ML20003D510

Nonproprietary Version of Environ Impact of Extended Burnup Fuel Cycles in Calvert Cliffs Units 1 & 2.

ML20003D510
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Site:	Calvert Cliffs
Issue date:	03/31/1981
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
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CEN-122(B)NP, NUDOCS 8103270494
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e LEGAL NOTICE

- This report was prepared as an account of work sponsored by Combustion Engineering, Inc. Neither Combustion Engineering nor any person acting on its behalf:

A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with 1 spect to the accuracy, completeness, or usefullness of the information contained in this report, or tha the use of any information, apparatus, .

method, or process oisclosed in this report may not infringe

' priwtely owned -rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report.

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- Environmental Impact of Extended Burnup Fuel Cycles in Calvert Cliffs Units 1 and 2 1.0 Introduction Baltimore Gas and Electric (BGLE) is planning to extend both the cycle length ar.d the discharge fuel burnup in Calvert Cliffs Units 1 and 2.

Since the original Calvert Cliffs Environmental Report was based on lower fuel burnups, this report has been prepared to address generically those environmental impacts which are affected by the eighteen-month extended discharge burnup fuel cycle. Detailed safety analysis results are not presented but will be addressed as required on a cycle-by-cycle basis in

'the reload submittals. The conclusion reached herein is that there are no significant or substantive increases in adverse environmental impacts due to the planned actions; in f act, there are several areas where the environmental impact of the Calvert Cliffs fuel cycle is reduced.

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The planned action is shown schematically in Figure I which is the projected operation schedule for Calvert Cliffs Units 1 and 2 for the next several years. Generally, through 1980, the units have been operat'ng on annual reload cycles with approximately a three-batch fuel management scheme. Beginning in late 1980 for Unit I and early 1981 for

. Unit 2, a transition is planned to an 18-month fuel cycle also with approximately three-batch fuel management. It is anticipated that the equilibrium batch average fuel discharge burnup will reach about 43,000 MWD /T and that the maximum rod average burnup will be about 50,000 Kr.'D/T.

This report is organized as follows. -The technical bases for assuring successful fuel performance at extended burnups-are discussed in Section

2. .The environmental impact duc to the effects of longer cycles and extended burnup on the uranium-fuel cycle are addressed in Section 3. In Section 4, the radiological effects of the eighteen-month extended burnup fuel. cycle are presented.- Section 5 suunarizes the above discussions and draws a conclusion regarding the environmental impact of the planned action.

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2.0 Fuel Performance Successful fuel performance is required for the environmental impact to remain acceptable as a result of increases in the fuel burnup. As discussed below, technical evaluations performed to date using available operational and experimental data indicate a high likelihood that the current C-E fuel design, or slight modifications thereof, can successfully be operated to higher burnups than are now standard practice without increasing the propensity for fuel rod failure.

C-E currently has over 220,000 standard fuel rods operating at a defect level of approximately 0.03 percent (Reference 1). Thus, far, over 18400 rods have been successfully operated to between 32,000 and 44,000 WD/T.

Since the upper limit of primary coolant activity (Technical 5pecification limit) for operation of Calvert Cliffs 1 and 2 is equivalent to approximately a 0.25 percent defect level, a significant margin exists to cover _the possibility of increased failures should they

' occur from extended burnup. However, this is not expected to be the case. As discussed in the following sections, C-E has in place numerous

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irradiation programs to demonstrate that fuel rods can be successfully operated to higher discharge exposures with the same or better performance than is typical of standard PWR fuel.

Since. the great majority of the technical bases for fuel design applicable to .the operating conditions at Calvert Cliffs is limited to burnups up to about 37,000 MWD /T, some technical uncertainties at extended burnups do exist in such areas as pellet clad interaction,

, external. clad corrosion, fuel assembly dimensional changes, and fission gas release. The first two areas are those which determine the effects of extended burnup on the propensity for fuel failure; the latter two areas are not directly related to fuel rod integrity but are included for completeness of the discussion.

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2.1 Pellet Clad Interation (PCI)

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PCI has - been identified as the cause of _ a limited number of ~ fuel rod failures due to rapid power changes (ramps) in light water reactors. The L

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4 propensity for fuel failures due to PCI is dependent on whether the fuel is in a condition susceptible to PCI, and the probability that a change in power of sufficient magnitude and rate will occur.

Considerable work is being done to define the major causes of PCI, that

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is, the effects of power change rate, final power, power increase and burnup. PCI failure has not been observed in fuel with very low burnups. Initially, fuel becomes more susceptible to PCI failure as burnup increases because of the closure of the as-fabricated gap between fuel and cladding as a result of clad creepdown, pellet swelling and relocation and because of the increased availability of iodine and cesium (fission products) which are considered to be the primary corrosive species contributing to stress corrosion cracking. [

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Currently, - the' Studsvik International Cooperative Over-Ramp Program and the CE/KWU Ramp Program at Petten have perhaps the most applicable data on ramp testing as a ' function of burnup (see Figure 2), [

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Figure 2 shows C-E's acquisition schedule for high burnup PCI data.

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Confirmation of acceptable behavior at high burnups will thus be available well in advance of the time when full scale extended burnups are attained in Calvert Cliffs.

2.2 External Clad Corrosion Clad corrosion involves a reaction of the clad with water, leading to the formation of an oxide on the external cladding surface. Although corrosion on the outside surface of PWR fuel rods has not been a cause of fuel rod failures, this phenomenon is of potential concern for extended burnup fuel since technical uncertainties exist when extrapolating present data to extended exposures. Figure 3 shows that corrosion data from Calvert Cliffs 1 for exposures up to 46,000 MWD /T are currently available. These data show no reason for not extending burnup because of external clad corrosion. ' Additionally, data from the joint EPRI/CE/KWU Corrosion Program will provide data for exposures of up to 50,000 MWD /T during 1981, well in advance of the time when full scale extended burnups are attained in Calvert Cliffs.

2.3 Fission Gas Release O

Extending the burnup increases the amount of released fission gas from

. the fuel. pellets which increases the internal fuel rod pressure. Should the pressure increase sufficiently, a possible concern arises with respect to the performance of the fuel cladding. []

]_However, even if future analyses show that this condition is predicted, the criterion of no clad lift-off will continue to be maintained. Specifically, the fuel design criterion will. continue to include a requirement that the cladding will not creep away from contact with the fuel. ~Thus, no adverse consequences

-would be expected. -

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The availability of data on the effect of high burnup on fission gas release is shown in Figure 4. As can be seen, data are already available up to 37,000 MWD /T and five sources of data in the 40-45,000 MWD /i range will be available well before full scale extended burnups are encountered in Calvert Cliffs.

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2.4 Fuel Assembly Dimensional Changes Extending fuel burnup gives rise to two concerns in the area c' fuel assembly dimensional changes, viz., bowing and reduction of axial clearance.

Due to a combination of stress relaxation, growth and creep, fuel has been found to bow. Excessive bowing could affect power peaking, DNS and ,

assembly handling. The extent of bowing in C-E fuel has been minor, even af ter burnups of up to 46,000 MWD /T as shown by recent data from the Calvert Cliffs-1 reactor (see Figure 3). C-E data obtained after 1, 2, 3 and 4 cycles of operation show that the effect of burnup on the extent of bowing beyond approximately two cycles of exposure is very small. In particular, the extent of fuel rod bowing appears .to saturate after 20,000 MWD /T. Therefore, it is expected that the incremental bowing due to extended burnup will not be significant.

Zircaloy clad fuel rods increase in length during reactor service due tt irradiation induced growth of the Zircaloy cladding. This is a well characterized effect and requires only the proper design allowance between the assembly upper end fitting and the top of the fuel rods.

Figure 3 shows the projected schedule for acquiring additional data on dimensional stability. This indicates that '.he C-E/EPRI/Calvert Cliffs-1

. program has already provided data up to 46,000 MWD /T. These data indicate no dimensional stability limitation in achieving high fuel assembly: burnups. In addition, data for exposures of up to 52,000 MWD /T will be available well before such burnups are attained in Calvert Cliffs.

In summary, significant factors affecting fuel rod integrity and design at extended burnup are continuing to be closely investigated. In addition, where technical uncertainties exist, C-E has been and is currently participating in extensive irradiation programs that provide

- data to assure successful fuel performance at extended fuel burnups. In all cases, an adequate high burnup data base will be available prior to achieving full scale high burnup in the Calvert Cliffs Units.

3.0 Uranium Fuel Cycle _

There is a trend in the industry toward eighteen-month fuel cycles. This trend is motivated in part by an anticipated improvement in plant capacity factor (due to the reduced number of refueling outages) and the associated decrease in required replacement power. The planned action for Calvert Cliffs is a change from annual to eighteen-month fuel cycles and an increase in fuel discharge exposure which is consistent with successful fuel performance. This section examines the differential environmental impact of the planned action and cencludes that the impact is favorable.

Reload fuel management schemes to date for Calvert Cliffs have employed

. batch average discharge exposures of about 29,000 MWD /T; the planned extended burnup design employs a batch average discharge exposure of up to about -43,000 MWD /T. For a given refueling interval, the hichar discharge fuel exposure plan shaws a sn. aller fraction of the ccre being replaced at each refueling. Table 1 provides a comparison of estimated fuel management information for the planrad 18-month cycles with extended fuel burnup versus operation of 18-month cycles with lower fuel burnups.

The low burnup design assumes that one-half of the core will be replaced at each 18-month refueling outage; fuel assemblies would be -burned for three years (similar to the current practice for annual cycles). The extended burnup design shown as;umes that one-third of the core is replaced at each 18-month refueling outage and that fuel assemblies are burned for four and one-half years (a 50% increase from current practice). This three-batch design yields the highest burnup for a range of fuel management plans being considered; des,igns with somewhat lower average burnups may actually be implemented in Calvert Cliffs.

The environs tal impact associated with the out-of-core uranium fuel cycle- includes -effects due to uranium mining and milling, UF 6

production, enrichment, fuel fabrication, transportation, and spent fuel storage and disposal (Reference 8). Implementation of the planned extended burnups will reduce the environmental impact of the Calvert -

Cliffs uranium fuel cych a consequence of lower head-end and tail-end ,

fuel cycle requirements. In comparing cycles of the same length, it is observed from Table 1 that increased fuel exposure is expected to reduce U038 requirements (and hence uranium mining and milling and UF 6 production requirements) by up to 13%, to reduce separative work

. requirements by up to 4%, and to reduce the number of assemblies fabricated (and hence conversion, fabrication, fuel transportation and spent _' fuel -storage' requirements) by up to 33%. Implementation of extended burnups will educe lend and water use and liquid, solid and

' radiological effluente associated with the above requirements and hence

. result in a re b ec environmental impact from these effects.

Cc Pison of - the eighteen-month extended burnup cycle with continued operation,of an annual low burnup cycle shows somewhat smaller reductions but still-results in an overall decreased environmental impact. As shown

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in Table 2, wnila the reduction in the number of assemblies fabricated and discharged continues at up to 33%, the reduction in U0 3g s

requirements for this comparison drops to 4%. The increase in separative work' requirements . entails an increased consumption of electrical energy

.. at the . enrichment 1 facility. However, this increase can be favorably compared with the increase in energy production which may be obtained by

. the reduced; number of refueling outages with an 18-month fuel cycle.*

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Two' aspects of. higher burnup fuel which should be discussed with regard, to storage design.~. requirements-=are . decay heat considerations and criticality-conditions. Decay heat considerations'should not adversely i

The additional SWUs consume 48 GWhe/ year but'the reactor produces an +

additional-231 GWhe/ year

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. affect fuel storage requirements. This is because the short term (first few weeks after reactor shutdown) decay heat production from high burnup fuel is virtually identical to the decay heat production from lower burnup fuel (since almost all of the decay heat produced in this time interval is from short-lived fission products which reach a saturated concentration during the first year of irradiation). Higher burnup fuel does not exhibit significantly higher decay heat pcoduction than lower burnup fuel until longer periods after shutdown at which time the decay heat production is significantly lower (by an order of magnitude or more) than the decay heat production at shorter time intervals after shutdown (see Figure 5). Heat removal requirements for the spent fuel storage pool are thus unaffected since cooling requirements are based principally on the earlier (larger) heating rates which are unaffected by increased fuel exposure. Since the initial enrichment of fuel designed for extended burnups is higher than that of fuel designed for lower burnups (see Table 1), criticality evaluations of shipping, handling, and storage actWities are being addressed in licensing proceedings separate from this report. As discussed above for fabrication, the successful implementation of higher burnup fuels will result in a reductinn in the number of spent fuel assemblies to be stored of up to 33%.

4.0 Radiological Impact of Eighteen-Month Extended Burnup Fuel Cycles This section provides an assessment of the radiological impact of operating Calvert Cliffs with a planned fuel cycle design consisting of an 18-month interval between refuelings and increased discharge exposures. Consideration is given primarily to fission product related nuct. ides since they are the only radioisotopes significantly affected by the planned fuel cycle. A comparison of radiological data for a low burnup . annual fuel cycle is made with values for the planned extended burnup fuel cycle. From this comparison, it has been determined that the source -terms are dominated by the short-lived, radienuclides, the concentrations of which saturate during the first cycle of irradiation.

Therefore,- with the high specific . activity short-lived nuclides constitut'ng the. predominant radiological concentrations and energy levels, ' uperation of the planned extended fuel cycle with the expected E ,

equivalent fuel performance (i.e., no increase in the propensity for fuel failures) has a negligible additional radiological impact. Furthermore,

, a basis for such operation will be the continued use of the same radiological activity technical specification limits as are presently

. being employed.

4.1 Plant Reieases l

An assessment has been made of the environmental impact of anticipated radiological releases from the Calvert Cliffs units operating with the planned eighteen-month extended burnup fuel cycle. The source of

, radiological activity for both gaseous and liquid effluents is the Reactor Coolant System (RCS). l i

The concentrations of nuclides in the RCS were determined through the use of a computer coded mathematical model developed by C E which solves the i mass balance for nuclide production and removal in the fuel pellet region and _the reactor coolant region. In the fuel pellet region, production

mechanisms include direct fission yield, parent fission product decay, and neutron act'ivation, while removal includes decay, neutron activation, l and escape to the coolant. In the coolant region, production mechanisms include escape from the fuel (through defective cladding), parent decay, and neutron activauon; removal is by decay, coolant purification, leakage, and by '.ed and bleed operations for startups, she+ downs, load follow, and

ectivity depletion.

For radionuclides deemed to be of concern with respect to environmental impact. Table 3 shows the incremental changes in RCS fission product nuclide concentration between the planned fuel cycle and the annual low burnup cycle calculated using the- above mathematical modelling.

Increases in the longer lived isotopes are due to the higher core average burnup. ' Generally, the short-lived isotopes reach the same equilibrium  !

concentration in either cycle; however, there are- increases in some of the shorter lived isotopes -due- to the- lower thermal flux in the more highly enriched core and'to the buildup of longer lived precursors.

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4.1.1 Gaseous Releases A summation of the total annual gaseous releases from the Calvert Cliffs units for the years 1978 and 1979 is provided in Table 4A as presented in the semiannual report to the NRC. As shown, the gaseous effluents are a small fraction of the plant technical specifications. The source of the radioactive gases is the RCS and thus the activity released from the plant is a function of the reactor coolant specific activity.

From Table 3, it is concluded that since the significant iodine isotope concentrations in the reactor coolant do not show an increase, the planned fuel cycle does not result in an increase in the site boundary thyroid dose. The increase in fission and activation gas effluent activity, primarily due to the increase in the Kr-85 reactor coolant concentration, would be approximately 6 percent and would remain a small fraction of the plant technical specification.

4.1.2' Liquid Releases The plant operating data summarized in Table 4B provides the quarterly liquid activity releases from the plant for the years 1978 and 1979. As shown, the liquid effluents are a cmall fraction of the applicable limit. As in the case of gaseous eff'euents, the liquid releases are a function of the activity in the reactor coolant.

It is anticipated that there will be an increase in tritium production o due to the higher soluble boron concentration required as a result of extending the fuel cycle length to 18 months. The radiological impact of tr! tium released in liquid effluents is negligible, however, since tritium contributes less than 1.0 percent to the total whole body dose.

The- increases in long-lived isotopes due to the planned fuel cycle will likewise contribute negligibly to the whole body dose which is, in any case,- well below the 40 CR190 lin. its. Table 4C summarizes the whole body dose contributions rest.lting from gaseous and liquid releases from the plant for the yea s 1978 and 1979. The thyroid dose projected to result from the liquio activ.;y releases under the planned fuel cycle.

will not change from that projected for the annual fuel cycle.

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• Plant Accidents 9

This section presents a general review of plant accidents reported in the FSAR and the Environmental Report. For the pur;,ose of discussion here, i plant accidents are divided into categories involving release of reactor

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coolant system activity, fuel rod gas gap activity and reactor core activity. The conclusion is that there is no significant incremental adverse impact caused- by the radiological effects of the planned eighteen-month extended discharge burnup fuel cycle.

4.2.1 Accidents Involving Release of RCS Activity Accidents in this category are those for which it has been determined that fuel cladding integrity is maintained during the accident.

Therefore, the activity available for release is dependent only upon the activity within the RCS..

Examination of Table 3 indicate an increase in RCS noble gas specific activity for Kr-85, Xe-131m, and X -135. However, the whole body dose is dominated by the high specific t.ctivity of Xe-133 which is not changed for the extended fuel cycie. The thyroid dose is determined by the iodine activity; however, the only iodine activity which increases is that of the long-lived I-129 which makes a neglibible contribution since it has a low specific activity (see Table 5). Therefore, it is concluded that there is no significant increase in radiological consequences due to extended burnup fuel for accidents involving release of RCS activity.

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' 4.2.2 Accidents Involving Fuel' Rod Gas-Gao Activity Release Accidents in this category involve loss of integrity of the fuel rod cladding and the subsequent release of the activity contained in the fuel rod gas gap. For accidents or transients which involve potential fuel I

. failure, the. quantities (curies) of radioactive gases and volatiles in the fuel gas gap.are pertinent.- Although larger quantities of long-lived fission -products - will - be present in the fuel, review of the applicable source . te' is ! indicates th't a the doses are .du'e almost exclusively to short-livs. Isotopes of radioactive gases-and volatiles. This is because

.the short-lived isotopes have high specific activities, while the low specific activities of long-lived isotopes result in a small contribution to the duae. Since these short-lived gases and volatiles reach equilibrium ccncentration after aDout one year of irradiation, there is

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no significant increase in site boundary dose due to the release of radioactive gases and volatiles from extended burnup fuel as compared to lower burnup fuel. ,

As an example, the radiological consequences of the fuel handling accident are addressed here because it is representative of this type of accident. Table 6 presents a comparison of the significant radioisotopes for this accident for low burnup (28,800 MWD /T) and extended burnup (43,200 MWD /T) spent fuel assemblies as calculated by the ORIGEN computer code ' (Reference 10). Since the dose is dominated by the high specific

' activity sf Xe-133 which has not changed, there is no significant change to the whole body dose from th.is accident. Furthermore, since there is no significant change in the iodine isotope I-131, there is no significant change in the thyroid dose.

4.2.3 Accidents Involving Release of Reactor Core Inventory The FSAR provides -an analysis of the radiological consequences of a postulated maximum hypothetical loss of coolant accident (LOCA).

Although increased fuel failures during the LOCA event are not

. anticipated, it should be noted that regulations require that site

' boundary doses be evaluated under the conservative assumption of 100%

. fuel failure; thus, any. tendency for increased furi failures during LOCA with ~ high burnup fuel will not alter the perceived consequences of the loss-of-coolant accident. This accident is postulated to result in the release of 100% of the noble ges, 50% of the iodine and 1% of all other fission product' isotopes present in the reactor core at the end of the fuel cycle. Site boundary dases are due almost exclusively to the short-lived fission products which reach saturated concentrations during the first year of irradiation. Thus, the inventory (curies) of isotopes which are'significant in the calculation of site boundary dose during the loss-of-coolant accident ' is virtually independent ' of burnup (see Table 7). In particular, when the source terms in Table 7 as calculated by

the ORIGEN computer code (Reference 10) are weighted by the Oose Factors of Reference 9, it is clear that there is no significant increase in the l

. dose due to either the noble gas cr the icaine source terms. It is thus

-concluded that there will be no additional significant radiological impact resulting from extended fuel cycle operatiens fcr these events as postulated in the FSAR.

4.3' Fuel Hand'ing During Refueling Dose rates in the spent fuel pool and refueling cavities are a function of the sources within the irradiated fuel and the fission and corrosion products in the pool water. The dose rate is dominated by the high specific activity shcrt-lived isotcpes which reach approximately the same equilibrium level. Furthermore, the dcse rate is required to be within technical specifications.

The actual dose is determined by the dose rate and the duration of the exposure. Since the planned acticn involves a refueling outage every eighteen months rather than every twelve conths, the dose component due to refueling is reduced by about one-third.

4.4 Transportation of Fuel The primary impact of extended fuel exposures on the transpcrtation of

. fuel to and from the -Calvert Cliffs nuclear power plant site is the substantial reduction in the number of fuel assemblies to -be

- transported.. As discussed in Section 3.0, a reduction in the number of fuel assemblies of up to 33% would be expected as a result of the planned

. actions. Somewhat higher enrichments (d 4.1 w/o U-235) will be utilized, and criticality evaluations of shipping, handling and stcrage activites are being addressed separately from this report. The dose and heat rates of newly- discharged fuel assemblies are dominated by the high specific activity short-lived isotopes whose ccncentrations are. unaltered by burnup ' extensions. . As the fuel cools dcwn, the centributicn of the

'long-lived. isotcpes becomes relatively more impcrtant, and slight increases in the _ccoldown tire are required to match the dose er heat

rate of lesser expesed fue'. Figure 6 shows, for exa ple, that a fuel assembly exposed to 42,000 n*D/T would require an additional 30 cays of cooldown to match the gama energy release cf fuel exposed to 28,000 L'D/T at 90 days caoldown. When the dose and heat rates have been matched, the envir9nmental impact of transporting extended burnup fuel depends only upon the number of fuel assemblies being transportec; as noted earlier, up to 33% fewer fuel asu lies are requireo as burnup is

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increased. There is therefore a reduced environmental impact due to the increase in fuel discharge exposure.

5.0 Sumary and Conclusion Baltimore Gas and Electric is planning to operate the Calvert Cliffs units on an 18-month fuel cycle with average fuel discharge burnups extended from about 29,000 L'D/T up to about 43,000 W'D/T. Fuel demonstration programs are in place to verify acceptable fuel performance at these extended burnups. The environmental impact of the Calvert JCliffs uranium fuel cycle is generally reduced due to reductions in uranium mining and- milling, UF pr ducticn, fuel fabrication, 6

transportation, and spent fuel storage cnd disposal. Radiological effects are not significantly changeo since doses are generally dcminated by. short-lived fission products which reach equilibrium levels at lower burnups.

Based on the above statements, it is concluded that the planned operation

'of Calvert Cliffs Units 1 and 2 on an 18-month cycle with extended fuel i

.. .burnup results in a net reduction in environmental impact.

1 References

1. M. G. Andraws, et al., "The Performance of C-E Fuel in Operating PWRs,"

TIS-6171, Aaril 30, 1979.

2.- " Final Report on the Studsvik Over-Ramp Project," STOR-37, (Proprietary) to be published.

3. " Examination of Calvert Cliffs I Test Fuel Assembly After Cycle 3,"

CE/EPRI Report NPSD-87, September 1979.

4. " Fort Calhoun Poolside Inspection Progra': at End of Cycles 4 and 5,"

CE/00E Report CEf;D-383, to be published.

5. "CE/KWU/EPRI Waterside Corrosi?q of Zircaloy Clad Fuel Rods, RP-1250-01, Task A, Review of Fuel Rod Waterside Corrosion Behavior," t;PSD-79, June 1979.

6. "CE/EPRI/KWU Waterside Corrosion of Zircaloy Clad Fuel Rods, Review of PWR Fuel Waterside Corrosion Behavior," EPRI-fiP-1472, August 1980.

7. " Gas Release and Microstructural Evaluation of Three-Cycle Fuel Rods from

-Calvert cliffs-1," CE/EPRI Report t;PSD-Il9, December 1980.

ss

. 8. 10 CFR Part 52, Table S-3.

. 9. " Numerical . Guides for Design Objectives and Limiting Conditions for  ;

Operation to. Meet lthe Criterion 'As Low as Practicable' for Radioactive ]

Material in ' Light-Water Cooled fluclear Power Reactor Effluents,"

- WASH-1258, Table A-4 of Annex A, July 1973.

10.. M. J. Bell, . "0RIGEti, - The ORtil Isotope Genera 'on and Depletion Code,"

ORNL-4628, May_1973.

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TABLE 1 Comparison of Fuel Management Information for 18-Month Low Burnup and Extended Burnup Fuel Cycles Extended Low Burnup Burnup 18-Month 18-Month Relative Cycles Cycles Chang.

Fraction of Core Replaced per Refueling 0.50 0.33 Batch Average Discharge Burnup (MWD /PTU) 28,800 43,200 Equilibrium Cycle Enrichment (w/o) 3.01 3.87 Number of Assemblies Fabricated / Discharged Equilibrium Cycle 108 72 30-Year Cumulative 2160 1440 -33%

U038 Requirements (ST)*

Equilibrium Reload 305 266 30-Year Cumulative 6096- 5309 -13%

Separative Work Requirements (103 Syg)*

Equilibrium Cycle 184 176 30-Year Cumulative 3673 3535 -4%

Heavy Metal Discharged in Spent Fuel (MTM)

Equilibrium Cycle 42 28 30-Year' Cumulative 839 560 -33%

• Tails Composition 0.2 w/o

TABLE 2 comparison of Fuel Management Information for Annual and Extended Fuel Cycles Extended Low Burnup Burnup Annual 18-Month Relative

. Cycles Cycles Change Fraction of Core Replaced per Refueling 0.33 0.33 Batch Average Discharge Burnup (MWD /MTU) 28,800 43,200 Equilibrium Cycle Enrichment (w/o) 2.76 3.87 Equilibrium Cycle 72 72 30-Year Cumulative 2160 1440 -33%

U038 Requirements (ST)*

Equilibrium Reload 185 266 30-Year Cumulative 5549 5309 -4%

3 Separative Work Requirements (10 5WU)*

Equilibrium Reload 107 176 30-Year Cumulative 3212 3535 +10%

Heavy Metal Discharged in Spent Fuel (MTM) 1 Equilibriuim Cycle 28 28 30-Year Cumulative 839 560 -33%

• Tails composition 0.2 w/o s

E TABLE 3 Incremental Changes in Reactor Coolant System Fission Product Specific Activities From the Low Burnup Annual Fuel Cycle to

, the Eighteen-Month Extendea Burnup Fuel Cycle Fractional Fractional Nuclide Change Nuclide Change Kr-85m 0.0 Te-129 0.01 Kr-85 0.62 1-129 0.27 Kr-87 0.0 1-131 0.0 Kr-88 -0.0 Te-132 0.0 Xe-131m 0.05 I-132 0.0 Xe-133 0.0 I-333 0.0 Xe 135 0.12 Te-134 0.0 Xe-138 0.0 I-134 0.0 Cs-134 0.19 Br-84 '0.0 1-135 0.0

' Rb-88 0.0 Cs-136 0.0 l Rb-89 0.0 Sr-89 0.19 Cs-137 0.24 Sr-90 0.53 Cs-138 0.0 Y-90 0.48 Ba-140 0.0 Sr-91 0.0 La-140 0.0 Y-91 .

0.13 Pr-143 0.0 Zr-95 0.23 Ce-144 0.15

% Mo-99 0.0 Ru-103 0.0 Ru-106 0.19 f

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TABLE 4A Total Annual Airborne Radiological Effluents

. 1978 A. Fission and Activation Gases 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr

1. Total Release, Ci. 2.2(+4) 8.4(+2) 2.9(+3) 1.7(+3)

-2. Average Release Rate, C1/sec 2.8(+3) 1.1(+2) 3.6(+2) 2.2(+2)

3. Percent of Quarterly Technical Specifi-cation Limit, % 1.6(+1) 6.2(-1) 2.1 1.3 B. Iodines

1. Total Release, Ci 5.7(-3) 1.4(-2, 5.3(-2) 5.0(-2)

2. Average Release Rate, Ci/sec 7.2(-4) 1.7(-3) 6.7(-3) 6.3(-3)

3. Percent of Quarterly Technical.Specifi-cation Limit, % 4.l(-1) 1.1 4.2 4.0 C. Particulates

1. Particulates Witn Half Lives. Greater Than 8 Days, ci 1.6(-3) 8.4(-4) 7.9(-3) 3.3(-3)

'2. Average Release Rate, C1/cc 2.l(-4) 1.i(-4) 1.0(-3) 4.2(-4)

3. Percent of Quarterly Technical Specifi- '

cation Limit, %

• Percent of technical specification limits for iodines includes particulates with half' lives greater than 8 days.

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TABLE AA (contindeo)

Total Annual Airborte Radiological Effluents

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1979 A. Fission and Activation Gases 1st Otr 2nd Qtr 3rd Otr 4th Qtr

l. Total Release, Ci

. 2.5(+3) 8.8(+2) 5.7(+3) 'l.0(+3)

2. Average Release Rate, Ci/sec 3.2(+2) 1.l(+2) 7.9(+2) 1.3(+2)

3. Percent of Quarterly

. Technical Specifi-cation Limit, % 2.2 3.0 4.0 7.3(-1)

B. Iodines

.l. Total Release, Ci 1.6(-1) 6.0(-2) 5.l(-2) 2.9(-2)

2. Average Release' Rate, Ci/sec 2.0(-2) 7.6(-3) 6.4(-3) 3.7(-3)

3. Percent of Quarterly

' Technical Specifi-cation Limit, % 1.3(+1) 4.7 4.0 2.3 C. Particulates -

1. Particulates With Half Lives Greater Than 8 Days, Ci 1.7 2.9(-2) 4.8(-2) 5.9(-3)

2. Average Release Rate, Ci/cc 2.1(-1) 3.7(-3) 6.l(-3) 7.4(-4)

3. Percent of Quarterly Technical Specifi-cation-Limit, %

~^

• Percent of technical specification limits for iodines' includes particulates with half lives greater than 8 days.

TABLE 4B Total Annual Liquid Radiolooical Effluents 1978 A. Fission and Activation Products 4th Qtr 1st Otr 2nd Otr 3rd Qtr

l. Total Release, Ci 1.6 6.4(-1) 1.5 2.3

2. Average Diluted Concecentra-tions Ci/sec 2.8(-9) 1.l(-9) 2.6(-9) 4.0(-9)

3. Percent of Applicable 4.0 Limit, % 2.8 1.1 2.6 B. Tritium

1. Total Release, Ci 8.4(+1) 1.l(+2) 1.4(+2) 1.3(+2) 2.-Average Diluted Concentra-tions, Ci/ml -1.5(-7) 1.8(,7) 1.3(-7) 2.2(-7)

3. Percent of Applicable Limit, % 4.8(-3) 6.l(-3) 7.8(-3) 7.4(-3)

.C. Dissolved and Entrained Gases 3.67 2.3 2.1(+1) 1.3

1. Total Release, C1

2. Average Diluted Concentra-tions, Ci/ml. 6.3(-9). 4.0(-9) 3.6(-8) 2.2(-9)

3. PerceEt'of Applicable Limit, % 2.l(-1) _ l.4(-1) 1.2 7.4(-2) s

TABLE 4B (continued)

Total Annual Liquid Radiological Effluents

.- 1979 A. Fission and Activation Products 1st Qtr 2nd Qtr 3rd Otr 4th Qtr

1. Total-Release, Ci 1.4 9.7(-1) 9.l(-1) 4.5

2. Average Diluted'Concecentra-tions Ci/sec 2.6(-9) 1.7(-9) 1.7(-9) 7.7(-9)

3. Percent of Applicable Limit, %- 2.6 1.7 1.7 7.7 B. Tritium

1. Total Release, Ci 1.8(+2) 1.2(+2) 1.3(+2) 1.3(+2)

~2. Average Diluted Concentra-tions, Ci/ml 2.4(-7) 2.l(-7) 2.4(-7) 3.1(-7)

3. Percent of Applicable

-Limit, % 8.0(-3) 7.0(-3) 7.9(-3) 1, .0(-2 )

C. - Dissolved and Entrained Gases

.l. Total Release, Ci 4.4 3.0 6.0' 2.3

2. Average Diluted Concentra-tions, Ci/ml 8.3(-9) 5.1(-9) 1.1(-8) 3.8(-9)

3. Percent of Applicable Limit, % 2.8(-1) 1.7(-1) 3.6(-1) 1.3(-1)

I 1

l i

i 1

1

--. _ l

Table 4C 1978 Gamma Immersion Dose at the Site Boundary in mrem Location- 1st Qtr. 2nd Qtr. 3rd Qtr. 4th Qtr. Total 9.37E-2 2.92E-3 2.53E-3 3.91E-3 1.03E-1 SE SSE 6.03E-2 2.93E-3 3.01E-3 4.99E-3 7.12E-2 S 4.83E-2 3.08E-3 1.88E-3 5.31E-3 5.86E-2 SSW 3.34E-2 3.06E-3 1.48E-3 5.69E-3 4.36E-2 SW l.41E-2 2.29E-3 2.70E-3 1.88E-3 2.10E-2 WSW 1.64E-2 1.70E-3 1.62E-3 1.77E-3 2.15E-2 W l.26E-2 1.41E-3 3.90E-3 9.27E-4 1.88E-2 i WNW l.41E-2 1.19E-3 4.41E-3 1.70E-3 2.14E-2 l.40E-2 1.20E-3 2.32E-3 7.26E-4 1.82E-2 NW 1978 Total Body Dose in mrem or Liquid Release

~

1st Qtr. 2nd Qtr. 3rd Qtr. 4th Qtr. TOTAL 3.48E-3 1.87E-3 2.81E-3 5.12E-3 1.33E-2

-4 m

l

Table 4C (continued) 1979 Gamma Immersion Dose at the Site Boundary in mrem Location 1st Otr. 2nd Otr 3rd Otr 4th Otr. Total

. SE 2.58E-2 8.86E-4 1.83E-3 6.88E-3 3.54E-2 SSE 5.53E-3 1.17E-3 3.26E-3 1.82E-3 1.18E-2 S 2.84E-3 1.21E 3 6.69E-3 1.20E-3 1.19E-2 SSW 4.53E-3 8.21E-4 2.00E-2 1.06E-3 2.64E-2 SW l.69E-3 4.75E-4 2.21E-2 5.46E-4 2.48E-2 WSW 2.02E-3 5.27E-4 1.34E-2 5.00E-5 1.60E-2 W 2.54E-3 5.56E-4 9.79E-3 3.49E-5 1.29E-2 WNW 3.35E-3 1.07E-3 4.55E-2 1.70E-4 5.01E-2 NW 3.88E-3 1.75E-3 4.62E-3 2.35E-4 1.05E-2 1979 Total Body Dose in mrem fce Liquid Releases 1st Otr. 2nd Otr. 3rd Otr. 4th Otr. TOTAL 3.,55E-3 1.20E-3 2.58E-3 7.55E-3 1.49E-2 l

e

TABLE 5 Calculated Reactor Coolant System Fission Product Specific Activities for the Annual Cycle

, Concentration in Concentration in Reacter Coolant Reactor Coolant Isotope ( p CT/cc) Isotope (hlCi/cc)

Br-84 0.0466 Xe-133 181

~

Kr-85m 1.49 Te-134 0.0262 Kr-85 0.885 I-134 0.62 Kr-87 0.81 Cs-134 0.10 Kr-88 2.6 I-135 2.7 Rb-88 2.55 Xe-135 7.53 Rb-39 0.064 Cs-136 2.55 x 10-2 Sr-89 5.07 x 10-3 Cs-137 0.32 Sr-90 -2.61 x 10-4 Xe-138 0.36 Y-90 1.02 x 10-3 Cs-138 0.69 Sr-91 3.56 x 10-3 Ba-140 6.11 x 10-3 Y-91 0.111 La-140 5.85 x 10-3 Zr-95 9.35 x 10-7 Ce-144 0.0040 Mo-99 2.03 Pr-143 0.0058 Ru-103 4.13 x 10-3 Ru-106' 2.48 x 10-4 Te-129 0.0251 1-129 7.21 x 10-8

- I-131 3.97 Xe-131m 1.48 Te-132 0.33 I-132' 1.09 I-133 5.66 4

^D

TABLE 6 i .

Activity Release to Spent Fuel Pool Water FSAR Activity Release (Curies)

Isotope 14 Outer Rods 176 Rods Percentage Increase

(( omplete for Extended Fuel Cycle

• Assembly)

Kr-85 4.90 x 102 2.86 x 103 40.5 Xe-131m .6.03 x 101 3.20 x 102 1,7 Xe-133 1.42 x 104 7.34 x 104 0.0

~

I-131 1.20 x 104 6.55 x 10 4 1.9

-* Increase. in specific activities for fuel assemblies at 43,200 MWD /T relative to fuel assemblies at 28,800 MWD /T.

.e i

9 e

g -, -c-- - e % w. y

L TABLE 7

~ Comparison of'the Calculated LOCA Release Source Terms (Curies)

Core Inventory for Core Inventory for the Eighteen Penth Fuel Cycle **  % Change

-Isotope the Annual Cycle

• 7.48+7 +1.3 I-131 7.38+7 +0.9 1.08+8 1.09+8 "I-132 1.53+8 -0.6 I 133 1.54+8 -1.2 1.66+8 1.64+8

-I-134 1.42+8 -0.7 I-135 1.43+8 1.97+7 -3.9 Kr-85m 2.05+7 +45.3 5.17+5 7.51+5 Kr-85 -5.0 3.81+7 3.62+7

'Kr-87 5.19+7 -4.4 Kr-88 5.43+7 5.24+5 +0.8 Xe-131m 5.20+5~ 0.0 1.53+8 1.53+8 Xe-133 +1.7

'3.00+7 5.05+7 Xe-135m 2.97+7 -1.0 Xe-135 -3.00+7- -2.4 1.25+8 1.22+8 Xe-138

~* Core average burnup at end of cycle is 19,200 L'D/T -

• • Core average burnup at end of cycle is 28,800 L'D/T

. 46 v -?

O

l l

l Figure 1 CALVERT CLIFFS UNITS 1 & 2 PLANNED OPERATING SCHEDULE I

l 1978 1979 1980 1981 1982 1983 l

UNIT 1 l

b I CYCLE 3 CYCLE 4 CYCLE S CYCLEG DATCH E BATCH F BATCH G BATCH H 9900 MWD /T 11,800 MWD /T 13,200 MWD /T 14,500 MWD /T UNIT 2 f CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5

! I CYCLE 1 BATCH F BATCH G BATCH D BATCH E DATCH 16,200 MWD /T 14,500 MWD /T 9800 MWD /T 11,200 MWD /T A, B & C l

e l

l l

i

- e BURNUP MILESTONES FOR FUEL IR R ADI ATIO N TESTS 1980 1981 1982 1983 1984 1985 1986 1987 8 I A E t t t I I A 1 I f I t t I I I 1 1 I I I PELLET CLAD INTERACTION (PCI)

[ (1) ]

STUDVIK SUPER RAMP PROGRAM 35 .40

[ .]

C-E/ DOE /KWU/KFA R AMP PROGRAM

>40 AT PETTEN C-E/ DOE PROGRAM IN FT. CALHOUN (SAVFUEL STARTUP PCI,14 x 14) 20 30 12 40 SG 60 C F./ DOE ARKANSAS SEGMENTS (16 x 16)*

I I

• f I f I I i f f f I l l I f t t t t t i I

• THESE PROGRAMS INCLUDE SEGMENTS E CONSERVATIVE ESTIMATE OF DATA AS SEEDS FOR POSSIBLE FUTURE

- AVAll ABILITY AT ESTIMATED RAMP TESTING BURNUP (GWD/T)

(1) STUDSVIK OVER RAMP PROJECT FINAL REPORT, STOR-37, /\

PROPRIETARY, TO BE PUBLISHED (REF. 2) /xx\ CALVERT CLIFFS LEAD ROD BURNUP IN DISCHARGED 3ATCH (2) PROPRIETARY DAT A NOW AVAILABLE (GWD/T)

Figure 2

m q - -

0- . . [

BURNUP MILESTONES FOR FUEL IRRADIATION TESTS .

1980 1981 1982 1983 1984 1985 198G 1987 I I I I I I I I I l t I i t i i I I t i I I I t DIMENSIONAL STABILITY

_ (1) _ (2). _

37 46 52 C-E/EPRI AT CALVERT CLIFFS 1 (14 x 14) i 12 l40 SG 60 C-E/EPRI PROGRAM IN ARKANSAS (16 x 16) l C-E/ DOE PROGRAM IN ARKANSAS (16 x 16) 40 SG 60 i

(3) - - -

C.E/ DOE PROGRAM IN FORT CALHOUN (14 x 14) 3G 49 56 t

l I

ZlRCALOY CORROSION _ _

C E/ DOE PROGRAM IN FORT CALHOUN (14 x 14) 49 56 ,

4 l

C E/KWU/EPRI WATERSIDE CORROSION -

(4)

PROGRAM 41 50 i

i C E/ DOE PROGRAM IN ARKANSAS (16 x 16) 40 56 GO C-E/EPRI AT CALVERT CLIFFS 1 (14 x 14) 4G 52 1 . . i i . . , , . . . . . . . . . . ._ 2 . . . .

(1) CE/EPRI REPORT NPSD 87 EXAMINATION OF CALVEST CLIFFS.

xx CONSERVATI CSTIMATE OF DATA TEST FUEL ASSEMBLY AF1ER CYCLE 3 SEPT.1979 (REF . 3 (> , AVAILABILITY AT ESTIMATED (2) D AT A UNDER EVALU ATION BURNUP (GWD/T)

(3) CE/00E REPORT CEND 383, FORT CALHOUN PQ0LSIDE 1%SPECTl0N PROGR AMS  ;

END-OF. CYCLES 4 AND 5,TO BE PUBLISHED (REF. 4 i CALVERT CLIFFS LEAD ROD BURNUP

, x 1 (4) (REFS. 5 and 6)

IN DISCHARGED BATCF' (GWD/T)

Figure 3  ;

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

BURNUP MILESTONES FOR FUEL IR R ADI ATION TESTS 1081 1902 1983 19M 1985 1986 1987 FISSION GAS RELEASE i {0 , , , , , , . , , , , , , , , , 1, , ,,

RAMP TEST AND STEADY STATE RELEASE 1

_ (1) _

STUDVIK SUPER RAMP PROGRAM, 35 >40 E .

J C-E/ DOE /XWU/KFA RAMP PROGRAM -

AT PETTEN >40

[ ]

STEADY STATE RELEASE ONLY (2)

C 37 46

_ E/EPRI AT CALVERT CLIFFS 1 (14 x 14)

E J C E/ DOE PROGRAM IN FT.CALHOUN (14 x 14) 49 56 C-E/ DOE PROGRAM IN ARKANSAS (16 x 16) 40 56 60 (1) PROPRIETARY DATA NOW AVAILABLE i i i e ii (2) CE/EPRI REPORT NPSD-119, GAS i i i i i i i e i i e i i e i e i e RELEASE AND MICROSTRUCTURAL -

EVALUATION OF THREE-CYCLE FUEL CONSERVATIVE ESTIMATE OF DATA RODS FROM CALVERT CLIFFS - 1, *5 AVAILABILITY AT ESTIMATED DEC.,1980 (REF. 7) BURNUP (GWD/T)

CALVERT CLIFFS LEAD ROD BURNUP Figure 4 Axx IN DISCHARGED tt ATCll (GWD/T)

Fir;ure 5 DECAY HEAT vs C O O L D O W f4 T i f.1 E 106 1 I I I l l

~

10 5 _

>s -

ca 2 -

w 5

m

[ 104 _

E: -

p  :

w -

t: -

u -

w O

_ 50 GWD/T

DISCilARGE SURNUP

_ 30 GWD/T

- DISCHARGE BURNUP

. 102 _ __

joi 1 l I I I I O 2 4 6 8 10 12 14 COOLDOWN TIME, YEARS r

Figure 6 GAMMA ENERGY FOR LOW BURNUP FUEL AND EXTENDED BURNUP FUEL vs COOLDOWN TIME 4

' I i i 10 17

- EXTENDED BURNUP FUEL .

(42 GWD/T) l LOW BURNUP FUEL (28 GWD/T)

Q -

s 6

E -

5.,

1 2

e -

b 5

s 4 ~

~

, N )

O l

f I I I 10 16 100 200 300 O

COOLDOWN Tif,AE, DAYS

_ . )