ML19254E830
| ML19254E830 | |
| Person / Time | |
|---|---|
| Issue date: | 10/16/1979 |
| From: | Meyer R Office of Nuclear Reactor Regulation |
| To: | Drey L AFFILIATION NOT ASSIGNED |
| References | |
| NRC-1061, NRC-1070, NUDOCS 7911020372 | |
| Download: ML19254E830 (11) | |
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NUCLEAR REGULATORY COMMISSION "g
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OCT i n igjf Mrs. Leo Drey 515 West Point Avenue University City, M0 63130
Dear Mrs. Orey:
Your letter to Dr. Picklesimer has been referred to us for a reply.
I have spoken to Dr. Picklesimer about his statement that "the maximum amount of failed fuel observed at any time in operating plants is much less than 0.1% by actual operating records and experience...." That statement, as you have discovered, is ambiguous, and one of its reason-able interpretations is incorrect. Dr. Picklesimer meant to say that on any given day in recent years, plant observations would indicate the p5ence of much less than 0.1% failed Zircaloy fuel rods.
Enclosed is Table 1-2 from a recent Westinghouse report, WCAP-8183, Revision 8, " Operational Experience With Westinghouse Cores- (up to December 31, 1978)," April 1979. This table reports a percentage of design-basis coolant activity. Since design-basis coolant activity corresponds to about 1", failed fuel, you can get the estimated cer-centage of fuel failures by dividing the tabulated numbers by 100.
Notice that the 0.4% and 0.7" values you quoted for Ginna and Beznau-l can in'eed be inferred from this table. Those particular faiTures were d
the result of a fabrication problem (excessive moisture) that was rectified in 1970. After that early fuel was discharged, you can see the improvements in activity levels.
Figure 5-3 frem that same report (also enclosed) shcws coolant levels from 1974 to present to be about 0.03 Ci/gm, which is equivalent to acproximately 0.02", failed fuel. This failure experience, as Dr. Picklesimer was trying to say, is much less than 0.1". failed fuel.
Fuel failure experience from the other reactor fuel suppliers is also available. Recent summaries of such infomation can be found in the American Nuclear Society conference proceedings frcm the topical meeting on Light Water Reactor Fuel Performance held in Portland on April 29 to May 3, 1979.
I have enclosed relevant pages from that document. Although we watch connerical fuel perfomance closely, the NRC does not routinely comoile fuel failure statistics since that information is available from the fuel suopliers.
1249 203 7 9110 2 03 7L
Mrs. Leo Drey 2
With regard to actual releases of radioactive materials from nuclear power plants, it is theoretically possible for noble gas releases to be as low as reported in NUREG-0077. That would have nothing to do with fuel rod design but rather with waste-gas decay tank capacity and primary system gas leaks. For a new plant with an empty storage tank and tight seals, low releases are cossible and the quoted releases are plausible.
I have not, however, investigated the cases you mentioned to check their accuracy because I believe the thrust of your question had to do with the fuel design.
Sincerely, Ralph 0. Meyer, Leader Raactor Fuels Section Core Performance Branch Division of Systems Safety cc:
M. L. Picklesimer F. Schroeder 1249 204 t
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TABLE 5-2 (Cont)
C00EANT ACTIVITY LEVEL ^
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Percentage of Design Basis Coolant Activity in Indicated Year, Quarter 1978 I
i PiANT 1
2 3
4 c
JOSE CABRERA (UEM) 0.68 0.30 2.98 3.22 c
c i
GINNA (RGE) 1.62 0.90 1.29 1.14 c
BEZNAU 1 (NOK) 1.29 1.25 0.76 0.98 l
C BEINAU 2 (NBK) 0.35 0.98 0.19 0.20 4
i MillAM A 1 (KEP) b b
b b
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O POINT BEACl! 1 (WEP) 1.05 0.14 1.45 0.86 c
c MillAMA 2 (MEP) b b
b b
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POINT BEACil 2 (WIS) 0.14 0.27 0.22 0.26 C
SilitRY l (VPA) 0.74 0.47 0.09 0.46 g
TURKEY POINT 3 (FPL) 0.50 '
O.37 0.24 0.33 SURRY 2 (VIR) 0.02 0.01 0.01 0.02 c
C ZION 1 (CWE) 1.20 1.19 1.10 0.42 l
b INDIAN PGINI 2 (IPP) 0.73 0.45 0.61 0.25 c
c IURKEY POINT 4 (FLA) 0.40 0.37 0.15 '
O.23 c PRAIRIE ISLAND 1 (NSP) 0.03 '
O.14 0.68 1.98 I'
C rv c
ca ZION 2 (COM) 0.32 0.47 c 0.29 0.46 i
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TAKAllAMA 1 (TAK) b b
b b
c KEWAUNEE (WPS) 0.08 0.12
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RINGilALS 2 (SSP) b b
b 0.29 PRAIRIE ISLAND 2 (NRP) 0.03 0.05 0.04 0.05 c
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,I TABLE S-2 (Cont; a
COOLANT ACTIVITY LEVEL jli Percentage of Design Basis toolant Activity in Indicated Year, Quarter 1978 1
2 3
4 PiANT c
D.
C.~ COOK 1 (AEP) 0.36 0.47 0.04 0.04 TROJAN (POR) 0.33 b
b b
c INDiff FolNT 3 (INT) 0.41 0.26
<0.01 0.11 llEAVER VALLEY.1 (DLW) 0.48 0.32 0.04 b
SALEM 1 (PSE)
<0.01 b
<0.01
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'l' KORI 1 (KOR) b b
b b
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J. M. FARLEY l (ALA) 0.03 0.02 0.01 0.05 0111 1 (OHI) b b
O. C. COOK 2 (AMP)
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NORTil ANNA 1 (VRA) 2.31 1.86 b
Olli 2 (0KB)
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l A declisie in the coolant activity level, due to any cause' other than a.
discharge of affected fuel at normal refuejing, is a reflection of r0 decreased release of fission product iodine.
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PERFORMANCE RECORD i
Average Coolant Activity Number of Rods l
Level (# CI/gm) in Service 5
1,000,000 I'4 900,000 lf 800,000
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1.2 mE
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EP 600,000
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j 400,000 0.6 0.4 300,000 if a
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70'71'72' 73'74 75'76' 77I I
I 78 Year FIGURE 5-3 PERFORMANCE RECORD - AVERAGE CCCLANT ACTIVITY LEVEL /NUMSER OF ROCS IN SERVICE g
1249 209 5-15 g
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being demonstrated in a SWR /3 '.
Thi s a -. :n ; g c. : : ;;;c 3r, en c rrc.
cperating strategy in which the control rod pattcr-i; 'ti.c: tr m y; the cycle so that control rod sequence exchange is eliminatcd.
Control rod move-ment to offset reactivity changes during power operstf on is limited to this fixed group of control rods.
In a Control Cell Core these control blades are 1-surrounded by low power fuel assemblies. Then, during normal plan' operation, the effect of power changes on fuel performance caused by control rod motion is t
minimized since only low power assemblies experience power shifts associated i
with control blade motion.
This design concept simplifies core operation and
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equally important has the potential to reduce fuel duty and increase thermal margins and capacity factors.
SUMMARY
OF FUEL PERFORMANCE The General Electric fuel experience encompasses a range of fuel designs from the old 7x7 fuel through the current 8x3R design. By mid-1979 GE will have produced and put into operation approximately 29,000 fuel assemblies.
This represents over 1,600,000 fuel rods with peak local exposures up to 40,000 mwd /t. Table IV summarizes the performance data from these different fuel types. These data clearly illustrate the effect of design, process and operating changes introduced over the last five years on fuel performance.
This extensive operating experience provides a sufficiently large data base to statistically verify the improved performance of the new fuel designs.
TABLE IV GE BWR/2-5 Fuel Experience Summary 7x7 7x7R 8x8 8x8R Cumulative Fuel Assemblies Loaded 10,289 5824 10731 1898 Assemblies Sipped at least Once 10,289 5793 5698 7
1,ead Sipped Batch Exposure (GWd/t) 25.6 22.3 22.9 12.5 Estimated Rod Failure Rate 0.98%
0.043%
' O.028%
0 References R. A. Proebstle, W.E. Saily and H.H. Klepfer, Cattent T,tenda in EuR FucL 1
Petfasance, ANS Topical Meeting - Water Reactor Fuel Performance, St. Charles, Illinois, May 9-11,1977.
0.0. Sneppard, Sailing ' aut Reach.t Fuel Red Pe.tfamance Evaintion s
P. tog.t:m, NEDC-24609, February 1979.
S.R. Specker, L.E. Fennern, R.E. Brown, R.L. Crowther and K.L. Stark, 3
Cantial Cell. Cc,te Imptoved Ocaign, Presented at American Nuclear Society Meeting, November 12-16, 1978, Washington, D.C.
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1249 210 3
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-d When combined with the lack of prepressurization, the fuel operated at higher temperatures than does =odern fuel at equivalent power. Since 1974, all of C-E's fuel has been prepressurized and has incorporated other design changes.
3 The pellet shape was changed to add chamfers and to reduce the length-to-diameter ratio from 1.7 to 1.2.
The pellet density was increased to 95.0::
E T.D., and tha fuel fabrication process was changed to reduce in-reactor j
densification. The wall thickness of the cladding was increased from 0.026 i
to 0.028 in. for added conservatism.
d W
The result of the above design changes can be evaluated by examining Figure 1.
Each of the operating plants are monitored frequently to estimate M
current defect levels. The sum of the highest defect levels for each plant f
has been plotted by year in Figure 1 in terms of the number of defective fuel J;
rods per each 10,000 rods in operation. The i=provement shown in the reli-ability level begins with the changeover to the current C-E design following the Maine Yankee experience. Some carryover of earlier fuel is evident from y+
fuel fabricated before 1974 and operating beyond that point.
The continuous improvement over the last five years from 3 defects in 10,000 rods to less than 1 defect in 10,000 rods is evidence of excellent fuel performance.
C-E's irradiation test progra=s are dedicated to maintaining this high level
)d 4
of reliability in light of the utilities' desires to increase: fuel maneuver-
-3t*
ability; reload fuel design flexibility; and batch-average discharge burnups.
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l S
EARLY DESIGN CURRENT DESIGN l
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<1 0.5 1971 1972 1973 197da 1975 1976 1977 1978 1979 m
0 TIME, YEARS M
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a MAINE YANKEi CORE 1 N
b AS CF FEBRUMY 1,1979 (19/225, CCC) 3 4
Fig. 1.
Hist - :- Of C-E Fuel Perfor=ance u
amus.
1249 211 i
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During an October 1978 outage at the Yankee Rowe reactor, a sipping pro-gram was conducted on forty (40) ENC 16 x 16 fu.1 ssse=blies that scrt being g
routinely discharged.
Peak asse=bly burnup exceeded 31,000 MWD /MTU.
The g}:
sipping program, conducted by Yankee Atomic personnel, identified four suspect
~
b assemblies.
6M In November 1978, a fuel examination was conducted by ENC personnel on 13 assemblies.
The of the discharged assemblies, including the four suspect examination consisted of underwater TV inspection with video tape recording.
Several of the assemblies were disassembled for = ore detailed observation of the interior rods.
Visible evidence of failure was detected in two of the four suspect assemblies.
In each case two peripheral fuel rods had suffered through-wall f retting wear in the vicinity of the top two spacer grids. The Yankee Rowe fuel assembly design has a unique offset corner on each of two sides.
All four failed rods were located in the vicinity of this offset corner location.
Review of the operating histories of these asse=blies shows that they had both i
been positioned at the periphery of the core during their first operating cycle and were then moved to the center of the core.
Furthermore, the side of each assembly where the f ailed rods were found had been the side adjacent to the core baffle during their first operating cycle. Although several of the other examined assemblies had also been located on the core periphery during the first irradiation cycle, no evidence of failure could be located in any of,
these fuel assemblies.
At this time the mechanism responsible for the f ailed fuel rods in Yankee Rowe is not fully understood. There was so=e evidence of mechanical da= age to the. f ailed assemblies which may have occurred af ter f abrication but before reactor irradiation. The observed f ailures are similar to fretting failures in other FWRs where the cause of fretting was traced to a coolant crossflow through a defective core baffle.
In these other cases; the baffle was of bolted construction and gaps had opened up at the baf fli joints.
In the Yankee Sowe reactor the baf fle is of welded construction and would seem less likely to have allowed significant crossflow. During the next scheduled Yankee Rowe outage, fuel assemblies which had occupied the ' ame core peri-s pheral positions as the discharged failed assembliks will be examined to determine whether coolant crossflow at those specific core locations sight be the cause of the failures observed.
1 Pri=ary coolant activity in other FWHs containing ENC fuel has been consistently below 0.2 pCi/cc, indicative of acceptable fuel perfor=ance.
CONCLUSIONS ENC has an active and on-going non-destructive program that is being carried out on both dc=estic and foreign FWRs and BWRs. Peak assembly burnups of 30,'.00 MWD /EU in a BWR and 32,700 MWD /EU in a PWR have been achieved.
Of the more than 200,000 ENC fuel rods which have been irradiated since 1971, only 11 individual rod failures have been identified.
Seven of these failures occurred in a 7 x 7 BWR design which has been discontinued.
Failures in the four PWR rods do not appear to have been related to an intrinsic design character.<*1c but rather to mechanical or hydraulic damage incurred prior to or during irradiation.
1249 212 y
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