ML20115D224

Rev 1 to HI-92880, Criticality SE of Comanche Peak Fuel Storage Facilities W/Fuel of 5% Enrichment

ML20115D224
Person / Time
Site:	Comanche Peak
Issue date:	09/29/1992
From:	Mitchell W, Sarah Turner HOLTEC INTERNATIONAL
To:
Shared Package
ML20115D217	List: ML20115D213 ML20115D219 ML20115D224
References
HI-92880, HI-92880-R01, HI-92880-R1, NUDOCS 9210200292
Download: ML20115D224 (141)

Text

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CRITICALITY SAFETY EVALUATION OF

/

COMANCHE PEAK FUEL STORAGE FACILITIES i

WITH FUEL OF 5% ENRICIDiENT 6

Prepared for the l TU. ELECTRIC COMPANY

, by i

Stanley E. Turner, PhD, PE Holtec Project 20860 3

Holtec Report HI-92880 i

i 230 Normandy Circle 2060 Fairfax Ave. -

Palm Harbor, FL 34683 Cherry Hill, NJ 08003 I

9210200292 921016 PDR ADOCK 05000445

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4 REVIEW AND CERTITICATION LOG

DOCUMENT NAMEt CRTTTc1LITY N ET'1 EVAIUh2 ION OF COMMANcME St ENRTCmfRNT HI-92880 _ 9f W ,[g M

! HOLTEC DOCUMENT I.D. No. e a/q

....,2 0 8 6 0 p gf HOLTEC PROJECT NUMBER [4t'

, CUSTOMER / CLIENT Texas Utilitiam e

REVISION BLOCK

)l QUALITY PROJECT

, ISSUE AUTHOR REVIEWER ASSURANCE MANAGER-

} No. & DATE & DATE & DATE & DATE i

S.E. TURNER W.Vernetson M. Soler 5.E. TURNER

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NOTE: Signatures and printed names ara required in tha - review  !

block. l l l l This document conforms to the requirements of the design 1

specification and the applicable sections of the governing codes. I

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TABLE OF CONTENTS

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . 1 2.0

SUMMARY

OF COMANCHE PEAK FUEL RACK DESIGNS . . . . . . 2 4 2.1 New Fuel Storage Rack Design. . . . . . . . . . 2 2.2 Spent Fuel S'torage Rack Design . . . . . . . . . 2 2.3 Fuel Assembly Specifications . . . . . . . . . 2 3.0 CRITICALITY ANALYSES . . . . . . . . . . . . . . . . 3 3,1 Introduction . . . . . . . . . . . . . . . . 3 3.2 Design Criteria . . . . . . . . . . . . . . . . . 3 3.2.1 New Fuel Storage Vault . . . . . . . . . . 3 3.2.2 Spent Fuel Storage Pool . . . . . . . . . . 4 3.3 Analytical Methods and Benchmark Experiments . . 4 3.4 New Fuel Vault Criticality Analysia . . . . . . . 5 Calculational Model 5 2.4.1 . . . . . . . . . . .

3.4.2 New Fuel Rack Analysis Results . . . . . . . 5 For Normal Conditions , , . . . . . . .

3.4.3 New Fuel Vault Under Accident . . . . . ..6 Conditions 3.5 Spent Fuel Rack Criticality Analysis . . . . . .. 6 3.5.1 Calculational Model . . . . . . . . . . . . 6 3.5.2 Spent Fuel Rack Analysis for . . . . . . . . 6 Normal Conditions 3.5.3 Spent Fuel Rack Under Accident . . . . . . . . 6 Conditions 4.0

SUMMARY

AND CONCLUSIONS . . . . . . . . . . . . . . . . g 5.0 References ... . .. . . . . . . . . . . . . . . . . 9 i

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LIST OF TABLES Table 1 UNCERTAINTIES FROM MANUFACTURING TOLERANCES USING CASMO-3.AND KEN.' "- . . . . ... . .. . 10

SUMMARY

OF CRITICALITY SAFETY ANALYSSS Table 2 -

. NEW FUEL VAULT - 5% ENRICHED: FUEL .. . . . . . . 11-Table 3

SUMMARY

OF CRITICALITY SAFETY ANALYSES-

. SPENT FUEL POOL - 5% W OFA ENRICHED FUEL. . . .. . 12-2 l

I LIST OF FIGURES i

i I Fig. 1 NEW FUEL VAULT LAYOUT AND CALCULATIONAL MODEL . . . 13 Fig. 2 CROSS-SECTION OF SPENT FUEL STORAGE CELL . . . . . 14 Fig.'3 REACTIVITY VARIATION WITH MODERATOR DENSITY-. . . . 15-i f

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1.0 INTRODW TION The present study was undertaken for the purpose of documenting the capability of the fuel storage facilities at Comanene Pcak to safely store fuel of 5% initial enrichment. The fuel storage facilities include the New Fuel Storage -Vault (NFV) ,- tha Spent- Fuel Storage Pool (SFP), and the In-Containment Storage- Rack.

Criticality safety analyses and accident evaluations for each of the fuel storage facilities at Comanche Peak are presented in this report. These calculations confirm that all storage f acilities can safely receive and store fuel up to 5% enrichmen't'4 including a ~

manutacturing tolerance of 0.05%) within the limits of 'the USNRC guidelines. -

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9 2.0

SUMMARY

OF COMANCHE PEAK FUEL RACK DESIGNS -

I 2.1 New Fuel Storage Rack Design The storage rack layout in the NFV is illustrated in Figure 1.

The racks consist of stainless _ steel boxes - (0.0747 -inch _ thick)_

located on a 21-inch lattice spacing.- The storage -boxes are-

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arranged in seven rows of two cells each as illustrated-in Figure-

[ 1. Normally, fuel is stored in the dry condition. - Examination of as-built dimensions indicate that the average lattice spacing i is 21.01.i 0.04 inches (two_ sided tolerance for 95%1 probability at the 95% confidence level), and the spacing between pairslof-

['~ ~ ' rows shown in Figure 1 is 36.0 1 0.133 inches (95%/95%).__These

values were used in-the evaluation of the small uncertainties in-

- reactivity due to manufecturing tolerances.

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2.2 Spent Fuel-Storage Rack Design l_ Two separate storage po'ois._ exist, both of the same cell' design.

! The nin-containment rack is a ' single 5 x 5 module for' temporary-l storage of fuel assemblies.- In tha~ storage pools, spent fuel is i stored underwater in ~ type 304: stainles'B'ateel racks. The spent

! fuel storage racks consist of square stainfecs steel boxes, nominally 9.000 inch inside dimension and 0.074"/ inches thick

i. located on a 16-inch lattice spacing.. Figure 2 illustrates'a L cross-section of the SFP ce165 l

l 2.3 Fuel Assembly Specifications

! Four 17 x 17 fuel assembly designs were considered in the

! criticality safety evaluation of the Comanche Peak storage ~

facilities. These included both the Westinghouse Optimized j Assembly _(OFA) and standard. design and.the Siemens large and small fuel' rod designs. . Initial calculations established that

[. the-Wastinghouse 17 x 17 (OFA)' design exhibits the highest reactivity,'and this fuel design was used for the remainder _of i

l the -alculations.

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3.0 Criticality Analyses 3.1 Introduction .

The storage facilities were analyzed to assure that the most reactive fuel assembly with 5% enrichment ould be safely stored withir the limits established by-USNRC guidelines. Applicable codes, standards, and regulations include-the'following:

- General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling. ,

- USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, New Fuel Storage, Rev. 3 - July 1981.

- USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3 - July 1981.

USNRC letter 'f April 14 1978, to all Power Reactor Licensees - OT Position tar Review and Acceptance of Spent Fael Store.ge and Handling Applications, including _

modification letter dated January 18, 1979.

USNRC Regulatory Guitte 1.13, Speni Tuel Storage Facility Design Basis, Rev. 2_(proposed), December 1981.

ANSI ANS-8.17-1984, Criticality Safety' Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

3.2 Design Criteria 3.2.1 New Fuel Storage-Vault The new fuel storage vault is intended for the receipt and storage of fresh fuel under normally dry, low reactivity conditions. To assure criticality safety under accident conditions and to conform to tlur requirements of General Design Criterion 62, two criteria, as defined in NUREG-0800, Standard Review Plan 9.1.1, must be satisfied. These criteria are as follows:

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- When fully loaded with *uel of the highest anticipated reactivity and flooded with clean unborated water, the maximum k,,, including uncertainties, shall not exceed a kg , of 6.95.

- With fuel of the highest anticipated reactivity in place and assuming the optimum hypothetical' low density moderation, the maximum kg, including uncertainties shall not exceed a k g, of 0.98.

i 3.2.2 Spent Fuel' Storage Pool The principal NRC guidance (and identification of requirements) for spent fuel storage racks is included in the April 14, 1978, NRC letter which provides the definitive interpretation of SRP 9.1-2 and Reg Guide 1.13. The limiting k g , for Vater filled storage facilities is 0.95 including uncertainties evaluated for

95% probability at the 95% confidence level. . The= water in'the i spent fuel storage pool normally contains soluble boron which would result in large subcriticality margins under normal operating conditions._ However, the NRC guidelines specify that~

the limiting k of 0.95 for normal storage be evaluated for the accidentcondiN$onofthelossofsolubleboron'. The double j contingency principle of ANSI N-16.1-1975 and of the April 1978 NRC letter allows credit for soluble boron under other t') normal or accident conditions since only a single independent accident need be considered at one time.

3.3 Analytical Methods and Benchmark Experiments In the fuel rack analyses, the primary criticality analyses.were i made with the three-dimensional Monte-Carlo code package NITAWL-KENO-5aa 2) using the 27-group SCALE

• cross-section library (3) and

! the Nordheim integral treatment for U-238 resonance shielding i effects. In addition, tolerance effects were determined with CASMO-3 0ba) a two dimensional multi-group transport theory code.

Benchmark calculations, presented in Appendix A, indicate a bias of 0.0101 1 0.0020 (95%/95%)") for NITAWL-KENO-Sa. Since the SCALE cross-section library as used-by NITAWL has scattering

matrices only at 20*F and . 277*7 C, a special routine was

! developed to interpolate the scattering matrix for temperatures between those currently in NITAWL.

1 i * " SCALE" is an acronym'for Standerdized Computer Analysis for i Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC.

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N Monte Carlo (KENO-Sa) calculations inherently incl;de a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty, a minimum of 250,000 neutron histories in 500 generations efL500 neutrons each, were accumulated in each_ calculation.

CASMO-3 has been benchmarked against eritical experiments with water gaps up tc 2.576 inches (Append _x A). However, when large water gaps are present (as in the Comanche Peak racks), CASMO-3 will underpredict or overpredict kg , depending upon the size of the water-gap and the number of mesh intervals used. Incremental-changes in kg , due to the small manufacturing tolerances can be calculated by CASMO-3 to-obtain estimates of'the associated uncertainties. A comparison of small incremental values calculated by CASMO-3 and KENO-Sa for the flooded NFV is given'in Table 1. CASMO-3 and KENO-Sa show the same trend with temperature (Appendix A). These data confirm that the incremental reactivity effects calculated by CASMO-3 are reasonable estimates of the uncertainties. Consequently, CASMO-3 was used to evaluate small incremental reactivity effects from manufacturing-tolerances.

3.4 New Fuel Vault Criticality Analyses 3.4.1 Calculational Model The calculational model used for the analyses is indicated by the dashed lines in Figure 1, conservatively using an infinite heray of 2x10 storage boxes. Each fuel rod (including cladding) and guide tube within each of the stainless steel boxes and the_ boxes themselves are explicitly described. The model describes the concrete reflector in one direction (as indicated in Figure 1) but uses reflecting boundary conditions in the other. direction (long-direction of the 2x10 storage cell array). This-effectively creates-an infinite array in_the wide' direction and is conservative since four of the rows have only 9 cells rather than the 10 described in the model.

3. 4. 2. New' Fuel Vault Analysis Results For Normal Conditions Fuel in the New Fuel Vault is normally stored in the dry condition. An infinite number of dry assemblies of this design would have a k g, < 0.5 for enrichments up to 5%.

3.4.3 New-Fuel Vault Under Accident Conditions-In_the new fuel vault, the accident conditions' considered are-(1) the optimum moderation density and .(2) _ flooded with clean unborated water. Figure 3 shows the variation of the calculated k g of the r.ew fuel storage vault with ater density.

' (mo,deration) . The maximumgk , occurs at 5.3%.of normal water-density.. Table 2 summarizes the calculated k g and uncertainties for both the low-density optimum moderation and the fully-flooded conditions, including calculational and manufacturing uncertainties. For these postulated accident conditions, the 4 maximum calcult ! 3d values of k g , of the new fuel storage vault

, are 0.972 at the optimum low-density moderating condition and-0.929 for the fully flooded condition. These values of k g, are within the USNRC guidelines (SRP 9.1.1) and indicate that fuel up to 5% enrichment can be safely stored.

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] 3.5 Spent Fuel Rack Criticality Analysis k

3.5.1 Calculational Model 1

The calculational model used for analysis of the spent fuel storage racks is shown in Figure 2. In the geometric model used in the calculations, each fuel rod and its cladding were described explicitly in both the CASMO-3 and KENO-Sa models.

Reflecting boundary conditions (sero neutron current) were used at the centerline of the water-gap between cells which=has the effect of creating an infinite array of storage: cells. In the KENO-Sa model, the active fuel length was used in the axial j direction, assuming a thick (30 cm) water reflector, top and l bottom.

3.5.2 Spent Fuel Rack Analysis Results for Normal Conditions

, CASMO-3 calculations were made of the effect of temperature on the keff of the rack. These calculations showed that the highest occurs at 40*C and, therefore, this temperature was used for k,1 a1 subsequent calculations.

Under normal storage conditions, the presence of soluble. boron in

the pool water-assures a very low value of k g, (approximately 0.70 at 2000 ppm).

3.5.3 Spent Fuel Rack Under Accident Conditions The results of the postulated loss of all soluble boron accident condition calculations, summarized in Table 3, indicate a maximum kg, of 0.946, including calculational and. manufacturing

. uncertainties (95% probability at the 95% confidence level).

, This is less than the regulatory limit of a k g, of 0.95.

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Under other accident conditions in the spent fuel pool, credit for t the soluble poison in tho water is acceptable (double contingency principle of the April 1978 USliRC lotter), and will assure that thn i required subcritical reactivity margin is maintained for normal conditions and all credible accidents. Based on KE110-5a calculations with and without boron, the prosence of 2000 ppm boron in the pool water reduced the k,,, by 0.251 6k.

KElio-Sa calculations of the accidental mis-loading of a fuel assembly of 5% enrichmont outside and adjacent to th's : storage rark were made in *he presence of 2000 ppm roluble boron. A r.ie-loaded fuel assembly could theoretically bo situated-in an insido corner adjacent to two-rack modules. Assuming all of the fuel assemblies were of the highest permissible reactivity, the calculated k,,, is 0.708, which is well below the liRC limit.

The soluble boron in the pool water vauld more 'than adequately compensate for a frel assembly of 5% enrAwhment dropped and assumed to came to rest on top of a filled rack modulo. The consequenccc of this accident have been assessed assuming 2000 ppm soluble boron witl* pool water. In its final assumed position lying on-top of the rack, the calculated k,,, is 0.697, which is well within the 11RC limits, 5

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I l 4.0 Summary and Conclusions i

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4 j Critj:ality safety analysos of the Comancho Peak fuel storago  :

i facilities resulted in krr values less than the regulatory limits e

i for all conditions considorod with fuel types. expected to be loaded in CPSES of up to 5.0% enrichment. Those facilities include both the now fuel vault, the in-containment storago rack, and the spent  !

fuel pool. 'rablos 2 and 3 summarizo the calculated k rr of the e facilition, confirming that the Comancho Peak fuel storage- ,

! f acilition can safely _ accommodato fuel of 5% onrichment or any fuel j of lower onrichmont or reactivity.

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5.0 REFERENCES

1. R.M. Westfall, et. al., "HITAWL-S Scale System Module for Performing Resonance Shielding and Working Library Production"

in SCALE: A Modular Code System for oerformina Standardized

ggpouter Analyses for Licensino Evaluation <, NUREG/CR-02OO,  ;

1979. I 4

2. L.M. Petrie and N.F. Landers," KENO Va. An Inproved Monte Carlo Criticality Program with Supergrouping" in Ecale! A Modular, 7

Code System for eerformina Standardized Comeuter Analyses for Licensina Evaluation, NUREG/CR-0200, 1979.

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3. R.M. Westfall et al., " SCALE! A Modular Code System for performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-02OO, 1979.

l 4. A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary) .

5. A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ME Transactigjlg, Vol. 26, 1 p. 604, 1977.

6. M. Edenius et al. , "CASMO Benchmark Report," Studovik/ RF 6293, Aktiebolaget Atomenergi, March 1978.

7. "CASMO-3 A Fuel Assembly Durnup Program, Users Manual",

Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986

8. M. Edenius and A. Ahlin, "CASMO-3 : New Features , Benchmarking,

and Advanced Applications", Nuclear Science end Enaineerina, 100, 342-351, (1988) j

9. M.G. Natrella, Experimental Statistics National Bureau of Standards, Handbook 91, August 1963.

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, Ul1 CERTAINTIES FROM MANUFACTURIl1G TOLERANCES USING CASMO-3 AND KENO-Sa i

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U02 dens;ty 0.0031 .0023 T

Lattico Spacing O.0000 .0004-

Enrichment 0.0015 .0019 s

! Box I.D. 0.0003 .0017 4

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SUMMARY

OF CRITICALITY SAFETY ANALYSES NEW FUEL VAULT - 5% ENRICllED FUEL (Under Accident conditions)

Cptimum(3) Flooded l4aderation Condition )

i Temperature for analysis 20ec (68eF) 20*C (68eF)

Reference k, (KENO-Sa) 0.9513 0.9143 Calculational bias, 6k 0.0101 0.0101 Uncertainties In the Bias") 0.0020 1 0.0020 KENO Statisticsd> i 0.0047 + 0 0022 Lattice spacing i_0.0006 Negligible Box I.D. i 0.0005 -1 0.0003 Spacing between rows 1 0.0010 NA SS thickness 0.0094 1 0.0016 Fuel enrichment 1 0.0011 1 0.0015 Fuel density 1 0.0017 1 0.0031 Eccentric fuel Negligible- Negligible Statistical combination 1 0.0110 1 0.0048 of uncertainties (2)

Total 0.9614 1 0.0110 _ 0.9244 1 0.0048 Maximum Reactivity (k,g) 0.972 0.929 (1). With two-sided factor for 95%/95% tolerance.

- (2) Square root of sum of squares.

. (3) 5.3% of normal water density.

W -e- wgr e - k eit T wm' 'wattWTNM' W WN--W WT N WO 7+#3 W e w+ *up-HI-=Ptrv%+h7wv**

Tablo 3 i

SUMMARY

OF CRITICALITY SAFETY ANALYSES i SPENT FUEL POOL - 5% W OFA ENRICllED FUEL (Without Soluble Boron) i SFP 1

Temperture for analysis 40'C (104'F)

Reference k, (KENO) 0.9310 Calculational bias, ok O.0101 l

Uncertainties U) i i In the Bias 03 0.0020 l KENO Statistics 1 0.0022

., Lattice spacing i 0.0003

! Box I. D. 1 0.0002 i SS thickness 1 0.0024 i Fuel enrichment 1 0.0019

! Fuel density 1 0.0021

Eccentric Fuel Nealicible 1

Statistical combination i 0.0048 of uncertainties Total 0.9411 1 0.0048 Maximum Reactivity (k,,,) 0.946 i

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i APPENDIX A BENCIIMARK CAlfUI.ATIONS s

by i

Stanley E. Turner, PhD, PE HOLTEC INTERNATIONAL i

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August, 1992 I

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l.0 INTRODUCTION AND

SUMMARY

The objective of this benchmarking study is to verify both the l NITAWL-KEN 05a (1,2) methodology with the 27 group SCALE cross-section library and the CASMO-3 code (3) for use in criticality safety calculations of spent fuel racks. Both calculational methods are i based on transport theory and have been benchmarked against critical  !

j experiments that simulate typical spent fuel storage rack designs as

! realistically as possible. Results of_these benchmark calculations

! with both methodologies are consistent with corresponding calculations i reported in the literature.

Benchmark calculations woro performed for. critical experiments that are representative of realistic fuel storage racks and poison worths.

Results of these calculations show that the 27 group (SCALE)

NITAWL-KENO-Sa calculations consistently underpredict the critical eigenvalue by 0.0101 +/- 0.0020 (with a two-sided tolerance f actor for 95% probability at a 95% confidence level).

Extensivu benchmarking calculations of critical experiments with ,

CASMO-3 have been reported (5), giving a bias in k*' of 0.0004 +/-

d 0.0011 for 37 cases. The 95%/95% bias for the CASMO-3' data -is 0.0000

. +/- 0.0027 (neglecting the small overprediction in k). CASMO-3 and NITAWL-KENO-Sa intercomparison calculations of infinite arrays of poisoned cell cen!igurations (representative of typical fuel storage rack designs) confirm that the reported bias is reasonable for use in CASMO-3 calculations. Additionally, Reference 5 documents good agreement between CASMO-3 calculations and measurements of heavy nuclide concentrations for Yankee core isotopics.

4 The benchmark calculations reported here confirm that either the 27 group (SCALE) NITAWL-KENO-5a or CASMO-3 calculations are acceptable for criticality analysis of spent fuel storage racks.

i For configurations with large water gaps (> 2 or 3_ inches), results of CASMO-3 calculations differ from corresponding KENO-Sa results.

The difference observed depends on the size of the water gap and the number of mesh intervals used=in the CASMO-3 calculation. However, as discussed in Section 3, CASMO-3 calculations will provide reasonable estimates of uncertainties for small perturbations of input parameters (e.g., manufacturing tolerances),

j 2.0NITAWL-KENO-Sa BENCHMARK CALCULATIONS Table 1 summarizes results of analysis of 4 series of Babcock & Wilcox (B&W) critical experiments (4), including some with absorber panels typical of a poisoned spent fuel rack. These critical experiments were calculated with KENO-Sa, the 27 group SCALE cross-section library, and the Nordheim resonance integral- treatment in NITAWL.

Dancof f factors used in NITAWL were cal'::ulated with Oak Ridge SUPERDAN routine (from the SCALE system of codes). The mean k,,, for thers calculations is 0.9899.

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4 Similar calculational bias has been reported by ORNL (7) for 54 l

, critical experiments. The bias for these mostly clean criticala j without strong absorbers is 0.0100 0.0015 (95%/95%). These

! published results are in good agreement with the results cbtained in the present analysis and lend further credence to the validity of the

27 group NITAWL-KENO-5a calculational model for use in spent fuel rack criticality analysis. Further, results reported in the present

evaluation show no abnormal deviations in k g, with intra-assembly i water gap, absorber panel worth, enrichment, or poison concentration

) and do not show the trends previously presented using the 123-group GAM-THERMOS cross-section library (8) .  ;

l

. Since the BtW critical experiments were mado with fuel of 2.459 %

U-235 enrichment, additional benchmarking calculations were performed-

! to confirm that enrichment is not_ a significant factor in the XENO-5a

, bias. The additional calculations were for a series of French 4

critical experiments (9) at 4.75% enrichment and for several BNWL i

criticals-(11) with 4.26 % enrichment.

l The results of the French criticals, presented in Table 2, show an

' overprediction of kg. Further, the calculated k show a trend i toward higher values ,with decreasing core size. ObL has reported

! similar results (10). These critical experiments are for very small cores, and the overprediction suggests that NITAWL-KENO-Sa has an inadequate treatment of the very large leakage from very small cores.

4 Since the analysis of fuel storage racks does not entail large neutron leakage, the observed inadeque:y will not affect fuel storage rack analysis.

L The results shown in Table 2 for the French criticals and BNWL

experiments (also small cores, but significantly larger than the i French criticals) suggest that enrichment has no effect on the KENO-Sa bias. Or, in the case of the French criticals, any small enrichment offect would yield a more conservative value of k g,.

(

! Subsequent CASMO-3 calculations (discussed in Section 5) indicate that enrichment has no significant ef fect on modeling critical experiments with CASMO-3.

~

3.0 INTERPOLATION ROUTINE A special routine was developed to interpolate the hydrogen scattering

, matrix for temperatures between 20 'C and 277 'C in NITAWL. This

! special routine corrects a deficiency noted in NRC Information Notice 91-66 (October -18, 1991. ) Benchmark calculations were made using

CASMO-3 for comparison based on the assumption that two independent methods of analysis would not exhibit the same error.

Results of thc. benchmark calculations shown in Table 3 confirm that l the trend with temperature obtained by both codes-is comparable over

'the range investigated. This agreement establishes the validity of ,

I ,

A-2 l

4 1

the interpolation routine used in conjunction with HITAWL-KENO-Sa to calculate k,,, for temperatures between 20 'c and 277 'C.

l l The deficiency in the hydrogen scattering matrix does not appear i except in the presence of large water gaps where the scattering matrix I is important. However, the value of k , from CASMO-3, in the presence

of a large water gap, differs from t$e KENO-Sa value. Table 3 and Figure 1 show results for both codes for a water gap of 2.6 inches.

The absolutt values.of k differ, but the trend with temperature is i very similar for both codes. The agreement in the trend with j temperature lends further credibility to the interpolation routino, j and also. supports the use of CASMO-3 to provide reasonable estimates

! of changes in k due to minor perturbations in input values (e.g.,

2 manufacturing tolerances). j 4.0 CLOSE PACKED ARRAYS )

The Bt.W close-packed series of critical experiments (12) simulate i consolidated fuel. These experiments .were analyzed using NITAWL-KENO-Sa. Results of these analyses, shown in Table 4, suggest j a slightly higher bias than that for fuel with normal lattice spacing.

ORNL (13) has obtained similar results. Because very few cases are

available for analysis, a maximum bias for close packed latticos of 0.0155 including uncertainties should be_used. This value of bias would conservatively encompass all but one of the cases measured.

l 5.0 THE CASMO-3 BENCHMARK CALCULATIONS The CASMO-3 code is a multigroup transport theory code utilizing transmission probabilities to accomplish two-dimensional calculations -,

of reactivity and depletion for BWR and-PWR -fuel assemblies. As such, CASMO-3 is well suited for criticality analysis of fuel storage racks -

since ' general practice treats these racks- as an infinite medium of i storage cells, neglecting leakage effects.

CASMO-3 is a modification of CASMO-2E code and has been. extensively ,

benchmarked against both mixed oxide and hot and cold critical experiments by Studsvik Energiteknik, Reported analyses (5) of 37 critical experiments indicate a mean k, , , of- 1.0004 +/- 0.0027 (95%/95%). To independently confirm the validity'of'CASMO-3 and to investigate any effects of enrichment, a series of calculations was made with CASMO-3 and with NITAWL-KENO-5a for identical poisoned

, storage cells representative of a typical spent fuel storage rack.

l Results of these intercomparison calculations

• shown 'in Table 5 are - -

within the normal statistical variation of KENO calculations. fSince

.two independent methods of analysis are not expected to have the same

error, the agreement between CASMO-3 and KENO-Sa indicate that fuel enrichment does not have a significant effect over the range of enrichment of power reactor fuel-(2.5% to 5%).

A-3

Results of these intercomparison calculations

• shown in Table 5 are within the normal statistical variation of KENO calculations.

Since two independent methods of analysis are not expected to have the same error, the agreement between CASMO-3 and KENO-Sa j udicate that fuel enrichment does not have a significant effect over the range of enrichment of power reactor fuel (2.5% to 5%).

A second series of CASMO-3-KENO-5a intercomparision calculations j consist of analysis of the contral cell only of 5 B&W experiments. 1 The calculated results, shown in Table 5, indicato a mean delta k difference that lies within the 95% confidence limit of the KENO-Sa i calculations.

values The combined between CASMO-3dataand in Table 5 show KENO-Sa a' mean difference in k ,95%/95%

of- 0.0003+/-0.0012(two-sideN confidence, CASMO yielding the higher value). Combined with the uncertainty in the KENO-Sa bias, the CASMO-3 bias is 0.0000 +/-

0.0023.

. Intercomparison between analytical methods is - a technique endorsed by Reg. Guide 5.14, " Validation of Calculational Methods for Nuclear Criticality Safety."

A-4

-)

4 I l 6.0 REFERENCE!? T^ A??ENDIX A i

l. Green. Lucious, Petrie, Ford, White, and Wright, "PSR /NITAWL-1 (code package) ;,'ITAWL Modular Code System For i Generating coupled Multigroup Neutron-Gamma Libraries from ENDF/B", ORNL-TM-3706, Oak Ridge National Laboratory, November 1975.

l R.M. Westf all et. al. , " SCALE: A Modular System for Perf orming  !

2.

Standardized Computer Analysis for Licensing Evaluation",

NUREG/CR-02OO, 1979.

3. A. Ahlin, M. Edenius, and H. Haggolom, "CASMO - -A Fuel Assemoly Burnup Program", AE-RF-76-4158, Studsvik report.

! A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analvsis", ANS Transactiens, Vol. 26,

p. 604, 1977.

4 "CASMO-3 A Fuel Assembly Burnup Program, Users Manual",

Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986

4. M.N. Baldwin et al., " Critical Experiments Supporting Close

- Proximity Water Storage of Power Reactor Fuel", B&W-1484-7, The Babccck & Wilcox Co., July 1979.

i

, 5. M. Edenius and A. Ahlin, "CASMO-3 : New Features , Benen=arking, and Advanced Applications", Nuclear Science and Encinee" inn, 100, 342-351, (1988)

. 6. M.G. Natrella, Exterimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

4

7. R.W. Westfall and J. H. Knight, " SCALE -System Cross-section Validation with Shipping-cask Critical Experiments", MIE j Transacticns, Vol. 33, p. 368, November 1979

8. S.E. Turner and M.K. Gurley, " Evaluation of NITAWL-KENO

Bench = ark Calculations for High ' Density Spent Fuel Storage Racks", Nuclear Science and Enaineerina, 80(2)

230-237, February 1982, i

A-5 P

+

y _ _ . , , . _ . . , . . _ , . _ , . , . , _ . . _ , , . . . . . , ,

i

9. J.C. Manaranchs, et, al. , " Dissolution and Storage Experiment

)

with 4.755 U-235 Enriched UOg Rods", Nuclear Technoicav, Vol. -

50, pp 146, Septancer 1980

10. A.M. Hathout, et. al., " Validation of Three Cross-section

', Libraries Used with the SCALE System for Criticality Analy-sis", Oak Ridge National Laboratory, !MREG/CR-1917, 1981.

11. S.R. Bierman, et. al., " Critical Separation between Sub-critical Clusters of 4.29 Wt. % MU Enriched UO Rods in Water with Fixed Neutron Poisons", Battelle Pacific Northwest

, Laboratories, NUREG/CR/0073, May 1978 (with August 1979 errata).

12. G.S. Hoovler, et al., " Critical Experi=ents Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins", B&W-1645-4, Babcock & Wilcox Company (1981).

13. R.M. Westf all, et al. , " Assessment of Criticality Computation-al Software Ior the U.S. Department of Energy Office of Civilian Radioactive Waste Management Applications", Section 6, Fuel Consolidation Applications , ORNL/CSD/TM-247 (undated) .

A-6

I t

Table 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KENO-Sa CALCULATIONS l OF B&W CRITICAL EXPERIMENTS Experiment Calculated a

. Number k,,,

I 0.9922 1 0.0006 i

II 0.9917 t 0.0005 III 0.9931 t 0.0005 j IX 0.9915 t 0.0006 4

l X 0.9903 i 0.0006 XI 0.9919 0.0005 l XII 0.9915 t 0.0006 l XIII 0.9945 1 0.0006 i

l XIV 0.9902 1 0.0006 XV 0.9836 t 0.0006 XVI 0-9863 1 0.0006 XVII 0.9875 1 0.0006 XVIII 0.9880 1 0.0006 XIX 0.9882 1 0.0005 l XX 0.9885 i 0.0006 XXI 0.9890 1 0.0006 Mean 0.9899 : 0.000703- t 0.0006(2)

Blas 0.0101 : 0.00200)

U) Standard Deviation of the Mean, calculated from the k,,, values.

(2) f[(o)/16 0)

With two-sided factor (K=2.903) for 95%/95% tolerance.

A-7

t t

, Table 2

RESULTS OF 27-GROUP (SCALE) NITAWL-KENO-Sa CALCUIATIONS

- OF FRENCH and B!TWL CRITICAL EXPERIMENTS French Experiments separation Critical calculated

, Distance, em Height, cm k, g

i 0 23.8 1.0302 1 0.0008 i 2.5 24.48 1.0278 i 0.0007

5.0 31.47 1.0168 6 0.0007 j 10.0 64.34 0.9998 1 0.0007 BNWL Experiments Calculated l

Case Expt. No. k, g

No Absorber 004/032 0.9942 1 0.0007 SS Plates (1.05 B) 009 0.9946 t 0.0007 SS Platos (1.62 B) 011 0.9979 i 0.0007 SS Plates (1.62 B) 012 0.9968 1 0.0007 SS Plates 013 0.9956 1 0.0007 SS Plates 014 0.9967 2 0.0007 Zr Plates 030 0.9955 1 0.-0007 Mean 0.9959 0.0013 A-8

i 1

i 1

) Tablo 3

$ Intercomparison of NITAWL-KENO-Sa (with Interpolation Routina)

' snd CASMO-3 Calculations at Various Temperatures Temeeratu m CASMO 3 W-N-KENO-BaW ,

4'c 1.2276 -  !

11.5'C 1.2322 1.2328 2 0.0015 e

25'c 1.2347 1.2360 1 0.0013

) 50'C 1.2432 1.2475 0.0014 l 75'c 1.2519 1.2569 1 0.0015 I 120*C 1.2701 1.2746 t 0.0014 5

i.

• Corrected for bias 9

l 3

a r

5 A-9 i

W

_ , , . . , - - - - . , . - . . _ _ _ _ . -.2.. _ . _ , , , . _ . . . . - ,.,,_....~,__,.__,.2.-,

Table 4 Reactivity Calculations for close-Packed critical Experiments calc. B&W Pin Module Boron Calculated 11 0 . Expt. Pitch Spacing conc. k, , ,

No. cm em ppm KS01 2500 Square 1.792 1156 0.9891 1 0.0005 1.4097 KS02 2505 Square 1.792 1068 0.9910 t 0.0005 1.4097 KS1 2485 Square 1.778 886 0.9845 1 0.0005 Touching KS2 2491 Square 1.778 746 0.9849 1 0.0005 Touching KT1 2452 Triang. 1.86 435 0.9845 0.0006 Touching i KT1A 2457 Triang. 1.86 335 0.9865 1 0.0006 Touching KT2 2464 Triang. 2.62 361 0.9827 1 0.0006-Touching KT3 2472 Triang. 3.39 121 1.0034 1 0.0006 Toucning A-10 l

i.

1

Table 5 RESULTS OF CASMO-3 AND NITAWL-KENO-Sa d

BENCIDiARK (INTERCOMPARISON) CALCULATIONS

Enrichmentd) k, 8k Wt. t U-235 NITAWL-KENO-5all) CASMO-3 CASMO-KENO 2.5 0.8376 1 0.0010 0.8386 +0.0010 3.0 0.8773 1 0.0010 0.8783 +0.0010 3.5 0.9106 1 0.0010 0.9097 -0.0009 4.0 0.9367 1 0.0011 0.9352 -0.0015 s

i 4.5 0.9563 1 0.0011 0.9565 +0.0002 i

5.0 0.9744 1 0.0011 0.9746 +0.0002 B&W Expt. No . (38 XIII 1.1021 1 0.0009 1.1008 -0.0013 XIV 1.0997 t 0.0008 1.1011 +0.0014 XV 1.1086 1 0.0008_ 1.1087 +0.0001 XVII 1.1158 t 0.0007 1.1168 +0.0010 XIX 1.1215 t 0.0007 1.1237 +0.0022 Hean 0.0003 : 0.0012 (2-sidad 95%/95%)

0)

Infinite array of assemblies typical of hig5-dens / ty spent fuel storage racks.

(33 k, f rom NITAWL-KENO-Sa corrected f or bias.

(3) Central Cell from B&W Critical Experiments A-Il

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ENCLOSURE 2 TO TXX-92468 HUREG/CR-5009 Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors

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NUREG/CR-5009 PNL-6258 RE,1S Assessment of the Use 4 of Extendec Burnup Fue.

in Light Water Power Reactors i

Manuscript Completed: January 1988 Date Published: February 1988 Prepared by D.A. Baker, W.J. Bailey, C.E. Beyer, F.C. Bold, J.J. Tawil Pacific Northwest Laboratory Richland, Washington 99352 Pre aared for Div.slon of Regulatory Applications Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington DC 20565 NRC FIN B2894

l M

{ ABSTRACT l This study has been conducted by Pacific Northwest Laboratory (a) for the U.S. Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water

power reactors. Some nuclear power' plant licensees have requested authoriza-tion to increase their current batch average burnup levels of 33 GWd/t uranium j to levels above 50 GWd/t. The environmental effects of extending fuel burnup i are discussed with respect to normal operations, to accident events, and to

)

the economic effects on the fuel cycle. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic-assessments.

) Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction in 1) total fuel requirements, 2) reactor downtime for fuel replacement,. 3) the number of fuel shipments to and from reactor sites, and 4) repository storage- requirements.-

i

- t 4

i i

i i

i 1

)

L i

i-1 (a) Pacific Northwest Laboratory is operated by Battelle Memorial Institute

~

for the U.S. Department of Energy.

111

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1 CONTENTS ABSTRACT .............................................................. 111 ACKNOWLEDGMENTS ....................................................... xi l 1

GLOSSARY .............................................................. xiii EXECUTIVE

SUMMARY

...................................................... xix

1.0 INTRODUCTION

..................................................... 1-1 4

1.1 DISCUSSION OF THE PETITION................................... 1-1 1.2 EXPLANATION OF EXTENDED BURNUP FUEL . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3 PROJECTED INDUSTRY USAGE OF EXTENDED BURNUP FUEL ............ 1-2 1.4 U.S. LICENSING EXPERIENCE OF REACTORS USING EXTENDED BURNUP FUEL ........................................ 1-6 1

1.5 FUEL OPERATING EXPERIENCE AT EXTENDED BURNUP ................ 1-7 2.0 PHYSICAL EFFECTS ................................................. 2-1 2.1 PHYSICAL EFFECTS ON FUEL AT EXTENDED BURNUP ................. 2-1 2.2 FISSION-PRODUCT RELEASE FROM FUEL AT EXTENDEL BURNUP ........ 2-6 1

1 3.0 ENVIRONMENTAL EFFECTS ............................................. 3-1 3.1 EFFECTS FROM NORMAL OPERATION ............................... 3-1 3.1.1 Mining and Milling ................................... 3-3 3.1.2 Convers ion and En richment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3

' 3.1.3 Fuel Fabrication ................................ 3-3 3.1.4 Refueling ....................................... .... .... 3-3 3.1.5 Reactor Operation .................................... 3-3 3.1.6 Transportation ....................................... 3-4 3.1.7 Waste Management ..................................... 3-7 3.1.8 Reprocessing ......................................... 3-8 3.2 EFFECTS FROM ACCIDENTS ...................................... 3-8 3.2.1 Fuel-Damage Accident ................................. 3-8 3.2.2 Fuel-Handling Accident ............................... 3-10 3.2.3 Transportation Accident .............................. 3-12 3.3

SUMMARY

OF ENVIRONMENTAL EFFECTS ............................ 3-13 4.0 ECONOMIC EFFECTS ................................................. 4-1 4.1 DATA ........................................................ 4-2 y

T 4.2 -FRONT-END EFFECTS ........................................... 4-4 s-4.2.1 Front-End Research ................................... 4-4 ,

s 4.2.2 Fuel Production ............ ......................... 4-5  !

4.2.3 Fuel Management ...................................... 4-11 4.2.4 Aggregate Front-End Effects .......................... 4-12 i

4.3 BACK-END EFFECTS ............................................ 4-13 i I'

4.3.1 Development and Evaluation ........................... 4-13 i

, 4.3.2 A t- re a c t o r S t o rag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13

4.3.3 Transportation ....................................... 4-14

4.3.4 Repository Storage ................................... 4-18 1 1

a 4.3.5 Back-End Total Costs ................................. 4-21 l 4.4 TOTAL FUEL.e.YCLE ............................................ 4-22 j

a 4.5 SENSITIVITY ANALYSIS ........................................ 4-24 REFERENCES ............................................................ 5-1 j APPENDIX A - FUEL ACTIVITY INVENTORY CALCULATIONAL PARAMETERS ......... A : .!

4 a

APPENDIX B - COST ADJUSTMENT METHODOLOGY .............................. B-1 l'

1 k

4 I

E y

1 l

4 t

4-s 4

vi 1

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FIGURES 1.1 Example of a BWR Fuel Assembly .................................... 1-8 1.2 Example of a PWR Fuel Assembly .................................... 1-9 2.1 Cross-Section of a Uranium Oxide Fuel Rod for a Commerci al Light Water Power Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 3.1 The Uranium Fuel Cycle ............................................ 3-2 5

1

(

em vii

+

i TABLES l

i S.1 Summary of Changes. in[ Radiological Impacts for Various Fuel- 1

Cycle Activities with the Implementation;of Extended Burnup Fuel ... xx 1 5.2 Sumns /y of Estimated Discounted _ Cost Savings from Extendec Fuel Burnup by Activity, 1985-2020 .................. xxii' i '

1.1 Proj ections of_ U.S. Nuclear Generating Capaci ty . . . . . . . . . . . . . . . . . . .

1-2 i i ,

1.2 Prciected Implementation Schedule for Extended Burnup Fuel Use .... 1-3 h 1.3 Alternate Forecasts of Batch Average Burnup Levels by Reactor Type and Date of Fuel Loading, 1980-2020 .................. 1-4.

, 1.4 Aggregate Average Burnup of Spent Fuel at Discharge l by Fuel Type and Year of Discharge from Reactor ................... 1-4 1.5 Forecasts of. Spent-Fuel Discharges and Burnup .

Levels by Implementation Schedule and Year, 1985-2020 ............. .1 1.6. Comparison of PWR Burnup Level Forecasts by Source and Year, 1980-2005 ..................................... 1 i 2.1 Fuel Assembly Performance Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2 Radionuclides Contributing toLEnvironmental Impacts from Normal Operational and- Accidental Releases' to Biosphere

Showing-the Change Factor at, Shutdown Resulting from an

Increase in Burnup Level'from 33 to 60.GWd/t- ..................... 2-8 3.1 Major Contributory Radionuclides for Burnups of 33 and 60 GWd/t Released into the Cooling System-During Normal Operation,

. and Thei r Associ ated Change - Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 l

3.2 Estimated Population. Doses to Workers and the General

, Public for Three Types of Spent-Fuel Carriers.

l for 33 and 60 GWd/ t Burnup Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 ... '3-6

, 3.3 Estimated Population Doses to Workers anu the General Public

~

for Solid Waste Shipments by Truck for-33 and 60 GWd/t_Burnup Fuel ....................................... '3-7.

3.4 -Summary of Changes in- Radiological Impacts for Various Waste--

Management Activities for Increasing Fuel Burnup from 33-to 60 GWd/t ..........................................................- 3-8 4

3.5 Summary _of Changes in Radiological Impacts from Reprocessing for-Increasl ag Fuel Burnup f rom 33 ; to 60 GWd/t . . . . . . . . . . . . . . . . . . . . . . . . 3-9 e

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I 3.6 Calculated Release Fractions in the Fuel-Cladding Gap of the Peak Fuel Rod During Nonnal Operation at Current and Extended Burnup Levels of 33 and 60 GWd/t Compared with Gap-Release Fractions Assumed in Regul atory Guides . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-12 3.7 Major Contributing Radionuclides Resulting from a Transportation Accident Involving the Release of Spent Fuel ..................... 3-13 3.8 Su unary of Changes in Radiological Impacts for Various Fuel-Cycle Activities with the Implementation of Extended Burnup Fuel ....... 3-14 4.1 Comparisons of Forecasted Spent-Fuel Discharges With and Without Implementation of Extended Burnup by Year, 1985-2020 .............. 4-2 4.2 Projected 00E Research and Developunt Expenditures for New Stended Burnup Projects by Year, 1987-1994 ....................... 4-5

• . lized Reference Reactor Year Uranium Ore Requirements r" and Without Reprocessing, by Burnup Level .................... 4-7 lized Reference Reactor Year Milling Requirements With lithout Reprocessing, by Burnup Level ......................... 4-8

.recasted Milling Costs by Year. 1985-2000 ....................... 4-8 4.6 Forecasted Conversion Costs by Year, 1985-2000 .................... 4-8 4.7 Annualized Reference Reactor Year Separative Work Unit Requitements With and Without Uranium Recycling, by Burnup Level....................................................... 4-9 4.8 Annualized Reference Reactor Year Fuel Requirements by Burnup Level .................................................. 4-11 49 Annualized Reference Reactor Year Transportation Requirements for Uni tradi ated Fuel , by Burnup Level . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-11 4.10 Back-End Transportation Costs, 1984-2020, by Repository Type a n d B u rn u p L e v e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 15 4.11 Estimated Back-End Transportation Costs, 1984-2020, by Spent-Fuel Aga a n d B u rn u p Le v e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 16 4.1" Annualized Reference Reactor Year Transportation Requi rements for Spent Fuel , by Burnup Level . . . . . . . . . . . . . . . . . . . . 4-18 4.13 Total Undiscounted Repository Costs by Spent-Fuel Ag e a nd Bu rnup Leve l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 19 4.14 Repository Costs, 1984-2020, by Burnup Level and Repository Type ......... .................................... 4-19 iX

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i 4.15 Total Discounted Repository Costs by Spent-Fuel Age and Burnup Level ....................................... 4-201

4.16 Estimated Discounted Repository Costs, 1984-2020 by Burnup Level and Repository Type ............,................ . 4-20 4-

4.17 Summary of. Estimated 1985 Discounted Back-End Cost Savings, 1985-2020, by. Data Source and Cost Ca tegory . . . . . . . . . . . . . . . . . . . . . 4-22

+

4.18 Estimated Savings Resulting from Implementation' o f - Ert nded Burnup by Category . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 .

4.19 Summary of Estimated Olscounted Cost Savings,- 1985-2020,

, f rom Extended Fuel Burnup, by Category . . . . . . . . . . . . . . . . . . . . . . . . . . 4-26 i

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i ACKNOWLEDGMENTS We wish to express our gratitude to Morton R. Fleishman, Office of Nuclear i discussion, and assistance in this. work. Regulatory Research, U.S. N technical peer review, V. ~L. Brouns for technical editing, and A. Jewell andW C. F. Schauls for-word processing.

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t GLOSSARY Absorbed dose - The amount of energy in the form of 'fonizing radiation absorbed in a unit mass of eatter. The unit is gray (Gy) = 1-joule / kilogram (kg) =

100 rad.

4 Actinide -- A radioactive nuclide having an ~ atomic number greater than that of radium, which is 88.

Activity - A_ measure of the rate at whicn. material is emitting nuclear.radia-tions; radioactivity is usually given in terms of the number of nuclear disintegrations occurring time. The unit of activity in becquerel is the a given quantity Bq) = 1(of matter over disintegration / a period of second = 27 picoeurie (pC1).

At-reactor storage.- Temporary storage of spent-fuel assemblies in pools of water at reactor sites.

Average individual - An individual of the general public whose habits are average for the general population.

Back end_- Pertaining to activities associated with the uranium fuel cycle af ter removal of spent fuel from a reactor, such as at-reactor storage, transportation to reprocessing plants or waste repositories, and final burial if not reprocessed. Also included is.deve17pment and evaluation related to these activities, i

Batch average burnup - The average (arithmetic mean) of the burnup of all fuel assemblies in the quartity of fuel to be replaced at a single fuel-replacement outage.

i Biotransfer factor - Ratio of- the concentration of activity in food crop to that of soil; concentration factor.

Boiling water reactor (BWR) - A light water reactor in which water, used as both coolant and moderator, is allowed to boil in the reactor core. The-resuiting steam can be used directly to drive a turbine. --

j Burnup - A measure of nuclear reactor fuel consumption; the; quantity of energy generated by' a unit amount of uranium fuel in a nuclear reactor, usually expressed as the average power. multiplied by the' days of; exposure in the core per unit mass of uranium: 1,000 megawatt-days / tonne (mwd /t) =

1 gigawatt-day / tonne (GWd/t) = 1 megawatt-day / kilogram (mwd /kg).

Capacity - The electrical power generating load for which a plant is rated.

Also the total capacity of a set of plants such-as the nuclear plants in the United States.

Capacity factor - The ratio of actual power generated to the amount that would have been generated, had the reactor operated at rated capacity for the same period, usually a year. i xiii I L. -

H m - . . . , m .vew.

Cladding - The outer metal jacket of nuclear fuel rods. It prevents the cor-rosion of the fuel and release of fission products into the coolant.

Class 9 accident - An accident that involves substantial physical deterioration of the fuel in the reactor core, including overheating to the point of melting, and involves deterioration of the containment system (filters, sprays, etc.) to perform its intended function of limiting the release of radioactive materials to the environment.

Conversion - The stxge of fuel production in which yellowcake (U3g0 )is converted to uranium hexafluoride (UF6 )*

Curie (Ci) - Old unit of activity defined g the r ant of a radioactive mate-rial that has an activity of 3.7 x 10 dis' agrations/second. Now being replaced by the becquerel. (See activity.)

Disaggregate - To separate into component parts.

Discharge burnup - Average burnup level of discharged fuel batch.

Distribution factor - The factor used in computing dose equivalent to allow for the nonuniform distribution of internally depo!,ited radionuclides.

' Dose - In thG report, " dose" is used to designate the dose equivalent in rem. (1 e dose equivalent.)

Dose commitmet) - The integrated dose that results from external exposure to, or an intve beginning M' of, radioactive material when dose is evaluated from the exposure or intake to a later time (usually 50 years).

Dose equivalent - The product of absorbed dose, quality factors, dose distri-bution factor, and other necessary modifying factors. The unit is the sievert (SV) = 100 rem.

Dose equivalent commitment - The infinite time integral of the per capita dose-equivalent rate in a given organ or tissue for a specified population.

Enrichment - The process by which the abundance of the isotope uranium-235 in natural uranium is increased above nomal (0.7%). For standard burnup levels, the enrichment level is around 3.5%. For extended burnup, the appropriate enrichment level is about 4% to 5%. The amour.t of enrichment processing is measured in SWUs (separative work units.)

Fabrication - Stage of fuel production in which the enriched fuel (UO 2

) is formed into pellets and placed into fuel-rod assemblies.

Fission products - The nuclei (fission fragments) formed by the fission of heavy elements, plus the nuclides formed by the radinactive dway of fission fragments.

xiv

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Front end - Pertaining to those activitie's of the uranium fuel cycle ending in reactor operation such as mining, processing, converting, and enriching the uranium, and the fabrication into fuel assemblies and final transpor-tation to the reactor site. Also ir luded is research and development related to these stages.

Fuel - Fissionable material used as the source of power when placed in a criti-cal arrangement in a nuclear reactor. The fuel is in the form of pellets contained in rods that are distributed in an array (typically square:

8 x 8, 17 x 17, etc.) contained within a fuel assembly.

Fuel cycle - The various activities or processes undergone by reactor fuel, from mining of the ore to storage of nuclear waste in a repository.

Fuel management - Decisions and operations relating to burnup of fuel in the .

reactor. These decisions concern such matters as the length of time fuel remains in reactor core, the frequency of fuel reloading,_ and the propor-tion of-total fuel replaced during reloading. .Another decision pertains to the placement of fuel assemblies of different age in the reactor core.

Fuel production - The five processing stages (mining, milling, conversion, enrichment, fabrication) necessary to produce reactor fuel.

Gap-release fraction - The fraction of the fission products that migrates from the fuel matrix to the region between fuel pellets and fuel cladding during normal operation and which would be available for immediate release into the cooling water in the event of clad damage.

Generating capacity - (See capacity.)

General population - Group or population not employed in occupations dealing with radioactive materials.

Half-life - The time _ required for the activity of a radionuclide to decay to I

half its valt.e; used as a measure of persistence of radioactive materials.

light water reactor (LWR) - A nuclear reactor that uses water as the primary coolant and moderator, with slightly enriched uranium as fuel. (See boiling water reactor and pressurized water reactor.)

Man-rem - Unit used to compare the effects of different amounts of radiation on groups of people. It is.obtained by multiplying the average dose equivalent to a given organ or tissue (rem) by the number of persons making up the population group. Sometimes given as person-rem.

Maximum individual (maximally exposed individual) - An individual of the general public whose locations and habits tend.to maximize his radiation dose, resulting in a dose higher than that received by other individuals in the general population.

Milling - Stage of fuel production in which uranium ore is purified into a more concentrated form called yellowcake.

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Nuclide - A species of atom having a specific mass, atomic number, and nuclear energy state.

Operating capacity - (See capacity factor.)

Peak rod average burnup - The rod average burnup level of the peak rod in a fuel batch or reactor core.

Pellet-cladding interaction (PCI) - Interaction between the fuel pellet and cladding that results in cladding stresses and chemical reactions th?t may lead to cladding rupture.

Population dose - The dose equivalent received by the genera: po radioactive materials. The unit is man-rem, or person-rem.pulation from Precursor - A nuclide that precedes that nuclide in a decay chain parent.

Pressurized water reactor (PWR) - A LWR in which heat is transferred from the core to a heat exchanger via water kept under high pressure, so that high temperatures can be maintained in the primary system without boiling the water. Steam to power the turbines is generated in the secondary circuit.

Quality factor (Q) - The factor by which the absorbed dose (rad or gray) is multiplied to obtain a quantity that expresses the effectiveness of the absorbed dose in causing biological damage on a common- scale for all ionizing radiation. In practice, Q is taken as unity for x rays, gamma rays, and beta particles. (See dose equivalent.)

Radiation (f onizing) - Particles and electromagnetic energy emitted by nucleac transformations that are capable of producing ions when interacting with matter; gamma rays and alpha and beta particles are examples.

Radioactivity - (See activity.)

Radionuclide - Any nuclide that is radioactive.

Radwaste - Waste that contains radioactive materials.

Reference reactor year (RRY) - One year of operation of a model 1000-megawatt-electrical (MWe) reactor that is assumed to be operating at its maximum capacity for one year with an on-stream plant factor of 80%.

Release fraction - (See gap-release fraction.)

Reprocessing - Process in which some of the remaining unused uranium is ex-tracted from used fuel to be used in the production of new fuel. This reduces the need for uranium ore. Reprocessing represents the closing of the fuel cycle.

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Rem Unit of' dose equivalent. The ab' sorbed dose of 1onizing radiation modified by the quality factor and -sometimes other factors that result in the same biological effect as 1 rad of radiation:-1 rem = 1' rad _for x, gamma, or beta radiation. (See_ dose equivalent and quality-factor.)-

Rod average burnup - The burnup level of a fuel rod _ averaged (arithmetic mean) over its active fuel length. '

Separative work unit (SWU) - A measure of. work required _to separate a quantity of isotopic mixture into two component parts, one having a-higher percen-tage of_ concentration percentage. The_ SWUof is the desiredinisotope expressed in11ts ofand theone having(a kilogram kg). lower-Tonne (t) - Unit of' mass. Sometimes called1 metric ton (MT) = 1 megagram (Mg) = 1,000 kilogram (kg) = 2,205 pounds.

Transuranium nuclide (transuranic) - A nuclide'having an atomic number greater than that of uranium, which is 92; usually-characterized by having a long half-life and low or no_ gamma emissions.

Yellowcake - The final precipitation formed in tha mining of uranium ores (UO2 )-

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EXECUTIVE

SUMMARY

This assessment has been performed in partial response to a petition for rulemaking to the U.S. Nuclear Regulatory Commission (NRC) requesting that regulations be amended to require preparation of an environmental impact state-ment on the environmental effects of the use of extended burnup fuel in the nation's-light water commercial power reactors (LWRs), and it has been performed in view of the anticipated increase in applications for the use of this fuel.

The overall findings of this assessment are that no significant adverse effects will be generated by increasing the present batch-average burnup level of 33 GWd/t uranium to 50 GWd/t or above, as long as the maximum rod average burnup level of any rod is no greater than 60 GWd/t.

Extensive studies of extended burnup fuels have been conducted under the direction of the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), with the participation of the fuel vendors nationwide and with the cooperation of several nuclear reactor utilities. These studies have shown that there is no loss in fuel integrity for rod average burnups reaching 60 GWd/t, as long as power levels for the fuel rods remain normal.

Activity inventory may increase for long-lived radionuclides of concern; how-ever, for short-lived fission products, the inventories will essentially remain the same. Of the longer lived fission products of concern, only cesium-134, cesium-137, and strontium-90 increase much with burnup (by factors of 2.5, 1.9, and 1.8, respectively). Neutron emission rate from transuranium isotopes will increase with burnup by a factor of 5.6. Fuel cladding gap-release i

fractions for volatile fission products may increase by factors of two, but will remain below regulatory guide assumptions on noble gases and iodines at current power levels.

If leakage of radionuclides from a fuel element occurs during operation, the activity is expected to be removed by the plant cooling-water cleanup system. No change in the licensed technical specifications pertaining to allowed cooling-water activity concentrations would be necessary. Thus, with extended burnup, little.or no increase in the release of radionuclides to the environment is expected during normal operation. Other parts of the fuel

' cycle would also not be adversely affected by changing to an extended burnup fuel utilization plan. The impacts on workers and the general population would actually be reduced because at higher burnups, outages for fuel changes will be less frequent, and fuel shipments to and from the reactor sites will be reduced, thus reducing exposure. Although the inventory of long-lived radionuclides in the spent fuel will increase, the amount of spent fuel removed from reactors each year will decrease. Table S.1 summarizes the impacts of-i normal operation.

Accidents that involve the damage or melting of the fuel in the reactor core and spent-fuel handling accidents were also revicwed. For accidents in which the core remains intact, the release wwld involve only volati'e fission products, and no increases in impacts would occur, bsmse tha radiontclides contributing most to the dose are short lived and thus. do in f acrease with burnup. For larger (severe) accidents in which an appreciable amount or cll of the fuel has melted and been released from the containment system into the xix 4

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TABLE S.I. Summary of Changes in Radiological Impacts for Various Fuel-Cycle Activities with the Implementation of Extended Burnup Fuel Activity  % Change Primary Contributors Reference (a)

Mining -5 Rn & daughters Table 3-11 Milling -5 Ra-226, Th-230, U &

daughters Table 3-13 Convers on -5 U & daughters, Ra-226 Th-230 Table 3-14 Enrichment 1 0 & daughters Table 3-15 Fabrication -45 Th-234, 0 & daughters Table 3-16 Refueling -45 Normal reactor 0(D) Cs, tritium, I-131, Table 3.1 of operation Kr, Sr-90, Xe this report Transportation:

Fresh fuel -45 U Table 4-3 Spent fuel -45 Fission products Page 4-4 Solid wastes 20 Corrosion, activation, fission products Page 4-19 Waste management See Table 3.4 Reprocessing -44 to -3 Tritium, C-14, Kr-85, U, I-129, I-133, Ru-106, Fission Products, Transuranics Table 3-24 (a) Tables and pages referenced are from Mauro et al.1985.

(b) There would be no change in the licen. ed technical specifications for coolin3-water activity.

biosphere (Class 9) only a few fission products and the actinides will increase in inventory with increased burnup. The fission products would increase by no more than a factor of two, and the actinides by no more than a factor of six (of those contributing to the dose). However, since these actinide have very small release fractions and biotransfer factors, the risks associated with the actinides would be insignificant compared to those associated with fission products such as cesium-137 and strontium-90. Furthermore, the factors of increases in the radioactive sources are less than the uncertainty involved in detemining the overall risk to the public.

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For tne fuel-handling accident, only the noble gases and icdines escaping the damaged cladding are of significance in the assessment of dose impacts to the population. For a peak rod at a burnup level of 60 GWd/t, the release fractions only increase by a factor of three to four for these radionuclides; however, they remain below those assumed in Regulatory Guide 1.25, with the exception of iodine-131. Because.the calculated iodine gap-release fraction is 20% greater for scme high-power fuel designs than the Regulatory Guide 1.25 (NRC 1972) assumed value of 0.10, the calculated thyroid doses with extended burnup fuel resulting from a fuel-handling accident could be 20%

higher than estimated using the guide.

Spent-fuel transportation accidents were reviewed. Activity inventory may increase by an overall factor of about three for long-lived radionuclides of concern (assuming a 5-year cooling period) when changing to extended burnup fuel. However, this increase is offset by a decrease in the number of fuel-transport trips required, such that the overall change would be a 50% increase in impact by changing to 60 GWd/t burnup.

The use of extended burnup fuel would reduce fuel requirements per unit of electricity. This translates directly into reduced requirements for the various materials and operations linked to fuel production (uranium mining,

, milling, conversion, separation, and fuel fabrication). The result of these reduced production requirements will be a significant reduction in cost, as well as a reduction in enviror. mental impacts from fuel-cycle operations required to support one year of reactor operation.

Although the discharged fuel at extended burnup is thermally slightly hotter, has increased neutron emission, and has more long-lived nuclides per unit mass compared with fuel that has not undergone extended burnup, the volume cf fuel discharged per unit time will be reduced; thus, although the waste contains a greater actinide and long-lived fission-product activity, there will be less of it. This ambiguous nature of the waste has an effect on all the back-end stages of the fuel cycle (at-reactor storage, transportation, and repository storage). The net result of these changes would be an increase in transportation shielding requirements, a reduction in the number of fuel ship-

' ments, smaller repository waste packages or increased spacing in the under-ground repository, and a reduction in future at-reactor storage requirements.

3ources used in this study indicate that a reduced volume of waste will result in a net raduction of at-reactor storage costs and transportation costs.

However, estimates of the effects on repository storage costs range from a substantial' increase to a substantial decrease, depending on repository building assumptions. Because of uncertainties regarding repository specifications at this time,-these estimates only indicate the range of likely effects.

A summary of the cost savings resulting from the implementation of extended fuel burnup is presented in Table S.2. -The Table shows the cost savings for different stages in the fuel cycle. The wide range given for some cost activ- .

ities, most notably repository costs, reflects differences in estimates among l various references. The total discounted cost savings resulting from the '

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~~ Summary of: Estimated' Discounted Cost Savingsi from Extended Fuel Burnup by Activity. 1985-2020

~

Savings-Fuel-CycleStage/ActivitV (millions of 1985$)

3- Front end of fuel cycle -

Research-and development -21

Fuel production and fuel management 2,552-4 Subtotal 2,531

[

4 Back'end of fuel cycle- 1 0*velopment and. evaluation 'O i

At-reactor storage 48

-Transportation-

.4 to 72

Repository -666_to 94'

Subtotal -614_to 214

{ TOTAL FUEL-CYCLE SAVINGS 11,917 to 2,745 d

5 implementation of extended' burnup for all reactors-in current operation for-i the period 1985 to 2020 is estimated .to be on the order of $2 billion.. Alr^

• all of this-amount comes from savings in the front end of tne-fuel cycle.-

Small cost savings are noteo in at-reactor storage and transport:of spent fuel. Sources disagreed as:to the. extent of effects of extended burnup on repository costs.--One source estimated a cost increase in this aret of from

$390 million to $670 million-(Weston-1985),1whereas data from-another source suggested a relatively modest decrease-in. repository costs'of between $78 mil-lion and $94;million (Dippold and Wampler 1984). -Because there are a number.

of severe ~ limitations in the source data, these estimates-must-be viewed with.

[

i caution.

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1.0 INTRODUCTION

This study was conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Comission (NRC) in partial response to a petition regarding the use of extended burnup nuclear fuel in the nation's light water power reactors (LWRs) and in anticipation of an increase in applications for the use of this fuel. Topics addressed in this assessment include the physical effects of extended burnup on the fuels, the environmental effects during both normal operations and accident events, and the economic effects oa the fuel cycle. This report also includes a glossary of technical terms used in the document and an appendix that provides activity inventories for various radionuclides at selected burnup levels.

1.1 DISCUSSION OF THE PETITION A petition for rulemaking was submitted on March 6, 1980, by Ms.-Catherine Quigg on behalf of Pollution and Environmental Problems, Inc. (Quigg 1980).

The petitioner's primary concern was the potential for the proliferation of the use of extended burnup fuel by the nation's commercial nuclear power reac-tors. Another concern expressed by the petitioner was the U.S. Department of Energy's (DOE's) progran ith various fuel vendors to test higher burnup fuel assemblies in the cores of some selected commercial power reactors. These activities, according to the petitioner, "could cause significant and widespread long- and short-term effects on the human environment" (Quigg 1980). The DOE program has been discussed in detail in a study issued by the DOE (1980a):

Environmental Assessment, 00E Program to Imorove Uranium Utilization in Light Water Reactors.

The petitioner requested that the NRC amend its regulations to require preparation of a generic environmental impact statement (EIS) on the use of high burnup nuclear fuels in commercial LWRs (see Glossary). This assessment has been conducted in partial response to that petition to determine if a significant environmental impact would be caused by the widespread use of extended burnup fuel, thus necessitating the need for development of a generic EIS.

i 1.2 EXPLANATION OF EXTENDED BURNUP FUEL Burnup is a measure of the energy generated by fuel in a nuclear reactor.

Burnup, which at times is also called exposure, is measured in megawatt-days per metric ton (tonne) of uranium (mwd /MTU) or gigawatt-days per tonne of uranium (GWd/t). In the literature, the term " megawatt-days per kilogram of uranium" (mwd /kg) is also found; 1 GWd/t = 1 mwd /kg = 1,000 mwd /MTU. Currently, fuel burnups are around 31 GWd/t for boiling water reactors (BWRs) and 36 GWd/t fcr pressurized water reactors (PWRs), averaged (weighted mean) over the batch of fuel assemblies to be replaced in the cure during a refueling outage (batch average) (sce Glossary for definitions of terms). Under an experimental program sponsored by the 00E, fuel rods manufactured by the various vendors have been exposed to burnups as high as 45 to 60 GWd/t in a few participating reactors.

The ultimate goal of this study is to determine the fuel behavior at peak rod burnups of 65 GWd/t (00E 1980a, p. 2), which would correspond to a batch average of around 50 GWd/t.

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The differences between "high" burnup fuel and " current" burnup fuel vary among fuel vendors. In general, high burnup fuel consists of higher uranium-235 enrichment (see Glossary) than current fuel. This usually results in higher fuel-rod power peaking in the core. The higher peaking is currently controlled by varying the enrichment radially and axially within the core and by using additional burnable poison material. Hardware design changes for extended burnup also vary from vendor to vendor, but some typical changes are larger fuel-rod plenums, smaller fuel-rod diameters, fuel-to-cladding (see Glossary) gap changes, and changes in tie-plate clearances from fuel rod to upper assembly.

1.3 PROJECTED INDUSTRY USAGE OF EXTENDED BURNUP FUEL The nationwide aggregate environmental and economic effests of using.

extended burnup nuclear fuel depend on the amount of electricity generated by nuclear power and the proportion of that amount that is generated by extended burnup fuel. Projections of U.S. nuclear generating capacity (see Glossary),

developed by the U.S. Department of Energy / Energy Information Administration (DOE /EIA), are given in Table 1.1. Note that tne DOE /EIA has released several sets of projections. The ones shown here are the latest available and, there-fore, will differ somewhat from those published earlier and from projections used by other sources referenced in this work. Table 1.1 presents four sets of projections reflecting alternative assumptions about future plant construc-tion and other factors. This report adopts the middle case as best reflecting the most probable future generating capacity, since it represents capacity growth near the middle range of the DOE /EIA projections.

Three prospective extended barnup implementation schedules are presented here. The first, taken from Murphie and Lang (1982), is shown in Table 1.2.

This schedule forecasts the rate of implementation in terms of the percentage of fuel supplied. According to this source, 55% of all U.S. vendor-supplied TA8LE 1.1. Projections of U.S. Nuclear Generating Capacity (net GWe at year end)

With Including New Orders No New Low Middle High Year Orders Case rase Case 1984 71 71 71 71 1985 71 77 80 85 1990 105 105 110 113  ;

1995 108 108 117 117 l 2000 106 106 116 119 l 2005 106 119 149 174 2010 104 129 182 237

=

2015 59 137 216 306 2020 46 144 248 377 i Source: DOE /EIA 1985, p. 24.

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TABLE 1.2. Projected Implementation Schedule for Extended Burnup Fuel Use

% of Total Year Feel Suoplied 1981 0 1985 11 1990 55 1995 80 2000 100 Source: Murphie anc Lang 1982.

fuel will be used for extended burnup fuel operation by 1990, with 100% imple-mentation by the year 2000. This usage is reached by a series of small (3,000-5,000 mwd /mt) [3 to 5 GWd/t] incremental increases in burnup being introduced at several-year intervals, with nearly full deployment of each increment following 5 to 6 years

' after commercial introduction. The schedule is also based on a frirly aggressive research, development and demonstration program; any reduction in support would delay the impleraentation schedule.

Commercial introduction and widespread availability at each increment in burnup are expected to closely follow technical demonstration

[Murphie and Lang 1982, p. 7-63] .

The average burnup of this fuel was not specified,-buc it can be expected to vary depending on the circumstances of the various utill+1es. However, Lang (1986) has recently stated that the DOE goals for burnup extension are 50 GWd/t (batch average) for PWRs and 45 GWd/t (batch average) for BWRs.

A second forecasted implementation schedule, developed by Franks and Geller (1986), is presented in Table 1.3. Shown in the table are alternate time patterns based oa different assumed scenarios: 1) no change in current burnup levels, 2) no-additional DOE-funded research on extended burnup,

3) likely implementation schedule consequent to additional DOE-funded research, and 4) maximum implementation rate schedule assuming DOE-funded research.

A third and more detailed set of forecasts of extended burnup implemen-tation and the amount of spent fuel generated is found in Weston (1985).

Table 1.4 presents expected aggregate average burnup levels for PWRs and BWRs, by year, for an implementation schedule in which utilities are assumed to introduce fuel designed for extended burnup as quickly as it becomes avail-able. The discharge burnup levels (see Glossary) are assumed to reach 60 GWd/t in 2005 for PWR fuel and 45 GWd/t in 2000 for BWR fuel (Veston 1985, p. 2-1).

Table 1.5 shows two sets of forecasts for annual spent-fuel discharges.

Both . sets are consistent with the DOE /EIA middle-case projections of U.S.

nuclear generating capacity. The first set shows the spent-fuel discharges 1-3

-. ..- _ . - - - . _ ~ - -- -- -. .

l TABLE 1.36 Alternate Forecasts of Batch Average Burnup Levels (GUd/t) by Reactor Type and Date of Fuel Loading, 1980-2020-

-Current Without New W1th New . Maximum Burnup Levels 'R&D R&D- Implementation' Date BWR PWR- BWR PWR BWR PWR .BWR PWR-Before 1/1/1980- 28.5 33 28.5 33 28.5 -33 28.5 33 Before .

1/1/1985- 28.5. 33 31. 39- 31 39 32-33 39.

! ,Before

1/1/1990 28.5 33 35 40 38 45 38-40 45 Before

1/1/1995 78.5 33 38 45 40 .50 40-42 50' 1

After_

1/1/2000 28.5 33 -38 50= 43. 55 42-45 'S5 i

Source: Franks and Geller 1986, Tables 3-3 and 4-2.

t TABLE 1.4. Aggregate Average Burnup 'of Spent; Fuel at Discharge by Fuel Type and Year of:

Discharge from Reactor

' Burnup (GWd/t)

Discharge Year PWR Fuel BWR Fuel i 1984 38 29 1990 48' 37 1995 52 41-

.2000 56 45

Beyond 1999 - NA -45 Beyond 2004- 6u 45-i Source: Weston 1985, p. 2-3..

NA = Not Available Note: ' Estimates here -are for source's- " peak -

i burnup' scenario" in which burnup levels i of 60 GWd/t for PWRs and 45 GWd/t for BWRs are' achieved as,quickly as;possible, i-1-4 i

i ,

~ . , - , , , . . ~ _ . _ . . . ,.

TABLE 1.5. Forecasts of Spent-Fuel Discharges (tonne of uranium) and Burnup Levels by Implementation Schedule and Year, 1985-2020 With Rapid Implementation Ofscharge Without of Extended Burnup Extended Burnup Burnup Level Discharge Year (t) (GWd/t) (t) 1985 1,250 39.7 900 1986 1,414 41.0 1,021 1987 1,650 43.0 1,187 1988 1,960 44.0 1,447 1989 2,110 46.4 1,553 1990 2,333 48.0 1,653 1995 2,733 52.0 1,527 2000 3,015 56.0 1,510 2005 3,443 60.0 1,873 2010 4,501 60.0 3,197 2015 4,294 60.0 2,632 2020 4,882 60.0 2,513 Source: Weston 1985, Table 2-5.

Note: Data are consistent with 00E/EIA middle case forecast. Burnup level (GWd/t) is the national average burnup at the time of discharge for PWR fusi, with no extended burnup. The second set of forecasts corresponds to the scenario in which extended burnup is adopted as quickly as possible. The columns showing fuel discharaes indicate clearly the decrease in the amount of fuel discharged as a result of the implementation of extended burnup.

level Table 1.6 presents a rough comparison of the three sets of PWR burnup forecasts. Several aspects of this table require explanation. The Murphie and Lang forecasts were converted to burnup levels based on Lang's (1986) most recent estimate of PWR goal (batch average) burnups of 50 GWd/t from the base level of 33 GWd/t. For example, the 1995 forecast of 80% of fuel supplied being extended burnup fuel was interpreted to mean that average burnups had increased 80% of the way from 33 GWd/t to 50 GWd/t. This inter-polation is done to compare the different forecasts of burnup levels on a consistent basis.

The and tent with both the likely schecale shown maximum for the Franksrates.

implementation and Geller data is consis-Slight liberties have also been taken with the reported dates. For example, the 1984 Weston figure is reported for 1985.

Generally, the implementation schedules are in close agreement. Murphie and Lang show full conversion to a.50 GWd/t burnup level by the year 2000, while the Franks and Geller schedule reaches 55 GWd/t. By the year 2005, Weston projects a peak burnup level of 60 GWd/t. There are differences in these forecasts other than the obvious differences in burnup level. For 1-5

_ TABLE 1.6 Comparison of- Pt!R Burnup Level: (GWd/t)

, Forecasts by_ Source and Year,- 1980-2005-f 2

Murphie Franks and- Roy F.

and Lang Geller Weston Year 1982 1986 1985-1980 33 33-1985 35 39 33[3)-

38 1990 42 45 48 1995 47 50 52 2000 50- 55 56 .

2005 50 55- 60 (a) Forecast actually given for 1984 by Weston (1985). .

' example, the Weston forecasts are dated by year of fuel discharge, whereas the Franks and Geller forecasts are dated by year of reactor loading. Conse-quently, .the Franks.and Geller forecasts differ from the Weston forecasts by 5 years because it takes approximately that-length of time from fuel loading to discharge for these burnups to be achieved in comercial power reactors.

The Murphie and Lang forecasts do not state whether they are based on the year of reactor _ loading or-on discharge burnup; however,. it is assumed that they are based onidischarge burnups because their. forecast closely matches-the observed discharge burnups -for 1985.

1.4 i

U.S. LICENSING EXPERIENCE OF REACTORS USING EXTENDED BURNUP FUEL t

Fuel has been supplied to reactors in the U.S. by five vendors: Babcock and Wilcox Co.; Combustion Engineering, Inc.; Exxon Nuclear Co., Inc.; General Electric Co.; and Westinghouse Electric Corp. These vendors have-submitted extended burnup topical reports to the NRC for review and approval. The reports I

summarize their hardware design criteria,' evaluation methods, and supporting-l i

data used for evaluating their designs. These reports have been-reviewed and approved by the NRC (Berkow 1985: Butcher 1985; Rossi 1986; and Thomas 1985c,d) l for their use in support of specific fuel design and plant applications for-operation at extended burnup levels. Consequently, these approvals have not been for specific fuel designs nor for specific plants- for operation at-extended burnup levels.

Specific fuel design and plant submittals_for extended burnup-operation will be reviewed and

approved'by the NRC based-on their.. individual-j merits.

i Two fuel vendors that manufacture BWRL fuel have recently1 submitted _new r

designs for fuel elements that would allow batch average burnups of between

! 40 and 45 Gdd/t. To economically achieve batch average burnups of greater

than 50 GWd/t in both BWRs and PWRs,' fuel enrichments _need to be increased to the extent that significant fuel-design changes will have to be implemented.

Consequently, for-the foreseeable future (i.e., within the next 10 years), it

' is not likely that batch average burnups will exceed 50 GWd/t.- All-vendor t

j 1-6

. - . , , ~ - -,w,

and utility submittals to the NRC and their NRC approvals have been for batch average burnups of less than 50 GWd/t.

Typical fuel reload orders for today's BWRs and PWRs have warranties for batch av3 rage burnups of about 31 GWd/t and 36 GWd/t, respectively (Nuclear Fuel 1984). Studies, based on economics, indicate that present optimum extended Burnups (discharge batch averages) for a 12-month cycle are typically about 45 GWd/t and 50 GWd/t for BWRs and PWRs, respectively, and about 10% higher for an 18-month cycle (Lang 1982a). It should be noted that optimum burnups based on economics can and most likely will change because the studies are dependent on front end and back end fuel-cycle costs that fluctuate. Economic studies have also been made of 24-month cycles (Rosenstein et al. 1986). The Electric Power Research Institute (EPRI) anticipates that fuel assemblies designed for improved extended buenup performance will not be available for full reloads until the late 1980s, and that burnup levels will gradually  !

increase by i to 2 GWd/t per year (Franklin 1982).

Projected benefits to the LWR fuel cycle from extended burnup are discussed by Murphie and Lang (1982) and Lang (1982b). Extended burnup, in ccebination with low-leakage fuel management (see Glossary) techniques, will benefit the nuclear fuel cycle by reducing the generation of spen

• fuel, the requirements for uranium resoure.as, the requirements for separatin work (see Glossary),

and the costs of the fuel cycle. The DOE sponsorea the pre.paration of a report (Matzte 1981) that describes the licensing assessment of PWR extended burnup fuel cycles. The effects of DOE-sponsored devalopmcnt on fuel cycles are described by Lang (1986).

1.5 FUEL OPERATING EXPERIENCE AT EXTENChD BURNUP The NRC publishes a series of snnual reports that provide descriptions of fuel performance in domestic e smmercial nuclear power plants. The seventh report (Bailey and Dunenfeld 1%6) covers fuel performance in calendar year 1984. One of the subjects d % cussed in the reports is high burnup fuel exper-ience. Extended burnup demonstrations sponsored by EPRI were completed in 1983 and 1984, and the usults are described in EPRI 1985. Information from these and other reports is discussed below.

In testimony given at hearings conducted on March 16 and 17,1982, by the subcommittee on Energy Research and Production (Congressional Information Bureau 1982), the DOE indicated that fuel from each of the five U.S. fuel vendors, equivalent tc 22 fuel assemblies, had been irradiated in seven commer-cial LWRs and th&t selected assemblies achieved burnups up to 47 GWd/t. Since that time, experience with high burnup fuel has increased significantly (Bailey and Dunenfeld 1986).

The terms " fuel assembly" and " fuel bundle" a . used interchangeably in the nuclear industry, although generally the former term is associated with fuel for PWRs and the latter with fuel for BWRs. A BWR fuel assembly consists of a fuel bundle and open-ended channel that encloses the bundle. The square array of rods making up a BWR fuel bundle is typically 7 x 7, 8 x 8, or 9 x 9.

The array of rods in a PWR fuel assembly is typically 14 x 14, 15 x 15, 16 x 16, 1-7

or 17 x 17. Figures 1.1 and 1.2 sho;! typical BWR and PWR fuel assemblies respectively.

Over 10,000 BWR fuel bundles have attained burnups greater than 25 GWd/t. Over 6,000 BWR fuel rods nave reached batch average exposures of 36 GWd/t, and over 350 have reached rod average exposures of 40 to 42 GWd/t.

BWR fuel has attained peak bundle exposures of 45.6 GWd/t and peak pellet-exposures of approximately 60 GWd/t (the peak pellet exposure achieved in 8 x 8 fuel bundles is 58.1 GWd/t) (Bailey and Dunenfeld 1986; Baily et al.

1985; Charnley 1985) Over 500,000 BWR fuel rods have reached peak pellet exposures greater than 30 GWd/t.

At least 798 PWR fuel assemblies have achieved burnups of 36 GWd/t or higher. Of those 798 PWR assemblies, 84 have reached 40 GWd/t or higher. Of the 84 assemblies, six have reached 48 GWd/t or higher, ar four of the six assemblies have reached 55 GWd/t or higher. At least 163,o28 PWR fuel rods f D

$ h t r

00

)

0 -i Upper Tie _ $

Plate l ll Fuel Cladding

,o E h 9 Fuel-Rod .

i Spacer Grid ,

' %, {

Removable Fuel Channel f--}

Lower Tie '

$4 Plate End Fitting _

(Nonle Fixture)

(

FIGURE 1.1. Example of a BWR Fuel. Assembly 1-8

I

Rod Cluster Control Assembly.

o Top

-Ai?/%

W ww &

ggg M Control Rod Noule_ ~ ~' M'"} y Yriffidifir#6 n...............

M . ,

+-- Fuel Rod E niini6 .hnn S. -

e k...

Fuel. Rod y 8Hs U s s..p p A

__

• Spacer Grid

-k,,[.....

e.................

"B1IIIIIIllIIIllll 77n v7pp u

g y .+ - Bottom Nozzle

+--0.21 m ---

FIGURE 1.2. Example of a PWR Fuel Assembly 1

have attained burnups of 36 GWd/t or higher, with 16,142 of the PWR fuei rods reaching 40 GWd/t or higher. At'least 1,200 PWR fuel rods have reached 48 GWd/t or higher, and of tnose, 816 have attained burnups of 52 GWd/t or higher.

Four PWR fuel assemblies achieved burnups of approximately 55 GWd/t. Leak testing (sipping) data confirmed that the assemblies were leak-tight, and examinations did not disclose any characteristics that were adverse to fuel performance (EPRI 1983). Results from the DOE-sponsored examination of Westinghouse fuel rods, which were irradiated in the CEN BP 3 reactor to peak pellet burnups of 74 GWd/t and rod average burnups up to 61.5 GWd/t, indicate that the fuel rods are leak-tight and exhibit no unusual performance features (Roberts 1982).

The experience through early 1982 suggests that no sudden or unexpected change in perfomance occurs in progressing to extended burnups (Roberts 1982).

4 -

Secondly, no fuel failures solely attributable to the lifetime of the fuel in 1-9

the reactor have been observed. Thirdly, performance of the fuel is more sensitive to power (i.e., linear heat generation rate) and operating tempera-tures of the cladding than to-lifetime in the reactor. _ General Electric has not experienced a single BWR rod- failure that^ can be related to a high burnup condition (Bally et al. 1985). A recent report (EPRI 1985) states that research on fuel properties at burnups up to 55.GWd/t has shown that no technical-limita-

'tions prevent'such increases (from PWR burnups of 33-GWd/t).- Standard vendor product line PWR fuel has been irradiated to burnups in excess of 55 GWd/t with no reported indication that burnup limits of such fuel _have been reached Babcoe.k & Wilcox observed a loss of cladding ductility (Andrews et al. 1985).

in one Oconee (PWR) fuel assembly that had attained a measured burnup of-50.2 GWd/t (NEI 1985, P The cladding. ductility loss occurred during the last (fifth)yecha et al.1985).

cycle of irradiation. However, the ductility was still found to be acceptable at these burnup levels based on the 1% ductility criteria specified in the NRC Standard Review Plan (NRC 1981).

The goal for DOE ad the nuclear industry is to achieve batch average 1986). These batch burnupsof45GWd/tforB% sand 50GWd/tforPWRs(Lana average burnups correspond to peak rod average burnups {seeGlossary)of50' to 55 GWd/t for BWRs and 55 to 60 GWd/t for PWRs._ In support of this goal, the DOE has sponsored several irradiation programs that have achieved assembly average burnups up to 46 GWd/t for BWRs and 55 GWd/t for PWRs (Lang 1986).

Average burnups for the peak rod in these assemblies were 49 GWd/t and 62 GWd/t, respectively. These irradiations, at extended burnup levels, have been-performed without fuel . failures and with satisfactory performance, as verified by both nondestructive poolside and destructive hot cell examination of the fuel after irradiation (Lang 1986).

The DOE- and EPRI-sponsored programs with the' fuel vendors and utilities have resulted in both BWR and PWR fuel rods being irradiated:to rod average burnups between 40 and 62 GWd/t. These rods have been subjected to extensive nondestructive and destructive tests to determine their relative performance at extended burnups. For example,-those rods that have experienced the highest burnups (around 62 GWd/t rod average) were nondestructively examined visually for surface anomalies and dimensional changes (fuel-rod diameters and lengths).

Also, leak testing, gamma activity scanning,' eddy-current testing for cladding anomalies, and waterside corrosion measurements;were performed. The destruc-tive examinations of fuel rods have included puncturing _ to determine _ the amount of fission gas release, fuel and cladding matallography, measurement of fuel-cladding' interaction, cladding-burst testr . ductility, and scanning electron microscopy to characterize the cladding .de and outside surfaces. -In summary, government and industry irradia. ion programs have provided an in--

depth data base of fuel performance experience at extended burnups.

4 L

1-10 4

(

2.0 PHYSICAL EFFECTS This chapter contains a discussion of the potential physical effects of extended burnup on the fuel and the fuel assembly and a discussion of fission-product release from high burnup fuel.

2.1 PHYSICAL EFFECTS ON FUEL AT EXTENDED BURNUP Potential physical effects of extended burnup on the fuel assembly and its components are presented in Table 2.1 as fuel performance considerations (Roberts 1982). (See Figures 1.1 and 1.2 for examples of fuel assemblies, and Figure 2.1 for an example of a fuel rod.) Included in Table 2.1 are the potential physical effects on the fuel pellets, the release of fission-product gases from the fuel, the fuel cladding, the fuel rod, the bowing of the fuci rod, the guide tubes for control rods, and the spacer grids. Each of these fuel performance considerations is discussed briefly below in relation to extended burnup operation.

Fuel-pellet swelling is a result of solid and gaseous fission-oroduct product *on in the fuel and, thus, increases with fuel burnup. This phenomenon is modeled ir aach fuel vendor's fuel performance code and is accounted for in design and safety analyses of extended burnup operation.

The rate of fission-gas release from the fuel increases at ex'.en di burnup levels, which results in higher internal fuel-rod pressures. This phenomeim is also modeled in each fuel vendor's ' fuel performance code.

The rate of cladding corrosion (waterside) also increases for some plant and fuel designs at extended burnup levels. This leads to a reduction in cladding wall thickness. This phenomenon is also modeled in each fuel vendor's fuel performance code.

Cladding hydriding is a reaction product of cladding oxidation (corrosion) and results in a reduction in cladding ductility; however, for normal operating conditions, cladding corrosion becomes a problem before hydriding significantly reduces cladding ductility. Cladding ductility has been examined at extended burnup leuls to ensure that it is consistent with the strain limits in the NRC Sthndard Review Plan (NRC 1981).

Cladding diametral creep (change in cladding diameter because of creep) is induced by cladding stresses that are generally compressive at moderate burnups; i.e., less than 2.5 to 30 GWd/t, which changes to a net tensile stress at extended burnups (greater than 35 GWd/t fur commercial fuel rods), resulting in an outward claddirig creep. The tensile stress at extended burnups generally results from fuel swelling; however, significant power increases at lower burnup can induce plastic deformation and creep as a result of fuel thermal expansion.

The NRC currently requires internal rod pressures to remain at sufficiently low levels to_ prevent a net outward creep of the-cladding because of this mechanism.

This prevents the fuel-cladding gap from opening up at extended burnups. The NRC Standard Review Plan (NRC 1981) conservatively bounds this situation by 2-1

u TABLE 2.1. Fuel Assembly Performance Considerations Item Consideration Desian/ Performance Impact Pellet Swelling Pellet swelling increases with burnup.

With cladding contact, this can result in pellet-to-cladding bonding and cladding tensile strain.

Fission gas Rod internal Fission-gas release increases with pressure increasing burnup. This is accept-able provided that the internal rod pressure criteria are not uceeded.

Cladding Corrosion / oxidation Increases in accelerating manner -

with increase in burnup because of adverse effect on cladding temperature.

Hydriding Of less concern than oxidation.

Creep At high burnups, cladding could be in tensile creep as a result of fuel swelling and higher fission-gas pressure.

Growth Cladding increases in length with increasing burnup. Clearance required to prevent rod interference with structural components.

Ductility Ductility decreases and yield stress increases with-increasing . fast neu-tron fluence. However, most property changes occur very early during irradiation, and the incremental-effect of higher burnups is not a major concern.

Fuel rod Flattening Avoidance of rod flattening is a licensing requirement. - Adequate prepressurization levels in fabri-cation must be ensured.

Fuel-rod bow Impact DNB(8 This has an impact mainly during the second cycle of operation. For current fuel designs, rod bow is-much reduced in frequency, magnitude, and potential impact on DNB margins.

2-2 1

TABLE 2.1. (Contd).-

Item Consideration Desian/ Performance Impact Fuel assembly Corrosion The fuel assembly Zircaloy guide guide tubes (hydriding) tubes operate at' lower temperature i and with no heat flux compared to--

i fuel rod tubes. Consequently,'cor-

' rosion.-(oxidation)-is not a major.

concern even though it is two sided.

i However, hydriding and subsequent loss of ambient ductility are con -

! siderations, because hydrogen uptake

.-is occurring through

both _ surfaces.

Growth Growth of- the guide: thimbles (overall; assembly growth):' occurs during irradi-ation, but is'quite small. Higher.--

-burnups could cause increased deflec-tion of the fuel assembly holddown spring.

Wei.r Wear of the guide thimble caused 'by vibration of control-rod clusters:

engaged in the fuel assembly.has i been previously observed. This is not burnup consideration per se.

Grid cell Stress relation The fuel rods are held in position by the grid spring forces. Relaxa-tion of these forces may result in l

t vibration of-the fuel rods and t

fretting wear in:the cladding at the point-of; realized spring contact.

! Hydriding For Zircaloy springs, embrittlement' -

from hydrogen uptake is an_ additional-consideration =for high burnup fuel.

! Growth For Zircaloy g*1d straps, growth

_ will occur under irradiation.--

e i-r Source: Roberts 1982.

(a) DNB'= departure from nucleate boiling.

l 2-3 4

u - <w ~,,, -

-e a.,-. +ver e, e , ,eev.-,-se ow+u-.~ -- , -

1 I

l l

C I

l -

Upper End Cap

=

N=

,  ; - Expansson Spring Insulator Wafer g

• Ai Fuel Claddsmg

' (Zircolov 2 on SWR Fuel.

Annulus ~

\

(Pellet. Cladding Geof [

( - .

e ,-

,~,

Fuel PeJiet a N ,

i

, ~ .  ;

, F

, ~ ,

, f W .

lMl Lower End Cao L .a FIGURE 2.1. Cross Section of a Uranium 0xide Fuel Rod for a Comercial Light Water Power Reactor limiting rod pressures to below primary system pressure. Cladding diametral creep is currently modeled by each fuel vendor in their fuel performance codes and other analytical approaches (e.g., in evaluating cladding flattening).

2-4

Cladding and guide tube axial growth is a result of irradiation damage and creep, which increases with burnup.

If the amount of growth is not properly accounted for in the feel rod and assembly design, fuel-rod or assembly bowing can result due to interference between the fuel rod and assembly structural is This phenomena components or between the assembly and reactor components.

modeled by the fuel vendors with analytical methods based on axial growth data from fuel rods and assemblies.

Cladding ductility decreases and yield stress increases with increasina neutron fluence. The ductility at the extended burnup levels anticipated and approved generically by NRC have been found to be within the strain limits suggested in the NRC Standard Review Plan (NRC 1981).

Fuel-rod flattening is a result of ax;al gap formation in the fuel column followed by cladding creep collapse into this gap. Cladding creep occurs This when primary system pressures are greater than internal rod pressures.

phencmenon is modeled by the fuel vendors with analytical methods and, thus, is accounteo for in design safety analysis of extended burnup operation.

Fuel-rod bow is caused by an nimuthal variation in the irradiation and thermal creep of the rod. If rod bow is sufficient to allow rods to come in close contact with each other during irradiation, a departure from nucleate boiling (DNB) and rod failure can result if rod powers are sufficiently high.

Current fuel designs limit the extent of rod bow, and data indicate that od bow saturates (i.e., it does not increase) at extended burnups. In addition, the lower power levels at extended burnups for current designs preclude the likelihood of a reduction in DNB margins.

Assembly guide-tube corrosion is not as significant a probier, as is the case for the fuel cladding because temperatures are lower. The guide-tubes have performed without problems or a significant degradation in ductility in several assemblies irradiated to extended burnups. ,

Assembl, guide-tube wear is due to coolant flow induced vibrations between the control tod ends oad the inner wall of the guide tubes. Guide-tube wear has been observed in some of the older fuel designs in specific plants, however, it appears to have been eliminated in current designs.

The relaxation of grid spring forces is due to irradiation and thermal creep, which can result in fuel rod vibrat' s and fretting wear of the cladding.

This problem has been observed in older den gns: However, it has not been observed in any of the extended burnup irradiations of current designs.

The embrittlement of grid s;. rings has not been observed in any of the fuel assemblies irradiated to extended burnups to date. Also, no problems have been observed in any of the extended burnup irradiations due to grid-strap growth.

Each of the above fuel performance considerations, alcng with ;. hose pro-vided in the NRC Standard Review Plan, have been exaMned and addressed by the NRC in their review (Berkew 1985: Butcher 1985; nossi 1986; Thomas 1985c,d) of each fuel vendor's t- Val reports on extended burnups (Babcock & Wilcox 1982; Exxon 1982; Com; m Ennineerirq 1984; General Electric 1982; and 2-5 i

3 Westinghouse 1982). These reviews have concluded that the computer codes and analysis methods used by each fuel vendor for evaluating the physical t.ffects of fue.1 perfomance have been adequately verified against data up to a parti-cular extended burnup ,1mit. This limit is directly related to the burnup level of the data and is deemed proprietary by many of the fuel vendors.

These data and analysis methods along with design limits will be used by each vendor to show that their particular fuel designs can operate safely w!thout fuel damage up to a limiting burnup level.

4 The NRC revie.:s have identified some performance considerations that need additional attention because of either 1) lack of appropriate data at extended burnups, or 2) design limits being approached at extended t'urnups. Those perforn.ance considerations that need additional data are cladding (fuel rod) and guide-tube growth at extended burnup. In addition, auida tubes (wall I thickness) need tc be examined for particular designs at extended burnups. l The fuel-rod cladding and guide-tube growth are of concern because they can )

- cause interference with assembly and reactor structural components resulting in fuel-rod and assembly bowing. The guide-tube wear appears to be plant and design-dependent and results from flow-induced vibrations between the control rod ends and the inner wall of the guide tubes. Recent cesign changes are i

believed to have eliminated this problem for those susceptible designs; however,  :

data are needed to confirm this at extended burnup. The lack of data for these fuel performance considerations is being addressed by each fuel vendor through experimental and lead test assembly irradiations before the operation of comercial fuel batches at extended burnup.

The perfomance consideration that is closest to the design limits of some fuel designs at extended burnup is the internal rod pressure for the peak licensed rod in a fuel batch. The increased rod pressure results from the increased fission-gas release from the fuel at extended burnup. Each vendor has stated that design and safety andyses will be submitted to NRC for each individual fuel design intended for extended burnup operation to demonstrate 1 that they meet all design and safety criteria. The maximum burnup achievable for each design will vary; however, the mar' mum extended burnup level will not exceed the burnup levels for which the computer codes and analysis methods were 'pprov2d.

4 2.2 FISSION-PRO ~0UCT RELEASE FROM FUEL AT EXTENDED BURNUP

, Two components are involved in determining the amount of radioactive fission gas released from the fuel. The first is the fission-peoduct (see Glossary) inventory in the fuel that is available for release, and the second is the fraction of these products that is actually released to the fuel cladding gap during nomal operation and various accident situations. It should be noted that several barriers exist in a commercial plant before these fission products can be released to the outside atmosphere. The fuel cladding, primary containment, and reactor containment are three of the major barriers. In.

addition to these barriers, there are many other physical and chemical avenues of holdup in the pathway from the fuel rod to the outside atmosphere. The various accident scenarios that can release fission products and their conse-quences are discussed in detail in Section 3.2.

2-6 i

6

4 The change in fission-product inventory in the fuel as a result of extended burnup operation is given in Table 2.2 as a change factor from the inventory calculated at a current burnup level of 33 GWd/t. The inventories and resulting change factors nave been calculated from the ORIGEN2 computer J

program (Croff 1980) for several fission products and actinides (see Glossary) i for the two burnup levels immediately following shutdown (i.e., no time for radioactive decay). The radionuclides given are those contributing most to risk from both normal operation and accidents. As ccn be seen from the table,

, only the long half-life (see Glossary) fission products, such as strontium-90, ruthenium-106, and cesium-134 and -137, increase much with burnup. ihere is, however, generally a larger increase for actinides.

' Because the fractional releases of the fission-product gases into the fuel-cladding gap have an important bearing on the effects of ruptures of the fuel cladding during normal operation and severe accidents, extensive studies of this phenomenon, both theoretical and experimental, have been conducted. The 00r extended burnup programs have obtained fission-gas release data for the no;M gases during normal operation et rod everage burnups (see Glossary) of up 1 62 GWd/t. These data have shown that fisstoa-gas release into the fuel-cla & g i,

gap increases with increasing burnup. A few fission-product release modeb have been developed to explain the increase in r31 ease with burnup. The ANS 5.4 fission-product release model (Turner et al.1982) is one of the more widely known of these models and was developed by an American Nuclear Society

^

group of experts as an "American National Standard for Calculating the Frac-tional Release of V91atile fission Products from Oxide fuel."

The release of volatile radioactive fission products from the fuel to the fuel-cladding gap is of graat importance in this study. The fuel-cladding gap-release fraction (see Glossary) from normal reactor operation is important for those accidents and off-nonnal events that do not involve fuel thermal transi-ents, but which can still lead to fuel-cladding leaks with subsequent releases of radionuclides to the environment (e.g., the fuel handling accident).

Calculations have been performed with the ANS 5.4 release model to deter-mine the maximum possible gap-release fraction of each of the volatile fission products. These results are presented in Section 3.2.2, where the environmental effects of the fuel-handling accident are discussed; however, a brief descrip-tion of these bounding calculations and how they compare to the majority of the gap-release fractions observed in commercial plants will be discussed in this rection, Tc bound the maximum possible releases for normal operation at extened burnup, the calculations performed in Section 3.2.2 are based on the maximun possible end-of-life rod pressure criteria epproved by the NRC, which in turi applies to the peak rod in a reactor core.

It should be noted that the fission-product release from the peak operating rod in sny given reactor core will be substantially greater than those from 95 to 99% of the fuel rods in a fuel batch at extended burnup. For example, 95 to 99% of the rods in any given fuel batch at a batch average burnup of 50 GWd/t have fission-gas (noble) release fractions between 0.015 and 0.029 (Pati and Garde 1905), whereas the calculated peak rod in the batch may have

release fractions two to five times this amount. At the current batch average 2-7 a

TABLE 2.2. Radionuclides Contributing to Environmental Impacts from Normal Operational and Accidental Releases to -

Biosphere Showing the Change Factor at Shutdown Resulting from an Increase in Burnup Level from 33 to 60 GWd/t Inventory (a)

(C1/t) , Change (b)

Nuclide Half-Life 33 60- Factor Tritium and Fission Products:.

H-3 12.3 y 5.4E2 9.4E2 1.7 Kr-87 76 m 4.8E5- 4.1E5 0.87 Kr-88 2.84 h 6.7E5 5.8E5 0.87-Sr-89 50.5 d 9.2E5 -8.0E5 0.87 Sr-90 29 y 7.6E4 1.3E5 1.8 Nb 35 d 1.7E6 1.6E6 0.95

-Zr-95 64 d 1.7E6- 1.6E6 0.95-Tc-99 213,000 y 1.3E1: 2.2E1 1.7:

Ru-103 39.4 d 1.6E6 1.7E6 1.1 Ru-106 368 d 5.4E5 7.5ES 1.4 I-131 8.0 d -1.0E6 1.0E6 - 1.0 1-132 2.3 h 1.5E6 1.4E6 0.99 I-133 20.8 h 2.0E6- 2.0E6 0.98 I-135 6.58 h 1.9E6 1.9E6 0.98

- Te-131 25 m 8.9E5 8.9E5 1.0 Te-132 78 h 1.4E6 1.4E6 0.99 Te-133 12.5 m 1.2E6 1.2E6 0.99 Xe-133 5.2 d 2.0E6 2.0E6 '0.98 Xe-135 9.1 h 4.1ES 4.4E5 1.1 Xe-135m 15.3 m 4.0E5 4.0E5 1.0

-Cs-134 2,06 y 1.5E5 3.7E5 2.5-Cs-137 30.2 y 1.0E5 1.8E5 1.9 Be-140 12.8 d 1.7E6 1.7E6 0.96 La-140 40.2 h 1.8E6 1.8E6 0.99-Ce-141 32.5 d 1.7E6 1.6E6 0.96-Ce-1?4 284.4 d 1.3E6 1.4E6 1.1-Actinides:

Np-239 2.35 d 2.1E7 2.1E7 1.0-Pu-238 87.7 y 2.0E3 7.1E3 3.5 Pu-241 14.7 y 1.2E5 1.7ES 1.4 .,

Pu-243 4.96 h 2.9E5- 7.0E5 2.4 An-241 .432 y 9.3E1 1.6E2 1.7 An-242 16 h 7.1E4 1.3E5 1. 8 --

An-244 10.1 h 4.1E3 1.5E4 3.6 Cm-242 163 d 3.5E4 8.2E4 2.4 Cm-244 18.1 y 1.4E3 8.1E3 5.8 (a) ORIGEN2(Croff1980). See Appendix A.

(b) Inventory for 60 GWd/t divided by; inventory for 33 GWd/t.

2-8

4 i

i l

burnup of 33 GWd/t the majority of the rods (95% or more) in a fuel batch wtth i

an extended burnup fuel design will have release fractions less than or equal l to 0.01, with the calculated peak rod having release fractions three to five l l

times this amount. )'

4

?

A calculation of the release of several radionuclides from the average rod i

in the core at extended burnup is provided in Section 3.1.5,-as well as estimates of average activity incret.se in the coolant from normal operation i with the introduction of extended burnup fuel.

I a

l l i .

1 1

5 a ,

i j  !

l.

l l

J i

i i

f

?

l l

i-i' 9 i

3.0 ENVIRONMENTAL EFFECTS This chaoter describes the effects on the environment of an increase in burnup of nuclear fuel used in comercial power reactors. The environmental effects are addressed in tems of the net change in radiological impacts, considering both normal and accident conditions, of various activities and processing operations of the uranium fuel cycle.

e The uranium fuel cycle, depicted in Figure 3.1, consists of mining of .

the natural uranium, processing the ore in mills to uranium oxide (U03 ellow- l cake), converting the yellowcake (see Glossary) to uranium hexafluor18e, 6 l gas, enriching this gas to higher concentrations of uranium-235, and converting l to UOp. The UO2 is pelletized, sintered, and inserted into rods in fuel fab- i rication plants (see Glossary). These rods are combined into fuel assemblies 1 and shipped to reactors as fuel to produce power. After the fuel reaches a

' specified burnup level, the spent-fuel assemblies are removed from the reactor and stoied at the site in water pools to cool both thermally and radioactively. 1 1

Finally, the spent-fuel assemblies are shipped to either a reprocessing plant (see Glost ry) or a high-level waste repository for final burial. At present the spent fuel has not been moved from the at-reactor storage (see Glossary) pools because no reprocessing plants are in operation and a high-level waste repository has not been constructed.

This discussion covers the incremental changes of radionuclide releases and their impacts that would be produced from both normal operation and acci-dents as a result of the increase in fuel burnup. The following points, which were discussed in Chapter 2.0, should be noted:

• Short-lived fission-product activity in the fuel itself is not appreciably nffected by the amount of burnup. However, the fission products released

into the fuel-cladding gap may increase with burnup, depending on the

- nuclide and accident scenario in question. These fission products are responsible for almost all of the exposure from the normal operation of the reactor or from accidental releases from the reactor.

! . In general, most lor.g-lived nuclides, such as long-lived fission products and actinides, will increase roughly in proportion to the amount of burnup.

However, compared to short-lived fission products, such as iodine-131, these long-lived nuclides contribute insignificantly to exMsures until the inventory of fission products has been aged a few years.

3.1 ELFECTS FROM NORMAL OPERATION In the following discussion, the radiological impacts of increased burnup for the normal operation of an LWR as well as the front and back ends of the nuclear fuel cycle are described. Subsections dealing with effects from parts of the fuel cycle that ore not part of react operation, such as mining and milling (see Glossary), conversion (see Glossary) and enrichment, fuel fabrica-tion, transportation, waste management, and reprocessing, are based on the recent Atomic Industrial Forum's National Environmental Studies Project report on The Environmental Consequences of Hiaher Fuel Burnup (Mauro et al.1985).

3-1 i

ugo

• Reactor p*i ,s Reactor Site Spent Fuel Storage lN N/ V N

Shipping Away From Reactor I

{/ i Spent Fuel Storage e--

Fuel lR Fabrication l. eprocessing iI

\

s s # [iI l

g / 9 al.

i / l.*-.= High Level Wastes N fj s I

'// / p*'g // I N

• Plutonium (PuO j )# / / / l

1. gv +

Enrichment 4 " " utsN [4/ / Interim

,y/ Storage 4

'/ I Conversion P Y to UFe Burial Ultimate g or Storage Disposal Milling I

Fue' cycle steps cur ontly Mining operational

- - - Steps currently not implemented pending operation of reprocessing plants and approval of recycling of mixed oxides as lightwater reactor fuel

/IGURE 3.1. The Uranium fuel Cycle 3-2

.-_ _ _ _ . _ _ .- _ _ - _ _ _ _ _ _ _.~.. _ . . . _ _ . _ . . _ . _ _ _

i I 3.1.1 Minino and Millina i

i

) The amount of uranium mined and milled would be reduced with extended burnup. Although the amount of fuel that would be produced would drop 45% in l inverse proportion to the change in burnup (33/60 a 0.55, or a decrease of j l

0.45), the quantity of ore mined would drop only 5% when the burnup level is

increasedfrom33to60GWd/t(Mauroetal.1985, Table 3-11). This behavior

was determined by performing detailed reactor-physicJ and fuel-cycle analyses to estimate ore requirements as a function of fuel burnup (Mauro et al. 1985).

I During milling, only a 5% decrease in the releases of radium-226, thorium-230, and uranium and daughters is expected when burnup level is increased to 60 GWd/t (Mauro et al. 1985, Table 3-13). Thus, the radiological impacts from these processes woulti be proportionally reduced.

t j 3.1.2 Conversion and Enrichment l The radiological impacts from converting yellowcake to UF would be reduced -l l

l 5%whenincreasingtheburnuplevelfromthepresent33to60bWd/t(Mauroet l di. 1985, Table 3-14) in the same proportion as the quantity of- ore mined.

These reductions arise because of the reduced releases of uranium _and daughters, j radium-226, and thorium-E30 into the environment. The radiological impacts of the uranium-235 enrichment process decrease slightly and then rise with j increasing burnup having a minimum impact at about 45 to 50 GWd/t. The overall

changefrom33to60GVd/tisanincreaseofabout1%(Mauroetal.1985, l Table 3-15). The radiological impact of this slight increase is not considered significant.

i

, 3.1.3 Fuel Fabrication Although the enrichment of -high-burnup fuel will-increase, any exposure to the fuel by fuel-plant workers or the surrounding general population (see Glossary) will be maintained at the same level by operational procedures if necessary. Thus, no change in r6diological impacts to workers is expected.

However, since the throughput decreases inversely with increasing burnup level, the radiological impacts from uranium and daughters and from thorium-234 releases would decrease proportionally (-45%) (Mauro et al.1985, Table .3-16).

3.1.4 Refuelina Impacts associated with refueling of the actors would decrease in pro _-

portion to the amount of fuel throughput, which is reduced inversoly to fuel burnup. Thus, the exposure to workers from refueling operations-is expected '

to decrease 451 when increasing from 33 to 60 GWd/t burnup.

3.1.5 Reactor Operation During normal reactor operation, the fuel cladding surrounding the fuel pellets p m ent radionuclides from entering the reactor cooling water. However, from tis. to time a fuel rod may become defective and release small- amounts of radionuclides into the cooling water. This radioactivity (see Glossary) is then quickly removed by the plant cooling-water cleanup system.

3-3

The normal reactor coolant contamination may be increased by the following three processes:

1. increase in the number of leaking fuel rods

2. increase in the activity (inventory) of radionuclides inside the fuel pellets

3. increase in the activity in the gap between the fuel pellets and the ciadding (gap activity).

Results from the studies discussed in Chapter 2.0 show that the integrity of the fuel rods undergoing extended burnup has not decreased as a result of increasing irradiation (Lang 1986). Thus, fuel-rod leakage is not expected to increase over that from normal burnup for rod average burnups up to 60 GWd/t.

However, a slight increase in inventory and fuel-cladding gap-release fractions will occur for some fission products in those rods at extended burnup.

If leakage occurs, the increased contamination would normally be removed from the reactor coolant by the plant cleanup system, and the reactor would be shut down if ex:essive radioactitit specification limits for the plant)y were detected Data (i.e., exceed technical for the radionuclides that con-stitute the greatest contribution to doses (see Glost,ary) to individuals and populations residing around LWRs are given in Table 3.1. This table shows the relative inventory change factor for each radionuclide due to an increase in burnup from 33 to 60 GWd/t, as presented earlier in Table 2.2. Also pre-sented are the release fraction and relative change factor foi each radionu-clide as calculated using the ANS 5.4 (Turner et al.1982) model for the average rod from a typical normal and extended burnup fuel design. As noted in Sec-tion 2.2, 95-to 99% of all fuel rods in a reactor core at this burnup level will have release fractions near these values. Also given in this table is the overall change factor for high burnup--the product of the inventory and release fraction factors due to the batch average rod increasing in burnup from 33 to 60 GWd/t.

As can be seen from Table 3.1, only cesium increases by a factor of four or more; however, the normal plant cleanup systems can easily handle this increase during nonnal operation. Thus it is anticipated that the environmental impact for normal operation will not change when extending fuel burnup.

3.1.6 Transpor_tation Transportation environmental ef fects respit from transporting fresh fuel assemblies to the reactor site, shipping spent fuel from the site to a repro-cessing or waste-isolation facility, and shipping solid waste to a low-level burial ground (Mauro et al.1985).

3-4

TABLE 3.1, Major Contributory Radionuclides for Burnups of 33 and 60 GWd/t Released into the Cooling System During Normal Operation, and Their Associated Change fators Release Inventory Release Fraction Overall Fraction Change (c) Change (d) 60 (b)

Change (a) Factor Nuclide Factor 33 Factor Cs-134 2.5 9.0E-3 1.7E-2 2 5 Cs-137 1.9 1.4E-2 2.7E-2 2 4 H-3 1.7 (e) (e) 1 2 I-131 1.0 5.0E-3 9.0E-3 2 2 Kr-87 0.87 1.6E-4 3.0E-4 2 2 Kr-88 0.87 2.0E-4 4.0E-4 2 2 Sr 1.8 (e) (e) 1 2 Xe-133 0.98 1.6E-3 3.0E-3 2 2 Xe-135 1.1- 4.0E-4' 8.0E-4 2 2 From Table 2.2 of this report.

The averapa rod for a typical high-burnup fuel design as of 1986.

Rele::: fraction at 60 GWd/t divided by release fraction at 33 GWd/t.

Product of inventory chang factor and release fraction change factor to one significant digit.

(e) The ANS 5.4 model is not capable of calculating release fractions for this radionuclide.

3.1.6.1 Transport of Fresh Fuel Assemblies The only difference in the activity inventory of fresh--fuel designed for high burnup compared to normal fuel would be due to the increase in uranium-235 enrichment from about 3.5% to around St. This increase would have an insignificant effect on exposures resulting from transport of the fuel from

, the fabrication plants to the reactor sites. The reduction in throughput brought about by the increase in burnup level would reduce truck shipments from six to four per year, which would reduce doses received by transportation workers and the general public by about 40%, from 0.014 to 0.0,9 man-rem (see Glossary)-(Mauro et &l. 1985, Table 4-3).

3.1.6.2 Shipment of Spent Fuel Spent fuel is shipped from the reactor site to a reprocessing plant or repository by truck, railroad car, or barge. The doses to the transport workers and population would be reduced as a result of changing to a higher burnup fuel. By increasing the storage time at the reactor-site for high burnt.p

- fuel, the radioactivity in the fuel elements can be allowed to decay so that.

the heat generated in a shipping cask and, hence, the number of fuel' assemblies carried per trip could remain unchanged.

An increase in the emission of neutrons by a factor of 5.6 is expected with.the use of 60 GWd/t fuel. This value was derived from the ORIGEN2 computer 3-5

code using parameters listed in Appendix A. These neutrons are primarily generated by the spontaneous emissions from americium-241, plutonium-238, and curium-244. With minor modifications to the shipping casks (such as extra shielding if necessary and borated cooling water), higher burnup fuel ma carried, while still adhering to the Department of Transportation (00T) y be regulatory limits for radiation (see Glossary) fields outside the cask. One shipping-cask firm that uses " dry" neutron shielding considers their casks to be able to handle fuel at an average burnup level of 55 GWd/t (Planell et al.

1983),

Another company using water for a neutron shield claims most of its casks will be acceptable up to this burnup level after minor physical modifica-tions 1983).or 1% boron is added to the water shield tank (Viebrock and Schreiber Because the number of fuel shipments would be inversely proportional to the increase in burnup (33/60 = 0.55), the radiation impact would be reduced by 45%, assuming the radiation field stayed the same (Mauro et al. 1985, p. 4-12). Table 3.2 shows that the impacts of transporting spent fuel when increas-ing burnup would be equally reduced by 45% for three modes of transport, since fewer trips would be needed per reactor year of operation.

3.1.6.3 Shipment of Solid Waste Solid waste consists of radioactive sludges and resins collecced from contaminated spent-fuel Storage water and reactor coolant. An increase in i solid waste would be caused by an increase in the contamination levels of the cooling water used for spent-fuel storage and reactor operation. Any increase in leaking fuel rods would thus contribute t3 the solid waste. Although che leakage rate of the fuel rods is expected ta remain the same for the higher l

burnup fuel (Lang 1986, Rubenstein and Tokar 1982), slight increases in cooling-water activity could occur through increased inventory and gap-release fraction TABLE 3.2.

Estimated Population Ooses to Workers and the General Public for Three Burnup fuel Types of Spent-Fuel Carriers for 33 and 60 GWd/t i

_ Spent-Fuel Carrier Truck Rail Ba rge 33 6'6~ 33 60 33' 60 Shipments per Year 60 33 10 6 5 3 Annual Dose (man-rem)

Workers 1.2 0.7 2. 1.2 0.04 0.02 General Public:

Onlookers 0.8 0.5 0.1 Along Route 0.1 -- --

1.1 0.6 0.2 0.11 0.03 0.02 TOTAL DOSE 3.1 1.8 2.3 1.4 0.07 0.04 Source: Mauro et al. 1985, Tables 4-6 through 4-8.

3-6

. - - . - . _ . - .. - - . . = - -

(seeTable3.1). Because this activity would need to be removed to keep the cooling water at the licensed technical specifications, an increase in solid waste would possibly result.

In consideration of this, Mauro et al. (1985, p.4-19) evaluated the effect of a 20% increase in solid waste for the case of extended burnup. Taking the trucking case as an example, Table 3.3 shows that the radiological impacts to workers and the general public from the shipment of solid waste would also increase by the same factor (20%) when changing to high burnup fuel. Similarly, doses would change by the same factor for rail and barge shipments.

3.1.7 W a ste Management

'Jsing high burnup fuel would reduce the annual throughput of spent fuel in proportion to the increase in burnt.p. Because of this reduction in through-put, impacts from spert-fuel storage and final disposal in a waste repository would gens ally be reduced. Table 3.4 summarizes the changes in radiological impacts from changing to high burnup fuel for various waste-management activi-ties. In general, the impacts decrease, except possibly for low-level waste

disposal, which could rise through the increase in reactor cooling-water decon-tamination. However, the longer cooling times that wnuld be required before shipment may result in the need for more reactor storage space. The general effect in waste management of increasing fuel burnup is a decrease in the

. impacts.

TABLE 3.3. Estimated Population Doses to Workers and the General Public for Solid Waste Shipments by Truck for 33 and 60 GWd/t Burnup Fuel Truck 33 60 IncreaseinWaste(%) --

20(a)

Shipments per Year 46 55 Annual Dose (man-rem)

Workers 1.0 1.1 General Public:

Onlookers 0.6 0.7 Along Route 0.4 0.5 TOTAL DOSE 2.0 2.3 Source

Mauro et al. 1985, Table 4-10.

(a) Arbitrarily 3.'.ected.

3-7

TABLE 3.4. Summary of Changes in Radiological Impacts for Various Waste-1 Management Activities for Increasing Fuel Burnup from 33 to 60 GWd/t-I Activity 1 Chance Primary Contributors Reference (a) i

Decommissioning -16 Ra-226, Th-230 Table 3-17

! & decontamination -28 U 4

-45 U & daughters

-12 Fission and activation products Low-level waste

disposal 0 to 20 Cs, corrosion, activa- p. 3-26 and [

tion, ar.d fission - p. 4-19 3 products 1

i High-level waste j disposal -45 Tritium, Kr-85, Table 3-20

fission products .i i

t and transuranics l At-reactor storage -8 Kr-85 Tat' '1 l Packaging spent fuel -8 Kr-85 . Table 4-22 I- (a) Sourcc: Mauro et al. 1985 3.1.8 Reprocessing f'

' During the reprocessing of spent fuel, the environmental impacts would be from the exposures of certain long-lived radionuclides in the fuel. These

nuclides would. in general, increase proportionally with burnup. However, the l~ amount of fuel undergoing reprocessing would be reduced-in proportion to the >

burnup.- The combination of these-two effects results in a' net reduction in '

i the exposure and, hence, a reduction in the radiological impacts (see _

j Table 3.5). Nohradiological impacts would be' reduced with increasing burnup j (Mauro et al. 198f, p. 3-34) in direct proportion to the fuel throughput.

4 i 3.2 EFFECTS-FROM ACCIDENTS i The following discussion considers uoth fai+1y 'arge accidents in which 2-thef fuel in the core. has been damaged.or melted und' a- smaller accident in -

which only one fuel rod ruptures. A transportation accident is also discussed.

c h 3.2.1 -Fuel-Damage Accident A' fuel-damage accident is one in which fuel-dcmageLoccurs as, for example, j

in a loss-of-coolant-accident (LOCA) in which-the emergency core cooling is-p

3-8L 7 4

i-4

--_i____J___.._-_______ ,,.b___.,_.---__. ,.m._w,_ 4,.v-rv.- =- + < - -m 'h rr - ~e --v-v"M-s -<e=e w ~

TABLE 3.5. Sumary of Changes In Radiological Impacts from Reprocessing for Increasing fuel Burnup from 33 to 60 GWd/t

% Chance ,_

Contributor

-22 C-14, Ru-106, U

-4 Tritium

-B Kr-85

-3 I-129

-44 1-131

-34 Fission Products and Transuranics Source: Mauro et al.1985, Table 3-24.

degraded. A LOCA is postulated to be the result of a major break in the primary coulant pipe so that the supply of cooling water to the reactor core is abrurcly 7 cut off. If emergency cooling water is not quiekly applieri, the fuel may be damaged or an appreciable fraction of the fuel in the core would melt, and, thus, radionuclides would be released from inside the failed fuel rods to the containment and, if not held up or filtered, to the biosphere.

During an accident in which the core is damaged but no appreciable melting of fuel occurs and the containment system works f airly well, as was experienced at the Three Mile Island reactor (TMI-2), only the most volatile radionuclides will be released from the fuel and possibly pass thr9 ugh _ filters to enter the biosphere. These nuclides include the short-lived fission products such as the iodines and noble gases: iodine-131, iodine-133, xenon-133, xenon-133m, xenon-135, xenon-135m, and krypton-88 (Rogovin 1980, Table 11-1). Because of the short half-lives of these nuclides, they do not increase with burnup. Thus, for this type of accident, no increase in accident consequences woald occur by using higher burnup fuel.

For the release of the fission-product gases from severe a:cidents that involve significant fuel melting (e.g., Class 9 accidents (see Glossary)],

not only are the most volatile radionuclides released, like the todines and noble gases, but a substantial percentage of other semivolatile and nonvolatile fission products are also released, such as cesium and actinides. The NRC source term work (Silberberg et al.1986) has addressed the release of fission products during Class 9 accidents. The percentage of fission-product inventory released from the fuel would not likely change as a result of extended burnup; however, the fission-product inventory in the fuel would change for the long half-life fission products and actinides, as shown in Tabic 2.2 The increase factors from this table show the ch,.nge in relative inventory of each of the radionuclides from the current burnup level of 33 GWd/t to that present at 60 GWd/t.

As can be seen from Table 2.2, nf the fission products, only strontium-90, ruthenium-106, and the cesiums increase much with burnup--less than a factor of three. The actinides increase mere with the increase in burnup. However, 3-9

I during an accident, these actinides contribute only minimally to doses, compared to the fission products. The main concern for these actinides would be from the standpoint of long-term effects of inhalation (lung dose) and ingestion of food products (vegetables, milk, and meat) raiud in or fed from fodder grown in contaminated soil.

Although the actinides, plutonium-238 and curium-244, increase more than twice with increase in burnup (3.5 and 5.8 times, respectively), they do not correspondingly contribute more to the ace.ident risk. This is because the accident risk from these actinides results primarily from the ingestion of food products produced on contaminated soil years after the accident. The fission products, cesium-117 and strontium-90, however, contribute much more to this risk becau:e the release fractions and plant biotransfer factors (see Glossary) are much smaller for plutonium and curium than for cesium and stron-tium. Thus the accident risk of Class 9 accidents is increased only by a

' actor of two with the increase in fuel burnup level from 33 to 60 GWd/t.

In addition, since there is an order of magnitude uncertainty in the risk estimates for accidents, the increased risk from the increases in actinides brought on by the higher burnup level are small compared to these uncertainties.

3.2.2 Fuel-Handling A:cident A fuel-handling accident could be caused by dropping a fuel assembly while removing the spent fuel from the core, placing the assembly in a spent fuel pool, or loading it for transport. This type of accident could result in the rupture of the fuel-rod cladding and subsequent release of volatile nuclides contained in the gap between the fuel and cladding. For current-burnup fuel, the fraction of volatile nuclides contained in this gap (gap fraction) has been conservatively assumed to be 10% (except krypton-85, which is 307.) in Regulatory Guide 1.25 (NRC 1972).

Recent studies by the 002 and the reactor fuel vendors (Lang 1986) have shown that fuel rods with peak burnups of 60 GWd/t suffer no more decrease in cladding integrity than rods of 33 GWd/t. Thus, when changing to higher burnup, the probability of fuel-handling accidents during fuel removal would not increase.

Because the fuel-handling accident is a single-assembly or individual-rod accident, and, in order to bound the possible releases from this type of acci-dent, we evaluated the release from the peak operating rod in a fuel-batch of a fuel design with high operating powers.

The NRC imposes a pressure limit on end-of-life internal fuel rod pressures for normal reactor operation. This limits the amount of gas that can be released for any given fuel design. Traditionally, the internal rod pressure limit has been set equal to the reactor system pressure; i.e., 2,250 psi for PWRs, and 1,050 psi for BWRs (NRC 1981). However, with the introduction of.

extended burnup operation, some fuel vendors have requested (Charnley 1983:

Aisher et al.1977) and NRC has approved (Thomas 1985a,b; Stoltz 1978) internal rod pressure limits that are vendor-specific.- These pressure limits are pro-prietary to the vendor.

3-10 t

e , . - - . . -

The burnup level at which the pressure limit is achieved is dependent on both the design of the reactor and the fuel rod. The expected release fraction of the stable (nonrad4 active) noble gases for the peak calculated rod in a fuel batch is about 0.14 (14%) for a high-power fuel design in both PWRs and BWRs. This estimated release fraction is based on the NRC-imposed pressure limits for normal operation at rod average burnups of 55 GWd/t and 60 GWd/t for BWRs and PWRs, respectively. This is less than the 30% release for krypton-85 thut is assumed in Regulatory Guide 1.25 (NRC 1972) for fuel-handling accidents and significantly less than the 100% release assumed in Regulatnry Guides 1.3 &nd 1.4 (NRC 1974a,b) for a LOCA. The amount of the krypton-85 isotope can conservatively be assumed to stay constant for the accidents con-sidered because of its long half-life (10.7 years). All other shorter half-life isotopes of the noble gases would have lower release amounts because of radioactive decay before their relcase from ihc fuel. For example, for an accident involving the handling of spent fuel, the NRC in Regulatory Guide 1.25 (NRC 1972) stipulates a 10% release for all radioactive noble gases and iodine with the exception of krypton-85, for which a 30% release is stipu-lated. Regulatory Guide 1.25 has been generically approved (Hulman 1982) for evaluating fuel that is to attain batch average burnups as high as 38 GWd/t discharge.

The release fractions calculated for the radioactive iodine and cesium are given in Table 3.6 and compared to those assumed in Regulatory Guides 1.25, and 1.3/1.4 for fuel-handling accidents and LOCA, respectively. As shown in Table 3.6, the release fractions calculated from this study for the shorter half-life isotopes are smaller than those for the longer half-life isotopes. As noted earlier, this is because of the significant amount of decay experienced by the short-lived isotopes before their release from the fuel. A comparison of the release fractions in Table 3.6 shows that the values calculated in this report for extended burnup fuel are all higher than those for the low burnup fuel.

However, the extended burnup release fractions are all lower than those assumed in the Regulatory Guides, with the exception of iodine-131. For this report, we have calculated a release fraction of 0.12 for iodine for a peak rod at a burnup level of 60 GWd/t, whereas Regulatory Guide 1.25 assumes a release fraction of 0.10 (normal burnup).

The release fractions for the shortar half-life isotopes in this report have been calculated from the 0.14 fractional release for stable noble gases using the ANS 5.4 fission-product release model (Turner et al.1982). It should be noted that the ANS 5.4 fission-product model considers only krypton, xenon, iodine, cesium, and tellurium fission products. As per ANS 5.4, the diffusion coefficients for iodine and cesium are factors of seven and two higher, respectively, than those for a ble gases.

Thus, we conclude that Regulatory Guide 1.25 procedures may be used for extended burnup fuel. These procedures give conservative values for noble gas release fractions that are above calculated values for peak rod burnups of 60 GWd/t, except for iodine-131, which may be up to 20% higher, 3-11

TABLE 3.6.

CalculatedReleaseFractionsintheFgj-CladdingGapofthe Peak Fuel Rod During Normal Operation {

at Current and '

Extended Burnup Levels of 33 and 60 GWd/t (rod average)

Compared with Gap-Release Fractions Assumed in Regulatory Guides Calculated Assumed in Assumeu in

' Peak Rod Reg. Guide 1.25 Reg. Guide 1.3/1.4 Nuclide ,33 _60 (Fuel-Handling Acc.) (LOCA)

Kr-85 and Stable Noble Gases 0.04 0.14 0.30 1 Kr-87 0.002 0.007 0.1 1 Kr-88 0.003 0.01 0.1 1 Xe-133 0.015 0.05 0.1 1 Xe-135 0.008 0.02 0.1 1 Cs-134 0.03 0.11 (b) (b)

Cs-131 0.05 0.17 (b) (b)

I-131 0.04 0.12 0.1 0.25 (e) Excludes accidents but includes nomal operational transients as allowed by the technical specifications.

(b) Not given 1. Reguietory Guides.

, 3.2.3 Transportation Accident Environmental impacts from an accident during the transport of tne higher burnup fuel from the reactor to a disposal site or reprocessing plant would increase with extended burnup. The activity of the fiasion products that contribute nost to a nuclear accident source tem, such as iodine-131, are not significant in transportation accidents because they would have decayed before transport begins. However, some of the long lived actinides would contribute to the risk. Table 3.7 shows the major radionuclides contributing to the inhalation impact (effective dose) of ar. accident of such magnitude that all (#f the activity inventory of the fuel is vaporized and inhaled. The activity inventory is that contained in spent fuel after 5 years cooling.

All other radionuclides in the fuel would contribute less than 1% of the dose.

The overall increase facter for the accident impact is 2.7 (sum of effective doses at 60 GWd/t divided by that at 33 GWd/t). The contributing nuclides listed in the table are similar to those previously found to be significant by Mattsen (1979). However, because the use of extended burnup fuel would reduce the need for fuel, the number _of shipments would also be reduced.

With fewer shipments, chances of an accident would be less. The above increase factor of 2.7 muld then be reduced by the factor of 33/60 to account for the reduced shipments:

i 2.7 x 33/60 = 1.5 t

3-12 1

b TABl.E 3.7. Major Contributing Radionuclides Resulting from a Trans-portation Accident Involving the Release of Spent Fuel 1reslation Eff. Does et Eff. Dose et Helf 4.lf e Dose Facter('Ii nventory (Cl/OO ) Increue 33 3d/t to Opd/t Nelide _ Q), _ _(eree/Ci) 33 50 Fetter ,[eres) zIO (eree) {I')

Pv.234 87.7 4.901 2.20 7.3E3 3.4 1.1E'l 30 3.505 40 Pv.241 14,7 1.100 9.2E4 1.3E5 1.0 1.0E.5 29 1.605 il 4 241 432 5.8E11 9.E2 1.3E3 1.8 6.204 15 8.504 10 Co 244 18.1 2.8E11 1.2ES 6.7E3 6.1 3.3E14 9 2.0 Ell 23 Pu 240 6111 6.0E11 5.1E2 6.3E2 1.3 3.004 9 3.9E14 -4 Pv 239 24110 5.8E11 3.1E2 3.1E2 1.2 1.804 6 2.104 2

$r.90 29 1.4E9 8.7E4 1.2E5 1.7 9.4U3 3 1.804 2 i

(s) Ostunsli and Rein 1944, Table D-7.

(b) Inventary af ter 5 yure of cooling. CRIQEN2(Croff1960), see Appendis A.

(c) Percentspe of contribution to the total due.

4 4 Thus, the-Increase in burnup from 33 to 60 GWd/t could increase by 50% l 1

the risk associated with a major accident involving 5-year-old spent fuel.

It should-be noted that the effective doses li:.ted in Table 3.7 are only calcu-lated to find the major nuclide contributors and in no way represent the dose to an individual or population group from an actual accident, since release l fractions or dilution factors were not included.

3.3

SUMMARY

OF ENVIRONMENTAL EFFECTS If leakage of radionuclides from a fuel element occurs during operation, the activity is expected to be removed by the plant cooling-water cleanup system. Thus, with extended burnup, little or no increase in the release of radionuclides to the environment is expected during normal operation. Other 4

parts of the fuel cycle would also not be adversely affected by changing to an i extended burnup fuel utilization plan. The impacts on workers and the general popuistion would actually be reduced, because at higher burnups, outages for fuel changes would be less frequent and fuel shipments to and from the reactor sites would be reduced, thus reducing exposure. Although_ the inventory of f long-lived radionuclides in the spent fuel would increase,.the amount of spent i fuel removed from reacters each year would decrease. Table 3.8 sunnarizes

! the impacts of normal operation, j Accidents that involve the damage or me '.ing of the fuel in the reactor core and spent-fuel handling accidents were also reviewed. For accidents in which the core remains intact, involving only volatile fission products .no increases in impacts would occur, since the radionuclides contributing most to the dose are short-lived and, thus, do_not increase with burnup. For more l l

severe accidents, ones in which an appreciable amount or_all of the_ fuel has melted and released from the containment system into the biosphere (Clas- 9),

only a few fission products and the actinides would increase in_ inventory

with increased burnup. The fissior sducts would increase by no more than a factor of two; the actinides by no mvre than a factor of six (of those con-tributing to the dose). However, since these actinides have very small release i

fractions and biotransfer factors, risks would br.. insignificant compared to I

3-13 i

l ,

TABLE 3.8. Summary of Changes in Radiological Impacts for Various Fuel-Cycle Activities with the implementation of Extended Burnup Fuel a Activity  % Change Primary Contributors Reference (4)

Mining -5 Rn & daughters Table 3-11 Milling -5 Ra-226, Th-230, V &

daughters Table 3-13 Conversion -5 V & daughters, Ra-226, Th-230 Table 3-14 Enrichment 1 0 & daughters Table 3-15 fabrication 45 Th-234, V & daughters Table 3-16 Refuelina -45 Normal reactor O(b) Cs, tritium, I-131, Table 3.1 of operation Kr, Sr-90, Xe this report Transportation:

Fresh fuel -45 U Table 4-3 Spent fuel -45 Fission products Pagt 4-4 Solid wastes 20 Corrosion, activation,

~

.ission products Page 4-19 Waste management See Table 3.4 Reprocessing -44 to 3 Tritium, C-14, Kr-85, U, I-129, I-133, Ru-106, fission products, transuranics Table 3-24 (a) Tables and pages referenced are from Mauro et al. 1985.

(b) There woald be no change in the l';ensed t chnical specifications covering cooling water activity.

x fission products such as cesium-137 and strontium-90. Furthermore, the factors of increases in the radioactive sources are less than the uncertainty involved in determining the overall risk to the public.

For the fuel-handling accident, only the noble gases and iodines escaping the damaged cladding are of significance in the assessment of dose impacts to workers involved. For a peak rod of a high-power fuel design at a burnup level of 60 GWd/t, the release fractions only increase by a factor of three to four for these radionuclides (see Table 3.6); however, they remain below 3-14 1

1

those assumed in Regulatory Guide 1.25, with the exception of iodine-131, Because the calculated iodine-131 gap-release fraction is 20% greater than the Regulatory Guide 1.25 (NRC 1972) assumed value of 0.10, the calculated thyroid doses with extended burnup fuel resulting from a fuel-handling accident could be 20% higher than estimated using the guide.

Spent-fuel transportation accidents were reviewed. Activity inventory may increase by an overall factor of about three for long-lived radionuclides of concern (assuming a 5-year cooling period) when changing to extended burnup fuel. However, this increase would be offset by a decrease in the number of trips, such that the overall change would be a 50% increase in impact by changing to 60 GWd/t burnup.

1 s

4 e

i 3-15 l

4.0 ECONOMIC EFFECTS This chapter examines the economic effects of extending reactor fuel burnup. The scope of this investigation is limited to the directing costs of electricity production, and, accordingly, no attempt is n9de to estimate indirect costs or benefits by imputing dollar values to such factors as changes in radionuclide inventory, accident characteristics, or risk. It is assumed that where indirect effects are potentially laroe changes in power-ger.erating operations will be instituted and, therefore, will be reflected explicitly in direct costs.

In general, the expected overall economic effect of using extended burnup fuel would be a reduction in costs, resulting primarily from the reduction in the total volume of fuel required. Implementing extended burnup in commercial PWR and BWR plants would reduce the requirements for uranium and fuel processing and waste disposal. The resulting reduction in electricity producticn coats outwetohs the associated cost increases in research and development and possible increases in waste storage. The principal finding is that extended burnup is expected to yield a net discounted cost savings on the order of $2 billion.

Most or all of this amount comes from the fuel production and other front-end activities of the fuel cycle.

In Section 4.1, problems e.oncerning the available data are discussed, and problem-solving approaches are outlined. Section 4.2 addresses the front end of the fuel cycle and covers fuel-related research, fuel production and management, and aggregate front-end effects. Section 4.3 contains a discussion of the effects of high burnup fuel on the back end of the fuel cycle, including development ano evaluation of waste management system activities, at-reactor storage of spent fuel, transport of spent fuel, repsitory storage, and waste repository costs. Section 4.4 sumarizes the net economic effects over the total fuel cycle. The factors to which the economic results are most sensitive are discussed in Section 4.5.

The analysis that follows is based on assumptions that were described in '

Chapter 1.0. These assumptions include projections of U.S. nuclear generat-ing capacity (see Table 1.1), the projected implementation schedule for extended fuel burnup (see Table 1.2), and projections of the aggregate average burnup of spent fuel (see Table 1.4). Forecasts of spent-fuel discharges and averaqe burnup levels (Table 1.5) were derived from these assumptione.

Table 4.1 provides an estimate of how the implementation of extended burnup will affect total spent-fuel discharges. [These values are based on theWeston(1985)datapresentedinTable1.4.] In Table 4.1,-the difference column shows for selected years the quantity reduction in spent fuel as a result of extended burnup; the ratio column presents the ratio of the quantity of spent fuel with extended burnup to the quantity of spent fuel without ex-tended burnup. The indicated time pattern shows the benefits of extended burnup in terms of reduced amounts of spent fuel generated--increasing through about the year 2005, then declining until about 2010, and finally rising sharply through the end of the forecast period.

4-1 l

l

1, TABLE 4.1.

Comparisons of Forecasted Spent-Fuel Discharges WithandWithout17gjementationofExtendedBurnup by Year, 1985-2020 Difference, Extended Burnup - No Extended Ratto, Extended Burnup/

Year Burnup (t uranium) No Extended Burnup 1985 -350 0.720 1986 -393 0.722 0.719 1987 -463 1988 -513 0.748 I 1989 -557 0.736 l 1990 -680 0.709 1995 -1,206 0.559 2000 -1,505 0.501 2005 -1,570 0.544 4 4 2010 -1,304 0.710 l 2015 -1,6f?. 0.613 2020 -2,369 0.515

]

(a) Derived from Weston 1985 (see Table 1.5 of this report).

4.1 DATA The scope of this work encompasses the review, analysis, and synthesis of previously published materials. Accordingly, no attempt is made to develop and apply a new cost-astimating methodology for extended burnup or to rerun

. cost- estimating programs used by the various sources. This section relies most I heavily on the tollowing sources: Mauro et al. (1985 , Franks and Geller (1986), Weston (1985), and Dippold and Wampler (1984)).

With a few exceptions, this approach is adequate to develop a reasonably clear picture of the likely effects of extended burnup. Three principal dif-ficulties were encountered, however. The first difficulty arose from the lack of detail in some of the available information. The second difficulty concerned discounting future costs to preserat values, and the third dealt with a recent revision in the DOE /EIA forecast of the middle-growth case of

U.S. nuclear-generating capacity. Brief descriptions of our responses to these

, problems are described in the text. while the technical procedures that were i applied to address them are contained in Appendix B of this report.

The first of these problems--the lack of detailed infor5 nation--affects much of the front end of the fuel cycle. In particular, while some cost and t other data are available for the separate stages of fuel production (mining, milling, conversion, enrichment, and fabrication), the detail is insufficient

, to estimate the effects of extended burnup on the individual stages. However,

.  ; aggregated data do permit estimation of the cost effects for fuel production e as a whole, l

4-2 l

d

The second problem concerned the discounting of future costs to the pre-sent. Throughout this report a 10% real discount rate is used. This rate is required by the U.S. Office of Management and Budget in the evaluation of time-distributed costs and benefits by agencies of the executive branch of the federal gcvernment (OMB 1972). However, several of the sources for thi.

report (Weston 1985; Dippold and Wampler 1984; Murphie and Lang 1982) implicitly use a discount rate of zero when they derive aggregate costs and benefits by adding undiscounted future amounts. Another source (Franks and Geller 1986) 4 uses a continuous discount rate of 7.813%.

This cost issue is troublesome because the time profile of these amounts  !

is not supplied by the source. The procedure used to deal with this problem was to start with information or assumptions about the time profile of the l annual money amounts. Then the total amount reported by the source was "de-discounted" (recalculated without the discount) and disaggregated (see Glossary) l back to a series of annual amounts.

)

The application of the above steps produced a set of estimated annual amounts that correspond to the original aggregated amount from the source.

These amounts were neat discounted at 10% and summed to arrive at a present ,

value consistent with the analysis of the original source and a 10% discount rate.

i For other figures with a similarly assumed time profile and original discount rate, repeating this entire procedure was unnecessary. The ratto of the adjusted present value to the original value yields a discount adjustment factor. Multiplying this factor by an original amount with a similar time profile and original discount eate gives the adjusted present value.

These discount adjustment factors vary according to how much the original discount are expectedratetodiffers occur.from 10% and how far into the future the annual amounts As an example, the Weston (1985) estimate of the cost i

~

savings in dry at-reactor storage from implementing extended burnup is $304 million. Multiplying this amount by the discount adjustment factor of 0.1565 yields a discounted present value of $47.6 million. However, the effects on the capital costs of the first repository ocuur mostly in earlier years. s This effect ;s reflected in the greater discount adjustment factor of 0.4914 for these costs.

Depending largely on the assumptions concerning the time profile of the amounts, discount adjustment factors range from lows of 0.0356 for costs relatad to repository closure and 0.0777 for operations costs of the second repository to 0.4914 for the capital cost < of the first repository.

The third major difficulty with the data provided by the primary sources is that they are based on an earlier DOE /EIA (1985) middle-growth forecast of U.S. nuclear generating capacity. The most recent middle-case forecast shows lower figures for nuclear generating capacity until the year 2005.

that year, the new forecast is higher than the old. The assumptionAfter that the aggregate effects of extended burnup are closely related to nationwide nuclear generating capacity imclies that the more recent forecast shifts these effects 4-3

into the more distant future. The discounted effect of the change in the forecast, therefore, is a reduction in the magnitude of the economic effect.c.

An estimate of the size of this reduction is developed by calculating the ratio of the discounted value of the new capacity figures to the discounted value of the old capacity figures. This procedure follnws from the assumption that economic effects in any year are proportional to the nuclear capacity for that year. The result of this calculation is a factor of 0.9771. This figure implies that the effect of the change in-forecast is to reduce the expected present value of the economic effects of extended burnup by about 2.M. In terms of the general accuracy of forecasts and the uncertainty in the discounting adjustments, this figure is not highly significant, but suggests that the effect of using data based on different revisions of the capacity forecasts is relatively minor. Therefore, except where specifically noted, no adjustment is made for the most recent forecast.

4.2 FRONT-END EFFECTS The primary source of information for front-end economic effects of ex-tended burnup is the Franks and Geller (1986) report, The Benefit of Extended Burnup in Fuel Cycle Cost. The front end of the fuel cycle includes researcli and development expenditures on reactor fuel assemblies, fuel production, and burnup of the fuel in the reactor. Implementing high levels of extended burnup in comercial reactorc may require additional fuel-related research expenditures.

I Fuel production spr.ns the operations of mining uranium ore, milling the ora to produce uranium oxide (U 0 ) or yellowcake, converting the yellowcake

to uranium hexafluoride (UF6), 3 8 enriching the uranium by increasing the per-centage of uranium-235, fabricating the fuel pellets and fuel-rod as,semblies, I and transporting the fuel to the reactor. During burnup cf the fuel in the reactor, heat is used to generate steam, wl.ich in turn drives generators to '

oroduce electricity.

The primary economic effect of extended fuel burnup on the front end of the fuel cycle is a significant reduction in the amount of fuel required to generate an equivalent amount of electricity. The fuel requirements are reduced in direct relation to the increase in average burnup. Thus, cost is reduced because of the diminished amount of fuel required. However, there is an offset-ting effect because the optimal enrichment of the Tual increases as the average burnup level increases. Generally, sources indicate that the cost reductions attributable to lessened fuel requirements will outweigh the additional expendi-tures for enrichmex: (Brown et al.1986; Delene 1984; Franks and Geller 1986; 4

, Murphie and Lang 1932).

! 4.2.1 Front-End Research While no insunnountable technological obstacles to extended burnup appear to exist, the attainment of batch average burnup levels of 60 GWd/t in commer-i :ial reactors will require, according to several sources, additional expendi-

! tures for research. For example both Franks and Geller (1986, pp. 1-1, 3-6)

and Roy F. Weston (1985, p. S-2), developed alternate sets of forecasts of q 4-4

burnup levels based on different assumptions regarding the level of DOE research funding. According to the Franks and Geller (1986, p. E-2) report:

00E has estimated the costs for the research and development program to provide new support for extended burnup as $35 million in as-spent dollars. In discounted 1985 dollars at a continuous discount rate of 7.813% per year this equals $22.3 million.

The schedule of these projected research expenditures is shown in Table 4.2.

Using the same DOE fig'tres, but a discount rate af lot instead of 7.813%,

yields a 1985 present value of $21.3 million.

4.2.2 Fuel Production The primary source of information concerning front-end costs is Franks and Geller (1986). Because the effects are reported as an aggregate over the whole front end of the fuel cycle, the information is insufficient to quantify the effects of extended burnup on each stage of fuel production and on fuel management. However, some aspects of the stages of fuel production are useful in inferring the general direction and relative magnitude of the economic effects of an encompassing change, such as implementing extended burnup. In particular, other available sources indicate that the various stages of fuel production (mining, milling, conversion, enrichment, fabrication, transport) can be characterized by relatively constant per unit costs over any foreseeable range of output (e.g., Franks and Geller 1986, p. 2-4; and Brown et al.1986,

p. 7 85). The enrichment stage is an exception in that costs are a function of the level of enrichment as well as the volume of UF6 , and the optimal enrich-ment level is higher with extended burnup.

TABl.E 4.2. Projected DOE Research and Development Expenditures for New Extend Projects by Year, 1987-1994g)Burnup Expenditure Year (millions of 1986$)

1987 6 1988 6 1989 3 1990 2 1991 3 1992 3 1993 6 1994 6 Source: Franks and Geller 1986,

p. E-5.

(a) Projections from this source do not extend beyond the year 1994 4-5

, _ _ . _ J

! 1his apparent absence of any significant economies or diseconomies of scale for most stages of fuel production is fortunate for our present purposes

for two reasons. First, it implies that the economic effects on any stage are roughly proportional to the change in the output of the stage. Second,

because the stages are all part of the same linear production stream, the

output of each stage will change by the same proportion. (See Mauro et al.

1985, pp. 2-4 through '2-15.) In other words, ignoring any increased enrichment levels, the implication is that if fuel requirements decline by 10%, a 10%

decline in the output at each production stage will occur, and total fuel costs will decline by 10%. This simplifies the analysis because the imple-mentation schedule for extended burnup determines the amount of fuel to be

produced, which in turn is proportional to costs. In practice, this relation-l ship cannot be expected to hold rigidly. Further, we are not suggesti_ng that prices are immune to change from other directions. In particular, reduced demand for fuel may result in prices being somewhat lower than they otherwise would be.

Another factor bearing on the level of fuel production is whether spent fuel is reprocessed. The extraction of some of the' remaining uranium-235 from used fuel reduces the need for new fuel production. However, because there is currently no comercial reprocessing, nor is there likely to be, at least in the near term, it is assumed in this work that there will be no re-i processing. One source (Mauro et al. 1985), however, where relevant, provides

< 1 two sets of data--one for the case with and one for the case without reprocess-

, ing. The reprocessing figures "are based on reprocessing of 35 HT of spent fuel per RRY [ reference reactor year (see glossary)], at a burnup of 33,000

, MWD /HT" (Mauro et al. 1985, p; 2-19).

1 4.2.2.1 Minino 3 A primary feature of and motivation for extended burnup is that it would i reduce the amount of fuel required to produce a given amount of electricity.

The diminished fuel requirements translate into a-decrease in the amount of uranium ore required to be mined, as well as a decrease in milling, conversion, and fabrication required. Estimates vary as to the magnitude of the resulting decline in uranium mining. One source states that typically, extended burnup

) "is expected to reduce uranium mining and milling and UF6 production require-

. 4 ments by 14-15 percent" (DOE 1980a, p. 9). A similar estimate--a 15% reduction i in uranium resource requirements--is provided by Murphie and Lang (1982, p.7-67). Even greater savings are reported by Brown et al. (1982). Brown

! and his coworkers indicate that employment of various optimizing measures t

could yield a potential reduction in uranium requirements on the order of 29%

. I for the Browns Ferry Unit 3 BWR (Brown et al.1986, pp. 7-71, 7-81).

In analyzing environmental imp? cts of extended burnup, another source,

the Atomic Industrial Forum, Inc. _ (Mauro 'et al .1985, pp. 3-13, '3-16),- comes i to a more conservative conclusion. This source suggests that as the burnup level is increased, the environmental impacts associated with mining decrease

in an asymptotic fashion (i.e., the rate of decrease lessens with increasing.

burnup). At the upper end of the range, an increase in burnup from 55 to 60 GWd/t would result in no change in environmental impact from mining. Over j the whole range from 33 to 60 GWd/t, there would be a 5% decrease in impacts.

e

! 4-6 t

4

This concluston follows from estimates of how a reactor's yearly requirements

< for uranium ors would change as average burnup is increased from 33 to 60 RRY Table 4.3 shows estimated uranium cre requirements for an RRY.

GWd/t's figur are intentionally high to ensure a conservative bias when estimating prospective environmental and other impacts.

4.2.2.2 Millina j

In the milling stage of fuel production, uranium ore is arocessed into a more concentrated form called yellowcake (c e Glossary). Witi extended fuel burnup, milling requirements would decline along with other stages of fuel production and would also depend on Mether spent fuel is reprocessed.

Table 4.4 provides estimated RRY milling requirements in tonnes of yellowcake.

In analyzing the economic effects of extended burnup on the front end of the fuel cycle, the Franks and Geller report used unit costs, as indicated in i Table 4.5.

f. i 4.2.2.3 Conversion l i

The procedure for converting yellowcake to UF6 is a straightforward chemi--

5 i

cal process, the output of which is closely proportional to the input (Mauro l

etal.1985,pp.2-13,6-7). Estimated conversion-work generally decreases at a decreasing rate with increases in the burnup level.

i Table 4.6 shows forecasts of conversion costs used in the Franks and 4 Geller report. These figures were developed in a manner similar to those in

Table 4.5. '

4 t

f TABLE 4.3. Annua 11 zed Deference Reactor Year Uranium Ore Requirements With and Without Reprocessing, by Burnup Level

' Uranium Ore (t)

Burnup With Without 2

' (GWd/t) Reprocessina e Reprocessina 33 222 280

35 221 273 i 40 217 241

45 213 237 50 212 236

!- 55- 211 234 1 60 211 ' 234 i Source: Mauro et al. 1985, p. 3-14. Uranium ore requirements without-reprocessing were-

. calculated from yellowcake requirements with reprocessing using the Mauro et al. '

assumption of a constant yellowcake-to-ore i ratio of 0.9.

4-7 e

w-b- vg, - - - - - -.+ g -- +p p yg _ , , pe -r

TABLE 4.4 Annualized Reference Reactor Year Milling Requirements With and Without Reprocessing, by Burnup Level Yellowcake (t)

Burnup Level With Without (GWd/t) _

Reprocessino Reprocessina 33 200 252 35 199 246 40 195 233 45 192 225 50 191 219 55 190 215 60 190 212 Source: Mauro et al. 1985, Tables 3-11 and 3-18.

1 TABLE 4.5. Forecasted Milling Conts by Year, 1985-2000 Yellowcake Cost Year (1985$/lb) 1985 30.50 1990 44.20 1995 51.60 2000 51.60 Source: Franks and Geller 1986, Table 2-1.

TABLE 4.6. Forecasted Conversion Costs b Year, 1985-2000 (1985$/kg UF6 Year Cost 1985 7.40 1990 7.40 1995 7.40 2000 9.50 Source: Franks and Geller 1986, Table 2-1.

4-8

. = _

t 4.2.2 4 Enrichment Fuel enrichment involves increasing the proportion of uranium-?.35 in the fuel. As Weston (1985, p. 3-4) states:-

Enrienment is~ the initial proportion of fisille uranium in the fuel.

The higher the burnup desired.from nuclear fuel, the higher the required enrichment. Normal enrichment for reactors is about 3.5 percent, and the . . . NRC limit {3 fuel mannfacturers to producing fuel with no more-than 5 percent enrichtnent for PWRs. As th'e fuel undergoes fission reactions-in the reactor,_ the enrichment diminis',es.

The consequent increase in separative work-necessary to achieve these higher-enrichment levels _ is illustrated in Table 4.7;,

The figures shown here.are for an RRY, which means'that they reflect the ;nax4 mum: levels likely to occur, since they were developed to estimate environinental. effects.

The prospect of new methods of fuel enrichment opens the possibility of significant reductions in. cost-in future years. According-to Delene (1984,

p. 2-E):

The current enrichment price is $135/SWU [ separative work unit].

This prict will escalate at approximately the rate of-inflation under current rules. However the: advent of the advanced centrifuges and the laser isotope separation hatz the prospect of-considerably lowering separative work costs. DOE projections show SWl' es declining to $70-100/SWU by the year 2000 doe to these a. 3 technologies.

TABLE 4.7 Annualized Reference Reactor' Year Separative Work Unit Requirements With and Without' Uranium Reprecessing, by-Burnup Level.

Separative Work Units Burnup Level With Without i

(GWd/t) R;eprocessing -- Reproce s s i ng -

33 134- 140 35- -134 138 40 133 135 45 132 133 SO 132 133 55- '134- 133 60 135- 134 l

Source: Mauro et al. 1985, Tables 3-11 and 3-18.

[a] Note that 10 CFR -51.52 limits enrichment to 4% by-weight, rather than the 5% ttated by Weston.(1985).

4-9 i 1 l

.- - . -- .. _. = - -.  :

TheFranksandGeller(1986)reporttakesaconservativeapproachinestimating front-end economic effects of extended burnup. A constant cost of $135 per GWU is assumed for all years.

4.2.2.5 Fabrication Extended burnup poses some possible technical problems with fabrication.

Fuel assemblies must be able to be retained in the reactor core for longer periods without failure. Principal concerns are increases in pressee in the

=

fuel rods caused by tne release of fission gases, cladding corrosion, fuel-pellet expansion, and fuel-rod integrity. Because of these technical concerns, most sourccs anticipate that fuel-pellet and fuel- td assembly operations will v experience an increase in cost.

A. cording to the Franks and Geller report '1986,pp.4-5,4-6):

Increases in fuel burnup may require te<:t- 41 modifications to the fuel designs that i' cream febricat<en w sts. . . . Because of possible increased to.,. cation costs, two cost schedules were used.

In the first the cost was kept constant at $230/kgU; in the second, it was allowed to increase with target discharge burnup.

Franks and Geller estimated a cost of $245 pu kg of uranium for PWR fuel 3

designed to undergo burnup levels in the over and up-to-50-GWd/t range.

Above this level, the estimated cost is $273 per kg of uranium. Franks and Geller also estimated a cost of $265 per kg of uranium for BWR fuel for burnups above 33 and up to 45 GWd/t.

y{l l This pattern of cost increases is confirmed by Delene (1984, p. 5):

. t Extended burnup fuel fabrication costs are assumed to cost from 10-25% more than that for standard fuel. For the purposes of this study, extended fuel burnup reference fabrication costs are assumed to increase by 1% cver the refuence standard fuel fabrication cost for each mwd /kg the fuel exceeds 30 mwd /kg.

The price level for fuel fabrication to be used at the 60 fiWd/t burnup level implied by Delene is higher than that indicated by Franks anc .eller--$299 per kg versus $273 per kg of uranium.

Offsetting the increased per unit price of fuel fabri. . ton is the decrease in the amount of fuel required at higher burnup levels. This reduction in terms et RRY rec:irements is shown in Table 4.8.

Applying the escalating Franks and Geller PWR fuel fabrication price schedule to tha fuel requirements in Table 4.8 gives an annual (RRY) fuel fabrication cost of about $8.1 rr.illion at a burnup level of 33 GWd/t. For a burnup of 60 GWd/t, the annual fuel fabricatica cost would decline by nearly

$3 million to $5.3 million, which is a decline of nearly 35t. Performing the same calcu;ations using the fuel fabrication price increase suggested by Delene, the annual fuel fabrication cost per RRY would be about $2.3 million to about

$5.8 million.

4-10 9

TABLE 4.8. Annualized Reference Reactor Year Fuel Requirements by Burnup Level Burnup Level Fuel Mass (GWd/t) (t) 33 35.0 35 33.0 40 28.9 45 25.7 50 23.1 55 21.0 00 19.3 Source: Mauro et al. 1985, Table 3-11.

4.2.2.6 Transportation The decrease in required fuel mass as a result of implementing extended burnup would translate directly into reduced transportation requirements as shown in Table 4.9.

TABLE 4.9. Annualized Reference Reactor Year Transportation Requirements for Unitradiated fuel, by Burnup Level Burnup Lc' 'l Number of (GWd/t'._ _ Truckloads 33 6 35 6 40 5 45 5 50 4 55 4 60 4 Source: Maaro et al .1985, Table 4-2.

4.2.3 Fuel Management Based on fuel-management consideratione slone, there would be savings from using extended burnup fuels (Delene 1984, r. 1). Extended burnup can be achieved either by extending the time betvaen-refuelings or by increasing the ,

number of batches while maintaining annual refueling, )

At a burnup of about 30 mwd /kg, a standard fuel batch stays in the I core for 3 cycles so there are 3 batcua in the core at one time. .  !

. . Alternatively, the number of batches in the core can be held j i

4-11 l

en

^

constant and the cycle time increased. This latter case leads to higher fuel enrichments as fuel burnup is increased than the case where the cycle length is held constant. The longer cycle length offers other advantages such as increased reactr availability brought about by the decreased number of shutdewns for ,sfueling [Delene 1984, p. 7].

4.2.4 Aggregate Front-End Effects Franks and Geller (1986) calculated the 1985 present value of front-end cost savings through the year 2020 as $500 million. This figure reflects costs resulting from DOE-funded research for the DOE /EIA (1985) middle-growth forecast of nuclear generating capacity. For the no-new-orders forecast, the estincted cost reductions amount to $490 million.

Unfortunately, while these Franks and Geller figures are the best available figures for front-end cost estimates of extended bu qup, they are inconsistent with others in the current DOE /EIA report. First, the Franks and Gel'.er esti-mates were calculated using a discount rate of 7.813% rather than 10%, as in the DOE /EIA report. Second, recent revisions in the DOE /EIA middle-growth forecasts are not reflected in Franks and Geller's estimates. Third, Franks and Geller adopted the convention of including at-reactor storage in the front i end, and, therefore, the totals ,hould have these costs subtracted. The last p and most significant problee with the Franks and Geller cost estimates is y that they incorporate a different pair of scenarios than the one with which

] this work is concerned. Franks and Gcller calculated the net cost savings of extended burnup with adaitional DOE-funded research versus extended burnup without this research. Of interest here, however, is extended burnup (with u additional DOE-funded research) versus no extended burnup.

N 3 While the Franks and Geller report provides a description of the methods

/I used, intermediate results are not supplied. In particular, no time pattern of costs or cost savings is given. Therefore, any_ adjustments are, to some extent, necessarily crude. A detailed description of the method used to adjust these data to a 10% discount rate and to adjust for revisions in the DOE /EIA forecasts is oiven in Appendix B, Section B.1.

Applying the procms described in Appendix B, Section B.1 gives an esti-mated cost savings of $2.6 billion for the years 1985 through 2020. This

amount is substantially higher than that estimated by Franks and Geller because

1) our schedule of extended t'Jrnup implementation assumes higher overage but..up levels than Franks and Geller's levels, and 2) the base-case burnup leveis used here are constant, whereas Franks and Geller's levels increase over time.

O

~~

Two simple adjustments remained. The first was to eliminate the cost of

.1 at-reactor storage. As noted below, the adjusted Weston (1985) estimate of at-

' reactor storage costs converted to 1985 dollars is $47.6 million. Second, the d cost savings needed to be reduced by the cost of the DOE-funded rerearch,

. which was calculated above at $21.3 million. Subtracting these amounts yields

a net cost savings of $2.53 billion for the front end of the fuel cycle.

l, H 4-12 i

! I

, I' 4.3 BACK-END EFFECTS i

In this work, "back-end" refers to those stages of'the fuel cycle starting Also with discharge from the reactor core through final repository storage. The included are the devel.opment and evaluation related to these stages.

cost effects of extended burnup on these stages of the fuel cycle are described i

l in this'section. .

\

Extending fuel-burnup changes the composition of radioactive isotopes in

i the spent fuel,-which in turn causes a change in the thermal and radiation-

' characteristics of the fuel. - The increased neutron activity of the waste may require more extensive shielding. If this is so, the capacity of the spent 4

I fuel transport _ casks may need to be reduced, and thus, the number of casks and Increased co.ts for these operations

! waste' shipments may need to be increased.The increased heat discharged by the extended on a per-cask basis would result.

j

' burnup spent fuel may necessitate greater spacing between fuel casks in the

, repository, implying a greater repository cost per-fuel cask.

Extended burnup generates a smaller volume of waste than current burnup levels, which would offset the cost increases described above. The total radio-

' activity per useit of power generated (i.e., the number of curies) that is

' processed in the repository would remain unaffected by extended fuel burnup (Weston 1985, p. 2-19). Therefore, while the increased-radioactivity and increased heat discharge would result in higher costs per cask, fewer casks would be needed, 4.3.1 Development and Eval,uation i

Development and evaluation (D&E) consist of all siting, design, develop-ment, testing, regulatory,: andiinstitutional activities associated with the waste-management system and account for a major portion of the total system costs.- Despite the size of this cost category, various sources indicate that extended burnup would have no significant effect on the level of costs: "D&E i, costs are considered to be on?y minimally affected_by changes in burnup" (00E l 1985a, p. 10). "The D&E costs represent about one-third of the c.st of the total waste-managetent system. . . . However, the D&E costs are not expected to be significantly affected by extended fuel burnup" .(Weston 1985, p.1-5).

4.3.2 At-Reactor Storage The Weston (1985) report estimates dry at-reactor storage (ORS) costs at-

}

$720 million for the'00E/EIA middle-case forecast with no increase in burnup (see Table 1.1 and accompanying discussion). At an average burnup level of Expressed in 1985 60 GWd/t, estimated DRS costs would decrease to $420 million.

i dollars, these amounts are $729 million and $425 million, respectively, with-a difference of $304 million. For alternative assumptions about the rate of implementation of extended burnup, Veston found that cost reductions for DRS "are nearly' proportional to the decrease in sp;nt-fuel generation rates" (W6ston 1985, p. 4-21).

4-13 t

i i- '

.Whereas Weston is unclear regarding the discounting procedure, his figures i

- appear to be calculated as sums of undiscounted annual amounts. Unfortunately, Weston does not give the actual costs over time for different burnup levels, so it is difficult to interpret his figures. However, a rough approximation of the discounted value of the cost savings can be made by applying the method-ology outlined in Section 4.1. Cost savings over time were estimated from a graph showing levels of required dry at-reactor storage by year and burnup level (Weston 1985, p. 3-7). The calculated discount adjustment factor 1s-

. 0.1565. By multiplying this factor by $304 million, an estimated present value of $47.6 million in cost savings for at-reactor storage was reached.

The reason the discount adjustment factor is so small is that the savings in this cost category do not begin until 1991, and the peak savings do not occur until 2004.

4.3.3' Transportation Following temporary storage at the reactor site, spent fuel would be transported to one of two nuclear waste repositories for-final disposal. The spent fuel would be shipped in special ca ns designed to meet requirements of-radiation shielding and heat dissipation. These casks would be a major com-ponent of transportation costs.

The shielding and heat dissipation requirements per unit mass of spent fuel increase with higher burnup levels. However, the higher burnup levels also result in a reduction in the amount of fuel discharged per unit of electricity generated. Increased duration of _at-reactor storage pemits the waste to cool and become less radioactive, thereby complying with the i

transportation-cask minimum specificattoi.s.

Different-authors hsve different views as.to hw cask volume will be affected by extended burnup. According to a 00E st idy-(00E 1985a, p. 7),

Generally, the currently available transportation casks could carry extended burnup fuel at their design capacity, unless limited by--

! criticality concerns. . . . The Nuclear Regulatory Connission-requires that transportation cask capacities be detemined under the i

assumption that the fuel to be transported is fresh unirradiated fuel. This may result in reduced cask capacity.

1'

Dippold'and Wampler found that cask volume, as part of i joint transportation and repository optimization analysis, would decrease with extended burnup (Dippold ana Wampler 1984, p. 33 and Table 6, p. 37). .Weston's principal finding concerning back-end transportation (Weston 1985, p. S-13) was that the )

l reduction in the quantity of spent fuel results in a comensurate I reduction in transportation costs because transportation-cask captci- l ties are generally not affected by extended-burnup fuel. Therafore, I transportation costs show a downward trend with increasing bu w .p. l

' l Two sets of- cost estimates on the effects of-extended burnup on back-end transportatior costs are available. . Table 4.10 sumarizes the Weston (1985) estimates of total back-end transportation costs for different burnup levels 4-14 1

, . _ _ . _ _ , . . - , _ - , _ , ~ . _ _-.-. .

.1 L

i F

. TABLE 4.10. Back-End Transportatio'n Costs, 1984-2020, by Repository Type and Burnup Level-

! (millionsof1985$,undiscounted).  ;

i

Repository Type-Salt Tuff Basalt and' and and l Burnup Level- Granite Granite - Granite- .

33 GWd/t:

Capital - 424 461 468

,. -- Operation 1,657 2,003 2,062 1

TOTAL 2,081 2,464 ' 2,530

, 60 GWd/t:

! Capital 271 298' 302

j. Operation _1,119- 1,329 h3,71 I

TOTAL _ 1,350 -1,626 1,673-l Cost Difference: '

! Capital 153- 163 166:

-Operation- 538 674- 691 t

TOTAL 731 838 857

[ Percent Decrease:

Capita 36.1 35.4 35.5 g Operation 32.5 33.6 -- 3 3 . 5 -

p

! TOTAL 35.1 34.0 33.9 Source: Weston.1985, pp.'4-11, 4-13, 4-15;.J Costs.

- - expressed in 1984$. in original source were

i converted to'1985$ using the gross national i product (GNP) implicit price deflator.

- Data:are consistent _with DOE /EIA middle-growth-

case.

p.

- and for different repository types. The total.back-end transportation costs.

are broken down -into .capitalc and operation categories. . . Capital includes " spent-2 i

fuelLassemblies" or shipping casks. Operation costsJare composed of cask-hauling, cask maintenance, and traffic management. Weston assumes that,: " con- -

sistent with tha . . -. Mission Plan -[00E _1985b], -there are two regional repost-

[ tories, one in the east and. one in- the west, 'and all- spent fuel is sent to 4

one of these repositories.s. . . . [The] eastern repositor

- crystalline rock-- (granite)* -(Weston 1985, pp. 3 to 3-2)y . TheiswesternLreposi-assumed to be :in -

tory is-assumed to be in either salt, tuff,' or basalt formations. Because p l repositories of.different geological type.would be located;in different parts

4-15 1

y - -- - , , y , y, -,,---,.-m r ., .,y 9 y, # w , ,.7y. ,, y,,,,.v..,.. ,,,m ....,m,,,,,,,va,.w-.,r, . . ,

.% mr , v.,

of the country, estimated shipping distances and, therefore, estimated trans-portation costs would differ by repository type. The a is assumed to be at least 5 years (Weston 1985, p. 3-1)ge of the spent fuel Table 4.11 presents a set of estimates for back-end transportation costs developed by Dippold and Wampler (1984). The capital and operation cost cate-gories are the same as in the Weston data. The lower costs associated with higher burnup levels in both the Weston and the Dippold an'J Wampler estimates result almost entirely from the need for fewer shipments as less spent fuel is discharged from reactors.

It is clear that there is a substantial difference in the magnitude of the costs and tha cost savings estimated by these two sources. This discrepancy exists even though both estimate sets were generated with versions of the TABLE 4.11. Estimated Back-End Transportation Costs, 1984-2020, by Spent-Fuel Age and Burnup Level (millions of 1985$, undiscounted)

Spent-Fuel Age (y)-

Burnup level 5 10 30 33 GWd/t:

Capital 132 121 87 Operation 701 645 462 TOTAL 832 766 549 60 GWC /t:

Capital 124 97 68

, Operation 659 519 365 TOTAL 783 616 433 Cost Difference:

Capital 8 28 18 Operation 42 126 98 TOTAL 50 150 116 Percent Decrease:

Capital 6.1 23.1 20.7 Operation 6.0 19.5 21.2 TOTAL 6.0 19.6 21.1 Source: Dippold and Wampler 1984, Appendix B, pp. 63 to 87. Costs expressed in 1983$

in original source were converted to 1985$ using GNP' implicit price deflator.

4-16 4

same computer program, WADCOM (Waste Disposal C_Ost Model) (Weston 1985, p. 4-4; Dippold and Wampler 1984, pp.19-21). Westoa used the multiple facilities version (WADCOM-MF), while Otppold and Warpler used a version consistent with their assumption of a single salt repository. The principal reason for the different cost figures in the two reports appears to be the assumed rates at which the repository or repositories are designed to receive spent fuel.

For the assumed single salt repository, Dippold and Wampler use a design receipt rate that is optimized for the reduced fuel stream from implementation of extended burnup. Thus, for a burnup level of 33 GWd/t, the design receipt rate is 3,000 t/y; for a burnup level of 50 GWd/t, the design receipt rate is 1,980 t/y; and for a t,urnup level of 60 GWd/t, the design receipt rate is 1,650 t/y (Dippold and Wampler 1984, Appendix B, pp. 63 to 87). In contrast, the Weston report assumes that the receipt rates at each of two repositories conform to the specifications of the Mission Plan (00E 1985b). Accordingly, design receipt rates for each of the two repositories rise in steps over time to a maximum of 3,000 t/y, making the inaximum total design receipt rate 6,000 t/y (Weston 1985, p. 1-4).

At a burnup level of 60 GWd/t, the design receipt rates assumed by Weston are from 2,2 to 3.6 times higher than those assumed by Dippold and Wampler.

This is roughly equivalent to the ratio of their respective estimates of trans-portation costs for a given burnup level, indicating that the difference in the cost estimates is largely due to the difference in assumed design receipt rates.

Another indication of the decline in transportation activity associated with increases in burnup levels is provided by another source, as shown in Table 4.12.

It is necessary to adjust the Weston and the Dippold and Wampler data from undiscounted sums of annual amounts to a 1985 present value calculated using _

a 10% discount rate. The procedure for accomplishing this is described in

, Appendix B, Section B.2. After adjustment, the 1985 present value estimates of the Weston transportation cost savings are $61.8 million, $70.8 million, and $72.4 million for salt and granite, tuff and granite, and basah; and granite repositories, respectively. Applying the same adjustment to the Dippold and Wampler data yields e::timates of the 1985 present value of transportation cost savings of $4.2 million, $12.7 million, and $9.8 million for spent fuel aged 5, 10, and 30 years, respectively. Again, these differences reflect different assumptions concerning repository design rates.

It is difficult to choose between these sets of estimates. First, it is unclear whether one or two repositories will be built. In addition, the repost-tory type (s) is not yet known, nor is the design receipt rate (s). Also, it should be noted that in order to reduce the inventory backlog of spent fuel, the total design receipt rate will have to exceed the total fuel discharge rate. For these reasons, we view the Weston and the Dippold and Wampler data as indicating the range of reasonable outcomes.

4-17

TABLE 4.12. Annualized Reference Reactor Year Tra'1sportation c Requirements for Spent Fuel, by Burnup Level Number of Shipments per Reactor Year Burnup Level (GWd/t) Truck Rail Barce(a) ,

60 10 5 33 57 10 5 35 50 9 5 40 44 8 4 45 4G 7 4 50 36 6 3 55 33 6 3 60 Source: Mauro et al. 1985, Table 4-5.

(a) No comercial cask currently exists for barges.

4.3.4 Repository Storsoe Extended burnup results in spent fuel with greater radioactivity and greater heat dissipation requirements per unit mass, but with fewer tonnes per kilowatt hour of electricity generated. As a result, though extended burnup reduces the number of assemblies, or quantity of-spent fuel, the areas of the repositories are still about the same size as that required for present burnup levels because the total decay heat dissipation requirements remain about the same based on the integrated fuel exposure or energy extraction (DOE 1985a).

Consequently, the analysis " indicates that the repository costs for all systems remain about the same as burnup increases if non integral [ sic] waste package loading is assumed" (DOE 1985a, p. 9).

A more optimistic conclusion is presented in the Dippold and Wampler (1984) report. Here, repository costs are found to decline continuously as bumup is increased from 33 GWd/t to 60 GWd/t and as spent feel age is increased from 5 to 30 years. These results are presented in Table 4.13.

Two qualifications must be applied to the costs shown in Table 4.13, however. The first is that these results are calculated on the assumption that all spent fuel will have the same burnup level. This assumption is not realistic because optimal burnup will differ among reactors, and effective burnup levelt, will differ among different fuel assemblies within the same fuel batch. The second qualification is that the dollar amounts shown are simple summations of a sequence of annual costs extending to the year 2020. Another set of repository cost estimates (from Weston 1985) is presented in Table 4.14.

Estimating the These costs are also totals of undiscounted annual amounts.

discounted present value of both these sets of costs is discussed in Appendix B, Section B.3.

Table 4.15 presents the resulting estimatas of the present value of the Dippold and Warpler repository costs. As expected, these figures imply that repository costs decline with spent-fuel age. This reduction occurs because 4-18 1 - - _ -

l TABLE 4.13. Total Undiscou.;ted Repository Cost bySpent-FuelAgeandBurnupLevel{a) l ('stilions of 1985$)

l Burnup level Spent-Fuel Age (y) l (GWd/t) 5 10 30 33 5,310 5,134 4,867 l 50 4,995 4,777 4,479 50 4,926 4,697 4,405 l Source: Dippold'and Wampler 1984, Appendix 8, pp. 63-87. Costs expressed in 1983$ in original source were converted to 1985$

using the GNP implicit price deflator.

(a) Total includes waste preparation and repository c >sts, but excludes at-reactor storage costs.

TABLE 4.14. Repository Costs, 1984-2020, by Burnup Level and Repository Type (millionsof1985$,undiscounted)

Burnup Level (GVd/t)

Repository Type 33 60 Salt 6,573 7,455 Granite 5,786 6,945 TOTAL 12,359 14,400 Tuff 6,719 8,649 Granite 5,786 6,945 TOTAL 12,505 15,594 Basalt 8,584 9,347 Granite 5,786 6,945

TOTAL 14,370 16,292 Source

Weston 1985, pp. 4-11, 4-13, 4-15.

Costs er. pressed-in 19841 in the original source were converted to 1985$ using the GNP implicit price deflator. Data are consistent with the DOE /EIA middle-growth case (see Table 1.1 and accompanying discussion).

l 4-19

3 i

i TABLE 4.15. TotalDiscountedRepositoryCosts(3)

by. Spent-Fuel' Age and Burnup Level (millionsof1985$)

Burnup Level Spent Fuel Age (y) -'

(GWd/t) 5 10 30 33 1081 1045 991

. 50 1017 972 912

60 1003 956 896 I

i Otfference (33-60) 78 89 94 l

Sourcet Derived from Dippold and Wampler 1984.

(a)_ Total includes waste preparation and i repository costs, but excludes at-reactor

,  !, storage costs.

j the shielding and heat dissipation _ requirements are reduced for fuel t'lat has.

been stored-at the reactor for longer periods. Note that repository costs

o decline with an increase in burnup level.

Table 4.16 shows the Weston cost estirates adjusted to_1985 present values.

These figures and the data derived from the Dippold and Wampler information imply different conclusions. Here the effects of increased burnup levels TABLE 4.16. Estimated Discounted Repository Costs, 1984-2020, by

Burnup Level and Repository Type (millions of 1985$)

Burrup Level (GWd/t) Difference-Repository Type 33 60 -(33-60)

Salt 1,568 1,778- -210-Granite 1,024 1,229 -205 TOTAL 2,592 3,007 --415 Tuff 1,602 2,063 -461 V Granite 1,024 1,229 -205

, TOTAL 2,626 3,292 -666

.I Basalt 2,047 2,229 -182 Granite 1,024 1,229 -205 7

TOTAL 3,071 3,458 -387' i

1 Source: Derived from Weston 1985.

4-20

1 I

I i

increase repository costs. .However, a closer review of the source reveals

! that no monotonic trend in costs occurs between the 33 and 60 GWd/t burnup

-l evel s . The burnup level at which repository costs arc at a minimum depends.

on the type of repository; but in all cases analyzed the minimum levr.! occurs at burnup levels above 33 GWd/t. Weston explains these results as follows i (Weston 1985, pp. 5-11 to S-13): l For repositories, tie primary effect of extended burnup is . ' eduction in the capacities of the waste packages because of the increased heat-generation rate per tonne of the fuel. When the heat-generation

rate per package is treated as a constraint, the number of spent-fuel assemblies that can be loaded in a waste package must be reduced to accommodate extended-burnup fuel. . .

l The underground area required for the disposal of the spent fuel

depends on the heat output of the total spent-fuel _ invento-v, rather

! than the total weight. Therefore, though the ' extended-burnup sce-narios show sinnificant reduction in the quantity of spent fuel, the 1 areal sizes o' > repositories will be equivalent to or greater l

than those rt,. for the base-burnup scenario [no extended burnup]

because of the i+.eated heat-generation rate per metric ton of uranium.

In addition, Weston suggests that-the reason for the divergence of his estimates from those of Dippold and War.pler is that they did not assume that-l all spent fuel would have the same burnup level (Weston 1985, p. S-13):

This range in burnups does not allow the optimization of repository designs for a single burnup level. Therefore, . realistic repository cost savings associated with extended burnup will.be less than those for a repository optimized for a single burnup level, such as those presented in a recent report by (Dippold and Wampler].

l 4.3.5 Back-End Total Costs Table 4.17 summarizes the for:Ochm estimates of back-end cost savings 4,

resulting from implementing exterded buri.90. Figures derivedt from both -the

Weston and the Dippold and Wampler data are shown. Because Dippold and Wampler do not provide an-estimate of at-reactor storage costs, the Weston-based figure 4 is inserted to allow calculation of total back-end cost savings.

s

The totals- for the two sources are quite different. The Weston data led

to an estimated net cost increase of from $267 to $548.million. In contrast, i the-estimates generated from the Dippold and Wampler figures show cost savings i

in the range of-$130.to a little over $150 million. Clearly, the Weston figures span a much larger range than the others. The differences between the two figures lie-primarily in the estimates of repository costs. As mentioned 1 above, Weston suggests that the main-reason forlthe difference in the estimates for this cost category is the different assumptions about the variation in burnup levels used in the calculations. In Dippold and Wampler all-spent fuel vas assumed to have the same burnup-level, allowing repository optimization

that was not possible under the Weston assumption of varied burnup levels.

! 4-21

, . - , ,,--.,-,,-..+,w.---.m- ,. -

TABLE 4.17. Summary of Estimated 1985 Discounted Back-End Cost Savings, 1985-2020, by Data Source and cost Category (millionsof1985$)

Weston (1985) Data Repository Type Dippold and Wampler Salt Tuff Basalt (1984) Data Cost and and and Spent-Fuel Age (y)

Activity Granite Granite Granite 5 10 30 At-Reactor Storage 47.6 47.6 47.6 [47.6](a) [47.6](a) [47.6](a)

Transportation 61.8 70.8 72.4 4.2 12.7 9.8 Repository -415.0 -666.0 -387.0 78.2 89.0 93.9, TOTAL -305.6 -547.6 -267.0 130.0 149.3 151.3 (a) Der;ved from Weston (1985) data for lack of Dippold and Wampler data for thir activity.

Another factor bearing on the two se;s of estimates is that the Weston i data are based on the assumption of two repositories each with a design receipt rate of 3,000 t/y, while the Dippold and Wampler data are based on a single repository with a-design receipt rate which declines with higher average burnup levels. As mentioned earlier, the current uncertair. ties with respect to repository specifications make it imprudent to reject one or the other set j of assumptions. For these reasons we interpret the estimates presented here

. as marking the range of likely outcomes, f

4.4 TOTAL FUEL CYCLE In a paper presented to the American Nuclear Society, Murphie and Lang (1982) estimated the net cost savings that could be expected from implementing C

extended burnup. These savings fall into three categories:

1. direct feel cost savings including those resulting from both the front end and the back end of the fuel cycle

2. reduced replacement powtr costs resulting from relaxed plant power maneuvering restrictions from the use of improved fuel [ pellet-cladding-interaction (PCI) resistant]

3. system cost savings from lengthened operating cycles facilitated by exte.nded burnup.

With respect to the first category, Hurphie and Lang (1982, p. 7-64) conclude the following:

I L

4-22

The direct benefit to the electric consumer of extended burnup in LWR's is a significant fuel cost savings. The projected 30-year levelizedfuelcyclecostsavingsof[00E1980b]wereescalatedto 1981 dollars and used in conjunction with the projected LWR capacities and implementation schedules. . . . This resulted in a projected cumulative savings of $8.0 billion for the United States . . . in constant 1981 dollars.

The second category refers to the fact that reactor operations are affected by potential PCI problems to the degree that capacity factor (see Glossary)

losses in the United States average about 3% for BWRs and 0.3% for PWRs (Murphie and Lang 1982, p. 7-64). Reductions la capacity lead to increased purchases of replacement power and, therefore, increased costs. Part of the research associated with extended fuel burnup includes the development of fuel with increased resistance to pellet-cladding interaction problems. Murphie and Lang (1982, p. 7-64) estimata that using this fuel will yield cost savings of

$2.5 billion (1981 dollars) through the year 2000.

The third category of cost savings relates to lengthened operating cycles.

Longer operating cycles inean that reactors are shut down for refueling less frequently. This leads to higner operating capacity factors, reduced refueling costs, and reduced costs for replacement power, The resulting savings are estimated at $1.4 billion in 1981 dollars (Murphie and Lang 1982, p. 7-65).

Table 4.18 reports these three savings categories in 1985 dollars.

3 TABLE 4.18. Estimated Savings Resulting from implementatio1 of Extended Burn by Category (oill:ons of 1985$)g)

Category Savinos Fuel Cycle 9.3 PCI-Resistant Fuel 2.9 l Extended Cycle _1. 6_

! TOTAL 13.8 l .

Source: Murphie and Lang 1982, pp. 7-64, 7-65.

(a) Costs expressed in 1981$

' in original source were converted to 1985$ using the GNP implicit price deflator.

These figures should be viewed with caution for two reasons. First, while the prospects of extended burnup add to the incentives to develop improved fuel, at least a portion of the cost savings in the second category could be realized independently of extended burnup. Second, the literature was unclear as to how or if the future cost savings were discounted. The total savings 4-23 l

5 l'

appear to be derived by summing the 'undiscounted_ annual amounts. Since the savings over time were not provided, the totals are'difficuit- to interpret..

l_ Because the annual savings will be. greater in later years and because-these savings accrue over a lengthy time pericd, the totals indicated may exceed the discounted sum by a considerable amount.- Discount. adjustment factors calculated elsewhere in this report for the sum of undiscounted annual amounts-

extending over a period of about 30 years range from around 0.18 to about i

0.30. If, for example, the appropriate discount adjustment factor is 0.25, fuel-cycle cost savings would be around $2.3 billion, and the total estimated by Murphie and Lang would be an adjusted $3.5 billion.

Combining cstimates of front-end and back-end cost savings developed here-gives a range of total discounted fuel-cycle cost savings of from $1.98 to

$2.68 billion. These amounts are strongly dominated by estimated savings in the front end of the fuel cycle. According to the estimates, chts segment of l

I the fuel cycle accounts for over 90% of tne savings.

4 4.5 SENSITIVITY ANALYSIS The foregoing estimates of _ the econonite. effects of implementing. extended burnup are based on a wide set of conditions. . Some conditions are particularly

important in determining the magnitude of the forecasted effects. This section

, presents.a discussion of the factors to which the economic effects of extended burnup are most sensitive.

Li j

i

' The front end of the fuel cycle accounts for the overwhelming bulk of the estimated reduction in total costs. However, these front-end' cost savings i

greatly depend on-assumed price levels,.particularly the uranium price. In fact, much of the ' original impetus for extended burnup came from the _ sharp rise in uranium prices in the 1970s. _ 8ecause extended burnup reduces the need for uranium, the dollar value of reduced uranium-requirements is directly

linked to the orice of uranium. This is stated succinctlyLby Delene: "The results . . . show a strong dependence of savings'on urantum price. The higher the price, the greater the savings" (Delene 1984, p. 21).

j The DOE notes that- the Franks and Geller results may be affected by uranium prices. Further, it is suggested that the price used by Franks and Geller may be high, resulting in an upward bias in the estimated benefits of extended burnup (DOE 1985a,-p. A.3):

An analysis of the Stoller [ Franks and Geller 1986] results -indicates that a-substantial fraction of the-predicted fuel cycle cost saving -

is due to savings in uranium costs. The uranium price projections were made to be representative of uranium purchased by_utilif't i

unde;*long-term _ contracts. The long-term contract prices used in-this study are considerably higher than the current spot' market price. Utilities make over 90 percent of- their uranium purchases

under long-term contracts. . . . [It] should be kept in mind that projected savings would become smaller or larger as uranium prices s

1 3

4-24 l

became lower or higher than the projected values. In the-Stoller

[ Franks and Geller] report.the sensitivity of fuel cycle' cost scvings with respect [to] unit cost values selected is unclear.

! Other pricc levels in the front end also have a bearing on cost effects.

As mentioned above, there are reasonably good prospects for a substantial decline in the cost of fuel enrichment. Delene, however, finds that "the sav4gs (appear] . . . to be relatively independent of enrichment price" i (Delene1984,p.21),.

t Aggregate savings may also potent: ally be sensitive to fuel-management methods. Work by Brown et al. (1986) indicates that extended burnup combined with the adoption of various optimization measures:in fuel management could produce improvements in ore utilization of up to 28.9% (Brown et al.1986, p.7-81).

At the back end of the fuel cycle, the most_important factor appears to

be the strict requirements for repository. design and operation._ If these requirements can be relaxed by aging the spent fuel, by' changing the minimum i repository specifications, or by technical improvement, a significant reduction in back-end costs would result. ,

' Finally, bechse this issue concerns-monetary cmounts extending well into the future, the magnitude of the economic effects is highly Wimdent on the l assorted time distribution of these amounts and the discourJ ne.

A sumary of-the cost savings resulting from implementing extended fue) burnup is presented in Toble 4.19.- Shown in the table are the cost savings

for different stages in the fuel cycle. The wide range given for some cost

rategories, most notably repository costs, reflects differences in, estimates

between primary sources. The total discounted cost savings for the period

! 1985 to 2020 resulting from implementing extended burnup is estimated to be l on the ordar of $2' billion. Almost all of this amount comes from savings in l-

-the front end of the cycle. Small cost savings art noted in at-reactor storage

and transport of spent fuel. Sources disagree as to the extent of effects of l extended burnup on repository costs. After adjustments, the Weston' data indi-l cate an increase in repository costs of around $390 to $670 million (See Table

4.16). Adjusted Dippold and Wampler data imply a modest decrease of around l $78 to $94 million-(See Table 4.15).- Because.the source data contain a number of-severe limitations, these estimates must be viewed with caution. Finally,

~

Murphie and Lang (1982), who only.pmvided a total estimate for the entire

fuel cycle, estimate total fuel-cycle costs at about $2.3 billion, which falls i -within the range given in Table 4.19.

l

[ 4-25 I _ ~- ._.1. . ___ 2 um r _ -. _. _ _

TABLE 4.19. Summary of Estimated Discounted Cost-Savings, 1985-2020, from Extended Fuel Burnup, by Category '

(millions =cf1985$)

Category Savinas

Front End of Fuel Cycle ~

Research and Development -21 -

t Fuel- Production and Fuel Management 2,552 Subtotal 2,531 ,

1 Back End of Fuel Cycle Development and Evaluation 0-At-Reactor Storage 48 l 72 i

Transportation 4 to

. Repository -666 to 94 Subtotal' -614 to 214 l

TOTAL-FUEL CYCLE. 1,917 to 2,745.-

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5.0 REFERENCE 5 Andrews, M. G., R. A. Mattie, and N. L. Sh piro, 1985. " Cutting PWR Cos.ts with Advanced In-Core Fuel Managenient Techniques." Nuclear Engineering' International 30(366):34-39.

Babcock and Wilcox. 1982. Extended Burnup Evaluation. Report BAW-10153(P),

Babcock and Wilcox, Lynchburg, Virginia.

Bailey, W. J., and M. S. Dunenfeld. 1986. Fuel Performance Annual Report for 1984. NUREG/CR-3950 (PNL-5210), Vol. 2 U.S Nuclear Regulatory Comission, Washington, D.C.

Baily, W. E., M. O. Marlowe, and R. A. Proebstle. 1985. " Trends in BWR Fuel Performance." In Proceedings of the American Nuclear Society Topical Meeting on Light Water Reactor Fuel Performance, pp.1-3 through 1-15. 00E/NE/34130-1, Vol.1, American Nuclear Society, Florida Section, Orlando, Florida.

Berkow, H. N., U.S. Nuclear Regulatory Commission. December 3,1985, letter to J. H. Taylor, Babcock and Wilcox Company.

Subject:

" Acceptance for Referencing of Licensing Topical Report BAW-10153P, Extended Burnup Evaluation." Available in NRC Public Document Room.

Brown, P. D. et al . 1986. " Impact of Extended Burnup on the BWR Fuel Cycle."

In Procedings of the American Nuclear Society Topical Meeting on Advances in Fuel Management, pp. 7-71 to 8-80. American Nuclear Society, Eastern Carolinas section, Pinehurst, North Carolina.

Butcher, E. J., U.S. Nuclear Regulatory Comission. October 10, 1985, letter to A. E. Lundvall, Jr., Baltimore Gas and Electric Co.

Subject:

" Safety Evaluation Report for Extended Burnup Operation of Combustion Engineering PWR Fuel, CENPD-269-P." Available in NRC Public Occument Room.

Charnley, J. S., General Electric Co. February 25, 1983, letter to F. J. Miraglia, U.S. Nuclear Regulatory Commission.

Subject:

" Proposed u Revision to GE Licensing Topical Report NEDE-24011-P-A." Available in NRC Public Document Room.

Charnley, J. S., General Electric Co. October 14, 1983, letter to M. S. Dunenfeld, U.S. Nuclear Regulatory Commission.

Subject:

"1984 Fuel Experience Report" (with attachment " Experience with BWR Fuel Through >

December 1984"). Available in NRC Public Document Room.

Combustion Engineering. 1984. Fxtended Burnup Ooeration of Combustion Engi-neering PWR Fuel. Report CENPD-269-P, Revision 1-P, Combustion Engineering, Inc., Windsor Connecticut.

5-1 l

Congressional:Infomation Bureau,. Inc. (CIB). 1982. " Industry Representatives Took Issue with Elements of the Department of Energy's_FY-1983 Budget for

- Nuclear Fission ' Programs as a House Science Subcomittee Continued its Consideration of-DOE's Funding Requests." In Atomic Eneroy Clearinahouse, 28(12):17. Congressional Infonnation Bureau, Inc., Washington, D.C.

Croff, A. G. 1980. 'ORIGEN2 'A Revised and Updated Version-of-the Oak Ridae Isotcpe Generation and Depletion Code. ORNL-5621, Oak Ridge National Labora-tory,. Oak Ridge, Tennessee.

L Delene,_J. G. 1984_(Draft).-. Fuel Cycle Cost Analysis of Extended Burnup F

Fuels. Oak Ridge National ~ Laboratory, Oak Ridge, Tennessee. Available in j NRC Public Document Room.

i Dippold, D. G., and J.-A. Wampler. 1984. Spent Fuel Burnup and Age:-

. Implications:for the Desion and Cost of a Waste Disposal System.-.

BMI/0NWI-561, Office of Nuclear Waste Isolation, Battelle Memorial Institute, Columbus, Ohio.

Electric Power Research Institute (EPRI). 1983. LWR Core MaterialsiPerformance

, Proqram:- Procress in 1981-1982. Report NP-3150 TR, Electric Power Research i Institute, Palo Alto, California.

! Electric Power Research Institute (EPRI).- .-1985.. LWR Core Materials Performance Program: Proaress in 1983-1984._ Report NP-4312-SR, Electric Power Research ji Institute, Palo Alto, California.

! Exxon Nuclear Company. - 1982. _, Qualification of Exxon Nuclear Fuel for Extended

Burnup. Report XN-NF-82-06,

Revision 1, Exxon Nuclear Company, Richland,

Washington.

L Franklin,'O. G.> 1982.- "The Schedule for- Extending FuelL Burnup." In Proceed- ,

j.ngs of the American Nuclear-Society Topical Meetina on LWR Extended Burnup -

Fuel Performance and: Utilization,.p. 8-10. DOE /NE/34087, _Vol. 2, American Nuclear Society, Virginia Section, Lynchburg,: Virginia.

Franks, W.;A., and L. Geller.- 1986. The._ Benefit of Extended'Burnup in Fuel Cycle Cost. . The S. M. Stoller Corporation, New York, New. York.

[

General Electric Company. 1982. Extended Burnup Evaluation Methodology.

Report NEDE-22148-P, General Electric Company, San Jose, California.

l Hulman,-L. G., U.S. Nuclear Regulatory Commission; December 20, 1982, memorandem to C.' H.= Berlinger, U.S. Nuclear Regulatory Commission.

,; Subjects -" Reload Reviews in DSI." Available in NRC Public Cocument Room.

a i _Lang, P.-M. 1982a. Contribution to " Panel Discussion of- Extended Burnup:.

How Far7: How Fast?" In Proceedings of the' American Nuclear Society Topical

4 Meeting 'n LWR Extended Burnup Fuel Performance and Utilization, p. 8-3.

DOE /NE/3dS7, Vol. 2, American-Nuclear Society,- Virginia Section, Lynchburg,

, [ Virginia.  :

I l c 5-2 l

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Langc P. M. 1982b. " Extending LWR Fuel Burnup." Nuclear Engineering Inter-national 27(329):14(TSSN0029-5507).

Lang, P. M. 1986. "Effect of 00E-Sponsored Development on LWR Fuel Cycles."

In Proceedings of the American Nuclear Society Topical Meeting on Advances in Fuel Management, pp. 7-61 to 7-69. American Nuclear Society, Eastern Carolinas Section, Pinehurst, North Carolina.

Mattsen, C. R., U.S. Nuclear Regulatory Commission. January 3, 1979, memorandum to Files.

Subject:

" Corrections to Data in ' Attachment 8' of Memo to Files Cated December 6,1978, Concerning Decay of Spent Fuel." Available in NRC Public Document aoom.

Matzie, R. A. 1981. Licensing Assessment of PWR Extended-Burnuo Fuel Cycles (Final Report). CEND-381, Combustion Engineering, Inc., Windsor, Connecticut.

Mauro, J. J., R. Eng, S. Marschke, and W. Chang. 1985. The Environmental Consequences of Higher Fuel Burn-up. P.epared for the National Environmental Studies Project, Atomic Industrial Forum, Inc., Bethesda, Maryland.

Murphie, W. E. , and P. M. Lang. 1982. " Projected National and World Bene-fits to the LWR Fuel Cycle from Extended Burnup." In Proceedings of the American Nuclear Society Topical Meeting on LWR Extended Burnup Fuel Per-formance and Utilization, pp. 7-61 through 7-69. 00E/NC/34087, Vol. 2, American duclear Society, Virginia Section, Lynchburg, Virginia.

Nuclear _ Engineering Internationel (NEI), 1985. " Putting the Heat on PWR Fuel," (editorial). 30(376):21-22.

Nuclear Fuel. 1984. "Ottinger Says Funding for Higher Burnups Would Cut Utility Waste Management Costs," (editorial). 9(4):5.

l Oztunali, 0. I., ana G. W. Roles. 1986. Update of Part 61 Impacts Analysis Methodology. NUREG/CR-4370, Vol. 1, U.S. Nuclear Regulatory Commission,

Washington, D.C.

Pati, S. R., and A. M. Garde. 1985. " Fission Gas Release from PWR Fuel Rods

at Extended Burnups." In Proceedings of the American Nuclear Society Topical Meeting on Light Water Reactor Fuel Performance, p. 4-19. 00E/NE/34130-1, Vol. 2, American Nuciear Society, Florida Section, Orlando, Florida.

Planell, J. R., M. Mason, and G. Guerra. 1983. Extended Fuel Burnup Demon-stration Program Topical Report - Transport Considerations for Transnuclear Casks. 00E/ET 34014-11, IN-E-4226, Transnuclear, Inc., White Plains, New York.

, Pyecha, T. D., G. M. Bain, W. A. McInteer, and C. M. Pham. 1985. " Waterside Cnrrosion of PWR Fuel Rods Through Burnups of 50,000 mwd /MTU." In _ Proceed-ings of American Nuclear Society Toolcal Meeting on Light Water Reactor Tuel Performance, pp. 3-17 through 3-35. 00E/NE/34130-1, Vol. 1, American Nuclear Society, Florida Section, Orlando, Florida.

1 i 5-3 l

Quigg, C., Pollution & Environmentsi Problems, Inc. March 6, 1980, letter to S. J. Chilk, U.S. Nuclear Regulatory Commission.

Subject:

" Petition for Rulemaking on Generic Impacts of High Burnup Nuclear Feel." Available in NRC Public Document Room.

Risher, D. H., et al. 1977. Safety Analysis for the Revised Feel Rod Internal Pressure Design Basis. WCAP-8963(Proprietary)andWCAP-8964(Non-Proprietary), Westinghouse Electric Corp., Pittsburgh, Pennsylvania.

Roberts, E. 1982. " Extended Burnup Considerations and Experience." In Proceedings of the American Nuclear Society Topical Meeting on LWR Extended Burnup Fuel Peri; marn and Utilization, pp. 1-1 througn 1-3. 00E/NE/34087, Vol.1, American helear Society, Virginia Section Lynchburg, Virginia.

Rogovin, M. (Dir). 1980. Three Mile Island - A Report to the Commissioners and to the Public. NUREG/CR-1250, Vol. II, Part 2, U.S. Nuclear Regulatory Comission, Washington, D.C.

Rosenstein R. G. et al. 1986. "24-Month Fuel Cycles." In Proceedings of the Amrican Nuclear Society Topical Meeting on Advances in Fuel Management.

American Nuclear Society, Eastern Carolinas Section, Pinehurst, North Carolina.

Rossi. C. E., U.S. Nuclear Regulatory Commission. July 18, 1986, letter to G. N. Ward, Exxon Nuclear Co.

Subject:

" Acceptance-for Referencing of Licensing Topical Report XN-NF-82-06(P), Rev.1, Qualification of Exxon Nuclear Fuel for Extended Burnup." Available in-NRC Public Document Room.

Roy F. Weston, Inc. (Weston). 1985. The -Effects of Extended Nuclear Fuel Burnuo on the Waste Management System. Roy F. Weston, Inc., Rockville.

Maryland.

Rebenstein, L. S., and M. Tokar. 1982. " Regulatory Perspective on Extended Burnup Fuel." In Proceedings of the American Nuclear Society Topical Meeting on LWR Extended Burnup Fuel Performance and Utilization, pp. 6-31 through 6-37. 00E/NE/34087, Vol. 2, American Nuclear Society, Virginia Section, Lynchburg, Virginia.

Silberberg, J. A., et al . 1986. Reassessment of the Technical Bases for Estimating Source Terms. NUREG-0956, U.S. Nuclear Regulatory Comission, Washington, D.C.

Stoltz, J. F., U.S. Nuclear Regulatory Commission. May 19, 1978, ietter to T. M. Anderson, Westinghouse Electric Corporation.

Subject:

" Safety Evaluation of WCAP-8963." Available in NRC Public Document Room.

Thomas, C. 0., U.S. Nuclear Regulatory Commission. March 1, 1985a, letter to J. S. Charnley, General Electric Co.

Subject:

" Acceptance for Referencing of Licensing Topical Report NEDE-24011-P, Amendment 7 to Revision 6, General Electric Standard Application for Reacto Fuel." Available in NRC Public Document Room.

5-4

i i

May 9,1985b. letter to Thomas, C. 0., U.S. Nuclear Regulatory Commission.

Subject:

" Acceptance for Referencing J. S. Charnley, General Electric Co.

of Licensing Topical Report NEDE-24011-P, Amendment 7'to Revision 6, General Electric Standard Application for Reactor Fuel: SER Page Changes for Clarification." Available in NRC Public Document Room.

August 13, 1985c, to Thomas, C. O., U.S. Nuclear Regulatory Comission. " Acceptance for Referencing J. S. Charnley, General Electric Co.

Subject:

of Licensing Topical Report NEDE-22148-P, Extended Burnup Evaluation Methodology." Available in NRC Public Document Room.

Thomas, C. O., U.S. Nuclear Regulatory Comission. October 11, 1985d, letter to E. P. Rahe, _ Westinghouse Electric Corp.

Subject:

"Acccptance for Referencing of Licensing Topical Report WCAP-10125P, Extended BurnLp Evaluation of Westinghouse Fuel." Available in NRC PuSTTc~uocument Room.

Turner, S. E., et al. 1982. BackgroundTTUerivation of ANS 5.4 Standard NUREG/CR-2507,)U.S. Nuclear Regulatory Fission-Product Release Mod _e_1.

Comission, Washington, D.C. sv /

1980a. Environmental Assessment, DOE Program V.S. Department of Energy (DOE). 00E/EA-0118, to Improve Uranium Utilization in Light Water Reactors.

U.S. Departmer' of Energy, Washington, D.C.

1980b. Nonproliferation Alternative Systems U.S. Department of Energy Assessment Program (NASAP)_.(DOE).Vol. IX of Reactor and Fuel Cycle Description.

00E/NE-0001/9, U.S. Department of Energy, Washington, D.C.

U.S. Department of Energy (DOE). 1985a. A Study of the Costs and Benefits of Extended Burnuo. Office of Nuclear Energy, Office of Civilian Radioactive waste Management, U.S. Department of Energy, Washington, D.C.

U.S. Department of Energy (DOE). 1985b. Mission Plan for the Civilian Radio-active Waste Management Program. 00E/RW-0005,-Vol. 1, II, and 111, Office of Civilian Radioactive Waste Management, U.S. Department of Energy, Washington, D.C.

U.S. Department of Energy / Energy Information Administration (DOE /EIA). 1985.

Commercial Nuclear Power: Prospects for the United States and the World.

00E/EIA-0438(85), U.S. Department of Energy, Washington, D.C.

U.S. Nuclear Regulatory Comission (NRC). 1972. Assumptions Used for Evalu-ating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling Storage Facility for Boiling and Pressurized Water Reactors. Regulatory Guide 1.25, U.S. Nuclear Regulatory Comission, Washington, D.C.

U.S. Nuclear Regulatory Comission (NRC). 1974a. Assumptions Used for Eval-uating the Potential Radiological Consequeaces of a Loss of Coolant Accident for Boiling Water Reactors. Regulatory Guide 1.3, Rev. 2, U.S. Nuclear Regulatory Comission, Washington, D.C.

5-5 i

i U.S.- Nuclear Regulatory Comission (NRC). 1974b. Assumptions Used for Evalu-atino the Potential Radiological Consequences of a Loss-of Coolant Accident '

' Tor Pressurized Water Reactors. Regulatory Guide 1.4, Rev. 2 U.S. Nuclear-Regulatory Comission, Washington, D.C.

i- U.S. Nuclear Regulatory Commission (NRC). 1981. Standard Review Plan for the j

Review of Safety An'alysis Reports for Nuclear Power Plants -- LWR- Edition.

NUREG/0800, Rev 2, U.S. Nuclear Regulatory Commission, Washingtcn, D.C.

U.S.OfficeofManagementandBudget(OMB). 1972. Discount Rates to be Used in Evaluating Time Dependent Costs and Benefits. Ci rcular A-94,- U.S. Of fice '

of Management and Budget,. Washington, D.C.

Viebrock, J. M., and R. A. Schreiber. 1983. Extended Fuel Burnup Demonstration

. Program - Nuclear Assurance Corporation Final Report. DOE /ET/34014-10, NAC-C-8327, Nuclear Assurance Corporation, Norcross, Georgia.

Westinghouse- Electric Corporation. ;1982. Extended Burnup Evaluation of Westinghouse Fuel. Report WCAP-1-10125, Westinghouse Electric Corporation, Pittsburgn, Pennsylvania.

4 P

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APPENDIX A FUEL ACTIVITY INVENTORY CALCULATIONAL PARAMETERS I

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l FUEL-ACTIVITY INVENTORY CALCULATIONAL PARAMETERS :1

! l l To detemine the activity inventory for-various burnup levels and times-I- after shutdown, the Oak Ridge Isotope-Generation and;0epletion Code,--0RIGEN2, i

was used (Croff 1980). From the resulting inventory, the changestin. activities for the various nuclides were. detemined as- the burnup _ level:was increased l from 33 to 60 GWd/t.: The . specific masses of the elements- and' uranium : isotopes-

_in grams per-tonne of fuel making;up a new fuel assembly are utilized in:the:

code. These parameters,'whteh are given in Table A.1, determine the resulting-

! activity levels in the fuel at variou:; exposure times. .Only thosel elements '

i having specific masses greater than 30 g/t are listed in the table. The i assumption.was made that the fuel was used in a typical pressprized water ,

reactor operating at a specific power level .of 37.5 MW/t.

i i

TABLE A.1. ORIGEN2 Input Specific' Masses (g/t) for Burnup Levels of 33 and 50_GWd/t.

t ,

f Element- 33 60 --

!' Carbon 129.46 129.46

! Nitrogen 55.88_ . -55.88-l  : Oxygen . 134,667 134,667 Aluminum 99.1 99.1-

' Silicon 138.03 1138.03 Phosphorous 307.52: 307.52

Titanium 108.12 - .108.12~

L Chromium 4,984.5- 4,984.5 4

Manganese 230.02- .230.02

Iron 9,442.6 9,442.6=

! Cobalt 272.3-. 72.3 l Nickel 9,572.3 .9,572.3-i Zirconium -218,341.8. ;218,341.8

Niobium- 710.8- 710.8 Molybdenus 394 - 394

Tin _

3,572 .3,572

Uranium-234 -

320 480

Uranium-235 35,450- 52,920- ,

l Uranium-238' 964,230 .946,600 p

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APPENDIX B COST ADJUSTMENT METHODOLOGY 4

1 4

- , , - , - - , . - , ~ , .

)

APPENDIX B COST ADJUSTMENT METHODOLOGY This appendix describes in detail the methods for adjusting data used in Chapter 4.0 to gauge the economic effects of implementing extended fuel burnup.

The most common adjustment was the conversion of the repori.2d cost data to a 1985 present value based on a 10% discount rate. Data from Weston (1985) and Dippold and Wampler (1984) are undiscounted sums of annual amounts. Cost figures from Franks and Geller (1986) are discounted at 7.813%. The adjustment for the front-end costs (Franks and Geller 1986) is described in Section B.1; the adjustment for the back-end costs (Weston 1985; Dippold and Wampler 1984) is described in Section B.2. In addition, adjustments to the repository storage cost data are described in Section B.3.

B.1 ADJUSTMENTS TO FRONT-END COST DATA As explained in Section 4.2.4 of this report, the Franks and Geller (1986) data on front-end costs are not completely consistent with the needs of this report. This section describes the steps taken to adjust the data for two of these problems. The first of these problems is that these data are based on a discount rate of 7.313% rather than the 10% rate assumed here. The second problem is that revisions in the DOE /EIA (1985) middle-growth orecasts are not reflected in the Franks and Geller data. In addition to these, two minor adjustments to the data are described in Section 4.2.4.

To adjust the Franks and Geller data to a 10% discount rate, the procedure outlined in Section 4.1 was used. First, a time profile for the Franks and Geller figures was generated. The method for doing this is based on the assump-tion that the cost savings are a positive function of both the total U.S.

nuclear generating capacity and the increase in average burnup levels.

This assumption is the basis for the following fomula, which is developed and used here to calculate an annual relative cost index:

' = Cj[(2/3)(E pj/B pj - 1) + (1/3)(Ebi/Bbi - 1)] (B.1) where I = index of relative cost savings i = year C = index of U.S. nuclear-generating capacity

~

E = average burnup level with extended burnup/research p = PWR B = base level average burnup level-(no extended burnup/no research) b = BWR.

! B-1

The accounts for the separate timef thepatterns relative aggie-of PWR a weights (2/3) and (1/3) are Franks and Geller's estimate o qate generating capacity of WRs and BWRs, respectively.

To estimate a series of indexes, Ij, that will have d Geller a time profile to the time profile of the cost savings imbedded within tothe d research theFranks an work, the ratio of the burnup level attained with DOE-funde One burnup level attaisied without the research was h calculated than one. The for each yea was subtracted from these ratios to give a base of zero t d rat U.S.ernuclear weighted sum of these ratios was scaled (multiplied) ilable. by forecas e generating capacity.are those used by Franks and Geller, rather k d than the la from the $550 million to establish a series of annu discount rate of 7.813% anr' a time profile corresponding to the calc indexes.

Af ter this sews of annual, undiscounted amounts was computed, d burnup/no y

' each annul 1 cost savings figure was adjusted to reflecti the extende extended burnup comparison, rather than the research/no This time, research com from whici Franks and Geller derived their figure.ing figures anothe however, the ratio of burnup levels was bam %without extended for U.S. nuclear generating capacity are used. set of ratios bwa This third ratio was Next, the second ratio was divided by the first. The result was an j tion then multiplied by the annual amount previously calcul and the change in basis of comparison. '

amounts were discounted at 10% to generate an estimate of Applying the above-described process gives an the 1985 pre value of the cost savings.

estimated cost savings of $2.6 billion.

B.2 ADJUSTMENTS TO SACK-END TRANSPORTATION COST DATA The data supplied by Weston (1985) and Dippold and Wampler (1984) back-end transportation costs and presented in Tah Thisbs 4.10 and 4.11 section 4.3.3 represent undiscounted totals for Thisthefollowsyearsthe 1984 ent of the 'eported cost savings using a 10% discount rate.

procedure cotlined in Section 4.1.

To celculate the present value, we needed to estimate the apprrnima time profile of transportation cost savings. i ies: the the eg.at savings in transportation are a function of two t me ser f planud total repository receipt Since the latter data rates are notand reported,the average fuel burnup we used insteadBecause transported fuel.

the average burnup level of fuel at the time of discharge.

B-2 i

level of the prior five years was used. For example, the average burnup level of fuel dischwged in the year 1995 was used as an estimate of the average burnup level of fuel shipped in the year ;000.

1 From these data we produced a time profile of the transportation cost savings. First, for each year, the average burnup level under the extended burnup scenario was divided by the average burnup level under the base (nonnal) burnup scenario. Then each of these ratios was multiplied by the planned repository receipt rate for that year. The result for any year is the reposi-tory receipt rate for an equivalent amourt of nomal-burnup fuel, where equival-er 'y is measured in tems of the quantity of electricity produced by the fuel.

The difference between the equivalent repository receipt rate and the planned repository receipt rate is a measure of the quantity of fuel that does ..ot have to be transported.

The equivalent repository receipt rates were then used to construct an index. First, the rates were normalized so that they sum to 1.0. If the rate in each year,1, is denoted by wi, then a discount adjustment factor, F, is derived from the folicwing fonaula:

F= Iwj/(1.1)I (B.2) 1 The value obtained for F is A 0845. The cost savings do not begin until th.

year 1998, when the first repository is scheduled to begin, and they increase each year until 2010.

B.3 ADJUSTMENTS TO REPOSITORY 37) RAGE COST DATA In Section 4.3.4 repository cost data from Weston (1985) and from Ofppold and Wampler (1984) are presented. Since these figures represent undiscounted l

totals, however, it was necessary to calculate an equivalent 1985 present value based on a 10% discount rate. The steps for doing this follow the method Two discount adjust-l outlined in Section 4.1 and are described in detail here.

ment factors were generated for the Weston cost estimates, one for each of the two repositories. A single discount adjustment factor was calculated for the Dippold aad Wampler cost figures.

Again, it was necessary to estimate a time profile for repository costs.

This was accomplished in the following manner u:ing the Weston figures, the capital costs for the first repository were assumed to be levelized over the period 1987 to 1998, which is before the first repository is schedule' to l

begin operating. The operations costs were levelized over the period 1998 to 2020, and the costs of repository closure were assigned to the year 2020.

This schedule of costs yields a discount adjustment factor of 0.2385 for the first repository. This factor was applied to tuff and basalt repository types as well as salt, because their cost schedules do not differ greatly.

l B-3 l

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A similar allocation was used for costs of the second (granite) repository.

Capital costs were levelfimi over the period 1990 through 2005; operations costs l

' over the period 2005 to 2020; and closure costs were assigned to the year

2020. The resulting discoun? The adjustment factor for costs associated with the discount adjustment factor for the combined second repository is 0.1770.

costs of the first and second repositories is 0.2036. This factor is applied to the Dippold and Wagler cost figures.

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1 NUREG/CR-5009 PNL-6258 DISTRIBUTION No. of No. of copies Copies OFFSITE ONSITE 10 M. R. Fleishman 3') Pacific Northwest Laboratory

~~

Office of Nuclear Regulatory Research U.S. Nuclear Regulatory W.

D. A.J. Baker Bailey (10)

Commission C. E. Beyer Washington, D.C. 20555 F. C. Bold V. L. Brouns J. B. Brown, Jr.

D. W. Dragnich M. D. Freshley M. J. Graham J. M. Hales P. C. Hays R. J. Hoe M. F. Mullen T. L. Page R. G. Scnrer.khise R. J. Sorenson J. J. Tawil Publishing Coordination (2)

Technical R? port Files (5)

Distr-1

. .-. ___ _ . . - _ . . -- . _ _ _ . . _ _ _ = _ _ - - . _ - - . . . - . . -

~

.g,.otou. ., .m a..... u oa, m -

...~-..c.....-,w..-4, 3".'a'

u. ,m.xv .o., o., e..n..... BIB UO G R APHIC D ATA S H E ETNUREG/CR.5009 PNL.6258

. i.,a . m . a , a . . . . a.;- '

ASSESSMENT OF THE USE OF EXTENDED BURNUP FUEL IN LIGHT WATER POWER REACTORS -

. o.n ...o., cc o no

o. ....  ;

~ (

i ....o.... January 1988 D.A. Baker F.C. Bold ' o*" " maa *

{

W.J. Bailey J.J. Tawil **'"

l C.E. Dever -

February 1988

, . . . . o. . . % o. a . . . , . . . ,. . . .. . . . . .s ,% . o o. . u . ...., < . c. ., ...c,,.w....,%...

Pacific Northwest Laboratory ,,,,,,,,,,,,,,,,,,

P.O. Box 999 Richland. WA 99352 '

. FIN 82894 o~.o-...m.................,....n,,.<.c, ...i,..o....o.,

Division of Regulatory Applications Office of Nuclear Regulatory Research U.S. Nuciear Regulatory Comission Technical Reoort

. n..oo ca . . . . . ,,, ,

Washington. DC 20555

.,,.u........,n

........c .

This study has been conducted by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Comission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch averige burnup levels of 33 GWd/t uranium be

, increased to above 50 GWd/t. The environmental effects of extending fuel burnup during nonnal operations and during accident events and the economic effects of c st changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for t :e environmental and economic assessments.

Environmentally. this burnup increase would have no significant impact over that of nonnal burnup. Economically. the increased burnup would have favorable effects.

consisting primarily of a reduction in 1) total fuel requirements. 2) rcactor downtime fer fuel replacement. 3) the number of fuel shipments to and from reactor sites. and 4) repository storage requirements.

.. o m ...... 6. .... .. m a.o ,onc... o..

light water reactor "*"*

extended fuel burnup fuel cycle Unlimited fue1 assembly '. ac .'m.a+cino?

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Unelassified a..-.,

Unclassified

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ENCLOSURE 3 TO TXX-92468 Letter from Ted C. Feigenbaum of New Hampshire Yankee to the NRC logged NYN-91049, dated March 18, 1991 L

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. - . , .. , , . , . . , . z,, :O .- , , . - . . - . . - - - . . , , . , -

- . - . - _ _ . - - - - - - _ _ . - - _ . ~ - - . - - _ - . . - . - - - - . -

I i j. .

! Now v Hampshire Ted C. Folgenbevm i Ptnident and chtel eseeve oliker i

NYN. 91049 March 18,1991 i

i United States Nuclear Regulatory Commission l

Washington, D.C. 2055$ J Attention: Docurnent Control Desk 4

~

Reference:

Facility Operating License No. NPF.86, Docket No. 50 443 Subjectt Request for License Amendmentt increased Enrichment of Reload Fuel Assemblies - ,

Ocotlemen

Pursuant to 10CFR$0.90, New Hampshire Yankee (NHY) hereby proposes to amend l

the Seabrook Station Operating License (Facility. Operating License NPF.86) by incorporating i

the proposed changes provided herein as Enclosure 1, into the Seabrcok Station Technical

{

Specifications. This request for 1. cense amendment is submitted in support of NHY's plans l

to extend the length of its operating cycle to eighteen months commencinagwith Cycle 3.

i The current Technical' Specification requirements pertaining to relona fuel assembly l

enrichment will not accommodate such an extension in the length of the operating cycle.

L The preposed changes involve an increase in the maximum carichment of reload fuel l assemblies authorized by Technical Specification 5.3.1 (Fuel Assemblies) to 5.0 weight percent

[ Uranium 235 from the cut'ent 3.5 weight percent Uranium ,135. Additionally, the proposed p

changes involve the addition of two new Technical Specifications, 3/4.9.13 (Spent Fuel i L Assembly.S.orage) and 3/4.9.14 (New Fuel Assembly Storage) and their associated' bases.

Technical Specification 3/4.9.13 and 3/4.9.14 specify the Limiting Conditions For Operation -

i and Surveillance Requirements associated with tbc storage of fuel assemblies in-the Spent i Fuel; Pool (existing fuei stcrage racks) and in the New Fuel Storage Vault (existing fuel i

storage racks). - The critical'y analyses for the Spent Fuel Fool' and New Fuel Storage Vault which underlie the proposed Limiting Cosditions for . Operation are - enclosed herein as

' Enclosure 2. New Hampshire Yankeo will develop procedures to implement the new Limiting ' .

< -Conditions for Operatic:1 and Surveillance Requirernents for the' Spent Fuel Fool and New Fuel Storage _ Vault' prior to receipt of reload fuel-assemblies with enrichments greater than-i 3.5. weight percent Uranium 235.

j i _,

O "4\9 i U\0\ 3&

l New Hampshire Yonke, Diston of Public Service Company of Nw Hampshire  !

+ . P.O. Box 300

• Soobrook, NH 03874

• 1elephon J603{ 474 9521 L _ -

4

1--603 4* l 93211 4413L -PAGE.002

AUGj!2;'.911 Ja15>- '

United States Nuclear P.egulatory Commission March 18,1991 Page two i

Attentiont Document Control Desk New Hampshire Yankee has reviewed the proposed changes utilliing the criteria specified in 10CFRSO.92 and has determined that the proposed charps do not involve a Significant Hazards Consideration pursuant thereto as discussed below:

1. The proposed changes do not involve a significant increase in the probability or j

consequences of an accident previously evaluated. There is no lucrease in the probability of a fuel assembly drop accident in the Spent Fuel Pool since the mass 1

of the fuel assembly does not increase when the fuel enrichment is increased. There is not a significant increase in the consequences of a fuel assembly drop accident in l

the Spent Fuel Pool since the fission product inventories in the fuel assemblics do l

i not change significantly due to an increase in the fuel enrichment. The existing FSAR analyses for the fuel assembly drop accident it.dicate that radiological consequences are well within 10CFR100 limits. Thl conclusion rernalns valid at the lucreased fuel assembly enrichment. There is no increase in the probabl!!ty or consequences of misplacing fuel assemblies in the Spent Fuel Pool because fuel assembly placement will be procedurally controlled and surveilled pursuant to the proposed Technical Specifications and criticality analyses demonstrate that the pool will remain subceltical assuming misplacetsent does occur. There is no increase la the probability or consequences of introducing optimum moderation conditions in the New Fuel Storage

• Vault as a result of a; increase in fuel enrichment. The New Fuel Storage Vault has been analyzed under a range of moderation conditions from fully flooded to optimum

' moti eration at the increased fuel enrichttent. These analyses demonstrate that the New Fuel Storage Vault remains suberitleal under tiese moderation conditions.

The proposed changes do not create the possibility of a new or different kind of 2.

accident t'aen previously evaluated Spent fuel handling accidents are not new or different types of accidents in that they are already analyted in the FSAR. Criticality' l

accidents in the New Fuel Storage Vault or Spent Fuel Fool ne not new or different types of accidents in that they are already analyzed in the FSAR for fuel enrichments up to 3.5 weight percent Uranium 235. Additional criticality analyses have been i performed for fuel enrichments up to 5.0 weight percent Uranium 235,

3. The proposed changes do not involve a significant reduction in a margin sf safety.

Criticality analyses have been performed which d monstrate that the New Fuel Storage Vault will remain suberitical under a range of moderatiot conditions from fully flooded to optimum moderation. Criticality analyses haw been performed which demonstrate that the Spent Fuct Pool will be at least five percent subcritical under a fuel assembly misplacement accident with soluble boron (2000 parts per million) present la the pool and v.ill remain subcritical with no soluble boren present.

New Hampshire Yankee requests approial of the Technical Specification changes proposed herein 'oy September 1,1991, as this is the apprnximate time at which NHY will be specifying the reload fuel for Cycle 3.

~ - ,

1 603 474 9521 4413 PAGE.003 AUG 12 '91 7 16

March 18,1991 UElted States Nuclear R:gulatory Commission Page three Attention: Document Control Desk Should you have any questions regarding this request or should you wish to have l

NHY representatives discuss the enclosed analyses at a tneeting, please coctact Mr. Terry L. Harpster at (603) 474 9521, extension 2765.

Very truly yours,

&V '

odn

' Ted C. Feig abaum Enclosures

, TCF:ALLhsl cc: Mr. Thomas T. Martin l I Reglonal Administrator I United States Nuclear Regulatory Commission l Region I 475 Allendale Road

' King of Prussia, PA 19406 i

! Mr. Gordon E. Edison, Sr. Project Mana6er Project Directorate 13 i Division of Reactor Projects U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. Noel Dudley NRC Senior Resident Inspector P.O. Box 1149

Seabrook, NH 03874 5

w t

4

.1 603.474 9521 4413 PAGE.004

' AUG 12 '91 , - _ - -

7 16 - . -,

-4

+

t i ,

New Hampshire Yankee March 18,1991 i

. i i

j ENCf_O$URE 1 TO NYN 91040 1 4

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l 1 603 474 9521'~4413- PAGE.005  !

7:16-Abd 12 . ' _ _91 -. ,_ , _ . _ _ . _ _ _ _ _ , _, _ _ , , _ _ ,

s fNDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION EGE 3/4 g-4 3/4.9.4 CONTAINMENT BUIuDING PENETRATIONS........................

l 3/4.9.5 COMMUNICATICN5........................................... 3/4 3/4 95 96 3/4.9.6 REFUELING MACHINE........................................

3/4 9-7

, 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREA 5. . . . . . . . . . . . . . . . . . *

! 3/4,9.8 kESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1......................................... 3/4 9 S low Water leve1..........................................

3/4 9 9 1

CONTAINMENT PURGE AND EXHAUST ? SOLATION SiSTEM...........

3/4 9-10 3/4.9.9 3/4 9-11 3/4.9.10 WAT E R I. EVE L - R EAC TO R VE 5 S E L . . . . . . . . . . . . . . . . . . . . . . . 3/4 . . . 9-12 t

3 /4. 9.11 W AT ER LEVEL - STO R AGE POO L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9 13 3/4.9 12 FUEL $TORAGE BUILDING EMERGENCY AIR CLEAN!N3 FYSTEM......

i 3/k.10 SPECIAL TEST EXCEPTIONS 3/4 10-1

! 3/4.10.1 SHUT 00WN MARGIN..........................................

3/4 10-2

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION t.lHITS...

3/4 10-3 4 3/4.10.3 PHYSICS TESTS............................................ 3/4 10 ) s/4.10.4 REACTOR COOLANT L00PS.................................... 3/4 10 5 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN.................... 3/4 10-6

C 3/4.10.6 REACT OR COO LANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

+ 1*.T .~

3/4.11 RADI0 ACTIVE EFFLUENTS

, 3/4 n.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 3/4 11-2

00se.........................-............................

! Liquid Radwaste Treatment System. ....................... 3/4 11-3

L i qui d Hol dup Tan ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-4 l

l 3/4.11.2 GASECUS EFFLUENT 5 00se Rate................................................

3/4 11-5 Dese - Noble Gases.......................................

3/4 11 6

! Dose - locine-131, Iodine-133, Tritium, and Radioactive

' Material in , Particulate Form............,................ 3/4 11-7 3/4 11-8 Gaseous Radwaste Treatment System.............. .........

Expl osive Gas Mixture - 5ys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-9 WA5TES................................, 3/4 1140 3/4.11.3 SOLIO RA010 ACTIVE 3/4 11-12 3/4.11.4 TOTAL 005E...............................................

i 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4 12 1 c.. . .

3/4.12.1 MOND0 RING PROGRAM.............. Z ............ ....... .

3 / Y 9 - Ma 5/q . 9.13 $f'fMT fufi A$$CPdt Y $rbMGC <s . rNtrzn. KNAlcnMNr"

~

f tGuRE' 3.9-I p'un. An u SW suAN&pcA stwr Fua AssCe6t Y 3rausc $l49'/7 ix . '

SEABROCK - UNIT 1 q [-f Sl4 ' *IY spf. 9. t+ Mv fah. AS5 f" 00' ST*Ma5 enas.cos aua is '91 7:17 MDMW D N t-sea'474 ssai 44 a

. -- -..- -. _ - - - .-_.- ---.- - . _ - _ - . _ - . =

fNDEX r

i

%,.s BASES Y PAGE

SECTION i.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM............ B 3/4 9 2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VE5SEL and 2

STORAGE P00L............................................ S 3/4 9-2 1

3/4.9.12 FUEL STORAGE BUILO!NG EMERGENCY AIR CLEAN!NG SYSTEM....... B 3/4 9-2

-+

_3/4.10 SPECIAL TEST DCEPTIONS

! 3/4.10.1 SMUT 00WN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS. . . . B 3/4 10-1 l B 3/4 10 1

3/4.10.3 PHYSICS TEST 5.............................................

J 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10 1 3/4.10. 5 POSITION INDICATION SYSTEM - SHUT 00WN. . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.6 REACTOR COOLANT L00P5..................................... B 3/4 10 1

, 3/4.11 RAD 10ACTEEjFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......................................... B ?/4 11 1 3/4.11.2 GASEOUS EFFLUENT 5........................................ B 3/4 11-2 3/4.11.3 SOLID RADICACTIVE WA5TES................................. B 3/4 11 5 B 3/4 11-5 i 3/4.11.4 TOTAL 005E...............................................

1

)NA3 4

'4.12 RADIOL 0filCAL ENVIRONMENTAL MONITORIN_G 3/4.12.1 MONITORING PR0 GRAM....................................... B 3/4 12-1 3/4.12.2 LANO USE CENSUS......... ................................ B 3/4 12-1 3/4.12.3 INTERLABORATORt COMPARI5ON PR0G MM....................... S 3/4 12-2 S.0 OESIGN FEATURES __

5.1 SITE I

5.1.1 EXCLUSION AREA.............................................. 5-1 5.1.2 LOW POPULATION 10NE.....*.................................... 51 1 5 1,3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS ANO LIQUID EFFLUENT 5. . . . . . . . . . . . . . . . . . . . 5-1 FIGURE S.1-1 5ITE AND EXCLUSION AREA B0VNDARY..................... 53 FIGURE S.1 2 LOW POPULATION 20NE..................................

35

FIGURE 5.1-3 LIQUID EFFLUENT DISCHARGE LOCATION................... S-7 l i

5. 2 CONTAINMENT CONFICURAT10H............................................... 5-1 5.2.1 5.2.2 OE SIGN P RESSURE AND TEMPERATURE. . . . .,. . . . . . . . . . . . . . . . . . . . . . . . 5-9 ,

i 1

S en r e m. nswei.r muaa- 6 2).+ g-3 h~ S/4. 9. n 3/4 9. it ^<w FkC. 435 V8W WAaC g 3lq 9 5

~

" SEA 3R00K - UNIT 1 .

xii i

! 603 474 9521 4413 PAGE.007

,AUG 12 '91

- 7:19.

l Add the Following Technical Specification to Section 3/4.9 l REFUELING OPERATIONS _

3/4,9.13 SPENT FUEL ASSEMBLY STORAGE

.l.lMITING CONDITION FOR OPERATION 3.9.13 Fuel assemklies stored in the spent fuel pool shall be placed in the spent fuel storage racks according to the critena shown In Figure 3.91.

APPL 1,CABlLITY: Whenever fuelis in the spent fuel pool.

ACTION:

a. With the requirements of the above specification not satisfied, suspend all other fuel movement within the spent fuel pool and move the non complying fuel assemblies to allowable locations in the spent fuel peci in accordance with Figure 3.91.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.13.1 The bumup of eac'= f 1 assembly to be stored in the spent fuel pool shall be determined from it'c measured curnup history pr!ct to stora9e in the spent fuel pool. A complete record of each assembly shall be maintained as long as that fuel assembly is retained on site.

4.9.13.2 After fuel assembly (iss) movement into or within the spant fuel pool, the position of the fuel assembly (les) that was (were) moved shall be checked and Independently verified to be in acccrdance with the criteria ln Ffgure 3.91.

l m

1 E03 474 95al 4413 DAGE.co9

.AuG 1291 7[19- . _ _ _ _ , _ .

6 b

25 I l l l ) ,,

i c u es .ior.o .nyoner. /

t (nust rolto stored next to 3

.' 3 must to stored next to 1 er empty locations

)

20 :

Twei 15 7'

G .

h-9., ,

4 1- --

/

/

10

/

e o I

Typet

, / I h

~

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w.31 0= ,- 3 o , o , , i .,

2.5 2.75 3.0 3.25 3,5 3.75 4.0 4.25 4.5 4.75 5.0 ItstialErvichmer.t(we IJ235)

Figurs' 3.9-1 Fuel Assembly Burnup vs, Initial Enrich::.ent For Spent Fuel Assembly-35.orage ,

1 603.474 9521_4413 PAGE.009 AUG 12 '91 7:19 - - - - , - _ . -_ _ - . , _ . _ . , . _ . .

u .

i 1

i Add the Following Technical Specification to Section 3/4.9

! REFUEUtG OPERATION.S 3/4.9.14 NEW FUEL ASSEMBLY STORAGE I

LIM: JING CONDITION FOR OPERATION 3.9.14 The new fuel storage vault may be maintained with a fullloading of 90 l assemblies with fuel ennchment up to 3.675 w/o "U. The loading must be reduced 8 <

to 81 assemblies for enrichments from 3.675 to 5.0 w/o "'u by limiting the fuel l assembly placemert in the central column of the New Fuel Storage Vault to every other location.

a

} &ntCAalLITY; Whenever fuelis .n the Neu Fuel Storage Vault.

, ACTION:

a. With the requirements of the above specification not satisfiad, suspend all other fuel movement within the New Fuel Storage Vault and move the non-complying fuel assemb!!es to alkwable locations in tne New Fuel Storage Vault in accordance with

- the requirements of the above Specification.

b. The provisions of Specification 3.0.3 are not applicable.

SURVE1LLANCE REQUIREMENTS i ' 4.9.14.1 After fuel assembly (los) movement into or within the New Fuol Storage Vault.

the position ci the new fuel assembly (les) that was (were) moved shall be checked and independently verified to be in acccrdance with the requirements of the above specification.

l 2

v 2

~

1 503 474 9521 4413 PAGE.010

, AUG _.12 ' '_91_ 7 j l 3 ,, --- _ _ _ _ __

\

Add Bases on Page B 3/4 9>3 3/4.913 Scent Fuel Assembiv Sterage Restrictions on placement of fuel assemblies of certain enrichments within the spent fuel poolis dictated by Figure 3.91. These restricticns ensure that the K, of '

the spent fuel pool will always remain less than 0.95 assuming the pool to be flooded with unborated water. The restrictions delineated in Figure 3.91 and the action statement are consistent with the criticality safety analysis periormed for the spent fuel pool as documented in the FSAR.

3/4.9.14 New Fu_el Assembly Stef; Log Restrictions on placement of fuel assembi:es of consin enrichments within the New Fuel Storage Arun is dictated Specification 3/4.9.14. These restrictions ensure that the k,, of the New Fuel Storage Area will always remain less than 0.05 assuming the area to be flooded with unborated water. In addition, these restrictions ensure that the k,d of the New Fuel Storage Area will always remain less than 0.98 when aqucous foam moderation le assumed. The restrictions delineated in Specification 3/4.9.14 and the action statement are consistent with tr.e criticality safety analysis performed for the spent fuel pool as documented in the FSAR.

k 1

1 AUG _ __11 ' 9,1 7:a0 ,

1 603 474 95al 4413 PAGE.oll

DESfGN FEATURES DESIGN PRESSURE AFD TEMPERATURE

,h+

@ 5.2.2 The containment builfiing is designed and shall be maintained for a maximum internal pressure of 52.0 psig and a temperature. of 296*F.

S.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly con-taining 264 fuel rods clad with Zircaloy-4. Each fuel red shall have a nominal active fusti length of 144 inches. The initial core loading shall have a maxistem enrichmentof3.15weightpercentU-235. Reload fuel shall be similar in phy-sical design to the initial core loading and shall have a maxieum enricament of

+r, weight percent U-235.

30 CONTROL R00 ASSEMBLIES 5.3.2 The core shall-containhfull-length control roc assent .ies. The full.-

length control rod asssmblies snall contain a nominal 142 inchts of absorber

material. The nominal values of absorber material shall be 80% si'.ver, 15% in-

. dium, and 5% cadmium. All control rods shall be. clad with stainless steel tubing.

- b.4 REACTOR COOLANT SYSTEM OE$1GN PRESSURE AND TEMPERATURE 5.t.1 Tk Reactor Coolant System is designed and shall be maintained:

a. In accordance v'th the Code requirements specified in Section 5.2 of the F5AR, with allewance for nomal degradation pursuant to the

applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and -

c. For a temperature of 550*F, except for the pressurizer wnich is 630*F.

1 VOLUME t

5.4.2- The total water and steam volume of the Reactor Coolant System is j 12,255 cueic feet at a ncminal T avg of SSB.5'F.

4 5.5 METEOROLOGICAL T0kER LOCATION 5.5.1 The meteorological tower shall be located as shown on Fipre 5.1-1.

' .+[

SEAP. ROOK - UHli 1 ,

59 g 9

L 1 603 474 9521 4413 PAGE.012-AUG l2 '91,- -7:21 _ , _

~

DESIGN rEATURES .

j 4

! 5.6 FUEL STORAGE E4.

2/ CRITICAt.ITY 3

5. 6.1.1 ina spent fuel storage racks are designed and shall be maintained a

with:

to les than or equal to .95when[codedwi l gg. a. A k,ff quivale A ---+ unbo, tad wat r', whic includes a conser tive allo nce of i f.% 3 9

for/uncertajdtiesasdescribe in Sectipn 4.3 of t e FSAR, a(d i b. A nominal 10.35 inch center-to-center distance between fuel j assemblies placed in the storage racks, ygg 5.6 .2 The .,ff for ew fuel or the rst core cadings4reddryi the j d~ spp[ntfuel torage eks sha) not e eed 0.98 .en aque s foam me ration i

afsumed. / t .

j ORAINAGE-5.6.2 The spent fuel storage pool is designed and shall be maintained to '

prevent inadvertent draining of the pool balow elevation 14 feet 6 inches.

I .

C CAPACITY

\

5.6.3 The spent fue1~ storage pool is designed and shall be maintained with a .

)

[sp storage capacity limited to no more tnan 1236 fuel assemblies. ,

5.7 COMPONENT CYCL.IC OR TRANstENT t.IMIT

! ' 7.1 The components identified in Table 5.7-1 are designed and shall be

! maintained within the cyclic or transient limits of Table 5.7-1. .

f I

i i

i

! ~

4 1

T b)

s. :

,_ , ,. f;UG.j 2 '9.11[7:21 _ _._. . _ . - . _ . . _ . 11603 474 9521-4413 PAGE.013- .

insert A 4

a. A k,, equivalent to less than or equ?! to 0.95 When flooded with unborated water, which includes margin for uncertalnty in calcu!ational methods and mechanical tolerances with a 95% probability at a 95% confidence level, i

! 1

Insert B ,

5.6.1.2 The new fuel storage racks are designed and shall be maintained with:

. i

a. A k,,, equivalent to less than er equal to 0.95 when ficoded with unkorated water, j which includes margin for uncenalnty in calculational methods and menanical 1 tolerances with a 95% probability at a 95% confidence level.

l i

b. A K,3 equivalent to less than or equal to 0.98 when aqueous foam modaration is l

assumed, wh?ch includes margin for uncenalnty in calculational mether.:s an 1

- mechanical tolerances with a 95% probability at a 95% confidence level, i 1

c. A nominal 21 inch center to center distance between fuel assemblies piaced in the i storage racks.

• t l

l i

TOTAL P.14 h ' ' 50 -

' A _._ . . . . . . _ , . _ , _!.[6.03 474 95al'4413 PAGE.014