United States Nuclear Regulatory Commission - Protecting People and the Environment

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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