ML20236E015
| ML20236E015 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/30/1987 |
| From: | Carey R, Collins L, Lowe A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20236D974 | List: |
| References | |
| 77-1168262, 77-1168262-00, BAW-1901-S01, BAW-1901-S1, NUDOCS 8707310183 | |
| Download: ML20236E015 (21) | |
Text
i BAW-1901 Supplement 1 June 1987 1
i ANALYSIS OF CAPSULE THIl-C GPU NUCLEAR THREE MILE ISLAND NUCLEAR STATION-UNIT 1
-- Reactor Vessel Material Surveillance Program --
Supplement 1 PRESSURE-TEMPERATURE LIMITS by A. L. Lowe, Jr., PE.
R. L. Carey j L. L. Collins J. W. Ewing W. A. Pavinich W. E. VanScooter K. K. Yoon f
B&W Document No. 77-1168262-00 q
BABC0CK & WILCOX Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 8707310183 870724 PDR P ADOCK 05000289 Ba4> COCK & WilCOE PDR a McDermott company
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CONTENTS Page 1.
INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 2. 1 DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY l PRESSURE-TEMPERATURE LIMITS . . . . . . . . . . . . . . . . . . . 2-I J q
- 3. DEVELOPMENT OF TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS l
. . . . . . . . . . . . . . . . . . . 3-1 I
- 4. CERTIFICATION i
. . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 APPENDIXES j
A. Revised Technical Specifications Pres Operating Limitation . . . . . . . . sure-Temperature
. . . . . . . . . . . . . . . . A-1 B.
References . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.1 l
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Babcock &wiscox a MCDermott Comparty
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- 1. INTRODUCTION This supplement contains the reactor coolant pressure boundary pressure-temperature limits which are normally a part of the report providing the results and conclusions from the testing and evaluation of a capsule from the reactor vessel material surveillance program. The original report for the analysis of capsule TMIl-C from Three Mile Island Nuclear Station Number 1 is BAW-19011 and contains the data which supports the development of the pressure-temperature limits for normal operation, both heatup and cooldown, !
inservice leak and hydrostatic tests and reactor core operation as reported in this supplement.
in addition, the revised technical specifications and pressure-temperature limits as adjusted for Three Mile Island Unit No. I through 10 EFPY are contained in Appendix A.
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1-1 Babcock &Wilcox a ucoermott company l
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- 2. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRESSURE-TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Three Mile Island Unit-1 are established in accordance with the requirements of 10CFR50, Appendix G.2 The methods and criteria employed to establish operating pressure and temperature limits are described in topical l report BAW-10046A.3 The objective of these limits is to prevent nonductile )
failure during any normal operating condition, including anticipated opera-tion occurrences and system hydrostatic tests. The loading conditions of interest include the following: I
- 1. Normal operathns, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests. j l
- 3. Reactor core operation. '
The major components of the RCPB have been analyzed in accordance with 10CFR50, Appendix G. The closure head region, tho reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of j the reactor vessel (and consequently of the RCPB) that regulate the pressure-temperature limits. Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods. The reactor vessel outlet nozzle also affects the pressure-temperature limit curves of the first several service periods.
This is due to the high local stresses at the inside corner of the nozzle, which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure,- the RTNDT of the beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained 2-1 Babcock &Wilcox a McDermott company
a, through a point-by-point comparison of the limits imposed by the closure head region, the outlet nozzle, and the beltline region. .The maximum allowable pressure is taken to be the lowest of the three calculated pressures.
The limit curves for Three Mile Island Unit-1 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the tenth EFPY. The tenth EFPY was selected because it is estimated that the third surveillance capsule will be withdrawn at the end of the refueling cycle when the estimated capsule fluence corresponds to approximately the T/4 end-of-life value. The time difference between the
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withdrawal of the third and the currently available surveillance capsule data provides adequate data time for re-establishing the operating pressure and temperature limits for the period of operation beyond the third surveillance capsule withdrawal.
The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10CFR50, Appendixes G and Hs For the other beltline region and RCPB materials for which the measured properties are not available, the unitradiated impact properties and residual elements, as originally established for the beltline region materials, are listed in Table 2-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced RT and the unirradiated 3T NDT. The NDT predicted RT is calculated using the respective neutron fluence and copper NDT and nickel contents. Figure 2-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall.
The supporting information for Figure 2-1 are the predicted fluences that have been demonstrated in BAW-19011 to be conservative. The design curves of Regulatory Guide 1.99, Rev. 2,4 were used to predict the radiation-induc'ed
. RT NDT values as a function o'f the material's copper and nickel content and neutron fluence.
The neutron fluences and adjusted RT values of the beltline region NDT materials at the end of the tenth full-power year are listed in Table 2-1.
The neutron fluences and adjusted RT values are given for .the 1/4T and NDT 3/4T vessel wall locations (T = wall thickness). The assumed RT NDT of the closure head region and the outlet nozzle steel forgings is 60F, in accord-ance with BAW-10046A.
2-2 Babcock & Wilson a McDermott company d _ _ _
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. i Figure 2-2 shows the reactor vessel's pressure-temperature limit curve for normal heatup. This figure also shows the core criticality limits as required by 10CFR50, ' Appendix G. Figures 2-3 and 2-4 show the vessel's pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively. All pressure-temperature l limit curves are applicable through ten EFPY. Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits !
1 of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below and to the j right of the limit curve. The reactor is not permitted to go critical until !
the pressure-temperature combinations are to the right of the criticality limit curve. To establish the pressure-temperature limits for protection against nonductile failure of the RCPB, the limits presented in Figures 2-2 l'
through 2-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel controlling the limit curves. This is necessary because the reactor vessel is the most limiting component of the RCPB.
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- 3. DEVELOPMENT OF TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS The pressure-temperature limits established for the technical specification were determined for selected heatup and cooldown rates by comparing the individual uncorrected pressure-temperature curves for the nozzle, beltline, and closure head over the operating temperature range of the reactor vessel i (see Section 2). The limiting pressure (minimum) at each temperature was selected as the basis for developing the maximum pressure for setting the {
actual operating limitations. Differential pressure corrections were then applied to the resulting limiting curves to account for the pressure dif -
ferential between the analyzed regions of the reactor vessel and the system pressure sensor on the reactor ecolant system hot leg. Instrumentation errors for pressure and temperature were then applied to the limits corrected for sensor location.
J The resulting corrected data points were plotted to obtain a bounding techni-cal specification curve for normal operations. Also, heatup and cooldown curves at various rates (OF/hr) over the various operating temperature ranges were combined into composite, bounding operating limit curves.
The resulting changes to the applicable technical specification section and the revised pressure-temperature curves for Three Mile Island Unit 1 are shown in AppenOx A. Figure A-1 describes the normal heatup and cooldown curves plus the criticality limits. Figure A-2 describes the inservice leak and hydrostatic test curves.
3-1 Babcock & Wilcox a McDermott company
i i
- 4. CERTIFICATION l l
The pressure-temperature operating limits for Three Mile Island Unit I reactor pressure vessel were calculated using approved procedures and established methods and techniques in accordance with the requirements of 10CFR50, Appendix G.
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A. L. Lo(e, Jr., P.E. (/ / Cate Sf Project Technical Manager
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This report has been reviewed for technical content and accuracy.
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73 rr L.1. (Gro'ss, P.E. (material properties)Dat6 -
M&SA Unit K. K. Yoon, .E. (fracture analysis)' Date/
M&SA Unit Verification of independent review.
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A. D. McKim, Manager 7h/D Datb M&SA Unit This report has been approved for release.
CAJ!. i ret J. F. Walters'
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APPENDIX A Revised Technical Specifications Pressure-Temperature Operating Limitations i
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3.1.2 PRESSURIZATION HEATUP AND C00LDOWN LIMITATIONS '
Aeolicability Applies to pressurizer, heatup and cooldown of the reactor coolant system. -
Ob.iective To assure that the temperature and pressure changes in the reactor coolant system do not cause cyclic loads in excess of design for reactor coolant system components.
Specification 3.2.1 For operations until ten effective full power years, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3.1-2 and are as follows:
Heatuo/Cooldown/ Criticality Limit Allowable corrbinations of pressure and temperature shall be to the right of and below the corresponding limit line in Figure 3.1-1.
Heatup and cooldown rates shall not exceed those shown on Figure 3.1-1. Reactor coolant pump combinations are limited to those noted on Figure 3.1-1.
The reactor must not be made critical until the pressure-temperature ty limit curve.
combinations are to the right of the criticali-Inservice Leak and Hydrostatic Testina Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-2. Heatup and Cooldown rates shall not exceed those shown on Figure 3.1-2.
Reactor Coolant pump combinations are limited to those noted on Figure 3.1-2.
3.1.2.2 The secondary side of the steam generator shall not be pressurized above 200 psia if the temperature of the steam generator shell is below 1000F.
3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 1000F in any one hour.
The spray shall not be used if the temperature difference than 4300F.
between the pressurizer and the spray fluid is greater 3.1.2.4 Prior to exceeding ten effective full power years of operation, Figures 3.1-1 and 3.1-2 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.B. The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference tempera-A-2 Babcock & Wilcox 4 MCDermott Comparty
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The basis for this predic-ture at the end of the service period.
tion shall be submitted for NRC staff raview in accordance with Specification 3.1.2.5. )
3.1.2.5 The updated technical specifications referred to in 3.1.2.4 shall be submitted for NRC review at least 90 days prior to the end of the !
service period. Appropriate additional NRC review time shall be i allowed for proposed technical specifications submitted in accord- J ance with 10 CFR 50, Appendix G, SJction V.C. I Bases All reactor coolant system components are designed to sithstand the l effects loads due to system temperature and pressure 1 changes.bj cyclicThese cyclic loads are introduced by unit load transi- j ents, reactor trips and unit heatup and cooldown operations. The J number of thermal and loading cycles used for design purposes are )
shown in Table 4-8 of the FSAR. The maximum unit he cooldown rates satisfy stress limits for cyclic operation.gp and The 200 psia pressure limit for the secondary side of the steam genera-tor at a temperature less gn 1000F satisfies stress levels for temperatures below the NDTT. j The unirradiated nil ductility reference temperature RT for the surveillance region materials were determined in accordab with 10 CFR 50, Appendices G and H. For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties were estimated using the methods described in BAW-10046A, Rev. 2.
As a result of fast neutron irradiation in the beltline region of the core, there will be an increase in the RT with accumulated nuclear operations. The adjusted reference tehratures have been calculated by adding the predicted radiation-induced shift in RT j and the unirradiated RT NDT foreachofthereactorcoolantbeltlE materials.
The predicted shift in RT was calculated using the respective neutron fluence after ten Efective full power years of operation l I
and the procedures defined n Regulatory Guide 1.99, Revision 2. the analysis of the reactor vessel material contained in the second surveillance capsule removed from Three Mile Island Nuclear Station Unit change I confirmed in impact that the current properties due totechniques used irradiation are for predictedge conservative.
Analysis average fastof the fluxsecond surveillance during cycles I throughcapsule indicates 4 was 1,41 x 10 gat tge n/cm -
see maximum at the pressure vessel wall. Extrapolation of the flux based on predicted fuel reload and burn-up conditions indicates that the maximum average fast neutron (E > 1 MeV) gux during ten full power years of operation will 1.38 x 10 n/cm2
-sec at the reactor vessel wall and 7.00 x 10ge n/cm2-sec at the 1/4 T location.
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The fast neutron operation, exposure during teggeffecgive full power years of therefore, is yg.40 x 10 n/cm at the reactor vessel ingdesurface,1.90x10 n/cm2 at the 1/4 T location, and 4.40 x 10 n/cm at the 3/4 T location.
Based on the predicted RT after ten effective full power years of operation, the pressure /therature limits of Figure 3.1-1 and 3.1-2 havebeenestabghedinaccordancewiththerequirementsof10CFR 50, Appendix G. The methods and criteria employed to establish the operating pressure and temperature limits are as described in BAW-10046A, Rev. 2. The protection against nonductile failure is assumed by maintaining the coolant pressure below the upper limits of these pressure temperature limit curves.
The pressure limit lines on Figures 3.1-1 and 3.1-2 have been established considering the folloning:
- a. A 25 psi error in measured pressure.
- b. A 120F error in measured temperature.
- c. System pressure is measured in either loop.
d.
Maximum differential pressure between the point of system pressure measurement and the limiting reactor vessel region for the allowable operating pump combinations.
The spray temperature difference restriction, based on a stress analysis of spray line nozzle is imposed to maintain the thermal stresses limit. at the pressurizer spray line nozzle below the design Temperature requirements for the steam generator correspond with the measured NOTT for the shell.
References
- 1. FSAR, Section 4.1.2.4.
- 2. ASME Boiler and Pressure Code,Section III, N-415.
- 3. FSAR, Section 4.3.10.5.
- 4. A. L. Lowe, Jr., et al., Analysis of Capsule TMI-IC, GPU Nuclear, Three Mile Island Nuclear Station Unit 1, Reactor Vessel Materials Surveillance Program, BAW .1901, Babcock &
Wilcox, Lynchburg, Virginia.
- 5. A. L. Lowe, Jr., et al., Analysis of Capsule TMI-IC, GPU Nuclear, Three Mile Island Nuclear Station Unit 1, Reactor Vessel Materials Surveillance Program, Supplement 1 Pressure-Temperature limits, BAW-1901. Sucolement 1, Babcock & Wilcox, Lynchburg, Virginia, June 1986.
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i APPENDIX B 1 References 1 l
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i 4
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- 1. A. L. Lowe, Jr., et al., Analyses of Capsule TMI-1C Metropolitan Edison Company, Three Mile Island Nuclear Station Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1901, Babcock & Wilcox, Lynchburg, !
Virginia, March 1986.
- 2. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness I Requirements for Light-Water Nuclear Power Reactors, Appendix G Fracture j Toughness Requirements, Federal Register, Vol. 48, No.104, May 17,1983. 1 1
- 3. H. S. Palme, et al., Methods of Compliance With Fracture Toughness and I Operational Requirements of Appendix G to 10CFR50, BAW-10046A, Rev.1, }
Babcock & Wilcox, Lynchburg, Virginia, July 1977.
- 4. U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor i Vessel Material, Draft Regulatory Guide 1.99, Revision 2, Dated February l 10, 1986.
1
- 5. A. L. Lowe, Jr. , et al., Pressurized Thermal Shock Evaluations in !
Accordance With 10CFR50.61 for Babcock & Wilcox Owners Group Reactor l Pressure Vessesi, BAW-1895, Babcock & Wilcox, Lynchburg, Virginia, !
January 1986. l
- 6. A. S. Heller and A. L. Lowe, Jr., Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds, BAW-1803, j Babcock & Wilcox, Lynchburg, Virginia, January 1984. '
i l 7. J. D. Aadland, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor !
l Vessel and Surveillance Program Materials Information, BAW-1820, Babcock l
& Wilcox, Lynchburg, Virginia, December 1984.
]
- 8. K. E. Moore and A. S. Heller, BAW 177-FA Reactor Vessel Beltline Weld Chemistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July '
1983.
]
- 9. K. E. Moore, et al . , Evaluation of the Atypical Weldment, BAW-10144A, Babcock & Wilcox, Lynchburg, Virginia, February 1980.
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