[Federal Register Volume 73, Number 54 (Wednesday, March 19, 2008)]
[Notices]
[Pages 14853-14856]
From the Federal Register Online via the Government Publishing 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