ML20107B588

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Application for Amend to License DPR-16,authorizing Use of Plant Type IV Fuel Bundles,Designed & Fabricated by Exxon Nuclear Co
ML20107B588
Person / Time
Site: Oyster Creek
Issue date: 04/04/1973
From: Sims R
JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML18039A986 List: ... further results
References
FOIA-95-258 NUDOCS 9604160446
Download: ML20107B588 (44)
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FACILITY CHANGE RFQUEST NO. 4 OYSTER. CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 Applicant hereby requests the Atomic Energy Commission to authorize:

1.

The use of Oyster Creek Type IV fuel bundles, designed and f abricated by the Exxon Nuclear Coc:pany, and 2.

Operation of the Oyster Creek reactor with the reload configura-tion described herein.

This change request includes a full description of the Type IV fuel bundles and the safety, transient, and accident analyses that demonstrate that these changes do not present significant safety hazards considerations 4

not described or implicit in the Oyster Creek Nuclear Generating Station i

Facility Description and Safety An'alysis Report and Amendments, thereto.

Therefore, there is reasonable assurance that the health and safety of the public will not be endangered by these changes.

i JERSEY CENTRAL POWER 6 LIGHT COMPANY t-

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By Vice President STATE OF NEW JERSEY

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l COUNTY OF MORRIS

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Sworn and subscribed to before me this /

  • day of

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1973.

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Notary Public MARION P. BAwlEC MOTARY PUBUC Of NEW JER0EY My Commission Expires Jan.21,1974 9604160446 960213 PDR FOIA DEKOK95-258 PDR h-e

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c Jersey Central P wd[l- & Light Company.

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.M ADISON AVENUE AT PUNCH BOWL RO AD e MoRRISTOWN, N.J, o7960 e 539 6111 April 4, 1973 I

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The Honorable Lawrence J. McNally Mayor, Lacey Township

' P. O. Box 475 l

Forked River, New Jersey 08731

Dear Mayor McNally:

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Enclosed herewith is one copy of Supplement No. 2 to Facility Change Request No. 4 for the Oyster Creek Nuclear Generating Station.

3 This Change Request was filed with the Atomic Energy Commission en f.pril 4, 1973.

ry truly yours, g

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R. H. Sims Vice President w-Enclosures 4

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UNITED STATES ' 0F AMERICA.

ATOMIC' ENERGY COM11SSION' IN Tile MATTER OF

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DOCKET No. 50-219 JERSEY CENTRAL POWER & LIGHT COMPMW:

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CERTIFICATE OF SERVICE This is to certify that a copy of Supplement No. 2 to Facility Change Request No. 4 for 'the Oyster Creek Nuclear Generating Station dated April 4.,1973, and filed with the United States Atomic Energy Commi sion on April 4,1973 has this 4th day of April 1973, been served on the Mayor of Lacey Township, Ocean County, New Jersey, by deposit in-the United States mail, addressed es follows:

The Honorable Lawrence J. McNally Mayor, Lacey Township P. O. Box 475 Forked River, New Jersey 08731 JERSEY CENTRAL POWER & LICllT COMP /WY

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By Vice President

' DATED:. April 4, 1973 i

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I INTRODUCTION This submittal describes the design of the Exxon Nuclear Puo2-UO2 i

' assemblies proposed for utilization in Jersey Central Power & Light Company's j

l Oyster Creek Nuclear Generating Station.- Loading of four lead assemblics of this design is scheduled for the Spring 1973 outage.

The design of these fuel assemblies (hereaf ter referred to as Type j '

IV) is compatible with the mechanical, thermal-hydraulic, and nuclear charact'er-1stics of the first core fuel assemblies (Type I) of the Oyster Creek Nuclear Generating Station as well as the other reload fuel assemblies (156 Type II, four Type III and 144 Type III E) which have been or will be loaded into the l

reactor.

Like the previous fuel, Type III E, the Type IV fuel bundic consists of a 7 x 7 array of rods. The center rod ie a spacer capture rod, filled with solid Zircaloy and four rods containing gadoliniem oxide aro provided for supplementary reactivity control. Type IV fuel diffcrs from Type III E fuel l.

l only in that 12 of the 48 fueled rods contain Pu 39 and Fu241 rather than 2

i U-235, as the fissile isotope.

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The mechanical, thermal-hydraulic, and nuclear characteristics of the Exxon Nuclear Type IV Fuel and the effects of using this fuel in the reactor

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core are discussed in Sections II through VII below.

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II TYPE IV FUEL DESIGN AND CHARACTERISTICS I

A.

MECHANICAL DESIGN OF THE FUEL ASSEMBLIES The Type IV fuel assembly, shown in Figure 1, is made up of 49 rods j

l in a 7 x7 square array with an active fuel length of 144 inches for all rods except for the center spacer capture rod, which is filled with solid Zircaloy.

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s The spacing between rods is maintained by seven Zircaloy spacers' c

containing Inconel springs equally spaced along the length of the fuel rods and by tie plates at the upper and lower ends of the rods. A centrally-located spacer capture rod maintains the ax!al location of the spacers.

The ends of all rods have extensions which fit into holes in the upper and lower tie plates.

Eight of the peripheral fuel rods, called tie t

rods, are threaded into the lower tie plate and pass through the upper tic

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plate where they are secured by lock washers and nuts.

The remaining rods are restrained within holes in the upper and lower tie plates.

Coil springs on each rod, captured by the upper fuel rod extensions (See Figure 1), act together to force the upper tie plate upward against' the tie rod nuts.

Each spring acts individually to assure that the fuel rod remains seated in the lower tie plate.

The icvel of enrichment is identified on the end surface of each fuel rod upper end cap by a letter code (See Figure 2).

During and after assembiv. the correct position of each rod is verified by comparison with a temolate which specifies the location of each rod type, and photographs are taken of the fuel rod ends as part of the quality control records.

In addition.

notches and a unique serial number are used to identify individual fuel rods for enrichment, poiren loading, and clad thickness. A detailed description of the assembly components is presented in Table I.

Spacers The spacers are designed to maintain the correct rod-to-rod spacing but allow for differential axial expansion.

The spacer uses an egg crate type design of criss-crossed narrow Zircaloy strips interlocked and welded together with a peripheral band to form a cell for each fuel rod.

Each fuel red is centered in its cell by en Inconel spring which holds the rod against support

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o-TABLE I COMPARISON OF MECilANICAL CHARACTERISTICS OF TYPE III E AND IV ASSEMBLIES Characteristieg Type III E Type IV FUEL MATERIALS UO2 UO2 and Puo2-UO2 Initial Enrichment, w/o fissile isotope:

Low 1.59 Medium Low Figure Hedium 2.42 2

liigh 2.87 Bundle Average Fissile 2.63 2.61 Pu isotopic composition (82.2% fissile) 77% Pu239, 5.2% Pu241 4

. Pellet Dish, % of Undished Volume-2.0 (in all rods) 2.0 (in all rods)

Average Pellet Density

% Theoretical Density.

93.5 &-94.5 93.5 & 94.5**

Number of Rods and 8-0.491 Pellet Diameter, in.-

11-0.489 See 7-1.488 Figure 22-0.468 2

Melting Point, 'F 5080 5080.- Uo2 5040 - Mixed oxid i

CLADDING MATERIAL Zr-2 Zr-2 Outside Diameter, in.

0.570 0.570 Number of Rods and 27-0.0355 27-0.0355**

I k'all Thickness, in.

22-0.0455 22-0.0455 i

4 FUEL RODS i

1 Active Length in.

Standard Rod 144 144 l

Gas Plenium Length, in.

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Standard Rod 10 5/8 10 5/8 Fill Gas llelium Helium 4 4

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T,e III E Type IV_

Characteristics 1

FUEL BUNDLE Geometry 7x7 7x7 Standard Rods / Bundle 44 44 Spacer Capture Rods / Bundle 1

1 Poison Rods / Bundle 4

4 Total Rods / Bundle 49 49 Rod Pitch, inches 0.738 0.738 Water to Fuel Bundle

. Ratio (Cold) 2.48 2.48

    • See Figure 2 for locations.

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dimples with sufficient force to minimize flow induced vibration Guide tabs-are provided on the upper edge of the spacers to avoid possible hang-up during

. channeling operations.

Upper and Lower Tie Plates (See Figure 1)

The upper tie plate is a flat perforated plate of cast and machined stainless steel.

It is designed to maintain the correct position of the fuel rods, provide flow passages and maintain the position of the fuel assembly within the reactor top grid assembly. A handle'is attached to the plate for 1

loading, unloading, and for general handling of the fuel assembly. A boss on the lifting bail points to the nearest control rod when the fuel assembly is correctly oriented in the reactor core.

l The lower tie plate consists of a flat perforated plate and an inlet box section all cast and machined as a single unit from stainless steel.

The perforated plate is designed to' support and maintain the correct pesition of the rods and to permit coolant flow through the fuel assembly.

The inlet box section distributes coolant from the assembly support section to the fuel rods.

i l

Fuel Pellets and Rods The fuel consists of compacted and sintered uranium dioxide or mixed l

Pu0 -UO2 (natural uranium) powders formed intu cylindrical pellets.

The nomina:

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density of the pellets is either 93.5 or 94.5 percent of theoretical density. /

pellets are dished p remove a nominal 2 percent of the fuel volume.

This volut l

is provided to accommodate fuel axial expansion and irradiation-induced swellins The fuel pellets are stacked in Zircaloy-2 cladding sealed by welding r

Zircaloy-2 caps into each end.

The atmosphere within the rods is helium. The requirement that the cladding be free-standing is met by specifying a minimum wall thickness of.033 inches. This thickness provides substantial margin 1.

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against cresp collapse of the' cladding due to primary 'ecolant pressure through-6 out the life of the fuel.

'The design of the fuel pellets and rods considered the potential for pellet densification, fuel swelling behavior, thermal expansion and distortion of the pellets, pellet-clad interaction, and fission gas release. Consideration of these factors at the maximum design peak burnup resulted in the selection of appropriate pellet dimensions, densities, dishing r.equirements, diametral gap, and fuel rod plenum.

A comparison of the mechanical characteristics of the Exxon Nuclear mixed oxide assemblies (Type IV) and Exxon Nuclear UO2 assemblies (Type.III E) is given in Table I.

t B.

NUCLEAR CHARACTERISTICS OF THE FUEL ASSEMBLIES l

Type IV assemblies are designed and fabricated to have a nominal 1

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average enrichr. ant of 2.61 w/o fissile isotopes.

Four of the UO2 fuel rods in l

the Type IV assembly contain 1.0 w/o gadolinium as a burnable poison. Tha fuel rod enrichment and poison distribution are given in Figure 2.

All pellets in the burnable poison rods contain gadolinium.

Calculated-infinite multiplication factors and other parameters of the Type IV fuel assemblies are compared to those of Type III E assemblies in i

Tables II and III.

As the calculations indicato, the differences between the l

designs are small.

In the absence of gadolinium, each of the temperature and void defects is slightly more negative with the mixed oxide fuel, such that the overall cold-to-hot operating defect in k= is about 0.6% k more negative.

In addition, the poison' worth of the gadolinium burnable poison and its effect on the reactivity defect is less in the mixed oxide-containing fuel bundle

-because the relative thermal flux in the poison rod locations is depressed by the presence of the adjacent mixed. oxide rods. - -

I FIGURE 2 OYSTER CREEK TYPE IV ASSEMBLY

,i OO@O@G0 O O O O "'O>@

O@@3,GD Q@

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OO 09,ODO@

f)@@@4@@j TR = Tie Rod bc = Spacer Capture Rod Rod 10 w/o Fissile Clad Pellet Pell.

liark Isotonga_ !!o. Rods liatorici_ Thickness, l'ils Density Diamete E

1.59 3

U02 4 5. E.

93.5

.468 +

L 1.59 2

UO2 35.5 93.5

.488 I 4b.5 93.5

.468 T IT 2.42 11 UO2 35.5 93.5

.488 T M

2.42 1

U02 II 2.87 8

UO 45.5 93.5

.468 T 2

H-2.87 6

U0p 35.5 94.5

.489 4 H*

2.87 1

U0j 35 5 94.5'

.4911 UO +Gdp03 35.5 94.5

.488 1 G

2.87 4

2 1

2.50 3

UO94Puo2 35.5 94.5

.488 +

1*

2.50 4

UD[24 Puog 35.5 94.5

.490[

2-3.20 2

UO2iFu02 35.5 94.5

.488 1 UO +Pu02 35.5 94.6

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'Z tion-Fucled 1

Zirc 25.b.

t TABLE II CALCULATED INFINITE MULTIPLICATION FACTORS (Z_ERO EXPOSURE)

E TYPE

,E TYPE IV ja, With Without With Without Gadolinium Gadoliniun Cado1*nium Gadolinium Cold - 68 F 1.1453 1.2926 1.1576 1.2849 300 F 1.1391 1.3010 1.1558 1.2926 549 F 1.1322 1.3067 1.1541 1.2950 o

8 Full Power - 0% Void 1.1271 1.3005 1.1485 1.2886 32% Void 1.1112 1.2842 1.1331 1.2687 64% Void 1.0816 1.2497 1.1071 1.2314

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CALCULATED REACTIVITY DEFECTS (ZERO EXPOSURE)

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TYPE III E TYPE'IV Defects. (Ak./k,)

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'wo/Gd w/Gd' wo/Gd f

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.I Doppler Defect (Ilot Standby to Full Power)

.0045 0047

.0048

-1.0049 Void Defect (0-32% Voids)

.0141'

-. 0125

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.0154' Temperature Defect P

(68-549 F)

.V114

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Control Rod Worth-

-(Cold).

.151 148

-:.149

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q, Fuel rod loca1' peaking factors for Type IV fuel are compared to those'for Type--III E fuel in Figure 3.

k'hile the substitution of the twelve Pu0 -containing fuel' rods for_UO2 fuel rods substancia11y increases'the power 2

in these rods, the Local Peak-to-Average poter is.tctually-calculated to be about 2 percent lower in the Type IV fuel.

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i C.

THEPJfAL AND HYDRAULIC CllARACTERISTICS OF Tile FUEL ASSEMBLIES A comparison of the thermal and hydraulic characteristics of Type I

III E and Type IV fuel assemblies is given in Table IV.

There are no differene.

between the heat transfer ~ characteristics of the designs.

1.

Maximum Fuel Temperature The calculated maximum fuel temperature (at 17.2 kw/f t maximum linear heat generstion rate) for Type IV fuel is 4260 F, which is the same as that calculated for Type III E fuel.

This maximum temperature is'found in the same location (a perimeter UO2 fuel rod) in both designs. A Pu0 -UO2 fuel 2

rod operating at the same linear heat generation rate would.have a maximum fuel temperature approximately 200 F lower, due to the change in the rod raalal power distribution resulting from the introduction of plutonium.- Since the maximum temperature is foun? la a UO2 fuel rod in both cases, the discussion of the margin to centerline malting, the effects cf exposure, and the effects of gadolinium are the same as given in Facility Changes Request No. 4, January 18, 1973.

2.

Flow Compatibility The Type IV fuel is hyd,rau11cally identical to Tvoc III E fuel.

The discussion of flow compatibility presented in Facility Change Request No. 4 also applies to Type IV fuel.

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FUEL ROD LOCAL POWER DISTRIBUTIONS TYPE III E AND TYPC-IV FUEL 32% Void 0 Exposure 1.156 1.050 1.185 1.100 1.126 1.219

.989 1.153 1.042 1.167 1.077 1.092 1.182

.965

-r 1.159 1.102

.493 1.039 1.179 1.123 1,133 1.087 467

.971 1.122 1.069

.952

.884

.905

.977 1.095

.863

.998 1.002 J.192 1.009 0.0

.850

.454 1.007 0.0

.999

.393

.915 Pc ISLAND

.861

.921 1.033 i

.966 1.178

.909 1

III E 1.039 1.123

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IV 1.190

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1.122 1.032 r

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TABLE IV COMPARISON OF THERMAL HYDRAULIC CHARACTERISTICS OF THE OYSTER CREEK TYPE III E AND IV ASSEMBLIES CORE CONDITIONS Rated Power, MWt 1930 Operating Pressure, psia 1035 Core Inlet Enthalpy, Btu /lb 517.3 0

Total Core Flow Rate, 10 lb/hr 161.0 Core L'eakage Flow, %

7 0

Effective Core Flow Rate, 10 lb/hr 56.73 Fraction of Power Generated in Fuel, %

96.7 Characteristics Type III E Type IV FUEL EESCRIPTION Number of Active Rods 48 48 Total Fuel Length, Ft/ Bundle 576 576 Hegt Transfer Surface Ft / Bundle 85.98 85.98 Bage Rod Flow Area Ft / Bundle

.1057

.1057 POWER PEAKINO FACTORS AND FUEL PERF0PJ1ANCE Technical Specificatfon Limit--

Total Peaking Factor 3.01 3.01 Assumed Local 1.29 1.29 Axial X Radial 2.33 2.33 The total peaking facto'r limit for Types III E and IV fuel is reduced to compensate for the introduction of the inert spacer capture rod, such that

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the peak linear rate generation rate is still limited to 17.2 ku/ft.

l Includes Engineering Factor of 1.05-I 14 -

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TABLE IV (Continued) l Characteristics.

Type III E Type IV Corresponding Maximum

. Heating Rate, kw/ft.

17.2 17.2 1

l

'. Corresponding Maximum j

11 eat Flux at Rated Powc;,

2 Btu /hr-ft 393,400 393,400

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Corresponding Clad i

Surface Temperature,F 568

.568 Corresponding. Critical Heat Flux Ratio

>2.0(XN-1)

->2.0(XN-3 i

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A.

3.' Minimum Critical llent Flux Ratio

-The minimum critical heat flux ratio discussion presented in Facility Chance Request'No. 4 also applies-to Type IV fuel.

D.

FUEL QUALITY ASSURANCE AND QUALITY CONTROL PROGRU!

See Facility Change Request No. 4.

III CYCLE 3 CORE CONFIGURATION AND ANALYSIS and IV SAFETY ANALYSIS Type IV fuel assemblies are designed to be substituted for Type III E fuel assemblies with an insignificant perturbation to the core neutronics and thermal-hydraulic analyses.

For Cycle 3 operation,' four (4) Type IV assemblies will be loaded in the core positions shown on Figure 4 or in equivalent quarter core symmetric positions where they replace four fresh Type III E assemblies.

Inspection of Tables II, III, and IV cicarly shows that the core analysis performed in Facility Change Request No. 4, where reloading with 148 Type III E fuel assemblies was assumed, will be valid if four Type IV assemblic are substituted for Ty'pe III E assemblies.

The small differences in assembly neutronic parameters will result in negligibly small differences-in core average parameters.

The maximum fuel rod power in Type IV fuel will be equal to or less than the maximum rod power in Type III E fuel in the same environ-ment, and the two types of assemblies are hydraulically identical.

Because of these features, the Safety Analyses presented in Section IV of Facility Change q

Request No. 4 also apply when the core loading includes the four Type IV 2

assemblies.

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OYSTER CREEK - CYCLE LOADING PATTERN FOR' LICENSING ANALYSES IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE.

IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IV IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE IIIE Type IIIE IIIE Fuel Bundle IV Type IV Fuel Bundle T

One Quarter of Symmetrical Core Loading Pattern 4,

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INCIDEN $NALYSIS

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A.

EUEL LOADING ERRORS 1.

Rod Loading Error _s Quality control and overchecks during fuel fabrication make misplace-ment of fuel carichment extremely unlikely.

Additional assurance against fuel rod mislocation in the Type IV assemblies is provided by making the shanks of the mixed oxide rod end caps larger in diameter than those on UOh rods, and correspondingly adjusting the penetrations on both upper and lower tic plates, such that it is impossible to mislocate a mixed oxide fuel rod into a UO2 rod location.

The worst remaining case of fuel rod mislocation is the possible loading of a high enriched U02 rod (2.87 w/o U-235) in place of the low enriche rod (1.59 w/o U-235) in the wide-wide corner of the assembly (See Figure 2).

The consequences of this loading error are the same as were discussed in Facility Change Request No. 4 for Type III E fuel.

2.

Fuel Ascembly Misorientation The analysis presented in Facility Change Request No. 4 also applies to thi submit t al.

B.

TRANSIENT ANALYSIS A review of the previously submitted transient analyses (1)(2)(3) has been conducted to ascertain the effect of utili ing the Type IV mixed oxide fuel as reload fuel.

Ta facilitate this evaluation, the behavior of an equili-brium cycle employing mixed oxide reload fuel has been compared to the behavior of an equilibrium cycle employing UO2 reload fuel.

Note that an equilibrium mixed oxide core serves to accentuate neutronic differences between the Pu02 and UO2 cores for transient analysis purposes.

The results of this evaluation are presented below.

(1) Amendment No. 65, Application for an Increase In Power Level, December 30, 1970.

(2)

Amendment No. 69, Additional Information in Support of Facility Change Request No. 2, May 26, 1972.

(3)

Facility Change Request No. 4, January 18, 1973. -

CD 1.

Parcmeters Significant to Transient Results Table V suwaarizes the reactor parameters calculated for the equili-briun cycles of UO2 and mixed oxide reloads, and compares these parameters to those used in previous analyses.

An inspection of the tabulated data indicates that the only potential point of concern is the larger negative void coefficien' of reactivity which is calculated for the mixed oxide core.

Differences in the remaining parameters are either in the conservative direction or are insignifi-cant.

The effects of these parameter differences on the results of various classes of transients are discussed in paragraph 2 below.

2.

Effects on Transients a.

_ Pressurization Transients Pressurization transients are characterized by forced decreases in void content, which result in a power level increase which is terminated either by a rod scram or by action of the negative reactivity coefficients.

The magnitude of the pressure increase is determined both by the rate at which steam flow is reduced and by the magnitude of the power level increase.

Sensitivity calcu-lations have been performed for the two pressurization transients of major interest (turbine trip without bypass and Safety Valve Sizing transients) *?ith the following results:

Turbine Trip Without Bypass Transient a

Analyses of this transient are performed for end-of-cycle conditions where the shape of the rod scram reactivity insertion curve maximizes the magnitude of the power excur-sion and accompanying power spike.

The sensitivity analysis was performed using the parameters given in Table V for end of cycle conditions for the equilibrium UO2 and mixed oxide -

y TAELE T PA*AMTERS FOR MAMIENT ANALTS1$

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Esed la Pree wriaatica Certent Calcelatieen 19M st Trasatest Core Core Cycle Cycle Cycle Cycle Cycle Cycle Qualibrium 00, Eq=111bri m 29 Analysis.

Re-eaalross*

1 19 2

2 2

2 3

1 amendment Amendment 65 69 SOL 3000 f%t/T BOC EOC BOC ECC suC EOC BOC TDC BM EaC

7. elayed h tros fractium

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.00547 (s.007)

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.00605

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.00500

.00493

.00563

.00492

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Not 48,~ M tron Lifetine. Leec 39.2 31.4 48.4 37.35 40.25 s40.6 40.2 40.6 39.7 Statd

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S /iff 'isid (a 1G')

-15.65

-11.47

-13.65

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- 9.55

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- 7.96

- 6.84

- 6.44

- 6.87

- 5.81

- 5.76

- 5.95

- 6.23

- 6.45

- 6.30

- 7.41

- 7.16

- 9.34

- 8.64 tcy;1er coefficieet at s332 Teida 5

Lk/s*F (a 10 3

- 1,19

- 1. 39

- 1.23 N.C.

- 1.16

- 1. 39

- 1.09

- 1.16

- 1.08

- 1.15

- 1.34

- 1.41

- 1,41

- 1.41 t/*F

-.166

-.254

-.176 M.C.

-.192

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-.295

-. 2 35

-.192

~. 2 34

-.254

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  • These scraz curves are as ehewa in Figure 5 for the EOC. but adjusted for differences in 6 according tez 8EOC 3 Leactivity insertion ($)7 EDC 3 Reactivity lasertion ($) m

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FIGURE 5 FULL POWER Sciud! CCRVES I

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(

)

1 fuel cycles, including a 25% factor of conservatism on the void coefficients of reactivity.

The calculated result of changing the parameters from the UO EOC case to the mixed 2

oxide EOC case is + 2 ps3 in the peak transient pressure, which is an insignificant difference, Safety Valve Sizing Transient e

Analyses of this transient are performed at beginning-cf-cycle conditions.

Since the incident assumptions include the absence of a control rod system scram, the result of the transient is largely controlled by the relative magnitu<

of the void and the Doppler coefficients of reactivity.

The coefficients are most unfavorable at beginning-of-cycle conditions.

The sensitivity analysis was performed using the parameters given in Table V for beginning-of-cycle cond; tions for the equilibrium UO2 and mixed oxide fuel cycles, including a 25% factor of conservatism on the void coefficic of~ reactivity.

The calculated result of changing the para -

meters from the UO2 BOL case :o the mixed oxide BOL case is

+ 2 psi in the peak transient pressure, which is an insigni-ficant difference.

i b.

Core Flow Reductions Sensitivity calculations reported in Facility Change Request No, for these transients included the range of values for the Doppler coefficient of reactivity calculated for,the equilibrium fuel cycles, and conservatively low values for the void coefficient of reactivity. Those results indicated an insignificant reduction in the MCHER results from the core flow reduction transients. -

(,i

)

?

=

c.

Core Flow Increases A core flow increase could result from the failure of a recircula tion pump flow controller which results in a rapid flow increase in one loop. The transient is charseterized by a redaction in ti void fraction in the core, an increase in reactivity and hence power due to the decreased void volume, and a mild increase in pressure due to the power increase. The transient effectively represents shifting from one equilibrium point to another on the flow vs. power level control line.

Sensitivity analyses indicato no significant difference between the UO2 and mixed oxide fuel cycles in the results of this transient.

d.

Loss of Feedwater and Excess Feedwater Flow As indicated in Facility Change Request No. 4, the results of these transients are not significantly affected by the change in n

core coefficients of reactivity.

e.

Rod Withdrawal Incident An inadvertant rod withdrawal is characterized by a local inserti of reactivity through removal of a rod blade counterbalanced by a local reduction in reactivity due to increased voids and fuel ter perature. As Tables III and V indicate, the negative void and te perature coefficients for both the equilibrium UO2 and mixed oxid cores are larger than in Cycle 3, while the rod worths are essen-tially unchanged (all expressed in terms of Ak/k).

Hence, the local power and core power transients for the equilibrium cores will be milder than for the Cycle 3 core for a given rod withdraw.

increment.

Hence, the results of the inadvertant rod withdrawal transient analysis presented in FCR No. 4 conservatively apply to the mixed oxide fuel cycle 23 -

l

('

I s

V1 ACCIDENT ANALYSI_S A.

LOSS OF COOLANT ACCIDENT ANALYSIS The consequences of a loss of coolant accident in the Oyster Creek reactor have been evaluated for Type Il Itcl.

The models used in this analysis comply with the AEC Interim Criteria.

The methods..d assumptions used in this analysis are described in Facility Change Request No. 4, except that the local power distribution used in the analysis has been changed to correspond to Type IV fuel at 4,000 MWD /T, which is the most limiting distribution for this analyt This local power distribution is shown on Fidure 6.

The results of the loss-of-coolant accident analysis for the design basis accident are:

Peak Clad Temperature:

2238 F

~

Local Peak Metal-Water P.caction 9%

Core-Average Metal-Water Reaction:

.25 %

B.

MAIN STEAM LINE BREAK ACCIDENT The analysis of the main steam line break accident depends on the operating therral-hydraulic parameters of the overall reactor such as the pressure, and the overall factors affecting the consequences, such as primary coolant activity.

Insertion of reload fuel will not change any of th'ese para-meters so the results of the analysis discussed in Amendment 65(1) will not change.

C.

REFUELING ACCIDENT The' analysis of the refueling accident involves the mechanical damage caused by a fuel bundle falling back onto the top of the core while it is being removed, and the subsequent release of radioactive fission products.

The severity of the consequences depends on the fission product inventory in the -

FIGURE 6 LOCAL POWER DISTRIBUTION I'OR LOCA CALCULATION OYSTER CREEK TYPE IV WD !! M 7____.___________._.----.-.------

l l#8 8 8 O O O GP I

I l 0999999 l 0038888 l

l C'OOOOGG I

l 0000888 1

l 000008G

!l lf)OOOOOGj u _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _____ _ _ _

b.

)

'. ' ~

l

(,.

l fuel and various factors affecting the amount and -kind of releases to the atmot 3

phere.

The radiological effects of plutonium recycle are discussed in Section i

VII below.

Based on those evaluations, it is apparent that utilization of mixed oxide fuel rather than UO2 fuel would have no significant adverse affect on.the consequences of a refueling accident, and that the previous assessment of this accident (FD & SAR Amendment 65) is conservative 1y' valid for Type IV fuel.

D.

ROD DROP ACCIDENT

-The analysis of a rod drop ~ accident under Cycle 3 core conditions was presented in Facility Change Request No. 4 where it was concluded that a large margin exists between the maximum in-sequence control rod worth (3.5 mk) and the dropped rod worth which would result in an enthalpy deposition of 280 calories / gram (% 14 mk).. Analyses have been performed to determine the consequences of this accident under equilibrium cycle conditions for both a UO2 and a mixed oxide fuel cycle. The results of these analyses are shown on Figure 7 where they are compared to the results for Cycle 3 conditions reported in Facility Change Request No. 4.

As the figure indicates, the calcu-lated enthalpy deposition is higher for both equilibrium cycles than for Cycle 3.

However, the conclusion reached in Facility Change Request No. 4 remains valid; the value of the dropped rod for which the peak fuel enthalpy deposition is 280 calories / gram is much higher than the maximum in-sequence rod worth

(> 11 mk vs. 3.5 mk).

m p

}

FIGURE 7 PEAK ENTHALPY vs. DROPPED ROD WORTH

/

/

r 900

/

s' EQUILIBRIUM UO

/

2 800

/

/

700

/

/

600

/

/

c'

/

"n

/

EQUILIBRIUM i

500

/

MIXED OXIDE l

s' 400 PEAK

/

ENTHALPY g-cal 300 BOC 3 8

200

/

100 1

0 O

2 4

6 8

10 12 14 3.6 18 20 22 24 26 28 30 32 34 DROT' PED ROD WORTH (mk)

VII,

OFFSITE(,110LOGICALEFFECTSCONSIDERATJ

)

Of f site. radiological ef fects resulting _ from the une of mixed oxid,e fuel, including differences in inventories of fission products and trans-uranfum isotopes, could arise.from either a severe reactor accident resulting-in an d

overheated core (Section A, below), or from normal reactor operation with failed l

. fuel which exposes the oxide fuel pellets to the reactor coolant (Section B, These effects are potentially different from those of urania fuel, due i

below).

to the differences in fission product and heavy isotope inventories (Sections i

C and D, below).

These effects are discussed in the following sections in terms i

4 of'the differences resulting from the replacement of urania fuel with mixed The analyses presented below are generic to Light Water Moderated i

oxide fuel.

The difference in detailed reactor parameters and and cooled power reactors.

configurations could affect the details of the analyses, but will not alter the i

i conclusion.

A.

HYPOTHETICAL ACCIDENT EVALUATION

]

The offsite radiological effects of plutonium have bean evaluated for j.

a hypothetical accident where,the barriers between all fuel pellets and the containment have been breached. The results of this analysis indicate that the offsite dose rates contributed by plutonium are negligible compared to the dose rates from fission products alone.

Since the latter normally comply with 10 CFR 100 criteria for such an accident, it is concluded that plutonium-bearing fuel will also comply with these criteria.

These studies are based on measured values of the vapor pressure of 5 w/o Puo2 in a Puo -UO2 mixture. Tabic VI lists this vapor pressure and that 2

of strontium oxide as functions of temperature.

Since core temperatures and core configurations during the hypothe-tical accident are not well defined, the basis chosen for an assessment of the effects of plutonium'is a comparison between plutonium effects and typical fission product effects. For this purpose, plutonium is compared to iodine-131

~

and strontium.--The relationship of the concentration of each material to the

- 28'-

~

l

?

(

i

~

Table VI Pu02 AND Sr0 VAPOR PRESSURES vs. TEMPERATURE Pu0 ")

Sr0(5) 2 T, 'C P, atm P, atm

-14

-11 1200 s 6 x 10 s 2 x 10

-0

~

1600 6 x 10 7 x 10 2000 4 x 10~

5 x 10~

-6

~

2200 4 x 10 6.5 x 10

-5

-2 2400 3.5 x 10 6 x 10 260d 2 x 10~

.35 2800 9 x 10~

> 1.0 4

J. E. Battles, et. al., "A Mass Spectrometric Investigation of the Volatization Behavior of (Uo gPu0.2) 0

-x," Plutonium 1970 and 2

Other Actinides, Nuclear rietallurgy, Vol.17, Ediced by w'.

N. Miner, October, 1970.

5 A. Classen and C. F. Vechemans, " Vapor Pressure Determination of Ba0, S ' and Ca0 and their Mixtures from Measurements of Evaporation,"

Z. PF Vol. 80, pp. 342-51 (1953).

(,)

)

concentration limits given in 10 CFR 20, Appendix B, Tr.ble II are used as a basis of comparison. The results of this comparative analysis are insensitive to the specific reactor design within broad limits.

Conservatively, the results of an analysis of a reactor operating at 3000 MWt, with the core exposed to 35,000 M'JD/MTM and a containment volume of 31,000 cubic feet, are assumed to apply and are discussed below. To estimate the quantities of I-131, strontium, and plutonium evolved, the "non-volatile" materials -(Sr-90 and Pu) are assumed to reach equilibrium vapor pressures in the containment volume.

Two hypothetical temperatures, 1600 C and 2300 C are chosen for illustrative i

purposes. To assure a large measure of conservatism in the conclusion, it is further assumed that the modest decontamination factors allowed for I-131 (plateout and rainout) also apply to the "non-volatile" materials (Sr-90 and Pu02). This assumes that all three materials are released to the environment in the same proportions as they are present in the containment volume.

Utilizin the above assumptions, the calculations indicate that:

The entire core inventory of iodine-131 is released to the con-tainment at both 1600 C and 2300 C.

0.0048% of the strontium is released to the containment at 1600 C, and all of the strontium is released at 2300 C.

Only 2.6 x 10~7% of the plutonium is released to the containment e

at 1600 C, and only 0.0034% of the plutonium at 2300 C.

The resulting activity concentration of the three materials in the containment volume are calculated to be:

)

1.f l

[f

)

Material

.1600 C 2300 C.

1-131 99 ci/t 99 ci/t

~

~

Sr-89 & 90 7 x 10 ci/t 14.5 ci/t

~9 Pu (all isotopes) 6.7 x 10 ci/t 1.7 x.10~

ci/t A measure of the biological importance'of each of these materials i

may be obtained by relating the above concentrations to their respective 10 CFR :

limits and normalizing these ratios to unity for I-131.

The results of this normalization are:

i Relative Biological Importance Material 1600 C 2300 C I-131 1.0 1.0

-5 Strontium 2 x 10

~

~

Plutonium 1 x 10 3 x 10 It may be seen.that, for those hypothetical accidents which might l

approach the limiting. guidelines in 10 CFR 100 with respect to deses to the thyroid from iodine-131 (300 rem), the doses from plutonium would be biologi-cally insignificant.

Fu:.Lher, any mechanisms uhich c.ight selectively minimize the offsite radiological effect.s of the "non-volatile" fission products such as higher plateout and. rainout factors or high efficiency filtering, would also l

j act to further diminish the radiological effect of plutonium.

Finally, the con-l centration of plutonium in a mixed oxide-fueled reactor core is only approxi-i l

mately a fact or of two to three greater than that in a l'0 -fueled core; hence, 2

j in view of the already minor relative importance of the plutonium radiological i

{

effects, the impact of substituting a mixed oxide for a UO2 fuel core is i

negligible.

i i

J '

t 4

c

(

)

-B.

PLUTONIUM RELEASE INTO PRIMARY COOLANT Corrosion studies of UO2 and Pu02 fuels in high temperature water indicate little difference in the corrosion characteristics of the fuels.6,7 A direct experiment in the Eng'neeri..g-Test Reactor to measure the release of radionuclides from defected mixed oxide fuel has been performed.8 The experiment utilized defected mixed oxide fuel (4 w/o Pu0 in UO ) which was 2

2 pre-irradiated to about 2000 MKD/1RM, then irradiated at about 12 kw/ft while defected for over 1000 MWD /MTM (Cycle I) and then finally irradiated at about 20 kw/ft (center melted) for almost 1000 MWD /MTM (Cycle II). The 250 C coolant in the loop was sampled during the periods of defected operation and analyzed for radionuclide content.

Table VII is a reproduction from the referenced report.8 Several observations and interpretations of the data can be made:

1) The ratio of plutonita to uranium in the coulant appears to be approximately the same as in the pellets, although the plutonium concentrations were below the threshold of accurate measurement. This tends to confirm that the removal mechanism is crosion/ corrosion and/or that the solubilities of the two oxides are approximately the same.
2) A comparison of selected isotopes (I-131, Cs-137, Pu-239, and U-238 to their respective 10 CFR 20 3imits is shown below.

It is apparent that the concentration of plutonium is very small 6

M. Zambernard, Development of Plutonium-Bearing Fuel Materials.

Prorress Report. Januarv 1 through March 31, 1963.

NUMEC-P-104, Nuclear Materials and Equipment Corp., Apollo, Penn., March 31, 1963.

7 N. Zambernard,. Development of Plutonium-Bearing Fuel Materials.

Progress Report, April 1 through June 30, 1963.

NUMEC-P-105, Nuclear Materials and 1:quipment Corp., Apollo, Penn., June 30,~1963.

8 M. D. Freshley, The Dnfect Performance of UO3-Puo? Thermal Reactor Ftnel, BNRL-SA-4138. Battelle Pacific Northwest Laboratories, November, 1971. :

8-q P

Table VII LOOP COOLANT ANALYSES OBTAINED DURING CYCLE II AND CYCLE III OPERATION OF THE MIXED-OXIDE PETECT EXPERIMENT (Ref. 8)

Element Cycle II Cycle III*

5

-Xe-133 1.61 x 10 d/s/ml 6 x 10 d/s/ml 4

3

-135-7.67 x 10 d/s/ml 2 x 10 d/s/ml 3

2 Kr-85m 3.54 x 10 d/s/mi 4 x 10 d/s/ml

-87 2.71 x 103 d/s/mi 0

1 x 10 d/s/mi 4

-88 1.23 x 10 d/s/ml 3

3 I-131 2.52 x 10 d/s/ml 1.6/10 d/s/ml

-132

.Present Present I

-133 4.49 x 10 d/s/mi 2 x 10 d/s/ml 3

3 3

-134 1.77 x 10 d/s/mi j

-133 2.58 x 10 d/s/mi 3

Te-131 Present 3

3

-132"

< 1 x 10 d/s/mi 1 x 10 d/s/mi Cs136

< 1 x 10 d/s/ nil

-137 1 x 10 d/s/mi 0

-138 1.16 x 10 d/s/ml 5 x 10 d/s/mi Rb-88 Present l

~

U-238 1.3 x 10 g/L 3.5 x 10~ g/t

-8 U-235 1.0 x 10 g/L 2.5 x 10 g/t

~

~0 Pu-239

< 1. 0 x 10 g/t

, < 1 x 10 g/t

~

c 4

}

  • The fission product analysis for the Cycle III sample is only semi-quantitative because of a malfunction of the counting equipment.

4 e

t j

i '

4 l

l l

(.

and that its relationship to its limiting concentration is far below those of the fission products.

Fission products will, therefore, control the e.anagement and releases of radioactive wa c. t e s.

Measured 10 CFR 20 Isotope pCi /m?.

Limit Ratio I-131 s.05 6 x 10-5 (insol) 830 3 x 10" (sol) 1.7 x 10 Cs-137 s.003 4 x 10-5 (insol) 75 2 x 10-5 (sol) 150 Pu-239

< 6. 2 x 10~

3 x 10-5 (insol)

<.02 5 x 10-6 (sol)

<.12

-0

-5

-6 U-238

.84 x 10 4

10 2 x 10 C.

PLUTONIUM AND TRANS-PLUTONIUM ISOTOPIC CONSIDERATIONS It is poccibic to ascess in gencrcl terr.s the potential 1cng-tcrra effect of plutonium recycle in terms of the relative inventory of trans-uranium isotopes which would be availcble if plutonium were either a) stored or b) recycled.

Calculations have been performed to identify the isotope buildup and burnout in both cases., assuming a typical light water reactor environment.

Table VIII gives the feed and discharge isotopic concentrations in typical UO 2 (3.1 w/o U-235) and reixed oxide (2.74 w/o Pu) fuel rods irradiated to an expo-sure of 30,000 M1lD/MTM. The trans-plutonium isotopes which could contribute significantly to a long-tern radiation ha::ard are included.

As a basis of comparison of the relative biological effect of recycling n not recycling plutonium, the radiological dose resulting from exposure to the trans-uranium isotopes contained in a unit quantity of each type of fuel has been computed.

Values for Relative Biological Effectiveness,

)

)

s

?

(

a TABLE VIII o

FEED AND DISCHARGE IS0 TOPICS (Grams / Metric Ton Metal)

UO Fuel Mixed Oxide 2

Iso. ope 1/2 life (yrs _)

In Out In Out Pu-238 89 102.7 433.1 680.9

-239 24,360 4,992.6 21,089.2 10,724.8

-240 6,760 1,986.8 8,389.7 8,262.9

-241 13 1,151.4 4,863.0 4,941.1 0

-242 3.79 x 10 395.5 1,669.4 2,902.1 6

Np-237 2.14 x 10 378.6 140.3 An-241 458 22.7 245.2

-243 7,650

.36 5.59 Cm-242

.45 11,3 142,8

-243 32

.07 1.21

-244 18.1 15.0 335.6 8

U-235 7.13 x 10 31,000 8,818.1 6,849.9 3,334.0

, )

i

(.

g'

' fractional deposition in the. bone (assumed to be the critica) organ), and the effective half-life of each isotope were taken from-ICRP Publication 2, 1959.

Table IX-shows the relative biological effectiveness of the quantities of trans-uranium. isotopes. given in Table. VIII, normalized to unity for the plutonium isotopes discharged with UO2 fuel:

TABLE IX RELATIVE RADIOLOGICAL HAZARDS T

Discharged UO2 Fuel - Pu Isotopes 1.0

- Trans Pu Isotopes

.115 Charged Mixed Oxide Fuel - Pu Isotopes 4.224 Discharged Mixed Oxide Fuel - Pu Isotopes 4.862

- Trans-Pu Isotopes 1.917 These vclues must be recast to reflect the relative effect of recycling a unit qucntity of plutoniu=, cs follows:

~

1) If plutonium is nor recycled (UO2 fuel is charged):

The amount of plutonium stored which would otherwise be recycled would be; 4.224 The amount of trans-plutonium isotopes stored which were discharged with the above quantity-of plutonium would be; '(4.224 x.115) 0'.460 J

i -

The amount of plutonium generated in the unit quantity of UO2 fuel at goal exposure would be; 1.000 The amount of trans-plutonium isotopes genera-i ted in the unit quantity UO2 fuel at goal l

exposure would be; 0.17 ;

5.799 t

4 i

1 t

I,

Y,.I

2) If plutonium is recycled; The. amount of plutonium stored would be; 0.0 The amount of trans-plutonium isotopes stored would be; 0.460 The amount of plutonium discharged with the unit quantity of mixed oxide fuel at goal exposure would be; 4.862 The amount of trans-plutonium isotopes generated in the mixed oxide fuel at goal exposure would be; 1.917 7.239 These results indicate that the total accumulation of plutonium and trans-plutonium isotopes, expressed in terms of their relative biological-effectivenss, will be 25 percent greater if plutonium is recycled than if it is only stored.

It may bc more valid to exclude from relative radiatica hazardo considerations those isotopes with quite short half lives (say, -less than 100 years) on the basis that environmental concerns center on whether controls can be aseured for several centuries, not merely years.

If this revision is made, the relative radiation hazard in the recycle case is 24% less than'in the non-recycle case. This reversal results from elimination of Pu-238, Pu-241, and the curium isotopes.

These results serve to pisce some perspective on the general effect of plutonium recycle with respect to the general risk of population exposure to the long-lived trans-uranium elements.

While the specific results will vary with the actual fuel design chosen, it appears that introduction of plutonium recycle in light water power reactors does not have a significant effect in this area.

t

L 1

D..

FISSION PRODUCT YIELD CONSIDERATIONS The distribution af fission products is a function of the atomic e

- mass of the fissioning isotope.

The three fissile isotopes of 15terest are U-235, Pu-239 and Pu-241. The fission product yields for these three (and other) isotopes were measured by Lisman, et. al.10 Their results are listed in Table X.

To provide an assessment of the importance of the changes-in fission product distribution without complicating the analysis with assumptions about release mechanisms, reactor operating levels and times, fission product decay time, fuel feed isotopics, enrichment levels, and exposures, the following

- assumptions were used:

The yields for each fissile isotope are considered separately.-

Biologically significant elements at each mass number are chosen

=

for consideration without regard to half-life.

Isotopes, half-lives, fractions deposited in each organ, and relative biologi-cal effectiveness are taken from ICRP Publication 2, 1959.

Ingestion by inhalation of soluble compounds is assumed.

Relative radiation doses are calculated on a unit fission basis l

.to allow comparison of the effects of each fissile isotope.

The results of this calculation, in terms of relative lifetime dose, normalized to unity for U-235, are shown in Table XI.

e l.

T 10 F. L. Lisman,'ets for Thermal Neutron Fissioned 232 Fission Yields of Over 40 et. al.

Fission Produ U, 235,.239 u, U

P and 241Pu, and Fast Reactor Fissioned 235U and 239 u.

Nuclear Science P

and Engineering,. 42, 191-214, 1970.

1 t i

1

(,

Tabic'X A

FISSION PRODUCT YIELDS'FROM U-235. Pu-239 and Pu-241 t

i Mass No. >

235 239 241 83

.529

.301 200 84 1.01

.487 353 85 1.33

.574 387 86 1.95

.770 601-87 2.57

~

1.0 741 88 3.61 1.35 954 90 5.93 2.09 1.53

'91' 5.92 2.52 1.82 92

5. 9 8 --

3.02 2.23-93 6.37 3.95 2.90 94 6.45 4.50 3.33 95 6.51 4.86 3.92 96 6.26 5.12 4.33 97 5.92 5.64 4.76 98 5.83 99 6.24 6.17 100 6.30 101 5.08 6.50 5.94 102 4.21 6.65 6.32 104 1.83 6.61 6.80 106

.389 4.55 6.08 125

(.013)*

(.072)*

.116

.042 131 2.86 3.60 3.15 j

132 4.27 5.09 4.64 133 6.76 7.18 6.71

-134 7.73 7.20 8.06 137 6.32

-6.74 6.60'

  • Values taken frota ANL-5800, Second Edition.

- 39'-

h q

.m

(.,

[

.Tcbic X (Cont.)

p.

Mass No.

235' 239 241 J138.

6.33 5.40 6.37.

140-6.35' 5.61 5.86 141 5.53 4

P 142 5.90 5.04 4.80.

j.

143 5.92 4.48 4.~48 144 5.45 3.78 4.13 4;'

145-

3.89
3. 03 '-

3.19.

146

-2.97 2.49 2.68-j 147 2.14

2.15 2.22

]-

148 --

1' 70 1.70

'1.89 149 1.01 1.24 1.43 150

.64

.965 1.16-i 151

.409

.811 l

152-

.213

.581

.725 154

.056

. 270

.370-i' 3

,1 k

i 4

e o

j i

f 4

  • l

g5_

Table XI U-235 Pu-239 Pu-241 Whole Body.

1.0

.73

.74 i

Thyroid 1.0 1.18

~ 1'. 0 7 Bone 1.0

.50

.47 The calculations show that the thyroid dose is dominated by the iodine isotopes. Since the mix of the~ iodine isotopes ingested is affected by many; variables, the iodine isotope affected the most is considered as the limiting case. This isotope is I-131, whose yields from Pu-239 and Pu-241 fissions are 26% and 10% higher, respectively, than the yield from U-235

-fission.

The calculatlous show the bone dose is dominated by the strontium isotopes and cerium-144.

If the do'se were-due entirely to Sr-90 ingestion,

-the values for Pu-239 and Pu-241 would be.35 and.26 normalized to unity for U-235.

If, on the other hand, the dose were due entirely to ingestion of i

cerium-144, the values for Pu-239 and Pu-241 would be.69 and.76.

Recognizing that the fission product inventory in an actual operating j

[

reactor core will be a mixture of yields from all the fissioning isotopes, and J

i J

l that i '

the variables of power level, exposure, operating time, decay time, and 2

release and dispersion mechanisms can result in very large variations in the relative quantities of the various fission products actually ingested, it can 3,

1:

l be concluded that any actual dose variations introduced by changing from UO2

+?

.to mixed oxide fuel will be insignificant.

1 e

.~

, -,~.,

,,,,e e

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