ML20002E025
| ML20002E025 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/03/1971 |
| From: | James Shea US ATOMIC ENERGY COMMISSION (AEC) |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20002E022 | List: |
| References | |
| NUDOCS 8101260133 | |
| Download: ML20002E025 (12) | |
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UNITED STATES I
k ATOMIC ENERGY COMMISSION
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WASHINGTON, D.C. 20545 "g. a f
March 3, 1971 9,
AAft4W bk t Do'. 5 -155 Files (
D. L. Zieb nn, ) Chief, ORB #2, DRL
'THRU:
EVALUATION OF TWO NEW CENTER'!ELT FUEL ASSEMBLIES FOR BIG ROCK POINT NUCLEAR REACTOR (CONSUMERS POWER COMPANY).
INTRODUCTION Consumers Power Company, by letter dated December 21, 1970, reauested approval of Proposed Change No. 24 to insert two new centemelt fuel bundles into the Big Rock Point Nuclear Reactor during the February 1971 refueling outage. The Consumers Power Company and the General Electric Company Nuclear Energy Division, have jointly undertaken the dontinuation of the centermelt irradiation testing program at Big Rock Point, a procran that was originally sponsored by EURATOM and the U. S. AEC, and teminated on June 30, 1969. Of the six original centermelt fuel assemblies, five were removed after the first 3-month cycle of irradiation pending destruc-tive evaluation of selected fuel rods. The sixth centemelt subassembly (D-50) was removed in May 1969 when suspected multiple fuel rod failures were confimed by visual inspection of the bundle. The cause of premature rod failures was tentatively attributed to accelerated corrosion due to clad overheating as a result of excessive crud deposition, predominantly a copper oxide. A supplemental report (6), based on hot cell examination of two fuel rods from the Intemediate Performance Centemelt Assembly that was irradiated for about one year achieving high power rod average exposures of 10,000 mwd /TU, indicates that the cause of severe clad deterio-ration was accelerated corrosion on the outside surface of the clad driven by local overheating of the clad. Grain growth in the zircaloy structure adjacent to the deterioration indicated temperatures of 1200-1300*F.
Failure was attributed to excessive crud deposition and high surface heat fluxes.
DESCRIPTION The two new centemelt fuel assemblies (D-57, an intemediate perfomance fuel assembly consisting of an 8 x 8 array of 0.570 inch 0.D. fuel rods, 16 of which are hot enough at rated power conditions to have incipient centerline fuel melting; D-56, an advanced perfomance fuel assembly con-sisting of a 7 x 7 array of 0.700 inch 0.D. fuel rods, 16 of which have definite but moderate center UO2 melting at rated power conditions) di r
from the original centemelt fuel assemblies that were approved by DRL and inserted into the Big Rock Point reactor in March 1968 in that:
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s 4 Filcs March 3, 1971 1.
stainless steel tubes with the same dimensions as fuel rod cladding are used:
a.
as corner structural posts for tying the upper and lower frames
- together, b.
to hold the five fuel rod spacers in each assembly at the proper elevation, and c.
to hold dummy rods containing cob' alt targets.
2.
there are only 16 high power rods in each bundle in contrast to 36 in the original 8 x 8 intermediate performance centermelt assemblies or 29 in the original 7 x 7 advanced performance centermelt assemblies.
We have prepared the following table from information provided by Consumers Power Company to identify the most important performance characteristics of the new centermelt fu fuelassemblieswehavealreadyapprovedy})assembliesincontrastto for the Big Rock Point core, and in this manner provide the basis for our evaluation.
For example - line 8 shows that the weight of UO2 for each b6ndle type has decreased more than can be accounted for by the omission of fuel in the corner positions, i.e.,
7.2%
for the 8 x 8 instead of 6% and 10.5% for the 7 x 7 instead of 8%.
1_ine_9_ shows a 16% decrease in U235 per bandle.
11ne__1_0 shows 9.5% decrease in fuel enrichment for the 8 x 8 fuel assembly; 7.5% decrease for the 7 x 7 fuel assembly.
lines 7, 1? and 13 show that 96% of the power was generated in 56% of the fuel (36 rods) in the original 8 x 8 in contrast to 44.5% of the power in the new 8 x 8 which is generated in 26.7% of the fuel (16 rods).
Similarly for the 8 x 8, 95.7% of the power was generated in 59.0% of the fuel (29 rods) compared with 55.60%
power generated in 35.60% of the fuel (16 rods).
_line 14 shows that the percentage of bundle power generated in the hottest rod of each bundle is about the same or up slightly.
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Filso A March 3, 1971
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I lin_e_ 18 shows that the ratio of power in adjacent rods is noticeably lower for the new fuel in contrast to the original centermelt assemblies and 1_ine 22 shows that there are fewer low power rods adja-cent to the highest power generation rod.
i line 21 shows that power generation in the original hot rods was slightly higher than anticipated and~
line 20 shows the ratio of old to new fuel rod peaking factors
' line 12 shows that the Technical Specifications MCliFR 1.5 is satisfied at the 122% steady-state p,9wer level.
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M Comparison of New and Original Centermelt Fuel r
1 2
3 4
7 x 7 Adw' 4ced 8 x 8 Intermediate Performance Centermelt Performance Ce ntermelt f
Original New Original New (Ref. 1)
Prop. Ch. 24 (Ref. 1)
Proo. Ch. 24
- 1) Rod' Diameter inches 0.570 0.570 0.700 0.700
- 2) Fuel Rods / Bundle 64 60 49 45
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- 3) Cobalt Targets in corner positions 0
4 0
4
- 4) Number Depleted fuel rods per bundle (low' power) 28 0
20 0
- 5) Number Natural UO2 rods per bundle (low 0
28 0
12 power) 1
- 6) Number Intermediate 4
Power (2% U235) rods 0
16 0
17 per bundle 4
- 7) Number High Power rods-per bundle 36 16 29 16
R 1
2 3
4
- 8) Weight UO / Bundle 140 167 2
kg 136 cale 126 cale 161 cale 144 calc
- 9) Weight U 0 / Bundle-kg 3.89 calc 3.26 calc 4.75 cale 3.98 cale 235 2
- 10) Average bundle 3.89 or 2.86 3.26 or 2.58 4.75 or 2.95 3.98 or 2.76 enrichment %
136 126 161 144
- 11) Percent total UO in 36(
)S6.3 16(IN)26.7 29(100)S9.2 16(100)35.6 E#
E E
high power rods R
- 12) MCilFR at 122% Power Multi Channel Model 1.56 (Ref 4c)*
1.54 (Ref 4c)*
Multi Rod Correlation 1.53 1.58 (Ref 5 a)
(Ref.Sa)
- 13) Percent Power Generated (Ref 2b)
(Ref 4 a&b)
(Ref 2a)
(Ref 4 a&b)
/
in Low power rods / bundle 4.0 11.1 4.3 15.85 Intermediate rods / bundle 0
44.4 0
28.55 liigh power rods / bundle 96.0 44.5 95.7 55 60
- 14) Percer.t bundle power generated in highest power rod 2.8 2.8 3.43 3.55 s
- MCIIFRs are reported for the intermediate performance assembly only. New thermal hydraulic correlations have been used to calculate the MCHFRs and therefore a direct comparison of MCHFRs is not valid.
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- 15) Variation of power in h1gh power generation rods within each bundle 8.5 2.0 7.5 3.0
- 16) High Power Rods 12 @ 4.3%
12 @ 4.3%
Enrichment - ea bundle 16 @ 5.0%
12 @ 5.0%
8 0 5.6%
5 0 5.6%
l 8 0 4.5%
8 0 5.0%
/
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l 4 0 5.6%
8 @ /,.5%
4 9 6.5%
- 17) Highest rod power 1.81-1.83 1.688 1.71-1.89 1.596 factor / bundle Lowest Rod Power Factor _
0.10-0.21 0.29 0.09-0.24 0.289 Bundle 1.688 or 5.83 1.596 or 5.5
- 18) Ratio Highest Rod Power Lowest Adjacent Rod 0.29 0.289 (Ref 2b)
(Ref 2a) 1.81 or 18.1 1.71 or 19.0 G
0.10 (Ref 3a)
(Ref 3a) 1.89or).9 1.83 or 8.7 0.21 0.24
- 19) Moderator /UO 2
W/F ratio 2.6 1.93
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3 4
- 20) Ratio Hot Rod Old 1.81 or 1.83 or 1.07-1.08 1.71 or 1.89 or 1.07 or 1.18 Hot Rod New 1.688 1.596
- 21) Ratio-Design liot R_od (Ref 3a) 1.83 or 1.01
'l.89 or 1.11 Actual Hot Rod 1.81 1.71 (Ref 2a, 2b)
- 22) Max. No. low power 6
1 low 6
4 rods adjacent to 3 intermediate hottest rod
- 23) Fuel Depletion av.
15,000 10,000 15,000 10,000 Bundle }Md/T liigh Power rods 20,000 15,000 calc 20,000 15,000 cale (Ref 2c)
(Ref 4d)
(Ref 2c)
(Ref 4d) e 4
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C Files March 3, 1971 EVALUATION The Intermediate Performance Fuel assembly, D-50 (an 8 x 8 array of fuel rods irradiated for a three-month period with five other centermelt fuel assemblies in the Big Rock Point reactor followed by an additional irradi-ation period of about eight months af ter the other five centermelt assemblies had been removed from the reactor to await destructive evaluation of fuel rod performance), will not be re-inserted into the core. This fuel assembly with hot rod average exposures of 10,000 mwd /TU has been permanently removed from the centermelt fuel irradiation program because of failure of many rods.
The failures are attributed to severe clad deterioration caused by accel-(6) erated corrosion on the clad surface where clad overheating had cccurred.
Similar failures were observed in normcl reload fuel that was irradiated during the same period.
Excessive crud deposition and high heat fluxes were reported to be the main factors involved in creating the high temper-ature condition. Neutrographs of one entire rod with incipient failure showed r31 hydriding of the cladding at the inside surfaces.
The absence of such hydriding is an indication that there was no problem with contam-Ination of the fuel with hydrogenous impurities.
In other words, examin-ation of fuel rods from the failed centermelt fuel assembly has revealed that the failures were not caused by impurities in the fuel, that clad temperatures were excessive and that there was an unusual accumulation of crud on the centermelt fuel rods and other reload fuel rods which also failed. We have concluded that these failures are not due to the high centerline fuel temperatures in the centermelt fuel assembly and we agree that the resulcant higher than normal heat fluxes can cause signi-ficant increases ir c:
Thelicenseehasstated[gjgsstherateofcrudaccumulation temperature <
is reduced.
that the new centermelt fuel assemblies will not be inserted into the Big Rock Point reactor until there is reasonable assurance that crud deposition on fuel rod surfaces has been significantly reduced to prevent local clad heating.
Based on the evidence presented, we agree that this is prudent. The licensee plans to insert the five centermelt assemblies remaining from the original centermelt irradiation program some time in the future af ter 1) chemical cleaning of the fuel rods to remove the crud accumulated during three months of in-core irradiation and 2) substitution of low power rods for the high power density rods in the outside rows of fuel rods.
The net effect of item 2 will be to reduce the number of high power rods from 188 in the six original assemblies to 170 in the 7 assemblies to be retained in the centermelt irradiation program. We have concluded that this reduction in the number of high power rods and compliance with the previously approved restriction that centermelt fuel assemblies be no closer than 16.5 inches center-to center reduces the accident consequences below those that were reviewed and accepted for the original six centermelt assemblies.
The licensee also has reported that the reactivity value for the new assemblies
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i Files March 3, 1971 is lower than the core average thereby reducing fuel rod reactivity worths and increasing shutdown margins. Based en line 9 of the table; we-concur that there is a significant reduction in U235 and therefore a cor-responding reduction in reactivity.
Similarly, the reactivity coefficients meltfuelassemblies(geslightlymorenegativethantheoriginalcenter-for the new assemblics and therefore acceptable since the severity of accidents will not be increased beyond the values calculated for the original six centermelt fuel assemblies. We concur that 1) the hot cell examinations of irradiated centermelt fuel have been completed a j) reported (3)(6) in accordance with the requirements of Amendment No. 1
, 2) the accumulation of crud on fuel rods should d2 crease with time, 3) the rate of crud accumula-tion on the centermelt fuel rods should be measured (at each refueling out-age), 4) the instrumented Reload-F assembly rod may give insight into the crud deposition problem, and 5) the operational experience with centermelt fuel so far warrants a continuation of the centermelt fuel irradiation program.
We have noted the following inconsistencies in the application. The MCHFRs at the.122% overpower condition are increased although water enthalpy must have increased because the_ rods adjacent to the high power rods generate significant power in contrast to the depleted fuel rods in the original centemelt assemblies. The Advanced Performance high power rods may not achieve the objective of molten fuel at the center because,) based on a 2 at 122% power (4g, the rated reported heat flux of 535,000 Btu /hr. ft heat flux is expected to be 440 000 Btu /hr.ft.2, too low to cause melting g rod ( h) at the start of life when such heat at the center of the fluxes are attainable We also note that clad failure has been attri-buted to a very localized clad overtem ture condition resulting in accelerated corrosion of the cladding of centemelt and reload fuel rods, ar.1 that a centributory factor, crud accumulation, has been 2dentified but the precisc cause of. local failures has not been determined. We are satisfied, however, that the failures were not caused by molten fuel conditions.
Additional infomation has been requested from the licensee to obtain greater accuracy in the comparative evaluation of the original and new centemelt perfomance characteristics but this information is not required to complete our safety evaluation. The general impression of some backing down from the original objec~tive cf definite center melting is evident. We have concluded that the hazards of operation with the two new centemelt assemblies and five of the original, chemically cleaned, and reconstituted assemblies are no greater than those considered in our previous evaluation of centemelt fuel assemblies for the Big Rock Point Nuclear reactor (1) and that the Technical Specifications may be changed to permit reactor operation with the two new centermelt fuel assemblies and five of the original centemelt fuel assemblies in the manner proposed by the licensee.
' A f
i Files Marcli 3, 1971 e
CONCLUSION The two new centermelt fuel assemblies and five cleaned and reconstituted centermelt= fuel assemblies together-will include 172 fuel rods that will operate with centermelting or near centermelting temperatures, 16 less than were contained ~1n the' original 'six centermelt' fuel. assemblies.. The total energy _in the high performance fuel ~ rods as a result of this change is lower than the value-considered in our original. evaluation. The mechanical desig.._and fuel distribution of the new. elements have been_ improved. We have therefore concluded that operation of the Big Rock Point Nuclear Reactor in the. manner proposed by Consumers Power Company will not increase the proba-
'bility of or change the consequences of the design basis accident nor docs it involve significant hazards considerations not described or implicit in the Safety Analysis Report for Amendment.No. l to the operating license or impair the effectiveness of_ engineered safety systems (Core Spray).
There is reasonable assurance that the health and safety of the public t
will not be endangered by operation of the Big Rock Point Nuclear Reactor with two new and five previously.irr diated centermelt fuel assemblies in the core and therefore,-pursuant to Section 50.59 of 10 CFR Part 50, the Technical Specifications of Facility License No..DPR-6 should be changed as proposed.
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\\sq[tmes J. SWea Operating Reacters Branch #2 Division of Reactor Licensing
Enclosure:
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References cc:
D. J. Skovholt DRL R.11. Vollmer, DRL D. L. Ziemann, DRL J. J. Shea, DRL R. M. Diggs, DRL Nary Jinks (2)
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References 1,
DRL approval to operate the' Big Rock Point reactor with six centermelt fuel bundles in the core.
' Amendment No. I to Facility Operating' License No. DPR-6 dated Nbrch 12, 1968.
Page 3_ -' Comparison with " Type C" Fuel.
a.
2.
Consumers Power Company Proposed Change No.13. dated May 26, 1967.
a.
. Figure 14 Individual Fuel Rod Relative Power at b.
. Figure 15 Beginning of Life c._
Page 8 - Centermelt fuel exposure.
- 3. ~Special Report No. 10 dated April 7, 1969
" Performance Evaluation of Centermelt Fuel af ter the First Period of Irradiation in the Big Rock Point Reactor":
Figure 2 - Page_16 -' Rod Local Power Factors
'a.
4.
Consumers Power Company Proposed Change _No. 24 dated December 21, 1970--
"Information on New Centermelt Fuel Assemblies":
a.
Figure 1, Rod Power Factors, b.
Table 1, Enrichment Distribution.
c.
Table 5, Thermal-Hydraulics at 122% Power Level.
d.
Bundle Average Exposure - reactivity and rod power, page 14.
e.
Reduction in crud deposition, page 4.
f.
Centermelt fuel assembly reactivity, page 14.
n g.
MCHFR, page 21.
h.
Figure 4.
k.
Cause of clad failure, page 23.
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Centermelt bundle power, page 17.
5.
Consumers Power Company - Answers to DRL questions - dated August 15, 1967.
a.
Table 1, MCHFRs, page 10.
6.
Special Report - February 1971
" Failure Analysis and Performance Eval-uation of Intermediate Performance Centermelt Fuel after 10,000 mwd /T Exposure":
a.
Page 4, no hydriding of the cladding at the inside surfaces.
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yAR 319A Docket bo. 50-155 Consumers Power Company ATIN.
Mr. Gerald J. Walke Nuclear Fuel Management Adminis tra tor 212 West Michigan Avenue Jackson, Michigan 49201 Change No. 24 Centlemen:
License No. DPR-6 We have reviewed your Proposed Change No. 24 dated December 21, 1970, to the Technical Specifications of Facility License No. DPR-6 for reactor operation with two new cantarmelt fuel bundles and five of the original six centeracit fuel bundles.
Tne new centermelt fuel bundles are different from the original center-melt assemblies in the following respects:
- 1) sheet metal corner angles have been eliminated, 2) removable cobalt targets have been placed in new stainless steel corner tubes, 3) there are fewer (16 compared with 29 and 36) high power fuel rods, and 4) rod-to-rod power Eradients have been reduced.
Before the five original centermelt fuel assemblies that were irradiated during April, May and June 1967 are returned to the Big Rock Point Nuclear Reactor, the crud accumulated during timt irradiation period will be removed by chemical cleaning and eight of the hign power rods in the outer rows will be replaced by low power rods.
With two new centermelt fuel bundles in the core, there will be a total of 32 high power rods at or near centermelt conditions at rated power. When the five original centermelt fuel assemblies have been reconstituted and rainserted into the Big Rock Point core, the total number of high power rods will be 172 compared with 188 for the six centermelt fuel bundles as originally fabricated and irradiated in the ?tg Rock Point core in March 1967.
5' We have concluded that the proposed change does not preaant significant hazards con <,iderations not described or implicit in Consumers' Safety Analysis Report and Proposed Change No.13 dated May 26, 1967, and approved by DRL Amendment No. I to the Facility Operating License No. DPR-6 dated March 12, 1968. There is reasonable assurance that the health and safety of the public will not be endangered by operation of the B1 i
T/NnotJ,s6 _ Rock Point Nuclear Reactor in the manner described by Consumtrs j
. M S.F 3 1971 Consumers Power Company Power Company with two new centermelt fuel bundles or with the two new centermelt fuel bundles and five of the original centermelt fuel bundles-as proposed.
Accordingly, pursuant to Section 50.59 of 10 CFR Part 50, the Technical Specifications of Facility License No. DPR-6 are hereby changed as indicated in Attachment A to this letter.
Sincerely, Pecer A. Morris, Director Division of Reactor Licensing Enclosure.
Attachnent A - Changes to Technical Specifications George F. Trowbridge, Esquire i-cc.
Distribution W. Dooly, DR ACRS (3)
R. Engelken, CO (2)
R. DeYoung, DRL H. Shapar, OGC R. S. Boyd, DRL N. Dube, DRL (5)
D. J. Skovholt, DRL J. R. Buchanan, ORNL R. H. Vollmer, DRL T. W. Laughlin, DTIE D. L. Eiemann, DRL Document Room J. J. Shea, DRL beocket File R. M. Diggs, DRL DR Reading DRL Reading Branch Reading QN e
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