G H K L 2-27 2-13 Iy.63 14.7 7 y I' 3; I 2-1 2-33 2-27 3-; 3-3 24C 2-34 3 */ 2.46 ty.64 12.ar CCy 5 00 l1.12 Q 2.34 I4 L __ lf M I6 IT 2-25 3-J;~ 3-4 2 41 241 3-7 3-V 23 b i4.95 K/7 5 77 I3.11 13.l.1 S yr 4.33 IQ ~ QPB _ G'T 7 247 3-9 2-52 3 -lo 3- // 1~45 3-I z 2-48 4 117h T.6G /3.2 T.17 r Y1 13 24 5 33 II.GY 1 @ { '3 /p I C -'~1 P 2-13 3-13l 2D 3 -/Y 2M 2 41 3-)5 246 3-I4 2-2 Y S iv.70 g.,,; 132b v.n 12.ar 12.r;~ g.or

13. x 4.u iva l

9 @ L__ 77 i ?CM fu 1 2-1 3-/7 Z-?T 3-/1 2-67 2-Ch 3-If 2 -b7 3-20 2 6 Iv32f '/.10 13.06 y.9y 12.90 12.37 st. 97 13.41 V73 !y.yl 3 @GS 4 io l 2-6 / 3-2 / za 3-zz J-23 2e/ 3-2y 2-t y 7 lll b3 Z.34 13.1f 1 44 Latt ILSf' [bb l177 &L 12 Il Q ~ IC 2-11 3-25 3-24 2-41 2-70 3-27 3-23' 2-yi 8 I4 12 Y.3 Y ['bf /3 bY I3.96 5 77

5. I 7 LE00 E

I I U 1 A 7A 9 3-21 2-Y1 2-71 3-30 3-31 2-72 2-Yo 3-3 =1 2-4 /y.49 l1.77 f.0 / fg 12.07 /y.73 1.46 ?R _J 2-Y7 2-/ IV.V3 ] IV.77 IN CORE FLUX MONITORS Q PLANT FUEL ASSEMBLY 11 UMBER xxx AVERAGE EXPOSURE (GWD/MTU) yyy FIGURE 3 - LACBWR CORE CONFIGURATION AIID EXPECTED FUEL ASSEMBLY EXPOSURE AT EtID-OF-CYCLE 5. CORE AVERAGE EXPOSURE IS 9,929 MWD /MTU. LAC-TR-067 110 108 uo3o R t 106 ts wo e C 104 o 3o C-s o 4o" 102 100 I i t t 0 2 4 6 8 10 Control Rod No, 11 Position (Modes Uithdrawn) PIGURE 4 - Rod Withdrawal Transient, BOC-6 Relative Core Power vs. Control Pod Position IAC-TR- 06 7 i~ A - Type I Fuel B - Type II Fuel C - Type III Fuel 2.2 2.0 N Ir.U l.8 Eo ss\\\\ M 1.6 ss s C N 1.4 ~ / A 1.32 -----_____---_____________. f f I I i 0 2 4 6 8 10 Control Rod IIo. 11 Position (!Iodes Withdrawn) FIGUPI 5 - Rod Uithdrawal Transient, BOC-6 MCPR vs. Control Pod Position IwN#8 . +, +.. 1a 4 2 , + 0 3 +. t. f',-, . i' j. I - / 5 O . i R t, 4i d + - \\.. 1 2 F i . +, t. 6. )U E T R 5 . t MU L i, l 3' /S E D O U + WP F G X + 0 E ( ) 4 + 2 C 2 E R - . i U N( R A A 1 i4 +.. , + - S A 2 O L I . +,. +. 4 ,4 P P I 4 .? .i,' X

+

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-32z- " POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 4.2.4.2.4 The LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the design LHGR of 11.94 kw/f t for Type I or Type II ( A-C ) fuel rods and 11.52 kw/ft for Type III (ENC) fuel rods during steady-state operation. APPLICABILITY: OPERATIONAL CONDITION 1.* ACTION: With the LHGR of any fuel rod exceeding the limit, initiate adjustment within 30 minutes so that the LHGR is below the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REOUIREMENTS 5.2.17.4 LHGR's shall be determined to be equal to or less than the limit by verifying that each control rod is within the control rod pattern and withdrawal sequence requirements during operation at > 25% RATED THERMAL POWER: a. At least once per 24 hours during steady state operation, and b. Each time power level of the reactor has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established.

• > 25% of RATED THERMAL POWER

-32bb-4.2.4.2 POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 4.2.4.2.1 and 5.2.17.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The specifications of these sections assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident (LOCA) will not exceed 2300 F in compliance with the limits established by the Interim Acceptance Criteria, June 1971 as applied to LACBWR stainless steel clad fuel, Reference 1, 2, 3, 4 and 7. The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat gener-ation rate of all the rods of a fuel assembly at any axial loca-tion and is dependent only secondarily on the rod to rod power distribution within an assembly. The PCT is conservatively calcul-ated assuming the reactor is operating at 102% full rated power and the axial tinea radial peaking factor is at a maximum value of 2.43 from 0-24 GWD/MTU maximum average planar exposure for any Type I, Type II or Type III fuel assembly. The factor is then reduced to 1.75 at 30 GWD/MTU exposure to prevent calculated failure of internal rods. Operation with peaking factors below these values at RATED THERMAL POWER will ensure that peak cladding 0 temperature during a LOCA will not exceed the limit of 2300 F for stainless steel fuel. The corresponding limiting Maximum AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) value for Type III is shown in Figure 4.2.4.2.1-1 and that for Type I and Type II fuel in Figure 4.2.4.2.1-2. The calculational precedures used to establish the MAPLHGR limits shown in these two figures are based on the loss-of-coolant acci-dent analyses performed by Gulf United Nuclear Corporation and Exxon Nuclear Company using approved calculational models consis-tent with the requirements of ECCS criteria as it applies to the LACBWR fuel. The effects of the following parameters were in-cluded: (1) Radial conduction within each fuel rod, (2) Rod and canister convection, (3) Rod thermal radiation among the rods and canister, (4) Gap conductance with exposure and densification, (5) Rod ballooning, and other mechanical and neutronic parameters. The peak cladding temperatures achieved during a postulated LOCA for Type I, Type II and Type III fuel are shown in Bases Figure 4.2.4.2.1-1. Operation at RATED THERMAL POWER with fuel assembly AVERAGE PLANAR HEAT GENERATION RATES below the MAPLHGR limits of Figures 4.2.4.2.1-1 and 4.2.4.2.1-2 will ensure that the PCT's will not exceed the 23000F limit. A list of the significant input parameters to the loss-of-coolant accident analysis is presented in Reference 4.

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-32dd-POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 The daily requirement for surveillance of the core APLHGR above 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The surveillance of core LHGR after power increases > 15% of RATED THERMAL POWER will assure that significant increases in APLHGR are determined. 4.1.4.1.1 and 5.2.17.2 THERMAL PONER-RECIRCULATION FLOW RELATIONSHIP The THERMAL POWER-recirculation flow limiting condition for steady-state operation has been conservatively set below the LSSS curve presented in Figure 4.0.2.2.1-1 of Specification 4.0.2.2.1. This specification also requires that the reactor not be operated in a steady-state condition above the RATED THERMAL POWER, 165 MWth, authorized in the NRC license for the facility. The limiting minimum flow setpoint of 30% of rated recirculation flow is sub-stantially above the natural circulation flow and the flow at which hydraulic instability occurs. A ratio of > 1.6 exists be-tween the low-flow scram setpoints and the instability-natural circulation flow. Therefore, adequate protection of the core against flow and core instability exists over the full power range of anticipated reactor operation as limited by the THERMAL POWER-recirculation flow operating region. The low flow limi.tation requires.Se establishment of a minimum flow of 30% of rated recirculation f13w before reactor startup. Operation in a steady state condition < 68.3% of RATED THERMAL POWER at the minimum 30% of RATED RECIRL:'iATION FLOW assures that the CPR remains above the minimum allowable r-'ve of 1.32 during an abnormal reactor transient (recirculation flow speedup is the most limiting). The steady state limiting CPR values corresponding to other operating conditions bound by the Power-Flow LSSS curve reported in Fig. 4.0.2.2.1-1 are defined in Fig. 4.2.4.2.3-1. The daily requirement for surveillance of the power to recirculation flow relation is sufficient since this relation shifts very slowly when there have not been significant power, flow or control rod changes. The surveillance of this relation after power increases 15% of RATED THERMAL POWER will assure that significant changes In the relation are determined.

4. 2. 4. 2. 3 and 5. 2.17. 3 MINIMUM CRITICAL POWER RATIO The required operating limit MINIMUM CRITICAL POWER RATIO (MCPR) at steady-state operating conditions as established in Specification 4.2.4.2.3 is derived from an analysis of abnormal operational trans-ients with the transient CRITICAL POWER RATIO > 1.32.

The CPR

-32ee-POWER DISTRIBUTION LIMITS R.SES FOR SECTIONS 4.2.4.2 and 5.2.17 MINIMUM CRITICAL POWER RATIO - (Continued) criterion of 1.32 was established, Reference 5, based on the XN-2 predicted power to the measured critical power, assuring better than 99% confidence, a 99% probability of avoiding boiling trans-ition. For any abnormal operating transient analysis, evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the MCPR limit of 1.32 at any time during the transient assuming Limiting Safety System Settings given in Specification 4.0.2.2. To assure that MCPR limit of 1.32 is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO. The type of transients evaluated with pressure increase, moderator temperature decrease, coolant inventory decrease, core coolant flow increase and decrease, and positive reactivity insertion. The limiting transient which determines the required steady-state limit during operation at RATED THERMAL POWER is the rod with-drawal transient. A minimum transient CPR of 1.32 is caused by steady state CPR's of 1.73, 1.49 and 1.50 for Type I, II, and III fuel respectively. A steady state MCPR of 1.73 has been accordingly established for Type I fuel at RATED THERMAL POWER and RATED RECIRCULATION FLOW conditions to ensure no penetration of the minimum allowed CPR value of 1.32. The corresponding steady state MCPR for Types II and III have been maintained at the Cycle 5 value of 1.59 for conservatism and continuity, as Cycle 6 maximum peaking factors are lower than these for Cycle 5. For core flow less than rated recirculation flow, the most limiting transient is the recirculation two-pump speedup caused by a failure of the motor-speed control system. The MCPR limit for steady state operation at flows less than rated recirculation flow are shown in Figure 4.2.4.2.3-1. MCPR values were calculated using a flow control line that corres-ponds to the Limiting Power-Flow line of Figure 4.0.2.2.1-1 which intersects 116.7% power at a maximum of 110% rated recirculation flow. The recirculation pumps are operated in the manual mode only, and the maximum core flow is limited by pump scoop tube travel and pump capacity.

-32ff-POWER DISTRIBUTION LIMITS DASES FOR SECTIONS 4.2.4.2 and 5.2.17 MINIMUM CRITICAL POWER RATIO - (Continued) The limiting bundle relative power was adjusted until the MCPR was equal to 1.32 at maximum flow. For additional conservatism, the transient was assumed to terminate at 120% of RATED THERMAL POWER rather than expected 316.7% power intercept point. Using this relative bundle power as a basis, the MCPR's were calculated at different flows. The calculated MCPR's for Type I, Type II and Type III fuel were then used to establish the curves shown in Figure 4.2.4.2.3-1. These curves represent the minimum allowable operating MCPR for the cost limiting fuel assemblies over the full core flow range of permissible operation. The nominal expected flow control line falls below the Limiting Power-Flow line and intersects the 100% of RATED THERMAL POWER-100% of rated recirculation flow intercept point. The terninal MCPR during a postulated transient from a normal low flow starting condition therefore would result in a MCPR > 1.32. In addition, an automatic reactor trip would be expected to occur due to power-flow and/or 120% overpower trip signals during the transient, but these were conservatively ignored for the analyses. The MCPR limit curves shown in Figure 4.2.4.2.3-1 are conservative and operation with greater MCPR values will assure that the MCPR limits will not be penetrated for the most severe operational transient. At THERMAL POWER less than or equal to 15% of the RATED THERMAL POWER, the moderator void content will be small even at minimum recirculation flow. For all designated control rod patterns under these conditions, the resulting MCPR value is in excess of require-ments by a considerable margin. With the low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for surveillance of the core MCPR is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The surveillance of core MCPR after power increases > 15% of RATED THERMAL POWER will assure that significant redEctions in MCPR are determined. 4.2.4.2.4 and 5.2.17.4 LINEAR HEAT GENERATION RATE The LINEAR HEAT GENERATION RATE (LHGR) specifications for Type I, Type II and Type III fuel assure that during steady-state operation, the peak LINEAR HEAT GENERATION RATE in any fuel rod is less than the design LINEAR HEAT GENERATION RATE. The specifications also assure sufficient margin to accommodate maximum centerline fuel temperatures less than the molting point during operational transients.

-32gg-POWER DISTRIBUTION LIMITS BASES FOR SECTIONS 4.2.4.2 and 5.2.17 LINEAR HEAT GENERATION RATE - (Continued) For Type I and Type II (A-C, fuel, the original design LINEAR HEAT GENERATION RATE specified by the fuel nanufacturer was conserva-tively reduced to 11.94 kw/ft to account for the effects of densifi-cation, power spikes and manufacturing factors. Por Type III (ENC) fuel, the design LINEAR HEAT GENERATION RATE of 11.52 kw/ft is also calculated with design conservatisms that are larger than the calculated axial densification effects plus manufacturing toler-ances and power spike effects, Reference 6. The daily requirement for surveillance of the core LHGR above 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The surveillance of core LHGR after power increases > 15% of RATED THERMAL POWER will assure that significant increases In LHGR are determined. 4.2.3.2.5 and 5.2.17.5 Maximum Average Fuel Assembly Exposure Fuel cladding integrity is a function of many parameters including fuel exposure, pellet clad interaction, THERMAL POWER, rate of change in power density, coolant chemistry, etc. Therefore, limiting fuel exposure will give additional assurance that the condition of any fuel assembly during operation will be satisfactory. The operating history of the LACBNR indicates that a limit of 15,000 MWD /MTU for the maximum average exposure of any fuel assembly is very conservative. During previous operation, a number of fuel assemblies have exceeded 15,000 MWD /MTU without any indication of failure and at the end of Cycle 3, EOC-3, four assemblies had ex-ceeded 18,000 MWD /MTU without failure. The average exposure of the 25 assemblies discharged at EOC-3 was 15,530 MUD /MTU and the peak exposure was 21,532 MWD /MTU. Gross fuel failure involving observ-able loss of fuel or clad has only occurred in three fuel assemblies, all during fuel Cycle 4, and all three of these assemblics were over 17,400 MWD /MTU exposure. The average exposure of the 32 assemblies discharged at EOC-4 was 16,459 MWD /MTU. Pellet-clad interaction is a well known and documented contributing factor to fuel rod failures. The presence of pellet cladding interaction has also been identified in post-irradiation examinations of fuel rods removed from LACBWR fuel elements. Fuel rods removed from fuel elements with average exposure up to 14,700 MWD /MTU have been exanined. The strength, ductility, and condition of the cladding in these rods was found to be adequate as determined by mechanical tests. The examination further confirmed that power history of the rods is of prire im-portance, though not the only factor in contributing to fuel rod

-32hh- ~ POWER DISTRIBUTION LIMITS ~~ ~ BASES FOR SECTIONS 4.2.4.2 and 5.2.17 MAXIMUM AVERAGE FUEL ASSEMBLf EXFOSURE - (Continued) 1 failure. A limit of 15,000 MWD /MTU fuel element average exp6sure is consistent with the results obtained from examinations ~ con-ducted on fuel elements with similar exposure history. During future operation the rate of withdrawal of control rods when the THERMAL POWER is above 25% of RATED THERMAL POWER will be reduced from that experienced during prior operation which will also significantly reduce the stresses in the fuel clad. Additional surveillance and limitations on coolant and off-gas activity will assure that operation does not continue with grossly failed fuel.

References:

1. " Technical Evaluation Adequacy of La Crosse Boiling Water Reactor Emergency Core Cooling System", Report SS-942, Gulf United Nuclear Corporation, May 31, 1972. 2. " Review of Densification Effects in La Crosse Boiling Water Reactor", Report SS-1085, Gulf United Nuclear Corpor-ation, May 15, 1973. 3. NRC Safety Evaluation Report, Letter Reid to Madgett, dated August 12, 1976. 4. "ECCS Analysis for Type II and Type III Fuels for the La Crosse Boiling Water Reactor", Exxon Nuclear Company, Inc., XN-NF-77-7, March 1977. 5. " Transient Analysis for LACBWR Reload Fuel", Response to Question 4, Nuclear Energy Services, Inc., Report GlA0025, February 18, 1977. 6. " Description of Exxon Type III Nuclear Fuel for Batch 1 Reload in the LACBUR", Dairyland Power Cooperative, LAC-3929, May 17, 1976. 7. Exxon Nuclear Co. Letter, J. A. White to C. W.

Angle,

Subject:

MAPLHGR Limits for Type I (Allis-Chalmers) Fuel, dated June 22, 1977. (Next page is page 33)}}