ML18064A468

Paper Entitled, Reactor Vessel Neutron Fluence Reduction Measures Taken at Palisades Presented at ANS Annual Meeting on 940619-23

ML18064A468
Person / Time
Site:	Palisades
Issue date:	06/19/1994
From:	Goralski G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 9411170177
Download: ML18064A468 (10)

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J' 2 '94 l 61 34 FROM CPCO PnLJSnOES SUPPOR PnGE.0'33 REACTOR VESSEL NEUTRON FLUENCE REDUCTION MEASURES TAKEN AT PALISADES GH6oralsk1 Consumers Power Company ANS ~nnual Heet1ng June 19-23, 1994

1.0 INTRODUCTION

The Palisades Nuclear Plant has been implementing a low leakage core design begipning with Cycle 8 to address the pressurized thermal shock (PTS) rule 1

• The fast neutron flux that rP.ar.he~ the rP.ar.tnr ve~~P.1 ha~

been sign1f1cantly reduced through fuel management. These flux roduction moasuros havo dohyod tho data for roaching tho PTS screening criteria, however, fuel management alone has not yet been successful in extending the PTS screening criteria date beyond the present end of operating licensed life, (EOL). This paper describes the fluence reduction moasuros that havo boon takon to dato at Palisados.

2.0 FLUENCE REDUCTION MEASURES TAKEN THROU&H FUEL MANA&EMENT 2.1

Background:

Cycles l through 7 (5-24-71 through 8-8-88)

The Palisades tore contains 204 fuel bundles. A three-batch fuel management plan was utilized for cycles l through 7 with one third of.

the core (68 assemblies) typically exchanged each refueling outage. One exception to this was Cyc1e 2 when all reload fuel was new. The original Cycle 1 discharged fuel was never reused due to growth c;onsiderations not being adequately co~ered by the assembly destgn*. New reload fuel for Cycles 3 through 7 was placed 1n all 60 core periphery locations. Eight of the new assemblies and the remaining reload ruel was positioned using a mixed central zone fuel management plan. This ru~l llldlltlyelllt!nt appt*oac.h minimize$ power peaking and generally provide~

the greatest thermal margin, but results in the max1mlB11 overall core llt~UtfUll 1~11kc1yt! carnJ rot ll~1Alr*u11 r1uA Lu Un: f't=ddur v~~s~l.

Tf1~ dVtfftlY~ \;yi.1~ ltm~tf1 rur* Cy\.:1~s 1 lhruu~h 7 WdS 372 ttff!t:L fvt! full power days (EFPO). Figure 1 shows how the key 48 core per;phery lui.;c&Lium; wtil~h dff~d rHt:Lur* vtisstil fuL neutron exposure were 1oaded with new fuel for Cycles 1 through 7.

( 9411170177 940619 f

' I

NOJ 2 '94 16:34 FROM CPCO PnLJSnDES SUPPOR Pn~E.004 2

2.2 Cycle 8 (11-29-88 through 9-15-90}

In July 1985 the 10 CFR 50.61 rule was published covering reactor vessel fracture toughness requirements for protection against PTS events. This rule established a screening criteria that was based on a correlation using reactor vessel weld and base metal chemistry and projected fast neutron exposure. Applicat;on of the rule resulted in Palisades exceeding the screening criteria for the reactor vess~l base metal before £0l. In August 1986 CPCo committed to initiate a reactor vessel fast neutron fluence reduction program beginning with the next cycle (Cycle 8). The fuel enrichment for Cycle 8 reload fuel had already been purchased, therefore, flexibility for Cycle 8 core design was limited.

To reduce tha fast neutron flux to the reactor vessel base metal, the overall peak value must be reduced. The peak occurs at eight core azimuthal locations at the octant def1ned angle of apprcximately 16 deqrees. While planning how to achieve the reQuired vessel flux .

reductions for Cycle 8, discussions with the NRC revealed that the PTS rule would likely change to that depicted in draft ReQulatory Guide 1.99 revision z:i. Evaluation of the impact of Regulatory Guide l.99 revision 2 verse 10 CFR 50.61 revealed th~t th~ 11m1t1ng rA~~tnr VAS~Al hPltlin~

material changed to the axial welds verse the previously limiting base

~h l.

• Tn rPd11r-P t.hP fA~t nP.Lltron flux to the reactor vessel axial welds, the peak values at octant defined angles of 0 and 30 degrees must be reduced.*

The method chosen to reduce th& fast neutron flux at thg limiting axial welds was to rebuild and use previously discharged fuel that had already

• been used in 3 cycles. The most recently discharged fuel was cnosen, Batch H. 16 Batch H assemblies were rebuilt during Cycle 7 operation by replacing 4 rows of fuel rods with stainless steel rods. Figure 2 shows how the key 48 core periphery locations were loaded with fuel for Cycle

8. Thermal margin gains reali2ed by the introduction of new Thermal Margin Monitors beginnihg with Cycle 8 allowed the cycle length to remain ~t the ~3~C approximate value as previous cycles, 37~ EFPD, despite the higher power peaking that results when more power i~ placed within the core interior by imp1cmcnting a 1ow 1cak~9c fuel m~n~gcmcnt scheme. 60 new Batch L assemblies were used in the core design.

~.3 Cycle 9 (3-15-91 through Z-6-92)

Cycle 9 core design was based on the assumption that 10 CFR 50.61 would

ncorporate the Regulatory Guide 1.99 revision 2 chemistry and fluence correlations. The anticipated revision did occur in May 1991. Cycle 9 core des'ign took further steps to reduce limiting and overall fast neutron flux reaching the reactor vessel. 16 three times burned Batch I asse~b1ies replaced the stainless steel H assemblies used during the previous cycle. These I assemblies also incorporated hafnium clusters to rurther suppress power and, in turn, rd

it m:ulr*uu:s frum Lf1~s~

assemblies. The hafnium clusters consisted of eight symmetrically located harnium rod~ *with th~ sc1111t ltmylt1 ds 11 fut!l ruli. AtJt.tiL1onally, the Cycle 9 core design placed two times burned assemblies in the remaining key core per*iphtff.Y lu1,;dlium;. N~w ful::!l nu longer ex1sted 1n any of the 48 key periphery locations. Figure 3 shows how the key 48 core periphe1*y .loc..ations were 1voJ~u wllt1 fut:l for Cyl:ll::! 9.

NOV !2 '94 16r35 FROM CPCO PnLJSnDE5 SUPPOR PnGE.005 3

The core design was now challenged with the need to p1ace many new assembhes Within the core interior. lh1s was compl1cated by the fact that Palisades was still maintaining an octant symnetr1c core design.

lhe resutt1ng Cycle 9 core design used 52 new Batch Massemblies with a lower than usual batch average enrichment (2.69 verse 3.26 w/o U-235).

Thermal margin gains were also needed. These margin gains were obtained from a new departure from nucleate boiling (DNB) correlation and by taking credit for increased PCS flow from a new measurement technique.

The steam generators were replaced for Cycle 9 operation, however, the core des;gn did not incorporate the resu1ting margin gain from higher PCS flow with the new steam generators. This was due to timing between the design of a reload and the decision to replace the steam generators.

The new DNB correlation margin gains were on1y applicable to the new

.Batch H fuel (and subsequent Batches) which-are high thennal perfonnance (HTP) design assemblies. The Cycle 9 length was 299 EFPD.

2.4 Cycle 10 (4-19-92 through 6-4-93)

Cycle 10 core design further reduced the peak flux value seen by the reactor vessel circumferential weld and reduced the flux value seen by the O degree axial welds. At the time Cycle 10 core was designed, the axial welds were limiting. However, the circumferential weld was not far behind. The flux value seen by the 30 degree welds was maintained at the Cycle 9 value. These further flux reductions were accomplished by the use of the same Batch I hafnium assemblies used during cycle 9 and by the use of S new Batch N shield assemblies. These new shield assemblies had two rows of stainless steel rods on each of two opposing sides of the assemblies. The fuel rods within these assemblies used 1.2 w/o'U-235. The 68 remaining new Batch N assemblies contained a 3.36 w/o U-235 batch average enrichment which was a step up from any previous P.nric.hmimt. 11*"::ic1 ;it P;il 1 ~;itfp,'. FigurP 4 'hnwe\ hnw t.hP kPy 4R t'.'OrP.

periphery 1Qcations were loaded with fuel for Cycle 10.

The implementation of the Cycle 10 core required use of the thermal margin gains achieved from the increased PCS flow due to the replacem9nt of the steam generators and the new DNB correlation mentioned above in tha Cycle g discussion. Another key item for Cycle 10 implementation was the change from octant symmetry to quarter core rotational symmetry.

This important change a11owed larger flexibility for placement of the new fuel assemblies within the core interior. The Cycle 10 length was 35i' EFPO.

2.6 Cyc1c 11 (11-8-93 through 2-96, planned)

The Cyc1c ll core design ww~ the fir~t core to attempt to implement what the industry refers to as an 18 month cycle. Management had the vision of Palisades operating at high capacity factors following the

• replacement of the steam generators. This vision meant that O&M savings incurred due to higher capacity factors could offset the higher fue1 costs as~ociated with higher enrichments that are required for longer operating cyc1e$.

NOV 2 '94 16136 FROM CPCC PnLJSnDES SUPPOR P()GE. e.06 e 4 The Cycle 11 core design originally planned to utilize the same Bate~ I assemblies used during Cycles 9 and 10, along with the Batch N shield assemblies. However, a failed fuel rod event that was discovered dur1ng the Cycle 10/11 refueling outage associated with a Batch I assembly required alternate shielding assemblies to replace the Batch I assemblies. 16 three times burned Batch L assemblies were used for this replacement. 14 fuel rods 1n each Batch L assembly were replaced by stainless steel rods. The Batch L rebuild effort was directed at the failed rod root cause efforti however, this would further reduce the fast neutron flux. The resu ting flux to the reactor vessel welds

-during Cycle 11 is estimated to be slightly less than that seen durir:g Cycle 10. The core design used 60 new Batch O assemblies with a 3.98 w/o U-235 batch average enrichment. Margin gain was realized by removal of a 2% mixed core DNB penalty now that all assemblies (except for the 16 low powered Batch L assemblies) were of the HTP des1qn. The c.vcle length is estimated at 421 EFPD.

3 *. 0 RESULTS figure 5 dapicts the relative powers for the key 48 core periphery locations for cycles 1-11. Table l shows the flux reduction results that have been achieved. Flux values at the reactor vessel clad/base rneta1 interface are given for the key azimuthal locations.

In June 1992 CPCo submitted a revised PTS screening criteria evaluation to tJtp NRL Tn t.tiat suhmithl r.hP.mist.ry valllP.s r.h;ingP.rl from thns~

previously submitted. The circumferent;al weld was now l;miting, with the alda1 welds net far behind when usin!J Cycle 9 flux rates to project EOL fluence. However, the PTS screening criteria was not exceeded until 2008 which is after the present licensed l;fe of March 2007. The NRC staff's review has applied two additional chemistry data points which makes the axial welds again limiting. An updated projection of ~hen the PTS screening criteria date would be exceeded when these two additional point~ 3~c included is a few years ~hort from reaching the year 2007.

fuel management alone may be able to allow the PTS screening criteria date to be dcl~ycd until 2007. A pl~n hat been developed to evaluate our options to resolve reactor vessel integrity concerns associated with PTS. [v~lu~tin9 further flux reduction m~asurcs through fuel management is part of the plan.

1l£f[R£NCES

1. tode of redera1 Regulations, 10CFRS0.61

2. J.f. Carew, L.Lois, "Reduction of Pressure Vessel Neutron rluence -

lndustry Response to 10CFR50.61", ANS Transactions June 2-6, 1991 J. Regu1atory Guide L99 Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", Hay 1988

... NOV 2 '94 16:37 FROM CPCO PnLJSnDES SUPPOR PnGE.007 e

FIGURE l TY?ICAL CYCLES l THROUGH ? P!RIPH!RAL LOADING PA'ITERN Number of Accumulated Cycles 270° 0 0 0 0 0 0

~

//30°.

Axial Weld Aiimuthal Location ALL P!RIPR£lAI. ASSEMBLIES All! FR!SH f'UEL

NOV 2 '94 16:::37 ~ROM CPCO PnLJSnDES SUPPOR Pf'IGE.008 e

FIGURE 2 CYCLE 8 PERIPHERAL LOADING PATTF.~N Stat.nleu Steel Shield Roda 180" Nu111beT of Accumulat d Cycles Axial Weld Azimuthal Location ASStHDLirs RtPRCSCNTa> DV 3 ARE THRICE eunNEO FUEL WITH SS SHIELD ROOS

.... .Nov 2 '94 16138 ~ROM CPCO PnLJSnDES SUP~CR PnGE:.009 e

FIGURE 3

~YCLE 9 PERIPHERAL LOADING PATTCRN Number of Accumu1~tcd Cycles z a 2

2 2 2 2 2

2 3 3 90* 210*

3 2

2 2 ..*

2 2 '

Axial Weld o*

Azimuth~l Location ASSEMBLIES REPRESENTED BY 3 ARE THRICE BURNED FUEL WI~H HAFNIUM ABSORBERS

NO~ 2 '94 161~8 ~POM CPCO PnLISntES SUPPCR FIGURE 4 CYCLE 10 PERIPHERAL LOADING PATTERN Stainless Steel Shielu Ruds Number of Accumulated Cycles z

2 2 270*

90~

Axh1 Weld Azimuthal Location ASSEMBLIES REPRESENTED BY 4 ARE FOUR TIMES BURNED FUEL WITH tiAF.NIUM ABSORBERS ASSlM~LltS REPRESENTED BY 0 ARE NEW SHIELD FUEL WITH SS SHIELD ROOS AND l.2 W/O ENRICHMENT

NOV 2 '94 16139 ~ROM CPCO PnLI5nDE5 5UP~OR PnGE.011 e

Figure 5 Core Octant Peripheral Assembly Relative Power Distribution C.vcles 1-7 Assembly Type/

Cycle 8 Number Of T;mes (Cycles)

C.vcle 9 Assembly Has Cycle 10 Previously Been Cycle 11 Burned

- 0 Degree l 2 3 4 5 6 7 8 Axial Weld 0.89 New 0.27 SS H 3x 0.37 Hf 3X 0.27 SS N OX 0.23. SS N lX 9 10 11 12 13 14 15 0.87 New 0.82 Nt!W 0.48 2X 0.37 zx 0.37 2X 16 17. 18 19 20 21 0.6~ New

• 0.68 New

• o.37 2X o.zz Hf 4x '-......

ss L 3X _16 Degree 0.24 Circ Weld 22 23 24 25 0.96 New n.65 2X 0.55 2X 0.51 2X 0.44 2X 26 27 . 28 0.97 0.61 New 0.96 0 .17 SS H JX 0.61 0.19 Hf 3x 0.56 0.18 Hf 4X 0.47 0.19 SS l 3X New New 2X 2X 10 OP.gree 2X Axial Weld

"" . NOV 2 '94 16: 39 ~ROM CPCO PnLISntES SUPCOR P()GE.012

. *~

e Tabh l

- Neutron ~ast Flux at Reactor Vessel Clad/~asa Metal Interfaci (EIO n/cm2-scic) location Cycle Cycle Cycle Cycle Cycle 1-7 8 g 10 11 0 Degree 4.87 2.15 2.0B I.51 1.42

• Axhl Weld 16 Degree 6.25 4.89 3.06 2.40 2.21 Circ Weld 30 Degree Axial Weld 4.79- 2.34 2.00 \ 2.ooJ 1.66 1'i~r* C, ,_f,. /() .,._ l...~r

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