Part 21 Report - 1997-404
ACCESSION #: 9705290350
May 22, 1997
Document Control Desk
ATTN: Chief, Planning, Program and Management Support Branch
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
10 CFR 21 Evaluation and Notification of Adequacy of ATRIUM**[Trademark]
-9 CHF Data Base
This letter is written notification of a reportable defect per 10 CFR
Part 21 reported to the NRC Operations Center by facsimile on May 22,
During NRC Engineering Inspection 9990081/97-01, the NRC determined that
the critical heat flux (CHF) data base for the ATRIUM**[Trademark]-9*_/
fuel design and other 9x9 fuel designs with internal water channels was
not extensive enough to adequately estimate the uncertainties for the
additive constants used in SPC's ANFB CHF correlation. A statistical
treatment of the existing relevant CHF data was developed by SPC to
estimate the uncertainties beyond the original CHF data ranges. These
estimated uncertainties are larger than the original additive constant
uncertainties. Revised safety limit calculations with the larger
additive constant uncertainties indicate that certain plants may have
operated with nonconservative safety limits. Based on these calculations
and a Part 21 evaluation, SPC has concluded the above described situation
represents a defect as defined in 10 CFR 21.3, "A condition ... that
could contribute to the exceeding of a safety limit..."
The affected BWR utilities have been kept informed, and the actions taken
and to be taken to address the issue are provided in the Attachment.
Very truly yours,
H. Donald Curet, Manager
cc: E. Y. Wang (NRR/DRPM/PECB)
*_/ATRIUM is a trademark of Siemens.
Siemens Power Corporation
Nuclear Division 2101 Horn Rapids Road Tel: (509) 375-8100
Manufacturing P.O. Box 130 Fax: (509) 375-8402
Richland, WA 99352-0130
(i) Name and address of the individual informing the Commission.
H. D. Curet, Manager, Product Licensing, Siemens Power
Corporation, 2101 Horn Rapids Road, Richland, WA 99352.
(ii.) Identification of the facility, the activity, or the basic
component supplied for such facility or such activity within
the United States which fails to comply or contains a defect.
The NRC determined the number of test points and the range of
conditions for critical heat flux experiments for the SPC
ATRIUM"-9*_/ fuel design and other 9x9 fuel designs with an
internal water channel are insufficient to justify the
uncertainty values for the additive constants used in the
safety limit determination. The uncertainty in the additive
constant for these fuel designs may be larger than previously
estimated. A larger uncertainty value may affect the safety
limit for BWRs containing reload quantities of SPC 9x9 fuel
designs with internal water channels.
(iii.) Identification of the firm constructing the facility or
supplying the basic component which fails to comply or contains
Siemens Power Corporation, Richland, WA.
(iv.) Nature of the defect or failure to comply and the safety hazard
which is created or could be created by such a defect or
failure to comply.
Statistically estimated additive constant uncertainties, in
lieu of uncertainties quantifiable with CHF data, are larger
than previously estimated for SPC 9x9 fuel designs with
internal water channels. These increased additive constant
uncertainties have resulted in the increase in the safety
limits for some licensees with SPC 9x9 fuel designs with
internal water channels.
(v.) The date on which the information of such defect or failure to
comply was obtained.
During NRC Engineering Inspection 9990081/97-01, the NRC
informed SPC the number of test points and the range of
conditions for critical heat flux experiments for SPC 9x9 fuel
designs with an internal water channel are insufficient to
justify the uncertainty values for the additive constants used
in the safety limit determination. The magnitude of an
estimated additive constant uncertainty was
*_/ATRIUM is a trademark of Siemens.
determined statistically April 17, 1997 using the methodology
submitted to the NRC and documented in ANF-1125, Supplement 1,
Appendix D, "ANFB Critical Power Correlation Uncertainty for
Limited Data Sets," April 1997. Due to the increase in the
estimated value of the uncertainty and its probable impact on
calculated safety limits, a Part 21 evaluation was initiated to
determine if a safety hazard existed or could be created.
(vi.) In the case of a basic component which fails to comply, the
number and the 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.
Safety limit calculations for current reactors (Quad Cities
Unit 2, Cycle 15; LaSalle Unit 2, Cycle 8 and Dresden Unit 3,
Cycle 15) that have received but have not yet operated with SPC
ATRIUM-9 fuel design were performed using the increased
additive constant uncertainty. The safety limit for Quad
Cities increased 0.01. The safety limits for LaSalle and
Dresden were unaffected.
The safety limit previously calculated by SPC for WNP-2, Cycle
11, which has been completed and included all SPC fuel, was
nonconservative in comparison to the safety limit calculated
using the increased additive constant uncertainties. Safety
limit calculations for fuel cycles subsequent to WNP-2, Cycle
11 were not the responsibility of SPC.
(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.
All affected licensees have been informed of the additive
constant uncertainty issue. SPC has submitted to the NRC a
statistical methodology (see item v. above) to estimate the
additive constant uncertainty. Until the NRC approves this
methodology, the statistically estimated increase in the
uncertainty is multiplied by two and added to the original
additive constant uncertainty. The larger uncertainty (i.e.,
increase multiplied by 2) is to be used in future safety limit
calculations until the NRC approves the uncertainty increase
estimated by the submitted statistical methodology or new CHF
data are obtained to quantify a new additive constant
(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.
In addition to providing revised safety limit calculations
where appropriate, SPC has advised its customers that it
intends to obtain additional CHF data to quantify the additive
constant uncertainty. If the new uncertainty is comparable to
the original uncertainty or the increase in the new uncertainty
is less than the presently estimated uncertainty increase
obtained by the statistical methodology, safety limits may be
reduced. Interactions with the NRC will be required to obtain
approval for the use of a new or reduced additive constant
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