ML20128L435

Spent Fuel Storage K 00 Conversion Analyses

ML20128L435
Person / Time
Site:	Peach Bottom
Issue date:	11/30/1992
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20128L425	List: ML20128L422 ML20128L432 ML20128L435
References
DFR-A-5449, DFR-A00-05449, GENE-512-92073, NUDOCS 9302190241
Download: ML20128L435 (8)

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GENuciaarEnergy GENE-512 92073 <

DRF A00-05449 November 1992-Peach Bottom Atomic Power Station C Spent Fuel Storage koo Conversion Analyses >

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SUMMARY

The Peach Bottom Atomic Power Station spent fuel storage rack is a high density storage type which utilizes Boraflex as the neutron poison absorber. De criticality safety analyses results (Reference 1) supporting the high density fuel storage rack design are based on demonstrating compliance to k,y s 0.95 -

using a 7x7 design basis fuel bundle with a uniform rod enrichment of 3.5 wt.% Uranium 235 and no credit for burnable poisons. The equivalent errichment loading for this fuel lattice is 17.3 grams of Uranium-235 per axial centimeter. The enrichment loading of 17.3 g U-235/cm was applied to the Peach ,

Bottom Atomic Power Station Tech Specs at the time of the spent fuel storage rack license approval.

Since it is acceptable to take credit for burnable poisons in the fuel (Reference 2), the maximum intinite lattice km is a tr.cre appropriate parameter than enrichment loading to demonstrate compliance to the fuel storage k,y criteria (s 0.95). The km parameter encompasses all fuel lattice parameters. He effect of high enrichment on the storage rack poison material has been accounted for by a bias correction. The use of a lattice km for demonstrating compliance to fuel storage criticality criteria has been used for all GE-supplied storage racks and is currently used for re-rack designs where the initial enrichment criteria has become limiting.

The infinite neutron multiplication factor (km) has been calculated for the design basis fuel bundle used in the criticality safety analysis for the Peach Bottom Atomic Power Station spent fuel storage racl.s. He calculation consists of an infinite array of design basis fuel in the cold uncontrolled reactor core geometry. The resulting km, including the critical benchmark blas, is 1.390 i 0.005 (t 2a).

Assuming a Normal distribution, a lower bound value can be calculated using 95/95% statistics. The 95/95% lower tolerance limit value was calculated to be km = 1.384. In addition, an enrichment bias correction of 0.022 was calculated to account for spectral effects associated with higher enrichment fuel lattices of up to 4.5 wt.% U-235. Herefore, it is recommended that an incore km limit of 1.384 minus 0.022 or 1.362 be used as the spent fuel storage compliance limit in place of the existing 17.3 g U-235/cm limit for the Peach Bottom Atomic Power Station high density spent fuel storage racks.

GE-supplied low density spent fuel racks have a km compliance limit of 1.30 (Reference 3). Therefore, the Peach Bottom Atomic Power Station spent fuel storage racks are less limiting than the origina! GE-supplied storage racks. The compliance check performed by GE for each GE supplied fuel bundle,-

including Gell fuel, is checked against the fuel lattice criteria of km < 1.30, which conservatively bounds the Peach Bottom Atomic Power Station spent fuel storage rack design.

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GENE-51242073 November 1992 ANALYTICAL. h1ETHOD The infinite lattice km calculation was performed with the General Electric h1 ERIT hionte Carlo neutron transport computer program.; The h1 ERIT program is a hionte Carlo program for solving the linear-neutron transport equation as a fixed source or an eigenvalue problem in three space dimensions. The cross sections in h1 ERIT are processed from the ENDF/B library in the multigroup and resonance parameter formats. Thermal scattering in water is represented by the Haywood kernel obtained from the ENDF/B library. The hfERIT program utilizes 190 full spectrum cross section energy groups. The types of reactions considered in hiERIT are fission, elastic, inelastic and (n,2n).

Neutron absorption is accounted for by reducing the w ight of the neutrons at each collision. When_the weight is reduced sufficiently, the neutrons are terminated by Russian roulette. As part of the solution, htERIT produces eigenvalue, micro- and macro-group fluxes, reaction rates, cross sections, and neutron balance by isotopes.

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METHOD OUALIFICATION Re h1 ERIT program has been thoroughly verified for programming, sampling procedures, particle '

tracking, random number generation, fission source distribution, statistical evaluation, resonance cross section evaluation, edits and other functions of the program. The overall performance of hiERIT and the cross section data was evaluated by comparison against critical experiments which include:

1. CSEWG thermal reactor benchmark problems:

TRX-1, TRX-2, ORNL-1, ORNL-2, PNL-1, PNL-2

2. Babcock and Wilcox Small Lattice Facility

3. Jersey Central Gamma Scan Experiments 4 BWR Gadolinia Critical Experiments

5. Battelle Critical Experiments with Fixed Neutron Poisons

6. Nippon Atomic Industrial Group (NAIG) Critical Experiments with BWR Control Rods -

The analyses of these benchmark calculations were performed using the ENDF/B-IV cross sections. He hiERIT program, based on the ENDF/B-IV cross sections, under-predicts k-effective by approximately1 0.5% for the high moderator-to-fuel ratio in water moderated uranium lattices. The Cross Section Evaluation Working Group (CSEWG) confirmed the same blas (Reference 4) for the ENDF/B IV cross sections. He hiERIT results reported in this summary include the Ak bias correction of 0,0054 i 0.0022 (1a).

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GENE-512-92073 November 1992 CALCULATIONS An incore infinite lattice neutron multiplication factor was calculated for the fuel bundle used in the Peach Bottom Atomic Power Station spent fuel storage rack nuclear safety analysis (Reference 1). De fuel bundle parameters used in the spent fuel storage analyses are contained in Table 1. De fuel bundle k-infinity was calculated using the reactor core geometry consistent with Peach Bottom Atomic Power Station without the presence of control rods. Since an unirradiated high enrichment lattice without burnable poisons is overmoderated in the reactor core geometry, the calculation was performed at 20"C, >

resulting in the minimum design basis km. De calculations were performed with 260 batches of 1000 neutrons, starting with a uniform fission source distribution. The first 10 batches were discarded to eliminate the bias resulting from the initial source distribution. The final results were based on the remaining 250 batches or 250,000 neutron histories. He resulting km value including the critical benchmark bias is 1.390 i 0.005 (2o).

t Assuming a Normal distribution, a lower bound value can be calculated using 95/95% statistics. It is recommended that the 95/95% lower tolerance limit values of km = 1.384 be used as the spent fuel -

storage compliance limit in place of the existing Uranium-235 enrichment limit for the Peach Bottom Atomic Power Station high density spent fuel storage racks.

A km correction factor is applied which accounts for the change in storage rack response to fuel designs of higher enrichments. Higher enrichments, i.e., greater than 3.5 wt.% U 235, result in a neutron spectrum change which reduces the effectiveness of the boron poison. The Peach Bottom Atomic Power Station spent fuel storage rack geometry was analyzed to determine the effect of enrichment. The analyses was performeo with a Gell type fuel lattice, assuming a uniform lattice average enrichment of.

4.5 wt.% U-235. The Gell type fuel lattice was analyzed in both the storage rack geometry and the incore geometry. Small amounts of gadolinia were added to reduce the beginning of life km until the intack km matched that of the design bases analyses using 7x7 fuel at 3.5 wt.% U-235. He corresponding enrichment bias is defined as the difference between the incore km for 3.5 wt.% and 4.5 wt % U-235 at the nominal calculated intack km = .918. Km for 4.5 wt.% U 235 was calculated with varying the number of gaiolinia poison rods in both the incore and the storage rack configuration. The data was fit with a least-squares polynomial curve. The calculated bias is 0.022 Ak and is illustrated in Figure 1.

Therefore, it is recommended that a 0.022 Ak bias be subtracted from the above incore km value in order to account for future high enrichment lattice designs. De lower 95/95 value of 1.384 minus the 0.022 enrichment bias results in a fuel storage compliance limit of maximum incore km s 1.362.

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4 GENE 51240073 November 1992 REFERENCES

1. Philadelphia Electric Company Peach Bottom Atomic Power Station Units 2 and 3 Updated Final Safety Analyses Report, Revision 10, January 1992.

2. " Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Phnts", ANSI /ANS 57.21983. *

3. " General Electric Standard Application for Reactor Fuel", NEDE 240ll P.A 10, Latest Revision,

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4. E.M. Bohn, et.al. (Ed),

• Benchmark Testing of ENDF/fs IV", ENDF 230, Vol. I, March 1976.

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'<T' GENE-51242073 November 1992 Table 1 Fuel Bundle Parameters PARAMETER VALUE Pellet 0.D. (inch) 0.487 Pellet Det.sity (%T.D.) 95.0 Fuel Rod O.D. (Inch) 0.563 - '

Fuel Rod I.D. (inch) 0.499  ;

Fuel Rod Array 7x7  :

Fuel Rod Pitch (inch) 0.738 Average U 235 Enrichment (wt.%) - 3.5 Rod Enrichment Distribution uniform Burnable Poisons none ,

Number of Water Rods none-Channel 1.D. (inch) 5.278 Channel 'Ihickness (inch) 0.100 J

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November 1992 1.5 1.45 -

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1.4 -

.E Enrichment 1 Bias E a 8

,g 1.35 -

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+ 7x7 (3.5 wt.% U-235) 1.3 -

N gel 1 (4.5 wt.% U 235) 1.25 - '

0.88 0.89 0.9 0.91 0.92 0.93 0.94 0.95 0.96 0.97 0.98- 0.99 Storage Rack k-infinity l

I Figure i Fuel Storage Rack Enrichment Bias l

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