United States Nuclear Regulatory Commission - Protecting People and the Environment


ACCESSION #: 9805190348



SIEMENS



May 11, 1998

NRC:98:030



Document Control Desk



ATTN: Chief, Planning, Program and Management Support Branch

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001



Interim Report of Evaluation of a Deviation Pursuant to 10 CFR

21.21(a)(2)



The following information is provided pursuant to the requirements of 10

CFR 21 to submit an interim report on issues that will not be completed

within 60 days of discovery.



An interim report for an issue under evaluation by Siemens Power

Corporation is enclosed:



Interim Report No. 98-003     "Steam Line Break"



Those SPC customers potentially impacted by this issue will be provided a

copy of this interim report.



If you have any questions or if I can be of further assistance, please

call me at (509)375-8757.





Very truly yours,



James F. Mallay, Director

Regulatory Affairs



/arn



Enclosure



cc:  Mr. E. Y. Wang (USNRC)

     Project No. 702



Siemens Power Corporation



Nuclear Division         2101 Horn Rapids Road    Tel:  (509) 375 8100

Engineering &            P.O. Box 130             Fax:  (509) 375 8402

Manufacturing            Richland, WA 99352-0130







                         Interim Report (98-003)



Subject:



Interim report of evaluation of a deviation.  pursuant to 10 CFR

21.21(a)(2)



Title:



Steam Line Break



Identification of Basic Activity:



PWR Steam Line Break Analysis



Basic Activity Supplied by:



Siemens Power Corporation - Nuclear Division



Nature of Deviation:



Three deviations have been identified that are applicable to either the

SPC Steam Line Break methodology or its application to individual plants.



The first deviation is that power distributions which may not bound

potential plant conditions during a Steam Line Break event have been

identified in the Steam Line Break analyses for some plants.  The

inadequate determination of power distributions results from the

following:



1.   variations in cycle to cycle radial peaking factors due to loading

     pattern changes which were not accounted for in the Steam Line Break

     analysis, and



2.   the iteration technique between the XCOBRA-IIIC thermal hydraulic

     code and the XTGPWR neutronics code may not ensure that a

     conservative power distribution is calculated for cases where the

     pumps are turned off.



A second deviation is that the iteration between XCOBRA-IIIC and XTGPWR

may result in non-conservative reactivity calculations with XTGPWR.  This

is a deviation in the SPC Steam Line Break methodology.



A third deviation is that the reactivity control system input for the

ANF-RELAP HFP analyses were constructed for some plants such that a step

change occurs in the Doppler reactivity when switching between two

reactivity models.  This deviation results in an under prediction of the

positive reactivity insertion due to the cool down of the fuel during the

Steam Line Break event.



An evaluation of whether these deviations represent reportable defects

under 10 CFR 21 has been initiated.  A schedule for completing the

evaluation is provided below.



Discovery Date:



March 13, 1998





Corrective Actions to Date:



Condition Report (CR) 6571 was initiated on March 13, 1998 and CR 6599 on

March 21, 1998.



A self-assessment is being performed to assess the deviations identified

in the condition reports and any other issues which may be identified by

a review of the Steam Line Break methodology.  A set of recommendations

from the self-assessment team will be issued during May.  It is apparent

from the initial self-assessment team findings that a change in the

approved Steam Line Break methodology will be necessary to correct the

identified deviation with respect to the iteration between XCOBRA-IIIC

and XTGPWR.



An evaluation of the impact of the combined effects has been made for two

representative plants.  The evaluations indicate that no significant

increase in fuel rod failures is expected from the correction of the

identified deviations.  The criteria for the Steam Line Break event is

the radiological criteria in 10 CFR 100.



Evaluation Completion Schedule Date:



December 1, 1998



*** END OF DOCUMENT ***

Page Last Reviewed/Updated Thursday, March 29, 2012