ML17334B113
| ML17334B113 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/31/1987 |
| From: | Leverant G, Nair P, Mike Williams SOUTHWEST RESEARCH INSTITUTE |
| To: | |
| Shared Package | |
| ML17325A184 | List: |
| References | |
| NUDOCS 8707020216 | |
| Download: ML17334B113 (221) | |
Text
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....9207280249 SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 REACTOR VESSEL MATERIALSURVEILLANCE PROGRAM FOR DONALD C. COOK UNIT NO. 2:
ANALYSIS OF CAPSULE X By P. K. Nair M. L. Williams (Consultant)
FINAL REPORT SwRI Project 06-8888 Indiana & Michigan Electric Company Donald C. Cook Nuclear Plant COOK PLANT Bridgeman, Michigan 49106 MED RECORD MED COPY SECTIOI'J EER c'i~!GPi!
May 1987 D~.TE
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QI'or i)A12 i(.'I.AHl '7 2 2, i~!Y(+I P'"lii<IANENT APPi";-OVED Approved:
fbi!0l,".1Ul;" RE7c.w 1 lON YRS.
Gerald R. Leverant, Director Department of Materials Sciences
ABSTRACT Capsule X, the third vessel material surveillance capsule removed from the Donald C. Cook Unit No. 2 nuclear power plant has been tested, and the results have been evaluated. The analysis of the data indicates that the pressure mater ial will retain adequate shelf toughness throughout the 32 EFPY design lifetime. Heatup and cooldown limit curves for normal operation have been developed for up to 12 effective full power years of operation.
TABLE OF CONTENTS
~Pa e LIST OF FIGURES 1V LIST OF TABLES 1.0
SUMMARY
OF RESULTS AND CONCLUSIONS
2.0 BACKGROUND
3.0 DESCRIPTION
OF MATERIAL SURVEILLANCE PROGRAM 4.0 TESTING OF SPECIMENS FROM CAPSULE X 12
- 4. 1 Shipment, Opening, and Inspection of Capsule 12 4.2 Neutron, Transport, and Dosimetry Analysis 13 4.3 Mechanical Property Tests 34 5.0 ANALYSIS OF RESULTS 47 6.0 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION OF DONALD C. COOK UNIT NO. 2
7.0 REFERENCES
61 APPENDIX A - Determination of Assembly-Mise Sour ce Distribution for Donald C. Cook Unit 2, Capsule X Analysis APPENDIX 8 - Descr iption of the 3-D Flux Synthesis Method APPENDIX C - Tensile Test Data Records
LIST OF FIGURES
~Fi ere ~Pa e Arrangement of Surveillance Capsules in the Pressure Vessel Vessel Material Surveillance Specimens Arrangement of Specimens in Capsule X R-8 Geometry for Donald C. Cook Unit 2. 15 Radiation Response of Donald CD Cook Unit No. 2 Vessel 41 Shell Plate C5521-2 (Longitudinal Orientation)
Radiation Response of Donald C. Cook Unit No. 2 Vessel 42 Shell Plate C5521-2 (Transverse Orientation)
Radiation Response of Donald C. Cook Unit No. 2 Reactor Vessel Heat-Affected Zone Material Radiation Response of Donald C. Cook Unit No. 2 Reactor 44 Vessel Weld Material Effect of Neutron Fluence on RTNDT Shift, Donald C. Cook 49 Unit No. 2 10 Dependence of C Upper Shelf Energy on Neutron Fluence, Donald C. Cook knit No. 2 11 Reactor Coolant System Pressure-Temperature Limits 57 Versus 100'F/Hour Rate Criticality Limit and Hydro-static Test Limit, 12 EFPY 12 Reactor Coolant System Pressure-Temperature Limits 58 Versus Cooldown Rates, 12 EFPY 13 Reactor Coolant System Pressure-Temperature Limits 59 Versus 100'F/Hour Rate Criticality Limit, and Hydro-static Test Limit, 32 EFPY (Ref. 17) 14 Reactor Coolant System Pressure-Temperature Limits 60 Versus Cooldown Rates, 32 EFPY (Ref. 17)
LIST OF TABLES Table ~Pa e 3.1 Donald C. Cook Unit No. 2 Reactor Vessel Surveillance Mater ials [12]
4.1 47-Group Energy Structure 16 4.2 Reaction Cross Sections (Barns) Used in Calculations For 17 Sequoyah Unit 1 4.3 Absolute Calculated Neutron Fluence Rate Spectra [)(E)] 18 At The Center of Surveillance Capsules (SC) For Donald C. Cook Unit 2 4.4 Calculated Saturated Activities At The Center Of 19 Surveillance Capsules For Donald C. Cook Unit 2 4.5 Donald C. Cook Unit 2 Spectrum-Averaged Cross Sections 19 At Center Of Surveillance Capsules (SC) 4.6 Azimuthal Variation of )(>1) In RPV Of Donald C. Cook Unit 2 20 Calculated Neutron Fluence Rate [)(E)] Spectra In Reactor 21.
0 Pressure Vessel At Peak Axial and Aximuthal Location (8 = 45') For Donald C. Cook Unit 2 4.8 Radial Gradient Of Fast Fluence Rate [)(E>1] Through 22 RPV, At Peak Azimuthal and Axial Locations In Donald C. Cook Unit 2 4.9 Calculated Fluence Rates And Lead Factors In Donald C. 23 Cook Unit 2 4.1O Equations and Definitions For Neutron Dosimetry Analysis 25 4.11 Constants For Processing Dosimetry Data 26
- 4. 12 Reactor Power-Time History For Donald C. Cook Unit 2 Capsule X 27 4.13 Correction Factors To Obtain Measured Saturated Activities 30 At Capsule X Centerline
'alculated 4.14 Unit 2 Saturated Midplane Surveillance Capsules Activities In Donald C. Cook 31
- 4. 15 Compar ison Of Measured and Calculated Saturated Activities 32 For Fast Threshold Detectors Thermal Neutron Fluence Rate In Capsule X 33
LIST OF TABLES (Continued)
Table ~Pa e 4.17 Comparison Of Fast Neutron Fluence Rates From Transport 35 Calculations and Dosimetry Measurements For Capsule X 4.18 Calculated Peak Fluences In Pressure Vessel Based on 35 Capsule X Dosimetry 4.19 Char py Impact Properties of Longitudinal Plate 37 Donald C. Cook Unit No. 2 Capsule X 4.20 Char py Impact Proper ties Of Transverse Plate 38 Donald C. Cook Unit No. 2 Capsule X 4.21 Charpy Impact Proper ties of HAZ Hater ial 39 Donald C. Cook Unit 2 Capsule X 4.22 Charpy Impact Properties Of Meld Metal 4O Donald C. Cook Unit 2 Capsule X 4.23 Effect of Irradiation On Capsule X Surveillance Mater lais 45 Donald C. Cook Unit No. 2 4.24 Tensile Properties Of Surveillance Materials, Donald C. 46 Cook Unit No. 2 5.1 Pr'ojected Values Of RTNDT For Donald C. Cook Unit No. 2 50 5.2 Reactor Vessel Sur veillance Capsule Removal Schedule [16 j 53 Donald C. Cook Unit No. 2
1.0
SUMMARY
OF RESULTS AND CONCLUSIONS The analysis of the third material surveillance capsule removed from the Donald C. Cook Unit No. 2 reactor pressure vessel led to the following conclusions:
(1) Based on a calculated neutron spectral distr ibution, Capsule X received a fast fluence of 1.002 x 10 9 neutrons/cm (E > 1 MeV) at its radial center line.
(2) The surveillance specimens of the core beltline materials experienced shifts in RTNDT of 70'F to 103'F as a result of exposure up to the 1986 refuelling outage.
(3) The core beltline plate mater ials exhibited the largest shifts in RTNDT. Since the intermediate shell plate material has the highest initial (unirradiated) RTNDT it will contr ol the heatup and cooldown limitations thr oughout the design lifetime of the pressure vessel.
(4) The estimated maximum neutron fluence of 3.406 x 10 neutrons/cm (E > 1 MeV) received by the vessel wall accrued in 5.273 effective full power years (EFPY). The pr ojected maximum neutron fluence after 32 EFPY is 2.067 x 10 neutrons/cm (E > 1 MeV). This estimate is based on the average fluence rate after 5.273 EFPY of operations.
(5) Based on the analyses of Capsules T, Y and X, the projected values of RTNDT for the Donald C. Cook Unit 2 vessel core beltline region, at the 1/4T and 3/4T positions after 12 EFPY of operation, are 146'F and 102'F, respectively. These values were used as the bases for computing revt.sed heat-up and cooldown limit curves for up to 12 EFPY of operation.
(6) Based on the analyses of Capsules T, Y and X, the values of RTNDT fot'he Donald C. Cook Unit 2 vessel core beltline region, at the 1/4T and 3/4T positions. after 32 EFPY of operation, are- projected to be 163'F and
130'F, respectively.
(7) The Donald C. Cook Unit No. 2 vessel plates, weld metal, and HAZ material located in the core beltline region are projected to retain sufficient toughness to meet the current requirements of 10CFR50 Appendix G throughout the design life of the unit.
2.0 BACKGROUND
The allowable loadings on nuclear pressure vessels are determined by applying the rules in Appendix G, "Fracture Toughness Requirements," of 10CFR50 [1]. In the case of pressure-retaining components made of ferr itic materials, the allowable loadings depend on the reference stress intensity factor (KIR) curve indexed to the reference nil ductility temperature (RTNDT) presented in Appendix G, "P. otection Against Non-Ductile Failure," of Section III of the ASME Code [2]. Further, the materials in the beltline region of the reactor vessel must be monitored for radiation-induced changes in RTNDT per the requirements of Appendix H, "Reactor Vessel Mater ial Surveillance Program Requirements," of 10CFR50.
The RTNDT is defined in paragraph NB-2331 of Section III of the ASME Code as the highest of the following temperatures:
(1) Dr op-weight Nil Ductility Temperature (DW-NDT) per ASTM E 208 [3];
(2) 60 deg F below the 50 ft-lb Charpy V-notch (Cv) temperature; (3) 60 deg F below the 35 mil C temoerature.
The RTNDT must be established for all materials, including weld metal and heat-affected zone (HAZ) material as well as base plates and forgings, which comprise the reactor coolant pr essur e boundary.
It is well established that ferr itic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron fluences in excess of 10 neutrons per cm (E > 1 MeV) [4]. Also, it has been established that tramp elements, particularly, copper and phosphorus, affect the radiation embrittlement response of ferr itic mater ials
[5-7]. The relationship between increase in RTNDT and copper content is
opening loading (MOL) fracture mechanics specimens. Current technology limitations result in the testing of these specimens at temperatures well below the minimum service temperature in order to obtain valid fracture mechanics data per ASTM E 399 [10], "Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials." Currently, these specimens are being stored pending an acceptable testing procedure like the J< fracture testing [11] has been defined.
This report describes the results obtained from testing the contents of Capsule X. These data and those obtained previously from Capsules T and Y are analyzed to estimate the radiation-induced changes in the mechanical properties of the pressure vessel at the time of the refuelling outage as well as predicting the changes expected to occur at selected times in the future operation of the Donald C. Cook Unit No. 2 power plant.
3.0 DESCRIPTION
OF MATERIAL SURVEILLANCE PROGRAM The Donald C. Cook Unit No, 2 material surveillance program is described in detail in MCAP 8512 [12], dated November 1975. Eight mater ials surveillance capsules were placed in the reactor vessel between the thermal shield and the vessel wall prior to startup, see Figure 1. The vertical center of each capsule is opposite the vertical center of the core.
The capsules each contain Char py V-notches, tensile, and WOL Specimens machined from the SA533 Gr B, CL 2 plate, weld metal, and heat-affected zone (HAZ) materials located at the core beltline. The chemistries and heat treatments of the vessel surveillance mater ials are summarized in Table 3.1. All test specimens were machined from the test mater ials at the quarter-thickness (1/4 T) location after performing a simulated postweld str ess-relieving treatment. Meld and HAZ specimens wer e machined fr om a stress-relieved weldment which joined sections of the intermediate and lower shell plates. HAZ specimens were obtained from the plate C5521-2 side of the weldment. The longitudinal base metal C specimens were or iented with their long axis parallel to the pr imary rolling direction and with V-notches perpendicular to the major plate surfaces. The transverse base metal Cv specimens wer e oriented with their long axis perpendicular to the pr imary rolling direction and with V-notches perpendicular to the major plate sur faces. Tensile specimens were machined with the longitudinal axis perpendicular to the plate primary rolling direction. The MOL specimens were machined with the simulated crack parallel to the pr imary rolling direction and perpendicular to the major plate surfaces. All mechanical test specimens, see Figure 2, were taken at least one plate thickness from the quenched edges of the plate material.
Capsule X contained 44 Charpy V-notched specimens (8 longitudinal and
X (220')
270 i l80'o V ('8>>) Y (320')
2 (356'i S (~.)
T (<0')
I U (140') 90'eac.or Veesel Therr.".al Si
'l'ore 3arrel FIGURE l.. ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE PRESSURE VESSEL
TABLE 3. 1 DONALD C. COOK UNIT NO. 2 REACTOR VESSEL SURVEILLANCE MATERIALS [ 12]
Heat Treatment Histor Shell Plate Material:
Heated to 1700 F for 4-1/2 hours. water quenched.
Heated to 1600 F for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, water quenched.
Tempered at 1250 F for 4-1/2 hour s, air cooled.
Stress relieved at 1150 F for 51-1/2 hours, furnace cooled.
Weldment:
Stress relieved at 1140 F for 9 hours, furnace coo1ed.
Chemical Composition (Percent)
Material C Mn P Si Ni Mo Cu Cr Plate C-5521-2 0.21 1.29 0.013 0.015 0.16 0.58 0.50 0.14 Plate C-5521-2 0.22 1.28 0.017 0.014 0.27 0.58 0.55 0.11 0.072 Weld Metal 0. 11 1.33 0.022 0.012 0.44 0.97 0.54 0.055 0.068 Weld Metal( 0.08 1.42 0.019 0.016 0.36 0.96 0.05 0.07 (a) Lukens Steel analysis.
(b) Westinghouse analysis.
(c) Chicago Bridge and Iron analysis.
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10 12 transver se from the plate material, plus 12 each from weld metal and HAZ material); 4 tensile specimens (2 plate and 2 weld metal); and 4 transverse plate WOL specimens. The specimen numbering system and location within Capsule X is shown in Figure 3.
Capsule X also was reported to contain the following dosimeters for determining the neutr on flux density:
Tar et Element Form ~Quan t it Ir on Bare wire Copper Bare wire Nickel Bare wire Cobalt (in aluminum) Bare wire Cobalt (in aluminum) Cd shielded wire Uranium-238 Cd shielded oxide Neptunium-237 Cd shielded oxide Two eutectic alloy thermal monitor s had been inserted in holes in the steel space s in Capsule X. One (located at the bottom) was 2.5$ Ag and 97.5$
Pb with a melting point of 579'F. The other (located at the top of the capsule) was 1.75$ Ag, 0.75$ Sn, and 97.5$ Pb having a melting point of 590'F.
11 TOP MT-7 LIT-8 TENSILE LIT-16 WOL MT-15 WOL MT-14 WOL WOL LIW 7 MW S TENSILE MW-47 MW-48 SPECIMEN CODE:
MT-47 MT-48 CHARPY MT PLATE MW-45 MW-46 C5221-2 Mr-45 MT-he CHARPY TRANSVERSE ML PLATE 213 C5221-2 LONGITUDINAL MW-43 MW-44 MT-43 MT-44 CHARPY MW WELD LIW-41 MW-42 METAL MT-41 LIT-42 CHARPY MH WELD MW-39 MW-40 HEAT MT-39 MT 40 CHARPY AFFECTED ZONE LIW>>37 MW-38 MT-37 MT-38 CHARPY MH-47 MH-48 ML-31 M1.-32 CHARPY MH 45 MH-46 ML-29 ML-30 CHARPY MH-43 MH 44 ML-27 M1.-28 CHARPY MH-41 MH 42 ML-25 M1.-28 CHARPY MH-39 MH40 BOTTOM MH-37 MH-38 CHARPY FIGURE 3. ARRANGEMENT OF SPECIMENS ZN CAPSULE X
12 4.0 TESTING OF SPECIMENS FROM CAPSULE X The capsule shipment, capsule opening, specimen testing, and reporting of results wene car ried out in accordance with the Project Plan for Donald C. Cook Unit No. 2 Reactor Vessel Irradiation Surveillance Program.
The SwRI Nuclear Projects Oper ating Procedures called out in this plan include:
(1) XI-MS-101-1, "Determination of Specific Activity and Analysis of Radiation Detector Specimens" (2) XI-MS-103-1, "Conducting Tension Tests on Metallic Specimens" (3) XI-MS-104-1, "Charpy Impact Tests on Metallic Specimens" (4) XIII-MS-103-1, "Opening Radiation Surveillance Capsules and Handling and Storing Specimens" (5) XIII-MS-104-2, "Shipment of Westinghouse PMR Vessel Material Surveillance Capsule Using SwRI Cask and Equipment" Copies of the above documents are on file at SwRI.
4.1 Shipment Ooenin and Ins ection of Capsule Southwest Research Institute pr epared Procedure XIII-MS-104-2 for the shipment of Capsule X to the SwRI laboratories. SwRI per sonnel severed the capsule from its extension tube, sectioned the extension tube into several lengths, and supervised the loading of the capsule and extension tube mater ials into the shipping cask for transpor t to San Antonio, Texas.
The capsule was opened and the contents identified and stored in accordance with Procedure XIII-MS-103-1. After sawing off the capsule ends, the long seam welds wer e milled off using a Bridgeport vertical milling machine. The top half of the capsule shell was removed and the specimens and spacer blocks were carefully removed and placed in indexed receptacles identifying each capsule location. After the disassembly had been completed, each specimen was carefully checked to insure agreement with the
13 identification and location as listed in MCAP 8512.[12] No discrepancies were found.
The thermal monitors and neutron dosimeter wires were removed from the holes in the spacers. The thermal monitors, contained in quartz vials, were examined and no melting was observed, thus indicating that the maximum temperature dur ing exposure of Capsule X did not exceed 579'F.
4.2 Neutron Trans crt and Dosimetr Anal sis As par t of the surveillance testing and evaluation program, the neutron transport and dosimetry analysis serves two purposes: ( 1) to determine the neutron fluence (E > 1.0 MeV) in the surveillance capsule where the metallurgical test specimens are located and (2) to determine the neutron fluence (E > 1.0 MeV) incident on and within the reactor pressure vessel (RPV).
The current methodology for RPV fluence determination is based on combining results of transport calculations with measured dosimeter activities. The transport calculations provide three important sets of data in the overall analysis: ( 1) spectrum-weighted, effective dosimeter cross sections, (2) lead factors for various locations in the RPV, and (3) fluence rates at locations of interest.
The calculated effective cross sections for different dosimeter s are divided into the measured reaction rates in order to obtain the fluence rate (E > 1.0 MeV) at the capsule location. The corresponding fluence rates at various depths into the RPV are obtained by dividing the capsule fluence rate by the appropriate lead factors. Both the effective cross sections and the lead factors depend only on ratios of computed results so that absolute
14 calculations are not r equired. The measur ed dosimeter activities pr ovide the fluence rate normalization. However, absolute fluence rates are calculated to compare with measurements to provide a measure of the uncer tainty involved in the RPV fluence determination pr ocedure.
4.2.1 Neutron Trans ort Anal sis A discrete or dinates calculation using the DOT [13] code was performed to obtain the radial (R) and azimuthal (0) fluence-rate distribution for the geometry shown in Figure 4. The inclusion of the surveillance capsules in the R-0 model is mandatory to account for the significant perturbation effects from the physical presence of the capsule.
The 47-group energy structure for the SAILOR[ 14] cross-section library is given in Table 4. 1. An S8 angular str uctur e and a P3 Legendr e cross-section expansion were used in the computations. The fine-group 0
dosimeter cross sections for the Cu(n,a) Co reaction were obtained from ENDF/B-V file and were collapsed to 47 groups using a fission plus 1/E weighting spectrum. The other reaction cross sections were taken from the SAILOR cross-section library. The reaction cross sections are given in Table 4.2.
The results of the transport calculations required for the RPV fluence analysis are presented in Tables 4.3 through 4.9. Table 4.3 contains the calculated absolute fluence-rate spectra for the centerline of the surveillance capsules and in Table 4.4 are the calculated saturated activities obtained by folding the results of Tables 4.3 and 4.2 The spectrum-average cross sections, Table 4.5, are obtained from the results of Tables 4.3 and 4.4. Table 4.6 shows that the peak fluence rates at the inner radius, 1/4-T, and 3/4-T locations are at the 8 = 45'zimuthal, and Table 4.7 are the group fluxes at the peak location. Table 4.8 shows the radial gradients of the fluence rates (E > 1.0 MeV) through the reactor pressure vessel. The peak
15 40 CAPSULES T, U, X, Y r
FORMER PLATE RPV P DOWNCOMER r
THERMAL SHlELD WATER GAP BARREL 4 CAPSULES S,V,W,Z FIGURE 4. R-0 Geometry foi Donald C. Cook Unit 2.
'6 TABLE 4.1 47-GROUP ENERGY STRUCTURE Group Lower energy Group Lower energy (MeV) (Mev) 14.19* 25 0. 183 12.21 26 0.111 10.00 27 0.0674 8.61 28 0.0409 7.41 29 0.0318 6.07 30 0.0261 4.97 31 0.0242 3.68 32 0.0219 3.01 33 0.0150 10 2.73 34 7.10 x 10 2.47 35 3.36 x 10 3 12 2.37 36 1.59 x 10 3 13 2.35 37 4.54 x 10-4 14 2. 23 38 2.14 x 10 4 15 1.92 39 1.01 x 10 4 16 1.65 40 3+73 x 10 17 1.35 41 1.07 x 10 18 1.00 42 5.04 x 10-6 19 0.821 43 1.86 x 10-6 20 0.743 44 8.76 x 10 7 21 0.608 45 4.14 x 10 7 22 0.498 46 1.00 x 10 23 0.369 1.00 x 10-11 24 0.298
- The upper energy of Group 1 is 17.33 MeV.
17 TABLE 4.2 REACTION CROSS SECTIONS (BARNS) USED IN CALCULATIONS FOR DONALD C. COOK UNIT 2 Group Energy U-238 Np-237 Fe-54 Ni-58 C0-63 (MeV) (n f) (n f) (n ) (n ) (n n) 1 1. 733E+01 1.275E+00 2. 535E+00 2.686E+01 2.962E-01 3.682E-02 2 1'.419E+Ol 1.086E+00 2.320E+00 4.137E-01 4.416E-01 4.540E-02 3 1.221E+01 9.844E-Ol 2.334E+00 5 '76E-01 6.103E-01 5.357E-02 4 1.000E+Ol 9.864E-01 2.329E+00 5.781E-01 6.588E-01 3.811E-02
' 8.607E+00 9.891E-01 2.248E+00 5.888E-01 6.553E-01 1.906E-02 6 7.408E+00 8.574E-01 1.965E+00 5.590E-01 6.285E-01 9.277E-03 7 6.065E+00 5.849E-01 1.520E+00 4.697E-01 5.365E-Ol 2.915E-03 8 4.966E+00 5.615E-01 1.538E+00 3.199E-01 3.917E-01 4.437E-04 9 3.679E+00 5.475E-01 1.638E+00 1.762E-01 2.287E-01 3.568E-05 10 3.012E+00 5.463E-01 1.680E+00 1.155E-01 1.658E-01 5.831E-06 11 2.725E+00 5.527E-01 1.697E+00 7.755E-02 1.131E-01 1.707E-06 12 2.466E+00 5.521E-01 1.695E+00 5.111E-02 9.308E-02 6.834E-07 13 2.365E+00 5.512E-01 1.694E+00 4.756E-02 9.232E-02 4.637E-07 14 2.346E+00 5.504E-01 1.693E+00 4.484E-02 8.614E-02 3.430E-07 15 2.231E+00 5.390E-01 1.677E+00 2.008E-02 4.661E-02 1.150E-07 16 1.920E+00 4.685E-01 1.645E+00 4.771E-03 2.660E-03 1.536E-08 17 1.653E+00 2.706E-01 1.604E+00 6.335E-04 1 ~ 337E>>02 0 18 1.353E+00 4.502E-02 1.543E+00 1.311E-05 4.438E-03 0 19 1.003E+00 1.102E-02 1.389E+00 0 5.023E-04 0 20 8.208E-01 2.881E-03 1.205E+00 0 1.729E-04 0 21 7.427E-01 1.397E-03 9.845E-01 0 4.914E-05 0 22 6.081E-01 5.378E-04 6.437E-01 0 7.673E-06 0 23 4.979E-01 1.502E-04 2.642E-01 0 8.903E-07 0 24 3.688E-01 8.333E-05 8 '00E-02 0 4.070E-08 0 25 2.972E-01 6.168E-05 3.552E-02 0 1.832E-15 0 26 1.832E-01 4.668E-05 2.043E-02 0 0 0 27 1.111E-01 4.015E-05 1.542E-02 0 0 0 28 6.738E-02 4.000E-05 1.228E-02 0 0 0 29 4.087E-02 6.176E-05 1.088E-02 0 0 0 30 3.183E-02 8.610E-05 1.023E-02 0 0 0 31 2.606E-02 8.700E-05 1.002E-02 0 0 0 32 2.418E-02 8.700E-05 9.906E-03 0 0 0 33 2.188E-02 8.700E-05 9.723E-03 0 0 0 34 1.503E-02 5.650E-05 1.004E-02 0 0 0 35 7.102E-03 4.860E-11 6.506E-03 0 0 0 36 3.355E-03 7.439E-10 8.716E-03 0 0 0 37 1.585E-03 4.199E-04 2.303E-02 0 0 0 38 4.540E-04 1.464E-08 3.701E-02 0 0 0 39 2. 144 E-04 1.044E-08 6. 129E-02 0 0 0 40 1.013E-04 1.243E-08 9. 027 E-02 0 0 0 41 3.727E-05 1 ~ 955E-08 2.296E-02 0 0 0 42 1.068E-05 3.086E-08 1.014E-02 0 0 0 43 5.043E-06 4.770E-08 4.011E-03 0 0 0 44 1.855E-06 7.171E-08 9.350E-03 0 0 0 45 8.764E-07 5.067E-08 1.407E-02 0 0 0 46 4.140E-07 1.881E-08 4.328E-03 0 0 0 47 1.000E-07 1.182E-09 8.332E-02 0 0 0
TABLE 4.3 ABSOLUTE CALCULATED NEUTRON FLUENCE RATE SPECTRA [4(E)) AT THE CENTER OF SURVEILLANCE CAPSULES (SC) FOR DONALD C. COOK UNIT 2 Group Upper Energy (MeV)
SC at 40'C 4(E)
- n'cm 2's"1 at 4 1 1.733E+01 6.93656E+06 5.76403E+06 2 1.419E+01 3.09479E+07 2.51896E+07 3 1.221E+01 1.27275E+08 9.75622E+07 4 1.000E+01 2.59658E+08 1.92220E+08 5 8.607E+00 4.64990E+08 3.27455E+08 6 7.408E+00 1.10830E+09 7.51266E+08 7 6.065E+00 1.59842E+09 1.00403E+09 8 4.966E+00 3.24363E+09 1.79877E+09 9 3.679E+00 2.93332E+09 1.45231E+09 10 3.012E+00 2.36696E+09 1.12970E+09 11 2.725E+00 2.89003E+09 1.33287E+09 12 2.466E+00 1.42825E+09 6.52104E+08 13 2.365E+00 4.42338E+08 1.98677E+08 14 2.346E+00 2.12501E+09 9.45496E+08 15 2.231E+00 5.48432E+09 2.41337E+09 16 1.920E+00 .7.12292E+09 2.98454K+09 17 1.653E+00 1.03149E+10 4.21588E+09 18 1.353E+00 2.05020E+10 7.93826E+09 19 1.003E+00 1.54321E+10 5.72833E+09 20 8.208E-01 6.80836E+09 2.54752E+09 21 7.427E-01 2.08115E+10 7.26207E+09 22 6.081E-01 1.90620E+10 6.55344E+09 23 4.979E-01 1.87027E+10 6.48139E+09 24 3.688E-01 1.87067E+10 6.28913E+09 25 2.972E-01 2.59350E+10 8.87760E+09 26 1.832E-01 2.32048E+10 7.80143E+09 27 1.111E-01 1.63390E+10 5.48592E+09 28 6.738E-02 1.52521E+10 5.10511E+09 29 4.087E-02 5.03766E+09 1.69700E+09 30 3.183E-02 1.71555E+09 6.14043E+08 31 2.606E-02 5.79265E+09 1.78767E+09 32 2.418E-02 3.69441E+09 1.19550E+09 33 2.188E-02 8.14806E+09 2.67201E+09
19 TABLE 4.4 CALCULATED SATURATED ACTIVITIES AT THE CENTER OF SURVEILLANCE CAPSULES FOR DONALD C. COOK UNIT 2 Surveillance Capsule Surveillance Capsule Reaction at at 4'Bq/g) 40'Bq/g) 54Fe(n,p)54Mn 1.535E+6 2.648E+6 5 Ni(n,p) Co 2.260E+7 4.054E+7 63Cu(n, a) 60Co 2.026E+5 2.867E+5 Np(n, f) 3 Cs 1.119E+7 2.749E+7 238U(n f)137Cs 1.561E+6 3.260E+6 TABLE 4.5 DONALD C. COOK UNIT 2 SPECTRUM-AVERAGED CROSS SECTIONS AT CENTER OF SURVEILLANCE CAPSULES (SC) a(barns) (
Reaction SC at 40 SC at 4'.0894 54Fe(n,p) 0.0678 58Ni(n~p) 0.0927 0.1174 Cu(n,n) 0.000700 0.00113 7Np(n,f) 2.763 2.558 238U(n f) 0.344 0.374 46Ti(n,p) 0.0152
/0 o(E)y(E)dE (1)
Jl $ (E)dE
20 TABLE 4.6 AZIMUTHAL VARIATION OF ](>1) IN RPV OF DONALD C. COOK UNIT y(E > 1.0 MeV) n/cm 's 0-T 1/4-T 3/4-T R ~ 219.78 R ~ 225.19 R ~ 236.142 1 1.56 9.480E+09 5.221E+09 1.028E+09 2 3.28 9.169E+09 5.176E+09 1.041E+09 3 4.00 9.025E+09 5.175E+09 1.052E+09 4 4.72 9.486E+09 5.037E+09 1.073E+09 5 5. 94 1.015E+10 5.597E+09 1.106E+09 6 8.00 1.085E+10 6.001E+09 1.175E+09 7 10.00 1.150E+10 6.375E+09 1.247E+09 8 12.00 1.217E+10 6.749E+09 1.320E+09 9 14.00 1.286E+10 7.122E+09 1.389E+09 0 16. 00 1.350E+10 7.466E+09 1.450E+09 11 18.00 1.402E+10 7.738E+09 1.497E+09 12 20.00 1.432E+10 7.883E+09 1.523E+09 13 21.50 1.427E+10 7.876E+09 1 ~ 527E+09 14 22.50 1.418E+10 7.839E+09 1.527E+09 15 23.50 1.408E+10 7.799E+09 1.526E+09 16 24.39 1.401E+10 7.779E+09 1.527E+09 17 25.02 1.399E+10 7.781E+09 1.530E+09 18 25.48 1.399E+10 7.784E+09 1.532E+09 19 26.31 1.399E+10 7;787E+09 1.537E+09 20 27.49 1.408E+10 7.847E+09 1.551E+09 21 28.30 1.424E+10 7.937E+09 1.568E+09 22 28.74 1.434E+10 7.990E+09 1.578E+09 23 29.48 1.449E+10 8.078E+09 1.597E+09 24 30.50 1.482E+10 8.251E+09 1.628E+09 25 31.50 1.522E+10 8.469E+09 1.666E+09 26 32.47 1.568E+10 8.712E+09 1.708E+09 27 33.47 1.620E+10 8.983E+09 1.754E+09 28 34.50 1.678E+10 9.277E+09 1,803E+09 29 35.25 1.722E+10 9.498E+09 1.837E+09 30 35.75 1.751E+10 9.630E+09 1.858E+09 31 36.25 1.778E+10 9.741E+09 1.877E+09 32 36.75 1.800E+10 9.828E+09 1,893E+09 33 37.25 1.815E+10 9.887E+09 1.907E+09 34 37.75 1.822E+10 9.908E+09 1.920E+09 35 38. 25 1.817E+10 9.900E+09 1.935E+09 36 38.81 1.804E+10 9.902E+09 1.954E+09 37 39.28 1.776E+10 9.924K+09 1.975E+09 38 39.66 1.766E+10 9.975E+09 1.994E+09 39 40.00 1.779E+10 1.006E+10 2.012E+09 40 40.34 1.802E+10 1.016E+10 2.028K+09 41 40.72 1.852E+10 1.032E+10 2.047E+09 42 41.05 1.899E+10 1.046E+10 2.064E+09 43 41.45 1.955E+10 1.066E+10 2.085E+09 c,g 41.92 2.008E+10- 1.090E+10 2.112E+09 42.39 2.047E+10 1. 112E+10 2.139E+09 46 42.87 2.075E+10 1.130E+10 2.165E+09 47 43.34 2.097E+10 1.144E+10 2.186E+09 48 43.82 2.112E+10 1.154E+10 2.203E+09 49 44.29 2.121E+10 1.161E+10 2.215E+09 50 44.76 2.125E+10 1.164E+10 2.221E+09
2'ABLE 4.7 CALCULATED NEUTRON FLUENCE RATE [$ (E) J SPECTRA IN REACTOR PRESSURE VESSEL AT PEAK AXIAL AND AXIMUTHAL LOCATION (6 45') FOR DONALD C. COOK UNIT 2
$ (E ) 1.0 MeV) n/cm 's Upper Energy 0-T 1/4-T 3/4-T Group (MeV) R = 219.78 R ~ 225.19 R = 236.142 1 1.733E+01 0.53166E+07 0.22286E+07 0.36063E+06 2 1.419E+01 0.23088E+08 0.97553E+07 0.15732E+07 3 1.221E+01 0.90374E+08 0.36426E+08 0.53124E+07 4 1.000E+01 0.17693E+09 0.70333E+08 0.96453E+07 5 8.607E+00 0.30438E+09 0.11754E+09 0.14818E+08 6 7.408E+00 0.71052E+09 0.26569E+09 0.30518E+08 7 6.065E+00 0.97912E+09 0.35272E+09 0.37525E+08 8 4'66 E+00 0.17730E+10 0.64140E+09 0.67721E+08 9 3.679E+00 0.13497E+10 0.53264E+09 0.63806E+08 0 3.012E+00 0.10299E+10 0.43784E+09 0.55198E+08 11 2.725E+00 0.11992E+10 0.53614E+09 0.70522E+08 12 2.466E+00 0.60323E+09 0.27104E+09 0.36044E+08 13 2.365E+00 0.17406E+09 0.84240E+08 0.12500E+08 14 2.346E+00 0.80461E+09 0.40595E+09 0.62522E+08 15 2.231E+00 0.19961E+10 0.10353E+10 0.15980E+09 16 1.920E+00 0.22153E+10 0.13200E+10 0.25036E+09 17 1.653E+00 0.30608E+10 0.19119E+10 0.38146K+09 18 1.353E+00 0.47574E+10 0.36067E+10 0.96084E+09 19 1.003E+00 0.31781E+10 0.27155E+10 0.92694E+09 20 8.208E-01 0.16647E+10 0.11772E+10 0.35203E+09 21 7.427E-01 0.43628E+10 0.46686E+10 0.19763E+10 22 6.081E-01 0.38778E+10 0.40155E+10 0.18109E+10 23 4.979E-01 0.42456E+10 0.45651E+10 0.20894E+10 24 3.688E-01 0.41077E+10 0.53608E+10 0.29320E+10 25 2.972E-01 0 '0974E+10 0.61226E+10 0.29813E+10 26 1.832E-01 0.55796E+10 0.62975E+10 0. 33266 E+10 27 1.111E-01 0.42564E+10 0.41358E+10 0.20823E+10 28 6.738E-02 0.37388E+10 0.33406E+10 0.15865E+10 29 4.087E-02 0.15103E+10 0.89469E+09 0.40075E+09 30 3.183E-02 0.99039E+09 0.28232E+09 0.12523E+09 31 2.606E-02 0.13253E+10 0.18702E+10 0.10917E+10 32 2.418E-02 0.90043E+09 0.11019E+10 0.71618E+09 33 2.188E-02 0.22970E+10 0.20128E+10 0.11316E+10
22 TABLE 4.8 RADIAL GRADIENT OF FAST FLUENCE RATE [P(E>1)J THROUGH RPV, AT PEAK AZIMUTHAL AND AXIAL LOCATIONS IN DONALD C. COOK UNIT 2 R(1) (cm) y(E>1) cm -s 219. 978 2. 109E+10 221.14 1.922E+10 222.92 1.572E+10 224.70 1.239E+10 226.48 9.649E+9 228.26 7.452E+9 230.04 5.721E+9 231.82 4.369E+9 233.60 3.316E+9 235.39 2.494E+9 237.17 1.849E+9 238.95 1.331E+9 240.73 8.723E+9 (1) RPV liner begins at R = 219.71 cm.
RPV begins at 220.25 and ends at 241.62 cm.
1/4-T ~ 225.19 cm.
3/4-T ~ 236.14 cm.
23 TABLE 4.9 CALCULATED FLUENCE RATES AND LEAD FACTORS IN DONALD C. COOK UNIT 2 Lead Factors Location . Radius Fluence Rate (cm) [n/(cm 2 '"1)] 4'apsule 40'apsule capsules ID s, V, W, Z (4') 211.41 2.746E+10 T$ U$ X$ Y (40 ) 211.41 6.245E+10 Vessel ID 219.71 2.125E+10 1.29 2.94 Vessel 1/4-T 225.19 1.164E+10 2.36 5.37 Vessel 3/4-T 236.14 2.221E+9 12.36 28.12
fluence rates at the inner radius, 1/4-T, and 3/4-T locations in Table 4.9 are obtained from Table 4.8 by interpolation (or extrapolation). The capsule fluence rates and the lead factors are also summarized in Table 4.9.
4.4.2 Neutron Dosimeter Testin and Anal sis The gamma activities of the dosimeters were determined in accordance with Procedure XI-MS-101-0 using an IT-5400 multi-channel analyzer and a Ge(Li) coaxial detector system. The calibration of the equipment was accomplished with Mn, Co, and 3 Cs radioactivity standards obtained from the U.S. Department of Commerce National Bureau of Standards. The dosimeter wires were weighed on a Mettler-Type H6T balance. All activities were corrected to the time-of-removal (TOR) at reactor shutdown.
The references for the procedures used in processing the dosimeter s ASTM E181-82, "Detector Calibration and Analysis Radionuclides" ASTM E261-77, "Determining Neutron flux, Fluence, and Spectra Radioactive Techniques" ASTM E262-85, Determining Thermal Neutron Flux by Radioactive Techniques ASTM-E263-82, "Determining Fast Neutron Flux by Radioactivation of Iron" ASTM E264-82, "Determining Fast Neutron Flux by Radioactivation of Nickel" ASTM E523-82, "Measuring Fast Neutron Flux Density of Radioactivation of Copper ASTM E704-84, "Determining Fast Neutron Flux Density by Radioactivation of Uranium-238" ASTM E705-84, "Determining Fast Neutron Flux Density by Radioactivation of Neptunium-237" The results of the neutron dosimetr y analysis pr ocedure are summar ized in Tables 4.10 to 4.16. The equations and definitions used for neutron dosimetry analysis are summarized in table 4. 10. The neutron
25 TABLE 4r 10 EQUATIONS AND DEFINITIONS FOR NEUTRON DOSVifETRY ANALYSIS
/ J tj)
~ NoF o(E)d(E)dEQ Pj(lm LTJ)e l(T (C. 1)
ATOg J j~L here ATOR ~ produce auc'Lide activitr ac and of irradiacion, bq/ag; o(E) r energy-dependeat accivacion cross seccioa (ca 2 ) for dosineccr a, 9(E) ~ energyHepcndent fluence race at survcillaace locacioa; Y product nuclide per rcactioa (fission yield);
l ~~ decay coastaat of the product auclidc (d 1);
Pj fraction of full pover during operacing period j; T. r lcagch of tine for irradiacion iatcrval j; k ~ rine froa beginaing of irradiation to cine of rcnovaL; r.. > elapsed cine fron beginning of irradiacion to cad of incerval j; N0 r nuaber of target scone pcr ag in dosinetcr; and J r nuaber of irradiatton intervaled .
+W (C.2)
ASAT J o(E)d(E)dg 0
vberc ASAT ~
s f reactioo race per target nucleus.
W
. o(E)d (K)dg 4(E)dg ASAT d(K>gt)
(4.3)
Kr.
vhete og t ~ cftcccivc spectrua-averaged cross seccion aad d(K)dE ~ flucncc race for neutrons vith energies greater than EtNeV(s(K>gt)).
t Substituting Eq. (2.2) ioto Eq. (2.1) and solving for ASAT, one obca'Las hgSAT ~ ATOR (C.4)
T NoT Q PJ(1-c " j)e " j) j~L Replacing ASAT in Eq. (2.4) by ASAT ia Eq. (2.3), ooc obtains d(E>E ) ~ ATOR (4.5)
J NoYog QPj(L-e LTj)e l(T-tj) jeL The tocal fluence is chen given by J
KE>gt) ~ d(E>gt) QPjTj (4.6) jR The h
~
Noc
'b charnel neutron fluence race (ptb) is daterained froa rbe bare cobalt activici<<s using Eq. (2.F) belov.
(
Cd P.(1 c-XTj)e-A(T- j) aod csdniua-covered (C.y) j-l were Ab bare cobalt activiry (dpsing),
AOd ~ cadniun-covered cobalr. activity (dps/ag),
No ~ nuaber of cobalt-59 scone per ag of cobalt, and oo i 3F.L barns.
Definit iona The Lead faccor (LF)r is defined as follovs oeucroa fluea<<e race (E>gr) at the capsule center
> d aaxtaua aeutron flucncc race at the PV saner radtus Thc saturacion tacror (SF) is given br 1
SF ~
J p P'(1-e jaL lTj)e l(T-cJ)
~A nore general defiaicion can bc stated by replacing the deaoaiaator by the uaxiuua neutron finance rate ac any point ia the pressure vessel (PV) ~
TABLE 4.11 CONSTANTS FOR PROCESSING DOSIMETRY DATA X-ray Branching Fission Atom Atomic Reaction N0 HalE-Life Intensity Yield Fraction Weight (atoms/mg) (day-1)
" Ti(n,p)4 Sc 1.018 x 10 83.85 d 8.261 x 10 0.9998 9 889 keV 0.081 47.90 0.9999 9 1120.keV 54Fe(n,p) 4Mn 6.254 x 10 312.50 d 2.218 x 10 3 0.9997 8 835 keV 0.058 55.847 5 Ni(n,p) Co 7 004 x 1018 70'85 d 9.783 x 10 0.9944 9 811 keV 0.6827 58.70 59Co(n,y) Co 1.022 x 10 5.271 y 3.600 x 10 0.9990 Q 1173 keV 1.0000 58.9332 0.9998 9 1332 keV 63Cu(n,a) Co 6.555 x 10 5.271 y 3.600 x 10 0.9990 Q 1173 keV 0.6917 63.546 0.9998 8 1332 keV 237Np(n,f) 3 Cs 2.540 x 10 30.17 y 6.290 x 10 0.8530 Q 662 ke'V 6.267 . 1.0000 237.0482 238U(n f)137Cs 2 '30 x 1018 30.17 y 6 '90 x 10 0 '530 9 662 keV F 000 1 0000 238 '508 CSi
27 TABLE 4.12 REACTOR POWER-TIME HISTORY FOR DONALD C. COOK UNIT 2 CAPSULE Fraction of Irradiation Decay T ale Operating Full Power* Interval T 11M Step Period P] T J
T-t 3 1 3/78 0.2437 10 2891 2 4/78 0.1544 30 2861 3 5/78 0.2594 31 2830 4 6/78 0.6382 30 2800 5 7/78 0.4396 31 2769 6 8/78 0.6066 31 2738 7 9/78 0.8531 30 2708 8 10/78 0.8825 31 2677 9 11/78 0.4808 30 2647 10 12/78 0.9257 31 2616 ll 1/79 0.9257 31 2585 12 z/79 0.9257 28 2557 13 3/79 0.9257 31 2526 14 4/79 0.9142 30 2496 15 5/79 0.5835 31 2465 16 6/79 0.0000 30 2435 17 7/79 0.9033 31 2404 18 8/79 0.9656 31 2373 19 9/79 0.9656 30 2343 20 10/79 0.5918 31 2312 21 11/79 0.0000 30 2282 22 12/79 0.0000 31 2251 23 1/80 0.4447 31 2220 24 Z/80 0.9191 29 2191 25 3/80 0.9191 31 2160 26 4/80 0.9191 30 2130 27 5/80 0.9191 31 2099 28 6/80 0.8272 30 2069 29 7/80 0.5926 31 2038 30 8/80 0.9669 31 2007 31 9/80 0.9669 30 1977 32 10/80 0.5614 31 1946 33 11/80 0.0000 30 1916 34 i 12/80 0.5979 31 1885 35 1/81 0.9782 31 1854 36 2/81 0.9782 28 1826 37 3/81 0.4418 31 1795 38 4/81 0.0000 30 1765 39 5/81 0.3525 31 1734 40 6/81 0.7806 30 1704 41 7/81 0.7201 31 1673
2v TABLE 4.12 (Continued)
REACTOR POWER-TIME HISTORY FOR DONALD C. COOK UNIT 2 CAPSULE Fraction of Irradiation Decay Time Operating Full Power* Interval Time Step Period P] Tj T-t'2 8/81 0.9516 31 1642 43 9/81 0.9516 30 1612 44 10/81 0. 1343 31 1581 45 11/81 0.9612 30 1551 46 12/81 0.9612 31 1520 47 1/82 0.9612 31 1489 48 2/82 0.9612 28 1461 49 3/82 0.4028 31 1430 50 4/82 0.9569 30 1400 51 5/82 0.9569 31 1369 52 6/82 0.9569 30 1339 53 7/82 0.9569 31 1308 54 8/82 0.4115 31 1277 55 9/82 0.9076 30 1247 56 10/82 0.9215 31 1216 57 11/82 0.6669 30 1186 58 12/82 0.0000 31 1155 59 1/83 0.1217 31 1124 60 2/83 0.9748 28 1096 61 3/83 0.9989 31 1065 62 4/83 0.9930 30 1035 63 5/83 0.9692 31 1004 64 6/83 0.7712 30 974 65 7/83 0.6673 31 943 66 8/83 0.9157 31 912 67 9/83 0.9172 30 882 68 10/83 0.4815 31 851 69 11/83 0.1659 30 821 70 12/83 0.9397 31 790 71 1/84 0.9623 31 759 72 2/84 0.9410 29 730 73 3/84 0.3054 31 699 74 4/84 0.0000 30 669 75 5/84 0.0000 31 638 76 6/84 0.0000 30 608 77 7/84 0.5424 31 577 78 8/84 0.9200 31 546 79 9/84 0.9430 30 516 80 10/84 0.9575 31 485 81 11/84 0.8472 30 455 82 12/84 0.4321 31 424
29 TABLE 4. 12 (Cont inued )
REACTOR POWER-TIME HISTORY FOR DONALD C. COOK UNIT 2 CAPSULE X Fraction of Irradiation Decay Time Opera t ing Full Power* Interval ~ * - -,T"lme-'
Step Period P~ Tg J
~
'-t 83 1/85 0.5208 31 393 84 2/85 0.9916 28 365 85 3/85 0.9764 31 334 86 4/85 0.9924 30 304 87 5/85 0.9986 31 273 88 6/85 0.9985 30 243 89 7/85 0.4295 31 212 90 8/85 0.0237 31 181 91 9/85 0.0000 30 151 92 10/85 0.0641 31 120 93 11/85 0.5437 30 90 94 12/85 0.7942 31 59 95 1/86 0.8000 31 28 2/86 0.5997 28 0
- Full power level for Cook Unit 2 is 3391 HWt. Time of removal is referenced to 2/28/86, 2400 hr.
30 TABLE 4.13 CORRECTION FACTORS TO OBTAIN MEASURED SATURATED ACTIVITIES AT CAPSULE X CENTERLINE Saturation Gradient Impurity Reaction Factor Factor Factor*
5 Fe(n,p) Mn 1.631 1.051 1.0 58Ni(n,p)58Co 1.720 1.164 1.0 63Cu(n,a) Co 2.340 0.9538 1.0 237Np(n f)137Cs 9.037 1.0 1.0 238U(n f)137Cs 9.037 1.0 1.0 39Co(n,y)60Co 2.340 1.164 1.0
- Impurities were assumed negligible.
TABLE 4.14 CALCULATED SATURATED MIDPLANE ACTIVITIES IN DONALD C. COOK UNIT 2 SURVEILLANCE CAPSULES Saturated Activities for Saturated Activities for Dosimeter Ca sule B /
40'urveillance Ca sule Bq/4'urveillance or Flux R=210.41 cm R=211.41 cm R=212.41 cm R=210.41 cm R=211.41 cm R=212.41 cm 54Fe(n,p) 4Mn 3. 240E+06 2. 648E+06 2. 170E+06 1.856E+06 1.535E+06 1.275E+06 58Ni(n,p) Co 4.953E+07 4.054E+07 3. 313E+07 2. 732E+07 2. 260E+07 1. 847 E+07 63Cu(n,a) Co 3.471E+05 2.867E+05 2.390E+05 2.428E+05 2.026E+05 1.704E+05 237Np(n~f)137Cs 3.279E+07 2.749E+07 2.234E+07 1.332E+07 1.119E+07 9.241E+06 238U(n,f)13 Cs 3.963E+06 3.260E+06 2.640E+06 1.880E+06 1.561E+06 1.286E+06 46Ti(n,p)46Sc 7.872E+05 6;454E+05 5.337E+05 5.114K+05 4.240E+05 3.545E+05
$ (E > 1.0 MeV) 7.544E+10 6.245E+10 5.048E+10 3.297E+10 2.746E+10 2.258E+10 g(E > 0.1 MeV) 2. 506 E+11 2. 111E+11 1.717E+11 9.354E+10 7.901E+10 6.521E+10
32 TABLE 4.15 COMPARISON OF MEASURED AND CALCULATED SATURATED ACTIVITIES FOR FAST THRESHOLD DETECTORS Time of Measured Calculated Calculated (C)
Removal Saturated Satura ted Divided by Radial Activity, Activity, Activity, Measures (E)
Reaction ID Location ATOR AE SAT AG SAT Activity (cm) (Bq/mg) (Bq/mg) (Bq/mg) (Bq/mg) 54Fe(n )54Mn Top 211.68 1.375E+3 Top-middle 211.68 1.407E+3 Middle 211.68 1.399E+3 Bottom-middle 211.68 1.423E+3 Bottom 211.68 1.367E+3 Average 1.394 a 0.023E+3 2.390E+3 2.648E+3 1.108 3Cu(n a) 0Co op-middle 211.18 1.197E+2 Middle 211.18 1.202E+2 Bottom-middle 211.18 1.216E+2 Average 1.205 ~ 0.010E+2 2.689E+2 2.867E+2 1.066 Ni(n ) Co Top-middle 212.18 1.837E+4 Middle 212. 18 1.808E+4 Bottom-middle 212.18 1.840E+4 Average 1.828 a 0.018E+4 3.660E+4 4.054E+4 1. 108 237N (n f)137Cs Middle 211.41 3.142E+3 2.839E+4 2.749E+4 0.9683 238U(n f)137Cs iddle 211.41 3.763E+2 3.400E+3 3.260E+3 0.9588
33 TABLE 4.16 THERMAL NEUTRON FLUENCE RATE IN CAPSULE X Thermal Fluence Saturated Activity (Bq/mg) Ra te Axial Location [n/(cm s 1) ]
Bare Cadmium-Covered Top Co 3.448E+07 1.445E+07 5. 283E+10 Bottom Co 3.402E+07 1.445E+07* 5. 16 1E+10 Average 5. 222 E+10
+Assumed to be same as top value.
34 dosimeter s and the constants used in processing the dosimeters are given in Table 4.11.~ ~ The reactor power-time history data given in Table 4.12 are used to calculate the saturation factors (see definition, Table 4.10) shown in Table 4. 13. In Table 4. 13, the gradient correction factors are obtained from the transport calculations given in Table 4. 14 and the impurity cor rection factors are assumed to be negligible. Each of the measured activities AT0R, Table 4. 15 are multiplied by the three appropriate correction factors in Table
- 4. 13 to obtain the measured saturated activities ASAT, for comparison with the calculated values. The results (Table 4.15) indicate that the calculated values are +1 1( to -4$ from the measured values. The thermal neutron fluence rates are given in Table 4.16 and are obtained using Eq. (4.7) from Table
- 4. 10. These values were too low to cause any significant burnin or burnout corrections. ~
4.2.3
~ ~ Results of Neutron Trans or t and Dosimetr Anal sis The comparison of the calculated and the derived fluence rates in Table 4. 17 indicates very good agreement: 6.019 x 10 from the measurements and 6.245 x 10 from the calculations. The derived fluence rate from the measurements is used to determine the fluences shown in Table 4. 18.
The assembly-wise source distribution for Donald C. Cook Unit 2 Capsule X analysis is provided in Appendix A. The three-dimensional (3-D) flux synthesis method used in this report is given in Appendix B.
4.3 Mechanical Pro crt Tests The irradiated Charpy V-notch specimens were tested on a calibrated" SATEC Model SI-1K 240 ft-lb, 16 ft/sec impact machine in accordance with Procedure XI-MS-104-1. The test temperatures, selected to develop the ductile-brittle transition and upper shelf r'egions, were obtained using a liquid conditioning
35 TABLE 4.17 COMPARISON OF FAST NEUTRON FLUENCE RATES FROM TRANSPORT CALCULATIONS AND DOSIMETRY MEASUREMENTS FOR CAPSULE X Calculated Measured Fluence Rate Divided by Saturated Derived from Ca lcu la ted Derived Reaction Activity Measurements Fluence Rate Fluence Rate (Bq/mg) [n/(cm 2.s 1)) (n/(cm 2 s 1) ]
Fe(n,p) Mn 2.390E+03 5.637E+10 6.245E+10 1.108 63Cu(n, 0) Co 2.689E+02 5.860E+10 6.245E+10 1.066 5 Ni(n,p) Co 3.660E+04 5.637E+10 6.245E+10 1. 108 3 Np(n,f) Cs 2.839E+04 6.452E+10 6.245E+10 0.9679 238U(n f)137Cs 3.400E+03 6.511E+10 6.245E+10 0.9591 Average 6.019 ~ 0.432E+10 6.245E+10 1.042 + 0.074 TABLE 4. 18 CALCULATED PEAK FLUENCES IN PRESSURE VESSEL BASED ON CAPSULE X DOSIMETRY
- 5. 273 EFPY 10 EFPY 15 EFPY 32 EFPY Location Fluence Fluence Fluence Fluence (n cm 2) (n cm 2) (n cm 2) (n cm 2)
Surveillance Capsule* 1. 002E+19 1.899E+19 2.849E+19 6.078E+19 Pressure Vessel IR 3. 406 E+ 18 6. 460 E+18 9.690E+18 2.067E+19 Pressure Vessel 1/4-T 1. 865E+18 3.538E+18 5.306E+18 1.132E+19 Pressure Vessel 3/4-T 3.562E+17 6.753E+17 1.013E+18 2.161E+18
- Based on averaged fluence rate derived from dosimetry measurements.
36 both monitor ed with a Fluke Model 2168A digital thermometer . The Charpy V-notch impact data obtained by SwRI on the specimens contained in Capsule X are presented in Tables 4.19 through 4.22. The shifts in the Charpy V-notch transition temperatures determined for the vessel plate, the weld metal and the HAZ mater ials are shown in Figures 5 through 8. The Capsule T and Y results are included for comparison.
A summary of the shifts in RTNDT determined at, the 30 ft-lb level as specified in Appendix G to 10 CFR 50 [ 1], and the reduction in C upper shelf energies for each material, is presented in Table 4.23.
Tensile tests were carried out in accordance with Procedure XI-MS-103-1 using a 22-kip capacity MTS Model 810 Material Test System equipped with an Instron Catalogue No. G-51-13A 2-in. strain gage extensometer and Hewlett Packard Model 7004B X-Y autographic recording equipment. Tensile tests on the plate material and the weld metal were run at 250 F and 550'F at a strain rate of 0.005 in/in/min. through the 0.2$ offset yield strength using servocontrol and ramp generator . The results, along with tensile data reported by Westinghouse on the unirradiated materials [ 12], are presented in Table 4.24. The load-strain records are included in Appendix C.
Testing of the WOL specimens was deferred at the request of Indiana 4 Michigan Electric Company. The specimens are in storage at the SwRI radiation laboratory.
Inspected and calibrated using specimens and procedures obtained from the Army Materials and Mechanics Research Center .
37 TABLE 4.19 CHARPY IMPACT PROPERTIES OF LONGITUDINAL PLATE DONALD C. COOK UNIT 2 CAPSULE X Southwest Research Institute Depar tment of Mater lais Sciences CHARPY TEST DATA SHEET MATERIAL - LONGITUDINAL Project No. 06-8888-001 Date 4/28/87 PHOTOGRAPH SPECIMEN TEMP ENERGY LATERAL FRACTURE NO. oF FT-LBS EXPANSION APPEARANCE ML-25 RT-71 17.0 .017 ML-26 +100 28.5 .026 ML-32 +125 30.5 .026 15 ML-27 +150 40.0 .037 30 ~ % w t'L-31
+175 70.0 .061 45 ML-28 +200 83.5 . 072 90 ML-29 +250 99.0 . 085 100 ML-30 +300 107.0 .085 300
,Q+)
TABLE )1.20 CHARPY IMPACT PROPERTIES OF TRANSVERSE PLATE DONALD C. COOK UNIT NO. 2 CAPSULE X Southwest Research Institute Department of Materials Sciences CHARPY TEST DATA SHEET MATERIAL TRANSVERSE Project No. 06-8888-001 Date 1!/28/87 PHOTOGRAPH SPECIMEN TEMP E?tERGY U\TERAL FRACTURE NO. 'F FT-LBS EXPANSION APPEARANCE Yi MT-48 + 50 8.0 F 007 0 MT-37 RT-7 1 14.5 .013 0 MT-38 +100 23.0 .022 15 MT-46 +100 20.5 .019 10 MT-47 +125 24.5 .024 10 MT-39 +150 30.0 .029 20 I' irg MT-40 +200 50.0 .048 30
<<)T-45 . 20i) 53.i) .050 ~0 I I
I i)T-44 i ".22 "> .60.
. l) <) 55 80 err 4[ ~ .2'.0 !
'>':" ~ r'a 3 N
39 TABLE 4.21 CHARPY IMPACT PROPERTIES OF HAZ MATERIAL DONALD C. COOK UNIT 2 CAPSULE X Southwest Research Institute Department of Materials Sciences CHARPY TEST DATA SHEET MATERIAL HAZ Project No. 06-8888-001 Date 4/28/87 PHOTOGRAPH SPECIMEN TEMP ENERGY LATERAL FRACTURE NO. oF FT-LBS . EXPANSION APPEARANCE MH-43 25 25.0 .018 10 MH-47 + 50 48.5 .039 45 H-37 RT+71 41.5 .036 40 MH-45 +100 64.5 .054 60 MH-38 +100 95.0 .068 70 MH-48 +125 117.0 .082 100 MH"42 +150 97.0 .067 80 MH"41 +200 100.0 F 081 100 MH-40 +200 71.0 .061 100 Mll-46 +225 110.0 .076 100 1
.'ill-44 +250 119.0 .083 100 11-39 +300 103.0 .080 100
4O TABLE 4.22 CHARPY IMPACT PROPERTIES OF MELD METAL DONALD C. COOK UNIT 2 CAPSULE X Southwest Resear ch Institute Department of Materials Sciences CHARPY TEST DATA SHEET MATERIAL - WELO Project No. 06-8888-001 Date 4/28/87 PHOTOGRAPH SPECINEN TEMP ENERGY LATERAL FRACTURE NO. 0F FT-LBS EXPANSION APPEARANCE X
MW-47 - 25 24.5 .022 10 MW-48 16.0 .018 MW-45 + 50 19.5 .017 10 MW-37 RT+71 24.0 .020 15 MW-38 +100 27.0 .030 25 MW-46 +125 61.5 .057 45 MW-40 +150 70.5 .064 100 MW"39 +200 75.5 ~ 069 100 MW-43 +200 61. 0 .058 MW-42 +250 64. 0 .061 100 I
i MW 41 +250 ee.o .057 too I
't MW-44 +300 68.5 .069 100 t
D li V
~ t ~ I~ ~ ~
l
C0 e:
':'parr 'ia'.e.
Capsule T
--e- -Camisole
,Caps le X Y
e'O 115'F 103'F Icier 0 100 200 300 400 530
~ 4 a pc e
804F ag.
/W r
~ e ~
r7 0
-:00 0 100 200 'Ov cs Tes: .e,.Fera.ere, 'c FIGURE 6. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 VESSEL SHELL PLATE C5521-2 (TRANSVERSE ORIENTATION)
43 16C
- Cade:
~
Un 1 r r ad la ted Capsule T
--o- Capsule Y Capsuie X
'e 75'F 4c.
4I 72'F 0
-100 100 2CQ 40C 5CC M'
-I I 0 684F
-lCC 100 200 420
.es Teaoerat"re, F FIGURE 7. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 REACTOR VESSEL MEAT-AFFECTED ZONE MATERIAL
44 10".
CoCe UAl trad lated Capsule
--o--Ca:sule ua Capsule X Y
r 0
../
a ssJ -7~ 60'F
/
70'F 0 100 2M 300 400 500 1CC C
sh PC 600F l
20 100 n JMV e CG 0 2CG xest ie~cerdture, F FIGURE S. RADIATION RESPONSE OF DONALD C. COOK UNIT NO. 2 REACTOR VESSEI WELD MATERIAL
TABLE 4.23 EFFECT OF IRRADIATION ON CAPSULE X SURVEILLANCE MATERIALS DONALD C. COOK UNIT NO. 2 Meld HAZ Trans. Plate Long Plate Criterion (1) Metal Haterial (2) C5521-2(3) C5521-2(3~5)
Transition Temperature Shift 8 50 ft-lb 60 F 75 F 115 F 105 F 8 30 ft-lb 70 F 72 F 103'F 95 F 8 35 mil 60 F 68'F 80'F 98 F (4) 70 F 72 F 103 F 95 F NDT Cv Upper Shelf Drop ll ft-lb 46 f t- lb 23 f t-lb 42 f t-lb (15%) (38%) (27%) (33%)
(1) Refer to Figures 4-7.
(2) Fluence = 8.53 x 1018 n/cm2, E > 1 MeV.
(3) Fluence = 1.05 x 10 n/cm , E > 1 MeV.
(4) Transition temperature shift at 30 ft-lb (46 ft-lb for longitudinal plate).
(5) Transition temperatures at 77 ft-lb, and 54 mils f17].
TABLE 4.24 TENSILE PROPERTIES OF SURVEILLANCE MATERIALS DONALD C. COOK UNIT NO. 2 Fracture Fracture Uniform Total Test Spec. Temp. 0.2% YS UTS Load Stress Elongation Elongation R.A.
Condition Material No. ('F) (ksi) (ksi) (lb) (ksi) (%)- (%) (/)
Capsule (a) Plate C5521-2 X MT-8 250 76.0 93.9 3588 156.0 15.0 18. 7 52.8 (Transverse) HT-7 550 72.1 92.3 3672 163.9 14.8 17.3 54.0 Weld Metal MW-8 210 79.9 94.5 3112 183.1 13. 9 21.4 65.3 HW-7 550 73.7 92.5 3148 166.6 11.4 18.8 61.4 (b) Plate C5521-2 Room 67.4 87.3 3200 161. 2 13.4 23.4 59.6 (Transverse) Room 65.4 85.9 2950 156.4 15.0 27.1 61.7 300 58.8 78.6 2650 146.1 13.0 22.6 63.1 300 60.5 79.5 2675 157.6 10.6 19.8 65.4 550 57.5 83.0 3225 142.1 11.5 19.0 53.8 553 58.9 83.1 3150 145.6 12.7 20.5 56.0 Weld Metal Room 75.7 93.2 2850 173.4 13.9 25. 7 66.8 Room 76.9 91.3 2950 178.8 12.2 22.6 66.6 300 70.7 88.0 2900 171.0 10.7 20.7 66.0 300 71.0 85.3 2875 179.0 10.3 21.2 67.5 550 70 ' 87.2 3160 157.2 10.1 19.2 59.6 550 68.2 87.8 3050 166.0 9.3 20.2 62.8 (a) Fluence = 1.002 x 10 m/cm , E ) 1 MeV.
(b) Unirradiated [12].
5.0 ANALYSIS OF RESULTS The analysis of data obtained from surveillance program specimens has the following goals:
( 1) Estimate the period of time over which the properties of the vessel beltline mater ials will meet the fracture toughness requirements of Appendix G of 10CFR50. This requires a projection of the measured reduction in C V
upper shelf energy to the vessel wall using knowledge of the energy and spatial distr ibution of the neutron flux and the dependence of C upper shelf energy on the neutron fluence.
(2) Develop heatup and cooldown curves to describe the operational limitations for selected periods of time. This requires a projection of the measured shift in RTNDT to the vessel wall using knowledge of the dependence of the shift in RTNDT on the neutron fluence and the energy and spatial distr ibution of the neutron flux.
The energy and spatial distribution of the neutron flux for Donald C.
Cook Unit No. 2 was calculated for Capsule X with a discrete ordinates transpor t Code. This analysis, predicted that the lead factor (ratio of fast flux at the capsule location to the maximum pressure vessel flux) was 2.94 at the capsule centerline, 3.09 for the core-side Charpy layer, and 2.50 for the vessel-side Charpy layer (see Table 4.9). This analysis also predicted that the fast flux at the 1/4T and 3/4T positions in the 8.5-in. pressure vessel wall would be 55( and 11) respectively of that at the vessel I.D.
A method for estimating the increase in RTNDT as a function of neutron fluence and chemistry is given in Regulatory Guide 1.99, Revision 1
[81. However, the Guide also permits interpolation between credible surveillance data and extrapolation by extending the response curves parallel
48 to the Guide trend curves. The data from Capsules T, Y and X are deemed to be credible because (1) the surveillance mater ials are judged to be controlling with regard to r adiation damage, (2) the scat ter in the tr ansver se plate and weld metal Char py data is small, and (3) the changes in yield strength are consistent with the Charpy curve shifts. Except for the longitudinal plate material, the slopes of the response curves constructed in Figure 9 are less than the square root of fluence utilized in Regulatory Guide 1.99. Although recent work [7] indicates that the square root of fluence dependence may be too high, the projected responses of the Donald C. Cook Unit No. 2 vessel beltline materials are based on the trend curves of Figure 9 which were constructed in accordance with Regulatory Guide 1.99 procedures.
The Donald C. Cook Unit No. 2 vessel plate surveillance material is more sensitive than the weld metal 'and HAZ surveillance materials to irradiation embr ittlement. Since the unir radiated values of RTNDT for the intermediate shell plate C5521-2 is higher than those of the weld and HAZ mater ials [16], the beltline region plate material is projected to control the adjusted value of RTNDT through the 32 EFPY design life of Donald C. Cook Unit No. 2. A summary of the projected values of RTNDT for 12 and 32 EFPY of operation of Donald C. Cook Unit No. 2, is presented in Table 5.1.
A method for estimating the reduction in C upper shelf energy as a function of neutr on fluence is also given in Regulatory Guide 1.99, Revision 1
[8]. The results from Capsules T [16], Y [17], and X are compared to a portion of Figure 2 of the Regulatory Guide 1.99, Revision 1, in Figure 10.
Although the shelf energy response of the weld surveillance material from Capsules X fall below them, the pr edictive trend curves of Regulatory Guide 1.99, Revision 1, will be used in this analysis for conservatism.. Response curves have been drawn through the HAZ Transverse Plate and Longitudinal plate
600 I I 400 Reg. Guide 1.99 Upper Limit e
200 100 I~ I l)i li 60 ,j I Code; l i jl III.
s[
- 0 Trans. Plate 40 G Long. Plate
+ Held Metal
- ~ HAZ Material I I I I I 'll 20 2 x lpll 1 pl8 1 pl 9 6 x1019 Neutron Fluence, nlcm2 (E > 1 MeV)
FIGURE 9. EFFECT OF t/EUTRpff FLUENCE Of) RTffDT SHIFT, Dpf(ALD C. COOK UNIT NO. 2
50 TABLE 5.1 PROJECTED VALUES OF RTNDT FOR DONALD C. COOK UNIT NO. 2 EFPY P.V. Mater ial Location hRTNDT Fluence (a) ARTRpT ~Ad
'. RTRpT 12 Plate C5521-2 I.D. 58'F(b) 7.8 x 10'8 101 159 1/4T 58'F 4.3 x 10 88 146 3/4T 58'F 8. x lp17 1 44 102 HAZ Material I.D. 20'F 7.8 x 1018 74 g4 1.4T 20'F 4.3 X lp18 63 83 3/4T 20'F 8.1 x 1017 31 51 Weld Metal I.D. O'( ) 7.8 x 10F 66 66 1/4T 4.3 x lp18 47 47 3/4T O'F 8. 1 x 1017 23 23 32 Plate C5521-2 I.D. 584F( ) 2.1 x 10 9 140 1g8 1/4T 58'F 1.1 x 10 9 105 163 3/4T 58'F 2.2 x 1018 72 130 HAZ Material I.D. 2. 1 x 1p19 113 133 3/4T'04F(b) 1/4T 20'F 1.1 x 10 9 84 104
(
204F 2.2 x 10 50 70 Weld Metal I.D. p'F( ) 2.1 x 10 108 108 1/4T O'F 1.1 x 10 80 80 3/4T OoF 2.2 x 1018 4p 40 (a) Neutrons/cm , E > 1 MeV.
(b) Reference 16.
(c) Estimated per Reference 18
l ~
60 I I
jl Ij
,I 'I' 40 Req. Guide 1.99 Upper Limit 5-g 20 Reg. Guide 1.99 C7l 5-QJ 0.15K Cu Plate C
W 0.05$ Cu Meld 10 I j
- Code: I k
I 0 Trans. Plate I cn Long. Plate
+ Weld Metal s.I
- ~ HAZ Material C) II 4
I 2 x 1P17 1P18 1P19 6 x lP Neutron Fluence, n/cm2 (E > 1 MeV)
FIGURE >o DEPENDEtlCE OF Cy UPPER SIIELF ENERGY 0th NEUTRON FLUENCE DptNLD C COOK UNIT Np 2
data since these results fall above the plate trend curve.
Refer ring to the conservative trend curves for 0.05$ Cu weld metal and the HAZ and plate response curves, the projected Cv shelf energies of the vessel materials are as follows:
o Plate C5521-2 (Unirradiated C Shelf = 86 ft-lb) 32 EFPY at I.D. 60 ft-lb (30$ reduction) 32 EFPY at 1/4T -- 63 ft-lb (27$ reduction) 32 EFPY at 3/4T 71 ft-lb ( 171 reduction)
Note: For shelf energies below the 0.15$ Cu plate curve the conservative plate curve is used.
o Weld Metal (Unirradiated C v Shelf = 75 ft-lb) 32 EFPY at I.D. -- 58 ft-lb (237 reduction) 32 EFPY at 1/4T 60 ft-lb (20$ reduction) 32 EFPY at 3/4T -- 65 ft-lb ( 13$ reduction) o HAZ Material (Unir radiated C v Shelf = 122 ft-lb) 32 EFPY at I.D. 68 ft-lb (44$ reduction) 32 EFPY at 1/4T 73 ft-lb (40$ reduction) 32 EFPY at 3/4T -- 100 ft-lb ( 18( reduction)
These projections indicate that the core beltline materials in the Donald C.
Cook Unit No. 2 pressure vessel mater ial will retain adequate shelf toughness throughout the 32 EFPY design lifetime.
The cur rent Donald C. Cook Unit No. 2 reactor vessel surveillance program removal schedule, revised to conform to ASTM 185-79 [9], is summarized in Table 5.2. Ther e are five capsules remaining in the vessel, of which three are standbys.
53 TABLE 5.2 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE [ 16]
DONALD C. COOK UNIT NO. 2 WOL Removal Equivalent Vessel
~Ca sule Material Time Fluence Weld Metal 1.08 EFPY(a) 3.4 EFPY at I.D.
Weld Metal 3 24 EFPY(b) 11 EFPY at I.D.
Trans. Plate 5.27 EFPY(') E.O.L. at 1/4T Weld Metal 9 EFPY E.O.L. at I.D.
Trans. Plate 32 EFPY E.O.L. at I.D.
Trans. Plate Standby Trans. Plate Standby Weld Metal Standby (a) Removed after core cycle 1.
(b) Removed after core cycle 3.
(c) Removed after core cycle 5.
6.0 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION OF DONALD C. COOK UNIT NO. 2 Donald C. Cook Unit No. 1 is a 3391 Mwt pr essur ized water r eactor operated by Indiana and Michigan Electr ic Company. The unit has bee provided with a reactor vessel material surveillance program as required by 10CFR50, Appendix H.
The third surveillance capsule (Capsule X) was removed during the 1986 refuelling outage. This capsule was tested as described in earlier sections of this report. In summary, these test results indicate that:
( 1) The RTNDT of the surveillance plate material in Capsule X increased 103'F as a result of exposure to a neutron fluence of 1.002 x 10 19 neutrons/cm (E > 1 MeV).
(2) Based on an analysis of the dosimeters in Capsule X, the vessel wall fluence at the I.D. was 3.406 x 10 neutrons/cm (E > 1 MeV) at the time of its removal.
(3) The maximum RTNDT after 12 effective full power years (EFPY) of operation was predicted to be 146'F at the 1/4T and 102'F at the 3/4T vessel wall locations, as controlled by the core beltline shell plate. These projections are comparable to those resulting from the evaluation of the data from capsule Y.
(4) The maximum RTNDT after 32 EFPY of operation was predicted to be 163'F at the 1/4T and 130'F at the 3/4T vessel wall locations, as controlled by the core beltline shell plate. These predictions are lower than that predicted from Capsule Y analysis.
The Unit No. 2 heatup and cooldown limit curves for 12 EFPY and 32 EFPY have been computed on the bases of (3) and (4) above. The following
55 pressure vessel contents were employed as input data in this analysis:
Vessel Inner Radius, 86.50 in., including cladding Vessel Outer Radius, r 95.2 in.
Operating Pressure, P 2235 psig Initial Temperature, T 70oF Final Temperature, Tf 550oF Effective Coolant Flow Rate, Q 134.6 x 10 lb/hr Effective Flow Area, A 26.72 ft2 Effective Hydraulic Diameter, D 15.05 in.
The SwRI computer program calculates the allowable pressure over the temperature range 70'F 550 F such that the r eference stress intensity factor, KIR, is always greater than the sum of twice KI (pressure induced) and KIt (thermal gradient induced) as dictated by Appendix G of the Code (2]. The current version of the SwRI program incorporates the physical property data specified by Appendix I of the Code through the 1982 Summer Adenda. The changes in thermal conductivity code allowables made in the ear ly 1980's reduced the calculated allowable pressure at coolant temperatures below about 200'F from that obtained when using the previously specified values.
Heatup curves were computed for a heatup rate of 100'F/hr. Since lower rates tend to raise the curve in the central region, these curves apply to all heating rates up to 100'F/hr. Cooldown curves were computed for cooldown rates of O'/hr (steady state), 20'F, 40'F/hr, 60'F/hr, and 100'F/hr. The 20'F/hr curve would apply to cooldown rates up to 20'F/hr; the 40'F/hr curve would apply to rates up to 40'F/hr; the 60'F/hr cur ve would apply to rates up to 60'F/hr; the 100'F/hr curve would apply to rates up to
100'F/hr.
The unit No. 2 heatup and cooldown curves developed for up to 12 EFPY after Capsule Y is identical to the Capsule X data. It is recommended that the current technical specification for 12 EFPY not be changed. These curves are reproduced in Figures 11 and 12. The limit curves developed in the Capsule Y repor t for 32 EFPY is conservative compared to the data generated here for Capsule X. These curves are reproduced in Figures 13 and 14.
2600
.REACTOR COOLANT SYSTEM HEATUP LIMITATIONS i::Ilj
'400 FOR FIRST 12 EFFECTIVE FULL POHER AP-.'PLICABLE I II
,'YEARS, (MtQGINS OF 60 PSIG AND 10'F ARE IH- I jCLUDED FOR POSSIBLE INSTRlNENT ERROR) jj 2200 I I
!il!, iJ!I Ili i I II!illlji I i LEAK TEST LIMIT I 2000 I MATERIAL PROPERTY BASIS I
1800 BASE l%TAL CU = 0.14<
(3/4T) ~ 102 oF. OPERATION OPERATION !
1400
".,ijj
- II
'!ij I
1200 Il I
II PRESSURE-TEMPERATlNE 1000 LIMIT FOR HEATUP RATES UP TO 100'F/HR 800 600 CRITICALITY LIMIT I 400 I I I
Ijj CAPS ULE Y TA I j jj} 'jl
}
200 1!jj 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLAI'lT SYSTEM TEN'ERATlNE < F> FIGURE 11. REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS 100'F/HR RATE, CRITICALITY LIMIT AHO IIYDROSTATIC TEST LIMIT, 12 EFPY I I I
2600
- REACTOR COOLANT SYSTEM COOLIXW LIMITATIONS:I APPLICABLE FOR FIRST 12 EFFECTIVE FULL PCNFR 2400 IN-YEARSS (MARGINS OF 60 PSIG AND 10'F ARE 2200
~
.:i.',il.:!I, fiif;fII I':
CLUDED FOR POSSIBLE INSTRlNENT ERROR,) I: I I ~ I
]
2000 I MATERIAL PROPERTY BASIS I 1800 =
'ASE METAL CU 0 ~ 14> UNACCEPTABLE INITIAL RTNDT = ~8 ~ OPERATION 1600 12 EFPY RTNDT (1/4T) = 146 F (3/4T) = 1020F 1400
>- 1200 PRESSURE-TBPERATU ACCEPTABK-LIMITS OPERATION 1000 SOO 8 IC 600 RATE F IIR Q~ 400 20~ a 40" . 200 1!!O ll'I I I II f:I m
~
I',ll
- I f
! I i
I
.,:I II I ~~ .I I
0 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYS/EM TEMPERATURE ( F) FIGURE 12. REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS COOLDOWtl RATES, 12 EFPY
2600 I:I: I ~',II REACTOR COOuelT SYSTEM HEATUP LIMITATIONS AP IPLICABLE FOR FIRST 32 EFFECTIVE FULL POWER
.:,it e STj'l il';
II I e
'l
~ I: ~
I i le
~
2400 I YEARS (MARGINS. OF 60 PSIG AND 10'F ARE IN- LIIIiT il.: li, 2200 CLUDED FOR POSSIBL'E INSTRINENT ERROR) le i i ii. I I:li e,e e
'Ill li ilill!!., I!,I!I! I!! II II ij.'i) 2000 I I
,i Il I
~
i le il lSOO BASE METAL CU = O.l4't j" ACCEPTABLE: INITIALRTNPT = 58 F e OPERATION 1600 32 EFPY RTNPT (1/4T) = 175oF 'INACCEPTAHLE i le (3/4TI = 135'F :. OPERATION e 1400 1200 1000 PRESSURE-TEMPERATURE 800 LIMIT FOR llEATUP RATES j UP TO 100'F/HR e Jj 5 600 I CR ITI CALITY T ij' I"'IMI I g 4oo j, I I ll 200 I", ii"., ill I.i.
~ e 60 100 150 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE ('F)
FIGURE 13. REACTOR COOLA>lT SYSTEM PRESSURE-TEMPERATURE LIMITS VERSUS 100'F/HR RATE, CRITICALITY LIMIT AND IIYDROSTATIC TEST LIMIT, 32 EFPY (Ref. 17)
2600
'I !I
.'j REACTOR COOLANT SYSTEM COOLDON) LIMITATIONS ! lj' jul II
- I APPLICABLE FOR FIRST 32 EFFECTIVE FULL PNER j I II 2400 YEARSe (NARGINS OF 60 PSIG AND 10'F ARE IN- jl CLUDED FOR POSSIBLE INSTRlNENT ERRORS) 2200 I 2000 II MATERIAL PROPERTY BAS I S j
1800 BASE %TAL CU = 0.14%
)
I INITIALRTNDT = 58'F'2 1600 EFPY RTNDT ( 1/4r) 3 4T
=1754'NACCEPTABLE 50F OPERATION 1400 0-1200 . ACCEPTABLF.
- OPERATI CH 1000 PRESSURE"lEMPERATURE
'IMITS
'g 800 1 j'. g . 800 Ij RATE..'.. /R'Il 0 400 20-. 40 .s. 60 200 100~ Iljl I't I III It I I . ~ I 0 60 100 ]50 200 250 300 350 400 450 AVERAGE REACTOR COOLANT SYSTEM TEI'IPERATURE ( F) FIGURE 14. REACTOR COOLANT SYSTEM PRESSURE-TEtfPERATURE LIMITS VERSUS COOLDONl RATES, 32 EFPY (Ref. 17)
61
7.0 REFERENCES
Title 10, Code of Federal Regulations, Part 50, "Licensing of Production and Utilization Facilities."
- 2. ASME Boiler and Pressure Vessel Code, Section III, "Nuclear Power Plant Components."
- 3. ASTM E 208-81, "Standard Method for Conducting Drop-Weight Test to Determine Ni-Ductility Transition Temperature of Ferritic Steels," 1982 Annual Book of ASTM Standards'.
Steele, L. E., and Serpan, C. Z., Jr., "Analysis of Reactor Vessel Radiation Effects Surveillance Programs," ASTM STP 481, December 1970.
- 5. Steele, L. E., "Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels," International Atomic Energy Agency, Technical Repor ts Series No. 163, 1975.
- 6. ASME Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1974 Edition.
7 ~ Randall, P. N., "NRC Perspective of Safety and Licensing Issues Regarding Reactor Vessel Steel Embr ittlement - Cr iteria for Trend Curve Development," presented at the American Nuclear Society Annual Meeting, Detroit, Michigan, June 14, 1983.
- 8. Regulatory Guide 1.99, Revision 1, Office of Standards Development, U.S.
Nuclear Regulatory Commission, April 1977. 9- ASTM E 185-79, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," 1981 Annual Book of ASTM Standards.
- 10. ASTM E 399-81, "Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Mater ials," 1982 Annual Book of ASTM Standards.
ASTM E 813-81, "Standar d Test Method for Toughness," 1982 Annual Book of ASTM Standards." JI, A Measur e of Fr acture
- 12. "American Electric Power Service Corporation Donald C. Cook Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8512, November 1975.
- 13. W. A. Rhoades and R. L. Childs, An U dated Version of the Dot 4 One-and Two-Dimensional Neutron/Photon Trans crt Code, ORNL-5851, Oak Ridge National Laboratory, Oak Ridge, TN, July 1982.
G. L. Simons and R. Roussin, SAILOR - A Coupled Cross Section Library for Light Water Reactors, DLC-76, RSIC. Donald C. Cook Unit No. 2 Technical Specifications.
0 62
7.0 REFERENCES
(continued)
- 16. Norris, E. B , "Reactor Vessel Material Surveillance Program for Donald c.
~
Cook Unit No. 2; Analysis of Capsule T," SwRI Report 06-5928, September 16, 1981.
- 17. Norris, E. B., "Reactor Vessel Material Surveillance Program for Donald C.
Cook Unit No. 2 Analysis of Capsule Y," SwRI Report 06-7244-002, February 1984.
- 18. US NRC Standar d Review Plan, NUREG-75/087, Section 5.3.2, Pr essure-Temperature Limits, November 24, 1975 '
APPENDIX A Determination of Assembly-Wise Source Distribution for Donald C. Cook Unit 2, Capsule X Analysis
DETERMINATION OF ASSEMBLY-WISE SOURCE DISTRIBUTION FOR DONALD C. COOK UNIT 2, CAPSULE X ANALYSIS Surveillance -capsule X was in the reactor for cycles 1-5. Table A.1 shows the cycle-average relative assembly-wise. power distribution for each of these five cycles. These values were obtained by averaging BOC, MOC, and EOC power distributions provided for each cycle. The resulting assembly-wise relative power distribution shown in the last column of Table A.1 formed the basis of the space-dependent source used in the transport calculations. The relative power values shown in this table were multiplied by a value of 17.6 MWth per assembly to obtain the absolute power produced by each assembly. Table A.2 shows the final absolute power produced by each assembly. Table A.2 shows the final absolute assembly-wise power distribution for a quarter core model (note that some assemblies appear as fractions in the quarter core, which reduces their absolute power produced). The absolute power values are converted to a neutron source by multiplying by the conversion factor of 8.163 x 10 neutrons/s per MW. A pin-wise intra-assembly distribution was used to represent the spatial power variation within each of the peripheral assemblies, while a flat distribution is used for interior assemblies. The relative pin-power distribution was provided by the Donald C. Cook Unit 2 sup-port staff. The normalized, space-dependent source distribution is then transformed to the DOT R8 mesh by using a computer program which performs the necessary interpolation and renormalization calculations. The output of this source routine, which includes a listing of the final DOT R9 spatial source distribution, is included.
~ ~ ~ ~
~ The source energy distribution corresponds to an ENDF/B-V Watt fission spectrum. ~
Table A.l. Cycle-Average Assembly Relative Power Distribution for Donald C. Cook Unit 2 CYCLE Zone Average
- 1. 146 0. 861 0.854 0.850 1.013 0.945 2* 1. 188 1.037 1.060 0.962 1. 139 1.077 3* 1. 151 0.968 l. 117 0.987 1. 183 1.081 4* 1. 205 1.135 1. 206 1.038 1.250 1.165 5* 1.117 0.988 1. 113 0.982 1.171 1.074 6* 1.123 1.073 1.079 1.070 1.186 1.106 7* 0.972 0.931 1.084 1.015 1.023 1.005 8* 0.731 0.944 0.873 0.855 0.944 ~0. 869 9* 1.192 1.031 1.047 0.974 1.146 1.078 10 1.151 0.964 1.083 1.064 1.187 1.090 11 1. 184 1.053 1.213 1 ~ 182 1.215 1. 169 12 1. 140 1.077 1.114 1. 066 1. 153 1. 110 13 1.173 1.218 1. 181 1. 185 1.239 l. 199 14 1.069 1.088 1.145 0.999 1. 138 1.088 15 1.039 1.166 1.120 1. 106 1.156 1. 117 16 0.751 0.928 0.851 0.759 0.955 ~0. 849 17* 1.167 0.980 1.122 0.997 1.187 1.091 18 1.189 1.066 1.216 1.183 1.220 1.175 19 1.143 1 ~ 012 1.110 1.089 1.234 1.118 20 1.199 1.237 1. 19.6 1.074 1.278 1.197 21 1.108 1.015 1.098 1.110 1.219 1.110 22 1.097 1.194 1.180 1.225 1.250 1.189 23 0.929 0.905 1.048 1.047 1.106 1.007 24 0.656 0.829 0.752 0.826 0.853 ~o. ass 25* 1.224 1.127 1.211 1.042 1.257 1.172 26 1.165 1.077 1.119 1.076 1. 163 1.120 27 1.201 1.242 1.199 l. 104 1 ~ 292 1.208 28 1. 139 1.011 0. 970 1.098 1.233 1.090 29 1. 134 1.178 1.125 1.244 1.216 1.179 30 1.036 0.942 1.034 1.073 1.183 1.054 31 0. 965 1.081 0.999 1.118 1.119 1.056 32 0. 545 0.556 0.423 0.563 0.459 ~0. 509 33* 1. 169 1.004 1.119 0.994 1.195 1.096 34 1. 199 1.233 1.193 1.198 1.265 1.218 35 1.127 1.026 1.017 1.121 1.226 1.103 36 1.146 1.184 1.127 1.249 1.258 1.193 37 1.166 0.912 1.052 1.038 1.216 1.077 38 0.983 0.984 0.955 1.173 1.215 1.062 39 0.814 0.901 0.781 0.767 0.773 ~0. 807 40* 1.095 1.045 1.075 1.062 1. 182 1.092 41 1.085 1.096 1.151 0.994 1.173 1. 100 42 1.148 1.194 1.191 1.217 1.253 1. 201 43 1.070 0.956 1.039 1.067 1.203 1. 067 44 1.019 0.986 0.941 1 ~ 182 1.210 1.068 45 0.973 1.051 0.893 1.014 1.007 0.988 46 0.497 0.547 0.401 0.404 0.389 0.448
*1/4 assembly in 1/4 core.
~1/2 assembly in 1/4 core. NOTZ: Circled values correspond to peripheral assemblies.
Table A.2. Absolute Assembly (i.e., Zone) Power for Donald C. Cook Unit 2 Total Power 3391 MWth No. of assemblies ~ 193 P per assembly
~ 3391 = 17.57 193 MW assembly Zone Relative Power Absolute Power ]~ 0.945 '.151 2* 1.077 9.461 3* 1.081 9.497 4* 1. 167 10.252 5* 1.074 9.435 6* 1.106 9.716 7* 1.005 8.829 8* 0.869 7.634 9* 1.078 9. 70 10 1.090 19.151 11 1.169 20. 539 12 1.110 19.503 13 1.199 21.066 14 1.088 19.116 15 1.117 19.626 16 0.849 14.917 17* 1.091 9.584 18 1.175 20.645 19 1.118 19.643 20 1.197 21.031 21 1. 110 19.503 22 1.189 20.891 23 1.007 24 0.783 13.757 25* 1.172 10.296 26 1.120 19.678 27 1.208 21.224 28 1.090 19.151 29 1.179 20.715 30 1.054 18.519 31 1.056 18.5 4 32 0.509 8.943 33* 1.096 9.628 34 1.122 19.710 35 1.103 19.380 36 1.193 20.961 37 1.077 18.923 38 1.062 18.659 39 0.807 ~14. 179 40* 1.092 9.593 41 1.100 19.327 42 1.201 21.102 43 1.067 18.747 44 1.068 18.765 45 0.988 17.359 46 0.448 7.871
~1/4 assembly in 1/4 core.
*1/2 assembly in 1/4 core.
NOTE: Circled values correspond to peripheral assemblies.
Figure A.l. Identification of Assembly Nomenclature Used in Source Determination 40 41 42 43 44 45 46 33 34 35 36 37 38 39 25 26 27 28 29 30 31 32 17 18 19 20 21 22 23 24 10 11 12 13 14 15 16 2 3 4 7 8
APPENDIX B Description of the 3-D Flux Synthesis Method
DESCRIPTION OF THE 3-D FLUX SYNTHESIS METHOD A 3-D (RBZ) flux distribution is synthesized using the following well established approximation: RZ (R,Z)
$ (R, e, Z) = pe (Re) 4 R(R)
ORE A(RZ) B.1 where 4R~ is the flux obtained from the RB.DOT calculation; and A(R,Z) " RZ ~ axial distribution function obtained by representing the
~R RZ flux = (QRZ) distribution and dividing it by the integral over Z of the RZ flux; i.e.,
4R J 4RZ dZ. Z In some previous studies, the RZ flux distribution was represented by the results obtained from a DOT RZ calculation, while the radial flux 4R was obtained from a one-dimensional calculation. However, it has been discovered that a simpler approximation gives similar results (within a few percent) as the result of these transport calculations for locations not outside of the RPV and near the reactor midplane. In this approach, we represent
-" (Z)
A(R Z) RZ B.2 J4RZ Z Z fZ P(>>dZ where P(Z) is the average axial distribution of power in the core. The func-tion P(Z) has been represented by 61 discrete nodal values provided by American Electric Power. These values, which are shown in Table B.1 and B.2, correspond to the average relative power for 61 six-centimeter nodes defined over the core height. Table B.l is the MOC axial distribution for a twice-burned peripheral assembly, while Table B.2 is for a fresh peripheral assembly.
Employing the expression in Eq. B.2, we find A(R,Z) = A(Z) AK =
' 1, 61 PKAZ i=1 Evaluating the denominator by summing the values in Tables B.l and B.2, and multiplying by hZ=6 gives PK axial flux factor for node K for burned assembly (PK taken from Table B.l)
PK AK ~ axial flux factor for node K for fresh assembly 15o ~ 8 (PK taken from Table B.2) The axial factors (AK) used in synthesizing the RSZ fluxes are also shown in Tables B.l and B.2. Note from these tables that the axial flux factors have different axial variations for the fresh and burned assemblies (indicating a difference in the relative flux shape). However, the peak value in each case is nearly identical (>>3.1 E-3), and occurs at approximately the same location (-35 inches below the midplane). The axial distribution is fairly flat in both cases, and varies by only about 10X over the middle 9 feet of the core. Since surveillance capsule X as well as the peak RPV flux are located oppo-site a twice-burned assembly, the axial distribution factors in Table B.1 are more appropriate for this analysis. In order to compute the 3-D flux or activity at some axial node i (corre-sponding to a height Z in Tables B.1 and B.2), for some R8 location one must
- 1. find the flux or activity at the appropriate (RI, 8J) location in the DOT RB run
- 2. find the axial flux factor at the appropriate node K
- 3. compute the 3-D value using expression
$ (RI OJ, ZI) 4R6(RI 6g)+AK
(*) For example, the reactor midplane corresponds to node 31. From Table B.l, it can be seen that the axial flux factor for node 31 is equal to 3.063 x 10 Therefore, all activities and fluxes in the DOT Re output should be multiplied by this factor in order to obtain the corresponding midplane values. All of the dosimeter results given in the tables presented previously correspond to midplane values obtained in this manner. The maximum values occur below the midplane and are obtained by using an axial factor of 3.143 x 10
Table B.l. Axial Distribution Factors for Burned Peripheral Assembly in Donald C. Cook Unit 2 Node Zk Ak (cm) (relative power) (axial flux factor) 1 3.0 0. 212 1.301,E-3 2 9.0 0. 212 1.301E-3 3 15.0 0. 268 1.645E-3 4 21.0 0.318 1.952E-3 5 27.0 0.359 2.204E>>3 6 33.0 0.386 2.369E-3 7 39.0 0.368 2.259E-3 8 45.0 0.411 2.523E-3 9 51.0 0.444 2.725E-3 10 57.0 0.456 2.799E-3 11 63.0 0.463 2.842E-3 12 '9.0 0.474 2.910E-3 13 75.0 0.477 2.928E-3 14 81.0 0.479 2.940E-3 15 87.0 0.470 2.885E-3 16 93.0 0.413 2.535E-3 17 99.0 0.470 2.885E-3 18 105.0 0.483 2.965E-3 19 111.0 0.488 2.995E-3 20 117.0 0.494 3.032E-3 21 123.0 0.496 3.045E-3 22 129.0 0.498 3.057E-3 23 135.0 0.494 3.032E-3 24 141.0 0.462 2.836E-3 25 147. 0 0.444 2.725E-3 26 153.0 0.488 2.995E-3 27 159.0 0.491 3.014E-3 28 165.0 0.496 3.045E-3 29 171.0 0.499 3.063E-3 30 177.0 0.501 3.075E-3 ~Mid lane 31 183.0 0.499 3.063E-3 32 189.0 0.493 3.026E-3 33 195.0 0.438 2.689E-3 34 201.0 0.476 2.922E-3 35 207.0 0.496 3.045E-3 36 213.0 0.498 3.057E-3 37 219.0 0.499 3.063E-3 38 225.0 0.504 3.094E-3 39 231.0 0.504 3.094E-3 40 237.0 0.503 3.088E-3 41 243.0 0.491 3.014E-3 42 249.0 0.438 2.689E-3 43 255.0 0.497 3.051E-3 44 261.0 -0.507 3.112E-3 45 267. 0 0.512 3.143E-3 46 273.0 0.512 3.143E-3
Table B.l. (continued) Node 21c Ak (c ) (relative power) (axial flux factor) 47 279.0 0.511 3.137E-3 48 285. 0 0.507 3.112E-3 49 291. 0 0.499 3.063E-3 50 297.0 0.462 2.836E-3 51 303.0 0.442 2.713E-3 52 309.0 0.484 2.971E-3 53 315.0 0.482 2.959E-3 54 321.0 0.477 2.928E-3 55 327.0 0.466 2.860E-3 56 333.0 0.449 2.756E-3 57 339.0 0.422 2.590E-3 58 345.0 0.381 2.339E-3 59 351.0 0.332 2.037E-3 60 357.0 0.266 1.632E-3 61 363.0 0.133 8.160E-4
Table B.2. Axial Distribution Factors for Fresh Peripheral Assembly in Donald C. Cook Unit 2 Node zk Ak (cm) (relative pover) (axial flux factor) 1 3.0 0. 174 1. 154E-3 2 9.0 0. 183 1.214E-3 3 15.0 0. 238 1.578E-3 21.0 0. 283 1.877E-3 5 27.0 0.320 2.122E-3 6 33.0 0.347 2.301E-3 7 39.0 0.348 2.308E-3 8 45.0 0.373 2.474E-3 9 51.0 0.403 2.673E-3 10 57.0 0.416 2.759E-3 11 63.0 0.427 2.832E-3 12 69.0 0.432 2.865E-3 13 75.0 0.434 2.878E-3 14 81.0 0.435 2.885E-3 15 87.0 0.428 2.839E-3 16 93.0 0.405 2.686E-3 17 99.0 0.431 2.858E-3 18 105.0 0.436 2.892E-3 19 111.0 0.438 2.905E-3 20 117.0 0.442 2.931E-3 21 123.0 0.444 2.945E-3 22 129.0 0.445 2.951E-3 23 135.0 0 ~ 444. 2.945E-3 24 141.0 0.420 2.786E-3 25 147.0 0.425 2.819E-3 26 153.0 0.450 2.984E-3 27 159.0 0.457 3.031E-3 28 165.0 0.458 3.038E-3 29 171.0 0.460 3.051E-3 30 177.0 0.459 3.044E-3 d lane 31 183.0 0.461 3.057E-3 32 189.0 0.454 3.011E-3 33 195.0 0.427 2.832E-3 34 201.0 0.451 2.991E-3 35 207.0 0.461 3.057E-3 36 213.0 0.464 3.077K-3 37 219.0 0.466 3.091E-3 38 225.0 0.467 3.097E-3 39 231.0 0.467 3.097E-3 40 237.0 0.465 3.084E-3 41 243.0 0.447 2.965E-3 42 249.0 0.436 2.892E-3 43 255.0 0.465 3.084E-3 44 261.0 0.473 3.137K-3 45 267.0 0.476 3.157E-3 46 273.0 0.478 3.170E-3
Table B.2. (continued) Node 2k Pk Ak (cm) (relative power) (axial flux factor) 47 279.0 0.478 3.170E-3 48 285.0 0.478 3. 170E-3 49 291.0 0.473 3.137E-3 50 297.0 0.442 2.931E-3 51 303.0 0.461 3.057E-3 52 309.0 0.466 3.091E-3 53 315.0 0.458 3.038E-3 54 321.0 0.450 2.984E-3 55 327.0 0.434 2.878E-3 56 333.0 0.413 2.739E-3 57 339.0 0.382 2.533E-3 58 345.0 .0.342 2.268E-3 59 351.0 0.286 1.897E-3 60 357.0 0.207 1.373E-3 Bottom 61 363.0 0.207 1.373E-3
APPENDIX C Tensile Test Data Records
Southwest Research Ins i ute Oepartment of Materials Sciences TENSILE TEST OATA SHE"i Specimen No. ri 8 pro ject Ho. Q8'-+w>4->-r r Test Temperature /N r Machine Ident. Strain Rate . aaMrw .ritYw.~ Data of Test 4/~rk7 Initial Oiameter . Z .n Final Oiameter .r+'7 Initial Final Area w/ 7;i" lltllgg Area L Final Gage Length g~ g ure:
.'~r~.'aximum Specimen Tempera Load Top T.C. 0.2~ Offset Load Middle T.C. Frac ure Load Bottom T.C. ~ /n"~ ll g.t g*. L U.T.S. = Maximum Load/Initial Area 0.2.. Y.S. = 0.2 Offset Load/Initial Area Frature Stress = Fracture Load/Final Area
" R.A. 100 (Init. Area-Final Area)/Init. Area " Total Elong. = 100 (Final G.L.-Init. G.L.)/Init. G.L. Pr', 5N .. Uniform Elong. = 100 (Elong. to Max. Load)/Init. G.L. Test Performed by: Calculations Per ormed by: ~ .W.~~~~~ (Oate Calculations Checked by: (Oate) ~/7/~~
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R.A. = 100 (Init. Area-Final Area)/Init. Area /j
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i /rg~ ~ / y~ Oate of Test eJ~ "J<W Ini.ial Oiameter Final Oiameter Ini.ial Area r VC7:n+ Final Area Initial Gage Lenctn Final Gage Length Specimen Tempe. a:ure: Maximum Load Top T.C. nP Q.Z.rac=ure Middle T.C. Load Hottom T.C. w47 along. to Max. Load , / 0.2".. Y.S. = 0.2'.l Offset Load/Initial Area 7', r> Frature Stress = Frac.ure Load/Final Area /4 .9',.~ /
",. R.A. = 100 (Init. Area-Final Area)/.nst. Area p +,'zn ".. To al long, = 100 (Final G.L.-Init. G.L.)/;r..'-. G.L. .. Uniform =long. = 100 (along. to Max. Load)/ n't. G.L.
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TABLE 4.1-12 OR COO COD S Unit 1 gor~oe~et gode dde da and Code Cases Reactor Vessel ASHE III*Class A 1965 Ed. through 1966 Winter Addenda, Code Cases 1332-2, 1358, 1339-'2, 1335, 1359-1, 1338-3, 1336 Full Length Control Rod ASHE III*Class A 1965 Ed. through 1966 Winter Drfve Mechanisms Addenda Steam Generators ASME IZZ* Class A 1965 Ed. through 1966 Winter Addenda Reactor- Coolant Pump No Code (Designed 1968 Edition Casings with ASME III Article 4 as a Guide) Pressurizer ASME III*CLass A 1965 Ed. through Winter 1966 Addenda, Code Cases 1401, 1459 Pressurizer Safety ASME III* 1968 Edition Valves Power Operated Rely.ef B-16.5 Valves Hain Reactor Coolant B31.1 1967 Edition System Piping Reactor Coolant System B-16.5 or HSS-SP-66, Valves and ASHE III, 1968 Edition* ASME Boiler and Pressure Vessel Code, Section IZI-Nuclear Vessels Repairs and replacements are conducted in " ;ordance with ASME Section XI 4.1-40 July 1991
V ~
8Hoc/~en~ 5 pa~ gg g TABLE 4.1-12 (cont'd.) Unit 2 . goragone~t ~od ddenda and Code Cases Reactor Vessel ASME IXI Class A 1968 Ed. (1968 Summer Addenda)-Coda Case 1335-4 Full Length Control ASME IXX Class A 1968 Ed.'No Add.) Rod Drive Mechanisms Steam Generators ASME IXX Class A 1968 Ed. through Minter 1968 Addenda, Code Cases 1401, 1498 for upper assemblies and 1983 Ed. through Summer 1984 for replacement lower assemblies Reactor Coolant Pump No Code (Desi~ed 1968 Edition through Casings with ASME XII Summer 1969 Addenda Artic3.e 4 as a Guide) Pressurizer ASME IXI* Class A 1965 Ed. through Wincer 1966 Addenda Pressurizer Safeey ASME IXI* 1968 Edition Valves Power Operated Relief B-16.5 Valves Main Reactor .Coolant B31.1 1967 Edition System Piping Reactor Coolant System B-16. 5 or MSS -SP -66, Valves and ASME IXX, 1968 Edition+
*ASME Boiler and Pressure Vessel Code, Section IIX - Nuclear Vessels Repairs and replacamenes are conducted in accordance with ASME Section XI 4.1-41 July 1991
EF ":. Sc MICHIGAN POWER COMPi J. ..r ~..u c Y I 1 C, . Hovaxiber 7, 1977 Donald C. Cook Nuclear Plant Qn3.t Ho' Docket Ho '0-315 DPR No~ 58 r I' Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U.S ~ Nuclear Regulatory Commission shington, D.C. 20555 ear Mr Casec T?d.s letter responds to Mrs Don K. Davis'etter oi 20, 1977 requesting reactor vessel material property information the Donald C. Cook Nuclear plant In our letter dated D uly 25, 1977, we informed you that we would need additional time to provide the requested information. Enclosed herewith are three (3) copies of a document entitled, "D. C Cook Unit Ho. 1 Reactor Vessel Material urveillance Program" which supplies the information requested. Very truly yours, ohn Tzdlingha t Vice Preside JT ~mam Sworn and subscribed to before me on this 7 day of November 1977 in New York County, New York Notary P lic GfiEGOiTY M. Gi:Z~Vilr Hatary Public. St:te at <'tew Yurit Ha. 31-46<3<31 GualiTied in New Yark County Commiss'an Expires Msrch 30, 19?5
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l:i " D. C. COOK UNIT NO. $ Rei'd)l.hs'es>>W'.Q? '.:;..
-REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM 'p p'. g r r s
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-: T;) The estimated maximum fluence (E >> I Mev) at the jnner Surface of the reactor vessel as of March 31, 1977 is 8.38 x 10 n/cm~.
- -;2.} .The effective full power years (EFPY) of operation accumulated as of March.31, 1977 is 1.34 EFPY.
'.: -'-.;.3. ) Fab~ication of the reactor. vessel was performed by Combustion Engineering, ~ r Inca P .:.',,:.4.) .
a.) Sketch of the reactor vessel showing materials in the belt1ine region
'...is shown in Figure l.
b.'). Information on each of the welds in the beltline region is shown in
~
.. Tables 1 through 4.
surveys c'.) Information on each of the plates in the beltline region is shown in Tables 4 through 8. Information relative to the weld and plate material in the material llance program is shown in Tables 1 through 3 and 5 through 8. a ~ I I ~ r .'
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, IDENTIFICATION AND LOCATION OF D. C.
FIGURE 1., COOK UNIT NO. 1
'. g .t+jj +'k p~p'g REACTOR VESSEL 8 ~f BELTLINE REGION WELD AND PLATE NTERIAL 1
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TA IDENTIFICATION AND LOCATION OF D. C'OOK VNIT N0.,1 VESSEL BELTLINE REGION MELD METAL': '-*".'.:. '. -, .",'.-,-,-,'.-..'lux Weld Wire Meld Location ttrdt tt t t Weld tt. t~ t ttl. ~T e Lot No. Post Meld Heat Treatment :: Nozzle Shell Submerged Arc Ml.14 B-4 Mod. '. 1125-1175'F-40HR-FC 13253 Lande 1092 3791 Vertical Seams (Tandem Mire) . B-4 Mod. .
'12008 1-442 A, 8 8( C Nozzle Shell to Submerged Arc Ml .18 'B-4 Mod. 20291 .'inde 1092 , 3833 1125-1175'F<<40HR-FC .;
Inter Shell rg Circle Seam ~' J 8-442 0r r ~~, Inter. Shell Submerged Arc M1.14 "B-4 Mod. 13253 Linde 1 092 .'791 ,:. 1125-11754F-40HR-FC ..: = Vertical Seams (Tandem Mire) B-4 Mod. : 12008 2-442 A, 8 5 C.
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Inter. to Lower Submerged Arc M1.42 .'. B-4 Mod. : IP3571 Linde 1 092- . 3958: ...'.1125-1175'F-40HR-FG -'- Shell Circle 9-442 Seam m
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S' Lower Shell Vertical Seams Submerged Arc (Tandem Mire) M1.14 Mod. B-4 Mod. 13253 12008.
. Lande 1 092 '; 3791 '-4
--;;..1125-1175'F-.40HR-FC ',;
3-442 A, 8 5 C Surveillance Submerged Arc 8-4 Mod. '13253 . Unde 1092 '. 3791 ~
.:. 1125-1175'F-40HR-FC -.: ,.'.
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IA: ':::.': CHEMICAL COMPOSITION 0 F YESSEL BELTLINE REGION MELO METAL
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5 Weld Mire Flux Mei ht Percent T~e Heat. No. i ~Te Lot No. C P S Si Ni 'Mo".' Cr . Cu "V 84 Mod. 13253 Linde 1092 3791 ,15 1.83 =..013 .015 .06 .72'45 ~
.04 ',07 84 Mod. 12008 'Linde 1092 3791. .13 1.92 ,010 .015 .05 .99 .51 .-:.06 .13 .
84 Miod. 20291 Linde 1092 3833 .16 1. 92 .008 .009 .03 .74 .51 84 Miod. IP3571 Linde 1092 3958 .12 1. 38 .017 : .009 .21 .82 .54 . ': .40 Surveillance Meld .26" 1.33 .023 .014 .18 .74 .44 .02 .
.27 .001
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- Mire Analysis - No As Deposited Meld Analysis was Performed ~ ~ ~, ~
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MECHANICAL PROPERTIES OF YESSEL BELTLINE REGION MELD METAL Energy Shelf Meld Mire Flux Lot No. TNDT oF at 10'F Ft-Lbs RTNPT oF Ft-Lbo YS 'TS KSI: KSI: ..El ong,'A T~e Heat No. .
'9.7 84 Mod.
84 Mod. 13253) 12008J Linde 1092 3791
'* 0* . 84,74,70 0* -- 63.3 80.1 27.5 84 Mod.
84 Mod. 20291 IP3571 Surveillance Meld Surveillance Meld Linde 1092 Linde 1092 CE Tests W Tests 3833 3958
>>70 0*
35,50,48 40,46,46 54~54~73 83,84,92
, >>56
-70 0*
115 111 ~
.'I 69.0 5>>>>,>>>>>81.9 70.5 67.1 88.0 84.0,'. '>> . 25.5 28.0.
26.8
. 67.1 69,4 69.2
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- Estimated per NRC Standard Review Plan Section 5.3.2 ~.-
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TABL XIMUM END-OF-LIFE FLUENCE AT VESSEL INNER MALL LOCATIONS I. h ~ ~ ~ I~ Plate or -. ..'Fluen e Plate or Meld Seam Location Seam No. :
~II C4
. Nozz le Shell Verti cal Seam 1-442A ;2.4 x 101 II II ll 1-442B 3.9 x 1017 II II II 1-442C '.9 x 101 Nozz le Shell to In ter. She ll Circle Seam "--.:.',t 8-442 7.3 x 1017 Inte r. Shell Verti cal Seam 2-442A 6.2 x 1018
'l II II II I
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2-442B 2-442C 1.1 x 1019 1.1 x 1019
'nte r. Shell to Lower Shel 1 Circle Seam 9-442 'I 2.0 x 1019 Lowe r Shell Vertic al Seam II 3-442A 1 1 x 1019 3-442B 6.2 x 1018 II 3-442C ~ ~ ~ ~
1.1 x 1019 Nozz le Shell Plate B4405-1 7,3 x 1017 II II B4405-2 7,3 x 10177 II II 84405-3 r' r 7.3 x 10 Inte r. Shell Plate B4406-1 2.0 x 10 II B4406-2 ~ ~ 2,.0 x 1019 II ' B4406-3 2.O x 1O19 Lowe II r Shell Plate II B4407-1 ~ 4I 2.0 x 10
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B4407-2 2.0 x 10119 1 B4407-3 2.0 x 10
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'DENTIF ION OF VESSEL BELTLINE REGION PLATE MATERIAL I
Mat'l. Heat 'Treatment ~Com onent Plate No .. 'eat No. No., ~Su I Ier 'ustenltlze ~Tem er .---.-'tress
.'Sec.
Relief ozzie Shell 84405-1 C3594 A5338 Cl. 1 Lukens 1600'F+50'F-4HR Mg 1225'F+25'F-4HR-AC . ))50'F+25'F-40HR-FC II II i 84405-2 C3594 A5338 Cl. 1 Lukens,. II II II II II 84405-3 C3872 A5338 Cl. 1 Lukens;,'- . . nter. Shell 84406-1 C1260 A5338 Cl. 1 .Lukens'.. II ...., II 84406-2 C3506 . A5338 Cl. 1 Lukens .- II II 84406-3
- C3506 A5338 Cl. 1 Lukens ':"
ower Shell 84407-1 C3929 A5338 Cl. 1 Lukens II II 84407-2 C3932 A5538 Cl. 1 Lukens II II 84407-3 C3929 A5538 Cl. 1 Lukens'....- Surve)llance Material same as Inter. Shell Plate B4406-3
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TABLE 6 CHE MICAL COMPOSITION OF VESSEL BELTLINE REGION PLATE MATERIAL h
..I Mei ht Percent Plate No. C Mn P S S$ N$ Mo Cu.42 84405-1 '21 .007 ..018 .26 .46 ".47 .14
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84405-2 .20 1.41 .006 .018 .25: - .45 .47 " .14 -' 84405-3 .24 30 .008 .013 .30 .48 .46 .14 84406-1 .25 1.17 .016 '025 .29 '52 ..49 '.12 84406-2 .24 1.41 .00& ': .015 .28 .50 .47 .15
'.49
84406-3 . 21'21 1.40 ..009 .015 .25 : .46 .1S 84407-1 1.35 .010 .014 .29 .55 : .53 . .14 84407-2 84407 3
.20
.22 1.25 1 .32
.012
.010
. .014
.014
.22
.24
.59
.50
.54
.55 'l4.12 84406-3* .24 1.40 .009 .015 .25 ; .49 , .46 .14
- Survei llance Plate Analysis Performed by West)nghouse
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c, MECHANICAL PROPERTIES OF VESSEL TLINE REGION PLATE MATERIAL::.
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TNDT RTNDT* Ft-L s YS'SI UTS, ':..- .,.; "RA ..' Plate No. oF r oF lOD NMWD* KSI 84405-1 10 2 134 87 , ~ 56,3 81.3 .-. 29.5 '.,';:,::.;-:.- 6&.l;-.'. 84405-2 0 34 142 92 62.9 .'-.:...::. 28. 5,::;.".':- 66.8 '5.8 84405-3 ~ 0 r I 40 123 80 64.4 86.4 ':'..":: 25.5 ".;;-"i-'. 66.5 84406-1
-10 -8 123 80 63.3
'helf.
86.3 .': '.'. 27.0::.'-..'.";".' 67.1 84406-2 -10 17 124 80. 5 .'7.2 89.7 .., ". "" 26.2:~"'::. ': 68.0 ' 84406-3 -10 27 121 78.5: 66.8 88.8 - " '. 26. 2:.:",;-':. 68. 0,."- ) 84407-1 -20 85.5 64.1 86.7
.'&.0'.:,:;:.'-':::,. '9.6 84407-2 84407-3
-20 0
-15 5
0 133 149 139 90.5 62.1 63.7 ~ 84.1 86.4 27.2 "; -;.-:. 70.6 27.2
".;:J,'.-; 69.7 '-'."'.
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- Estimated from Data in the Major Morking Direc 'ion (MUD ) per NRC'St andard Review Plan Sect)on 5.3.2 'I
~
Q TABLE 8 F MECHANICAL PROPERTIES OF SURVEILLANCE PLATE 5 OTHER BELTLINE P LATES PERFORMED BY MESTINGHOUSE '".-: . Shelf Ener h NDT
'F NDT s YS UTS Elong. " ~
RA ~ Plate No.. oF lOD NMMD KSI KSI 84406-1 5 83 84406-2 84406-3 84407-1 33 40 28 130 96 98 '68.4 90.4 '.." 27.5 ":: -'..'0.0 't 103 ~ I m 84407-2 -12 126 84407-3 38 108 ~
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gPgchm. 4 '7 f4+ top l5 Provide the following information for the pressuze vessel:
- l. A schematic of the reactoz vessel showing all welds in the belt-line region. Welds should be identified by a shop contzol number (such as a pzoceduze qualification number) and the heat of filler metal, type and batch number of flux, etc.
- 2. For each of the above welds, and for welds in the vessel material surveillance programs, an identification of the welding process (sub arc, electroslag, manual metal arc, etc.). Also, a listing of the following information on each of these welds: chemical composition (particularly Cu, P and S content), drop weight T~ , RT , upper shelf Charpy energy and tensile properties.
- 3. The maximum end of life fluence at the vessel I.D. for each weld in the beltline.
Reference NRC letter dated Hay 20, 1977 to Fw. John Tillinghast, Vice President, Indiana and Michigan Electric Company on the above subject and addi-tional requested information. For Donald C. Cook Unit 2 reactor vessel the response to the above question and to the additional requested information in the referenced letter is provided below:
- l. Not Applicable.
2 Not Applicable.
- 3. Chicago Bridge and Iron.
4 a A sketch of the reactor vessel showing all material welds in the beltline region is shown in Figure 1.
- b. Information relative to each of the welds in the beltline region is shown in Tables 1 through 4.
121. 2-1 Appendix g jgENDHEHT 77 Untt 2 JULY, 1977
- c. Information relative to each of the plates in the beltline region is shown in Tables 4 through 7.
5~ Information relative to the weld and plate material included in the vessel material surveillance program is shown in Tables 1 through 3 and:5 through 7. 121.2-2 Appendix g AMENDMENT 7y Vnit 2 ~ULY. 1S77
gggckmrnf 7 Figure gl21.2-1 p~~ 3+ID Reactor Vessel Beltline Re ion Melds D. C. Cook Unit 2 Oo Plate C5556-L Plate C5521-2 80'700 Soo 80o I80o CORE Oo Plate C5540-2 C) 270o .SOo Plate C5592-1 I80o MAX. EHD OF I.IFE WELD ORIENT. WELD LOCATIOH FLUENCE H/cm< VERTICAL l70o 4 350o 7.7 x IOI8 VERTICAL 90o 4 270o 6.3 x IOI8 CIRCUMFERENTIAL INTER. TO LONER SHELL 2.0 x IO 9 Appendix g 321. 2-3 Unit 2 SlENDHENT 77 dULY, 1977
TABLE 1 IDENTIFICATION OF REACTOR VESSEL BELTLINE REGION WELD MATERIAL Welding Weld Weld Wire Flux Process ~oal. No. ~Te Heat No. ~Te Lot No. Post Weld Neat Tr Inter. Shell Sub. Arc* MPS-1323-2F4F6 ADCOM TNMM S3986 LINDE 124 934 1125-1150'9-62 2/2 NES-PC (Vertical Seams) Inter. to Lower Shell (Circle Seam) Lower Shell (Vertical Seains) Surveillance Meld 1115-1165'F-9 HRS-FC
<Welds fabricated using both single and tandem wires Ps
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c+ fD fO CL We TABLE 2
~ Oc BELTLINE REGION WELD MATERIAL CllEMICAL COMPOSITION WELD WIRE FLUX WEIGIIT PERCENT TYPE -
IIEAT NO. LOT NO C Mn P S Si, Ni Mo Cr Co ADCOMINllM S39B6 Line 124 934 (Single Wire) .080 1.42 .019 ~ 016 ~ 36 .96 ~ 07 ~ 05 (Tandem Wire) ~ 092 1.46 ~ 019 .015 ;35 ~ 97 ~ 53 ~ 07 ~ 06 SURVEILLANCE WELD ~ 110 1.33 .022 ,012 ~ 44! ~ 97 .54 ~ 07 .055
TABLE 3 HECHANICAL PROPERTIES OF BELTLINE REGION MELD MATERIAL WELD MIRE FLUX S}IELF NDT NDT ENERGY YS UTS ELONG Bh TYPE NEAT NO. TYPE LOT NO. F 'F FT-LBB KBI KSI X X ADCOMIN19f S3986 LINDE 124 934 (Single Wire) 27* 77* 71.8 86.5 30.0 68.6 (Tandem Wire) 27* 77* 74 ' 91.2 25.5 66 0 SURVEILLANCE MELD <<40 27 77 76. 3 92. 3 24. 2 66. 7
*Estimated from surveillance weld data I
g6acl ~a~I- 7 p~p)$ /D MAXIMUM ENDMF-LIFE FLUENCE AT INNER MALL REACTOR VESSEL LOCATIONS FLUENCE (n/cm 2 ) later. Shell 7 ' x 10 (Ve'rtical Seams) Inter.,Shell to Lower Shell 2 0 x 10 (Circle Seam) Lower Shell 6 3 K 10 (Vertical Seams) Inter 6 Lower Shell Plates 20xlo Appendix g Unit 2 121.2-7 AMENDblBlT 77 JULY, ]g77
TABLE 5 IDENTIFICATION OP BELTLINE REGION PLATE MATERIAL PLATE MATUAL COMPONENT CODE NO. HEAT NO. SPEC SUPPLIER HEAT TREATMENT Inter. Shell 10>>1 C5556-2 A533B,CLol LUKENS 1650-1750'P-5HR-WQ 1550-1650 F-4 3/4 HR-WQ 1200-1300 P-5HR-AC 1100-i175'P-62 1/2 1R-PC Inter. Shell 10-2 C5521-2 .A533B,CL,1 LUKENS 1650-1750'P-4 1/2 HR-WQ
'550-1650'P-5HR-WQ 1200-1300'P-4 1/2 HR-AC 1100-1175 F-62 1/2 HR-PC Lower Shell 9-1 C5540-2 A533B, CL+ 1 LUKENS 1650-1750'P-4 1/2 HR-WQ 1550-1650'P-5HR-WQ 1200-1300'F-4 1/2 HR-AC 1100>>1175'F-62 1/2 .HR-FC Lower Shell 9-2 C5592-1 A533B, CL, 1 LUKENS 1650-1750'P-4 1/2 HR-WQ 1550-1650 F-4 1/2 HR-gQ 1200-1300'P-4 1/2 HR-AC 1100-1175'P-62 1/2 HR-PC Sunreillance Plate C5521-2 A533B,CL,1 LUKENS 1650-1750'P-4 1/2 HR-WQ 1550-1650'P-5HR-WQ 1200-1300'P-4 1/2 HR-AC 1125-1175'F- 51 1/2 1R>>PC S!
cA, ~
TABLE 6 CHEMICAL COMPOSITION OP BELTLINE REGION PLATE MATERIAL PLATE PLATE HEIGHT PERCENT CODE NO. HEAT'O. LOCATION C S Si Ni Mo Cu 10-1 C5556-2 TOP ,24 1.34 ~ 012 ~ 015 .19 .56 ~ 55 .14 BOT. ~ 21 1.38 ~ 014 ~ 014 ~ 18 ~ 58 .55 ~ 15 10-2 C5521-2 TOP ~ 22 lo28 ~ 012 ~ 016 , 18 .57 ;54 ~ 14 BOT, ~ 21 1,29 ~ 013 .015 ~ 16 ~ 58 .50 ~ 14 C5540-2 TOP ~ 21 1+31 .015 ~ 014 ~ 20 ~ 64 ~ 57 .11 BOT, ~ 19 1.34 .011 ~ 015 . ~ 18 ~ 63 .56 .10 9-2 C5592-1 TOP ~ 20 1 35 .010 ~ 015 .19 ,60 .53 ~ 14 BOT, .20 1,25 ~ 012 ~ 014 ol8 .57 ~ 50 .14 SURVEILLANCE PLATE .22 1 28 . 017 ~ 014 ~ 27 .58 ~ 55 ~ ll
TABLE 7 MECHANICAL PROPERTIES OF BELTLINE REGION PLATE MATERIAL SHELF PLATE NDT NDT . ENERGY YS UTS ELONG ~ Rh CODE NO. HEAT NO. 'P F PT-LBS KSI KSI X X 10<<1 C5556-2 0 90. 67.2 87 ' 25 ' 10-2 C5521-2 10 38 86 64.5 85 ' 25;5 9-1 C5540-2 -20 -20 110 65.8 85 7 26 ' 9-2 C5592-1 -20 20 103 70.0 88,1 24.5 SURVEILLANCE PLATE 10 38 . 86 66.4 86 ' 25.2 60. 6
U g ff'4 "/t or~~~ 7 g c INDIANA & MICHIGAN POWER COMPANY P. 0, BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 July 3, 1979 AEP:NRC:00097C Donald C. Cook Nuclear Plant Unit No ~ 1 Docket No. 50-315 License No. DPR-58 Mr. James G. Keppler, Director U.S. Nuclea~ Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
References:
(;) NRC IE BULLETIN NOS. 78-12, 78-12A, 78-128 "ATYPICAL WELD MATERIAL REACTOR PRESSURE VESSELS" (2) "COMBUSTION ENGINEERING REPORT IN COMPLIANCE WITH NRC IE BULLETIN 78-12, DATED JUNE 8, 1979
Dear Hr. Keppler:
This letter and its attachments are in response to the above referenced I.E. Bulletins as they apply to Unit 1 of the Donald C. Cook Nuclear Plant. Combustion Engineering, manufacturer of the reactor vessel for Unit 1 has submitted to the NRC, on June 8, 1979, a generic report (ref-erence 2) providing the required weld material information on all reactor vessels fabricated by them. Westinghouse and American Electric Power have reviewed the above referenced report and concluded that it represents adequately the data for the weldment material used in the reactor vessel of
Ae.t': Nii.: Uoi) 9/i
//Qcci'ng fg r3 ~ 7/P Unit 1 of the Donald C. Cook Nuclear Plant. Westinghouse has noted some discrepancies in the Combustion Engineering report..These are editorial in nature and will be submitted to the NRC as a revision by Combustion Engineering, Inc.
Very truly yours, ohn E. Dolan ice President Attachments:
- 1) Combustion Engineering letter to NRC dated June 8, 1979
- 2) Combustion Engineering review certification letter dated June 8, 1979
- 3) Westinghouse letter to AEP dated 6/25/79 cc: R. C. Callen G. Charnoff D. V. Shaller - 8ridgman R. S. Hunter R. W. Jurgensen
Hr. J. G. Keppler, Director ~ PEP: NRC 0097C egg ~'F)i~ ~i.g'- g
~ ~~/4 bc: S. J. Milioti/J. I. Castresana/T. Satyan R. F. Hering/S. H. Steinhart/J. A. Kobyra H. N. Scherer, Jr.
R. F. Kroeger J. F. Stietzel - Bridgman D. Migginton - NRC Cook Plant Region III Resident Inspector AEP:NRC:00097C R. C. Kopeiow/J. R. Jensen OC-N-6015.3.1
Q g f:pv<'pi ~~i ..i".s ~'>> g<~" ~ to i I ~
~ Comcu!ion
)QQQ PivSPOCi i'ui tiOOd
.Windsor. Gonhcc:icui 0609'ci r..r.'.irq. Ihc Telex 9 3"97 Attachment AEP:HRC:00097C 1
- ..-- POWER +gzi c l~m c ~ '7
'~ SYSTEViS p'du~~' + c>+IN June 8, 1979 LD-79-036 Hr. Harold D. Thornburg Division of Reactor Cons;i uction Inspection Office of Inspection ard Enforce.-: nt U. S. Nuclear Peguiatory Commission Washington, D. C. .20555
Subject:
I5E 8ulletin 78-12, "Atypical iteld t',aterial in Peactor Pressure Vessel ';.'elds"
Dear tir. Thornburg:
Enclosed please find;hree (3 copies of a docu",ent entitled "Infer:-'on Reque ted;y V*:- Bulletin 7o-12, ~typical 'i'eld i"-terial in Re=ctoi Pres-sure Vessel 'h'elds." This repoi t is beir," s..';.'.i tted directly to the i'=: b. Cc;:bustion Ei::-.ine-ering as perr:itt ". by::.:'olc ..n. A to the,"ullet n. It is expecto"',a holders oi Cons'ruct o:: Per;.-.i-.s and u>".ei*ating Lic:..".ses v:il', re:.ere:.ce this'eport in responding to the bulletin on the;r individual deci,e:s. Should you have any c"estior.s, please feel free to call r,",e or fl; . E. H. Kennedy of my staff a" (203)6"'3 1911, extension 2o2G. Very truly yours, COf18USTIOll EiiG!.",EER!,'lG, IilC. Licensirg l',anager AES:dag Enclosure
l .J Li C.E power Systems Tel, 615;26~ 463: Attachment 2 CprnbUstlon F ngineering. tnc AEP:HRC:00097C ~ ~ 9 ] t Q/, Vie I I Street 1 Cttgttanooga. Tennessee 3y't02 rki5&c-'.~~ H g PODER SYS I Ei',S June 8r 1979 I hereb'i certha 7S- 12 and 73- 12' ". s the record se,.rch rcruirL d b J I.E D;tile tin cot iplc"ed aild t"a>>, to theM~ be st of: 'y knot;ilcde e and ~el t ne re. o -t. I t 'E ~ n Jut%>> M r 1979 r e:ltit'ed, In: o r:"..;. >> 'tl:xe 'u "nd Etlforce:t:e::t tulle 70-1~ A t, pica 1 t'c ld la te r i a i ' Rcac tor Pressure kressels", o tie a>>aQ 'ca' e, ate 'ials uscG in the 'acr'cation o ti".'e r 0 1 lo'w'lng reactor ves el: C-E Contract r,'o.: 23366 Uti1 i ty/Si t;e: Indiana- '.ichic'an Electric Co. Donald Cook <1 N. A. Stone, Jr., tlanager Nuclear Quali". Assurance Chat tat>oo>a '.nuclear Op ration"
AEP:t(RC:00097C Westinghouse P+~h~~.y p H<N:lear Service Ow<sio<< Pri ver Systems Electric Corporation Company <axn<S pilisD<'<g'< pennsy<vzn<a 15230 June 25, 1979 AEP-79-17 Hr. J. R. Jensen t techani ca 1 Eng i iii i i ing Di v i s i on American Electric Power Service Corp. 2 Broadway Hew York, HY 1000)
Dear t'Ir. Jensen:
NRC IE BULLETIttS ='78-12 5 ;-.'78-12A "At~4 <1 t eld immaterial in Reactor P.essure ",essel 'lds" Based upon our t<< tinical evaluation of the in,ormation contained in th c n r c rep<<<<o"piled t v Co-'.bustion Engine ring, Inc. to satisfy the requir.;,.en.s ore-sented in the U '. <<uclear Regulator" Cor.:;,.ission IE Bulletins ="78-12 ~~ =7:-<2.-'.,
>lestinghouse ha" onclud"d that the 'weld raterial data and other required 'n-'-r-rt'r p nt. l,o the D.C. Cook Unit 1 reactoi vessel are included in Co.":3"5-tion Engineer ing, Inc. report.
This repoi't has I<reviously been submitted to ie U.S. Nuclear Regulatory Co:.-;.'s-sion> as eviden>> '<l by Co.",:bustion Engineering, inc. transmittal let er o-. Ju,". 1979 to <<e "S '<<<<: lear Regulatory Cor<i7ission, a copy of which is enclosed for your informatioil, Additionally, w" tiave enclosed for your files a copy of Combustion Enein crine, Inc. letter to ". ",tinghouse, dated June 5, 1979 and attached certific.=t'.on stating tha" .h~ ioner'.c report submitted to US l<uclear Regulatory conta'ns c'.a-:. for the D.C. Cool; Unit i-eactor vessel. 1 Mestinghouse au<lated the content of the subject report against the AS:E Coco and H E-Spec. r('iiiire, nts or th D.C. Cook Unit 'eactor vess 1 built '". Combustion Enoiii i in Inc. The report contains data pertaining to the D.C. Cook Unit I <,oac.. vessel ard is consi"ered to be in co."iipli=.nce with the US hRC Bull< i iris and ';<estinghouse requirements. However, some apparent errors were not< I in thc report. These discrepancies, vere broucht to t'.".e attention of C<><.,! ii:tion Engineering, Inc. and Coirbustion Engireering, Iiic. is currently ev.<t<<~tiflg tileol They have agreed to resolve the cor.:ments:o
~
Mestingnous s'<i i' lction a<id will submit revised pages for the repoi.t to ..':e Vuclear Regula'<<I v Commission and l(estinghouse at a later date.
-'1 Q Q I',r, J. R. Jersen June 25, 1979
++4'c!pm,+g AEP- 79-1 7 1
A$g 1<<j+ In addition to the data supplied by Combustion Engineering, Inc. in the subJect report,westinghouse has developed surveillance v<eldrent data. This data is contained in the following report, vihich has previously been transmitted to you: O.C. Cook Unit 1, MCAP'8047, dated Harch, 1973 As stated in their report Combustion Engineering, Inc. does not maintain archive material for the fields represented by this report. In addition, Westinghouse inventoried -our archive surveillance <;el dment material and none exists for the D.C. Cook Unit 1 reactor vessel. .In conclusion, .this letter provides assurance that the D.C. Cook Unit 1 reactor vessel is +overed in the subject report, and fulfills l!estinghouse's obligations relative to the Reactor Yessel for'y k:,ericar Electric
'ld flaterial Program contracted Poi;er Service Corporation.
E r A copy of the Co;..bustion Engineering, Inc, generic report applicable to the D.C. Cook Unit 1 reactor vessel is submitted, or your records. Sincerely, (/ F. t/oon, t!ana g e r Eastern Region 8 Ht(I Support JDC/ej at tachmen ts cc: D. Y. Shaller* R. H. Jurgensen* J. G.'ern*
*without attachment
~ I C-8 Power Systems Tel. 615(265-463; ~urn Combustion Engineering, Inc.
911 W. Main Street Chattanooga, Tennessee
~rgp~ Biv-9 ~F I +
374Q2, 8'p) I c g POWER Lt .M 5 SYSTEMS NFORMATION REQUESTED BY NUCLEAR REGULATORY COMMISSION INSPECTION & ENFORCEMENT BULLETIN NO. 78-12 "ATYPICAL WELD MATERIAL IN REACTOR PRESSURE VESSEL WELDS"
INFORMATION REQUESTED BY NUCLEAR REGULATORY COMMISSION INSPECTION & ENFORCEMENT BULLETIN NO'8-12 "ATYPICAL WELD MATERIAL IN REACTOR PRESSURE VESSEL WELDS" Prepared by COMBUSTION ENGINEERING, INC. NUCLEAR POWER SYSTEMS June', 1979
Page 1 of 4 V REACTOR PRESSURE VESSELS FABRICATED BY CR1BUSTEON ENGINEERING, INC. Ajjkc4~ + g.
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-.;. -C-E CUSTOMER ASME CODE OWNER SITE CT HO.
- =- 164 General Electirc I & VIII, 1962 Niagara Mohawk Nine Mile Point Pl
'=-264 General Electric I & VIIE, W-63 Jersey Central Oyster Creek 17765 Westinghouse III, W-65 Consolidated Edison Co. Indian Point /f2 19865 General Electric IIE, S-65 Northeast Utilities Millstone 81 2966A CEiPD - Windsor III, 1965 Consumers Public Power Palisades 3266 Wes tinghouse EII, W-65 Public Service of N. J. Salem !31 3366 Westinghouse III, W-65 Consolidated Edison Co. Indian Point 83 6866 Westinghouse III, W-65 Carolina P&L Robinson 82 21366 General Electric III, W-66 Consumers Public Power Cooper Site 66 General Electric EIE, W-66 Boston Edison Co. Pilgrim 21566 General Electric III W-66 Power Authority State N.Y. Fitzpatrick 23066 Westinghouse III, W-66 Pacific Gas & Electric Diablo Canyon 81 23366 Westinghouse III, W-66 Indiana-Michigan Elec. Co. Donald Cook i'P.l 71166 CEMD - Windsor IEI, W-67 Omaha Ft. Calhoun 2067 Westinghouse III, 'rT-66 Public Service of N. J. Salem 82 2167 Westinghouse III, S-71 Duke Power Company McGuire 81 2667 General Electric III% S-69 Detroit Edison Fermi 2867 General Elect'ric EII, W-69 Commonwealth Edison LaSalle 3067 General Electric IIE, S-68 Long Island Lighting Co. Shoreham 3167 General Electric IIE, W-66 Southern Services H tch 81 i 67 CENPD Windsor III, W<<67 Baltimore Gas & Electric. Calvert Cliff 73167 CENPD Windsor EEI, W-67 Baltimore Gas & Electric Calvert Cliff 74167 CENPD Windsor EII, W-67 Florida Power & Light St. Lucie I
SUMMARY
OF WELD MATER ID TEST WIRE/FLUX MELDING MATERIALS NmmER AND DATES OF TESTS WIRE/FLUX OR ELECTRODE C-E REFER . WIRE/ELECTRODE FLUX WELD DEPOSIT TEST PLATES CODE ATTACHED HEAT/LOT NO. OF NO. NON-CONFORM. VENDOR TYPE VENDOR TYPE LOT NO. DATE(S) REPORT NO. TESTS ADCOH I.IHDE 1092 3947 4-1-70 Ml. 37 12008'05414 RACO 3 1 INDE 1092 '947 Ml. 37 RACO 3 33A277 LIHDE 1092 3947 4-8-70 M1.38 Reid-Avery )BQf 305424 LINDE 1092 3947 4-10-70 ill.39 Reid-Avery IIHH 305414 LINDE 1092 3951 5-4-70 M1.40 ADCOH 12008 LIHDE 1092 3951 5-11-70 M1.41 Reid-Aver HMH 305414 LINDE 1092 3951 M1.41
-vr 305414 LINDE 1092 395f) 6-2-70 M1.42 Reid-Aver 1P3571 LINDE 1092 3958 NA M1.42 Reil-Aver I IHH 1P3571 LINDE 1092 3958 M1.43 6-9-70 l)1.43 Reid-Aver 1P 3571 LIHDE 1092 3958 Reid-Avery HMI 1P3571 LIHDE 1092 3958 6-3>>70 Ml.44 Reid-Avery )IHH 305414 LIHDE 1092 3958 6-3-70 M1.44 ADCOH 27204 LINDE 124 3687 7-11-67 E1.01 Nh IIHH 51989 LINDE 124 3687 E1.01 ADCOH )IHM 27204 LIHDE 124 3687 10-10-67 E1.02 Reid-Aver IBIH 348009 LIHDE 124 3687 2-28-68 E1.03 Reid-Avery IBBI 349009 ).IHDE 124 3688 2-7-69 El.04 NA IBIH A-8746 LIHDE 124 3688 5-7-69 E1.05 NA IB IH A>>8746 I.INDE 124 3878 9-10-69 El.06 Reid-Aver IBIH 33A277 LIHDE 124 3878 10-29-69 E1.07 Page 6 of 21
- WIRE/FLUX INDEX Heat of Wire ~Plux T ne Lot Test Results 646B428 Linde 80 8174 Page 1 661H577 Linde 80 8174 Page 1 86054-B Arcos B-5 4D4F Page 2 4D5F 1248 Areas B-5 4K13F Page 3 5458 Linde 80 8208 Page 4 V-5214 Areas B-5 5613F Page 5 39B196 Linde 1092 3617 Page 6 34B009 Linde 80 8405 Page 7
'7204 Linde 1092 3724 Page 8 12420 Linde 1092 3724 Page 8 13253 Linde 1092 3724 Page 9 13253 & 12008 Linde 1092 3774 Page 10 20291 Linde 1092 3791 Page 11 7114 Linde 1092 3833 Page 12 8746 Linde 1092 3854 Page 13 IP2809 Linde 1092 3854 Page 14 IP2815 Linde 1092 3854 Pages 15 & 16 21935 Linde 1092 3869 Pages 17 thru 19 33A277 Linde 1092 3869 & 8651 Pages 20 & 21 305424 Linde 1092 3889 Pages 22 & 23 O 305414 IP3571 885T40 90099 Linde Linde Linde Linde 1092 1092 0091 0091 3947 3958 3922 3922 Pages Pages Pages Pages 24 26 28 30 25 27 29 31 35C191 Linde 0091 3922 Page 32 90136 Linde 0091 3977 Pages 33 & 34 10120 Linde 0091 3999 Pages 35 & 36 10137 Linde 0091 3999 Pages 37 & 38 6329637 Linde 0091 3999 Pages 39 & 40 51874 Linde 0091 3458 Pages 41 & 42 51876 Linde 0091 3458 Pages 43 & 44 51907 Linde 0091 3458 Pages 45 & 46 606L40 Linde 0091 3489 & 3458 Pages 47 thru 49 51922 Linde 0091 3489 Pages 50 & 51 .51923 Linde 0091 3489 Pages 52 & 53 51912 Linde 0091 3490 Pages 54 thru 56 3P4767 ;Linde 0091 3490 Pages 57 & 58 83640 Linde 0091 3490 ~
Pages 59 & 60 83642 Linde 0091 3536 Pages 61 & 62 83653 Linde 0091 3536 Pages 63 & 64 83648 Linde 0091 3536 Pages 65 &'6 4P5174 Linde 0091 1122 Pages 67 & 68 83637 & 83650 Linde 0091 1122 Pages 69 thru 71 5P5622'3646 Linde 0091 1122 Pages 72 & 73 Linde 0091 1122 Pages 74 & 75 2P5755 Linde 0091 1122 Pages 76 & 77 4P6052 Linde 0091 0145 Pages 78 & 79 87005 Linde 0091 0145 Pages 80 & 81 87600 Linde 0091 0145 Pages 82 & 83 88118 Linde 0091 0145 Pages 84 & 85
fROM ~OhTC VFelding Material Qualification Metallurgical Research aad
-... to Requirements of ASME .Development Department
. Section
' A-32255 IIl - Chattanooga
~
. 810560 June 9, 1970 I
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The following test data is for 3/16" diameter bhre wire, type B-4. MOD.,
. heat number 1P3571 (tandem), flux type 1092, lot number 3958.
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' vreld deposit was made using the above heat of wire and lot of Qux. 'A'eldinq
~
eras done in accoraar ce with C. E.'A'eiding Procedure Specificat'on SAA"33-H3 ~
~,'he completed v eldment was given a post weld heat treatment of 1150'F 25'F
. for 40 hours ard iurnace cooled to 600 F.
~ ~
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Charov V-Notch impacts est Code Pt bs. + 10'F Reouirement s VZ 79, 68, 64 '0 Ft.,Ebs. @+10 F
~ ~ ~
AllWeld Metal . 505 Ter.sile 0 Yield Strength Ultimate Tensile Elongation in Reduction of ~..... KSI Strenath KSI 2 II 'rea /0
~ ~ .'0.5 86.8 27.0 67.0
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INDIANA II MICHIGAN POWER COMPANY P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 June 1, 1979 AEP:NRC:00097 Donald C.Cook Nuclear Plant, Unit No. 2 Docket No. 50-316 License Nos. DPR-74 Mr. James G. Keppler, Director U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
References:
(1) NRC IE BULLETIN NOS. 78-12, 78-12A, 78-12B ATYPICAL WELD MATERIAL IN REACTOR PRESSURE VESSELS (2) "CHICAGO BRIDGE 5 IRON COMPANY REPORT IN COMP I-ANCE WITH THE NRC BULLETINS 78-12 AND 78-12A", DAT-D APRIL 24,1979
Dear Mr. Keppler:
This letter and its attachments are in response to the above referenced I.E. Bulletins as they apply to Unit No. 2 of the D.C. Cook Nuclear Plant. Chicago Bridge 8 Iron (CB8 I), manufacturer of the reactor vessel for Unit 2, has submitted to the NRC, on April 24, 1979, a generic report (reference 2) providing the required weld material information on all reactor vessels fabricated by CBE I. Westinghouse and American Electric Power have reviewed the above referenced .report and concluded that it represents adequately the data for the weldment material used in the reactor. vessel of Unit No. 2 of the Donald C. Cook Nuclear Plant. Weld-ment material that might be used for verification purposes, is available in the archives of the Westinghouse Electric Corporation.
Hr. James G. Keppler, Director AEP:NRC:00097 y As stated in our letter No, AEP:NRC:000978 dated May 21, 1979, the above information for Donald C. Cook Unit No. 1 reactor vessel will be submitted by July 2, 1979 Very truly yours JED:em ohn E. Dolan ice President Attacnments:
- 1) CBIII review certification letter to the NRC dated 4/24/79
- 2) C88 I letter to the NRC dated 4/24/79
- 3) Westinghouse letter to AEP dated 5/23/79 cc: R. C. Callen G. Charnoff D. Y. Shaller-Bridgman R. W. Jurgensen
Jpp 9 Hr. J. G. Keppler, Director AEP:NRC:00097 bc:S.J. Mi lioti/J. I, Castresana/T.Satyan R. F. Hering/S. H. Steinhart H. N. Scherer, Jr. R. F. Kroeger J. F. Stietzel-Bridgman D. Higginton-NRC Cook Plant Region III Resident Inspector AEP:NRC:00097 DC-N-Gois.a t R. C. Kopelow/J. Jensen
'Chicago Bridge & Iron Company
'rat: MEiir r'>+
8="00.-"~iroank". north goo x '00rg oUs:on r ad Roustc~. T~;;as 77040 The documentation and information required by NRC Bulletins 78-l2 and 78-I2A, and Westinghouse PO //546-MVC-40I 945-MN for CBI Contract $$ 68-3262 Vessel D. C. Cook II are contained in the attached report. Welding consumables were re-reviewed against the original requirements in accordance with the above listed documents. No deviations were found. Based upon our records, I certify, to the best of my knowledge, this report is correct. Ralph E. Kelley Date Manager, CQA Services
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ATTACHMENT 2 Akir4~jP~ WP
' Ega'p'hicago Bridge 5 tron Company .6000 Fairbanks north Houston road p o box 40066 Houston, Texas 77040 telephone 7i3. 466 7661
.April 24, 1979 Office of Inspecti'on & Enforcement'.
S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. G. W. Reinmuth RE: NRC BULLETINS 78-12 6 78-12A Gentlemen: .In accordance with the above listed Bulletins and requirements from Westinghouse and General Electric, enclosed is one copy of our report. il This report includes information from all completed Reactor Vessels constructed by Chicago Bridge & Iron Co. Very truly yours, CH CAGO BRIDGE 6 IRON CO. Ralph E. Kelley, M ager CQA Services Houston Operations REK:mks Enclosure I'
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0 I ATTACHMENT 3 Nuctear Service Oivision Westinghouse Water Reactor Electric Corporation Olvislons Box 2728 Pittsburgh Pennsytvanta 15230 May 23, 1979 AEP-79-10 Mechanica E gineering Division American 1 ctric Power Service Corp. 2 Broadway New York, NY 10004
Dear Mr. Jensen:
NRC IE Bulletins 878-12 & 878-12A "At ical Weld Material in Reactor Pressure Yessel Melds" ~ Based upon our technical evaluation of the information contained in the generic report compiled by Chicago Bridge & Iron Company to satisfy the requirements presented in the U.S. Nuclear Regulatory Commission IE Bulle-tins f78-12 and f78-12A, Mestinghouse has concluded that the weld material data and other required information pertinent to the D.C. Cook Unit 2 reactor vessel are included in Chicago Bridge & Iron's report. This report has previously been submitted to the U.S. Nuclear Regulatory Commission, as evidenced by Chicago Bridge & Iron Company's transmittal letter of. April 24, 1979 to the U.S. Nuclear Regulatory Commission, a copy of which is enclosed for your information.'dditionally, we have enclosed for. your files a copy of Chicago Bridge & Iron Company's letter to Westinghouse, dated April 24, 1979, providing further confirmation that the generic report prepared by"'vendor includes records pertaining to the O.C. Cook Unit 2 reactor vessel. The Chi'cago Bridge & Iron certifications stating that the report contains data for the O.C. Cook Unit 2 reactor vessel is included in Part 2 of the report. Hestinghouse audited the subject report against the ASME "and M E-Spec. requirements for the D.C. Cook Unit 2 reactor vessel built by Chicago Bridge & Iron. The report contains data pertaining to the D.C. Cook Unit 2 reactor vessel and is considered to be in compliance with the U.S; Nuclear Regulatory Commission bulletins and Westinghouse requirements. In addition to the data supplied by Chicago Bridge & Iron Company in the subject report, 'llestinghouse has developed surveillance weldment data. This data is contained in the following report, which has previously been transmitted to you: O.C. Cook Unit 2, MCAP-8512, dated November, 1975
7 J. R. Jensen 2 May 23, 1 79 As stated in their report Chicago Bridge 5 Iron Company has no archive material for the welds represented by this report. Westinghouse inven-toried our archive weldment material which could be used for verification purposes on the O.C. Cook Unit 2 reactor vessel. This material consists of one full thickness weldment made up of weld wire from heat number 53986 and Linde Flux 124 from lot number 934. In conclusion, this letter provides assurance that the O.C. Cook Unit 2 reactor vessel is covered in the subject report, and fulfills Westinghouse's obligations relative to the Reactor Vessel Weld Material Program contracted for by American Electric Power Service Corporation. A copy of the Chicago Bridge and Iron generic report applicable to the D.C. Cook Unit 2 is submitted for your records. Sincerely, oon, Manager Eastern Service Region JDC/pl Attachments cc: O.V. Shaller R.W. Jurgensen J.G... Kern
i A/4.g,p y 3 I'~ >
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wp ~ /P Chicago Bridge 5 Iron ompany GGGO Fairb".."ks ".<"". Houstct: roa" p 0 Qox +CGGG HousIcsI. TexGS i.G-'0 1~ il)0e i sy, t C3E3K PLANT MED RECQRD - MED GQPY CHICAGO BRIDGE & IRON COMPANY ENGINEER ~+~ DATE REPORT IN COMPLIANCE WITH THE H PLANiT L(r ETIAM'iE CATE TC PI 'iIIT, C3 NQN P'Rt'Al<~! IT MININUH RETENTION YRS. NUCLEAR REGULATORY COMMISSION BULLETINS 78-12 & 78-I 2A Report prepored by V-2I-7 Ralph E. Kelley Date Mgr., CQA Services IL
PART I LIST OF REACTOR VESSELS INCLUDED WESTINGHOUSE VESSELS CBI CONTRACT VESSEL 68-3262 DC Cook II 68-3780 Trojan 7I-2631 Virgil C. Summer I 7I-2632 Shearon Harris I 7I-2633 Shearon Harris II GENERAL ELECTRIC VESSELS 9-5624 hhonticell o 9-620I Vermont Yankee 68-247I Brunswick I 68-2472 Brunswick I I 68-3331 Susquehanna I 68-3332 Susquehanna II 69-2967 Duane Arnold 69-4824 Quad Cities II (CBI Portion) 69-4962 Peach Bottom II (CBI Portion) 69-5128 Peach Bottom III (CBI Portion) 69-540I Limerick I 69-5402 Limerick II 69-557I Zimmer I 73-6735 CI inton I
Chicago Bridge 8 Iron Company 3.CO.=circa.".r.. ncr ". Hcus:c.". road
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h ~ i) t The documentation and information required by NRC Bulletins 78-l2 and 78- I 2A, and Westinghouse PO 8546-MVC-40 I 945-MN for CBI Contract /3 68-3262 Vessel D. C. Cook II are contained in the attached report. Welding consumables were re-reviewed against the original requirements in accordance with the above listed documents. No deviations were found. Based upon our records, I certify, to the best of my knowledge, this report is correct. Ralph E. Kelley Date Manager, CQA Services
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I NUCLI:.Rlh RECORD INDEX D ESC(( IPT ION I vrllulrut hl umber I Wllllll)Cf 0 I PrlgrS WIRE WIRE WIRE FLUX FLUX SIZE HEAT NO. RUN OR LOT TEST NO. SPECIFICATIONS M (t/i b (- (II I J P r'C Ib'c + l( .w t g tmtQ lo 2 ~% OCQ z
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MED RECORD MEO COPY Dilte SECTION P>>t~ ENGINEER (. I. I":I I I'r I I I I i i I Si(I>>lro)< C
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CHICAGO BRIDGE & IRON COMPANY 1500 N. 50TH ST. P. 0. BOX 277, BBRMlNGHAM, ALABAMA35202 vv
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TWX B10-733 3554 -: "-\ Western Union WUX
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~... Area Code: 205 595-1191 '
I'ERTIFICATE OF ANALYSIS I PURCHASE ORDER NUMBER: . MECHANICAL TESTS: .
.'est Number: PT 200 A Heat Treatment 50=.Hours 9 1125/1150 Type Electrode: Adcom lNtlM/Linde 124 Farenheit Trade Name: 'dcom ltlH;i >(ire Tensile Properties 9 Room 3/16" Type: ,505" Temp.'..':..'iameter:
Flux Lot Number: 3877-Run 934-Linde" 12CJTS 89,000 PSI ~
!;.::~-:,t teire Heat Number: S3986 " YLP 70,100 PS I X Elongation in 2 'inches = 23.5 ~
CHEMICAL TESTS X Reduction of. Area;:= 65:.. '.. Carbon. Manganese.
~
.101 1.49
.,' Impact Proper ties Type: Charpy Vee Notch Orientation: To Iield Direction ..
Ch romi um. Ni eke 1...
.12
.92 )))'.-,
Test Temperature + 1-0' 1 li Si con. .41 Foot- lbs. 67:.5, 6~, 65
'Shear Columbium. ,004 5 60, 60, 55 Tantalum... Lateral Expansion 61, 58, 52
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Molybdenum, e ~ .53 Tungsten.'. Copper. .05 v Titanium. .~ Phosphorus . ~ ~ .022 Sulfur. '. .016
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CHICAGO BRIDGE AND IRON COMPANY Birmingham Materials Lab oratory ',.: ', ', i;. e
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~ 0 ~ 8OX 13308, MEMPHIS, TENNESSEE 3e't13 CERTIFICATE OF NALYSIS 901 947-311'a Purchase Order Number: MECHEiNICAL TEST RESULTS M30506-3262/3780 Test Number: WO 5 337C (Tandem Ni re) Heat Treatment 1150'F +25'-50'F Type Electrode: Adcom 1N)~J"./Linde 124 for 62 1/2 Hours (20 x 150) Flux Tensile Properties Trade Name: Adcom lNMM Type: . 505 'y Electrode Diameter: 3/16" UTS 92 g 000 PS I Lot L~umber: YLP 78,800 PSI Heat Number: S39 86 % Elongation in 2 inches = 26i Flux Batch Number: Run 934 Reduction of Area = 57.3 Lot 3878 CHEI1IC..L TEST RESULTS Impact P ope r tie s Carbon ~ ~ ~ ~ ~ ~ ~ ~ . 089 Type: Charpy Vee Notch Manganese..... l. 47 Orientation: ~ to Held Direc"ion hromium...... Test Temperature +10'F c)wel ~ ~ ~ ~ ~ o o~o .90 Foot Lbs. 39, 53, 38 Sl 1 leon
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(gg.rI'p;(. ~!~ <!lgi'!e.i ~KAn~a(V CFIICAGO RRIDGH Ez, C ~ IH,ON COMPANY HE~5 ++1~ P ~ O. 8OX 13308e MEMPHIS'EtlNESSEE 38113 hga CERTIFlCATE OF ANALYSIS gp /+ /0 Purchase Order Number: MECHANICAL TEST RESULT M30506-3262/3780 Test Number: NO I337C (Single Wire) Heat Treatment 1150'; +25'-50'F Type Electrode: Adcom 1NMM/Linde 124 for 62 1/2 Hours Tensile Properties Trade Name: Adcom 1NtB1(20 x 150) Flux .505"p Type: Electrode Diameter: 3/16 "gf UTS 89, 500 P -I Lot Nulnber: YLP 74,300 PSI Heat Dumber: S3986 4 Elongation in 2 inches = 27% Flux Batch Number: Run 934 Reduction of Area = 675 Lot 3878 % CHEt4ICAL TEST RESULTS Impact Properties Carbon o ~ ~ ~ ~ ~ ~ ~ .076 Type: Charpy Vee Notch Manganese..... l. 44 Orientation: to Neld Direction hromium .10 Test Temperature +10'F. ckel......g. .81 Foot Lbs. 50, 49, 62 Silicon.......J .46 Lateral Fxpansion 45, 44, 53 Columbium..... % Shear 35, 35, 40 Tantalum...... Mol; bdenum.... .50 Tungsten....'.. Coppel o ~ Tit.anium......
~ ~ ~ ~ ~ ~ .06 III ofThisthe material conforms to AStlE CODE, Paragraph SECTIOtl N511.3 Phosphorus'. . 026 Sulfur........ .017 Vanadi.um......
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CitiCAGO HaiDC;8 tt: IRON CoxIpAi> 2TT. BtRMtN GRAM. At-ABAMA 35202 Y'500 N 50TH 5 T. P. O. BOX TWX 810-733-3654 Western Vnion-WUX Area Coae: 205 595-1191 CERTIFICATE OF ANALYSIS PURCHASE ORDER NU<BER: MECHANICAL TESTS Tes t Number: PTa200-Single '.lire Heat Treatment62-1/2hours91125/1150 Type Electrode: Adcom Inmm/Linde 124 Trade Name: Adcom Inmm ';lire Tensile Properties At Room TemPeratur Di arne ter: 3/16 "g 0 505 llg Lot Number: 387>" Run 934" Li.nde 1%".:,'UTS '6,500 'lire Heat Number: S-3g 86 YLP 71,800 Elongati on i'n 2 inches= 30.0" CHE)I I CAL TESTS Reducti on of Area. = 68.'6 Nickel Carbon. 0.080. Impact Prooertit s Type: Ch a r Py Vee tlo tch Manganese. Chromi um.
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CHICAGO BRIDGE N PROX CO>rPAr'V 1500 n BOTH ST. P. 0 BOX 27'7. BiRMINQ>AM, ALABAMA 35202 TWX 810-733-3654 Wesiern iJn.on WVX Area Cooe 205 595-119'i CENTI FI CATE OF Ai'lALYSIS PURCHASE ORDER tlU:<GER'ECHAtll CAL TESTS Tes t tlumber: PT g2'30-Tanden ':lire Heat Treatment 62-1/2 hours 31125/ Type Electrode: Adcom Innm/Linde 124 1150eF Trade tlar. e: Adcom Inmm ';li re Tensile Propertiesht Room Temperature Diameter: 3/16 "g ~ Type: 0,5'35 "9 Lot. Number: 3876- Pun 934T Lande'~41'.1'~UTS 91,2.'3:3 ---" ';li-e Heat Number.: S 39B6 -YLP 74,700 Elongation in 2 inches=25.5', CHEMICAL TESTS Reduction of Area. =66.0 h Carbon. 0.092 Impact Properties Manganese. 1.46 Type: Charpy 1t'ee !totch Ch romi um. 0.07 Ori entati on: 4 to Me] d Oi recti on Nickel. 0:97 Test Temperature t'~ us, 10'F Si li con. 0. 35 Foot- lbs. 41- 4 5-. 46 Columbium. X Shear 50-55-55 Tan ta um. 1 Lateral Expansion 49-44-41 Molybdenum... 0.53 Tungs ten. Copper. 0. 06-Titanium. Phosphorus... 0.019 Sul fur. 0.015 Vanadium. I ron. Schae f f le r Fe rri te. This material conforms to Section II I of the AS;!E CODE, Paragraph .'l511.3 CHICAGO BRIDGE ANO IRON COMPANY Birmingham Materials Laboratory P By H ~4M ~~ g Date 5-/Z-8 In charge of Testing for Materi als Evaluati o'
NUC L EAR R ECORD IND EX 0 ESC R IPT ION Oocument Number BARE WIRE AND Number of Pages FLUX CORED WIRE ape El'ectrode Size Heat No. Lot No. Specifications At c-csrt I NrrtM x'AB.L-4 gl ~LC S L+Q(hz Rat.C COPIES of documents covered by this index are ccr tificd to be true copies Office Code Date Class>f>cat>on Signature Foidcr of Contract Num! >c>
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POST OFFICE BOX 2SECS - PHOI<C C43 1121 CUSTOACE( S ORO .R NO. ADCO'.'RDER NO. DATL SHIPPED SPECIF ICATION H-102('s01 761 !l- I'3- 'P? EO Chicaco Bride e 6 Iron MAR)~ED: ITE t:< CONSlSTI:lG OF ,'zc '- TO Box 13308 3C'32" x 36" 1Y,."-~1 Yse..p.lis, Tenn. 36113 GE<vTLEie'sE>a'".E IHEra BY CFP<TIFY THAT fdATERIAL f'<EFEP.:.ED 10 ABOVE COi~.'FOl<i"<S.TO TH PHYSICAL AlaaD CIIEfealCr<L TESTS AS FOLLO': S hf!D IS lfi ACCOr<DAf CE 'V!ITH SPECIFICATIOs.'S:- I IE C. a:.. ! s . ! s. ! o. ! c . ! e:. Cu. t,;n Fc. Al Ca 966 ~ 16 1.97! . 07 !. 012,010 ..010 !1. 07 .03 .006 ~ 55 a a"
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/II Reed: Sg. tpO, Resy....,....,. Xtt. Nuclear Services f.y'ox Westinghouse lstater Reactor Person: tote@ation Division Electric Corporatfon Oivisions 2128 Pittsoury Pennsylvania 15230.2l28 AEP-85-641 June 14, 1985 Mr. H. P. Alexich, Vice Presid CIerlct t" caaK PLANT and Director Nuclear Operations MED RECORD - MED COPY American Electric Power Service Corporation SEGTIQN One Riverside Plaza ENG1NEER Columbus, Ohio 43216. DATF El PLANT LIFETIME AHERICAN ELECTRIC POHER SERVICE CORPORATION D. C. COOK UNIT I DATE TO PLANT,~ ~
Reactor Vessel Beltline Reaion Held Chemistr Cl NON PERMANENT MINIMuMRETENTIDH YRS.
Dear Hr. Alexich:
A review of the weld wire and flux used to fabricate the weld seams in the core beltline region of the D. C. Cook Unit I reactor vessel was conducted per the request of D. Hafer of American Electric Power Service Corporation to determine the as deposited copper, nickel and phosphorous content of the as deposited weld seams. The circumferential girth seam betwe n the intermediate and lower shell is considered to be the limiting weld seam in the vessel. This seam was fabricated with weld wire heat number IP3571 and Linde 1092 flux lot number 3958. Eight separate chemica'I analyses are known to have been performed on this combination of the wire and flux and the results are presented below: Source Cu Ni P CE Held gua1 ification Test (Single Hire) .40 .82 . 017 CE He I d gual ification Test (Tandem Hire) .37 .75 .01 T Kewaunee Uni rradiated Surveillance Held .20 .77 .016 Maine Yankee Unirradiated Surveillance Held .36 .78 .015 Maine Yankee Irradiated Charpy Specimen .25 .70 .030 Maine Yankee Irradiated Charpy Specimen .25 .66 .020 Maine Yankee Irradiated Charpy Specimen .33 .71 .040 Haine Yankee Irradiated Charpy Specimen .33 .70 .040 Average .31 .74 .024 Sased upon the above data, it is Hestinghouse's recommendation that the average of the ab'ove data points be used for the Cu and Ni content, since this would be more realistic than using any single data point. This approach has been accepted by the NRC on other applications.
0 AEP-85-641 Mr. M. P. Alexich June 14, 1985 A+gchmenT IO p~ Z,qis ~ The phosphorous content reported for the irradiated specimens considered to be highly suspect. Hestinghouse considers the average of the four unirradiated values (.016 WTX) to be a realistic phosphorous content for the weld. j The longitudinal weld seams in the hei tline region of the vessel were made with a tandem submerged arc.process using weld wire heats 12008 and 13253 with Linde 1092 flux lot 3791. Ho as deposited weld chemistry exists for this combination o'f wi'res and flux. Four other tandem welds which contained wire heat number 12008 showed as deposited copper contents of 0.19 to 27'X. The surveillance weld which was made from wire 13253 and, Linde 1092 flux lot 3791 and which has a copper content of 0.27'L is considered to be highly.,'.representative of the longitudinal weld seams and the use of its chemi str$ for the longitudinal weld seams appears appropriate. The application of new copper and nickel values to the beltline region girth weld seam of the D. C. Cook reactor vessel will not result in the vessel exceeding the PTS screening limits imposed by the NRC. Please call should you require more information Very truly yours, g A. P. Suda, Manager Great Lakes Area Projects Department APS/debi 4496f:12 cc: M. P. Alexich, 1L D. Hafer, 1L N. G. Smith, .1L J. Feinstein, 'lL
) gyi iS le lexical cc:- 4r P. A T. 0. Argenta UNITED STATES P. A. Barrett NUCLEAR R EGULATORY COMMISSlON S. J ~ Brewer WASHINGTON, O. C. 20555 J. G. Feinstein S. P. Kiementowicz June 9, 1989 R. F. Kroeger J. F. Kurgan 50"RR -.;-'. 0. H. Malin Docket No. J. J. Markowsky Mr. Milton P. Alexich, Vice President R. I. Pawliger Indiana Michigan-.Power Company J. B. Shinnock S. H. Steinhart c/o American Electric Power Service 0. H. Williams, Jr. Corporation 1 Riverside Ohio 43216 Plaza-'olumbus, Dear Mr. Alexich.
SUBJECT:
AMENDMENT NO. 126 TO FACILITY OPERATING LICENSE NO. DPR-58 (TAC NO', 71062) The Commission has issued the enclosed Amendment No. 126 to Facility Operating License No. OPR-58 for the D. C. Cook Nuclear Plant, Unit No. l. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 14, 1988 and supplements dated December 30, 1988, and June 5, 1989. This amendment, revises the TSs to allow operation of future reload cycles of O. C. Cook Unit 1; at reduced pimary coolant system temperature and pressure co'nditions. The reduced temperature and pressure (RTP) conditions will decrease the steam generator U-tube stress corrosion cracking of the type observed at D. C. Cook Unit 2. A copy of our related Safety Evaluation is also enclosed. Notice of Issuance wi 11 be inc1uded in the Commission' biweekly Federal ~Re ister notice. Sincerely, tC!LI,u.Q..1 Qjh.Jig John F. Stang, Project Manager Project Directorate III-1 Division of Reactor Projects " III, IV, V 8 Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- l. Amendment No. 126 to DPR-58
- 2. Safety Evaluation cc w/enclosures:
See next page
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o4 ~o UNITED STATES NUCLEAR REGULATORY COMMISSION X WASHINGTON, D. C. 20655 3 r 0 /~ 1 ~
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. INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 126 License No. DPR-58
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Indiana Michigan Power Company (the licensee) dated October 14, 1988 as supplemented December 30, 1988, and June 5, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;
- 8. The facility wi 11 operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health.
and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2. C.(2) of Facility Operating License No.
DPR"58 is'ereby amended to read as follows: Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 126, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Lawrence A. Yandell, Acting Director Project Directorate III-1 Division of Reactor Projects-III, IV, V 8 Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 9, 1989
Hr. Hilton Alexich Donald C. Cook Nuclear Plant Indiana Michigan Power Company CC: Regional Administrator, Region III Mr. S. 8'rewer U.S. Nuclear Regulatory Commission American Electric Power 799 Roosevelt Road Service Corporation Glen Ellyn, Il 1 inois 60137 1 Riverside;Plaza Columbus, Ohio 43216 Attorney General Department of Attorney General 525 West Ottawa Street Lansing, Michigan 48913 Township Super visor Lake Township Hall Post Of fi ce Box 818 Br i dgeman, Michigan 49106 W. G. Smith, Jr., Plant Manager Donald C. Cook Nuclear Plant Post Office Box 458 Bridgman, Michigan 49106 ,U.S. Nuclear Regulatory Commission Resident Inspectors Office 7700 Red Arrow Highway Stevensvil le, Michigan 49127 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W. Washington, OC 20037 Mayor, City of Bridgeman Post Office Box 366 Bridgeman, Michigan 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, Michigan 48909 Nuclea~ Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of. Public Health 3500 N. Logan Street Post Office Box 30035 Lansing, Michigan 48909
~4 c+~ Illy~ I O~ UNITED STATES A O NUCLEAR REGULATORY CQMMlSSION C WASHINGTON, D. C. 20555 e
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.126 TO FACILITY OPERATING LICENSE NO. OPR"58 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT UNIT NO. 1 DOCKET NO. 50-315
- 1. 0 INTRODUCTION By letter dated October 14, 1988, as supplemented December 30, 1988, and June 5, 1989,the Indiana Michigan Power Company (the licensee) requested an amendment to the Technical Specifications (TSs) appended to Facility Operating License No. DPR-58 for the Donald C. Cook Nuclear Plant, Unit No. 1. The proposed amendment would permit the operation of future reload cycles of Unit 1 at reduced primary system temperature and pressur e conditions. The reduced temperature and pressure (RTP) conditions will decrease the steam generator
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U-tube stress corrosion cracking of the type observed at the D. C. Cook Nuclear
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I Plant, Unit 2.
~
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~ The licensee's contractor (Westinghouse) has determined that
~ ~
this RTP program should more than double the time to reach a given level of .
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steam generator U-tube corrosion in comparison to the original temperatures and
~ ~ ~ ~
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pressure. D. C. Cook, Unit 1 is presently licensed to operate at 3250 MWt, which is rated thermal power defined by Definition 1.3 of the Technical Specifications. Some transient and accident analyses are performed at" a higher power level to position Unit 1 for a potential power uprating. However, not all of the analyses have been performed at this higher power level. The small break loss-of-coolant accident (LOCA) analysis was, for example performed at a power of level of 3250 MWt with the high head safety injection cross-tie valve shut and at 3588 MWt for all other analyzed plant conditions. The staff's review of the RTP program for Unit 1 did not consider any issues related to a future power uprating. The licensee performed analyses and evaluations to support the RTP program for D. C. Cook, Unit 1. The licensee's efforts addressed full rated thermal po~e~ operation (3250 MWt) with a range of vessel average temperature between 547 F and 576.3 F. Two discrete values of the pressure, 2100 psia and 2250 psia, were used in the analyses and evaluations. The analyses and evaluations support a maximum average tube plugging level of 10K, with a peak steam generator tube plugging level of 15K. The licensee will select the desired operating temperature and the pressure on a cycle-by-cycle basis. The licensee performed the safety analyses and evaluations at conservatively high power levels and high primary system temperatures in order to position
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both of the O. C. Cook units for future power uprating and in order to support
~
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potential future operation of Unit 2 at reduced temperatures and pressure.
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The potential uprated power for Unit 1 that is partially supported by this analysis and evaluation is 3425 MWt, which corresponds to a reactor power level of 3413 MWt. The design power capability parameters are given in Table 2. 1-1 of Reference 2.
- 2. 0 EVALUATION
- 2. 1 NUCLEAR STEAM SUPPLY SYSTEM NSSS
- 2. 1. 1 Lar e.and Small Break LOCA Anal ses The licensee performed a large break LOCA analysis using the 1981 version of the Mestinghouse ECCS Evaluation Model, which uses the BASH computer code.
The analysis assumptions include a total peaking factor, F , of 2. 15, a hot channel enthalpy rise factor, F-delta H, of 1.55, 10K safety injection flow degradation, a reactor power level of 3413 MWt, and 15K uniform steam generator tube plugging level. A range of hot-leg temperatures of 580. 7'F to 611. 2'F and a range of cold-leg temperatures of 513. 3 F to 546. 2'F, consistent with the temperature range of the RTP program, were considered in the analysis. In the analysis, the reactor coolant system pressure was varied to justify plant operation at either 2100 psia or 2250 psia. A large-break LOCA analysis was also performed with the RHR. cross-tie valve closed. For this case, a reduced core power of 3250 NMt was used to compensate for the reduction in safety injection flow caused by the closed RHR cross-tie valve. For those limiting pressure and temperature conditions which produced the largest peak clad temperature, a full break spectrum of discharge coefficients was performed. The limiting break size was determined to be a cold-leg guillotine break with a discharge coefficient, C , of 0. 6, a hot-leg temperature of 611. 2 F and a primary system pressure o) 2250 psia, assuming maximum safety injection flow. The peak clad temperature was calculated to be 2180. 5'F. Based on these results, the requirements of 10 CFR 50.46 have been met for the Unit 1 large-break LOCA analysis. The licensee performed a small-break LOCA analysis using the Mestinghouse small-break ECCS Evaluation Model, which uses the NOTRUMP code. The analysis assumptions included a total peaking factor of 2.32, a hot channel enthalpy rise factor of 1.55, safety injection flow rates based on pump performance curves degraded 10K below design head and including the effect of closure of the high head safety injection cross-tie valve, and a uniform 15K steam generator tube plugging level. The analysis was performed at a core power level of 3250 MWt, a range of operating core average temperatures of 547'F to 581.3 F, and reactor pressure of either 2100 psia or 2250 psia. All other plant conditions were analyzed at a power of 3588 HMt. The licensee analyzed a spectrum of cold-leg breaks at the limiting reactor coolant system temperature and pressure conditions. The limiting break size from this analysis was then analyzed at other temperature and pressure points of the operating range. The limiting case was determined to be a three-inch diameter cold-leg break at a pressure of 2100 psia and at a core average temperature of 5474F. This limiting . break resulted in a peak clad temperature of 2122 F. Based on these results, the requirements of 10 CFR 50.46 have been met for the Unit 1 small-break LOCA analysis. The licensee reviewed the effect of the RTP program on the post-LOCA hot-leg recirculation time to prevent boron precipitation. This time is affected by power level and various systems'ater volumes and boron concentrations. Because these systems'ater volumes and boron concentrations are not affected by the RTP program, there is no effect on the post-LOCA hot-leg switchover time.
0 The licensee reviewed the effect of the RTP program on the post-LOCA hydrogen generation rates. The assumption of 120 F maximum normal operations containment temperature bounds, for the analysis of record, the effect of the primary system temperature changes of the RTP program on the post-LOCA hydrogen generation rates.
- 2. 1.2 Non-LOCA Transients and Accidents The licensee has evaluated the impact of the RTP program on the non-LOCA events presented in Chapter 14 of the D. C. Cook, Unit 1 FSAR. The approved reload core design methodology and design codes were used. The evaluations were performed to support the operation of Unit 1 at a core power of 3250 MMt over a vessel average temperature range between 547'F and 576.3'F at a primary system pressure of either 2100 psia or 2250 psia. The evaluation assumes a steam generator tube plugging level of 10K, with a peak steam generator tube plugging level of 15K. The non-LOCA safety evaluation supports the parameters of the RTP program with the exceptions of the steamline break mass and energy releases outside containment, which were evaluated at a full power vessel average temperature no greater than the current D. C. Cook Unit 1 full power average temperature, T , of 567.8'F.
avg'he evaluation performed by the licensee also considered the parameters for a potential uprating of Unit 1 to reactor core power level of 3413 MMt, with a vessel average temperature range between 547 F and 578.7'F at a primary system pressure of'ither 2100 psia or 2250 psia. The steam generator tube plugging level is assumed to be the same as for the RTP program. Even though the non-LOCA evaluation may have been performed for the uprated core power and its associated parameters, the staff's review of this license amendment does not address a D. C. Cook Unit 1 power uprating. The licensee revised certain reactor trip and engineered safeguards features (ESF) setpoints to provide adequate operating margins for the RTP operating conditions. Revised reactor trip setpoints were incorporated in the overtemperature-delta T (OTDT) and overpower-delta T (OPDT) trip functions. The revised ESF setpoints affects the low steamline pressure value of the high-high steamline flow coincident with a low steamline pressure actuation logic. The new OPDT and OTDT reactor trip setpoints were developed by the licensee for a new set of core thermal safety limits for the RTP program at a reactor core power level of 3413 MMt. The approved setpoint methodology of Reference 3 was used. For those events analyzed with the approved Improved Thermal Design Procedure (ITDP)., Reference 4, a safety-limit value of 1.45 was used for the Departure from Nucleate Boiling Ratio (DNBR). This is conservative compared to the design DNBR value of 1.32 for a thimble cell and 1,33 for a typical cell required to meet the DNB design basis. In the safety analysis for D. C. Cook, Unit 1, the licensee assumed the high pressurizer water level trip setpoint of 100K (nominal reactor setpoint). Furthermore, the reference average temperature used in the OPDT and OTDT trip setpoint equations are rescaled to the full power average temperature each time the cycle average temperature is changed. Similarly, the appropriate value of primary system pressure of either 2100 or 2250 psia was used in the two trip setpoint equations. For the revised ESF setpoint of the high-high steamline flow coincident with low steamline pressure, the low steamline pressure setpoint was lowered from 600 psig to 500 psig to accommodate the range of conditions of the RTP program and a potential power uprating.
- 2. 1.3 Steamline Break Mass/Ener Releases The current mass and energy releases for the inside containment analysis is based on analyses performed for Cook Unit 2, which are also applicable to Cook Unit 1. Data are represented in Chapter 14 of the FSAR for Unit 2 at power levels of 0, 30, 70, and 100% power. For the "at power" analyses, the initial primary system temperature and secondary steam pressures of the RTP program are lower than those in the Unit 2 FSAR analyses. The mass blowdown rate is dependent on steam pressure and since the steam pressure wi 11 be -less than-the- current-analyses, the initial mass blowdown rate wi 11 be lower.---The..lower steamline pressure setpoint (500 psig) of the ESF actuation signal does not significantly impact the analysis because the lead-lag compensation results in a steamline pressure signal which anticipates the rapid decrease in pressure caused by a steamline break. Based on these considerations, the licensee concludes that the RTP program wi 11 result in a lower integrated energy release into containment and that the data used in the Unit 2 FSAR remains bounding.
A study was performed for Unit 1 of the mass and energy release outside containment to address equipment qualification issues (Ref. 5). Cases at 70% and 100% power were analyzed. The analysis presented in Reference 5 assumed . the full power vessel average temperature to be 567.8'F ~ Any reduction in full power T from the analyzed T and the associated reduction in initial s earn pressure Pill result in less lkmlting releases. The low steamline pressure value assumed in the analysis supports the reduced value of the setpoint to 500 psig. The increased level of steam generator tube plugging is acceptable because the analysis assumed better heat transfer characteristics. The licensee concludes that the current mass and energy release analysis is acceptable for the RTP program as long as the full power T is equal to or less than 567.8 F. avg
- 2. l. 4 Startu of an Inactive Loo The licensee evaluated the startup of an inactive loop event. This event cannot occur above the P-7 permissive setpoint of 10% power as restricted by the Technical Specifications. The parameters assumed in the FSAR analysis for three-pump operation at 10% power remain bounding for the parameters for 10%
power condition. The licensee concludes, therefore, that the conclusions presented in the FSAR remain valid.
- 2. 1.5 Uncontrolled Rod Bank Withdrawal from a Subcritical Condition The uncontrolled rod bank withdrawal from a subcritical condition transient causes a power excursion. This power excursion is terminated, after a fast power rise, by the negative Doppler reactivity coefficient of the fuel, and a reactor trip on source, intermediate, or power range flux instrumentation. The power excursion results in a heatup of the moderator/coolant and the fuel. The analysis used a reacti~ity insertion rate of 75 pcm (note that one pcm is equal to a reactivity of 10 delta K/K). This reactivity insertion rate is greater than for the simultaneous withdrawal of the two sequential control banks having the greatest combined worth at the maximum speed of 45 inches/minute. The neutron flux overshoots the nominal full power value; however, the peak heat flux is much less than the full power nominal value because of the inherent thermal lag of the fuel. The analysis, with the reduced system pressure of 2100 psia, yields the minimum value of'NBR. The analysis is performed using the Standard Thermal Design Procedure (STOP). The W-3 ONB correlation was issued to evaluate ONBR in the span between the lower non-mixing vane grid and
the first mixing vane grid. The MRB-1 ONB correlation is applied to the remainder of the fuel assembly. From the analysis performed, the licensee concludes that the ONB design bases are met for all regions of the core, and therefore, the conclusions in the FSAR remain applicable for a reduction in nominal systea,pressure to 2100 psia.
- 2. 1.6 Uncontrolled Control Rod Assembl Bank Withdrawal at Power The uncontrolled rod bank withdrawal from a power condition transient leads to a power increase. The transient results in an increase in the core heat flux and an increase in the reactor moderator/coolant temperature. The reduction in pressure for the RTP program is non-conservative with respect to ONB. In addition, a revised Overtemperature Oelta-T setpoint equation is being assumed in the Cook Unit 1 analyses. The Power Range High Neutron Flux and Overtempera-ture Oelta-T reactor trips provide the primary protection against ONB. Both minimum and maximum reactivity cases were analyzed over a range of reactivity insertion rates. The licensee provided quantitative results for the maximum reactivity feedback case for power levels of 10K, 60K, and 100K power for a range of reactivity insertion rates. The results indicate that the ONBR limit is met for all the cases.
The licensee examined a number of cases associated with the pressurizer water volume transient caused by an uncontrolled control rod assembly bank withdrawal-at-power event. It was determined that credit for high pressurizer water level reactor trip was required to prevent the pressurizer from filling. The licensee assumed a value of 100K narrow range span (NRS) for the high pressurizer water level reactor trip setpoint. A time delay of 2 seconds was assumed for trip actuation unti 1 rod motion becomes adequate to terminate the transient. Thus the high neutron flux and overtemperature-delta T reactor trips provide adequate protection over the range of possible reactivity insertion rates in that the minimum value of ONBR remains above the safety-limit ONBR value. In addition, the high pressurizer water level reactor trip prevents the pressurizer from filling.
- 2. 1.7 Rod Cluster Assembl Misali nment The rod cluster control assembly misalignment events consist of three separate events: (1) a dropped control rod, (2) a dropped control bank, and (3) a statically misaligned control rod. These events were reanalyzed because the reduction in pressure for the RTP program is nonconservative with respect to the ONB transient. A dropped control rod or control bank may be detected in the following manner: (1) by a sudden drop in the core power as seen by the nuclear instrumentation system; (2) by an asymmetric power distribution as seen by the excore neutron detectors or the core exit thermocouples; (3) by rod bottom signal; (4) by the rod position deviation monitor; and (5) by rod position indicators. A misaligned control rod may be detected in the following manner; (1) by an asymmetric power distribution as seen by the excore neutron detectors or the core exit thermocouples; (2) by the rod position deviation monitor; and (3) by rod position indicators. The resolution of the rod position indicator channel's +5 percent or +12 steps (+7.5 inches). Oeviation of any control rod from its group by twice this distance (+24 steps or t15 inches) will not cause power distribution worse than the design limits. The rod position deviation monitor provides an alarm before a rod deviation can exceed + 24 steps or + 15 inches.
The dropped rod event was analyzed using an approved methodology (Ref. 6). A dropped rod or rods from the same group will result in a negative reactivity insertion which may be detected by the negative neutron flux rate trip circuitry. If detected, i reactor trip occurs in about 2.5 seconds. For those dropped rod events for wMch a reactor trip occurs, the core is not adversely impacted because the rapid decrease in reactor power will reach an equilibrium value dependent on the reactivity feedback or control bank withdrawal (if in automatic control). The limiting case for this class of events is the case with the reactor in automatic control. For this case a power overshoot occurs before an equilibrium power condition is reached. The licensee states that, using the methodology of Reference 6, all analyzed cases result in ONBR values which are within the safety-limit ONBR value. The licensee states that a dropped rod bank results in a reactivity insertion of at least 500 pcm. This will be detected by the negative neutron flux rate trip circuitry and cause a reactor trip within about 2.5 seconds of the initial motion of the rod bank. Power decreases rapidly and there is, therefore, no adverse impact on the reactor core. The most severe misalignment cases, with respect to ONBR, are those in which one control rod is fully inserted or where control bank "0" is fully inserted but with one control rod fully withdrawn. Multiple alarms alert the operator before adverse conditions are reached. The control bank can be inserted to its insertion limit with any control rod fully withdrawn without ONBR falling below the safety-limit ONBR value, as shown by analysis. An evaluation performed by the licensee indicates that control rod banks other than the control bank would give less severe results. For the case with one rod fully inserted, ONBR remains above the safety-limit ONBR value. For all cases following identification of a control rod misalignment, the operator is required to perform actions in accordance with plant Technical Specifications and procedures.
- 2. 1.8 Chemical and Volume Control S stem Malfunction The boron dilution event was analyzed by the licensee for startup and power operation. The analysis is performed to show that sufficient time is available to the operator to determine the cause of the dilution event and take corrective action before the shutdown margin is lost. The licensee reports that 45 minutes is available for Mode 1 (power operation) and 68 minutes for, Modes 2 or 3 (startup or hot standby conditions) (Ref. 7).
- 2. 1.9 Loss of Reactor Coolant Flow The loss-of-flow transient causes the reactor power to increase until the reactor trips on either a low-flow trip signal or reactor coolant pump power supply undervoltage signal. The reactor power increase causes a reactor moderator/coolant temperature increase. This initial coolant temperature increase causes a positive reactivity insertion because of the positive moderator temperature coefficient. The licensee analyzed both a partial loss-of-flow (loss of one pump with four coolant loops in operation) transient and a complete loss-of-flow transient (loss of four pumps with four coolant loops in operation).
For the partial loss-of-flow transient, the reactor is assumed to be tripped on a low-flow signal. For a complete loss-of-flow transient, the reactor is assumed to be tripped on a pump undervoltage signal. For either event, the average and hot channel heat fluxes do not increase significantly above their initial values and the ONBR remains above the safety-limit ONBR value.
- 2. l. 10 Locked Rotor Accident The locked rotor accident causes a rapid reduction in the fluid flow through the affected loop. The reactor trips on a low-flow signal which rapidly reduces the neutron flux upon control rod insertion. Control rod motion starts 1 second after the flow in the affected loop reaches 87K of its nominal value. The licensee evaluated this accident assuming that offsite power is available. No credit is taken for the pressure-reducing effect of the pressurizer relief valves, pressurizer spray, steam dump, or controlled feedwater flow after reactor trip. The licensee performed an analysis to determine the ONB transient and to demonstrate that the peak system pressure and the peak clad temperature remain below limit values. The peak reactor coolant system pressure of 2588 psia reached during the transient is less than that which would cause stresses to exceed the faulted conditions stress limits. The peak clad temperature reached is 1959'F. Less than 4.5X of the fuel rods in the most limiting fuel assembly reach values of DNBR less than the safety-limit DNBR value. These results indicate that the RTP program assumptions give acceptable consequences for the locked rotor accident.
- 2. 1. 11 Loss of External Electrical Load The loss-of-external-electrical-load event was analyzed by the licensee to show the adequacy of pressure-relieving devices and to demonstrate core protection.
This reanalysis was necessary because of changes in reactor pressure and temperature conditions for the RTP program and because of changes to the Overtemperature-Delta T reactor trip setpoint equation. Maximum and minimum o reactivity feedback cases were examined, with the case analyzed with and without credit for pressurizer sprays and power-operated relief valves. For the minimum reactivity feedback case with pressurizer pressure control, the reactor trips on a high pressurizer pressure signal. For the maximum reactivity feedback case with pressurizer pressure control, the reactor trips on a low-low steam generator water level signal. For the minimum reactivity feedback case without pressurizer pressure control, the reactor trips on a high pressurizer pressure signal. For all four cases, the minimum value of ONBR remains well above the safety-limit ONBR value and the Overtemperature-Delta T setpoint was not reached. The analysis confirms that the conclusions of the FSAR remain valid for this event for the RTP program.
- 2. 1. 12 Loss of Normal Feedwater Flow The loss-of-normal-feedwater-flow event was analyzed by the licensee to show that the auxiliary feedwater system is capable of removing the stored and decay heat, thus preventing overpressurization of the reactor coolant system or uncovering the core, and returning the plant to a safe condition. The reanalysis was based on a positive moderator temperature coefficient. A conservative decay heat model based on the ANSI/ANS-5. 1-1979 decay heat standard (Ref. 8) was used. Pressurizer power operated relief valves and the maximum pressurizer spray flow rate were assumed to be available since a lower pressure results in a greater system expansion. The initial pressurizer water level was assumed to be at the maximum nominal setpoint of 62K narrow range span. Reactor trip occurred when the low-low steam generator water level trip setpoint was reached. The results of the analysis show that a loss of normal feedwater does not adversely affect the reactor core, the reactor coolant system, or the steam system, and that the auxiliary feedwater system is sufficient to prevent watet relief through the pressurizer relief or safety valves. The pressurizer does
not fill and, therefore, the conclusions of the FSAR remain valid for this event, including RTP conditions. 2.1.13 Excessive Heat Removal Oue to Feedwater S stem Malfunctions The excessive-heat-removal event due to feedwater system malfunction was analyzed by the licensee to demonstrate core protection. This analysis was necessary because of changes in reactor core temperatures and pressure for the RTP program and because of changes to the OTDT and OPDT trip setpoints. This event is an excessive-feedwater-addition event caused by a control system malfunction or an operator error which allows a feedwater control valve to open fully. The licensee analyzed both full power and hot zero power cases. Both cases assumed a conservatively large negative moderator temperature coefficient. The full power case assumed the reactor was in automatic or manual control. The Improved Thermal Design Procedure (ITOP) of Reference 4 was used in the analysis. For the accidental full opening of one feedwater control valve with the reactor at hot-zero power conditions, the licensee determined that the maximum reactivity insertion rate is less than the maximum reactivity insertion rate analyzed in the Uncontrolled-Rod-Cluster-Assembly-Bank-Withdrawal-at-Subcritical-Condition event. Thus, this hot-zero power case is bounded by the results obtained previously for the other event. In addition, if the event were to occur at a hot-zero power and an exactly critical condition, the power range high neutron flux trip (low setting) of about 25K of nominal full power will trip the reactor. The hot-full po~er case with the reactor in automatic control is more severe than the case with the reactor in manual control. For all excessive feedwater cases, continuous addition of cold feedwater is prevented by automatic closure of all feedwater isolation valves on steam generator high-high level signal. A turbine trip is then initiated and a reactor trip on a turbine trip is then assumed. The results presented by the licensee demonstrate the safe response of Cook Unit 1 to the event, at hot-full power and in automatic control, with the ONBR remaining well above the safety-limit ONBR value.
- 2. l. 14 Excessive Increase in Secondar Steam Flow The excessive-increase-in-secondary-steam-.flow event was analyzed by the licensee to demonstrate core protection. This event is an overpower transient for which the fuel temperatur e will rise. It was analyzed because of reactor core temperature and pressure changes for the RTP program and because of changes to the OTOT and OPOT setpoints. The Cook Unit 1 reactor control system is designed to accommodate a 10K step load increase and a 5X-per-minute ramp load increase over the range of 15 to 100 percent of full power. Load increase in excess of these rates would probably result in a reactor trip. Four cases were analyzed by the licensee. These included minimum and maximum reactivity feedback cases with each case analyzed for both manual and automatic reactor control. for the minimum reactivity feedback cases, a zero moderator temperature coefficient was assumed to bound the positive moderator temperature coefficient..
For al'1 the cases, no credit was taken for the pressurizer heaters. The analyses used the ITDP of References 4. The studies show that the reactor reaches a new equilibrium condition for all the cases studied, with ONBR remaining well above the safety-limit ONBR value. The operators would follow normal plant procedures to reduce power to an acceptable value to conclude the event.
I I
2.1.15 Loss of all AC Power to the Plant Auxiliaries The loss-of"all-AC-power-to-the-plant-auxiliaries event was analyzed to demonstrate the adequacy of the heat removal capability of the auxiliary feedwater system. This transient is the limiting transient with respect to the possibility of pressurizer overfill. This event is more severe than the loss-of-load event because the loss of AC power results in a flow coastdown due to the loss of all four reactor coolant pumps. This results in a reduced capacity of the primary coolant to remove heat from the core. A positive moderator temperature coefficient was assumed in the analysis. A conservative decay heat model based on the ANSI/ANS-5.1-1979 decay heat standard (Ref. 8) was used. No credit is taken for the immediate release of the control rods caused by the loss of offsite power. Instead a reactor trip is assumed to occur on a steam generator low-low level signal. Pressurizer power operated relief valves and the maximum pressurizer spray flow rate was assumed to be available since a lower pressure results in a greater system expansion. The initial pressurizer water level is assumed to be at the maximum nominal setpoint of 62K narrow range span plus uncertainties of 5X narrow range span. The results demonstrate that natural circulation flow is sufficient to provide adequate decay heat removal following reactor trip and reactor coolant pump coastdown. The pressurizer does not fill. Thus, the loss of AC power does not adversely affect the core, the reactor coolant system, or the steam system, and the auxiliary feedwater system is sufficient to prevent water relief through the pressurizer relief or safety valves.
- 2. 1. 16 Steaml ine Break The steamline break accident was analyzed by the licensee to assess the impact of the reduced reactor coolant system pesssur e of the RTP program and the low steam pressure setpoint (lowered from 600 psig to 500 psig) of the coincidence logic with high-high steam flow for steamline isolation and safety injection actuation. An end-of-life shutdown margin of 1.6X delta K/K for no load, equilibrium xenon conditions, with the most reactive control rod stuck in its fully withdrawn position, was assumed. A negative moderator temperature coefficient corresponding to the end-of-line rodded core was assumed. The licensee evaluated four combinations of break sizes and initial plant conditions to determine the core power transient which can result from large area pipe breaks. The first case was the complete severance of a pipe downstream of the steam flow restrictor with the plant at no-load conditions and all reactor coolant pumps running. The second case was the complete severance of a pipe inside the containment at the outlet of the steam generator with the plant at no-load conditions and all reactor coolant pumps running. The third case is the same as the first case with the loss of offsite power simultaneous with the generation of a Safety Injection Signal (loss of offsite power results in reactor coolant pump coastdown). The fourth case is the same as the second case with loss of offsite power simultaneous with the generation of a Safety Injection Signal. A fifth case was performed to show that the ONBR remains above the safety-limit ONBR value in the event of the spurious opening of a steam dump or relief valve. The licensee determined that the first case was the limiting case, that is, the double-ended rupture of a main steam pipe located upstream of the flow restrictor with offsite power available and at no-load conditions.
The results indicate that the core becomes critical with the control rods inserted (however, with the most reactive control rod stuck out) before boron solution at 2400 ppm enters the reactor coolant system. The core power peaks at less than the nominal full core power. The ONB analysis showed that the
minimum DNBR remained above the safety limit ONBR value, even though this event is classified as an accident with fuel rods undergoing.DNB not precluded. The analysis performed by the licensee demonstrates that a steamline break accident will not resu'ft in unacceptable consequences.
- 2. 1. 17 Ru ture of Control Rod Orive Mechanism Housin Rod E'ection Accident The rod ejection accident is analyzed at full power and hot, zero-power conditions for both beginning-of-cycle (BOC) and end-of-cycle (EOC). The analysis used ejected rod worth and tr ansients peaking factors that are conservative. Reactor protection for a rod ejection is provided by neutron flux trip, high and low setting, and by the high rate of neutron flux increase trip. The analysis modeled the high neutron flux trip only. The maximum fuel temperature and enthalpy occurred for hot, full-power BOC case. The peak fuel enthalpy was, however, below 200 cal/gm for all the cases analyzed. For the hot, full-power cases, the amount of fuel melting in the hot pellet was less than 10K. Because fuel and clad temperatures and the fuel enthalpy do not exceed the FSAR limits, the conclusions of the FSAR remain valid.
Based on a review of the licensee's evaluation and analysis of the non-LOCA transients and accidents (2. 1.3 through 2. l. 17) for the reduced temperature and pressure operation (the RTP program), the staff concludes that they are acceptable because (1) approved methodologies and computer codes have been used, and (2) all applicable safety criteria have been met. This review is based on (1) a full power vessel average temperature of less than or equal 'to 567.8'F, (2) a steam generator tube plugging level of 10K with a peak tube plugging level of 15K, and (3) .the minimum measured flow requirement of 91,600 gpm per loo'p is met.
- 2. 1. 18 Steam Generator Tube Ru ture SGTR) Accident The licensee analyzed the steam generator tube rupture (SGTR) event for Cook Unit 1 using methodology and assumptions consistent with those used for the Cook FSAR SGTR analysis. The range of parameters associated with a future rerating program and the RTP program were used in sensitivity analyses to assess the impact of these programs on the primary-to-secondary break flow and the steam released to the atmosphere by the affected steam generator. These two factors affect the radiological consequences of an SGTR accident. In addition, the licensee's evaluation of the radiological doses considers the effect of the noble gas concentrations. The licensee states that the results of the analyses show that the doses remain within a small fraction (10K) of the 10 CFR Part 100 guidelines for both the thyroid and whole body doses. Since the worst case doses are within the 10 CFR Part 100 guidelines, the staff concludes that the analysis of the SGTR is acceptable
- 2. 1.19 Fuel Structural Evaluation The fuel assembly Cook Unit 1 because lift and buoyancy forces are increased for the RTP program at a reduction in reactor coolant system temperature of about 20'F will increase the coolant density by about 3X. The licensee evaluated this force incr ease against the fuel assembly allowable holddown load. The results of the evaluation show that the increased force is well within the minimum spring holddown force design margin. In addition, the licensee determined that the cold-leg break remains the most limiting pipe rupture transient with respect to lateral and vertical hydraulic forces. Based on the licensee's review, the staff concludes that the 15xl5 fuel assembly design remains acceptable.
The fuel rod design was evaluated to assess the impact of future rerating. The licensee determined that the rod internal pressure criterion will continue to be the more important factor in fuel burnup capabilities. The fuel will also undergo more severe fuel duty because of the uprated power. The licensee plans to perform cycle-specific verification for each reload to assure that all fuel rod design criteria are met. 2;1;20-.-Justification for. Pressurizer Level The purpose of the Pressurizer High Level Limit is to ensure that a steam bubble is present in the pressurizer prior to power operation to minimize the consequences of overpressure transients and the possibility of passing water through the relief and safety valves. The safety analysis assumes a maximum water volume which corresponds to about 65K indicated level. This nominal indicated level is maintained during normal operation by the pressurizer .level control system. The licensee (and the fuel supplier - Westinghouse) recommends the use of 92K for the Pressurizer High Level trip limit. They state that this new trip limit wi 11 still ensure the presence of a steam bubble in the pressurizer. The pressurizer level will, however, be controlled to the nominal value. For normal operations (Condition I event), the reactor parameters, including the pressurizer level, do no significantly deviate from their nominal values. The licensee concludes that, for the pressurizer level to exceed the nominal level, a transient or accident must occur for which protective action is provided by the
~ ~ ~
Reactor Protection System. Any other possible conditions for which the nominal
~
level would be exceeded before and during a transient would require a
~
~
transient'r transients beyond those usually considered for an FSAR type of analysis. The
~
~
staff concludes on the basis of the licensee's evaluat>on that a Pressurizer High Level Trip of 92~ is acceptable. 2.2 BALANCE OF PLANT SYSTEMS The licensee states that balance of plant (BOP) systems and components were analyzed for the effects of operation at reduced temperature and pressure conditions. The secondary side conditions for these analyses were determined using the Performance Evaluation and Power System Efficiencies (PEPSE) heat balance data (14.20 E6 lb/hr main steam flow and main feed flow). The systems reviewed were the non safety-related secondary side power generating and nonpower generating systems. Included in the licensee's analysis were portions of the main feedwater, main steam, steam generator blowdown (SGBS), component cooling water (CCWS), auxiliary feedwater (AFS), heating, ventilation, and air conditioning (HVAC), service water, waste disposal, fire protection, radiation monitoring, and spent fuel pool (SFP) cooling and cleanup systems. The performance of the above BOP systems was evaluated at the reduced temperature and pressure by using the new primary side NSSS data (14.20E6 lb/hr main steam and main feed flow, and 434'F main feed temperature) furnished by Westinghouse.
~
The licensee states that the impact on containment pressures and temperatures following a postulated design basis main steam line break was evaluated and its
~ ~
effect on equipment qualification was verified.~ The flooding analysis in safety-
~
related areas of the plant as a result of a postulated pipe break was reevaluated due to the slight increase in flow rates in the main feed, condensate, and main
~
steam systems. The turbine-generator system was also evaluted to confirm its integrity and performance at the increased steam volumetric flow rate and to verify that the original turbine missile analysis remains valid.
~
The licensee's analysis of BOP system performance provided the following findings
~ ~
concerning the RTP conditions at the present licensed power level of 3250 HMt
~ ~ ~
NSSS power: ~ (a) The: capability of the safety-related portion of the main feedwater system will not be affected and will continue to perform its safety function because the proposed RTP conditions are bounded by the existing main feedwater system design. The licensee's analysis of the pressure/temperature rating conditions for the system confirms that pressure boundary integrity will not be affected. In addition, the main feedwater system isolation valve closure time is not affected by the RTP"imposed conditions. (b) The capability of the steam generator blowdown system to remove impurities from the secondary side remains essentially the same for the RTP-imposed conditions during normal operation based on the exsisting design. (c) The reactor makeup water system's (HSM) capability to provide demineralized water for makeup and flushing operations throughout the NSSS auxilliaries, the radwaste systems, and fuel pool cooling and cleanup system is not chaIlenged because the existing system design is based on the worst case demand which bounds the RTP conditions. (d) The licensee confirmed that safety-related equipment will not be affected by changes in the flooding analysis due to the RTP conditions. Flooding in the auxiliary building due to failure of nonseismic Class I piping has been reviewed. The licensee analyzed systems having access to large water volumes and/or potentially large flowrates were considered as discussed in the FSAR. The only such system is the main feedwater system. Since the changes in flow in the main feedwater system are still within the design limits, the results concerning flooding discussed in the FSAR are still applicable. Flooding in the containment is slightly increased due to the larger initial water mass in the reactor coolant system because of the higher density at the reduced temperature. This change was found to be within the volume margins used to determine the maximum flood-up elevation. The containment flooding evaluation in the FSAR remains valid at the RTP-induced conditions. (e) The adequacy of the AFM system for accident mitigation was demonstrated in the Mestinghouse accident analysis performed in support'f the RTP program under the following scenarios:
- 1. Loss of main feedwater
- 2. Loss of offsite power
- 3. Hain steam line rupture Each accident analysis demonstrated acceptance criteria such as system overpressure limits or ONB limits. The AFM system's ability for design basis accident decay heat removal calculated in the RTP analysis is unaffected.
13 As evaluated in the RTP analysis, the heat loads in both the primary and secondary systems due to reactor decay heat remain unchanged. Therefore, the Component Cooling Water System (CCWS) analysis and service water system (SWS) analysis in the FSAR remain valid. (g) For main steam line breaks inside the containment structure, the pressure and temperature will remain within the bounds of the peak pressure and temperature used in the evaluation of containment performance. The initial primary temperatures and secondary steam pressures under the RTP conditions will be lower than those used in the FSAR analysis. The licensee has confirmed that containment environmental qualification of equipment inside containment is not affected. (h) The superheated mass and energy release analysis outside containment was evaluated to address equipment qualification issues. The primary temperatures and secondary steam pressures resulting from the RTP conditions will be lower than those used in the FSAR'analysis. The mass and energy release will be lower and operation with RTP will result in lower temperatures in the break areas. As such, the current superheat mass and energy release analysis outside containment remains bounding provided the full power vessel average temperature is restricted to the currently-licensed 567.8'F and below. The secondary pressure conditions assumed ir. the high energy steam line break analysis wi 11 be lower than those presented in the FSAR. These bound the proposed RTP conditions and therefore the current anaIysis is sufficient. The primary function of the spent fuel pool cooling system (SFPCS) is to remove decay heat that is generated by the elements stored in the pool. Decay heat generation is proportional to the amount of radioactive decay in the elements stored in the pool which is proportional to the reactor power history. Since the plant's rated power level of 3250 NWt remains unchanged, the demand on the SFPCS is not increased. The purification function is controlled by SFPCS demineralization and filtration rates that are not affected by the RTP conditions. (k) The fire protection systems and fire hazards are independent of the plant operating characteristics with the exception of the slightly increased current requirements for the electric motor driven pumps in the primary system. The increased load is due to the more dense water being pumped under the RTP conditions. The increased current required is small and therefore is not considered to be a fire hazard. The licensee confirmed that BOP systems have the capability to maintain plant operation under the RTP-induced conditions without modification to the existing design. The staff has reviewed the FSAR and licensee submittals in order to verify that safety-related BOP system performance capability, as analyzed, bounds the
changes in design basis accident assumptions created by the RTP operation. The staff has confirmed that safety-related BOP system design capability, flooding protection, aad equipment qualifications are bounded for the proposed rerating and therefore are considered acceptable as is. Based on the above, the staff concludes that the proposed license amendment for the D.C. Cook Nuclear Plant Unit 1 concerning the Reduced Temperature and Pressure is within the existing safety-related BOP system design capability for design basis accident mitigation and, therefore, the staff's previous approval against the applicable licensing criteria for the main steam system, main feed system, CCWS, SWS, AFS, MSW, SGBS, SFPCS, flooding protection, containment performance, and equipment qualifications remain valid. The staff, therefore, finds the BOP systems concerned acceptable for continued operation at the proposed reduced temperature and pressure. 2.3 REACTOR VESSEL ANO VESSEL INTERNALS The reactor vessel is designed to the ASME Boiler and Pressure Vessel Code, Section III (1965 Edition with addenda through the winter 1966). The licensee has determined that the operation of the reactor vessel under the most limiting conditions of the RTP rerating is acceptable fot its original 40-year design objective. All of the stress intensity and usage factor limits of the applicable code for the Unit 1 reactor vessel are still satisfied when the RTP is incorporated, with the exception of the 3Sm limit for the Control Rod Orive Motor (CROM) housings and outlet nozzle safe end. However, the code permits exceeding the 3Sm limit provided plastic or elastic/plastic analysis criteria are met. The licensee's review of the reactor vessels internals for the RTP program included three seperate areas: a thermal/hydraulic assessment, a RCCA drop time evaluation, and a structural assessment. Force increases were calculated for the upper core plate, across the core barrel, and in the upper internals near the outlet nozzles. In these areas the existing margin was determined to be sufficient to accommodate the increased stresses. The results of this review indicate that the original reactor internals components remain in compliance with the current design require-ments when operating at the new range of primary temperatures and pressures. The PTS rule requires that at the end-of-life of the reactor vessel, the projected reference temperature (calculated by the method given in 10 CFR 50.61(b)(2), RT/pts) value for the materials in the reactor vessel beltline be less than the screening criterion in 10 CFR 50.61(b)(2). The RT/pts value is dependent upon the initial reference temperature, margins for uncertainty in the initial reference temperature and calculational procedures, the amounts of nickel and copper in the material, and the neutron fluence at the end-of-life of the reactor vessel. Of these properties, only neutron fluence is affected by rerating with RTP. Since the colder coolant in the downcomer region is more dense and thus provides for a more efficient neutron shield for the reactor vessel, fluence estimates are lower than those at current operating conditions. All other properties are independent of the RTP-induced conditions. The effects of NRC Generic letter 88-11, dated July 12, 1988, regarding Regulatory Guide 1.99 Rev. 2 were evaluated by Westinghouse and determined to not be significant for RTP. The effect of RTP will be incorporated by the licensee in future PTS submittals.
An evaluation was performed to determine the impact of RTP rerating on the applicability of the PTS screening criteria in terms of vessel failure. A probabilistic fracture mechanics sensitivity study of limiting PTS transient characteristics, starting from a lower operating temperature, showed that the conditional probability of reactor vessel failure will not be adversely affected. Therefore, the overall risk of vessel failure will not be adversely impacted, meaning that the screening criteria in the PTS Rule are still applicable for the O.C. Cook Nuclear Plant Unit 1 reactor vessel 'relative to rerated conditions. Analysis of the CROM housings and the outlet nozzle safe end shows the maximum range of primary plus secondary stress intensity exceed the 3Sm limit. The licensee, however, performed a simplified elastic/plastic analysis in accordance with paragraph NB-3228.3 of the ASME Boiler and Pressure Vessel Code, Section III (1971 or later edition) and the higher range of stress intensity is justified. Therefore, based on the licensee's reviews and analysis of the above portions of the reactor vessel and internals, the staff concludes that the conditons imposed on the reactor vessel and internals by the RTP rerating are acceptable. 2.4 TURBINE MISSILES The FSAR turbine missile analysis is based on a low pressure turbine failure. The licensee's analysis of the slightly changed steam conditions entering the low pressure turbine shows that the probabilty of a low pressure turbine missile is virtually unaffected. The factors that directly or indirectly cause stress corrosion cracking in the low pressure turbine wheels are steam pressure and temperature, mass flow rate, steam moisture content, water chemistry, oxygen level, and turbine speed. The licensee reported that changes in these factors are negligible due to the RTP-induced conditions. The only noticeable change that the staff can determine is a 1.0X increase in the steam flow rate. The staff's conclusion, based on the licensee's review, is that the turbine missile hazard is neglibily affected by the RTP conditons and is, therefore, acceptable. 2.5 PLANT STRUCTURAL ANO THERMAL DESIGN The NSSS review consisted of comparing the existing NSSS design with the performance requirements at the rerated RTP conditions. The current components of the Cook Unit I/model 51 steam generators continue to satisfy the requirements of the ASME B8PV Code, Section III,(the code applicable for the design of the Cook Nuclear Plant Unit 1), for this program. In addition, thermal hydraulic evaluations of the steam generators show acceptable stability and circulation ratios at the RTP rerated conditions. Circulation ratio is primarily a function of power, which is unchanged, therefore is itself virtually unchanged. The dampening factor characterizes the thermal and hydraulic stability of the steam generator. Mestinghouse has determined that all dampening factors are negative at nearly the same value as the current operating conditions. A negative dampening factor indicates a stable device. Since the code requirements continue to be satisfied, and since stability and circulation ratios have been determined by Mestinghouse to be
within the design criteria, the staff concludes that RTP operation is acceptable for the Model'1 steam generators. The pressurizer. structural analysis was performed by modifying the original O. C. Cook Nuclear Plant Pressurizer analysis ("Model 51 Series Pressurizer Report" ). The analysis was performed to the requirements of the ASHE Code 1968 Edition, which is the design basis for the O.C. Cook Nuclear Units. The only ASME Code requirement affected by the transient modifications was fatigue. The limiting components for fatigue usage factors are the upper shell and the spray nozzle, which are calculated to be 0.97 and 0.99 respectively. These remain, however, within the ASME acceptance criteria of 1.0 and are, therefore, acceptable to the staff. Reactor coolant pump hydraulics and motor adequacy were reviewed for the proposed RTP conditions by Westinghouse. The increased hot horsepower and stator temperature conditions are within the NEHA Class B limits. A review of generic Reactor Coolant Pump stress reports for model 93A pumps by Westinghouse finds that all the design requirements provide adequate bounding of the RTP-induced conditions and, therefore, the staff finds this acceptable. Oue to lower temperatures from the RTP program, the RCS will not expand as much as currently designed. This will result in support gaps being present in locations that were previously zero. The small gaps in the support structure may result in increased dynamic loading (both seismic and LOCA) in localized areas. The overall LOCA loadings on the RCS, however, remain approximately the same for the following reasons: The lower RCS temperatures yield lower thermal loadings.
- 2. The 0. C. Cook Nuclear Plant has a leak before break design methodology which allows the faulted condition evaluation to proceed without having to consider loadings from postulated breaks in the primary loop piping.
The seismic margin available for this plant is also significant which means that there are no components in the system which are close to their allowable stresses. Based on the above, the temperatures associated with the RTP rerating are, therefore, acceptable to the staff for the loop piping, the loop supports, and the primary equipment nozzles. The effects of the O. C. Cook Nuclear Plant RTP rerating on the operability and design basis analysis of the CROM's of Unit 1 were reviewed. The RTP rerating does not affect the operability or service duration of the CROM latch assembly, drive rod, or coil stack. The CROM latch assembly and drive rod were originally designed for 650'F, and the design basis stress and fatigue calculations remain representative for these components since the components are exposed to the hot leg temperature, which has not increased. The coil stack is located on the outside of the pressure housing which is subject to ambient containment temperatures, which have not changed. An evaluation was performed on the impact of the RTP rerated operating conditions on the structural analysis of the CROM pressure housing. The component of the pressure housing which experiences the greatest stress range and has the highest fatigue usage factor is the upper canopy. This is the pressure housing seal weld between the rod travel housing and the cap. Mestinghouse provided a review on the impact of the differences
between the original normal and upset condition transients and those of the RTP on the code allowable stress levels and fatigue usage factors. The results of the evaluation are: The maximum stress intensity range is equal to 109,960 psi, which is less than the maximum allowable range of thermal stress of 127,105 psi which was previously found to be acceptable.
- 2. The total fatigue usage factor is equal to 0.672, which is less than the allowable limit of 1.0 (ASME Section III, 1971 Edit>on).
The staff concludes, based on licensee evaluations, that the impact of the RTP program on the CRDM's is within design criteria and, therefore, is found to be acceptable.
- 2. 6 CONTAINMENT EVALUATION Short-Term Containment Res onse As part of the analysis to support RTP operation, the reactor cavity and loop subcompartments short-term pressurization in the event of a break of large coolant piping or a steam line was reanalyzed by Westinghouse.. In some of those areas, the analyzed pressure exceeded the structural limits as expressed in the FSAR. These structures were reevaluated using the peak pressures obtained from the RTP analysis, WCAP 11902 (ref.2), to confirm that the acceptance criteria of Section 5.2.2.3 of the updated FSAR, titled "Containment Design Stress Criteria,"
were met. The original design of the containment included a number of considerations of which the subcompartment pressures were but one. For example, radiation shielding requirements may have dictated a thicker concrete slab than was necessary from a structural perspective. The actual capacity is generally-greater than the design pressures stated in the FSAR, and is further increased due to the fact that the materials used are stronger than the required minimum design strengths. In the RTP structural review, advantage was taken of these greater capacities by performing manual or finite element evaluations of the affected structural elements. The greater material. strengths were used in the analysis where appropriate. Loo Subcom artments The containment building subcompartments are the fully or partially enclosed spaces within the containment which contain high energy piping. The subcompartments are designed to limit the adverse effects of a postulated high energy pipe rupture. The results of the short term containment analyses and evaluations for the D.C. Cook Nuclear Plant Unit 1 demonstrate that, for the pressurizer enclosure, the fan accumulator room, and the steam generator enclosure, the resulting peak pressures remain below the allowable design peak pressures'or the loop compartments, the peak calculated pressures at the RTP rerated conditions are higher than the FSAR design allowables. For these areas, structural evaluations were performed as discussed above for the revised peak pressures, and the structural adequacy of the containment subcompartments have been confirmed (Ref. 10) as follows:
1 V 'e Differential Pressure
~ ~
Node 1 or 6 to Node 25
,This is the
~ ~
differential pressure
~ ~
from the reactor coolant loop compartments adjacent ta the refueling canal nodes 1 or 6 across the operating deck to the upper containment. Original Design pressure 16.6 psi Original Calculated pressure 14.1 psi New Calculated pressure 18.7 psi The licensee demonstrated the increased differential pressure to be acceptable by review of existing computer analysis of the reactor coolant pump hatch covers and reevaluation of the operating deck load carrying capacity. Differential Pressure Node 2 or 5 to Node 25 This is the differential pressure across the operating deck from the reactor coolant loop compartments located 90 degrees from the refueling canal to the upper containment. Original Design pressure 12.0 psi Original Calculated pressure 10.6 psi New Calculated pressure 13.0 psi The licensee demonstrates the increased differential pressure to be acceptable by comparison to Node 1 and Node 6 areas, The slabs in both areas are the same. Peak Shell Pressure This is the differential pressure across the containment shell to the outside, for nodes located in the ice condenser inlet areas closest to the refueling canal. Original Design pressure 12.0 psi Original Calculated pressure 10.8 psi New Calculated pressure 14.0 psi The licensee demonstrates the increased pressure to be acceptable by evaluation on a localized basis. The containment shell can handle pressures well in excess of the overall 12 psi design pressure. The average pressure over the structurally significant portion of the containment shell surrounding and including these nodes is smaller than the 12 psi containment shell design pressure. Reactor Cavit The reactor cavity is the structure surrounding the reactor with penetrations for the main coolant piping. This structure is designed to limit the adverse effects of the initial pressure response to a loss of coolant accident. The results of the reactor cavity analysis and evaluations for the D. C. Cook . Nuclear Plant Unit 1 demonstrate that, for the reactor vessel annulus and pipe annulus, the resulting*peak pressures at the RTP rerated conditions are within the FSAR design allowables. For the upper and lower reactor cavities the peak calculated pressures under RTP conditions exceeded the structural design pressures (Ref. 2, Sections 3. 7. 2 and 3. 7. 3) as stated in the FSAR. For these
0 areas, structural evaluations were performed for the revised peak pressures, and the structural adequacy of the containment subcompartment has been confirmed (Ref.~ 10) as follows:
~
Missile Shield Refuelin Canal Bulkhead Blocks and U er Reactor Cavit al sf erentla ressures The upper reactor cavity walls surround the reactor head. The missile shields and the refueling canal bulkheads are blocks separating the upper reactor cavity from upper containment. The missile shield is bolted down during operation, and is removable for refueling. The refueling canal bulkheads fit snugly in grooves in the upper reactor cavity walls.
~Ci II11 W i1 Sli 11 and Bulkheads Original Oesign pressure 48.0 psi 48.0 psi Original Calculated pressure 44.1 psi 44.1 psi New Calculated pressure 48.4 psi 54.3 psi The licensee demonstrates the increased pressure for the cavity wall to be acceptable by finite element analysis of the entire upper reactor cavity wall.
The licensee has demonstrated the increased pressure for the missile shields and the bulkheads to be acceptable by manual calculation. The test cylinder break strength of the concrete, which is higher than the design strength, was also taken into consideration. Peak Lower Cavit Pressure This is the cavity located under the reactor vessel. The peak pressure is used in the structural analysis rather than the differential pressure since most of the cavity walls are in the foundation mat. Original Oesign pressure 15.0 psi Original Calculated pressure 13.8 psi New Calculated pressure 18.5 psi The licensee demonstrated that the increased pressures are acceptable by manual calulation. The staff concludes, based on the licensee's demonstration, that the 0. C. Cook Nuclear Plant's design basis pertaining to containment short term response, as stated in Chapter 5.2.7.3 of the FSAR, is adequate for RTP operation, and therefore, is acceptable. The licensee must update the FSAR to reflect the higher structural design values. Lon Term Containment Pressure The long term peak containment pressure analysis supports operation with the RHR crosstie valves closed at a power level of 3425 NMt for both Units 1 and 2 containment structure.. This analysis contained additional justification for operation under the RTP conditions (Ref. 11) and was approved by the staff Safety Evaluation dated January 30, 1989 (Ref. 12).
I H I I 2.7
~ NUCLEAR,~ PROCESS AND POST-ACCIDENT SAMPLING SYSTEMS The Nuclear Sampling System (NSS) fs designed to provide representative samples for laboratory. analyses used to guide the operation of various primary and secondary systems throughout the plant during normal operation. Since reduction of sample pressure and temperature, when necessary, is already being done by heat exchangers and needle valves, the parameters associated with the RTP program do not affect the performance of the NSS. With no power upratfng, the source term remains unchanged. Therefore, the staff concludes that operation under RTP conditions fs acceptable for the NSS.
The staff finds that, since no power uprating is being proposed at this time, there is an insignificant effect on the post-accident containment thermal conditions and therefore the existing post-accident sampling system remains adequate and is acceptable. Operation under RTP conditions results in slight reductions in secondary sfde temperatures and pressures with no change in the source term. The staff concludes that the change ca'n be accommodated by the process sampling system without causing degradation of their performance, and fs, therefore, acceptable. 2.8 ELECTRIC SYSTEMS DESIGN Operation under RTP conditions results in minor changes to the. heat balance. The only impact noted on the electrical systems is the slight increase in motor current for the motors used as prime movers of primary coolant. The required power fs increased by the higher densities encountered due to the RTP program. The licensee has reviewed cable penetratfons, busses, and motor ratings to conclude that there is sufficient design margin to handle the increased load. The staff finds, based on the licensee's evaluation, that the proposed RTP program minimally affects the electric power system and associated loads and fs therefore, acceptable. 3.0 TECHNICAL SPECIFICATIONS
- 1. Definition 1.38 on design thermal power fs being deleted on page 1-7 of the Technical Specifications (TS's) because there fs no longer a single design thermal power at which all the transient and accident analyses have been performed. The licensed power level for Cook 1 remains 3,250 MWt. This change is acceptable.
- 2. Table 1-3 on page 1-10 is being deleted because ft previously gave information on the analyses performed at the design thermal power.
This change fs acceptable because the definition of design thermal power is being deleted also.
- 3. Figure 2.1-1 on page 2-2 fs being revised to reflect the revised DNBR safety limit of 1.45. This change is acceptable because it is supported by the safety analysis.
- 4. The pressurizer pressure low setpoint (Item 9 of Table 2.2-1 on page 2-5) fs increased by 10 psig. This is acceptable because ft was assumed in the large- and small-break LOCA analyses.
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- 3. 0
~ TECHNICAL SPECIFICATIONS
- 1. ~ Oefinition 1.38 on design thermal power is being deleted on page 1-7
~ ~ ~
of the Technical Specifications (TS's) because there is no longer a single design thermal power at which all the transient and accident analyses have been performed. The licensed power level for Cook 1 remains 3,250 HHt. This change is acceptable.
- 2. Table 1-3 on page 1-10 is being deleted because it previously gave information on the analyses performed at the design thermal power.
This change is acceptable because the definition of design thermal power is being deleted also.
- 3. Figure 2.1.-1 on page 2-2 is being revised to reflect the revised ONBR safety limit of 1.45. This change is acceptable because it is supported by the safety analysis.
- 4. The pressurizer pressure low setpoint (Item 9 of Table 2.2-1 on 2-5) is increased by 10 psig. This is acceptable because it was page assumed in the large- and small-break LOCA analyses.
The Overtemperature-Oelta T trip setpoint equation (pages 2-7 and 2-8) is being revised in terms of rated thermal power rather than design thermal power. In addition, this revised OTDT trip setpoint protects the core safety limits of Figure 2. 1-1. This change is acceptable because it is supported by the non-LOCA safety analyses.
- 6. The Overpower-Oelta T trip setpoint equation (page 2-9) is being revised to reflect the revised core safety limits of Figure 2.1-1.
This equation is also being defined in terms of the indicated T at rated thermal power. These changes are acceptable because tile are supported by the safety analysis for the RTP program.
- 7. Technical Specification 3.2.2 on page 3/4 2-5 is being revised from a maximum F of 2. 10 to 2. 15. This change is acceptable because it is supportetl by the large-break LOCA analysis. The F values for Exxon fuel are being deleted because this fuel will n3 longer be used at Cook Unit 1.
- 8. The K(Z) curve applicable to Exxon fuel (page 3/4 2-7) is being deleted. This is acceptable because Exxon fuel will no longer be used at Cook Unit 1.
- 9. The K(Z) curve for Mestinghouse fuel (page 3/4 2-8) is being revised. This is acceptable because it is supported by the new LOCA analysis for Cook Unit l.
- 10. The F-Oelta H limit applicable to Exxon fuel (page 3/4 2-9) is being deleted. This is acceptable because Exxon fuel will no longer be used at Cook Unit l.
Table 3.2-1 on page 3/4 2-14 on ONB parameters is being revised. T must be less than or equal to 570.9'F, the pressurizer
l
-22" pressure must be less than or equal to 2050 psig, and the reactor coolant system total flow rate must be greater than or equal to 366,400 gpm. These changes are acceptable because they reflect the safety analysis for the RTP program.
Technical Specification 3.2.6 on page 3/4 2-15 is being revised to change F in the APL limit to 2. 15. This change is acceptable because 3t reflects the new F limit of Specification 3.2.2. The limits on APL applicable to E)xon fuel are being deleted because Exxon fuel will no longer be used at Cook Unit l. Functional Units 2 and 11 of Table 3.2-2 on page 3/4 3-10 are being changed. Functional Unit 2 incorporates an editorial change to indicate that the response time is applicable to both the high and low setpoints of the Power Range Neutron Flux trip. This change is acceptable because it is editorial in nature. Functional Unit being changed from a response time of "not applicable" to "equal to ll is or less than 2 seconds." This is acceptable because this trip on pressurizer water. level-high was modeled in the analysis of the control rod withdrawal-at-power event. Functional Units 1.f and 4.d of Table 3.3-4 on pages 3/4 3-24 and 3/4 3-26 are being changed to decrease the steamline pressure low setpoint by 100 psig. These changes are acceptable because they are supported by the steamline break analysis and the steamline break mass and energy evaluations. Tec'hnical Specification 3.4.4 on page 3/4 4-6 is being revised to 92K of span. This change is acceptable because the safety analysis. it is supported by Technical Specification 3.5. l.b on page 3/4 5-1 is being revised from an accumulator borated minimum water volume of 929 to 921 cubic feet. This change is acceptable because consistent with the LOCA analysis for Cook Unit 1. it is Surveillance Requirement 4. 5. 2.f is being revised to reduce the discharge pressure of the safety injection pump and the residual heat removal pump. These changes are acceptable because they are consistent with the LOCA analyses. Surveillance Requirement 4.5.2.h is being revised by adding a requirement to verify that the charging pump discharge coefficient is within a specified range following ECCS modifications. The footnote is broken into four parts for clarity. This change is acceptable because it ensures that the flow delivered to the core by the charging pumps in the event of a LOCA is within the analyzed values. Surveillance Requirement 4.7.1.2 on, page 3/4 7-5 is being revised to change the discharge pressure requirements of the motor and turbine driven auxiliary feedwater pumps to 1375 psig and 1285 psig, respectively. This corresponds to a 5X degradation of the pumps
,~
l 23 from the manufacturer's pump head curve. These changes are acceptable because they are consistent with the changes for the RTP progl'am>>
- 20. Basis page B 2-1(a) is being changed to incorporate the design limit and safety analysis limit DNBR values. The DNB limits for Exxon fuel are being deleted since Exxon fuel is no longer used at Cook Unit 1. The design limit and safety analysis limit DNBR values are acceptable because they are consistent with the RTP program.
- 21. Basis page B 2-2 is being revised to delete reference to F-Delta H for Exxon fuel and to design thermal power. These changes are acceptable because references to both items have been deleted in the Specifications.
- 22. Bases page B 2-4 is being revised to reflect the changes to the Overtemperature-Delta T trip function. The changes are acceptable because they reflect changes made to the Specifications.
- 23. Bases page B 2-5 is being revised to reflect the changes to the Overpower-Delta T trip function and the pressurizer water level-high tr'ip. These changes are acceptable because they reflect changes to the Specifications.
- 24. Bases page B 3/4 2-1 is being revised to replace the minimum DNBR value of 1.69 by the words "the safety limit DNBR". This change is acceptable because it will avoid changes to the Bases if the safety limit DNBR value is changed.
- 25. Surveillance Requirement 4.1. 1.5.b is being changed to require T determination of T every 30 minutes when the reactor is criti87 and- T is less tQP 545'F. This change is supported by Reference 9 and 57lows a full power T of 550'F for Cook Unit 1 Cycle ll without requiring a monitor)(II every 30 minutes while at full power, which the previous value of 551'F would have required. This change is acceptable because the intent of maintaining the minimum coolant temperature for criticality of Specification 3.1. 1.5 is preserved.
4.0 ENYIRONNENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal ~Re later on June 9, 1989 ( 94 FR 24774). Accordin917, based upanut ie envsronmental assessment, we have determined that the issuance of he amendment will not have a significant effect on the quality of the human environment.
- 5. 0 CONCLUSION
. The staff has reviewed the request by the Indiana and Michigan Power Company to operate the Donald C. Cook Nuclear Plant Unit 1 at the reduced temperatures and pressures of the RTP program. Reactor operation is restricted to an upper limit on T of 567.8'F because the steamline break mass and energy release inside con$ kfnment was not reanalyzed as part of the RTP program. Although the
"24"
~
safety analysis was performed at power ratings which would support a possible
~ ~ ~
power uprating for Cook Unit 1, power uprating is not addressed in the staff's
~ ~ ~ ~
review. The power of O.C. Cook Nuclear Plant Unit 1 is limited to the present
~ ~ ~ ~
~ ~ ~
rated thermal'ower of 3250 MMt. Based on its review, the staff concludes that
~
appropriate material was submitted and that normal operation and the transients and accidents that were evaluated and analyzed are acceptable. The Technical Specifications submitted for this license amendment suitably reflect the necessary modifications for the operation of Cook Unit l. The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission s regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Oate: 'June 9, 1989 Principal Contributors: Dan Fieno John Stang, NRR Anthony Gody, NRR
6.0 REFERENCES
- 1. Letter (AEP:NRC:1067) from M. P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated October 14, 1988.
- 2. "Reduced Temperature and Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 - Licensing Report," D. L. Cecchett and D. B. Augustine, WCAP-11902, October 1988.
- 3. Ellenberger S.L., et al., "Design Bases for the Thermal Overpower-Delta T and Thermal Overtemperature-Delta T Trip Functions," WCAP-8746, March 1977.
- 4. Che'lerner, H.; Boman, L.H.; Sharp, D.R., "Improved Thermal Design Procedures," WCAP-8567, July 1975.
- 5. Butler, J. C., and Love, D.S., "Steamline Break Mass/Energy Releases for Equipment qualification Outside Containment," WCAP-10961, Rev. 1 (proprietary) and WCAP-11184 (nonproprietary), October 1985.
- 6. Morita, T,, et al., "Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10297-P-A (proprietary) and WCAP-10298-A (nonproprietary),
June 1983.
- 7. Letter (AEP:NRC: 10678) from M. P, Alexich (Indiana and Michigan Power Company) to the USNRC, dated February 6, 1989.
- 8. "American National Standard for Decay Heat Power in Light Water Reactors," ANSI/ANS-5. 1-1979, August 1979.
- 9. Letter (AEP:NRC: 1067A) from M. P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated December 30, 1988.
- 10. Letter (AEP:NRC: 1067C) from M. P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated March 14, 1989.
- 11. Letter (AEP:NRC: 1024D) from M. P. Alexich to T. E. Murley (NRC), dated August 22, 1988. Includes WCAP-11908, "Containment Integrity Analysis for Donald C. Cook Nuclear Plants, Units 1 and 2."
- 12. Letter, J. F. Stang (NRC) to M. P. Alexich (IMECo), dated January 30, 1989.}}