ML19319D875
| ML19319D875 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/20/1977 |
| From: | Goller K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19319D872 | List: |
| References | |
| NUDOCS 8003270531 | |
| Download: ML19319D875 (9) | |
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UNITED STATES sNUCLEAR REGULATORY COMMISSION s
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s, SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING L,LCENSE Amendment No. 15 License No. DPR-54 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment Dy Sacramento Municipal Utility District (the licensee) dated July 9, 1976, as supplemented Auaust 24, 1976 and December 17, 1976, comolies 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; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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.
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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-54 is hereby amended to read as follows:
'(2). Technical Specifications The Technical Specifications contained in Appendices A and B, as revised throuoh Amendment No.15, 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 COPMISSION AU#Gd4 Karl R. Goller, Assistant Director for Operatina Reactors Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 20, 1977 e
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ATTACHMENT TO LICENSE AMENDMENT NO.
15 FACILITY OPERATING LICENSE NO. DPR-54 DOCKET NO.'50-312
' Revise Appendix A as follows:
Remove Pages Insert Paaes 3-3a 3-3a 3-4 34 4-10 & 4-12 4 4-12e 4
New pages_and changes on the revised pages are shown by marginal-lines i
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u RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Cenditions for Operation 3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 100 F/hr.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 410 F.
3.1.2.6 Within two years of power operation, figures 3.1.2-1 and 3.1.2-2 accordance with appropriate criteria accepted shall be updated in by the NRC.
Bases n
the effects of All reactor coolant system components are designed to withst cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by uni", load transients, reactor trips and unit heatup and cooldown operations. The number of thermal and loading cycles used.'or design in table 4.1-1 of the FSAR.
The maximum unit heatup and purposes are shown100 F per hour satisfies stress limits for cyclic operation.(2) cooldown rate of The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 130 F satisfies stress levels for temperatures below the DTT.(3)
The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximum NDTT value of 10 F has been determined based on Charpy V-r.tch tests.
The maximum NOTT value obtained for the steam generator shell material and welds was 70 F.
Figures 3.1.2-1 and 3.1.2-2 contain the linitin(g)reactorcoolant system pressure-temperature relationship for operation at DTT and below to assure that stress levels are low enough to preclude brittle fracture. These stress levels and their bases are defined in paragraph 4.3.3 of the FSAR.
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RANCHO SECO UNIT 1 TECHNirAL SPECIFICATIONS Limiting Conditions for Operation Bases (continued)
- As a result of f ast neutron irradiation in the region of the core, there will be an The predicted i
NDTT increase in the NDTT with accumulated nuclear operation.exposureisshownonfigur The actual increase for the 40 year shift in NDTT will be determined periodically during plant operation by testing of irradiated vessel material samples located in Davis Besse Unit 1.(5)
The results of the irradiated sample testing will be evaluated and compared to the design curve (figure 4.3-2 of FSAR) being used to predict the increase in transition temperature.
The design value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 3.0 x 1010 n/cm sec at 2,772 MWt rated power and an integrated exposure of 3.0 x 10l9 2
n/cm2 for 40 years operation.(6)
The calculated maximum values are 2.4 x 1010 n/cm2 sec and 2.4 x 10I9 n/cm2 integrated exposure for 40 years operation at 80 percent load.f4)
Figure 3.1.2-1 is based on the design value which is considerably higher than the calculated value.
The DTT value for figure 3.1.2-1 is based on the projected NDTT at the end of the first two years of operation.
During these two years, the energy output has been conservatively estimated to be 1.8 x 106 thermal megawatt days, which is equivalent to 655 days at 2,772 iiWt core power.
The projected f ast neutron exposure of the reactor vessel for the two years is 1.7 x 10I8 n/cm2 which is based on the 1.8 x 106 thermal megawatt days and the design value for fast neutron exposure.
in NDTT will be established periodically during plant operation by The oc t ua l shif t testing vessel material samples.
Samples are irradiated by securing them near the inside wall of the vessel in the core area of Davis Besse Unit 1.
To compensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown.
The NDTT shift and the magnitude of the thermal and pressure stresses are sensitive to integrated reactor power and not to instantaneous power level.
Figures 3.1.2-1 and 3.1.2-2 are applicable to reactor core thermal ratings up to 2,772 MWt.
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Amendment No; 7, 15
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Standards 4.2 REACTOR COOLANT SYSTEM SURVEILLANCE Applicability Applits to the reactor vessel, the reactor coolant system and its components.
Objective To establish examinations whereby the reactor coolant system and component integrity is monitored.
Specification 4.2.1 The reactor vessel material irradiation surveillance specimens removed from the reactor vessel at approximately 170 effective full power days shall be installed, irradiated in and withdrawn from the Davis-Besse Unit No. I reactor vessel in accordance with the schedule shown in Table 4.2-1.
Following withdrawal of each capsule listed in Table 4.2-1, SMUD shall be responsible for testing the specimens and submitting a report of test resulte in accordance with 10 CFR 50, Appendix H.
4.2.2 An Inservice inspection shall be made conforming as closely as design permits to the rules of the ASME Boiler and Pressure Vessel Code Section XI, Rules for Inservice inspection of Nuclear Reactor Coolant Systems with revisions approved as of June 1973, tables15-261,15-251 and 515-240 of this Code will be used as a guide for determining the examination frequencies and the applicable specific areas to be examined.
The-in-spection interval will be ten years.
As part of the inservice inspec-tion, hydrostatic tests will be performed as prescribed under Section 15-500 of this Code.
4.2.3 A preoperational examiration will be made to include all the items that would normally be completed throughout the inspection interval.
This survey will establish initial system integrity and provide a baseline for future testing.
4.2.4 Each reactor coolant pump motor flywheel will be inspected volumetrically during the ten year inspection interval.
One hundred percent of the fly-wheel will be examined. All flywheels received a one hundred percent ultrasonic examination prior to installation on the motor.
Because the reactor coolant system was not designed to meet the require-ments of Section XI fo the ASME Boller and Pressure Vessel Code, complete compliance is not feasible or practical.
However, access for inservice inspection has been considered and design modifications made where practical.
Therefore, where possible,Section XI of this Code will be utilized in
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the -conduct of this program.
Table 4.2-2 itemizes those areas where complete compliance with the code is not possible because of specific design and construction' details.
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RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Standards 4.2.5 if as a result of any of these inspections, defects are found to develop, further examinations will be made as needed to determine the exact condition.
Following evaluation of this evidence, a decision will be made to the effect upon plant safety and the requirements for repairs.
4.2.6 Records of each inspection shall be kept to permit evaluation and future comparison.
Periodic consideration will be given to incorporation of new or improved 4.2.7 inspection techniques into the surveillance program.
4.2.8 A report or application for license amendment shall be s_bmitts.' to th NRC within 90 days af.er the occurre nee of any of the following:
1.
Failure of Davis-Besse Unit No. I to achieve commercial operation at 100% power by January 1,1978, or 2.
Beginning one year after attainment of commercial I
operation at 100% power, any tims that Davis-Besse Unit l
No.1 fails to maintain a cumulative reactor utilization j
factor of greater than 65%.
i The report shall provide justification for continued operation of Rancho Seco with the reactor vessel surveillance program conducted i
at Davis-Besse Unit No. 1 or the application for license amendment
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shall propose an alternative program for conduct of the Rancho Seco reactor vessel surveillance program.
Bases Irradiation surveillance provides the capability of determining radiation induced changes in the mechanical and impact properties in the region of the reactor vessel surrounding the core.
Test specimens of base metal, deposited weld metal and the heat-affected zone are installed in capsule assemblies placed inside the i
vessel.
In'accordance with the schedules of Table 4.2-1 specimens will be re-moved; and a series of drop weight tests, Charpy impact tests and tension tests f
will be conducted.
Threshold neutron flux detectors and maximum temperature detectors will be installed with the specimens.
Changes in nil-ductility transition temperature wl'1 be determined, and appropriate alteration to plant operating parameters will be made.
Preoperational and inservice inspections emphasize areas of highest stress-concentration and probability of failure.
The area predominantly selected v
for these examinations are welds and the adjacent metal.
Examination of the welds is of ten by a volumetric (ultrasonic or radiography) method which, when performed, examines surrounding base metal and the weld heat-affected zone.
Both testing methods will use present state-of-the-art equipment operated by highly trained personnel qualified within the requirements of the applicable codes.
RANCH 0'SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards To assure the availability of adequate surveillance data.for the Rancho Seco I reactor vessel, a program has been developed to moni tor the i rradiation No.
the Davis Besse No, i reactor, and of the surveillance specimen capsules at Fluence compare this to the irradiation of the Rancho Seco No.1 reactor vessel.
estimates which are cons'ervative in the appropriate direction are used for this The f requency of monitoring varies depending on the known neutron comparison.
This provides lead factor between the capsules and the reactor vessel.
fluence cmple time for anticipating problems and initiating corrective action should
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operation of the host reactor be seriously delayed.
For the purpose of Technical Specification 4.2.8, the definition of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term " commercial operation".
Cumulative reactor utilizarion factor is defined as:
[(Cumulative thermal megawatt hours since attainment of commercial operation at 100% power) x 100] + ((licensed thercial power) x (cumulative hours since attainment of commercial operation at 100!. power)].
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Amendinent No. 15
Table 4.2-1 Rancho Seco CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 CAPSULE INSERTION / WITHDRAWAL RSI-B Withdraw at end of first cycle.
RSI-E Insert at end of first cycle, withdraw at end of tenth cycle.
RSI-D Withdraw at end of second cycle.
RSI-A Insert at end of second cycle, withdraw at end of seventh cycle.
RSI-C Withdraw at end of twelfth cycle.
RSI-F Withdraw at end of ninth cycle.
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AmendmentNo.j/',
15 4-12a
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