United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 97-15: Reporting of Errors and Changes in Large-Break/Small-Break Loss-of-Coolant Evaluation Models of Fuel Vendors and Compliance with 10CFR50.46(a)(3)

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                         WASHINGTON, D.C.  20555-0001

                                 April 4, 1997

                                 BREAK LOSS-OF-COOLANT ACCIDENT EVALUATION
                                 MODELS OF FUEL VENDORS AND COMPLIANCE WITH
                                 10 CFR 50.46(a)(3)


All holders of operating licenses or construction permits for nuclear power
reactors and all reactor fuel vendors.


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees about two recent staff findings related to the
review of large-break (LB) loss-of-coolant accident (LOCA) emergency core
cooling system (ECCS) analysis evaluation model changes and also to remind
licensees and reactor fuel vendors of the requirements contained in Section
50.46(a)(3) of Title 10 of the Code of Federal Regulations [10 CFR
50.46(a)(3)] concerning the reporting of ECCS cooling model changes and
errors.  It is expected that recipients will review the information for
applicability to their facilities and consider actions, as appropriate, to
avoid similar problems.  However, suggestions contained in this information
notice are not NRC requirements; therefore, no specific action or written
response is required.

Description of Circumstances

Recently identified changes and errors in Siemens Power Corporation (SPC,
formerly Exxon Nuclear) and General Electric (GE) LBLOCA analysis models have
led to a series of 30-day reports and 10 CFR 50.72 reports as required by
10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for
Light Water Nuclear Power Reactors." 

SPC LBLOCA ECCS Evaluation Model Changes

The SPC LBLOCA ECCS model, TOODEE2, was approved by the NRC staff to meet the
requirements of 10 CFR 50.46 in a letter dated July 8, 1986 [Accession number
8607150319], from D. M. Crutchfield (NRC) to G. Ward (Exxon).  In 1991, SPC
had made changes to the NRC-approved fuel cooling test facility (FCTF) reflood
heat transfer coefficient correlation used in TOODEE2.

During August 1995, the NRC met with SPC about the LBLOCA ECCS evaluation
model.  As a result of that meeting, the staff sent a letter to SPC, dated
November 13, 1995 [9511150211], that identified problems concerning changes in
the TOODEE2 computer code 

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specifically related to the 1991 changes to the NRC-approved FCTF reflood heat
transfer coefficient correlation and the significance of the code changes. 
The staff then requested in a letter dated March 13, 1996 [9603150002], that
SPC formally submit to the staff for its review and approval all model
revisions and corrections implemented in TOODEE2 since the staff's approval of
the code in July 1986.

On June 2, 1996, SPC submitted topical report XN-NF-82-20, "EXEM/PWR Large
Break LOCA ECCS TOODEE2 Updates," Revision 1, Supplement 5 [9606260239], which
described the updates made in the TOODEE2 computer code between 1986 and 1991. 
TOODEE2 is part of the evaluation model used by SPC for pressurized-water
reactors.  The staff has completed its review of this report and has concluded
that the proposed LBLOCA-ECCS model (i.e., the 1991 model) is not acceptable
and the previously approved model (i.e., the 1986 model) contains an
unacceptable error.  This information was formally communicated to SPC in a
safety evaluation enclosed in a letter dated November 29, 1996 [9612040294].

After concluding that the 1991 model was unacceptable, the staff met with SPC
and those licensees using SPC's LBLOCA evaluation model on October 16, 1996,
to discuss the unacceptable error in the 1986 model.  The staff also requested
and received information from the licensees that demonstrated that they were
in compliance with 10 CFR 50.46 (see meeting summary dated November 5, 1996

Public Service Electric & Gas (PSE&G) Audit of GE 

During a recent licensee-conducted quality assurance (QA) audit of the fuel
vendor (GE - Wilmington, North Carolina), PSE&G, the licensee of Hope Creek
Nuclear Generating Station, identified a weakness in GE's tracking of errors
and changes in the LOCA evaluation models.  Between 1990 and 1995, information
sent to the licensee indicated that there had been no known impact on the
calculated peak cladding temperature (PCT).  Earlier in 1996, two impacts had
been reported by GE to the licensee and when reviewing the handling of this
information during the audit, three additional impacts not previously reported
to the licensee were discovered, dating back to 1990, 1992, and 1993.  In
addition, the audit determined that GE had not been tracking the cumulative
impact of errors and changes on the PCT as expected by the licensee.  The
cumulative PCT impact was previously known to be 35 �F     (19 �C); however,
on the basis of the errors identified during the audit, the value is now
raised to 100 �F (56 �C) exceeding the 50 �F (28 �C) reporting threshold.  The
licensee's recalculated PCT still remains below the ECCS acceptance criteria
of 2200 �F (1200 �C).

In a letter to the NRC dated February 17, 1997 [9703060067], GE characterized
the licensee-identified weakness as an issue of timeliness of notifications to
utilities of errors and changes in the LOCA evaluation models.  Furthermore,
notification about changes or errors identified during the 1990 to 1995 period
were provided by GE on an annual basis.  Because notification by GE to
boiling-water reactor BWR licensees on individual impacts less than 50 �F (28
�C) were not provided as they occurred, the BWR licensees did not have the
required information to fully comply with the requirements of 10 CFR 50.46
[specifically the requirement to report within 30 days a cumulative PCT impact
greater than 50 �F (28 �C)]. .                                                                IN 97-15
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Although the LOCA analyses are performed by the fuel vendors, licensees are
responsible for compliance with the regulations related to the LOCA analysis,
that is, 10 CFR 50.46(a).  Section 50.46(a)(1)(i) requires licensees to
calculate ECCS cooling performance with an acceptable evaluation model.  The
staff's recent interactions with the licensees using the SPC's LBLOCA
methodology (the review experience of the SPC LOCA evaluation model changes)
and the Hope Creek QA audit indicate that licensees may not be closely
monitoring the work of their respective fuel vendors.  When the error in the
1986 model was discovered and when SPC changed the TOODEE2 code in 1991, the
resulting changes in the PCT were, in some cases, significant, and the
responsible licensees were not aware of the significant changes. 
"Significant" is defined in 10 CFR 50.46(a)(3)(i) as follows:  "a significant
change or error is one which 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 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."

Licensees may not be performing adequate assessments of errors when they are
aware of them.  Furthermore, licensees' audits of SPC's evaluation model
changes appear to have been ineffective in identifying the technical
inadequacy of the changes.  It should be noted that 10 CFR 50.46 allows fuel
vendors or licensees to make evaluation model changes without the staff's
prior approval; however, the licensees are responsible for identifying any
deficiencies in the change process and reporting them to the NRC staff
accordingly.  In addition,  the licensee determines whether the changes are

It also appears that licensees may not be monitoring the cumulative effect of
the evaluation model changes.  In a given year, the impact of the evaluation
model change may be less than 50 �F (28 �C) on the limiting PCT calculated
with the last acceptable model and hence the change is not significant.  But
the impact of the evaluation model changes over several years together can
exceed 50 �F (28 �C) and, therefore, will be reportable as significant.

Section 50.46 places the responsibility for the reporting of evaluation model
changes on the limiting PCT calculated with the last acceptable model on the
licensees.  Some licensees have apparently considered that the annual
notifications sent by the fuel vendor are sufficient to meet the requirements
under 10 CFR 50.46(a)(3)(ii).  Specifically, "the applicant or licensee shall
report the nature of the change or error and its estimated effect on the
limiting ECCS analysis to the Commission at least annually as specified in
.50.4.  If the change or error is significant, the applicant or licensee shall
provide this report within 30 days...."   The notifications submitted by the
fuel vendors will not satisfy these reporting requirements; however, licensees
are allowed to refer to the vendor's annual notifications.  As stated in    
10 CFR Part 50, Appendix B, Section VII, "The effectiveness of the control of
quality by contractors and subcontractors shall be assessed by the applicant
or designee at intervals consistent with the importance, complexity, and
quantity of the product or services."
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In summary, licensees are reminded that to meet the ECCS acceptance criteria
their responsibilities include:

(1) Section 50.46(a)(1)(i) requires licensees to calculate ECCS cooling
    performance with an acceptable evaluation model.

(2) Section 50.46(a)(3)(ii) requires licensees to report changes and/or
    errors and their estimated effects on the limiting ECCS analysis to the
    Commission at least annually, and if the change or error is significant,
    the licensee shall provide this report within 30 days.  

(3) Individual licensees are responsible to assess effectiveness of the
    control of quality of ECCS evaluation models provided by the vendors as
    required by Part 50, Appendix B.  Meaningful technical audits may be
    necessary to meet Appendix B.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.

                                          signed by

                                       Thomas T. Martin, Director
                                       Division of Reactor Program Management
                                       Office of Nuclear Reactor Regulation

Technical contacts:  George Thomas, NRR         Joseph L. Staudenmeier, NRR
                     (301) 415-1814             (301) 415-2869
                     E-mail:  gxt@nrc.gov       E-mail:  jls4@nrc.gov

                     Eric Benner, NRR
                     (301) 415-1171
                     E-mail:  ejb1@nrc.gov
Page Last Reviewed/Updated Tuesday, December 03, 2013