Information Notice 2009-23, Nuclear Fuel Thermal Conductivity Degradation
| ML091550527 | |
| Person / Time | |
|---|---|
| Issue date: | 10/08/2009 |
| From: | Brach E, Mcginty T, Tracy G NRC/NMSS/SFST, Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking |
| To: | |
| Powell, E, 301-415-4052 | |
| References | |
| IN-09-023 | |
| Download: ML091550527 (5) | |
ML091550527 UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS
OFFICE OF NEW REACTORS
WASHINGTON, DC 20555-0001
October 8, 2009
NRC INFORMATION NOTICE 2009-23:
NUCLEAR FUEL THERMAL CONDUCTIVITY
DEGRADATION
ADDRESSEES
All holders of operating licenses and construction permits for nuclear power reactors under the
provisions of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing
of Production and Utilization Facilities, except those who have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor vessel. All current
and potential applicants for an early site permit, combined license, or standard design
certification for a nuclear power plant under the provisions of 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. All holders of, and applicants for, a
certificate of compliance for a spent nuclear fuel transportation package under the provisions of
10 CFR Part 71, Packaging and Transportation of Radioactive Material. All holders of a
certificate of compliance for a spent fuel storage cask and all holders of a license for an
independent spent fuel storage installation under the provisions of 10 CFR Part 72, Licensing
Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive
Waste, and Reactor-Related Greater Than Class C Waste.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to notify
addressees of information related to the impact of irradiation on fuel thermal conductivity. The
NRC expects the recipients to review the information within this IN for applicability to their
facilities and consider actions, as appropriate, for their facility. However, suggestions contained
in this IN are not NRC requirements; therefore, no specific action or written response is
required.
DESCRIPTION OF CIRCUMSTANCES
It is well understood that irradiation damage and the progressive buildup of fission products in
fuel pellets result in reduced thermal conductivity of the pellets. However, thermal performance
codes approved by NRC before 1999 did not include this reduction in thermal conductivity with
increasing irradiation because earlier test data were inconclusive as to the significance of the
effect.
Measurements collected from an instrumented assembly at the Halden ultra-high-burnup
experiment during the 1990s have indicated steady degradation in the thermal conductivity of
uranium fuel pellets with increasing exposure. These data indicate a degradation of
approximately 5 to 7 percent for every 10 gigawatt-days per metric tonne of exposure. On the basis of these experimental data, the NRC updated its confirmatory fuel thermal- mechanical performance tool, FRAPCON, to include a new model for predicting fuel thermal
conductivity as a function of exposure. NUREG/CR-6534, Volume 1, FRAPCON- 3:
Modifications to Fuel Rod Material Properties and Performance Models for High-Burnup
Application, issued October 1997, discusses these bases and model updates.
Beginning in 1999, several reactor fuel vendors submitted improved fuel thermal models to the
NRC for review and approval. These new models incorporate updates to the fuel thermal
conductivity models that account for degradation caused by irradiation. The improved vendor
models generally considered experimental qualification data that were substantially similar to
the data considered in NUREG/CR-6534. However, the staff is aware that models that do not
account for the effect of degradation are still used to perform safety analyses.
BACKGROUND
Licensees employ a series of computer codes to analyze plant behavior in the safety analyses
they perform to demonstrate compliance with the Commissions regulations. The computational
approach models various physical processes to predict transient and accident events. These
models simulate reactor conditions for postulated events and compare predicted plant
performance to applicable regulatory criteria.
The simulation of the fuel element is an integral part of the safety analysis. Within the analysis, the fuel pellet thermal conductivity model determines the rate at which heat is transferred from
the fuel pellet, first to the gas gap, then to the fuel cladding, and subsequently to the coolant. A
lower fuel pellet conductivity results in higher fuel temperatures at a given linear heat-generation
rate. Therefore, the analytical prediction of the fuel thermal conductivity will affect the results of
several types of safety analyses. Any codes used for safety analyses that incorporate data
starting at the fuel rod level and generated by the pre-1999 models may mischaracterize the
expected plant performance.
DISCUSSION
General Design Criterion (GDC) 10, Reactor Design, in Appendix A, General Design Criteria
for Nuclear Power Plants, to 10 CFR Part 50, establishes that licensees should not exceed
specified acceptable fuel design limits (SAFDLs) during any condition of normal operation, including the effects of anticipated operational occurrences, to ensure that the fuel is not
damaged. Also, the general requirements to maintain control rod insertability and core
coolability appear in the GDC (e.g., GDC 27, Combined Reactivity Control Systems Capability, and 35, Emergency Core Cooling). In particular, 10 CFR 50.46, Acceptance Criteria for
Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, provides the
specific coolability requirements for a loss-of-coolant accident. In addition, 10 CFR 50.46(a)(3)
specifies requirements for evaluating and reporting each change to, or error discovered in, an
acceptable evaluation model.
Technical specifications require licensees to submit a report on core operating limits that
incorporates the revised cycle-specific parameters resulting from the new core configuration
implemented during the refueling outage. Technical specifications require that the analytical
methods used to determine the core operating limits be those previously reviewed and approved by the NRC. Licensees rely on computer codes for fuel performance calculations and
to perform safety analyses. Within the scope of reload licensing evaluations, they use these
computer codes to establish cycle operating limits to ensure that all applicable requirements
(e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, emergency core cooling
system limits, and nuclear design limits) are met.
If the pre-1999 methods misrepresent fuel thermal conductivity, calculated margins to SAFDLs
and other limits may be less conservative than previously understood.
GENERIC APPLICABILITY
Safety analyses performed for reactors using pre-1999 methods may be less conservative than
previously understood.
Lower fuel pellet conductivity does not appear to significantly influence spent nuclear fuel
cladding temperatures that are typically estimated for aged spent nuclear fuel during dry cask
storage and transportation operations. However, an increase in estimated cladding
temperatures could challenge small thermal margins in the design bases for certified or licensed
spent nuclear fuel storage casks and certified spent nuclear fuel transportation packages.
CONTACT
This IN requires no specific action or written response. Licensees should refer any questions
about this notice to the technical contacts listed below or to the appropriate project manager in
the Office of Nuclear Reactor Regulation. Combined license applicants should refer any
questions about this notice to the technical contact listed below or to the appropriate project
manager in the Office of New Reactors.
/RA by RLorson for/
/RA/
E. William Brach, Director
Timothy J. McGinty, Director
Division of Spent Fuel Storage
Division of Policy and Rulemaking
and Transportation
Office of Nuclear Reactor Regulation
Office of Nuclear Material Safety
and Safeguards
/RA/
Glenn Tracy, Director
Division of Construction Inspection
and Operational Programs
Office of New Reactors
Technical Contacts: Anthony J. Mendiola, NRR
301-415-1054
301-415-1296
E-mail: Anthony.Mendiola@nrc.gov E-mail: Peter.Yarsky@nrc.gov
301-415-1193
E-mail: Joseph.Donoghue@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
CONTACT
This IN requires no specific action or written response. Licensees should refer any questions
about this notice to the technical contacts listed below or to the appropriate project manager in
the Office of Nuclear Reactor Regulation. Combined license applicants should refer any
questions about this notice to the technical contact listed below or to the appropriate project
manager in the Office of New Reactors.
/RA by RLorson for/
/RA/
E. William Brach, Director
Timothy J. McGinty, Director
Division of Spent Fuel Storage
Division of Policy and Rulemaking
and Transportation
Office of Nuclear Reactor Regulation
Office of Nuclear Material Safety
and Safeguards
/RA/
Glenn Tracy, Director
Division of Construction Inspection
and Operational Programs
Office of New Reactors
Technical Contacts: Anthony J. Mendiola, NRR
301-415-1054
301-415-1296
E-mail: Anthony.Mendiola@nrc.gov E-mail: Peter.Yarsky@nrc.gov
301-415-1193
E-mail: Joseph.Donoghue@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
ADAMS Accession Number: ML091550527
TAC No. ME1392 OFFICE
TECH EDITOR
BC:SNPB:DSS
D:DSS
NAME
EPowell
KAzariah-Kribbs(e-mail)
AMendiola
WRuland
DATE
7/28/09
8/6/09
7/29/09
8/3/09 OFFICE
BC:DSRA:NRO RO
D:DSRA:NRO
PGCB:DPR
PGCB:DPR
NAME
JDonoghue (e-mail)
CAder (e-mail)
DBeaulieu
CHawes
DATE
10/1/09 e-mail
10/01/09
10/2/09
10/05/09 OFFICE
BC:PGCB:DPR
D:DPR
D:DCIP
D:DSFST
NAME
MMurphy
TMcGinty
GTracy
EWBrach (Lorson for)
DATE
10/06/09
10/07/09
10/06/09
10/07/09 OFFICIAL RECORD COPY