ML20206H282
| ML20206H282 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/03/1999 |
| From: | Stephen Dembek NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20206H284 | List: |
| References | |
| NUDOCS 9905110134 | |
| Download: ML20206H282 (15) | |
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9 Ji UNITED STATES
.o NUCLEAR REGULATORY COMMISSION If 8
WASHINGTON, D.C. 20555 0001
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PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 j
DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.133 License No. DPR-30 i
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated September 3,1998, as supplemented by letters dated January 22,1999.,
4 February 5,1999, and March 17,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will 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-80 is hereby amended to read as follows:
i 9905110134 990503 i
PDR ADOCK 05000275 P
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l (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in' Appendix B, as revised through Amendment No. 133, i
are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Stephen De bek, Chief, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: May 3, 1999 l
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UNITED STATES
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PACIFIC GAS AND ELECTRIC COMPANY
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DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.131 License No. DPR-82 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated September 3,1998, as supplemented by letters dated January 22,1999, February 5,1999, and March 17,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, 4
B.
The facility will 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 l
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 i
Operating License No. DPR-82 is hereby amended to read as follows:
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l (2) Technical Soecifications The Technical Specifications cohtained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 131, are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
- 3..
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from date ofissuance.
FOR THE NUCLEAR REGULATORY COMMISSION Stephen Dembek, Chief, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications
- Date of Issuance: May 3, 1999 4
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i ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO.133 TO FACILITY OPERATING LICENSE NO. DPR-80 i
AND AMENDMENT NO. 131 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by amendment number and contain marginallines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT viii viii 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 8 3/4 4-7 8 3/4 4-7 j
B 3/4 4-12 B 3/4 4-12 B 3/4 4-15 B 3/4 4-15 B 3/4 4-16 B 3/4 4-16 i
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INDEX LIMITING CONDITION 3 FOR OPERATION AND SVRVEILLANCE RE0VIREMENTS SECTION EA(if 3/4.3 INSTRUMENTATION (continued)
Explosive Gas Monitoring Instrumentation.................
3/4 3-59 3/4.3.4 DELETED l
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1
. REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................
3/4 4-1 Hot Standby..............................................
3/4 4-2 Hot Shutdown..............................................
3/4 4-3 Cold Shutdown - Loops F111ed.............................
3/4 4-5 Cold shutdown - Loops Not F111ed................,........
3/4 4-6 3/4.4.2 SAFETY VALVES Operating................................................
3/4 4-8 3/4.4.3 PRESSURIZER..............................................
3/4 4-9 3/4.4.4 RELIEF VALVES...........................................
3/4 4-10 3/4.4.5 STEAM GENERATORS........................................
3/4 4-11 DIABLO CANYON - UNITS 1 & 2 vii Unit 1 - Amendment No. 57,75,00,120 Unit 2 - Amendment No. 55,74,07,118
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.4 REACTOR COOLANT SYSTEM (continued]
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.,
3/4 4-16 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION...
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4 4-18 Leakage Detection Systems..
3/4 4-19 Operational Leakage...
. TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION 3/4 4-21 VALVES.......
3/4 4-25 3/4.4.8 SPECIFIC ACTIVITY..
FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY 3/4 4-27
> 1 pCl/ GRAM DOSE EQUIVALENT l-131....
TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE
.. 3/4 4-28 AND ANALYSIS PROGRAM.........
3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS....
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
l 3/4 4-31 APPLICABLE UP TO 16 EFPY....,..........
FIGURE 3.4 3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -
l 3/4 4-32 APPLICABLE UP TO 16 EFPY..
... 3/4 4-35 Overpressure Protection Systems...............
DIABLO CANYON - UNITS 1 & 2 viii Unit 1 - Amendment 54,00.100,133 Unit 2 - Amendment 50,0-',00,131
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Leak Test Limit i
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for heatup rates up to 60 'F/hr.
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2000 No allowance included for possible I
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0 SO 100 150 200 250 300 350 400
'450 RCS TEMPERATURE (*F)
Controlling Material:
1/4T:
Unit 1 Lower Shell Longitudinal Weld 3-442 C.
RTwt 9 1/4T = 183.7'F.
3/4T:
Unit 2 Intermediate Shell Plate B5454-2. RTwt 9 3/4T = 151.4*F.
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY l
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DIABLO CANYON. UNITS 1 & 2 3/4431 Unit 1 - Amendment No.54-100,133 Unit 2 - Amendment No. 53-99131
2500
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Applicable for service through 16 EFPY for 2250 -
cooldown rates up to 100'F/hr.
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No allowance included for possible instrument " ~__._~:~ ~ -" ~.
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0 50 100 150 200 250 300 350 400 450 RCS FLUID TEMPERATURE (deg. F)
Controlling Haterial:
1/4T: Unit 1 Lower Shell Longitudinal Weld 3-442 C.
RTndt 9 1/4T = 183.7*F.
3/4T: Unit 2 Intermediate Shell Plate 85454-2.
RTndt 9 3/4T = 151.4'F.
FIGURE 3.4 3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY l
DIABLO CANYON - UNITS 1 & 2 3/4 4-32 Unit 1 - Amendment No.54-400,133 Unit 2 - Amendment No. 53-99.131 l
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- REACTOR COOLANT SYSTEM BASES
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3/4 4.9 PRESSURE / TEMPERATURE LIMITS f
'The temperature and pressure changes during heatup and cooldown are limited to be-consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section i
Ill, Appendix G, and Section XI, Appendix G:
1.
The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
a.
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and b.
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
2.
These limit lines shall be calculated periodically using methods provided below,
- 3.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generatoris below 70*F, 4.
Deleted 5.
' System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI. Allowable pressures and temperatures for inservice leak and hydrostatic tests are given in Figure 3.4-2.
~ 6.
The criticality limit on Figure 3.4-2 is based on the minimum allowable temperature of 314*F for an inservice hydrostatic test of 110% of operating pressure.
.l
' The fracture toughness testing of the ferritic materials in the reactor vessel was performed
- in accordance with the 1966 Edition for Unit 1 and the 1968 Edition for Unit 2 of the ASME Boiler and Pressure Vessel Code, Section Ill. These properties are then evaluated in i
accordance with the NRC Standard Review Plan.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, RT, at the end of 16 effective full power years (EFPY) of_ service life.
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-7 Unit 1 - Amendment No. 54,00,100,133 Unit 2 - Amendment No. 53,07,00,131
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THIS PAGE INTENTIONALLY DELETED.
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DIABLO CANYON - UNITS 1 & 2 g 3/4 4.g Amendment Nos'. 54 and 53
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THIS PAGE INTENT 10NALLY DELETED.
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DIABLO CANYON - UNITS 1 & 2 B 3/4 4-11 Amendment Nos. 54 and 53 L
- BASES PRESSURE / TEMPERATURE LIMITS (Continued)
The selection of such a limiting RTwor assures that all components in the Reactor Coolant l
System will be operated conservatively in accordance with applicable Code requirements.
1he reactor vessel materials have been tested to determine their initial RTuor; the results of these tests are shown in the FSAR Update. Reactor operation and resultant fast neutron (E greater than 1 MeV) Irradiation can cause an increase in the RTuor. Therefore, an adjusted reference temperature, based upon the fluence, copper content and nickel content of the materialin question, can be predicted using value of ARTsar computed by Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," for the maximum l
neutron fluence at the locations of interest. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTuor at the end of 16 EFPY.
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Values of ARTuor determined in this manner will be used until the results from the material surveillance program, evaluated according to ASTM E185-82, can be used. Capsules will be removed in accordance with the requirements of ASTM E185 and 10 CFR Part 50, Appendix H.
The surveillance specimen withdrawal schedule will be maintained in the FSAR Update. The lead factor represents the relationship between the fast neutron flux density at the location of
-the capsule and the inner wall of the reactor vessel. The heatup and cooldown curves must be recalculated when the ARTuor determined from the surveillance capsule exceeds the calculated ARTuor for the equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section XI of the ASME Boiler and l
- Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in the following paragraphs.
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 semi-elliptical 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 or outside surface of the vessel wall.
The dimensions of this postulated crack, referred to in Appendix G of ASME Section XI as the maximum postulated defect, 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 non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the DIABLO CANYON - UNITS 1 & 2 B 3/4 4-12 Unit 1 - Amendment No. 54rt00,133 Unit 2 - Amendment No. 59;99,131
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BASES PRESSURE / TEMPERATURE LIMITS (Continued) 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 portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.
l Unlike the situation at the vessel inside surface, the thermal gradients established at the outside l,
surface during heatup produce stresses which are tensile in nature and thus tend to reinforce L
any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, l
since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a l
lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis, i
Following the generation of pressure-temperature curves for both the steady-state and l
finite heatup rate situations, the final limit curves are produced as follows. A composite curve is i
constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three l
values taken from the curves under consideration.
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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 l
controlling condition switches from the inside to the outside and the pressure limit must at all l
times be based on analysis of the most critical criterion.
LOW TEMPERATURE OVERPRESSURE PROTECTION l
l The OPERABILITY of both Class 1 PORVs or an RCS vent opening of at least 2.07 l
square inches ensures that the RCS will be protected from pressure transients that could L
exceed 110% of the limits of Appendix G to 10 CFR Part 50 when operating at low l
l
' temperatures. Low temperature is defined as less than or equal to the reactor coolant l
temperature corresponding to a reactor vessel wall temperature of RTuor + 50'F, where RTsor l
is evaluated at the beltline location (1/4T), which is controlling in the Appendix G Pressure-Temperature (60*F/hr heatup) limits. These pressure and temperature requirements are consistent with the guidelines and definitions in ASME Code Case N-514. The LTOP enable l
temperature applicable through 16 EFPY is 270*F.
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DIABLO CANYON - UNITS 1 & 2 B 3/4 4-15 Unit 1 - Amendment No. 01,00,100,133 Unit 2 - Amendment No. 00,07,00,131
1 8_EACTOR COOLANT SYSTEM
_ BASES LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued)
OPERABILITY of the PORVs for LTOP use requires a lift setting of less than or equal to 435-psig. This setpoint ensures that either Class 1 PORV has adequate relieving capability to protect the RCS from overpressurization for all anticipated transients, concurrent with any single active failure. The limiting transient for LTOP is a mass injection event based on the combined ECCS injection line flow from one centrifugal charging pump and the positive displacement pump, into a water-sold RCS, with letdown isolated. The 435 psig setpoint was determined for this event based on a single, OPERABLE PORV, reactor service less than or equal to 16 EFPY, and administrative controls on RCP operation, charging pump operability, and the ECCS injection flow path. The instrument uncertainties are not included in the Technical Specification setpoints. Uncertainties associated with LTOP instrumentation were determined in accordance with the guidance provided in WCAP-14040-NP-A. An allowance for the pressure uncertainty is provided by administrative controls as discussed above.
The Maximum Allowed PORV Setpoint for the LTOPs will be modified, if required, based on the results of examinations of the reactor vessel materialirradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H. The surveillance specimen withdrawal schedule is maintained in the FSAR Update.
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-16 Unit 1 - Amendment 00,100,123,133 Unit 2 - Amendment 97,00,121,131 i