Part 21 Report - 1996-561
ACCESSION #: 9610240239
ABB
October 18, 1996
LD-96-045
Document Control Desk
U. S. Nuclear Regulatory Commission
Washington, D.C. 20555
Subject: Report of a Defect Pursuant to 10 CFR 21 Regarding the
Application of Certain Aspects of ABB-CE Safety Analysis
Methodology
Dear Sir:
The purpose of this letter is to notify the Nuclear Regulatory Commission
of a defect, as defined in 10 CFR 21 - "Reporting of Defects and
Noncompliance". The identified defect involves the application of
certain aspects of the ABB Combustion Engineering (ABB-CE) reload safety
analysis methodology.
Specifically, the identified defect concerns the screening methodology
used by ABB-CE to assess the continued conservatism and applicability of
the DNBR probability distribution function (pdf) from cycle-to-cycle.
This assessment in turn provides assurance that the fuel DNBR Specified
Acceptable Fuel Design Limit (SAFDL) will not be violated. It is
important to point out that the current ABB-CE thermal-hydraulic
Statistical Combination of Uncertainties (SCU) methodology used to
explicitly determine the pdf input to the calculation of the DNBR SAFDL
is not in question. That is, if the SCU analysis methods are explicitly
applied to determine the pdf input to the DNBR SAFDL calculation, the SCU
pdf input to the DNBR SAFDL calculation is correct. Rather, the defect
involves the evaluation criterion used during cycle specific screening to
determine whether or not an explicit SCU pdf re-analysis is required to
verify the continued conservatism during the upcoming cycle for the DNBR
SAFDL of record. Based upon an evaluation, ABB-CE has concluded that the
above described situation represents a defect as defined in 10 CFR 21.3,
i.e., "A condition or circumstance ... that could contribute to the
exceeding of a safety limit...". In accordance with 10 CFR 21.21 (c)(4),
the Enclosure to this letter summarizes the information available to ABB-
CE at this time.
ABB Combustion Engineering Nuclear Power
Combustion Engineering, P.O. Box 500 Telephone (860) 688-
Inc. 200 Day Hill Rd. 1911
Windsor, CT 06095- Fax (860) 285-5203
0500
Document Control Desk LD-96-045
18 October, 1996 Page 2
If you have any questions, please feel free to contact me or Mr. Chuck
Molnar of my staff at (203) 285-5205.
Very truly yours,
COMBUSTION ENGINEERING, INC.
Dr. Ian C. Rickard, Director
Operations Licensing
Enclosure: As stated
cc: M.A. Barnoski (ABB-CE)
Enclosure to LD-96-045
Page 1 of 4
ABB Combustion Engineering Nuclear Operations
10 CFR 21 Report of a Defect or Failure to Comply
The following information is provided pursuant to the requirements
identified in 10 CFR 21.21 (c)(4):
(i) Name and address of the individual(s) informing the Commission:
Dr. I. C. Rickard, Director
Operations Licensing
ABB Combustion Engineering Nuclear Operations
2000 Day Hill Road
Windsor, CT 06095-0500
(ii) Identification of the facility the activity or the basic component
supplied or such facility or such activity within the United States
which fails to comply or contains a defect:
The activity for which this report is being filed is the DNBR
probability distribution function (pdf) safety analysis
screening methodology used by ABB-CE to verify the continued
conservatism of the analysis of record for those plants for
which ABB-CE performs reload safety analyses.
The current ABB-CE thermal-hydraulic Statistical Combination of
Uncertainties (SCU) methodology used to determine the pdf input
to the calculation of the DNBR Specified Acceptable Fuel Design
Limit (SAFDL) is not in question. If the SCU analysis methods
are applied to determine the pdf input to the DNBR SAFDL
calculation, the SCU pdf input to the DNBR SAFDL calculation
would have no safety concern. Rather, the reported defect
involves the evaluation criterion used during cycle specific
screening to determine whether explicit SCU pdf re-analysis is
required to verify the continued conservatism of the DNBR SAFDL
of record for the upcoming cycle.
The SAFDL on DNBR currently contains an allowance which
accounts for uncertainties in parameters (e.g. inlet flow, rod
diameter, calculation uncertainty, rod pitch, etc.) which are
not explicitly included in the DNBR on-line calculation for
digital plants and the setpoints for analog plants. The
effects of all these uncertainties are statistically combined
according to SCU methodology to produce a pdf which is used to
determine the DNBR SAFDL. The pdf is calculated using the
sensitivities of DNBR to various parameters. The most
significant component sensitivity is the sensitivity of DNBR to
the inlet flow in the limiting assembly.
The current screening methodology for assessing the continued
conservatism of the DNBR pdf is based on a two stage screening
process. In the first stage, the limiting assembly inlet flow
factor (and the flow factors in the four adjacent assemblies to
the limiting assembly) and pin-to-node power ratios for the
upcoming cycle are compared with the corresponding values from
the Analysis Of Record (AOR). If the limiting
Enclosure to LD-96-045
Page 2 of 4
assembly inlet flow factors and pin-to-node power ratios are
higher for the upcoming cycle than the corresponding values
from the AOR, no further DNBR pdf analysis is performed and the
AOR DNBR pdf is concluded to remain applicable for the upcoming
cycle. If the limiting assembly inlet flow factors and/or pin-
to-node power ratios for the upcoming cycle are lower than the
corresponding values from the AOR, the second screening
criterion is applied. The second criterion compares the
CETOP/TORC penalty factors derived as part of the 4-pump TORC
analysis. If the reference cycle CETOP/TORC penalty factors
are lower than those for the upcoming cycle over the range of
interest for the upcoming cycle, then the pdf calculated using
the reference cycle TORC model is concluded to continue to be
bounding. If the above described screening process concludes
that the AOR DNBR pdf may not remain applicable, explicit TORC
analyses are performed to either establish the applicability of
the AOR pdf to the upcoming cycle or calculate a pdf that does
apply to the upcoming cycle.
Recent preliminary calculations for Palo Verde-1 Cycle 7 cast
doubt on the sufficiency of this screening methodology for
confirming the applicability of the DNBR pdf for all core
designs. Specifically, explicit TORC calculations for a core
design which was not ultimately chosen for the upcoming cycle
demonstrated that the pdf, and hence DNBR SAFDL, was not
bounded by the reference case, even though the reference
CETOP/TORC penalty factors were demonstrated to be conservative
using the screening methodology described above.
Further evaluation has shown that the screening methodology may
not be sufficient for more recent fuel managements that exhibit
the characteristic of a very flat assembly power distribution,
which forces the limiting or "hot" subchannel more interior to
the assembly. A review of the thermal-hydraulic models and SCU
methodology, particularly for the more recent fuel managements
exhibiting the characteristics identified above, has shown that
the models and methodology are sound, except for the screening
methodology. As such, this situation represents a condition of
circumstance that could contribute to exceeding the DNBR safety
limit defined by plant technical specifications.
It should be noted that although the identified defect may
impact the DNBR SAFDL for specific core designs, it does not
imply that the margin to the SAFDL is non-conservative in all
cases. For instance, conservatism in the CETOP/TORC penalties
in the AOR may compensate and provide conservative margin to
the DNBR SAFDL.
(iii) Identification of the firm constructing the facility or supplying
the basic component which fails to comply or contains a defect.
Combustion Engineering, Inc.
2000 Day Hill Road
Windsor, CT 06095-0500
(iv) Nature of the defect or failure to comply and the safety hazard
which is created or could be created by such defect or failure to
comply:
Enclosure to LD-96-045
Page 3 of 4
The defect identified involves a deficiency in the screening
methodology used to verify the continued conservatism of the
DNBR pdf for the safety analysis of record such that a non-
conservative conclusion may be reached. As such, this
situation represents "a condition or circumstance ... that
could contribute to the exceeding of a safety limit as defined
in the defect definition of 10 CFR 21.3.
(v) The date on which the information of such defect or failure to
comply was obtained:
ABB-CE determined on 17 October, 1996 that a defect as defined
in 10 CFR 21.3 did in fact exist.
(vi) In the case of a basic component which contains a defect or fails to
comply, the number and location of all such components in use at,
supplied for, or being supplied for one or more facilities or
activities subject to the regulations in this part.
The defect applies to the DNBR pdf safety analysis screening
methodology used by ABB-CE to verify the continued conservatism
of the analysis of record for those plants for which ABB-CE
performs reload safety analyses. This screening methodology
was applied to ABB-CE designed nuclear power plants for which
ABB-CE continues to provide fuel and thermal-hydraulic safety
analyses utilizing the SCU methodology. Specifically, the
nuclear power plants potentially affected by this defect
include;
Arkansas Nuclear One Unit 2
Calvert Cliffs Units 1 and 2
Palo Verde Units 1,2, and 3
San Onofre Units 2 and 3
St. Lucie Unit 2, and
Waterford Steam Electric Station Unit 3
For these plants, ABB-CE has determined that the AOR for all
except Palo Verde Unit 3, Cycle 6 remain bounding and no
compensatory measures are required to avoid violation of the
DNBR safety limit. For Palo Verde Unit 3, Cycle 6 operation it
was ascertained that the screening methodology did in fact
result in a potentially nonconservative DNBR SAFDL value.
Compensatory measures were identified and provided to the
utility (Arizona Public Services) to avoid exceeding the DNBR
safety limit. It is ABB-CE's understanding that Arizona Public
Services has taken the steps necessary to implement the
compensatory measures provided.
For the ABB-CE designed Omaha Public Power District (OPPD) Ft.
Calhoun nuclear power station, although ABB-CE is not currently
the fuel and/or safety analysis vendor, ABB-CE safety analysis
methodologies are employed by OPPD to perform its own safety
analyses in-house. ABB-CE has insufficient information
available to make a meaningful determination of applicability
in this case. As such, OPPD has been informed of the condition
reported herein advising that an applicability evaluation of
this situation is warranted pursuant to 10 CFR 21.
Enclosure to LD-96-045
Page 4 of 4
ABB-CE designed nuclear power plants, for which ABB-CE is not
currently the fuel and/or safety analysis vendor and,
therefore, which are not affected by the identified defect
include;
Millstone Unit 2
Maine Yankee
Palisades
St. Lucie Unit 1
(vii) The corrective action which has been, is being, or will be taken;
the name of the individual or organization responsible for the
action; and the length of time that has been or will be taken to
complete the action:
The defective screening methodology has been replaced by a
revised screening methodology which reduces the degree of
engineering judgment applied and imposes specific guidelines
which must be satisfied prior to use of the revised screening
methodology. If the specified guidelines cannot be satisfied,
the SCU analysis methods will be explicitly applied to
determine the pdf input to the DNBR SAFDL calculation.
(viii) Any advice related to the defect or failure to comply about the
facility, activity, or basic component that has been, is being,
or will be given to purchasers or licensees:
As mentioned in item (vi), ABB-CE has informed Arizona Public
Services of compensatory actions to assure that the Palo Verde
Unit 3, Cycle 6 plant operation does not exceed the DNBR SAFDL.
Additionally, ABB-CE has notified OPPD of its inability to
evaluate this situation for the Ft. Calhoun nuclear power
station and advising that they evaluate their situation
pursuant to 10 CFR 21.
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