Information Notice 2011-21, Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation
| ML113430785 | |
| Person / Time | |
|---|---|
| Issue date: | 12/13/2011 |
| From: | Laura Dudes, Mcginty T Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking |
| To: | |
| alexion T W, NRR/DPR 415-1326 | |
| References | |
| TAC ME7686 IN-11-021 | |
| Download: ML113430785 (6) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NEW REACTORS
WASHINGTON, DC 20555-0001 December 13, 2011 NRC INFORMATION NOTICE 2011-21: REALISTIC EMERGENCY CORE COOLING
SYSTEM EVALUATION MODEL EFFECTS
RESULTING FROM NUCLEAR FUEL THERMAL
CONDUCTIVITY DEGRADATION
ADDRESSEES
All holders of an operating license or construction permit for a nuclear power reactor under
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 holders of or applicants for an early site permit, standard design certification, standard
design approval, manufacturing license, or combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to notify
addressees of recent information obtained concerning the impact of irradiation on fuel thermal
conductivity, and its potential to cause errors (specifically, higher predicted peak clad
temperature results) in realistic emergency core cooling system (ECCS) evaluation models.
[This IN uses the term error consistent with applicable NRC regulations described below
regarding the effect of any change to or error in an acceptable evaluation model. The
modeling error in this IN was discovered as a result of research that provided new data and
greater understanding of the phenomenon and as such, the word error in this IN is not intended
to convey culpability.] 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
An NRC licensee recently sponsored an analysis to determine the effect that accounting for fuel
thermal conductivity degradation due to irradiation would have on the results of a realistic
emergency core cooling evaluation that it was proposing to implement. The analysis
determined that the effect would be significant in that the predicted peak fuel cladding
temperature from the most severe postulated loss-of-coolant accident would increase by more
than 50 degrees Fahrenheit (°F).
The licensee asked for this analysis during the NRC staffs review of its request to implement
the ASTRUM realistic Westinghouse Electric Company (WEC) ECCS evaluation model. The
licensees analysis addressed issues discussed in IN 2009-23, Nuclear Fuel Thermal
Conductivity Degradation, dated October 8, 2009. As discussed in IN 2009-23, currently
approved fuel performance codes that provide input to realistic ECCS models may not account
for fuel thermal conductivity degradation.
Because the licensee currently uses an evaluation model that treats fuel thermal conductivity
degradation differently than does the evaluation model that it is proposing to use, there is no
immediate safety concern for this licensee. Also, because the results of the licensees analysis
do not cause its proposed results to exceed the acceptance criteria at 10 CFR 50.46(b), the
NRCs review of the licensees request is not affected.
BACKGROUND
IN 2009-23 describes how legacy fuel performance codes may overpredict fuel rod thermal
conductivity at higher burn-ups based on new experimental data.
Since the NRCs issuance of IN 2009-23, the vendors of fuel performance analysis codes, and
the downstream safety analyses that rely on their results, have engaged with the NRC through
numerous public meetings and written correspondence. The vendors have adjusted legacy
codes so that the codes correlate better with more recent fuel performance data. The vendors, and the NRC licensees that use these vendors analytic methods, have been working to quantify
the impact of the issue identified in IN 2009-23 on downstream safety analyses.
The operating experience described above indicates that the realistic emergency core cooling
evaluation models developed by WEC, which rely on the Fuel Rod Performance and Design
(PAD) Code for fuel thermal mechanical performance data, are susceptible to errors of similar
magnitude to the plant-specific results described above. As described by 10 CFR 50.46(a)(3)(i),
these errors may be significant. Note that the analytic treatment of fuel burnup differs among
the realistic ECCS evaluation models, and the impact described in this IN pertains to results
obtained using the ASTRUM model.
DISCUSSION
Based on the conclusions reached by this licensees analyses, the NRC believes that correcting
for the effect of this error could cause a number of plant-specific ECCS evaluations to predict
significantly higher peak cladding temperatures that exceed the 10 CFR 50.46(b)(1) acceptance
criterion.
WEC has indicated that the approved evaluation models contain substantial conservatisms that
would more than compensate for this issue, and that as a result, there is no issue of immediate
safety concern. The NRC is currently verifying WECs claims that the evaluation model conservatisms compensate for the ECCS evaluation model error caused by fuel thermal
conductivity degradation.
WEC has stated that the impact of modeling thermal conductivity degradation needs to be
considered in the realistic evaluation model methodology because research has provided new
data and greater understanding of the phenomenon. According to the vendor, explicit
incorporation of thermal conductivity degradation modeling would represent an enhancement to
analytic capabilities and not correction of an error. The NRC staff does not agree with WECs
characterization.
Applicable Regulatory Requirements
In 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water
Nuclear Power Reactors, the NRC provides the requirements for ECCSs and their evaluations;
specifically, 10 CFR 50.46(a)(1)(i) requires the following of realistic ECCS models:
Each boiling or pressurized light-water nuclear power reactor must be provided with
an emergency core cooling system (ECCS) that must be designed so that its calculated
cooling performance following postulated loss-of-coolant accidents conforms to the
criteria set forth in paragraph (b) of this section. ECCS cooling performance must be
calculated in accordance with an acceptable evaluation model
According to 10 CFR 50.46(a)(1)(i), an acceptable evaluation model has the following
characteristics:
the analytical technique realistically describes the behavior of the reactor system
during a loss-of-coolant accident
when the calculated ECCS cooling performance is compared to the criteria set forth in
paragraph (b) of this section, there is a high level of probability that the criteria would not
be exceeded.
According to 10 CFR 50.46(b)(1), the calculated maximum fuel element cladding temperature
shall not exceed 2,200 °F.
Under 10 CFR 50.46(a)(3)(i), the NRC requires licensees to estimate the effect of any change to
or error in an acceptable evaluation model or in the application of such a model to determine
whether the change or error is significant. For the purposes of 10 CFR 50.46, a significant
change or error is one that results in a calculated peak fuel cladding temperature different by
more than 50 °F from the temperature calculated for the limiting transient using the last
acceptable model, or that is a cumulation of changes and errors such that the sum of the
absolute magnitudes of the respective temperature changes is greater than 50 °F. Finally, 10 CFR 50.46(a)(3)(ii) promulgates requirements for reporting estimated changes to or
errors in ECCS evaluation models, or applications thereof, to the Commission, stating the
following:
If the change or error is significant, the applicant or licensee shall provide this report
within 30 days and include with the report a proposed schedule for providing a
reanalysis or taking other action as may be needed to show compliance with §50.46 requirements
Any change or error correction that results in a calculated ECCS performance that does
not conform to the criteria set forth in paragraph (b) of this section is a reportable event
as described in §§ 50.55(e), 50.72, and 50.73. The affected licensee shall propose
immediate steps to demonstrate compliance or bring plant design or operation into
compliance with § 50.46 requirements.
GENERIC IMPLICATIONS
The NRC has initiated conversations with WEC to address the generic implications of this
information. Specifically, these discussions involve other susceptible plants and licensee
requirements as provided in 10 CFR 50.46. The NRC will continue considering the safety and
regulatory aspects of the information.
CONCLUSION
A potentially significant, as described in 10 CFR 50.46(a)(3)(i), ECCS evaluation model error
has been identified. Licensees using WEC realistic ECCS evaluation models may wish to
contact the vendor for assistance in estimating the effect this error may have on plant-specific
ECCS evaluation results.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or to the appropriate project managers in the Office
of Nuclear Reactor Regulation or Office of New Reactors.
/RA/ /RA/
Timothy J. McGinty, Director Laura A. Dudes, Director
Division of Policy and Rulemaking Division of Construction Inspection
Office of Nuclear Reactor Regulation and Operational Programs
Office of New Reactors
Technical Contacts: Benjamin Parks, NRR
301-415-6472 E-mail: Benjamin.Parks@nrc.gov
Yi-Hsiung Hsii, NRO
301-415-2877 E-mail: Yi-Hsiung.Hsii@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library.
ML113430785 *via e-mail **via phone TAC No. ME7686 OFFICE NRR/DSS/SRXB Tech Editor BC:NRR/DSS/SRXB BC:NRR/DSS/SNPB D:NRR/DSS
NAME BParks* JDougherty* AUlses* AMendiola WRuland*
DATE 12/12/11 12/12/11 12/12/11 12/12/11 12/13/11 OFFICE D:NRO/DSA D:RES/DSA LA:PGCB:NRR PM:PGCB:NRR BC:PGCB:NRR
NAME CAder** KGibson CHawes TAlexion SRosenberg
DATE 12/13/11 12/12/11 12/13/11 12/13/11 12/13/11 OFFICE D:DCIP:NRO D:DPR:NRR
NAME LDudes TMcGinty
OFFICE 12/13/11 12/13/11