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. 
IN 2009-23 

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On the basis of these experimental data, the NRC updated its confirmatory fuel thermalmechanical 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 Commission’s 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 
IN 2009-23 

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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. 
IN 2009-23 

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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 Peter Yarsky, NRR 

 301-415-1054 301-415-1296 

 E-mail: Anthony.Mendiola@nrc.gov E-mail: Peter.Yarsky@nrc.gov

 Joseph Donoghue, NRO 

 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. 
IN 2009-23 

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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 Peter Yarsky, NRR 

 301-415-1054 301-415-1296 

 E-mail: Anthony.Mendiola@nrc.gov E-mail: Peter.Yarsky@nrc.gov

 Joseph Donoghue, NRO 

 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 NRO 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