ML20032A146

Nonproprietary Version of Supplementary Sar,Palisades Gadolinia Demonstration Program,Cycle 4

ML20032A146
Person / Time
Site:	Palisades
Issue date:	09/25/1981
From:	Nielsen L SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML18046B000	List: ML18046A999 ML20032A146
References
XN-NF-79-61(NP), NUDOCS 8110280366
Download: ML20032A146 (25)

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1 XN NF 79-Bl[NP) g l

I SUPPLEMENTARY SAFETY ANALYSIS REPORT I

PALISADES GAD 0LINIA DEMONSTRATION PROGRAM I

CYCLE 4 I

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Issue Date: 09/25/81 I

SUPPLEMENTARY SAFETY ANALYSIS REPORT pal.ISADES GADOLINIA DEMONSTRATION PROGRAM CYCLE 4 Prepared By:,'

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L. A. Nielsen, Unit Manager Ddte PWR Neutronics

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Prepared By:

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G. C. Cooke, Manager Date Plant Transient Analysis Y

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Concur:

~R. B. Stot'It, Managt*r.

Date Neutronics and Fuel' anagement Concur:

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'f/Ziff/

I J. N. Morgar// Manager ~/

Date Licensing atfd Safety Engineering

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Concur:

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'G.'F.'OWsley,Magager Reload Fuel Licensing

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Approve:

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G. K. Yofer, Manager Date Fuel Engineering and Technical Services E'

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E(ON NUCLEAR COMPANY,Inc.

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E NUCLEAR REGUL ATORY COMMISSION DISCLAIMER r

I IMPORT ANT NOTICE REGARDING CONTENTS AND USE OF THIS(CUMENT PLE ASE READ CAREFULLY W

t.I This technical r epor t was tiersved through researcti and devokipment prograrns sconsored by Exxon Nuclear Company, f oc.

It is emnq sub mitted by Exxon Nuc tear to the USNRC as part of a ter hnical coe' j

bu tion to faciiitate safety analyses by brensees of the I!SNRC wh ch l

utilize Exxon Nuclear icivicatn1 r elo.wi fuel or other t ech n.c.d serv.res providnt by Exson Nuclear for ti+t water power reactors anit it is f ree arai cor re. ' a the twst of Exxon f auclear's 1 rnswinige, ir,'o r t na tion.

arut tiehef.

the inf ormation containoi herein snay be usal by the USN PC m i:3 review of i b.s r epor t, anit ',y I xnsees or app!a ar;ts t.. *i a s t!.

I U%RC which are customers of Exxon Nuclear in their da mtration of cornahance with the USN RC's regulations.

W thout derogatmg from the foregoing, neither Exxon Nuc! ear rv any perion acting on its bel alf.

4 A Mak es.niy w cirr ari t y, ex tiress its ini; >1 ris t.

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c on'pleteness t il.it iUll i AWilasf M1I Hn lli4% i$t ii a lii tt i i l, i >f lll.il ilie ily at

.piy in t or n i.t t son. al ipa s.it us, sne flw ul, of l.f i n ev. aled i. ell in this ilis une n! will ait inf riniy lit iv.itef y i wswwl reiht,,

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Assumes ar y liabihties with resp.r t to t he use o f, o r fv >r darrages result nq fror-the use of, any mfor n ution, ap para tu s, met huf. or process thsclosed in this dorumen!.

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-i-XN-NF-79-61(NP) i,I TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

AND

SUMMARY

1 i.

2.0 BACKGROUND

2 1

j 3.0 NEUTRONIC ANALYSIS.......................

6 3.1 GAD 0LINIA BEARING FUEL CELL CROSS SECTION 6

I 3.2 GADOLINIA BEARING FUEL ASSEMBLY CALCULATION 7

3.3 CONE ANALYSIS 8

I 4.0 THERMAL DEhiGN--(Gd 0 - U0 R00 TEMPERATURE)..........

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t 5.0 ECCS ANALYSIS 13 APPENDIX A REFERENCES........................ 20 1

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L_IST OF TABLES Table Page 4.1 COMPARIS0N OF LHGR VALUES FOR 4 W/0 Gd - UO FUEL RODS....

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LIJT OF FIGURES lI i

Figure Page 2.1 RATIO 0F CALCULATED TO MEA'aVRED LOCAL POWER DISTRIBUTION 1

OYSTER CREEK LEAD FUEL ASSEMBLY - 3,800 MWD /MTU

.2 PALISADES CYCLE 3 GAD 0LINIA POISONED ASSEMBLIES l

POWER DENSITY-PREDICTION VS MEASURED 5

ll 3.I PIN CELL INFINITE MULTIPLICATION FACTOR VS EXPOSURE 9

l 3.2 CYCLE 4 POWER DISTRIBUTION SENSITIVITY TO GAD 0LINIA WORTH AT

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500 MWD /MT AR0, HFP 10 l_

5.1 GAD 0LINIA ROD RELATIVE POWER VERSUS PEAK 00 R0D EXPOSURE 16 2

l 5.2 GADOLINIA ROD PELATIVE BURNUP VERSUS PEAK U0 R0D EXPOSURE..

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5.3 COMPARIS0N OF FUEL TEMPERATURES VERSUS EXPOSURE 18 5.4 COMPARISON OF FUEL ROD INTERNAL GAS QUANTITIES VERSUS EXPOSURE 19

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1.0 INTRODUCTION

AND SUMMARf The Batch H reload for Cycle 4 of the Palisades Reactor will consist of 68 assemblies with a batch average U235 enrichment of 3.27 w/o.

Included in the reload will be thirty-two (32) fuel pins which contain 4.00 w/o Gd 0 in UO enriched to 2.69 w/o U-235.

The' gadolinia bearing i

23 2

pins will be distributed equally among four (4) fuel assemblies.

The purpose of inserting tne 4 w/o Gd 0 bearing fuel rods in Batch H is to 23 demonstrate the application of a concentration of gadolinia bearing fuel s

in a pressurized water reactor (PWR).

The number of pins constitutes a sufficient quantii.y of gadolinia bearing fuel rods in the reload to control the power distribution and correspond to a measurable quantity of core reactivity.

The four assemblies with the thirty-two gadolinia poisoned fuel rods at 4.00 w/o Gd 0 are replacing thirty-two 3.43 w/o 23 U-235 fuel pins.

The continuing gadolinia demonstration program in Palisades will enhanse the PWR data base obtained from the current Palisades and Prairie Island Unit 1 programs and provide the experience

g which will allow transition in the future from demonstration quantities

g of gadolinia to production quantities of gadolinia.

The incorporation of four assemblies with a high concentration of gadolinia into the cycle 4 core does not significantly change the opera-ting parameters of the core.

Sofety analysis of the cycle 4 core and Batch H fuel with the Gd 0 containing assemblies have been performed.

23 Results of the analyses show that the gadolinia rods do not alter the safety limits or margins for the Palisades Satch H fuel design.

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' XN-NF-79-61 (NP)

2.0 BACKGROUND

Gadolinia bearing fuel (UO - Gd 0 ) supplied by Exxon Nuclear Company 2

23 (ENC) has undergone irradiation in BWR's for several years.

In addition, Exxon Nuclear Company urrently has small quantities of gadolinia bearing uranium fuel rods under irradiation in the Palisades and Prairie Island Unit 1 PWR Nuclear Plants.

A substantial number of Exxon Nuclear supplied BWR fuel assemblies con-taining gadolinia as a burnable poisen have been irradiated to high burnups.

The gadolinia is contained ia several fuel rods in each assembly and is uni-formly blended with the enriched UO '

2 Typical irradiated fuel assemblies have been examined during the reactor refueling outages.

The examinations have includ2d visual examinations, fuel rod diameter measurements, fuel rod length measurements, and gamma scar measure-ments.

The fuel examinations performed to date, including fuel rods containing gadolinia, have revealed no abnormalities.

The gamma scan measurements have demonstrated the accuracy of the ENC calculational methods to predict the depletion of the gadolinia. A comparison of calculated and measured local power distributions for a BWR fuel assembly is shown on Figure 2.1.

The calculated powers in the gadolinia rods compare well with the measured powers.

Currently in Palisades, there are a total of 32 rods distributed among eight assemblies with 1.00 w/o gadolinia in the UO fuel pellets. These rods 2

a weie ioaded in the Palisades reactor at the start of the present operating cycle (Cycle 3).

Comparisons of measured assembly power to predicted assembly power in the gadolinia poisoned assemblies shown in Figure 2.2 show power a

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f differences of less than variance from the ENC predictien. These assemblies will continue to be closely :nonitored and compared to ENC predictions throughout the cycle.

In che Prairie Island Unit 1 Nuclear Power plant, there are currently 64 gadolinia poisoned fuel rods being irradiated. These rods were loaded in 16 assemblies at the beginning of the current operating cycle (Cycle 5). Measured power dis'.ributions show power differences in the gadolinia bearing assemblies of only to variance from the ENC prediction. The closc agreement of predicted and measured values of g

Prairie Island Unit 1 is similar to that for Palisades.

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Figure 2.1 Ratio of Calculated to Measured Local Power Distribution Oyster Creek Lead Fuel Asserably - 3,800 f1WD/MTU I

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Finure 2.2 Palisades Cycle 3 i

Gadolinia Poisoned Assemblies j

Power Density Prediction vs measured i

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XONF-79-61(hP) 3.0 NEUTRONIC ANALYSIS I

The neutronics calculations for the gadolinia bearing rods to be loaded in Palisades are based on standard Exxon Nuclear Company methods.

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Modeling techniques developed for the present gadolinia loading in Palisades are used.

The UO2 - Gd 023 fuel cell cross sections are calculated with a multigroup transport theory code which includes the effect of the surrounding cells on the neutron energy spectrum.

From this calculation, transport cor-I rected diffusion theory cross sections are developed for a discrete pin cell.

These cross sections may then be input directly in a discrete mesh ccre model or alternately into single assenbly calculations from which flux weighted cross sections are calculated for use in a nodal code.

3.1 Gadolinia Bearing Fcel Cell CrossSection I

The gadolinia bearing fuel cell was depleted and cross sections generated with the XPINI4) code.

XPIN calculates infinite lattice parameters by multigroup transport theory.

I This " super cell" is depleted and the cross sections in the central cell collapsed to two groups.

In Figure 3.1 the infinite multiplication factor for a 2.69 w/o enriched pin cell con-taining 4 w/o gadlinia is compared as a function of exposure to a similar cell with no gadolinia.

Effective two group diffusion cross sections are developed using a rectangular representation of the super cell.

The super cell is modeled with a discrete mesh (one mesh interval per pin pitch) diffusion theery cal-culation using PDQ.( ) The fasc group cross sections are taken directly from the XPIN, Tin cell, while the thermal group absorption and fission cross I

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ XN-NF-79-61(NP) k sections are corrected until the diffusion theory rea' ; ion rates (fast and thermal) match those predicted with transport theor'.

I 3.2 Gadolinia Bearing Fuel Assembly Calculatio_r3 Assembly calculations have been done to determine a desirable distribution of gadolinia bearing fuel rods within an asserr51y. A discrete mesh, diffusion theory PDQ representation was utilized.

With this model, both the number and location of UO - 6d 02 3 pins can be studied in detail.

2 Based on the assembly calculations, and on core calculations to be discussed later, an assembly loading configuration has been determined.

For the Palisades gadolinium demonstration program, eight standard fuel rods in the gadolinium bearing fuel assemblies were replaced with 2.69 w/o UO rods containing 4 w/o gadolinia.

On an assembly basis the worth 2

of the gadolinia is predicted to be at beginning of life. At an assembly exposure of 4,000 MWD /MT, the poison worth has diminished to about at 8,000 MWD /MT, the poison worth has diminshed to about and has become indistinguishable by an assembly exposure of about j

10,000 MWC/MT.

The effect of the UO -Gd 02 3 pins on assembly local peaking 2

was studied for the Reload H design.

Included were fresh fuel assemblies containing B C - Al 0 burnable poison and assemblies containing U02-4 23 Gd 0 Pi"S' 23 I

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XN-NF-79-61 ( NP ) I 3.3 Core Analysis l

The Cycle 4 reference core design has been analyzed with the 3-D l

reactor simultor XTG(6)

With the planned loading, the gadolinia poison is predicted to be equivalent to about of soluble boron at beginning of cycle.

I The Cycle 4 fuel loading pattern has been designed such that I

even if the gadolinia poison worth is significantly smaller or signifi-cantly larger than expected, a desirable core power distribution will be i

achiesed.

The effects of off-nominal gadolinia poison worths have been studied by varying the poison worth while maintaining a fixed loading pattern.

Three cases were considered and the core depletion character-istics studied.

The base case is the predicted core behavior, if the gadolinia worth and burnout is accurately calculated. The off-nominal extremes of poison worth were bracketed by setting the beginning of cycle (B0C) poison worth to ene-half the nominal worth for one case and for the othe extreme, the BOC gadolinia worth was assumed to be 50%

more than predicted.

These postulated core configurations were depleted and the resulting pcwer distribution compared.

As expected, there are noticeable differences between the nominal and extremes in the BOL relative power cistribution.

However, as shown in Figure 3.2, the power shape associated with either extreme is acceptable. The power distributions for these three cases at equili-brium conditions (500 MWD /MTU) are given in Figure 3.2.

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Figure 3.2 Cycle 4 Power Distribution Sensitivity to Gadolinia Worth at 500 MWD /MT AR0, HFP I

I XN-NF-79-61(NP) 4.0 THERMAL DESIGN--(Gd 0 - U0 Rod Temperatures) 33 g

An analysis of the Palisades poison rods (4.00 w/o Gd 0 ) was performed 23 to determine allowable LHGR values which preclude pellet centerline melt as a function of pellet exposure. The impact upon picnt operation was then determined by comparison of the limiting LHGR values anticipated for the poison rods during the life of the fuel.

The results of the comparison indicate that sufficient margin exists between pellet centerline melt and maximum anticipated LHGR values for the poison rods up to pellet exposures of 40,000 MWD /MT.

The allowable LHGR values to preclude pellet centeriine melt were deter-mined using models identified in the ENC gadolinia fuels topical report ( }

which includes the effect of Gd 0 on the thermal properties of the fuel 23 pellets.

This report indicates the melting temperature for a 4.00 w/o poison rod to be at beginning-of-life.

The melting temperature was assumed to degrade at the same rate as a fuel pellet containing no Gd 0 23 LHGR values corresponding to T at several exposures melt appear in Table 4.1.

Also shown in Table 4.1 are the maximum calculated transient LHGR values indicating a minimum margin of occurring at end of life.

I In order to determine the impact of the poison rods upon reactor oper-ation, the results of the plant transient analysis (10) for the Palisades reactor were examined to determine the peak kw/ft anticipated for the rod withdrawal transient and compared against the li.,it values. The values of the peak kw/ft for the rod withdrawal transient were conservatively evalu-ated and the values appear in Table 4.1.

As the results in Table 4.1 indicate, sufficient margin exists between T

and the maximum anticipated kw/ft to preclude centerline melt for the melt poison pellets.

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.! 5 Table 4.1 Comparison of LHGR Values for i

4 w/o Gd2 - U02 Fuel Rods

{ g Pellet Exposure g

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(kw/ft)* Transient 10,000 17.9 l

20,000 17,9 30,000 1/,9 40,000 17.9 t

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• Single rod withdrawal transient, EOC conditions,126% power overshoot r

i F = 1.45, F = 1.75, F = 1.0, F - 1.03.

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XN-NF-79-61(NP) 5.0 ECCS ANALYSIS This section establishes that the 4.0 wt % gadolinia (Gd 0 ) bearing 23 fuel rods (referred to as Gd 0 rods) to be included in the Palisades H 23 gadolinia assemblies are not limiting rods in a LOCA, and hence, do not impact the ECCS allowable total peaking limits applicable to the Palisades H design. The use of a lower enrichment for the Gd 9 rods in the Palisades 23 H gadolinia assemblies precludes these rods from being the limiting rods in a LOCA.

The lower enrichment (2.69% for the Gd 0 rods versus 3.43%

23 in assemblies without gadolinia) reduces both the power and burnup of the Gd 0 rods relative to the UO rods in the assembly.

The lower power 23 2

and burnup lead to corresponding reductions in fuel temperature and fission gas relea;e.

~ Figure 5.1 shows the ratio of Gd 0 rod power to peak UO rod power 23 2

versus the exposure of the peak UO rod.

The U0 rod considered is the 2

2 ECCS limiting (peak power) UO rod in the assembly.

Figure 5.1 shtvs that 2

the Gd 0 burnable isotopes are largcly depleted when the peak U0 rod 23 2

reaches a burnup of 15,000 MWD /MTM (8000 MWD /MTM burnup on the Gd 0 23 rod).

Beyond this point, the relative power of the Gd 0 rod increases 23 slowly from about 80% of the UO rod power at a U0 rod burnup of 15000 2

2 MWD /MTM to just over 99% of the UO rod power at end-of-life.

2 r d burnup to Figure 5.2 provides the corresponding ratio of Cd 023 peak UO rod burnup versus the exposure of the peak U0 rod.

The burnup 2

2 rod of the Gd 0 r d is considerably less than that of the peak U02 23 rod burnup is only 80%

throughout life and even at end-of-life the Gd 023 that of the UO rod.

2 A comparison of steady-state peak pellet volume average temperatures (i.e., stored energy) between the Gd 0 rod and the peak power UO rod 23 2

as a function of the U0 r d peak pellet burnup is shown in Figure 5.3.

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XN-NF-79-61(NP)

The results correspond to the ECCS total peaking limit, F, of 2.76 for q

the limiting U0 rod in the Palisades H design (14.93 kw/ft total power, 2

14.61 kw/f t heat reledse in the f uel). At low exposures, the Gd 0 rod 23 has much lower stcred energy than the UO rod because of its reduced 2

power (Figure 5.1).

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is due to the reduced thermal conductivity for Gd 0 bearing pellets 23 identified in Reference 7.

I The Gd 0 rod will have a lower peak 23 clad temperature (PCT) in the LOCA than the limiting UO2 rod.

Figure 5.4 compares the internal gas quantity (gram moles) within the free volume of the Gd 0 and peak UO rods.

At 'aw exposure, the 23 2

rod is associated with sorbec slightly higher gas quantity of the U02 gas release which, in turn, is associated with the comparatively high rod at beginning-of-life and low fuel average temperature of the UO2 and Gd 0 rods exposure.

The increasing gas quantity for both the U02 23 at high exposure results from the burnup dependent enhanced fission gas I

release model specified by the NRC.

Figure 5.4 shows that for U02 red peak pellet burnops in excess of 20,000 MWD /MTM, the fission gas release enhancement effect on the pin ilternal gas quantity at any time is much less for the Gd 0 rod than for the limiting UO red. This is uecause 23 2

the burnup of the Gd 0 r d is only 65-80% that of the peak UO2 rod 22 (Figure 5.2).

The lower Gd 0 r d fission gas release is a further factor which 23 makes the Gd 0 r d less limiting than the peak power UO rod. At high 23 2

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_ XN-NF-79-61(NP)

.I exposures (> 30,000 MWD /MTM peak pellet burnup) the calculated PCT in the LOCA increases due to increased cladding strain and associated increases in metal water reaction.

The increased strain stems from increased fuel rod internal pressure due to enhanced fission gas release.

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l XN-NF-79-61(NP) APPENDIX A REFERENCES I

1.

F. B. Skogen, ' Exxon Nuclear Neutronic Design Method for Pressurized Water Reactors," XN-75-27 Exxon Nuclear Company, Ju'e, 1975.

2.

Supplement 1 to Reference 1, September,1976 3.

Supplement 2 to Reference 1, December, 1977.

l 4

W. W. Porath, A. H. Robinson, a d D. R. Skeen, "XPIN-The Exxon Nuclear Revised HAMBUR Users Manual", XN-CC-26, Rev.1 Exxon Nuclear Company, December, 1375.

5.

W. R. Cladwell, "PD;,'7 Reference Manual", WAPD-TM-687, Westinghouse Electric Corporation, January, iP5.

6.

R. B. Stout, "XTG-A Two-Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing", XN-CC-28, Rev. 3, Exxon Nuclear Company, August, 1975.

7.

L. D. Gerrald, and G. C. Cooke, "Gadolinia Fuel Properties for LWR Fuel Safety Evaluation," XN-NF-79-56 (Proprietary), Exxon Nuclear Company, July, 1979.

l 8.

J. A. Christenseri, et al., " Melting Point of Irradiated Uranium Dioxide," Trans. American Nuclear Society, 7(2),390-391,1964.

9.

J. A. Christensen, et al, " Melting Point of Irradiated UO," WCAP-6065.

2 10.

XN-NF-78-18, " Plant Transient Analysis of the Palisades Reactor for Operation at 2530 MWt," July, 1977.

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Issue Date: 09/26/81 l

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SUPPLEMENTARY SAFETY ANALYSIS REPORT i

PALISADES GADOLINIA DEMONSTRATION FROGRAM 4

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1 Distribution f

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L. A. Nielsen G. F. Owsley 1

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I Document Control (5)

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