ML20153F196

Vessel Fluence Reduction Fuel Cycle Study

ML20153F196
Person / Time
Site:	Rancho Seco
Issue date:	12/31/1985
From:	Hassan H, Rodes J, Walters J BABCOCK & WILCOX CO.
To:
Shared Package
ML20153F180	List: ML20153F175 ML20153F196
References
BAW-1884, TAC-59976, NUDOCS 8602250333
Download: ML20153F196 (34)

Text

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BAW-1884 rl December 1985 I

I Vessel Fluence Reduction I Fuel Cycle Study I

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I BAW-1884 December 1985 I

I I VESSEL FLUENCE REDUCTION FUEL CYCLE STUDY I J. R. Rodes Fuel Management Analysis and Services I

Prepared for ~'

DUKE POWER COMPANY SACRAMENTO MUNICIPAL UTILITY DISTRICT 422 South Church Street 6201 "S" Street Charlotte, North Carolina 28242 P. O. Box 15830 Sacramento, California 95813 I

I Reviewed and Approved by:

S .MCh Date\klekb Manager, Fuel Engineering '

W G, / Date //fof<%

Manager, Fuel Management '

I Analysis & Services ll.4.(b $ Date llhl76 I P jec't Manager, Engineering Services /

I BABC0CK & WILC0X Nuclear Power Division I P. O. Box 10935 Lynchburg, Virginia 24506-0935 I Babcock &WHcom a MCDermott Company

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I CONTENTS Page

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

2. FUEL SHUFFLE SCHEMES . . . . . . . . . . . . . . . . . . . . . . . 2-1

3. CALCULATIONAL PROCEDURE ..................... 3-1

4. RESULTS ............................. 4-1 4.1. P0Q07 and FCYCLS Resul ts . . . . . . . . . . . . . . . . . . 4-1 5 4.2. ADJ Results ........................ 4-3 4.3. Vessel Fl uence Resul ts . . . . . . . . . . . . . . . . . . . 4-3

5. OPERATING LIMIT ASSESSMENT . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Base Limits ........................ 5-1 I 5.2.

5.1.1. LOCA Based Limits . . . . . . . . . . . . . . . . . .

5.1.2. IC-DNB Based Limits . . . . . . . . . . . . . . . . .

Margin Improvers . . . . . . . . . . . . . . . . . . . . . .

5-1 5-1 5-2

6.

SUMMARY

AND CONCLUSIONS ..................... 6-1

7. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 List of Tables Table 3-1. Design Criteria and Guidelines . . . . . . . . . . . . . . . . . 3-3 4-1. Comparison of Vessel Fluence Reduction Schemes . . . . . . . . . 4-5 I 4-2.

4-3.

Fast Flux at the Vessel Wall . . . . . . . . . . . . . . . . . .

Estimate of Additicnal Years to Reach Screening Criteria for Oconee 1 and Rancho Seco .............

4-6 4-7 5-1. Margin Improvers . . . . . . . . . . . . . . . . . . . . . . . . 5-2 I

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I List of Figures Figure Page l

2-1. In-Out-In Fuel Loading Diagram . . ............... 2-2 2-2. In-In-Out Fuel Loading Diagram . . . . . . . . . . . . . . . . . 2-2 g 3-1. Vessel Fluence Calculations .................. 3-4 4-8 5

l 4-1. LBP Cycle D, Assembly Burnup Distribution ...........

4-2. VLL Cycle D, Assembly Burnup Distribution ........... 4-9 4-3. ULLNAT1 Core Loading and Assembly Burnup Distribution ..... 4-10 l 4-4. ULLNAT2 Core Loading and Assembly Burnup Distribution ..... 4-11 4-5. ULLBP1 Core Loading and Assembly Burnup Distribution . . . . . . 4-12 4-6. ULLHBU Core Loading and Assembly Burnup Distribution . . . . . . 4-13 $

4-7. LBP Loadings . . . . . . . . . . . . . . . . . . . . . . . . . . 4-14 5 4-8. Cycle Average RPDs for LBP and VLL Reference Cycles ...... 4-15 4-9. Cycle Average Assembly RPDs for Vessel Fluence Reduction Schemes ....................... 4-16 I

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I I 1. INTRODUCTION This vessel fluence reduction fuel cycle study is Phase 2 of a three-phase project designed to reduce high-energy (>1.0 MeV) neutron fluence on reactor vessel weld material . Of concern is the ability of the vessel weld material to conservatively withstand pressurized thermal shock (PTS), while undergoing increasing embrittlement induced by high-energy neutrons, over the planned I lifetime of the plant.

In Phase 1, a computer code, ADJ, was developed to economicrlly correlate the I power production in specific fuel assemblies to the fast flux at the reactor vessel inner wall .1 ADJ uses specially prepared PDQ07 power distribution data, combined with a data file of adjoint fluxes from 00T computer runs generated for specific azimuthal angles relative to the core major axis, to calculate both the fast flux at a specific weld location and the fraction of flux contributed by each assembly. The weld locations considered were at 0, 11, 14, and 19 degrees relative to the core major axis. These angles represent weld locations of interest for the Rancho Seco, Three Mile Island Unit 1, and Oconee Unit I reactor vessels.

Phase 2, reported herein, consisted of developing several fuel cycle loading patterns to specifically reduce the fast neutron fluence at the aforementioned weld locations through reducing peripheral assembly power densities over that I achieved with the very low leakage (VLL) fuel management scheme reported in reference 2. This was achieved by placing 1) highly burned fuel, 2) fresh lumped burnable poison (LBP) in burned fuel, or 3) fresh fuel containing

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natural uranium in peripheral assembly locations H15, K15, and L15 (and l

symmetric locations). Then, the ADJ code was used to assess the fluence reduction for eacn fuel cycle shuffle scheme. In addition, an analysis of the Technical Specification operating limits was addressed in section 5. The detailed calculations for the Phase 2 analysis are documented in reference 3.

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I Phase 3 will consist of plant- and cycle-specific work to be determined after the completion of Phases 1 and 2. This would relate the fuel cycle design to g reach a given fluence reduction to the specific characteristics of each plant. 5 I

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2. FUEL SHUFFLE SCHEMES I

The basis for all core shuffle patterns was the equilibrium VLL Cycle D \

developed in reference 2. This scheme employs an in-in-out fuel shuffle arrangement. Fresh fuel, typically containing LBP to control power peaking, is placed in the core interior, intermixed with once-burned fuel assemblies, i Twice-burned fuel is placed on the core periphery. This arrangement achieves a significant rede: tion in core radial neutron leakage relative to the in-out-in (LBP or low leakage) shuffle scheme predominantly used in all B&W operating plants. Typical examples of LBP and VLL fuel shuffle schemes are illustrated in Figures 2-1 and 2-2, respectively.

I Before investigating specific vessel fluence reduction shuffle schemes, the

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resul ts produced in Phase 1 were evaluated to determine the relative contribution of each fuel assembly to the total fluence at a given weld location1 . The weld locations analyzed in Phase 1 were limited to angles of 0 to 19 degrees relative to the major axis (specifically 0, 11, 14, and 19 degrees). Using the average power distribution of Oconee 1, cycles 3 through l

7 as a reference case , assemblies H15, K15, and L15 (and symmetrics) were found to contribute most of the fast fluence. The combined total contribution I of these assemblies ranged from 86% at 0 degrees to 63% at 19 degrees off the major axis. Therefore, emphasis was placed on reducing the power only in HIS, K15, and L15, rather than at other locations on the periphery.

Starting with the base VLL shuffle scheme, several methods were employed to reduce the power in locations H15, K15, and L15: 1) highly burned fuel, 2) burned fuel containing fresh burnable poison rod assemblies (BPRAs), and 3) natural uranium in place of enriched uranium.

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I Figure 2-1. In-Out-In (LBP) Fuel Figure 2-2. In-In-Out (VLL) Fuel Loading Diagram Loading Diagram I l

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3. CALCULATIONAL PROCEDURE The calculations performed for each fuel shuffle scheme proceeded as shown in Figure 3-1. The FCYCLS code was used to develop most of the shuffle schemes for subsequent analysis with P0Q07. FCYCLS is a fast-running, one neutron energy group, nuclear analysis tool for calculating two-dimensional radial I assembly average power and burnup distributions. Various trial shuffle schemes were depleted for one cycle. Those schemes showing promise for low peripheral assembly power densities and acceptably low maximum peaking were lI subsequently evaluated with PDQ07.

g The P0007 model employed for this study uses two neutron energy groups to E calculate two-dimensional pin-by-pin power distributions in quarter-core geometry. Each shuffle scheme was depleted to 415 effective full power days (EFPD) for comparison to the base VLL scheme. The design criteria listed in Table 3-1 were used as a guideline for this study and are consistent with that used in reference 2 (VLL scheme), with the exception of assuming a higher allowable maximum fuel assembly burnup. The key parameters compared in the base VLL scheme and the various vessel fluence reduction schemes included peripheral assembly peaking, maximum peak pin, maximum assembly burnup, and I cycle length. In addition, an assessment of the Specification operating limits is addressed in section 5. Each shuffle scheme impact on Technical was depleted for one cycle. One cycle was deemed sufficient to establish the relative merits of each shuffle scheme relative to the VLL. The determination of equilibrium cycle lengths and power peaking is highly dependent on specific plant and cycle conditions and would be performed in Phase 3.

Vessel fluences were calculated from P0Q07 power distributions through a three-step process. The PINPOW code was used to convert PDQ07 partition powers to pin powers. Then, the pin powers for each depletion time step were input to the SORREL code. SORREL calculates the cycle-average pin powers and converts these pin powers froa x-y geometry to r-theta geometry. ADJ then I

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combines the r-theta pin power distributions from SORREL with the adjoint flux I

from a 00T-prepared data file and calculates the fast flux at a given weld location and the fraction of flux contributed by each fuel assembly. Further details concerning ADJ can be found in reference 1.

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I I Table 3-1. Design Criteria and Guidelines I 1. Peak pin power calculational limits Inner flow zone - 1.587 Outer flow zone - 1.507

2. Maximum FA burnup calculational limit s50,000 mwd /mtU

3. Depletion of reactivity conditions l

1a 17 ppm boron APSRs inserted (L12 and symmetric locations)

Hot full power, equilibrium xenon

4. No thermal-hydraulic feedback

5. Constant feed batch configuration of 64-FAs at 3.36 wt % U-235.

, 6. All full-length control rods fullv withdrawn.

7. Power level - 2568 MWt.

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I Figure 3-1. Vessel Fluence Calculations Base VLL F

I FCYCLS I

PDQ07 e I Evaluate Peaking g

& Cycle Lifetime 3 y

PINPOW g

SORREL y

ADJ - - - DOT

! Adjoint Fluxes I

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4. RESULTS I

This section describes the P0007 and ADJ results for the various vessel l

fluence reduction schemes relative to the base VLL design described in reference 2. In addition, similar results from the LBP shuffle scheme 2 are included since this type of scheme is typical of that currently used in B&W operating plants. The core loading diagrams for each of these schemes are

'I shown in Figures 4-1 and 4-2.

4.1. P0007 and FCYCLS Results Four vessel fluence reduction schemes were investigated in detail. Each scheme was a basic modification to the VLL scheme, with the specific intent to reduce the power and flux in locations H15, K15, and L15 (and symmetrics). In I two schemes, designated as ULLNAT1 and ULLNAT2, assemblies containing natural uranium, rather than conventional enriched uranium, were used to reduce peripheral powers. Another scheme used fresh LBP inserted in the burned fuel in these locations (ULLBP1), and the fourth scheme used very high burnup fuel (45,000 mwd /mtV) to lower peaking (ULLHBU). The core loading diagrams for each of these schemes are shown in Figures 4-3 through 4-6 and the LBP loadings are shown in Figure 4-7.

Each scheme was depleted to 415 EFPD, then compared to the VLL scheme. Table 4-1 shows the key parameters of maximum pin peak, K15 and L15 average RPO (indicative of the peripheral peaking most affecting vessel welds of interest), maximum assembly burnup, cycle length impact relative to the VLL I scheme, and relative negative imbalance limit impact.

comparison, data from the LBP shuffle scheme are also included. Detailed For additional power peaking data for each vessel fluence reduction scheme can be found i..

reference 3. Figures 4-8 and 4-9 give the cycle-average assembly radial RPDs (eighth-core) for the LBP and VLL schemes, and the vessel fluence reduction schemes, respectively.

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I As shown in Table 4-1 and Figure 4-9, each fluence reduction scheme, particularly the natural uranium scheme designated ULLNAT1, produced substantially lower peripheral peaking than the VLL. The peripheral peaking, K15 and L15 average of 0.178, is the lowest of all patterns investigated. While the peak pin (1.551) meets the peaking guidelines given in Table 3-1, subsequent evaluation (see section 5) suggests that this scheme would require more restrictive operational limits. Another natural uranium pattern, ULLNAT2, was developed that would achieve a peak pin comparable to the VLL scheme, but at the expense of slightly higher peripheral power and shorter cycle length.

The LBP scheme, ULLBP1, with 1.8 wt % B4 C BPRAs produced a peak pin of 1.534 l

This peak is higher than desirable but through pattern optimization the peak pin could be reduced to that comparable to the VLL, yet maintain approximately the same peripheral power. The high-burnup scheme, ULLHBU, was depleted with H15, K15, and L15 starting with assembly burnups of 45,000 mwd /mtU (3.36 wt %

U-235 initial enrichment). The FCYCLS code was used to calculate this cycle -

because of the ease with which high-burnup fuel could be modelled. Since FCYCLS is a nodal code, it only calculates assembly average RPDs. However, by careful comparison to the P0Q07 calculations in the other shuffle schemes, an accurate estimate of 1.505 for the pin peak was made. The peripheral RPD of 0.246 for VLLHBU is comparable to the 0.228 value of the ULLBPI scheme.

Starting with an assembly burnup of 45,000 mwd /mtu, this scheme naturally produced the highest end-of-cycle (E0C) assembly burnup of all schemes with 48,753 mwd /mtV. Burnup limits for fuel assembly designs currently undergcing irradiation do not allow burnups this high, but burnup limits for future assembly designs can allow limits in excess of 50,000 mwd /mtV. Al ternatively, g fuel of lower initial enrichment and burnup could provide the equivalent B reactivity for lowering peripheral RPDs.

All four schemes exhibited cycle lengths of 10 to 15 EFF0 shorter than the VLL because of the reduced reactivity contribution from H15, K15, and L15. The 10 to 15 EFPD should be viewed with caution since the vessel fluence reduction schemes are not " equilibrium" cycles and are therefore not directly comparable to the VLL. However, it gives an indication of the relative cycle lengths achievable between the various designs.

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I I 4.2. ADJ Results I Following the PDQ calculations, ADJ was run for each scheme to calculate the fast flux (>1.0 MeV) at angles of 0,11,14, and 19 degrees relative to the major axis. The fluxes for the LBP and VLL designs were also computed for I compari son , as were the results obtained from reference 1 for the average power distribution of Oconee Unit 1 cycles 3 through 7. These cycles of Oconee represent a composite of out-in and LBP type shuffle schenes and are I

typical of the average peripheral peaking experienced by many operating reactors to date. The resulting fast flux data for the schemes above are compared in Table 4-2. Also shown is the ratio of calculated flux to the corresponding value of both the VLL scheme and the average of Oconee 1 cycles 3 through 7. These ratios illustrate the substantial reductions achieved both with the VLL relative to earlier fuel management schemes, and I the var ious vessel fluence reduction schemes relative to the VLL. For example, the ULLNAT2 pattern was 20 to 30% lower than the VLL scheme and 60 to 66% lower than the average of Oconee 1 cycles 3 through 7 between the angles of 0 to 19 degrees. The preceding fluxes calculated with ADJ assume 2568 MWt operation and a flat axial core power shape. In addition to overall flux reduction, these results indicate that the angular position of peak vessel flux changes from approximately 11 to 19 degrees (and possibly higherangles). Extending the ADJ results to actual, yet conservative, operating conditions is addressed in the next section.

4.3. Vessel Fluence Results This section presents specific examples of the vessel fluences calculated for the various fuel shuffle schemes and the impact of these schemes on vessel li fetime , relative to Nuclear Regulatory Commission calculations (NRC) presented in their "NRC Staff Evaluation of Pressurized Thermal Shcck."4 Rancho Seco and Oconee 1 were selected for specific analysis; however, the ADJ I results can easily be applied to other B&W plants as well.

In the previous section, the fluxes calculated with ADJ were based on 2568 MWt operation and two-dimensional RPDs. Extending ADJ to actual operating conditions requires the addition of an axial shape factor, normalization to I

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actual rated power, and corrections to account for measured versus calculated fluences obtained from specimen capsule analyses. For illustrative purposes only, a conservative average axial shape factor of 1.17 and a 1.26 correction factor based on specimen capsule analyses 5 were applied. The ADJ calculated fluxes were converted to maximum fluences per effective full power year (EFPY) at the limiting weld locations for Oconee Unit 1 (19 degrees) and Rancho Seco (14 degrees). Using the calculated additional fluence required to reach the screening criteria for Oconee Unit 1 and Rancho Seco (updated from reference 4 from December 31, 1981 to January 1, 1986, assuming operation with an LBP low-leakage shuffle scheme), the EFPY needed to reach these criteria were calculated and compared to the remaining EFPY for each plant. For Oconee 1, the remaining EFPY was calculated assuming a 32-EFPY lifetime, and for Rancho Seco, an 80% capacity factor was assumed for g operation over the remainder of the licensed operating period (October 11, 5 2008). The results for each scheme are shown in Table 4-3 for Oconee Unit 1 and Rancho Seco. For both Oconee Unit 1 and Rancho Seco, the results show that converting to the base VLL scheme is sufficient to increase the EFPY to reach the screening criteria well above the remaining plant lifetime.

Consequently, further vessel fluence reduction may not be necessary.

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Table 4-1. Comparison of Vessel Fluence Reduction Schemes (a)

. Maximum Relative Relative Tech.

Assembly cycle length Spec. negative Code Maximum K15 and L15 RPO Burnup, impact, imbalance limit Pattern used pin peak avg. for cycle mwd /mtU EFP0 reduction, %

Reference Patterns L5P P0Q07 1.483 0.484 41112 -5 -1.2 VLL P0Q07 1.501 0.300 42095 --

0.0 Natural Uranium in H15, K15, L15 ULLNATI P0Q07 1.551 0.178 42384 -10 3.3 ULLNAT2 P0Q07 1.511 0.186 44129 -15 0.7 a

Ei Fresh LBP in H15, KIS, L15 ULLBP1 P0Q07 1.534 0.228 42344 -10 2.2 High-Burnup Fuel in H15, K15, L15 ULLHBU FCYCLS 1.505 0.246 48753 -10 0.3 m, k yg (a)64 feed, 3.36 wt % U-235.

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s Table 4-2. Fast Flux (>1.0 MeV) at the Vessel Wall Fast Flux Ratios Shuffle Angle, Fast Flux, Relative to Relative to Scheme degrees n/cm2-secx10~9 VLL Oconee 1 Cy 3-7 g

Oconee 1 Cy 3-7 Average 0 9.417 1.97 3

11 9.979 2.05 5

14 9.920 2,06 --

19 9.321 a

1.98 --

g LBP 0 7.168 1.50 0.76 11 7.421 1.52 0.74 14 7.291 1.51 0.73 19 6.748 1.43 0.72 VLL 0 4.770 --

0.51 11 4.869 --

0.49 14 4.822 --

0.49 19 4.713 --

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ULLNAT1 0 3.171 0.66 0.34 11 3.343 ,

0.69 0.34 14 3.413 4

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5 19 3.669 0.78 0.39 ULLNAT2 0 3.239 '

O.68 0.34 11 3.371 0.69 0.34 14 3.446 0.72 0.35 19 3.752 / 0.80 0.40 ULLBP1 0 3.736 0.78 0.40 11 3.891 0.80 0.39 14 3.918 0.81 E

0.40 5 19 4.043 '

O.86 0.43 ULLHBU(a) 0 4.031 0 85 0.43 11 4.198 0.8ti 0.42 14 4.227 0.88 0.43 19 4.362 0.93 0.47 (a) Estimated using ULLBP1 x 0.246/0.228 I

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I I Table 4-3. Estimate of Additional Years to Reach Screening Criteria (#}

for Oconee 1 (19 degrees) and Rancho Seco (14 degrees)

Add'l fluence EFPY to reach Left Fluence

( Shuffle screencri}gria, 2 Fluenge/EgY, EFPY to reach in plant reduction n/cm x10 factor Scheme n/cm'x10 screen. crit. life Oconee 1 (8.30 EFPY)

NRCestimate(b)0.955 0.054 17.69 23.70 1.34 LBP low leak 0.955 0.03137 30.44 23.70 0.78 VLL 0.955 0.02191 43.59 23.70 0.54 l

I ULLNAT1 0.955 0.01706 55.98 23.70 0.42 ULLNAT2 0.955 0.01744 54.76 23.70 0.43 ULLBP1 0.955 0.01880 50.80 23.70 0.47 ULLHBU 0.955 0.02028(c) 47.09 23.70 0.50 Rancho Seco (5.32 EFPY)

NRC estimate (b) 0.500 0.058 8.62 18.22(d) 2.11 LBP low leak 0.500 0.03659 13.66 18. 2?. 1.33 VLL 0.500 0.02420 20.66 18.22 0.88 ULLNAT1 0.500 0.01713 29.19 18.22 0.62 ULLNAT2 0.500 0.01729 28.92 18.22 0.63 ULLBP1 0.500 0.01966 25.43 18.22 0.72 ULLHBU 0.500 0.02121(c) 23.57 18.22 0.77

(") Screening criteria are as of January 1,1986.

(b) NRC value from the PTS report updated to January 1, 1986.

(c) Estimated.

(d) Based on the October 11, 2008 license expiration and an 80% capacity factor.

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I Figure 4-1. LBP Cycle D, Assembly Burnup Distribution 8 9 10 11 12 13 14 15 l nB I 12 l nB l u l nB l u i 12 l 12 i Hl 30341 l 17353 l 24596 l 0l 24002 l 0l 17366 l 17770 l

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TO T410 Delta Edit Fuel Initial Burnup Burnup Burnup Batch Set Assembly wt% U-235 MM/mtU MNd/mtU MM/mtU l 11B 89 49 3.36 23067 37074 14007 12 90 64 3.36 15995 24759 8764 13 91 64 3.36 0 15997 15997 CORE 53 177 3.36 12169 25000 12831 l I I

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I Figure 4-2. VLL Cycle D, Assembly Burnup Distribution 8 9 10 11 12 13 14 15 I na l 12 l na l u l ns I u i 12 I nn i I Hl l

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l l l l l l Pl 16965 l 0[ 17249 l 17531 l 27500 l l 29273 l 14440 l 27279 l 25015 l 31882 l I I I I l na l na l nB IVIICID Batch ID I Rl 29538 l 31635 l 34891 lVLICID TO Burnup (MNd/mtU) l 34405 l 36192 l 38109 lVLICID T415 Burnup (NNd/mtU) l TO T415 Delta l Edit Fuel Initial Eurnup Burnup Burnup l Batch Set Assembly wt% U-235 MNd/mtU MNd/mtU MNd/mtU llB 89 49 3.36 29251 36803 7552 12 90 64 3.36 16314 30019 13705 8 13 91 64 3.36 0 16430 16430 CDRE 53 177 3.36 M996 26984 12987 I

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Figure 4-3. ULLNAT1 Core Loading and Assembly Burnup Distribution 8 9 10 11 12 13 14 15 l na l 12 l na l u l na i 13 l 12 l 13A l Hl 29919 l 14375 l 26534 l 0l 26482,l 0l 16967 l 0l g l 42384 l 29797 l 41211 l 18564 l 41505 l 17683 l 28954 l 2735 l 5 I I l l l l l I I l 12 l 12 l 13 l 12 l 13 l 12 l u l 13A l Kl 14375 l 17897 l 0l 17950 l 0l 15851 l 0l 0l l 29798 l 33918 l 18590 l 35061 l 18038 l 31661 l 14095 l 2685 l l 1 I I I I I I I l llB l 13 l 12 l 13 l 11B l 13 l 12 l 13A I E Ll 26534 l 0l 17954 l 0l 25152 l 0l 17308 l 0l 3 l 41215 l 18602 l 35500 l 18464 l 38500 l 16060 l 27184 l 1930 l I

13 I

12 I

13 I

12 I

13 l

12 12 I a Ml l

0l l

17962 l l

0l l

17945 l l

0l l

12167 l l

17537 l l

5 l 18569 l 35078 l 18458 l 34834 l 17771 l 26302 l 25005 l Nl l

l na 26482 l l

I u 0l l

l na 25140 l l

l u 0l l

l 12 14373 l 1

l u 0l l

l na 27528 l 1

l l l 41507 l 18037 l 38474 l 17710 l 29637 l 12487 l 31972 l 3 I I I I I l l 5 l u l 12 l u l 12 l 13 l na I ol Ol 15833 l 0l 12123 l 0l 29918 l

.l 17686 l 31647 l 16051 l 26226 l 12447 l 35174 l 1 1 I l l l l 12 l 13 l 12 l 12 l na l Pl 16965 l 0l 17249 l 17531 l 27500 1 E l 28955 l 14099 l 27133 l 24993 l 31942 l 5 I I I I l 13A l UA l 13A lUIINATl Batch ID g Rl 0l 0l 0 lULUGT1 TO Burnup (NWd/mtU) 1931 [UIINAT1 T415 Burnup (NWd/mtU) 5 l 2736 l 2686 l TO T415 Delta Edit Riel Initial Burnup Burnup Burnup Batch Set Assembly wt% U-235 NNd/mtU NNd/mtU NWd/mtU llB 89 29 3.36 26998 37155 10158 12 90 64 3.36 16314 30183 13869 g 13 91 64 3.36 0 16698 16698 5

13A 8 20 0.711 0 2394 2394 CORE 53 177 3.36 10322 23309 12987 l

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I Figure 4-4. ULLNAT2 Core Loading and Assembly Burnup Distribution l

8 9 10 11 12 13 14 15 l nB l 12 l nB l 13 l nB l u l 12 l uA 1 Hl 29919 l 16967 l 26534 l 0l 29918 l 0l 14375 l 0l l 41838 l 31699 l 40927 l 18264 l 44119 l 17503 l 26953 l 2861 l I I I -l l l 1 l- l l, 12 12 13 u l l l 1 12 l 13 1 12 I l uA l

, Kl 16965 l 17897 l 0l 17950 l 0l 15851 l 0l 0l l l 31702 l 33399 l 17940 l 34754 l 17984 l 31878 l 14537 l 2799 l I I I I l 1 I I i I nB l u l nB i u l nB l 13 l 12 1 13A l Ll 26534 l 0l 26482 l 0l 25152 l 0l 17308 l 0l l 40945 l 17980 l 41484 l 17965 l 38568 l 16419 l 27499 l 2008 l ,

I Ml I

l u 0l i

l u 17962 l l

l u 0l l-l u 17945 l l

l u 0l I

l u 12167 l I

l u 17537 l l

l l 18287 l 34797 l 17989 l 34684 l 17805 l 26645 l 25272 l I 1 l 11B I

l u I

l 11B I

[ 13 l

l 12 l

l 13 l

l 11B l

l Nl 29918 l 0l 25140 l 0l 14373 l 0l 27528 l I l 1

l 44129 l u

I I

17995 l 38550 l nB l

l u l

l 17745 l 12 l

l 30005 l 13 l

l 12932 l 12 l

l 32211 l I ol l

1 17511 1 0l 1

15833 l 31868 l I

0l 164u l I

n123 l 26570 l I

0l 12888 l I

17954 l 24796 l I 12 l 13 12 12 nB I

l l l l Pl 14373 l 0l 17249 l 17531 l 27500 l l 26958 l 14542 l 27448 l 25260 l 32179 l 1 I I I I Rl l

l 13A 2862 l 01 l 13A 0l 2800 l l 13A l ULT.NAT2 Batch ID 0 l ULT.NAT2 TO Burnup (Mil /mtU) 2009 l ULT.NAT2 T415 Burnup (Mil /mtU)

TO T415 Delta Edit Fuel Initial Burnup Burnup Burnup Batch Set Assembly wt% U-235 NWd/mtU Mt1/mtU Mil /mtU llB 89 29 3.36 26998 38415 11418 12 90 64 3.36 16314 29596 13282 I 13 91 64 3.36 0 16682 16682 13A 8 20 0.711 0 2496 2496 l CORE 53 177 3.36 10322 23309 12987 l I

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'I 4-11 Babcock & WHcox a McDermott company L

I Figure 4-5. ULLBP1 Core Loading and Assembly Burnup Distribution 8 9 10 11 12 13 14 15 l llB l 12 l 11B l 13 i 11B l 13 l 12 l llB l Hl 29919 l 14375 l 26534 l 0l 26482 l 0l 16967 l 29542 l l 42344 l 29746 l 41150 l 18450 l 41310 l 17058 l 28804 l 33183 l 1 I I I I I I I I i 12 l 12 l 13 l 12 l D l 12 l 13 l llB l Kl 14375 [ 17897 l 0l 17950 l 0l 15851 l 0l 31649 l l 29747 l 33862 l 18515 l 34970 l 17880 l 31498 l 13946 l 35087 l l l l l 1 I I I I l 11B l 13 l 12 l 13 l 11B l 13 l 12 l llB  ! g L[ 26534 l 0l 17954 l 0l 25152 1 0l 17308 l 34874 I E l 41154 l 18527 l 35432 l 18383 l 38434 l 16009 l 27102 l 37353 l l

l I I I I I I I l 13 l 12 l u l 12 l 13 l 12 l 12 l Ml 0l 17962 l 0l 17945 l 0l 12167 l 17537 l l 18455 l 34987 l 18377 l 34770 l 17709 l 26269 l 25025 l l 1 I I I I I I E I llB l 13 l 11B l D l 12 l D l 11B l 5 Nl 26482 l 0l 25140 l 0l 14373 l 0l 27528 l l 41312 l 17880 l 38409 l 17647 l 29589 l 12454 l 31967 l 3 I

13 I

12 I

u I

12 I

u I

11B I E I l I l l l l ol 0l 15833 l 0l 12123 [ 0l 29918 l l 17061 l 31404 l 15999 l 26194 l 12414 l 35158 l 1 I i 1 1 I i 12 l 13 l 12 l 12 l llB l Pl 16965 l 0l 17249 l 17531 l 27500 I E l l 28805 l u950 l 27050 l 25013 l 31937 l 3 l 1 I I l llB 11B 11B l l l lUIIEP1 Batch ID l 34891 lUIIEP1 TO Burnup (mwd /mtU)

Rl 29538 l 31635 l l 33181 l 35075 l 37371 lUIIBP1 T415 Burnup (MNd/mtU)

TO T415 Delta Edit Ebel Initial Bwmup Burnup Burnup Batch Set Assembly wt% U-235 Mt1/mtU NNd/mtU M*3/mtU llB 89 49 3.36 29251 36492 7241 12 90 64 3.36 16314 30112 13798 13 91 64 3.36 0 16575 16575 CDRE 53 177 3.36 13996 26984 12987 1

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I 4-12 Babcock &WHcox a McDermott company

I I Figure 4-6. ULLfiBU Core Loading and Assembly Burnup Distribution 8 9 10 11 12 13 14 13 l 11B l 12 l HB l 13 l 11B [ 13 l 12 l HB l t

I I Hl l

I 29919 l 41800 l l-16967 l 31586 l l

26534 l 40825 l l

0l 18351 l 1

29918 l 44015 l l

0l 17547 l l

14375 l 26843 l I

45000 l 48753 l I

12 I Kl l

l 16967 l l 12 17897 l 31586 l 33281 l l 13 0l 18038 l l 12 17950 l 34509 l l 13 0l 17862 l l 12 15851 1 31662 l l 13 01 13718 l l 11B 45000 l 48650 l l

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1 1 I I I I I l nB l 13 l nB l 13 l nB l 13 l 12 l HB l Ll 26534 l 0l 26534 l 0l 25152 [ 0l 17308 l 45000 l l 40825 l 18038 l 41351 l 17838 l 38667 l 16327 l 27435 l 47732 l I Ml I

l 13 0l I

l 12 17950 l l

I 13 0l I

l 12 17954 l l

l 13 0l 1

l 12

-l 12167 l l 12 17537 l I

l I l I

l 18351 l nB l-l 34509 l 13 1

l 17838 l nB I

l 34274 l 13 I

l 17591 l 12 I

1 26470 l 13 l-l 25473 l nB 1

l Nl 29918 l 0l 25152 l 0l 14375 l 0[ 27528 l I l I

l 44015 l 13 I

l 17862 l 12 I

38667 l 13 I

17591 l 12 I

29999 l 13 I

12926 l I

32303 l l l l l 12 l I o1 l

1 0l 15851 l 17547 l 1

31662 l 1

0l 16327 l I

12167 l 26470 l I

0l 12926 l l

17954 l 24997 l l 12 l 13 l 12 l 12 nB l l Pl 14375 l 0l 17308 l 17537 l 27528 l l 26843 l 13718 l 27435 l 25473 l 32303 l I I I l I

l llB l 11B l 11B lULGBU Batch ID Rl 45000 l 45000 l 45000 lUIIHBU TO Burnup (Mil /mtU) l 48753 l 48650 l 47732 lULGBU T415 Burnup (mwd /mtU)

I TO T415 Delta Fuel Initial Burnup Burnup Burnup Batch Assembly wt% U-235 Mt1/mtU mwd /mtU Mti/mtU llB 49 3.36 34353 42457 8104 12 64 3.36 16322 29505 13183 I 13 64 3.36 0 16531 16531 CDRE 177 3.36 15412 28399 12988 I

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4-13 Babcock & Wilcox a McDermott company

' I Figure 4-7. LBP Loadings 8 9 10 11 12 13 14 15 1.1 0.8 1.8 1.4 H 0.8 1.8 1.8 1.4 1.1 1.1 0.5 K 1.8 1.4 0 1.8 1.4 0 1.R 1.4 0 1.1 0.8 l m

l 1.8 0.8 1.8 0.8 1.8 0.8 1*1 M

1.1 1.1 1.4 0.2 LBP N 0.2 VLL and ULLBPI 0.2 ULLNATI g 0.5 ULLNAT2 and ULLHBU g 0

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4-14 Babcock &Wilcox a McDermott company

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I Figure 4-8. Cycle Average RPDs for LBP e.nd VLL Reference Cycles I 8 9 10 11 12 13 14 15 I H 0.839 0.938 1.037 1.160 1.084 1.104 1,385 1.395 1.160 1.128 1.353 1.314 1.009 0.947 0.548 0.375 I K 1.027 1.345 1.151 1.354 1.129 1.116 0.556 1.205 1.398 1.288 1.359 1.204 1.112 0.351 I

1.152 1.353 1.028 1.244 0.907 0.411 L

1.321 1.390 1.008 1.226 0.772 0.248 I M 1.134 1.273 1.262 1.342 0.998 1.073 0.634 0.577 I N 0.987 0.930 0.412 0.338 1.153 0.944 I 0.499 0 0.397 I

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4-15 Babcock & Wilcox a McDermott company

I Figure 4-9. Cycle Average Assembly RPDs for Vessel Fluence Reduction Schemes 8 9 10 11 12 13 14 15 0.960 1.187 1.130 1.429 1.157 1.362 0.923 0.211 0.918 1.134 1.108 1.406 1.093 1.348 0.969 0.220 E H 0.957 1.184 1.125 1.421 1.142 1.313 0.911 0.280 5 0.915 1.126 1.100 1.413 1.085 1.351 0.960 0.289 1.234 1.431 1.318 1.389 1.217 1.085 0.207 K 1.194 1.381 1.294 1.385 1.234 1.119 0.216 1.229 1.426 1.311 1.377 1.205 1.074 0.265 1.185 1.392 1.275 1.375 1.217 1.056 0.281 1.351 1.422 1.028 1.237 0.760 0.149 L 1.155 1.383 1.033 1.264 0.785 0.155 1.346 1.415 1.023 1.233 0.754 0.191 1.141 1.374 1.041 1.257 0.780 0.210 1.300 1.368 1.098 0.575 M 1.289 1.371 1.115 0.596 E 1.296 1.365 1.086 0.577 5 1.257 1.355 1.101 0.611 1.175 0.961 0.342 E N 1.204 0.996 0.361 5 1.172 0.959 0.342 1.203 0.995 0.368 0.405 ULLNATI e 0 0.527 ULLNAT2 0.403 ULLBP1 0.542 ULLHBU I

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l 4-16 Babcock & Wilcox a McDermott Company

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I l S. OPERATING LIMIT ASSESSMENT I 5.1. Base Limits The most restrictive power peaking-related Technical Specification limits are normally the negative imbalance limits at full power. The two criteria that I determine these are the loss-of-coolant accident (LOCA) kW/f t and the initial condition (IC) departure from nucleate boiling (DNB) maximum allowable peaking (MAP) limits. Each of these criteria will be discussed separately.

5.1.1. LOCA Based Limits I The LOCA kW/ft criteria normally determine the negative imbalance limit. A 1%

increase in the radial power peak will reduce the imbalance limit by approximately 1%. Therefore, the impact on a specific fuel cycle can be estimated by reducing the current offset limit by the amount of margin between 1.501 and the pin peaks given in Table 4-1. For example, if the current limit i is -12% imbalance and the ULLNAT1 option (1.551 pin peak) is used, then the margin difference between 1.551 and 1.501 is 3.3% and the new imbalance limit I will be approximately -12 + 3.3 = -8.7% imbalance. A pin peak of 1.501 is used as tne base since it, like the values in Table 4-1, reflects no thermal feedback. In this example, if a limit of -10% imbalance is required for l operation, then a margin improver that affects LOCA margin by at least 1.3% is needed.

5.1.2. IC-DNB Based Limits The IC-DNB peaking margin is not a strong function of imbalance. Therefore, a l

lack of positive IC-ONB margin will tend to limit the maximum. achievable power

'I rather than restrict the imbalance limits. In this evaluation, neqative margin can be understood as a reduction in the rated power level. The maximum acceptable pin paking for fuel cycle designs is 1.50 I 0.02. This value reflects the closed-channel DNB analysis and no thermal feedback in the I

5-1 Babcock & Wilcox a McDermott company

I peaking evaluation. The 0.02 variation accounts for the variation of the actual margin due to loading differences from one fuel cycle to the next. g Elimination of this unknown requires a three-dimensional evaluation of a a specific fuel cycle.

If the ULLNAT1 option is considered with respect to IC-DNB margin, the margin difference is 3.3%. Since a variation of 2% in the fuel cycle design limit is possible, a 5% IC-DNB should be considered unless each specific fuel cycle design is evaluated for margin during the fuel cycle design process.

5.2. Margin Improvers Margin improvers are available for most of the vessel fluence reduction schemes. The following list of improvers is provided to allow the choice to be tailored to the specific needs of the utility. Each type of limiting criteria needs to be addressed, but the same amount of margin improvement is not necessarily needed for each type, as shown in the examples above.

Table 5-1. Margin Improvers I

Improvement Option LOCA IC-DNB Crossflow with Design Peak Modification 0%- 4%

Statistical Core Design 0% 5%

Fixed Margin Technical Specification 10% 5%

Imbalance Error Reanalysis 2% 0%

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6.

SUMMARY

AND CONCLUSIONS I This report examined several fuel management schemes for reducing the fast flux to the reactor vessel wall for a specified range of weld locations.

These schemes built upon the results attained from the VLL design by further reducing peripheral assembly pcwers in locations symmetric to HIS, K15, and L15. From an analysis of these schemes, the following conclusions are drawn:

1. The base VLL design offers a substantial vessel fluence reduction relative to typical previous fuel cycles, and may be sufficient for some utilities seeking to substantially reduce vessel fluence.

2. Shuffle schemes providing additional fluence reductions up to 30% lower than the VLL, and over 60% lower relative to typical previous cycles, can be implemented without unusual design modifications.

3. These fluer.ce reductions can be realized without significantly affecting Technical Specification operating limits, relative to the VLL design.

However, as addressed in reference 2, implementing the VLL design may entail additional core-specific physics, thermal-hydraulic, and safety analysis work relative to current fuel management schemes. The same is true for the schemes examined in this report. l

4. Further reducing vessel fluence while maintaining maximum pin peaking l comparable to the VLL design inherently requires fuel loadings that j shorten cycle length, for a given feed batch size and enrichment. Cycle l lengths may be reduced up to a maximum of 10 to 15 EFPD relative to that I attainable with the VLL design.

5. Selectively reducing peripheral assembly RPDs can shift the angular location of the peak vessel flux, possibly causing another weld location to be limiting. In addition, decisions regarding placement of vessel I

6-1 Babcock & Wilcox a McDermott company

I cavity dosimetry at the peak flux location may be influenced by the type of vessel fluence reduction scheme to be implemented.

Implementing one of the vessel fluence reduction schemes addressed in this report will necessitate the careful plant-specific evaluation of the desired level of fluence reduction relative to the economic considerations associated with

- Potentially shorter cycle lengths than attainable with the VLL.

- Irradiating fuel assemblies to very high burnup.

- Fabricating fuel assemblies containing natural uranium (or tails). Due to the very low incremental burnup that these assemblies would experience each cycle, the maximum residency time would be much longer than for typical fuel assemblies. The evaluation of maximum achievable residency time would need to be addressed in future mechanical design analyses.

- Alternative schemes, such as combination of very highly burned fuel and fresh LBP, may result in fluence reductions approaching that attained with natural uranium.

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7. REFERENCES

1. N. L. Snidow, Development of Correlation Between Power in Peripheral Fuel Assemblies and Fast Flux at the Reactor Vessel Wall, BAW-1864, Babcock & Wilcox, Lynchburg, Virginia, January 1985.

2. Very Low Leakage Fuel Shuffle Scheme Development Report, BAW-1769, Babcock & Wilcox, Lynchburg, Virginia, January 1983.

I 3. J R. Redes, Vessel Fluence Fuel Cycle Study, 32-1156660-00, Babcock

& Wilcox, Lynchburg, Virginia, July 1982.

4. Enclosure A, NRC Staff Evaluation of Pressurized lhermal Shock, l SECY-82-465, November 1982.  ;

5. Analysis of Capsule OCl-A, Duke Power Company, Oconee Nuclear Power Station Unit 1, BAW-1837, Babcock & Wilcox, Lynchburg, Virginia, l August 1984.

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