ML20059L113

Forwards Copy of GE Rept Providing Update Re GE Experience W/Bwr Fuel Through Dec 1989

ML20059L113
Person / Time
Site:	05000605
Issue date:	03/06/1991
From:	Charnley J GENERAL ELECTRIC CO.
To:	Rosalyn Jones Office of Nuclear Reactor Regulation
Shared Package
ML20058L685	List: LD-90-093, Forwards Info Re Performance of CE Fuel During 1989 ML20058L681 ML20059L113 ML20059L129 ML20059L134 ML20059L142 ML20059L162 ML20059L179
References
JSC-91-005, JSC-91-5, MFN-026-91, MFN-26-91, NUDOCS 9402030185
Download: ML20059L113 (16)

Text

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O GE EVuclear Energy

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March 6,1991 MFN 026-91 JSC 91-005 US Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Mail Station F1-137 Washington, DC 20555 ATTENTION:

R. C. Jones, Jr.

Chief Reactor Systems Branch

SUBJECT:

EXPERIENCE WITil BWR FUEL T11ROUGil DECEMBER 1989 Gentlemen:

Enclosed is a copy of the GE report providing an update of GE's experience with BWR fuel through December 1989. It is being sent to you at your request for use in the preparation of your annual fuel performance report.

Please contact Gary Jones on (408) 925-1516 if you have any questions.

Sincerely.g n nr #

J. S. Charnley, Manager Fuel Licensing (408) 925-3697 M/C 188 Enclosure JSC:jz CC:

L S. Gifford G. G. Jones S. Wu (NRC) 7402030185 911025 Pla ADOCK 05000605 A

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GE EXPERIENCE WITH BWR FUEL THROUGH DECEMBER 1989 P-W E

5 1

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4 I.

Introduction This information report provides an updated review of GE experience with production and developmental BWR Zircaloy-clad UO fuel rods through December 1989. This experience 2

includes successful commercial reactor operation of fuel bundles to greater than 45,000 mwd /MTU bundle average exposure (approximately 60,000 mwd /MTU peak pellet expo-sure).

The performance of GE 8X8 fuel types continues to be highly successful as demonstrated by an overall fuel rod reliability rate from 1974 to the end of 1989 of greater than 99.98%.

II.

GE BWR Fuel Experience Base As of December 31,1989, over 3.8 million GE 8X8 fuel type production Zircaloy-clad UO2 fuel rods were in, or had completed, operation in commercial BWRs. Figure 1 shows cumu-lative fuel rods in GE 8X8 fuel bundles loaded as a function of calendar year. As of Decem-ber 31,1989, over 1.54 million GE fuel rods were in operation. Figure 2 illustrates GE's core loadings at the end of 1989 by fuel type. As of December 31,1989, GE had loaded approxi-mately 1.17 million pellet-clad interaction (PCI) resistant barrier fuel rods in commercial BWR's. The GE fuel manufactunng facility in Wilmmgton, North Carolina, is producing 100% of its 1990 load as barrier fuel, demonstrating the overall customer acceptance of this fuel design.

In 1989, eighteen domestic and eight overseas GE BWR plants containing GE fuel had re-fueling outages with over 3800 new GE 8X8 fuel bundles loaded. Over 50% (or 12 reloads)-

of this new fuel loaded was GE's latest production fuel design (GE8X8EB and GE8X8NB),

including the introduction of GE8X8NB in a commercial reload.

III.

In-Reactor Surveillance Programs and Summhry of Surveillance Results One of the most important aspects of the GE fuel design process is the in-reactor perform-ance monitoring of a design before and after its introduction. In keeping with the GE philos-ophy of test-beforesise, lead use assemblies (LUAs) containing selected key design features are used to demonstrate the satisfactory performance of these features and to provide lead experience for future production fuel. The fuel surveillance program adopted by GE and ac-cepted by the NRC is described in References I through 4.

A summary of GE's LUA surveillance program is contained in Table 1. Examination results are pmvided below:

A.

Barrier Fuel Procram The goal of this program was the demonstration of a Pellet-Cladding Interaction (PCI) resistant fuel under conditions which would provide statistically significant results. The PCI resistant fuel features the barrier concept to protect the fuel cladding from failure caused by PCI. The barrier fuel program consisted of four LUAs, loaded into Quad Ci-ties-1 over 10 years ago in 1979 at the beginning of cycle 5, and a demonstration reload 2

of 144 bundles with Z -lined cladding placed into the core at Quad Citie>-2 in 1981 at the beginning of cycle 6.

The barrier LUA's at Quad Cities-1 operated for 4 or 5 cycles and underwent four pool-side examinations consisting of visual mspections and non-destructive testing of se-lected fuel rods. These examinations revealed that the bundles and individual fuel rods exhibited characteristics typical of nonnal operation.

The Quad Cities-2 barrier fuel program was designed to subject the barrier cladding fuel to significant power increases in order to demonstrate the PCI resistance of barrier fuel.

Two power increase demonstrations were performed; the first in 1983 at the end of cycle 6 and the second in 1985 at the end of cycle 7. Sixteen barrier bundles were involved in each demonstration. During the following plant outage, all demonstration barrier bundles were evaluated by vacuum offgas sipping and determined to be sound. Subse-quent to the power increase demonstrations, all PCIOMR operating restrictions were re-moved from the barrier fuel bundles in the core. Plant offgas surveillance indicates that all fuel bundles in the core continue to operate reliably. Of the 144 bundles in the reload,,

32 operated for 3 cycles,80 operated for 4 cycles, and the remaining 32 bundles are operating in their 5th cycle.

B.

Improved Desien Feature Lead Use Assemblies Several LUAs have been.fesigned and placed in operation for the purpose of obtaining experience and performance data on new product design features. Rese LUAs have un-dergone extensive preirradiation characterization, with plans for interim poolside exami-nations. These Improved Design Feature LUAs include:

1.

1983 Lead Use Assemblies Four LUAs were loaded into Peach Bottom-3 in 1983 at the beginning of cycle 6.

Design features tested include improved spacer and upper tie plate, axial zoning of gadolinia, cladding thickness, pellet dimensions, and fuel rod helium prepressunza-tion. The first poolside examination of these bundles was completed in August 1985 after one cycle of operation and showed characteristics typical of normal operation.

The second poolside examination was completed in November 1987 after two cycles of operation and showed characteristics typical of two cycles of normal oper-ation. Peach Bottom-3 just recently retumed to service (December 1989) since its refueling outage at the end of cycle 7 in April 1987.

2.

1984 Lead Use Assemblies Five LUAs were loaded into Duane Amold in 1985 at the beginmng of cycle 8.

Features tested include water rod configuration, impmved spacer and upper tie plate, cladding surface treatment, axial zoning of gadolinia, fuel rod helium prepres-surization, pellet dimensions, and pellet density. The first poolside examination of these bundles was completed in April 1987, after one cycle of operation, and 3

showed characteristics typical of normal operation. The second poolside examina-tion was completed in October 1988, after two cycles of operation, and showed charactenstics typical of two cycles of nonnal operation. The next poolside exami-nation is scheduled in 1990 after the third cycle of operation.

3.

1987 Lead Use Assemblies Four LUAs were loaded into Hatch-1 in 1987 at the beginning of cycle 11. These fuel assemblies represent lead use GE8X8NB production fuel. The first poolside examination of these bundles was completed in October 1988 after one cycle of operation and showed characteristics typical of normal operation with no evidence of Crud-Induced Localized Corrosion (CILC). The next poolside examination of these bundles is scheduled in 1990 after the second cycle of operation.

4.

Cladding Corrosion Performance Lead Use Assemblies Six LUAs were loaded into Hatch-2 in early 1988 at the beginning of cycle 8 and six LUAs were loaded in Hatch-1 in late 1988 at the beginning of cycle 12. Fe,a-tures tested include cladding material, heat treatment, and surface conditLaing.

These two reactors have historically exhibited highly variable cladding corrosion performance, even for cladding material taken from the same tubing lot but irra-diated in the two different reactors. In late 1989, three LUAs that included tubing teing the most recent corrosion improvement processes were examined (after one cyce of operr. tion) from the Hatch-2 reactor. These LUAs reflected bundle average expoures up to 13,000 mwd /MTU Visual inspection of several improved GE tub-ing processes revealed little or no visible nodular corrosion along the full length of the fuel rods. The next poolside examination of the Hatch-2 bundles is scheduled in 1991 after the second cycle of operation, while the first poolside examination of the Hatch-1 bundles is scheduled in 1990.

5.

GE8X8NB-1 Channel Lead Use Assemblies Four LUAs were loaded into Cooper in 1988 at the beginning of cycle 12. 'Ihese LUAs represent lead use of GE8X8NB-1 production fuel bundle design features.

The first poolside examination of these bundles was completed in April 1989 after one cycle of operation and showed characteristics typical of normal operation. The next poolside examination of these bundles is scheduled in 1990 after the second cycle of operation.

6.

1987 Lead Use Assemblies Four LUAs were loaded into Peach Bottom-2 in 1989 at the beginning of cycle 8.

These fuel assemblies represent lead use GE8X8NB production fuel. The first pool-side examination of these bundles is scheduled in 1991 after the first cycle of opera-tion.

IV.

Generic Fuel Performance Mechanisms 4

Pellet-cladding interaction (PCI) and crud-induced localized corrosion (CILC) are the only cladding perforation mechanisms that have affected fuel performance in recent periods. As described below, product improvements have been developed that wili essentially eliminate these two fuel rod failure mechanisms.

A.

Pellet-Cladding Interaction Light water reactor (LWR) nuclear fuel is susceptible to fuel md cladding perforation, commonly called PCI failure, when subjected to fast power increases at moderate to high exposures. Operational procedures which involve slow approaches to power have essentially, but not completely, eliminated PCI failures in LWRs but at the cost of reac-tor capacity factor loeses. Zirconium barrier fuel was invented by GE as a material solu-tion to the PCI failure problem. Extensive test reactor and laboratory tests along with successful in-< ore power ramp demonstrations in the Quad Cities Unit 2 power reactor have shown that Zr-barrier fuel is convincingly failure resistant. Barrier fuel was com-mercially introduced by GE in 1983. The Zr-barrier fuel commercial experience funher confirms the effectiveness of this fuel design concept with not a single PCI induced Zr-

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barrier fuel rod failure in greater than 680,000 barrier fuel rods completing at least one reactor cycle of operation. PCI failures are expected to be elmunated within the next few years as the population of non-barrier fuel (34% of all GE fuel currently in operation is non-barrier) is discharged.

B.

Crud-Induced LocaH7ed Corrosion In 1979, an unexpected low-level failure mecharusm oflocalized fuel rod cladding cor-rosion was revealed in some BWRs. Poolside examination of the failed fuel rods re-vealed plant corrosion product (crud) scale deposits with high copper concentrations.

The natum of the failures led to identification of special conditions of environment, op-erational history, and material-susceptibility that must occur simultaneously to cause failure. These crud-induced localized corrosion (CILC) failures have been limited to plants with copper alloy condenser tubes and filter deminerahzer condensate cleanup systems.

Fuel examinations, surveillance, and extensive research have led to a practical under-standing of this mechanism A reproducible out-of-reactor test for measunng the sus-ceptibility of Zircaloy to in-reactor nodular corrosion was developed by GE and corre-lated to in-reactor performance (Reference 5). This test confhmed a previously undetected variability in the susceptibility of Zircaloy to in-reactor nodular corrosion.

This test has been patented and made available to the industry on a non-profit basis through the ASTM.

Manufacturmg processes have been developed that both improve the corrosion resis-tance of the incoming material produced by the Zircaloy vendors and funher ensure that R

improved corrosion resistance is maintained thmughout the fabrication processing to yield final size fuel rod cladding that is more resistant to in-reactor nodular corrosion.

5

4 i

These processes have been implemented in the production of all GE fuel to provide a high degree of assurance that adequate corrosion resistant properties are achieved.

V.

Conclusions GE has developed a substantial fuel experience base that, coupled with an aggressive fuel surveillance program, has provided significant feedback on statistically significant numbers of fuel rods with regard to the performance effectiveness of design, operational and manufac-turing changes. It is concluded that the experience gained with GE production and develop-mental fuel continues to demonstrate the high reliability of the GE designed BWP fuel.

VI. References 1.

J. S. Chamley (GE) to C. H. Berlinger (NRC), " Post Irradiation Fuel Surveillance Pro-gram" November 23,1983.

2.

J. S. Charnley (GE) to L. S. Rubenstein (NRC), " Fuel Stuveillance Program" February 29,1985.

3.

J. S. Chamley (GE) to L. S. Rubenstein (NRC), " Additional Details Regarding Fuel Sur-veillance Program", May 25,1984.

4.

L. S. Rubenstein (NRC) to R. L. Gridley (GE), " Acceptance of GE Proposed Fuel Sur-veillance Program", June 27,1984.

5.

B. Cheng H. A. Levin, R. B. Adamson, M. O. Marlowe, V. L. Monroe, " Development of a Sensitive and Reproducible Steam Test for Zircaloy Nodular Corrosion", ASTM 7th Intemational Conference on Zirconium in the Nuclear Industry, Strasbourg, France, June 24-27,1985.

l l

6

i i

Table 1 Summary of Ongoing Lead Use Assembly Surveillance Programs Bundle Number of Average Number Completed Exposure At of Cycles of Last Outage Program Reactor Bundles Operation (GWd/MTU)

Objectives Banier LUA's Quad Cities-1 1

5 43 Barrier Cladding 1983 LUA's Peach Bottom-3 4

2 24 Improved design features 1984 LUA's Duane Amold 5

2 28 Improved design features 1987 LUNs Hatch-1 4

1 12 Lead Use GE8X8NB Corrosion Hatch-2 6

1 13 Clad Mat'l Performance Process Variables GE8X8NB-1 Cooper 4

1 8

Lead Use Channel GE8X8NB-1 LUA's Features Corrosion Hatch-1 6

Clad Mat'l Performance Process Variables 1987 LUNs Peach Bottom-2 4

Lead Use GE8X8NB 7

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ADVANCED NUCLEAR FUELS CORPORATION m ononu w o u,oso rouca rau.nzna uo. u sn w osso skwsj 31b c M illt X lb 2075 RAC:072:89 October 2, 1989 i

Chief Reactor Systems Branch Division of Engineering and System Technology.

' Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555

Subject:

ANF Annual Fuel Performance Renort

Reference:

Letter, M.

W.

Hodges (USNRC) to B.

Copeland (ANF), " Fuel Performance Annual Report",

May 9, 1989.

Attached is the Advanced Nuclear Fuels Annual Fuel Performance Report for 1988 as requested in the referen.ed letter.

Please contact.me if there are questions or if further information is needed.

Sincerely, hY R. A. Copeland Manager, Reload Licensing cc:

Dr. S. L. Wu (USNRC)

I I

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f ADVANCED NUCLEAR FUELS 1988 FUEL PERFORMANCE 1

As of December 31,-1988 ANF fuel had -been loaded into 43 comme light water reactors in the United States Europe, and Asia, including rcial and 21 PWRs.

ANF fuel has also been supplied to the LOFT test reactor s.

By the end of ' 1988, a total of 14,128 fuel assemblies comprisi 1,729,592 fuel rods had been irradiated.

Of these, 8,616 assemblies were irradiated.in BWRs and 5,512 assemblies were irradiated in PWRs.

ANF experience is summarized in Table 1.

burnup is shown in Figure 1.

The distribution of ANF fuel versus DWRs and seven PWRs during the year.New reloads were placed into serv t

The highest exposures reached by ANF fuel to date are 50.0 CWD/

Tihange-1 PWR in Belgium, and 41.1 CVD/HTU at the Big Rock Point-BWR.

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fuel rod integrity remained better than 99.994%.

Nif" In 1988, the failure rate from fuel related and unknown causes was 0.0007% in BW s.-

ANF recently adopted use of the INP0 Fuel Reliability Indicator (F a general measure of fuel performance.

s The FRI for PWRs -is the iodine-131 coolant activity level normalized to a standard cleanup system flow rate corrected for tramp uranium.

Lower FRI values are indicative of fewer failed rods in the core.

The FRI distribution for ANF fuel is shown in Figure 2 and is derived from the 1988 yearly average for each reactor that Nif fuel in the core.

operated with is 1.02 x 10~4 The median value for all reactors. containing ANF fuel This compares well with the overall INP0 median of 4.8 x 10'3 pci/gm.

The five year trend in the ANF FRI indicates a continued improveme in fuel performance.

i Data have not been compiled to establish a fission gas-related FRI f BWRs.

However, the failure rate from fuel-related and unknown causes is low or.

for BWRs than for PWRs operating with ANF fuel.

er BWR 9x9 fuel assemblies and PWR 17x17 assemblies reached burnups during 1988.

BWR 9x9 fuel and PWR 17x17 fuel reached assembly averag

more resistant to failure because of reduced linear heat generation rates.

Lower.' fuel temperatures, less fission gas release, decreased pellet-clad interaction,-and lower clad stresses are the benefits.

Leaks in cladding attributable to causes other than fuel design or manufacturing, when they could, be adequately examined, were found - to. be accompanied by an abnormal operational condition.

Two such phenomena that were ' identified as the cause of failures were:

(1) damage (from handling) to q

a critical part of a fuel assembly, e.g. a spacer, with resultant fretting, or (2) trapping or lodging of a piece of debris from the coolant stream where it could cause fretting of the cladding.

In 1988, fuel rod failure rates from such causes were zero for BWRs and 0.0025% for PWRs.

Afif cladding continued to show good corrosion performance in all reacto'r environments based on corrosion data collected during 1988.

These data'were obtained at four PWRs and five BWRs including one reactor characterized as,

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CILC susceptible.

Severe fretting wear was discovered on the lower end cap shanks of spacer capture rods (SCRs) in one European BWR during 1988. Although 127 fuel assemblies were affected, no fuel failures' occurred.

By the end of 1988, all available fuel had been repaired and returned-to service.

Repair activities.

continued in 1989 for fuel that was in the reactor at the end of 1988.

fuel examinations at other reactors, operating with essentially the same -Mif fuel design, have demonstrated that the problem is limited to the one reactor.

The fretting cccurred due to specific flow conditions present in this BWR.

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' TABLE 1.

SUMMARY

OF.ANF FUEL EXPERIDiCE THROUCH 12/31/88-A.

FUEL ASSEMBLIES 4

in Core Discharned Reactor Max. Burnup Max. Burnup Total f

Tyne Otra n t i t y CWD/MTU 1

Ouantity CWD/HTV__

Otrantity

-BWR 6468 37.4 2148 41.1 1 8616 PWR 7007 43.3 3505 50.0 5517 Total 8475 5653 14128 t

B.

TUEL RODS l

Reactor Tvon in Cnre Discharand-Total.

BWR 436,883 131,926 568,809 PWR 44?.671 708.11?

1.150.783 i

Total 879,554 840,038 1,719,592 7

I Avera9e of extended burnup rods transferred to a new host fuel assembly.

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3000 E 3 PWR ASSEMBUES 2500 --

EZ BWR ASSCUBUES 6

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FlCURE 1 Distributton of Irrodloted Advanced Nuclear Tuol by Exposure Throu0h the End of 1988

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Annual Distribution for PWR's 1988

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