South Carolina Electric & Gas Company; Virgil C. Summer Nuclear Station, Unit No. 1; Exemption, 14853-14856 [E8-5513]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 54 (Wednesday, March 19, 2008)]
[Notices]
[Pages 14853-14856]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-5513]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-395]


South Carolina Electric & Gas Company; Virgil C. Summer Nuclear 
Station, Unit No. 1; Exemption

1.0 Background

    The South Carolina Electric & Gas Company (SCE&G, the licensee) is 
the holder of the Renewed Facility Operating License No. NPF-12 which 
authorizes operation of the Virgil C. Summer Nuclear Station, Unit No. 
1 (VCSNS). The license provides, among other things, that the facility 
is subject to all rules, regulations, and orders of the Nuclear 
Regulatory Commission (NRC or the Commission) now or hereafter in 
effect.
    The facility consists of a pressurized-water reactor located in 
Fairfield County in South Carolina.

2.0 Request/Action

    Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), 
Section 50.12, ``Specific Exemptions,'' SCE&G has, by letters dated May 
31 and October 11, 2007, requested an exemption from 10 CFR 50.46, 
``Acceptance Criteria for Emergency Core Cooling Systems for Light-
Water Nuclear Power Reactors,'' and Appendix K to 10 CFR 50, ``ECCS 
Evaluation Models,'' (Appendix K). The regulation in 10 CFR 50.46 
contains acceptance criteria for emergency core cooling system (ECCS) 
for reactors fueled with zircaloy or ZIRLOTM cladding. In 
addition, Appendix K requires that the Baker-Just equation be used to 
predict the rates of energy release, hydrogen concentration, and 
cladding oxidation from the metal-water reaction. The exemption request 
relates solely to the specific types of cladding material specified in 
these regulations. As written, the regulations presume the use of 
zircaloy or ZIRLOTM fuel rod cladding. Thus, an exemption 
from the requirements of 10 CFR 50.46, and Appendix K is needed to 
irradiate a lead test assembly (LTA) comprised of different cladding 
alloys at VCSNS.
    The exemptions requested by the licensee would allow the use of one 
LTA containing either all Optimized ZIRLOTM fuel rod 
cladding or a combination of Optimized ZIRLOTM and 
AXIOMTM fuel rod cladding to continue to be irradiated up to 
a burnup of 75 gigawatt days per metric ton uranium (GWd/MTU).
    Previously, by letter dated January 14, 2005, the NRC staff 
approved the irradiation of four LTAs containing fuel rods with 
Optimized ZIRLOTM and several different developmental clad 
(AXIOMTM) alloys. That exemption was contingent on the fuel 
rod burnup remaining within the applicable licensed limits, which for 
burnup, was a value of 62 GWd/MTU. The licensee inserted those LTAs 
into VCSNS for irradiation in fuel cycles 16 and 17. In the licensee's 
letters of May 31 and October 11, 2007, the licensee requested an 
exemption to continue the irradiation of one of the four LTAs for a 
third operating cycle. This LTA would be irradiated in fuel cycle 18 in 
order to gain high burnup experience. The licensee requested to 
irradiate the LTA to a peak rod average of up to 75 GWd/MTU.
    The licensee also requested an exemption from 10 CFR 50.44, 
``Combustible gas control for nuclear power reactors.'' The requested 
exemption from 10 CFR 50.44 is not being considered further by the NRC 
staff because revisions were made to 10 CFR 50.44 (68 FR 54123; 
September 16, 2003), such that it does not refer to specific types of 
zirconium cladding, thus removing the need for such an exemption.

3.0 Discussion

    Pursuant to 10 CFR 50.12, the Commission may, upon application by 
any interested person or upon its own initiative, grant exemptions from 
the requirements of 10 CFR Part 50, when (1) the exemptions are 
authorized by law, will not present an undue risk to public health or 
safety, and are consistent with the common defense and security; and 
(2) when special circumstances are present. Under Section 50.12(a)(2) 
of 10 CFR, special circumstances include, among other things, when 
application of the specific regulation in the particular circumstance 
would not serve, or is not necessary to achieve, the underlying purpose 
of the rule.

Authorized by Law

    This exemption would allow the licensee to re-insert one LTA 
containing either all Optimized ZIRLOTM fuel rod cladding or 
a combination of Optimized ZIRLOTM and AXIOMTM 
fuel rod cladding that does not meet the definition of Zircaloy or 
ZIRLOTM as specified by 10 CFR 50.46, and Appendix K, into 
the core of VCSNS

[[Page 14854]]

during fuel cycle 18. As stated above, 10 CFR 50.12 allows the NRC to 
grant exemptions from the requirements of 10 CFR Part 50. The NRC staff 
has determined that granting of the licensee's proposed exemption will 
not result in a violation of the Atomic Energy Act of 1954, as amended, 
or the Commission's regulations. Therefore, the exemption is authorized 
by law.

No Undue Risk to Public Health and Safety

    In regard to the fuel mechanical design, the SCE&G exemption 
request relates solely to the specific types of cladding material 
specified in the regulations. No new or altered design limits for 
purposes of 10 CFR 50, Appendix A, General Design Criterion 10, 
``Reactor Design'', need to be applied or are required for this 
program. Following VCSNS Cycle 17, post-irradiation examinations (PIE) 
will be completed on the LTAs to verify acceptable performance and to 
validate fuel performance model predictions. These models, tuned to the 
latest PIE data, will be used to ensure that all design criteria are 
satisfied up to the projected end of cycle 18 (EOC18) burnup. The 
licensee states that if either the PIE shows anomalous behavior or 
predicted performance is outside acceptable bounds, the LTA will not be 
inserted into Cycle 18. Based upon the limited number of advanced alloy 
fuel rods, the PIE (which would detect anomalous behavior), and the use 
of approved models (tuned to the latest PIE data) to ensure that all 
design criteria remain satisfied, the NRC staff finds the LTA 
mechanical design acceptable for VCSNS. In regard to the core reload 
and accident analysis, the NRC staff finds that, based on current LTA 
performance and testing to date, it is not anticipated that any of the 
advanced cladding fuel rods would fail during normal operation and 
anticipated operational events. In the event of unforeseen failures in 
this limited population, plant instrumentation is capable of detecting 
increased reactor coolant activity, and reasonable operator action 
would ensure TS limits would not be violated. Further, due to their 
limited number, failure of the advanced alloy fuel rods during an 
accident would neither challenge docketed dose consequences nor 
coolable geometry. The licensee will continue to use approved core 
physics and reload methodologies to model the LTA up to the projected 
EOC18 burnup. The NRC staff finds the use of these methods acceptable.
    The licensee stated in its May 31, 2007 letter that the assessment 
contained in Westinghouse Commercial Atomic Power-12610-P-A, ``VANTAGE 
+ Fuel Assembly Reference Core Report,'' dated April 1995, concluded 
that the fuel handling accident (FHA) thyroid doses are not adversely 
affected by extended burnup. However, the amount of fission gas release 
(from the fuel pellet) is sensitive to burnup and power history. As 
such, the fission product gap inventory may be affected by the higher 
burnup and power history of the LTA. The NRC staff requested additional 
information (RAI) regarding the limited empirical database of fission 
gas measurements at 75 GWd/MTU burnups, to be able to verify that the 
FHA dose analysis is not impacted. The licensee's October 11, 2007 
response identified a number of conservatisms within the existing dose 
calculations which, if credited, could result in a significant 
reduction in the limiting FHA dose for the extended burnup LTA and thus 
compensate for the uncertainty in fission product gap inventory within 
the high burnup LTA rods. These included the pool decontamination 
factor, the relative power factor for this particular LTA in fuel cycle 
18, the thyroid dose conversion factors, offloading time, reactor 
building purge isolation, and mechanical fuel damage due to impact. 
Consistent with Regulatory Guide (RG) 1.25, ``Assumptions Used for 
Evaluating the Potential Radiological Consequences of a Fuel Handling 
Accident in the Fuel Handling and Storage Facility for Boiling and 
Pressurized Water Reactors (Safety Guide 25),'' an overall effective 
decontamination factor of 100 is used in the current analysis to 
determine the percentage of iodine activity within the fuel rod gap 
that is released to the reactor building atmosphere. As described in 
the UFSAR Section 15.4.5.1.2.2, this value is a factor of five or more 
below the expected value. The licensee stated that although not fully 
credited, this conservatism is recognized in Appendix B to RG 1.195, 
``Methods and Assumptions for Evaluating Radiological Consequences of 
Design Basis Accidents (DBA) at Light-Water Nuclear Power Reactors'', 
which outlines an acceptable methodology for evaluating the 
radiological consequences of a FHA. Provided the depth of the water 
above the damaged fuel is 23 feet or greater, the accepted 
decontamination factors for the elemental and organic species of iodine 
are 400 and 1, respectively, giving an overall effective 
decontamination factor of 200 (i.e., 99.5 percent of the total iodine 
release from the damaged rods is retained by the water). The NRC staff 
confirms that VCSNS Technical Specifications (TSs) 3.7.10 and 3.9.7 
require the water level to be a minimum of 23 feet for the spent fuel 
pool and the reactor vessel during refueling, respectively. Because of 
these controls, the NRC staff is confident that the overall effective 
decontamination factor will not increase above 200. If the RG 1.195 
overall effective decontamination factor is credited within the VCSNS 
FHA analysis, the calculated thyroid dose would decrease by 50 percent. 
The NRC staff finds that the licensee has appropriately applied RG 
1.195, Appendix B, and that this conservatism exists in the current 
licensing basis FHA analysis.
    The licensee presented information showing that the relative 
assembly power factor for both the LTA and the assembly impacted by the 
LTA during an FHA will not approach the 1.7 peaking limit assumed in 
the VCSNS FSAR analysis. The assumptions in RG 1.195 are conservative 
to account for the fact that in a general analysis, it is unknown which 
assembly out of any assembly in the core may be dropped. Therefore, the 
highest peaking factor out of all the assemblies in the core and the 
highest burnup out of all the assemblies in the core are assumed to be 
applied in the same postulated dropped assembly. One assembly would be 
unlikely to have both the highest burnup and the highest peaking 
factor. Therefore, in this specific case, with more realistic and 
appropriate relative assembly powers credited for both the LTA and 
other potentially impacted assemblies, the licensee states the limiting 
dose would decrease by approximately 37 percent. Although relative 
assembly powers are not generally credited in DBA radiological 
consequences analyses, the NRC staff finds that the specific situation 
described above does show that conservatism exists in the current 
licensing basis FHA analysis when compared to the expected impact of 
dropping the extended burnup LTA.
    As regards the thyroid dose conversion factors, the current VCSNS 
dose analysis for the FHA is conservatively based on thyroid dose 
conversion factors from ``Calculation of Distance Factors for Power and 
Test Reactor Sites,'' TID-14844, March 1962. If conversion factors from 
International Commission on Radiation Protection, ICRP-30, ``Limits for 
Intakes of Radionuclides by Workers,'' 1980, were used instead, the 
licensee states that this would result in approximately a 29 percent 
reduction in the limiting dose. Use of ICRP-30 thyroid dose conversion 
factors is acceptable to the staff as documented in RG 1.195. The NRC 
staff accepts that this conservatism exists in

[[Page 14855]]

the current licensing basis FHA analysis.
    For LTA offloading time, the licensee discussed the additional 
decay time that would be expected for the movement of the extended 
burnup LTA as compared to the DBA dose analysis assumption. The VCSNS 
TSs allow a core offload to begin no sooner than 72 hours after 
shutdown. The licensee presented a basis for concluding that, in actual 
practice, core offload would begin no sooner than 144 hours, which 
would further reduce the radiological doses from a DBA. However, 
because the licensee did not provide how it would control the expected 
144 hours to start core offload (i.e. TS, procedural change, etc.), the 
NRC staff finds that this conservatism can not be credited. Following a 
postulated accident inside the reactor building, the radioactivity is 
assumed to be released to the environment through the reactor building 
purge system, and if the system isolates before release to the 
environment, it likely would significantly reduce the FHA dose. 
However, since the system is not fully safety grade, the staff finds 
that this conservatism can not be credited in this analysis.
    As regards the mechanical fuel damage due to an FHA, the VCSNS FSAR 
analysis assumes all rods of the dropped assembly and 50 rods on an 
impacted assembly fail. The licensee states that this is a very 
conservative assumption given the broad spectrum of loads (e.g., 
shipping, thermal, deadweight, loss-of-coolant accident, and seismic 
loads) considered and the resulting high structural strength of the 
fuel assembly and other core components. In its October 11, 2007, RAI 
response, the licensee stated that the irradiated fuel assembly drop 
events have also yielded no increase in local area dose rates. The NRC 
staff agrees with the licensee that the amount of assumed cladding 
failure per RG 1.195 guidance is intended to be generally conservative, 
based on industry experience, but it is not expected to be any more or 
less conservative for the extended burnup LTA than for any other type 
of fuel.
    Contingent on these conservatisms being applicable only to the one 
LTA, the NRC staff finds that the acceptable conservatisms identified 
do compensate for the uncertainties in the gap fractions. Therefore, 
the fission product gap inventory assumed in the current licensing 
basis FHA radiological assessment remains bounding for the extended 
burnup LTA.
    For accidents other than FHA, even though extended burnup to 75 
GWD/MTU for the one LTA would cause a variation in the core inventory 
compared to the current fuel, there are no significant increases to 
isotopes that are major contributors to accident doses. Thus, the NRC 
staff finds that current licensing basis DBA results remain bounding 
for estimated offsite and control room operator doses and the radiation 
dose limitations of Part 100 and GDC-19 will not be exceeded. The NRC 
staff finds that the licensee used assumptions, inputs, and methods 
that are consistent with the conservative regulatory requirements and 
guidance identified above. Based on the VCSNS current licensing bases, 
and the acceptable conservatisms discussed above, the NRC staff finds 
with reasonable assurance, that the licensee's estimates of the 
exclusion area boundary, low-population zone, and control room doses 
will continue to comply with the applicable regulatory criteria. 
Therefore, the proposed extension of the fuel rod average burnup limit 
for one LTA is acceptable with regard to the radiological consequences 
of postulated design basis accidents.
    The underlying purpose of 10 CFR 50.46 is to establish acceptance 
criteria for ECCS performance. The applicability of these ECCS 
acceptance criteria has been demonstrated by Westinghouse. Ring 
compression tests performed by Westinghouse on Optimized 
ZIRLOTM (documented in Appendix B of Addendum 1 to WCAP-
12610-P-A) demonstrate an acceptable retention of Post-LOCA ductility 
up to 10 CFR 50.46 limits of 2200 degrees Farenheit and 17 percent 
equivalent cladding reacted (ECR). Based on an ongoing LOCA research 
program at Argonne National Laboratory, cladding corrosion has a more 
significant impact on post-quench ductility than fuel rod burnup. The 
oxidation measurements provided by the licensee illustrate that the 
oxide thickness (and associated hydrogen pickup) for an LTA up to 75 
GWd/MTU would be below the measured oxide for both Zircaloy-4 and 
ZIRLOTM at current burnup limits. Hence, the effect of 
corrosion on the LTA fuel rods up to the higher burnup would not 
invalidate the applicability of the ECCS acceptance criteria for 
Optimized ZIRLOTM. Due to their limited number, any change 
in the Post-LOCA ductility characteristics of the advanced alloy fuel 
rods (relative to the 2200 degrees Farenheit peak cladding temperature 
and 17 percent ECR) would not challenge core coolable geometry. 
Utilizing currently approved LOCA models and methods, Westinghouse will 
perform cycle-specific reload evaluations to ensure that the LTA 
satisfies 10 CFR 50.46 acceptance criteria. Therefore, the exemption to 
expand the application of 10 CFR 50.46 to include Optimized 
ZIRLOTM is acceptable.
    Paragraph I.A.5 of Appendix K states that the rates of energy, 
hydrogen concentration, and cladding oxidation from the metal-water 
reaction shall be calculated using the Baker-Just equation. Since the 
Baker-Just equation presumes the use of zircaloy clad fuel, strict 
application of the rule would not permit use of the equation for the 
LTA cladding for determining acceptable fuel performance. Metal-water 
reaction tests performed by Westinghouse on Optimized 
ZIRLOTM (documented in Appendix B of Addendum 1 to WCAP-
12610-P-A) demonstrate conservative reaction rates relative to the 
Baker-Just equation. As for the limited advanced alloy fuel rods, their 
similar material composition is expected to yield similar high 
temperature metal-water reaction rates. The reaction rate should not be 
impacted by the higher burnup. Thus, application of Appendix K, 
Paragraph I.A.5, is not necessary to achieve its underlying purpose in 
these circumstances.
    Based upon results of metal-water reaction tests and ring-
compression tests which ensure the applicability of ECCS models and 
acceptance criteria, the limited number and anticipated performance of 
the advanced cladding fuel rods, and the use of approved LOCA models to 
ensure that the LTAs satisfy 10 CFR 50.46 acceptance criteria, the 
staff finds it acceptable to grant an exemption from the requirements 
of 10 CFR 50.46, and Appendix K to 10 CFR Part 50 for the use of an LTA 
up to 75 GWd/MTU in the VCSNS.

Consistent With Common Defense and Security

    The proposed exemption would allow the use of one LTA with advanced 
cladding materials. This change to the plant core configuration has no 
relation to security issues. Therefore, the common defense and security 
is not impacted by this exemption.

Special Circumstances

    Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii), 
are present whenever application of the regulation in the particular 
circumstances is not necessary to achieve the underlying purpose of the 
rule. The underlying purpose of 10 CFR 50.44 is to ensure that means 
are provided for the control of hydrogen gas that may be generated 
following a LOCA. The underlying purpose of 10 CFR 50.46 and Appendix K 
to 10 CFR Part 50 is to establish

[[Page 14856]]

acceptance criteria for ECCS performance. The wording of the 
regulations in 10 CFR 50.46 and Appendix K is not directly applicable 
to these advanced cladding alloys, even though the evaluations 
discussed above show that the intent of the regulations are met. 
Therefore, since the underlying purposes of 10 CFR 50.46 and Appendix K 
are achieved with the use of these advanced cladding alloys, the 
special circumstances required by 10 CFR 50.12(a)(2)(ii) for granting 
of an exemption from 10 CFR 50.46 and Appendix K exist.

4.0 Conclusion

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the exemption is authorized by law, will not present an undue 
risk to the public health and safety, and is consistent with the common 
defense and security. Also, special circumstances are present. 
Therefore, the Commission hereby grants SCE&G exemptions from the 
requirements of 10 CFR 50.46, and 10 CFR Part 50, Appendix K, to allow 
one LTA containing either all Optimized ZIRLOTM fuel rods or 
a combination of Optimized ZIRLOTM and AXIOMTM 
fuel rods to continue to be irradiated up to a burnup of 75 GWd/MTU.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will not have a significant effect on the 
quality of the human environment (73 FR 10069; February 25, 2008).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 13th day of March 2008.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E8-5513 Filed 3-18-08; 8:45 am]
BILLING CODE 7590-01-P