...

FENOC

by user

on
Category: Documents
2

views

Report

Comments

Description

Transcript

FENOC
FENOC
_MIN%
FrStErgy Nuclear Operating Company
5501 North State Route 2
Oak Harbor, Ohio 43449
419-321-7599
Fax: 419-321-7582
LOw W. Myers
Chief Operating Officer
Docket Number 50-346
License Number NPF-3
Serial Number 2968
August 13, 2003
United States Nuclear Regulatory Commission
Document Control Desk
Washington, D. C. 20555-0001
Subject:
Davis-Besse Nuclear Power Station
Transmittal of Final Report on Examination of the Reactor Vessel Head
Degradation, and Safety Significance Assessment Update
Ladies and Gentlemen:
Enclosed for your information is BWX Technologies, Inc., BWXT Services, Inc. Report
1140-025-02-24, "Examination of the Reactor Vessel (RV) Head Degradation at Davis-Besse,"
June 2003, which was prepared for Framatome ANP, Inc. on behalf of the FirstEnergy Nuclear
Operating Company (FENOC). This report describes the laboratory examinations performed on
a portion of the degraded reactor vessel head and two control rod drive mechanism (CRDM)
nozzles removed from the Davis-Besse Nuclear Power Station, Unit No. 1. The examinations
included visual inspections, dye penetrant testing, scanning electron microscopy, energy
dispersive spectroscopy, metallography, and Knoop microhardness. Detailed dimensional
measurements of the exposed stainless steel cladding area in the Nozzle 3 cavity were also taken.
These examinations were performed in accordance with a detailed plan that was approved by the
NRC staff. Preliminary results of the examinations were previously shared with the NRC staff.
On April 8, 2002 (Serial Number 1-1268), FENOC submitted a safety significance assessment of
the degraded reactor vessel head. On July 20, 2002 (Serial Number 1-1282), FENOC provided
additional information regarding the estimated area sizes of exposed clad material that could
potentially cause the cladding to fail at normal operating pressure. This information was
presented in a calculation performed for FENOC by Structural Integrity Associates (SIA). The
calculation was later updated and the results were provided to the NRC by letter dated
November 18, 2002 (Serial Number 1-1290). The November 18, 2002 letter noted that the
extracted cavity area of the reactor vessel head was under analysis at a laboratory, and that this
analysis would provide additional data on clad thickness, exposed clad area, and cracking. The
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Page 2
November 18 letter further noted that FENOC planned to submit a revision of the safety
significance assessment based on the updated data, when available. As noted above, the BWXT
report formally provides this data.
Based on discussions with SIA, the impact of the differences between the dimensional data
utilized in the safety significance assessment calculation and the more exact data contained in the
BWXT report, even considering the effect of cracking discovered in the exposed clad area
(which was not considered in the calculation), should be minimal, and therefore, a revision to the
safety significance assessment calculation is not warranted. The details of this qualitative
assessment are provided in Enclosure 2.
No formal response to this letter is required or requested. However, should you have any
questions or require additional information, please contact Mr. Kevin L. Ostrowski, Manager Regulatory Affairs, at (419) 321-8450.
Very truly yours,
MKL
Enclosures
cc:
Regional Administrator, NRC Region III
J. B. Hopkins, NRC/NRR Senior Project Manager
C. S. Thomas, NRC Region II, DB- I Senior Resident Inspector
Utility Radiological Safety Board
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure I
EXAMINATION OF THE REACTOR VESSEL (RV) HEAD DEGRADATION
AT DAVIS-BESSE
BWX TECHNOLOGIES, INC.
BWNXT SERVICES, INC.
REPORT 1140-025-02-24
JUNE 2003
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 1
IMPACT OF LABORATORY EXAMINATION RESULTS
ON THE
SAFETY SIGNIFICANCE ASSESSMENT
OF THE
DEGRADED REACTOR VESSEL HEAD
Summary of Existing Analyses and Results
As part of the safety assessment evaluation of the wastage found in the vicinity of CRDM
penetration No. 3, elastic-plastic finite element analyses were performed in References 1, 2, and 3
to determine the design margins in the exposed cladding.
*
*
*
*
Two bounding clad thicknesses (0.297 and 0.125 inch) were considered in this evaluation. The
evaluation assumed a defect-free cladding.
A very conservative stress-strain curve was used in the evaluation. A plot of the stress-strain
curve is shown in Figure 1.
The analysis was based on the geometry of the exposed cavity found by non-destructive
examination (NDE) at the time of the analysis. As stated in Reference 2, in the initial
evaluation in Reference 1, an exposed cladding area of 20.5 sq. inches was modeled. In the
subsequent evaluation in Reference 2, an exposed cladding area twice this value was modeled.
Finally in Reference 3, a range of cladding exposed areas from 20.5 to 82 square inches was
considered.
The results of the evaluation are summarized in Figures 2 and 3 for the two clad thickness
cases considered. As can be seen from these Figures, significant margins exist for the exposed
cladding area of 20.5 square inches even for the cladding thickness of 0.125 inches regardless
of the failure criterion used. As shown in Figures 2 and 3, two failure criteria were used in the
analyses; one based on plastic instability and the other based on the maximum strain limited to
11.15%.
Results of Destructive Examination
Pertinent information from the destructive examination performed by BWX Technologies Inc. in
Reference 4 is summarized below.
*
*
The actual exposed area of the cladding is approximately 16.5 sq. inches.
The actual stress-strain curve of the cladding material is much tougher (measured by the area
under the stress-strain curve) and more ductile (measured by the failure strain) than that used in
the analysis. A comparison of the two curves is shown in Figure 4. It should be noted that the
engineering stress-strain curves from Reference 4 have been converted to true stress-strain
curves for the comparison in Figure 4.
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 2
*
*
Seventy-eight thickness readings were taken of the exposed cladding cavity. The average
thickness was 0.256 inches with a minimum reading of 0.202 inches and a maximum thickness
of 0.314 inches.
Cracking was identified on the exposed surface of the cladding, which followed a mixed
interdendritic/intergranular path. The cracking occurred at the middle of the cavity and
extended a maximum of 0.057 inches below the exposed cladding surface. At the location of
the cracks, the minimum clad thickness is at least 0.236 inches. From the photograph shown in
Figure 3.6.1 of Reference 4, the largest crack was estimated to be approximately 0.5 inches in
length.
Assessment of the Effect of Observed Flaws On Existing Analyses
*
*
*
Because of the ductile nature of the actual cladding material as demonstrated by its high
toughness and ductility, the cladding is expected to fail by net section plastic collapse. As
such, the crack tip stress intensity factor is not of essence and only the remaining ligament of
the cladding is critical to the determination of the failure pressure.
The cladding thickness at the location of the cracks is 0.236 inches and the maximum crack
depth, at its present stage of development, is 0.057 inches, and therefore, the remaining
ligament is 0.179 inches. This thickness is greater than the minimum clad thickness of
0.125 inches analyzed in support of the safety assessment. The results of the analysis of the
0.125 inches cladding should at least bound this actual remaining ligament of the cladding.
This is even conservative considering the fact that the maximum crack length is only limited to
about 0.5 inches in length and therefore covers only a fraction of the exposed area of the
cladding.
The area of the exposed cavity used in the analyses in References 1, 2 and 3 to support the
safety evaluation (20.5 sq. inches) is greater than the actual measured area of 16.5 sq. inches in
Reference 4. On an elastic basis, the failure pressure is expected to be inversely proportional to
the square of the diameter of the exposed cladding and hence, the failure pressure for the actual
exposed area is expected to be at least 20% greater than calculated with the larger area in the
analyses.
Conclusion
The above observations support the conclusion that the observed cracks in the cladding should not
have any major impact on the existing safety evaluation. Because of the ductile nature of the
cladding material, it is expected to fail by net section plastic collapse. The failure pressure is
expected to be higher than the results of the 0.125 inches cladding case considered in the existing
safety evaluation.
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 3
References
1. Structural Integrity Associates Calculation No. W-DB-OIQ-301, Rev. 1, "Elastic-Plastic Finite
Element Analysis of Davis-Besse RPV Head Wastage Cavity."
2. Structural Integrity Associates Calculation No. W-DB-O1Q-302, Rev. 0, "Elastic-Plastic Finite
Element Analysis of Enlarged Davis-Besse RPV Head Wastage Cavity."
3. Structural Integrity Associates Calculation No. W-DB-Ol Q-305, Rev. 1, "Elastic-Plastic Finite
Element Analysis of Davis-Besse RPV Head Wastage Cavity with Different Enlarged Areas
and Thicknesses."
4. BWX Technologies Inc. Document No. 1140-025-02-24, "Examination of the Reactor Vessel
Head Degradation at Davis-Besse," June 2003.
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 4
True Stress-Strain Curve Data for 308/308L Clad Material
(Used in Analyses)
80.0
70.0
60.0
c
.X
50.0
0
on
2 40.0
2 30.0
20.0
10.0
0.0
0%
2%
4%
6%
8%
10%
12%
True Strain
Figure 1. True Stress-Strain Curve for Stainless Steel Type 308/308L Clad Materials
Used in Analyses
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 5
Failure Pressure Versus enlarged Exposed Clad Area (0.125 Inch Clad Thickness)
4000
3500
3000
. 2500
E200
'A* 2000
AL
at
1500
IL
1000
500
0
10
20
30
40
50
60
70
80
90
2
Exposed Clad Area, In
Figure 2. Failure Pressure Versus Cavity Exposed Area (Clad Thickness = 0.125 inches)
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 6
Failure Pressure Versus Enlarged Exposed Clad Area (0.297 Inch Clad Thickness)
8000
7000
F
6000
c 5000
I-
0
01 4000
= 3000
.
AL
2000
-.-11.15% Strain Criteria
-4-Instability
- -Operating Pressure, 2185 psig
. ...
Linear Fit
1000
0
110
20
30
40
50
60
Exposed Clad Area,
70
In
80
90
100
110
2
Figure 3. Failure Pressure Versus Cavity Exposed Area (Clad Thickness = 0.297 inches)
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 2
Page 7
Comparison of True Stress-Strain Curve Data
80.0
70.0
60.0
_x 50.0
00
-
co
*
Analysis
It~~~~~~~~~~~~-*
Test B2C2A1_
Test B2C2A2
h~~~~~~~~~~~~
30.0
_
L..
I-
20.010.0
0.0 0.0000
0.0500
0.1000
0.1500
0.2000
0.2500
0.3000
True Strain
Figure 4. Comparison of Actual Stress-Strain Curve to That Used in the Analyses
Docket Number 50-346
License Number NPF-3
Serial Number 2968
Enclosure 3
COMMITMENT LIST
THE FOLLOWING LIST IDENTIFIES THOSE ACTIONS COMMITTED TO BY THE
DAVIS-BESSE NUCLEAR POWER STATION (DBNPS) IN THIS DOCUMENT. ANY
OTHER ACTIONS DISCUSSED IN THE SUBMITTAL REPRESENT INTENDED OR
PLANNED ACTIONS BY THE DBNPS. THEY ARE DESCRIBED ONLY FOR
INFORMATION AND ARE NOT REGULATORY COMMITMENTS. PLEASE NOTIFY
THE MANAGER - REGULATORY AFFAIRS (419-321-8450) AT THE DBNPS OF ANY
QUESTIONS REGARDING THIS DOCUMENT OR ANY ASSOCIATED REGULATORY
COMMITMENTS.
COMMITMENTS
DUE DATE
None
N/A
Fly UP