[Federal Register Volume 68, Number 179 (Tuesday, September 16, 2003)]
[Notices]
[Pages 54246-54249]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-23556]


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NUCLEAR REGULATORY COMMISSION

[Docket No. STN 50-454]


Exelon Generation Company, LLC, Byron Station, Unit No. 1; 
Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering 
issuance of an exemption to Title 10 of the Code of Federal Regulations 
(10 CFR) Part 50, for Facility Operating License No. NPF-37 issued to 
Exelon Generation Company, LLC, (Exelon or the licensee), for operation 
of the Byron Station, Unit No. 1, located in Ogle County, Illinois. 
Therefore, pursuant to 10 CFR 51.21, the NRC is issuing this 
environmental assessment and finding of no significant impact.

Environmental Assessment

Identification of Proposed Action

    The proposed action would allow the use of a limited number of fuel 
rods with ZIRLO\TM\ cladding that has a tin content lower than the 
currently licensed tin content range for ZIRLO\TM\ in one lead test 
assembly (LTA) (i.e., LTA M09E). The licensee has also requested 
approval to irradiate two LTAs (i.e., M09E and M12E) that contain low-
tin ZIRLO\TM\ clad fuel rods and two ``standard'' Westinghouse 17x17 
VANTAGE+ ZIRLO\TM\ assemblies (i.e., M10E and M11E) up to 69,000 MWD/
MTU for Byron, Unit 1 Cycle 13 (B1C13). The burnup limits are not part 
of the technical specifications (TS), but are design bases limits for 
the fuel cladding, and limit the current fuel rod-average burnup to 
less than or equal to 60,000 MWD/MTU. The proposed action is in 
accordance with the licensee's application dated January 17, 2003, as 
supplemented by letter dated March 24, 2003. The licensee has indicated 
that it intends to submit an amendment request with respect to an 
increase in the rod-average burnup.

The Need for the Proposed Action

    Available industry data indicates that corrosion resistance of 
nuclear fuel cladding improves for cladding with a low tin content. The 
optimum tin level provides a reduced corrosion rate while maintaining 
the benefits of mechanical strength and resistance to accelerated 
corrosion from abnormal chemistry conditions. In addition, fuel rod 
corrosion/temperature feedback effects have become more limiting with 
respect to fuel rod design criteria. By reducing the associated 
corrosion buildup and, thus, minimizing temperature feedback effects, 
additional margin to fuel rod internal pressure design criteria can be 
obtained.
    As part of a program to address these issues, Westinghouse Electric 
Company (Westinghouse), has developed an LTA program in cooperation 
with Exelon that includes ZIRLO\TM\ fuel cladding with a tin content 
lower than the currently licensed range for ZIRLO\TM\. Use of fuel rods 
using such low-tin cladding requires exemptions from 10 CFR 50.44, 
``Standards for combustible gas control system in light-water-cooled 
power reactors''; 10 CFR 50.46, ``Acceptance criteria for emergency 
core cooling systems for light-water nuclear power reactors''; and 
Appendix K to 10 CFR Part 50, ``ECCS Evaluation Models.''
    In addition, the basis for approval of ZIRLO\TM\ cladding used in 
the Byron core is provided in an NRC safety evaluation addressed to 
Westinghouse, ``Acceptance for Referencing of Topical Report WCAP-
12610, `VANTAGE+ Fuel Assembly Reference Core Report,' '' dated July 1, 
1991. The safety evaluation approved the use of the VANTAGE+ fuel 
design that was described in WCAP-12610-P-A, and found its use 
acceptable up to a rod-average burnup of 60,000 MWD/MTU. Use of the 
VANTAGE+ fuel design in the Byron core beyond that burnup level has not 
been approved yet because of uncertainty in changes in the gap-release 
fraction associated with increasing fuel burnup. The present methods 
for assessing fission gas releases have not been validated with actual 
data at higher peak-rod burnups. Therefore, part of the Westinghouse 
LTA program includes acquisition of actual operating data through the 
limited use of fuel rods in the Byron Unit 1 core to obtain burnup 
levels higher than 60,000 MWD/MTU that will be examined at the end of 
the Byron Unit 1, Cycle 13 (B1C13) fuel cycle.
    Two LTAs (i.e., LTA M09E and M12E) were in use in Byron Unit 2, 
Cycle 10 (B2C10). These LTAs are composed of low-tin and standard 
composition ZIRLO\TM\ cladding. The licensee modified one of the LTAs 
(M09E) to include fresh fuel rods with ZIRLO\TM\ cladding that has a 
tin content lower than that of the ZIRLO\TM\ cladding of the currently 
licensed fuel. No fuel rods were replaced in LTA M12E. Both LTAs will 
be used in Byron Unit 1 Cycle 13

[[Page 54247]]

(B1C13) in non-limiting core locations. In addition, the licensee 
proposes to irradiate two standard 17x17 VANTAGE+ ZIRLO\TM\ assemblies 
(i.e., M10E and M11E) in Byron, Unit 1 Cycle 13 (B1C13), also in non-
limiting core locations. At the end of B2C10, the approximate assembly 
average burnup is expected to be 51,094 MWD/MTU for LTA M09E, 51,123 
MWD/MTU for LTA M12E, 51,457 MWD/MTU for LTA M10E, and 51,423 MWD/MTU 
for LTA M11E.
    The licensee has requested that it (1) be authorized to use the 
modified LTA M09E in Byron, Unit 1 Cycle 13 (B1C13) to obtain data on 
both the use of low-tin ZIRLO\TM\ and high burnup operation (up to 
69,000 MWD/MTU), and (2) be authorized to irradiate the other three 
assemblies (M10E, M11E, and M12E) up to 69,000 MWD/MTU to obtain data 
on the effects of high burnup operation. The proposed irradiation of 
these fuel assemblies does not require a change to the TS; however; 
this burnup will exceed the current design basis limit for the fuel 
cladding of 60,000 MWD/MTU for peak fuel rod-average burnup.
    Irradiation of these four LTAs to a higher burnup will provide data 
on fuel and materials performance that will support industry goals of 
extending the current fuel burnup limits and will provide additional 
insight regarding gap-release fraction related to fuel performance 
behavior at high burnups. The data will also help confirm the 
applicability of nuclear design and fuel performance models at high 
burnups.

Environmental Impacts of the Proposed Action

Background
    In its previous environmental assessments concerning fuel burnup, 
the Commission relied on the results of a study conducted for the NRC 
by Pacific Northwest Laboratories. The results of the study were 
documented in detail in the report, ``Assessment of the Use of Extended 
Burnup Fuel in Light Water Power Reactors'' (NUREG/CR-5009, PNL-6258, 
February 1988). The overall findings of this study showed there were no 
significant adverse effects that would result from increasing the 
batch-average burnup level of 33,000 MWD/MTU to 50,000 MWD/MTU or above 
as long as the maximum rod average burnup level of any fuel rod was no 
greater than 60,000 MWD/MTU. Furthermore, based on the above study and 
the report, ``The Environmental Consequences of Higher Fuel Burn-up,'' 
(AIF/NESP-032), issued by the Atomic Industrial Forum, the NRC staff 
concluded that the environmental impacts summarized in Table S-3 of 10 
CFR 51.51 and in Table S-4 of 10 CFR 51.52 for a burnup level of 33,000 
MWD/MTU are conservative and bound the corresponding impacts for burnup 
levels up to 60,000 MWD/MTU and uranium-235 enrichments up to 5 percent 
by weight.\1\
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    \1\ See ``Extended Burnup Fuel Use in Commerical LWRs; 
Environmental Assessment and Finding of No Significant Impact,'' 53 
FR 6040, February 29, 1988.
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    In this environmental assessment regarding the impacts of the use 
of extended burnup fuel beyond 60,000 MWD/MTU, the Commission is also 
relying on the results of an updated study conducted for it by the 
Pacific Northwest National Laboratory (PNNL) entitled, ``Environmental 
Effects of Extending Fuel Burnup Above 60 GWd/MTU,'' (NUREG/CR-6703, 
PNNL-13257, January 2001). This report represents an update to NUREG/
CR-5009. Although the study evaluated the environmental impacts of high 
burnup fuel up to 75,000 MWD/MTU, certain aspects of the review were 
limited to evaluating the impacts of extended burnup up to 62,000 MWD/
MTU because of the need for additional data about the effect of 
extended burn-up on gap-release fractions. During the study, all 
aspects of the fuel-cycle were considered, from mining, milling, 
conversion, enrichment and fabrication through normal reactor 
operation, transportation, waste management, and storage of spent fuel.
Environmental Impacts
    The NRC has completed its evaluation of the proposed action and 
concludes that there are no significant environmental impacts 
associated with (1) using LTA M09E with fuel rods composed of ZIRLO\TM\ 
cladding that has a tin content lower than the currently licensed tin 
content range for ZIRLO\TM\, and (2) irradiating four fuel assemblies 
(M09E, M10E, M11E, and M12E) to a burnup of 69,000 MWD/MTU. The 
following is a summary of the staff's evaluation:
    The extended burnup assemblies will have a different mix of fission 
and activation product radionuclides than the rest of the core. The 
activities of short-lived fission products will tend to remain constant 
or decrease slightly, while activities associated with activation 
products and actinides tend to increase with increasing burnup. As 
discussed in Attachment 2 to the licensee's January 17, 2003, request, 
although there are variations in core inventories of isotopes due to 
extended burnup, there are no significant increases of isotopes that 
are major contributors to accident doses. In addition, the four fuel 
assemblies will only contribute a small variation in the isotopic 
population of the entire core (193 assemblies). Thus, with extended 
burnup of the four assemblies and their placement in non-limiting core 
locations, no significant increase in the release of radionuclides to 
the environment is expected during normal operation. In addition, no 
change is being requested by Exelon in the licensed technical 
specifications pertaining to allowed cooling-water activity 
concentrations. If leakage of radionuclides from the extended burnup 
fuel assemblies occurs during operation, then the radioactive material 
is expected to be removed by the plant cooling water cleanup system.
    Using the modified LTA M09E in B1C13 with low-tin ZIRLO\TM\ 
cladding and irradiating the four fuel assemblies to a burnup of 69,000 
MWD/MTU will not result in changes in the operation or configuration of 
the facility. There will be no change in the level of controls or 
methodology used for processing radioactive effluents or handling solid 
radioactive waste, nor will the proposal result in any change in the 
normal radiation levels within the plant. Accordingly, the impacts on 
workers and the general population would not be significant because of 
the small radiological effect of the four extended-burnup assemblies.
Environmental Impacts of Potential Accidents
    Accidents that involve the damage or melting of the fuel in the 
reactor core and spent-fuel handling accidents were also evaluated in 
NUREG/CR-6703. The accidents considered were a loss-of-coolant accident 
(LOCA), a steam generator tube rupture, and a fuel-handling accident. 
In addition, Exelon addressed both LOCA and non-LOCA events in 
Attachment 2 to the January 17, 2003 request.
    For LOCAs, the amount of radionuclides that would be released from 
the core (1) is proportional to the amount of radionuclides in the core 
and (2) is not significantly affected by the gap-release fraction. The 
gap-release fraction is a small contribution to the amount of 
radionuclides available for release when the fuel is severally damaged. 
Any increase in the amount of some longer-lived radionuclides available 
for release from the four LTAs (1) will be small and (2) will not 
result in a significant increase in the overall core inventory of 
radionuclides. Therefore, there would be no significant increase in the 
previously calculated

[[Page 54248]]

dose from a LOCA and the dose would remain below regulatory limits.
    The pressurized-water reactor (PWR) steam generator tube rupture 
accident involves direct release of radioactive material from 
contaminated reactor coolant to the environment. As discussed 
previously, no change is being requested by Exelon in the licensed 
technical specifications pertaining to allowed cooling-water activity 
concentrations. The maximum coolant activity is regulated through 
technical specifications that are independent of fuel burnup. 
Therefore, the gap-release fraction does not significantly affect the 
amount of radionuclides available for release during a steam generator 
tube rupture. Therefore, there would be no significant increase in the 
previously calculated dose from a steam generator tube rupture and the 
calculated dose would remain below regulatory limits.
    The scenario postulated to evaluate potential fuel-handling 
accidents involves a direct release of gap activity to the environment. 
The assumptions regarding gap activity are based on guidance in 
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for 
Evaluating Design Basis Accidents at Nuclear Power Reactors'' and 
NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power 
Plants''; the gap activity consists primarily of the noble gases, 
iodines, and cesiums. The only isotopes that contribute significant 
fractions of the committed effective dose equivalent and thyroid doses 
are \131\I and \134\Cs. Similarly, the only isotopes that contribute 
significant fractions of the deep dose are \132\I and \133\Xe. The 
inventory of iodine, the primary dose contributor, decreases with 
increasing burnup. However, gap-release fraction increases as burnup 
increases; this in turn, would increase the calculated dose from a fuel 
handling accident involving one of the four assembles addressed in this 
exemption. As discussed earlier and outlined in NUREG/CR-6703, 
additional information is needed to assess the relationship between 
gap-release fraction and burnup beyond 60,000 MWD/MTU to 75,000 MWD/
MTU. However, based on the trend of the gap-release fraction from 
33,000 MWD/MTU to 60,000 MWD/MTU, the increase in gap-release fraction 
as burnup increases from 60,000 MWD/MTU to 69,000 MWD/MTU is expected 
to be small. Therefore, the staff concludes (1) that the increase in 
the previously calculated dose resulting from a fuel-handling accident 
involving one of the assemblies would not be significant and (2) that 
the dose would remain below regulatory limits.
Environmental Impacts of Transportation
    The environmental effects of incident-free spent fuel 
transportation were also evaluated in NUREG/CR-6703. Incident-free 
transportation refers to transportation activities in which shipments 
of radioactive material reach their destination without releasing any 
radioactive cargo to the environment. The vast majority of radioactive 
shipments are expected to reach their destination without experiencing 
an accident or incident, or releasing any cargo. The incident-free 
impacts from these normal, routine shipments arise from the low levels 
of radiation that are emitted externally from the shipping container. 
Although Federal regulations in 10 CFR part 71 and 49 CFR Part 173 
impose constraints on radioactive material shipments, some radiation is 
not entirely shielded by the shipping container and exposes nearby 
persons to low levels of radiation. Based on the analyses presented in 
NUREG/CR-6703, the staff concludes that doses associated with incident-
free transportation of spent fuel with burnup to 75,000 MWD/MTU are 
bounded by the doses given in 10 CFR 51.52, Table S-4, for all regions 
of the country if dose rates from the shipping casks are maintained 
within regulatory limits.
    Additionally, the environmental effects of spent fuel 
transportation accidents were also evaluated in NUREG/CR-6703. Accident 
risks are the product of the likelihood of an accident involving a 
spent-fuel shipment and the consequences of a release of radioactive 
material resulting from the accident. The consequences of such a 
transportation accident are represented by the population dose from a 
release of radioactive material, given that an accident occurs that 
leads to a breach in the shipping cask's containment systems. The 
consequences are a function of the total amount of radioactive material 
in the shipment, the fraction that escapes from the shipping cask, the 
transport of radioactive material to humans, and the characteristics of 
the exposed population. Considering the uncertainties in the data and 
computational methods, the overall changes in transportation accident 
risks due to increasing fuel burnup of the four fuel assemblies are not 
significant. The calculated doses resulting from a spent fuel 
transportation accident will remain below regulatory limits, and no 
significant increase in the environmental effects of spent-fuel 
transportation accidents are expected.
Non-Radiological Impacts
    With regard to potential non-radiological impacts, the proposed 
action does not have a potential to affect any historic sites. It does 
not affect non-radiological plant effluents and has no other 
environmental impact. Therefore, there are no significant non-
radiological environmental impacts associated with the proposed action.
Summary
    Based on the staff's independent assessment discussed above, the 
NRC concludes that there will be no significant environmental impacts 
associated with (1) using LTA M09E with fuel rods composed of ZIRLO\TM\ 
cladding that has a tin content lower than the currently licensed tin 
content range for ZIRLO\TM\, and (2) irradiating the four fuel 
assemblies (M09E, M10E, M11E, and M12E) to a burnup of 69,000 MWD/MTU.

Environmental Impacts of the Alternatives to the Proposed Action

    As an alternative to the proposed action, the staff considered 
denial of the proposed action (i.e., the ``no action'' alternative). 
Denial of the application would result in no change in current 
environmental impacts. The environmental impacts of the proposed action 
and the alternative action are similar. However, it would deny to the 
licensee and the NRC operational data on low-tin content ZIRLO\TM\ and 
the performance of fuel at extended burnup conditions.

Alternative Use of Resources

    The action does not involve the use of any different resources than 
those previously considered in the Final Environmental Statement for 
the Byron Station, Unit Nos. 1 and 2, dated April 30, 1982.

Agencies and Persons Consulted

    On July 9, 2003, the staff consulted with the Illinois State 
official, Frank Niziolek, of the Illinois Department of Nuclear Safety, 
regarding the environmental impact of the proposed action. The State 
official had no comments.

Finding of No Significant Impact

    On the basis of the foregoing environmental assessment, the NRC 
staff concludes that (1) allowing use of an LTA (i.e., LTA M09E) with a 
limited number of replacement fuel rods with ZIRLO\TM\ cladding that 
has a tin content lower than the currently licensed tin content range 
for ZIRLO\TM\, and (2) permitting irradiation of four fuel assemblies 
(M09E, M10E, M11E, and

[[Page 54249]]

M12E) to a burnup of 69,000 MWD/MTU, will not have a significant effect 
on the quality of the human environment. Accordingly, the NRC has 
determined not to prepare an environmental impact statement for the 
proposed actions.
    For further details with respect to the proposed action, see the 
licensee's letters dated January 17 and March 24, 2003. Documents may 
be examined, and/or copied for a fee, at the NRC Public Document Room, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible 
electronically from the ADAMS Public Library component of NRC's Web 
site, http://www.nrc.gov (the Public Electronic Reading Room). If you 
do not have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC Public Document Room (PDR) 
Reference staff at 1 (800) 397-4209, or (301) 415-4737, or by e-mail to 
[email protected].

    Dated at Rockville, Maryland, this 9th day of September, 2003.

    For the Nuclear Regulatory Commission.
Anthony J. Mendiola,
Chief, Section 2, Project Directorate III, Division of Licensing 
Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 03-23556 Filed 9-15-03; 8:45 am]
BILLING CODE 7590-01-P