ML20004F554
| ML20004F554 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/01/1981 |
| From: | Raffety S DAIRYLAND POWER COOPERATIVE |
| To: | |
| Shared Package | |
| ML20004F551 | List: |
| References | |
| LAC-TR-096, LAC-TR-96, NUDOCS 8106190066 | |
| Download: ML20004F554 (24) | |
Text
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, to DPC Letter LAC-7572 Dairyland Power Cooperative Report: LAC-TR-096, "LACBWR Cycle 6 Fuel Performance and Finalized Refueling Plan for Cycle 7" t~
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LAC-TR-096-LACBWR CYCLE 6 FUEL PERFORMANCE AND FINALIZFD REFUELING PLAN FOR CYCLE 7 S. J. RAFFETY DAIRYLAND POWER COOPERATIVE Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
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! LAC-TR-096~
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SUMMARY
LACBWR Fuel Cycle _'6 ended on1 NovemberT 9, 1980, afterLapproximately 17.5 months operation.. The incremental core exposure-during the
. cycle was 7339' MWD /MTU and the EOCLcore average exposure was-ll,542 MWD /MTU. 'The cycle was limited by depletion'of excess reactivity
~
'resulting-from fuel burnup.
'Throughout the cycle the off-gas activity and primary coolant gross B/y, a,
I-131:and Dose Equivalent-131 activities exhibited rela-
.tively constant, low 'ralues indicating very little if any fuel clad degradation.
(See Figure 2.)
During.the refueling outage each fuel assembly was removed from
-the core and examined.
No fuel deformation or clad defects were observed.
One assembly produced a weak indication of fission gas leakage during dry sipping (28 times background) and-this' assembly probably contains a fuel rod with a small clad defect.
In general, the fuel: condition at the end of Cycle 6 was much better than at the end of previous LACBWR cycles.
See Table I for a historical summary of LACBWR fuel performance.
The LACBWR core configuration for Cycle 7 consists of 24 fresh Type IIIfassemblies, 46-previously exposed Type III assemblies, and 2 previously exposed Type I: assemblies, all in Zircaloy shroud cans.
The core average exposure at BOC-7 is 6528 MWD /MTU.
Under the present Technical Specification-limit of 15,600 MWD /MTU
'for any assembly'the expected length of Cycle 7 is 208 full power days.! If the Technical Specification burnup restriction can be eliminated,.the cycle length could be approximately 263 full power days.
2.0 CYCLE 6 ANALYSES 2.1 Cycle 11istory Cycle 6 began on May 25, 1979, and ended on November 9, 1980, for a total of 534 days.
During this period the energy produced was equivalent to 356.1 full power days.
The exposure distribution in the fuel at the EOC 6 is shown in Figure 1.
The average exposure at EOC uas.ll,542 MWD /MTU and the cycle length was 7339 MWD /MTU.
The cycle was limited by depletion of excess reactivity resulting from fuel burnup.
When the cycle was terminated the control rods were fully withdrawn and the maximum achievable power had coasted down to 75.5 % of rated power.
Figure 2 shows a power histogram for Cycle 6 along with off gas activity and primary coolant gross B /y, a,
I-131 and Dose Equiva-lent I-131' activities.
The radioactivity in the coolant and off-gas of a reactor is a sensitive indicator of fuel clad integrity.
s:
- *a a
- LAC-TR- 0 9'6 a
As can be:seen-from'the plots-inLFigure~ 2~all o'f;these pa'ameters-i J
r exhibited relatively: constant,- low values'in the LACBWR during-
~
=Cyclec6) indicating very little if any; clad _ degradation during the cycle. -The majority of the fission product-activities.that were, observed;in~the reactor. coolant or off-gas during Cycle 6 probably toriginated'from fissioning of fuel material'from previously-failed
-fuel that was deposited as' crud on the surface of-fuel rods and other: core components.
The infrequent spikes in = activity in the reactor-coolant were also due._to fuel material still in the system from previous. failures being resuspended in the coolant by reactor shutdown-~ transients and the restarting of-coolant pumps,fetc.
_ -2.2-Fu'el-Inspection Each fuel assembly was removed _from the reactor and examined in the
~
' spent. fuel storage pool with an underwater TV camera and by direct
. visual observation.
All assemblies that-had been in the reactor for more than one fuel' cycle plus the 2 Type I (A-C) assemblies
- were examined for fission gas release by dry sipping.
After com-pletion of the inspections all fuel was stored in the new storage racks in'the spent fuel storage pool while other core internals inspection and core component replacement was accomplished.
2.2.1 General Appearance The general appearance of the fuel was very good.
No deformation
.of fuel rods or other assembly components were observed and no clad-defects were evident.
As in previous cycles the older, high expo-sure fuel assemblies from low power peripheral core positions exhi -
bited relatively heavy crud deposi ts on the fuel rods, especially near the bottom of the assemblier, and considerable flaking off of this crud la3er was evident.
The newer fuel and fuel from high power core pr ;itions exhibited auch less crud on the_ fuel rods.
All of the fuel assembly nozzle s and lower fuel rod support grids were very clean with essentially no crud deposits evident.
2.2.2 Dry Sipping Results All of the fuel assemblies that had been in the reactor for more than one fuel cycle plus the 2 Type I (A-C) assemblies (46 assemblies total) were examined for fission gas release by dry sipping.
Only one possible leaker was identified and it produced only a relatively weak indication, approximately 28 times background.
2.2.3 Comparison with Previous Cycles The fuel condition at the end of Cycle 6 was much better than at the end of any prior LACBWR fuel cycle.
No visible fuel rod defects were evident in any of the fuel.
One Type II ( A-C) fuel assembly produced a relatively small indication of fission gas leakage during dry sipping and probably has one fuel rod with a small through wall defect. J
1 9
LAC-TR-096 The historical performance of the LACBWR fuel from Cycle 1 through
. Cycle 6 is presented in Table I.
As can be seen in this-table, Cycle.6 was one of the longer LACBWR fuel cycles at 7,339 MWD /MTU.
TheLend-of-cycle core average exposure, maximum assembly average exposure, and. average exposure of assemblies discharged compares favorably with previous cycles, but is still considerably below
. design values or those values desirable for greater economy.
The cycle operating conditions were'approximately average for the LACBWR with 50 startups (% 29 to heating power only), 14 scrams while at power, and 7 normal shutdowns from power.
Seven startups were from cold conditions.
The end-of-cycle off gas and primary system activities for each cycle are also listed in Table I and it'is evident that at the EOC-6, all of these indicators of fuel condition'are considerably lower than at the end of previous cycles.
Operational experience to date indicates that the Type III (Exxon) fuel is less prone to failure ind will give better and longer service than the Type I and Type II (A-C) fuel.
3.0 FINALIZED LACBWR RELOAD PLAN FOR CYCLE 7 3.1 Core Configuration for Cycle 7 Twelve Type II fuel assemblies with an average exposure of 15,885 MWD /MTU and twelve Type III assemblies with an average exposure of 13,859 MWD /MTU were discharged at the end of Cycle 6 The average exposure of the 24 assemblies was 14,872 MWD /MTU.
The assemblies discharged along with their core location during Cycle 6 and their exposure are listed in Table II.
The core configuration for Cycle 7 consists of 24 fresh Type III assemblies intermixed in a checkerboard fashion with 46 previously exposed Type III assemblies and 2 previously exposed Type I assem-blies as shown in Figure 3.
The core average exposure at BOC-7 is 6528 MWD /MTU.
During the reconfiguration of the LACBWR core for Cycle 7, eight stainless steel fuel shrouds and 28 high exposure Zircaloy fuel shrouds were removed and replaced by 36 new Zircaloy shrouds.
During Cycle 7, there will be no stainless steel fuel shrouds in the fueled region of the LACPWR core.
4 3.2 Expected Length of Fuel Cycle 7 The detailed studies of the burnup of Cycle 7 with the TRILUX code show that the exposure of the lead assembly not on the periphery of the core reaches the present Technical specificatior' limit of 15,600 L.0-TR-096 MWD /MTU when the' core average exposure is approximately 10,889 MWD /MTU. :The corresponding length of Cycle 7 is 208. full power-days or 346 days at a capacity factor of 60%.
With no. limit:on fuel exposure (or at least a higher-limit) and coast down to approximately 85% of rated power, the fuel cycle Lcould be extended by at,least two months to a core average exposure'of approximately 12,050 MWD /MTU.
This exposure corres-ponds to a cycle length of 263 FP' days or 439 days at a capacity
. factor.of 60%.
The expected fuel exposure distribution near the end of Cycle 7' is shown;on' Figure 4. j
p-I r
k TABL FUEL PERFORMANCE CYCLE 1 CYCLE lA CONDITION 11 JULY 6 7-14 OCT. 72-19 AUG. 72 30 MAR. 73 CORE AVG. EXPOSURE (MWDAITU) 0 - 9968 83667 - 11,107 CYCLE LENGTII (MWDAITU) 9,968 2,440 MAX. BOC ASSEMBLY AVG. EXP.
0 12,810 MAX. EOC ASSEMBLY AVG. EXP.
12,810 15,770 NO. OF ASSY. DISCH.
8 26 AVG. EXPOSURE OF DISCII.
11,490 14,360 NO. OF ASSY, DISCH. WITf! E 15,000 0
10 NO. OF ASSY. DEFECTIVE 8 PROBABLE 20 NO. CF ASSY. WITH VISIBLE DEFEC',
1 15 MAX. NO. uF VISIBLY DEFECTIVE RODS PER ASSEMBLY l
9 T TAL NO. OF VISIBLY DEFECTIVE RODS 3
42 ESTIMATED TOTAL NO. OF DEFECTIVE RODS NA 47 NO. OF ASSY WITil FEEL DISPLACED DURING OPERATION O
O AVG. EXPOSURE OF DEFEC*IVE ASSY.
11,490 13,986 EXP. RANGE OF DEFECTIVE ASSY'S 11,034 - 12,136 12,264 - 15,205 MAX. EXPOSURE OF GMD ASSY.
12,810 15,770 SCPMtS WI!ILE AT PO4ER NA 4
COLD SilUTDCWi:S NA 2
1 TOTAL NO. CF STARTLTS/TO HEATING RANGE ONLY I
9 (NO. OF ASSY./ TYPE) 72/I 72/I.
FUEL TYPES
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2 LAC-TR-096 4
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'IN THE LACBWR CYCLE 2' CYCLE 3 CYCLE 4 CYCLE 5-CYCLE 6
'25 JUNE 7 F-21-DEC. 73-11 AUG. 75-9 MAR. 78-25 May 79-3 NOV.-73 9 MAY 75 11 MAY 77 25 MAR. 79 9 Nov. 80 l6,251 - 7,953 3,928 - 11,269 5,906 --12,833 5,763 - 9,729 4203 - 11,542 1,702 7,341 6,927 3,966 7,339
'15,300 15,660 -
11,580 12,589 13,065 I
'16,740 21,532 19,642 14,889 16,688
- 25 25 32 28 24 11,591 15,530 16,459 13,966 14,872 i
3 9
27 0
12 23 10 26 17 1 PROBABLE 11-4 6
7 0
l l'
.8 6
9 4
o l'
36 18 19 12 0
48 24 40 22 1
l 0
0 3
0 o
i 11,190 16,691 16,774 13,880 16,6P0 2,200 - 15,300 13,528 - 21,532 12,042 - 19,642 11,925 - 14,889 NA 16,740 18,982 17,361 14,713 16,658 l
7 14 9
12 14 l
2 5
5 5
6 l
l
-l-53/27 32/ 11 23/7 50/29 20/8 48/I, 24/II 24/I, 48/II 72/II -.{
40/II, 32/III
' 2/I, 12/II, 58/III TABLE I i L_
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~ _ _ -
I TABLE I -
4 FUEL PERI ORMANCI CYCLE 1 CYCLE lA I
CONDITION 11 JULY 67-14 OCT. 72-19 AUG. 72 _
30 MAR. 73 STARTUPS FROM CCLD CONDITION ROD MOVEMENT RESTRICTIONS Rod Interchange Rod Interchange Allowed at. Full Allowed at Full Power Pover PCWER ESCALATION RESTRICTIONS None Above 10% Power 1%/ Min. With 10 Min. Hold Af ter Each 5% Increase.
EMD OF CYCLE CONDITIONS:
I POWER (% OF RATED POWER) 97%
98%
3
% 320 N 900 OFF-GAS ACTIVITY AFTER 150 ft HOLDUP TANK (C1/ DAY)
PRIMARY SYSTEM GROSS S/y N 1.0
% 4.0 ACTIVITY (11Ci/g)
PRIMARY SYSTEM I-131
% 69.9E-3 ACTIVITY (pCL/g)
PRIMARY SYSTEM DOSE EQUIVALENT I-131 ACTIVITY (pCi/g)
PRIMARY SYSTEM a ACTIVITY N O.35E-6 N 3.5E-6 (pC1/g)
- Values in parentheses are estimated for 98% reactor power.
+
LAC-TR-096 (Cont'd)
- IN TIIE LACBWR CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 25 JUNE 73-21 DEC. 73-11 AUG. 75-9 MIA 78-25 MAY 79-3 NOV. 73 9 MAY 75 11 MAY 77 25 MAR. 79 9 NOV. 80 3
6 6~
6 7
Fo Rod Inter-No Rod Inter-No Rod Inter-No Rod Inter-No Rod Inter-changes Allowed changes Allowed changes Alloweu chages Allowed changes Allowed Above 10% Power Above 10% Power Above 10% Power 5%/Hr. Up To 50%
5%/Hr. Up to 50%
1%/ Min, with 10 10%/Hr. After 10%/Hr. Also, Power. 1%/Ilr.
Power. 1%/Hr.
Min. Hold After July 1974, Re-On Initial Es-Above 50% Power.
Above 50% Power.
l'ach 5% Increase.
duced to 1%/Hr.
calation of All Cold Startups All Cold Startups A.so, 5%/ Day Above 60% Power.
Cycle 10%/ Day 10%/ Day From 30%
10%/ Day from 30%
Above 57% Power Also, Cn Initial From 40% to 601 lte 50% Power and to 50% Power and On First Escala-Escalation of and 5%/0ay 5%/ Day Above 50%
5%/ Day Above 50%
tion of Fuel Cycle, 10%/ Day Above 60% Power.
Power.
Above 25%
Power. Above 25%
- Cycle, from 40% to 60%
Power, No Control Power, No Control and 5%/ Day Above Rod Withdrawn Rod Withdrawn 60% Power.
.vore Than 4"/Hr.
More Than 4"/Hr.
82%
85%
56%
49%
76%
N720 N450 N1000
.N290 N200 g
(N860) *
(N519) *
(N1750) *
(N580) *
(N361)
- 7 N2.1 N2.0 N13.0 N3.1 N.8 (N2. 5 ) *
(N2. 3) *
(N22.8)
(N6.2)*
(N1.0)*
N47.9E-3 N14.7E-3 N37.4E-3 N 9.3E-3 N2.lE-3 (N57.2E-3*
(N6 5. 4E-3) *
(N18.6E-3)*
(N2.1E-3)*
N3.9E-2 N1.5E-2 (N7.8E-2)
(N1.9E-2)*
N 1.2E-6
% 3.5E-6
% 66.lE-6
%.85E-6
%.08E-6 TABLE I - (Cont'd) j_Q 'T K _. ((
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1 TABLE"II~
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'LACBWR Fuel Assemblies Discharged atLEnd!offCycle=6 -
' Core Position Exposure, MWD /MTU Fuel ^ Assembly No.
~ ~
2-58*
-F-1 116,688-2 A-6 16,658' N'
2-67
~E-1
'16,320-
~
2-66 E-10 16,310 2-54 F-10 16,309 12-37
-L 15,734 2.71-A-5 15,704 2-50 L-6 15,589 2-64 B-9 15,452 1
2-57.
K-2 15,407 2-61
~B-2 15,230 2-60 K-9 15,223
- 3 C-6_
14,334 3-6 H-5 14,276 3-15 D-7 14,167 3-18 G-4 14,142
~
3 D-4 13,974 3-19 G-7 13,953 26 C-5 13,945 3-7 H-6 13,884 3-24 E-8 13,540 3-9' F-3 13,533 3-23 E-5 13,281 3-10 F-6 13,275
. Average exposure of discharged assemblies is 14,872 MWD /MTU.
- Assembly 2-58 probably contains a defective fuel rod as indicated by a weah dry sipping signal (N28 times background).
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IO 2h 14.31 IMI IN CORE FLUX MONITORS Q PLANT NORTH Fuel Assembly Number xxx Average Exposure (GWD/MTU) yyy Denotes Stainless Steel Shroud Cans L FIGURE 1 - LACBWR Core Configuration and Fuel Assembly Exposure at End-of-Cycle 6, November 9, 1980.
EOC Core Average Expo-sure is 11,542 MWD /MTU.
(S-1A and S-2A are startup neutron sources) -
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- 32:a - L (This Page is Left Blank Intentionally) Deleted 4.2.4.2.5 Deleted 5.2.17.5 Amendment No.
{y,. ~ - 32gg - g . :s? W POWER DISTRIBUTION LIMITS BASES FOR SECTION 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 manufacturer was conserva-tively reduced to 11.94 kw/ft to account for the effects of dens-- ification, power-spikes.and manufacturing factors. For 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 tolerances and. power spike effects, References 6 and 7. 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 f.ncreases In LHGR are determined.
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 Bciling Water Reactor", Report SS-1085, Gulf United Nuclear Corporation, 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 81A0025, February 18, 1977. 6. " Description of Exxon Type III Nuclear Fuel for Batch 1 Reload in the LACBWR", Dairyland Power Cooperative, LAC-3929, May 17, 1976. 7. Exxon Nuclear Co. Ietter, 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) Amendment No. t}}