ML20211N972
| ML20211N972 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 12/05/1986 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20211N947 | List: |
| References | |
| 2431K, NUDOCS 8612180358 | |
| Download: ML20211N972 (18) | |
Text
{{#Wiki_filter:. ATTACHENT 1 PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATION SECTIONS 3.3.2 and 4.3.2 PRESSURIZATION AND SYSTEM INTEGRITY i PAGES MODIFIED vii viii 79 84 85 86 87 88 89 90 91 92 93 PAGES ADDED ! 93a 8612180358 861205 PDR ADOCK 05000295 p PDR l 2431K
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F13 ure Page 3.3.2-1 Reactor Coolant System Heatup Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limitations 85 3.3.2-3 Fast Neutron Fluence (E > 1 MeV) as a Function of 86 Full Power Service Life (EFPY) for Zion Unit 1 3.3.2-4 Fast Neutron Fluence (E > 1 MeV) as a Function of 87 Full Power Service Life (EFPY) for Zion Unit 2 3.3.6-1 Dose Equivalent I-131 RC Limit versus Percent of Rated Thermal Power 124c
- 3. 4 -1 High Steam Line Flow Setpoint 131a 3.11-1 Restricted Area Boundry 226 6.1-1 Minimum Shift Crew Composition 327 LIST OF FIGURES (Continued) vil 10580/10590
_ _ - . - _ _ _ _ - _ _ _ _ - _ . . ~ . _ _ - . . . - . __ .- -._ _ _- -_ - . -. . .. Trblo Page. 1.1 Operational Modes 6b . 1.2 Surveillance Frequency Notation 6c 3.1-1 Reactor Protection System-Limiting Operations Conditions and Setpoints 30 3.1-2 Reactor Protection System Instrument Numbers 33 i j 3.3.2-1 Zion Unit 1 Reactor Vessel Toughness Data 88 3.3.2-2 Zion Unit 2 Reactor Vessel Toughness Data 89 3.3.4-1 In Service Inspection Program 106 1 4 3.3.5-1 Reactor Coolant Systems and Chemistry Specifications 122 ] 3.4-1 Engineered Safeguards Actuation System-Limiting Conditions on 129 j Operation and Setpoints i j 3.4-2 Engineered Safeguards System Instrument Numbers 132 4 3.7-1 Neutron Flux High Trip Points with Steam Generator Safety Valves 160a j Inoperable - Four Loop Operation 3.7-2 Neutron Flux High Trip Points with Steam Generator Safety Valves 160b 1 Inoperable - Three Loop Operation 3.11-1 Maximum Permissible Concentration of Dissolved or Entrained Noble 226a Gases Releases from the Site to Unrestricted Areas in Liquid Effluents j 3.11-2 Radioactive Liquid Effluent Monitoring Instrumentation . 228 3.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation 236 3.14-1 Radiation Monitoring Instrumentation 251 ) . 3.15-1 Equipment Requirement with Inoperative 4KV E.S.S. Bus 268 i l 3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus 269 1 l LIST OF TABLES 1 1150t/11511 v111
LIMITING CON 01110N f0R OPERATION -' SURVEILLANCE REQUIREMECT' 3.3.? PRESSURIZAllC3 AND SYSTEM INTEGR11Y 4.3.2 PRESSURIZAll0N AND SYSIEM INTEGRilY . A. Heatup and Cooldown A. The reactor coolant temperature and pressure The Reactor Coolant System temperature and shall be determined to be within the limits at least once per 30 minutes during system pressure (with the exception of the heatup, cooldown, and inservice leak and pressurizer) shall be limited in accordance hydrostatic testing. operations. with the limit lines shown in Figures 3.3.2-1 and 3.3.2-2 during heatup, cooldown and inservice leak and hydrostatic testing with:
- 1. a. A maximum heatup rate of 20*F/hr appilcable up to and including 180*F RCS indicated temperature. A maximum heatup rate of 60*F/hr applicable for RCS indicated temperatures greater than 180*F.
- b. A maximum cooldown of 100,*F in any I hour period.
- c. A maximum temperature change of <10*F in any 1 hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit .
curves. APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least MODE 3 within the next 6 hours and reduce RCS Tgyg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours. 10580/10590 79 065FA
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0 50 900 tSO 200 250 300 3SO 400 4SO SW l INDICATED TEMPERATURE (DED. F) i l The 20*F/hr heatup rate is applicable for all RCS indicated temperatures. The 60*F/hr heatup rate is applicable for RCS indicated temperatures greater than 180*F. l Zion 1 and 2 Reactor Coolant System Heatup Limitations applicable for up l to 15 EFPY and Heatup Rates up to 20*F/hr and 60*F/hr. Curves contain margins of 10*F and 60 psig for possible instrument errors.
- REACTOR COOLANT SYSTEM HEATUP LIMITATIONS I
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Bases 3.3.2 & 4.3.2 FRAClURE 100GHNESS PROPERllES The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boller and Pressure Vessel Code, Section 111, Appendix G, and 10CFR50 Appendix G. The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASIM E185-73, and in accordance with additional reactor vessel requirements, lhese properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section til of the ASME Boller and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and j Cooldown Limit Curves", Apr il 1975. 1 j Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference j temperature, RTNDT, at the end of 15 ef fective full power years (EFPY) of service life. The 15 EFPY service life period is chosen such that the limiting RTNOT at the 1/4T location in the core region is greater than the l RTNDT of the limiting unirradiated material. The selection of such a limiting RTNDI assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RINDT; the resultr of these tests are 4 shown in Tables 3.3.2-1 and 3.3.2-2. Reactor operation and resultant fast neutron (E greater than 1 MeV) trradiation can cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the l fluence, copper content, and nickel content of the material in question, can be predicted using Regulatory Guide 1.99, Revision 2. " Effects of Residual Elements on Predicted Radiation Damage to Rea,ctor Vessel Materials." The heatup and cooldown limit curves of Figures 3.3.2-1 and 3.3.2-2 include predicted adjustments for this shift in i RT NDT at the end of 15 EFPY as well as adjustments for possible errors in the pressure and temperature sensing 1 instruments. i i Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived f rom Appendix G in Section ill of the ASME Boller and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear ' elastic fracture mechanics (LEFM) technology. In the calculation procedures, a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in . I Appendix G of ASME Section 111 as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. fo assure that the radiation embrittlement ef fects are accounted for in the calculation of the limit curves, the most limiting value of ARI the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shif t, NDT, corresponding to the end of the period for which heatup and cooldown curves are generated. 10580/10590 90 4 0651A
^
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l ' The ASME approach for calculating the allowable limit curves for various heatup and cooldown' rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the references stress intensity f actor, Ki p for the. metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The KIR curve is given by the equation: KIR = 26.78 + 1.223 exp [0.014.5(T-RINDT + 160)] (1) Where: KIR is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RTNOT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: C Kgn + Kit i KIR (2) 1 Where: Kgg = the stress intensity factor caused by membrane (pressure) stress, Kit = the stress intensity factor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RTNOT of the material, 1, l C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations. At any time during the heatup and cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal ) stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, Kit, for the reference flaw is computed. From Equation (2) the pressure stress
- intensity factors are obtained and, from these, the allowable pressures are calculated.
1 ) Figures 3.3.2-1 and 3.3.2-2 define limits to assure prevention of rion-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the j heatup and cooldown rates that can be achieved over certain pressure-temperature ranges. i The leak test limit curve shown on the heatup curves (Fig. 7.3.2-1) represent minimum temperature requirements at ! the leak test pressure specified by ASME Section 111 and the NRC Standard Review Plan NUREG-0800. I l Allowable combinations of pressure and temperature for specified temperature change rates are below and to the j right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation. i i 10580/10590 91 j 0651A i 1 ;
HEATUP ** 1hree separate calculations are required to determine the limit curves for finite heatup rates. As is done in the - cooldown analysis, allowable pressure-temperature relationships are developed f or steady-state conditions as well ~ as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thernal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Kgp for the 1/4T crack during heatup is lower than the K IR for the 1/41 crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the ef fects of compressive thermal stresses and dif ferent KIR's for steady-state and finite heatup rates do not of fset each other and the pressure-temperature curve based on steady State conditions no longer represents a lower bound of all similar curves f or finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the ! lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second position of the heatup analysis concerns the calculation of pressure-temperature limitations for the i case in which a 1/41 deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the I thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature i and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on I both the rate of heatup and the time (or coolant temperature) along the heatup ravp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperatur e, the allowable pressure is taken to be the lesser of the three values taken f rom the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches f rom the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. 4 i lhe heatup rates of the Reactor Coolant System are limited to 20*F/hr for RCS indicated temperatures equal or less than IBO*F and 60*F/hr for RCS indicated temperatures greater than 180*F to comply with the requirements of 10CFR50 Appendix G. Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. i C00lDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which i 105H0/10590 92 j 0657A i 1
- increase with increasing cooldown rates. A110wIble pressure-temperature relattens are generated fer both -
steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed ~ for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the
- material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of Kgg at the 1/4T location for finite cooldown rates than for steady-state operation.
Furthennore, if conditions exist such that the increase in Kgg exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period. PRESSURIZER LIMITS Although the pressurizer operates in temperature ranges above those for which there is reason for concern of ' nonductile f ailure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. , HYDROSTATIC TESTING LIMIT CURVE ~ I Allowable pressure-temperature relationships for leak and hydrostatic testing are also calculated using methods ! } derived from Non-Mandatory Appendix G2000 in Section til of the ASME Boiler and Pressure Vessel Code. The approach specified is the same as described for heatup and cooldown limits except that the safety factor on Kgg { is reduced to 1.5 and there are no significant thermal transients or gradients. Thus the governing equation for ] the leak and hydrostatic testing analysis is: 3 1.5 K ig < Kgg INADVERTANT SAFETY INJECTION In the event of an inadvertant safety injection actuation, the affected reactor will trip immediately, place the reactor in the hot shutdown condition. Af ter 60 seconds safety injection may be reset and injection terminated as required. An inspection of the primary system while at hot shutdown will prevent possible degradations in the primary system from undergoing further immediate thermal shock imposed during a cooldown. If degradations in the , j ' primary system are discovered, an orderly controlled cooldown will be planned to minimize the effects of thermal shock on these degradations on the af fected unit. I ) 10580/10590 93
- 0657A l
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REFERENCES-
- 1. ASME Boiler and Pressure Vessel Code, Section 111, 1976 Summer Addenda. *
- 2. WCAP-7924-A, " Basis for lleatup and Cooldown Limit Curves", April 1975.
- 3. ASME Boller and Pressure Vessel Code,'Section 111, N-331.
- 4. ASME Boller and Pressure Vessel Code, Section 111, N-415.
- 5. FSAR, Chapter 4.3. ,
- 6. WCAP-8724, "ASME 111, Appendix G Analysis of the Commonwealth Edison Company Zion Unit 1 Reactor Vessel".
- 7. WCAP-8727, "ASME 111, Appendix G Analysis of the Commonwealth Edison Company Zion Unit 2 Reactor Vessel".
- 8. WCAP-10677, " Adjoint Flux Program for Zion Units 1 and 2".
- 9. Regulatory Guide 1.99 Revision 2.
- 10. Code of Federal Regulations, 100FR50 Appendix G, " Fracture Toughness Requirements."
- 11. WCAP-ll247, "lieatup and Cooldown Limit Curves for the Commonwealth Edison Company Zion Units 1 and 2 Reactor Vessel".
- 12. WCAP-10962, " Zion Units 1 and 2 Reactor' Vessel Fluence and RTp13 Evaluations".
- 13. CWE-86-563, " Low Temperature Overpressure Protection System Setpoing Analysis", August 26, 1986.-
- 14. CWE-865-588, Low Temperature Overpressure Protection System Setpoing Analysis Extensions" October 27, 1986.-
10500/10590 93a 0657A
ATTACHMENT 2 DESCRIPTION OF PROPOSED CHANCES Technical Specification Sections 3.3.2, and 4.3.2 contain the reactor coolant system heatup and cooldown limitation curves that are presently applicable to 8 effective full power years (EPFY). Zion Units 1 and 2 are approaching 8 EFPY and require revisions to the curves for continued operation. Additionally, portions of sections 3.3.2, and 4.3.2 are being updated to reflect current methods, conditions, and analyses used in generating the new heatup and cooldown curves. Specific reasons for the updating follows. - EXPLANATIONS OF CHANGES ? , Bases: Changes to the heatup, cooldown curves have been made to confona with requirements of the Standardized Tech Specs. Page 79, Specification 3.3.2.A.1.a Heatup rates are changed to reflect the revised heatup curve (p. 84) to extend curve applicability to 15 EFPY. Page 84 The heatup curve is being updated to derive the 15 EFPY curve as addressed in WCAP-11247. Page 85 The cooldown curve is being updated to derive the 15 EFPY curve as addressed in WCAP-11247. Page 86 The current figure 3.3.2-3 has been removed and replaced with a Neutron Fluence vs EFPY figure for Unit 1. Information pertaining to the old figure is contained in WCAP's-10902, 11247, and 7924-A. This information is required only for deriving the limitation curves and is used in the WCAP's. Page 87 Figure 3.3.2-4 is being replaced by more accurate curves for Unit 2 which were used in the analysis of the 15 EFPY heatup, cooldown curves. Actual derivation of the fluence curves is explained in WCAP-10902 and are referenced in WCAP-11247. This information also applies to the new figure 3.3.2-3.
r-Page 88, 89 Table 3.3.2-1 is being replaced by more current information used in deriving the limitation curves. The values in the 50 ft-lb/35 MIL and TRANS USE columns are different due to the use of methods in NUREG-0800. These values replace the old ones which were generated by Westinghouse and were accepted by the NRC at that time. Some of the values in the RT NDT column changed due to influence of the 50 ft-lb/35 MIL column. The table currently in the Tech Specs is only for Unit 1; no information was available at that time for Unit 2. These new tables are referenced on page 90 of the proposed Tech Specs. Page 90-93 The bases section on fracture toughness properties is being changed to conform with the Standardized Tech Specs. Changes in this section were also made to correspond with the proposed LCO section (e.g. 15 EFPY curves, figures, tables). It should also be noted that equation 2.0 K ig
+ 1.25 kit 19KIR has been modified to reflect the equation contained in the Standardized Tech Specs.
Page 93a The listing of references has been updated. i ir r l l l 2431K ! - =_
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- e ATTACHMENT 3 WCAP-11247 l
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