ML20039F858
| ML20039F858 | |
| Person / Time | |
|---|---|
| Site: | Palisades, Fort Calhoun  |
| Issue date: | 12/31/1981 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML13308A045 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-A, NUDOCS 8201130481 | |
| Download: ML20039F858 (37) | |
Text
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5 U SEENDfXA EVALUATION OF PRESSURIZED THERMAL SH0CK EFFECTS DUE TO SMALL BREAK LOCA'S WITH LOSS Of FEEDWATER FOR THE FORT CALHOUN REACTOR VESSEL Prepared for OMAHA PUBLIC POWER DISTRICT NUCLEAR OWER SYSTEMS DIVISION
=
POWER P. i hi
- SYSTEMS i
COMBUSTION ENGINEERING INC 820113048'l
4 i
LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENCINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:
A.
MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT (TJFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
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. - - ~ -
ABSTRACT This Appendix to CEN-189 provides the plant-specific evaluation of pressurized thermal shock effects due to small break LOCA's with extended loss of feedwater for the Fort Calhoun reactor vessel.
It is concluded that crack initiation would not occur for the transients considered for more than 32 effective full power years, which is assumed to represent full plant life, i
i
CEN-189 Appendix A TABLE OF CONTENTS SECTION TITLE PAGE-ABSTRACT A1.
PURPOSE Al
. A2.
SCOPE Al A3, INTRODUCTION Al A4.
THERMAL HYORAULIC ANALYSES Al AS.
FLUENCE DISTRIBUTIONS A2 A6.
MATERIAL DROPERTIES Als A7.
VESSEL INTEGRITY EVALUATIONS A21 A8.
CONCLUSIONS A32 11
1Al.0 PURPOSE This Appendix provides the plant-specific evaluation of pressurized' thermal shock effects of the SB LOCA + LOFW transients presented in the main body of the CEN-189 report for the Fort Calhoun reactor vessel.
A2.0 SCOPE The scope of this Appendix is limited to the evaluation of the SB LOCA +
LOFW transients presented in CEN-189, as applied to the Fort Calhoun reactor vessel.
Other C-E NSSS reactor vessels are reported in separate Appendices.
A
3.0 INTRODUCTION
This Appendix to CEN-189 was prepared by C-E for Omaha -Public Power District for their use in responding to Item II.K.2.13 of NUREG-0737 for the Fort Calhoun reactor vessel.
This Appendix is intended to be a companion to the CEN-189 report.
The transients evaluated in this Appendix are those reported in Chapter 4.0 of the main report. Chapter A5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter A6 reports the plant-specific material properties and change of properties due to-irradiation, based on the methods of Chapter 6.0 of the report.
Chapter A7 reports
. the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter A6.
A4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to cerform the plant-specific vessel evaluation reported in this Appendix are those reported in Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used which could be eliminated by performing more detailed plant-specific thermal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.
Al
AS. Fort Calhoun Fluence Distribution Omaha Public Power District supplied the reactor power history and detailed radial power distributions needed to update the azimuthal fluence distribu-tion from the time at which the surveillance capsule was removed to December 31, 1981.
l Through December 31, 1981 a cumulative energy generation of 66,655,037 Megawatt hours was quoted as shown in Table A5-1.
Defining full power as 1420 Megawatts-themal (Mwt) this energy output yields 5.36 Effective Full Power Years (EFPY). According to the surveillance capsule analysis (AS-1) the peak fast neutron fluence on the reactor vessel was 3.4 x1018 (n/cm2) after 2.59 EFPY at 1420 Mwt. This implies a rate of accumulation in the 2
18 (n/cm ) per EFPY at 1420 Mwt. As a result the peak fluence of 1.31 x 10 value of the peak wall fluence as of December 31,1981 (5.36 EFPY) is 2
7.04 x 1018(n/cm),
A comparison of the surveillance capsule dosimetry results with the neutron energy spectrum calculated at the surveillance capsule position using a DOT-RO model yielded a calculated value within 4% of that quoted in the surveillance capsule report. Therefore there is good confidence in the surveillance capsule analysis results.
The full power level was cefined as 1420 Mwt because this was the full power level during the exposure experienced by the surveillance capsule and most cf the time period up to December 31, 1981. For extrapolation to future times the full power level is referenced to 1500 tot and an annenximate estimate of the increased fluence accumulation rate is obtained by multiplying the previous rate by the ratio of the power densities (1500/1420 = 1.056).
A summary of the results obtained from the surveillance capsule analysis report is shown in Table A5-2.
TABLE A5-2 Peak Vessel Effective Full Peak Fluence Full Power Level Power Years AccumylationRate Wall F}uence (n/cm )
(Mwt)
EFPY n/cm per EFPY 3.4 x 1018 1420 2.59 1.31 x 1018 7.04 x 1018 1420 5.36 (12/31/81) 1.31 x 1018 A2
The azimuthal shape of the fluence distribution was obtained by updating the 00T-R9 azimuthal distribution corresponding to the surveillance capsule analysis (2.59 EFPY) to December 31, 1981 (5.36 EFPY). The adjustment factors were calculated using the SHADRAC code as described in Section 5.2.2.
The detailed radial power distribution corresponding to 5.36 EFPY was obtained by combining the power distributions supplied by Omaha Public Power District for cycles 4 through 7 with the distribution used to represent the time up to the end of cycle 3 in the surveillance capsule analysis. The nodalization of the power distribution is shown in Figures A5-1 and AS-2.
The detailed power distributions for cycles 4 through 7 are shown in Figures-AS-3 through AS-6.
The resulting azimuthal fluence distribution is shown in Figure A5-7.
The 00 reference point for the azimuthal distribution is shown in Figure A5-8.
One-eighth core symmetry was assumed. The axial and radial fluence distributions in the reactor vessel are obtained from D0T-RZ calculations and are as shown in Figures A5-9 and A5-10, respectively.
The fluence d:stributions were applied as described in the following sections of this report.
References:
A5-1.
Omaha Public Power District Fort Calhoun. Station Unit No.1, Evaluation of Irradiated Capsule W-225, Combustion Engineering, TR-0-MCM.001 Rev. 1, August 1980.
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Fort Calhoun Power History TABLE A5-1 Power Level Cycle EFPD (MW h)
MW-HR t
1 307.56 1420 10,481,698 2
371.58 1420 12,663,429 3
265.08 1420 9,033,881 4
280.65 1420 9,564,550 5
360.18 1420 12,274,999 6
77.05 1420 2,625,937 6
278.07 1500 10,010,543 I
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APPENDIX A FORT CALHOUN A.6 MATERIAL PRCPERTIES The methods used to develop and evaluate the materials for the Fort Calhoun reactor vessel are described in Section 6.0 in the main body of the report. The chemistry data (nickel, copper, and phosphorus content) and initial (pre-irradiation) toughness properties of the reactor vessel shell course plates and welds are summarized in Table A6-1.
In cases where the chemistry exceeded the Regulatory Guide 1.99 prediction limits (0.35%
and 0.012% P), those upper limit values were used in the reference tempera-ture shift calculations.
In cases where the weld metal nickel content was not determined, it was conservatively estimated using information on the type of wire (eg., high MnMo versus MnMcNi wire) or the weld process (inclusion of Ni-200 wire during weld deposition). For the Fort Calhoun weldments, the weld inspection records and welding certification reports indicated that all the welds could be ex-pected to contain high nickel (greater than 0.30 w/o), so the nickel content was conservatively estimated to be 0.99 w/o as indicated in Table A6-1.
The toughness properties given in Table A6-1 are the drop weight NDTT (if determined) and the initial reference temperature, RTNDT. For the plate materials, the RT was determined using transversely oriented Charpy impact NDT specimens or by converting longitudinal impact data using Branch Technical Position MTEB 5-2*.
For the weld material, the RTNDT was estimated using the weld qualification test results benchmarked to the surveillance weld for the vessel, as discussed in Section 6.0 and described below.
The individual weld qualification test results (three Charpy impact specimens tested at +10F) are listed in Table A6-2.
Each weld which exhibited an average Charpy energy of 57 f t-lo or greater (the average Charpy energy for the surveillance weld at 10F) was considered to be at least as tough as the surveillance weld; i.e.,
that weld seam RT was -50F or less. For those NDT weld qualification test results exhibiting an average Charpy energy less than 57 ft-lb, the RINDT was increased by an amount equivalent to the temperature
- " Fracture Toughness Requirements for Older Plants," U.S. Atomic Energy Ccamission, Regulatory Standard Review Plan.
A6-1 A15 1
difference between the average Charpy energy transition curve for the sur-veillance weld and the average Charpy energy for the vessel weld test results.
In effect, the temperature at which 50 ft-lb or better exists was determined, and the RTNDT was astablished at a temperature 60F below that value.
A " map" of the cylindrical portion of the Fort Calhoun reactor vessel is given in Figure A6-1.
It shows the locations of the plates and welds listed in Table A6-1 and their corresponding values of initial RTNDT (F) located within a rectangle on the Figure. RT values for the vertical weld NDT seams (designated 1-410, 2-410, and 3-410) are shown at a single seam but apply to all three vertical seams in a given shell course.
Included in the Figure are the locations of the inlet and outlet nozzles, the core midplane, and the extremities of the active core.
Figure A6-2 is a map of adjusted RTNDT values for important locations at the inner surface of the Fort Calhoun vessel predicted for December 31, 1981.
The predictions are based on the best estimate neutron fluence, 0.704 x 10 n/cm (E>lMeV), (corresponding to 5.36 effective full power years at peak flux location on the inside surface of the reactor vessel), the initial RTNDT and copper, phosphorus, and nickel contents given in Table A6-1, and the normalized neutron flux profiles given in Section A.S.
The values of adjusted NDT (initial RTNDT plus predicted shift) are located in rectangles adjacent RT to the plate and weld designations. The RT values apply to the inner sur-NDT face of the vessel in the region indicated by a circle. The circled regions generally represent areas of peak neutron flux for a given weld seam or plate, i
e A6-2 A16
TABLE A6-1 FORT CALHOUN REACTOR VESSEL MATERIALS Product Material Drop Weight Initial Chemical Content (2)
Form Identification NDIT(F1 RTrlDTJ F_),
N_ic_kel_ ' Copper Phosphorus b
Plate D-4801 -1
-20 0'
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-10'8 O.50 0.ll 0.01 0 Plate D-4801-2
-30 b
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Plate D-4802-1
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c Plate D-4802-2
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Weld 3-410 N/A
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fC N/A Not Available Determined using Branch Technical Position MTEB 5-2 a
i b Estimated based on average for Fort Calhoun plates having reported analyses c Surveillance program data d Estimated (see text and Table A6-2)
Upper bound for coated electrodes (see Table 6-3, Main Report) ef Estimated Ni content (high nickel type wire or weld process)
Requlatory Guide 1.99 upper bound prediction limit g
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TABLE A6-2 FORT Call!0lRI REACTOR VESSEL WELD SEAll TOUGHilESS DATA d
Charpy Qualification Test Results Average Energy Estimated Weld Seam at 10 F (ft-lb) at 10 F (ft-lb)
RTNDT ( F) 1-410 A/C 79, 36, 30 31.7
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-50 3-410 A/C 60, 64, 56 60.0
-50 b
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- A.7.0 Fort Calhoun Vessel Integrity The fracture mechanics analysis is performed using the plant specific properties of the Fort Calhoun vessel. The attenuation of the peak fluence value is considered in three dimensions (r, z, 9), and the superposition of the fluence profile and the weld geometry map is used in calculating the predicted RT value at all points in the vessel riDT as a function of Effective Full Power Years (EFPY). This information is i
used in locating the points in_the vessel having the highest RT at NDT each of the three axial sections of interest:
- 1) middle of core, z 129.46 in.
=
2) top of core, z 65 in.
=
- 3) above-core, z 40 in,
=
where z is the axial distance below the centerline of the nozzle.
From the predicted RT values, the material toughness properties K and f4DT IC K
are determined from the calculated temperatures for the SBLOCA +
Ia LOFW transients using the method described in Section 7.6.
Critical crack depth diagrams are constructed from the applied K vs crack depth y
curves and the calculated material toughness curves. By performing the same fracture mechanics analysis a number of times for increasing plant life (EFPY) the integrity of the Fort Calhoun vessel for the SBLOCA + LOFW transient is evaluated.
A.7.1 Summary of Physics and Materials Data Input to Fracture fiechanics Analysis A detailed survey was performed on the combined fluence and material properties maps of the Fort Calhoun vessel to determine the most critical locations in terms of radiation embrittlement. The properties are considered independently at the three axial sections. At each section, the combination of fluence and materials data were evaluated for a large number of points around the circumference. The adjusted RT values at the inner vessel radius were compared, and the location flDT with the highest RT value was used in the fracture mechanics analysis.
fiDT At the mid-core level the location of highest RT occurs in the flDT weld material at an azimuthal angle of 0 degrees. The fluence factor at this location is.92 of the peak fluence in the vessel.
A21
1 i
The materials data at this point are as follows:
.99 PCT.
Ni
=
.35 PCT.
Cu
=
.012 PCT.
P
=
-20%
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=
NDT 19 At the 12/31/81 level of 5.4 EFPY, and peak fluence of.709 x 10 n/cm2 (E > 1 MEv), this corresponds to a point fluence of.652 'x 1019 2
n/cm and an adjusted surface RT valueof2399.
riDT At the top cf core level the location of highest RT occurs in the t4DT weld material at an azimuthal angle of 0 degrees. The fluence factor at this location in the vessel is.29 of the peak fluence. The materials data at this point are as follows:
.99 PCT.
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=
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Cu
=
.012 PCT.
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=
-20%
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=
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19 n/cm (Ep1 MeV), this corresponds to a point fluence of.209 x 10 2
0 n/cm and an adjuste'd surface RT value of 120 F.
f1DT At the above-core level (about halfway between the top of core and the inlet nozzle), the location of highest RT occurs in the plate material NDT at an azimuthal angle of 45 degrees. The fluence factor at this point is.006 of the peak fluence in the vessel. The materials data for this point are as follows:
.50 PCT.
Ni
=
.110 PCT.
Cu
=
.010 PCT.
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=
-10%
Initial RT
=
NDT A22
At the 12/31/81 level of 5.4 EFPY, and peak fluence of.709 x 10 '
1 19 n/cm2 (E >l MeV), this corresponds to a point fluence of.004 x 10 2
n/cm and an adjusted surface RT valueof-59.
NDT This represents the materials information available at the time of the analysis. Lower initial weld metal RT values were subsequently NDT justified by additional testing and analysis. The use of the present values therefore provide a conservative evaluation of vessel integrity.
A.7.2 Results of Fracture Mechanics Analysis for SBLOCA + LOFW --+ Open PORV'S (Case 4)
The stress analysis for this case is presented in Section 7.8.1 of the report. The fracture mechanics analyses were performed for this case using the Fort Calhoun vessel properties and predicted fluence levels up to the assumed end-of-life condition of 32 EFPY. The critical crack depth diagram at the mid-core level of the vessel for 32 EFPY is given in Figure A.7-1..For times greater than 65 minutes in f
thb transient, K is calculated to exceed the initiation toughness, g
KIC, for a range of initial flaw sizes. However, from the plot of K vs time shown in Figure 7.14 of the report it is seen that warm-y prestressing wodid occur after 10 minutes in the transient, beyond which time K is continually decreasing. Thus, no crack initiation would g
occur under these circumstances. The upper shelf toughness line indicates 200 ksi %.. This represents the upper the flaw depths for which K
=
y i
limit of applicability for linear elastic fracture mechanics. A ductile failure mechanism would be expected for crack sizes above this limit.
The fact that warm-prestressing precludes crack initiation prevents initially small flaws from extending into that region.
l The critical crack depth diagram for the top of core level at 32 EFPY is shown in Figure A.7-2.
For this case, also, initial flaws within a certain range of depth are calculated to exceed the level of initiation toughness after 72 minutes in the transient.
From the plot of K vs time for the top of core level in Figure 7.15 it is seen y
that warm-prestressing occurs after 10 minutes in the transient. Thus, no crack initiation would occur under these conditions at the top of core level of the vessel.
A23
Figure A.7-3 shows the critical crack depth diagram at the above core level of the vessel for 32 EFPY.
It is apparent from this figure that the calculated stress intensities are below both the initiation and arrest toughness levels, thus there is no potential for brittle crack initiation in the vessel above the top of the core for this transient. This is because of the relatively low fluences at this height on the vessel wall.
A.7.3 Results of Fracture Mechanics Analysis for SBLOCA + LOFW ---+
Restoration of Feedwater (Case 5).
The stress analysis for this transient is presented in Section 7.8.2 of the report.
Fracture mechanics analyses were performed using the Fort Calhoun vessel properties with various levels of accumulated fluence up to the assumed end-of-life condition of 32 EFPY. The critical crack depth diagram at the mid-core level of the vessel for 32 EFPY is given in Figure A.7-4.
The calculated stress intensity values exceed the arrest toughness after 70 minutes, and a small initiation region is apparent at 99 minutes in the transient. The fact that warm-prestressing occurs for this transient after 78 minutes, as shown in the plot of K vs. time in Figure 7.17 of the report, indicates g
that crack initiation would not occur under these conditions. The 200ksiT12representstheupper upper shelf toughness line for K;
=
limit of applicability of LEFM. A ductile failure mechanism would be expected for crack sizes above this limit.
In this case, warm-prestressing prevents initially small flaws from extending into that range.
The critical crack depth diagram for the top of core level at 32 EFPY is given in Figure A.7-5.
Similarly, the diagram for the above the core level of the vessel at 32 EFPY is shown in Figure A.7-6.
Both of these figures indicate that the initiation toughness level is not exceeded at these locations in the vessel throughout the expected plant life for this transient loading condition.
A24
A.7.4 Conclusion These results demonstrate that no crack initiation would challenge the integrity of the Fort Calhoun vessel throughout the assumed plant life for the SBLOCA + LOFW transient with recovery of feedwater, and for the SBLOCA + LOFW transient where the PORV's are opened.
A25
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8.0 CONCLUSION
S This Appendix to CEN-189 provides the results of analytical evaluations of pressurized thermal shock effects on the Fort Calhoun reactor vessel for cases of a SBLOCA + LOFW, in response to the requirements of Item II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-uation based on remedial actions to prevent inadeouate core cooling:
1.
SBLOCA + LOFW + POR'V's opened after 10 minutes 2.
SBLOCA + LOFW + Aux. FW reinstated after 30 minutes Thermal-hydraulic system transient calculations were performed on a reference-plant basis, as reported in CEN-189 with the parameter variations over the range representing all operating plants. Four different cases were analyzed for each of the two different scenarios defined above, for a total of eight cases. The most challenging of each of the two different scenarios was analyzed using linear elastic fracture mechanics methods to determine the critical crack tip stress intensity values for comparison to plant specific materials properties at various times in plant life. The effect of the warm prestress phenomenon is identified where applicable for each transient, and credited where appropriate.
In this Appendix, the results of plant specific neutron fluence pro-file calculations are superimposed on plant specific material proper-ties to define vessel capability versus plant life. The results of the generic LEFM analyses were evaluated using the plant specific material properties.
It is concluded that crack initiation would not occur due to the 53LOCA + LOFW transients considered, for more than 32 effective full power years of operation, which is assumed to represent full plant life.
A32
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