ML19309B072
| ML19309B072 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/17/1976 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19309B066 | List: |
| References | |
| NUDOCS 8004020677 | |
| Download: ML19309B072 (8) | |
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G' RANCHO SECO UNIT-1 TECHNICAL SPECIFICATIONS Limiting Conditions 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 l
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 shall be updated in accordance with appropriate criteria recepted by the NRC.
43 Bases Allreactorcoolantsystemcomponentsaredesignedtowithstynjtheeffectsof I
cyclic loads due to sy!, tem temperature and pressure changes.
These cyclic loads are introduced by unit load transients, reactor trips and unit heatup and cooldown operations.
The number of thermal and loading cycles used for design purposes are shown in table 4.1-1 of the FSAR.
The maximum unit heatup and 1
cooldown rate of 100 F per hour satisfies stress limits for cyclic' operation.(2)
The 200 psig pressure limit for the secondary side of the steam generator at a l
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-notch tests.
The maximum NDTT value obtained for j
the steam generator shell material and welds was 70 F.
-i Figures.3.1.2-1 and 3.1.2-2 contain the limiting) reactor coolant system pressure-temperature' relationship for operation at DTT M 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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Proposed Amendment No. 43 3-3a I
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RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (continued)
As a-result of fast neutron irradiation in the-region of the core, there will be an increase in the NDTT with accumulated nuclear operation.
The predicted m9ximum NDTT
' increase for the 40 year exposure is shown on figure 4.3-1 of the FSAR.(41 The actual 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 45 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 desi n value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 0 n/cm sec at 2,772 MWt rated power and an integrated exposure of 3.0 x 1039 2
3 0 x 10 2
n/cm2 for 40 yea { operation.(6)
The calculated maximum values are 2.4 x 1010 n/cm_
sec aryd 2.4 x 10 n/cm2 integrated exposure for 40 years operation at 80 percent load.14)
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 MWt core power.
The projected fast neutron exposure of the reactor vessel for the two years is 1.7 x 1018 n/cm2 which is based on the 1.8 x 106 thermal megawatt days and the design value for fast neutron exposure.
The actual shift in NDTT will be established periodically during plant operation by 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 43 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 t
to integrated reactor power and not to instantaneous power level.
Figures 3.I'.2-1 and 3.1.2-2 are applicable to reactor core thermal ratings up to 2,772 MWt.
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i Proposed Amendment.No. 43 I
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,7 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.2 REACTOR COOLANT SYSTEM SURVElLLANCE Applicability App 1Ies 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 program for irradiation survelliance of the reactor vessel materials to monitor changes in the mechanical and impact properties shall be per-formed as described in paragraph 4.4.4 of the FSAR.
Removal of specimens from capsules within the Davis Basse Unit I reactor shall be as scheduled 43 in table 4.2-1.
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 appilcable 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-i ments of Section XI fo the ASME IBoiler 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 the conduct of this program. Table 4.2-2 Itemizes those areas where complete. comp 11ance with the code is not possible because of specific design and construction details.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Survel11ance 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.
4.2 7 Periodic consideration will be given to incorporation ~of new or Improved inspection techniques into the surveillance program.
i 4.2.8 Surveillance specimens will be withdrawn at approximately 170 effective full power days and reinserted at the initial fuel cycle of Davis Besse 43 Unit i reactor. The capsules will be withdrawn from the Davis Besse reactor in accordance with the Table 4.2-1 schedule.
4.2.9 Starting with initial operation of cycle 2, the reactor vessel survelliance specimen capsule lead time shall be determined at the frequencies specified 43A* {
in table 4.2 'la. Subsequent to cycle 4 operation, the minimum lead time is 2.5 EFPY.
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 met 6 and the heat-affected zone are installed in capsule assemblies placed inside the 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 will be conducted.
Threshold neutron flux detectors and maximum temperature detectors will be installed with the specimens.
Changes in nil-ductility transition temperature will 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 for these examinations are welds and the adjacent metal.
Examination of the welds is often 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 appilcable codes.
Proposad 4-11 Amendment No. 43
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q RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Survel11ance Standards To assure the availabl11ty of adequate surveillance data for the Rancho Seco No.1 reactor vesse1 ~, a. program has been developed to monitor the i rradiation of the surveillance specimen capsules at the Davis Besse No. I reactor, and compare this to the irradiation of the Rancho Seco No.1 reactor vessel.
Fluence estimates which are conservative in the appropriate direction are used for this comparison. The frequency of monitoring varies depending on the known neutron fluence lead factor between the capsules and the reactor vessel.
This provides ample time for anticipating problems and initiating corrective action should 43A.1 operation of the host reactor be seriously delayed. The requirenent that the lead factor be 2.5 EFPY by the end of Rancho Seco No.1, cycle 4, or corrective action be developed provides assurance that survelliance date will be available in a timely manner to allow revisions to Technical Specification 3.1.2.3 The lead factor of 2.5 EFPY is based on a.8 capacity factor and thus provides over 3 calendar years for consideration and implementation of all alternatives.
The requirement of a factor of 2.5 EFPY lead time need not be implemented prior to cycle 4 operation since Technical Specification '3.1.2.3 will be reviewed for adequacy and updated if needed.
This based on recent surveillance capsule test results, which Justify operation for a period in excess of 4 cycles of ope rat i on.
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i Proposed 4-12 Amendment No. 43
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p' s-RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE'4.2-1 DAVIS BESSE UNIT I 43 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE OF RANCHO SECO SPECIMENS Fi rs t Capsu l e-----------------------At the t ime when p red i cted shi f t of C V adjusted fracture energy curve is approximately 50 F or.at one-fourth service life,- whichever is earlier.
Second and Third Capsule------------At approximately one-third and two-thirds of the time interval between first and fourth capsule withdrawal.
Fourth Capsule----------------------Three-fourths of service life.
Fifth Capsule-----------------------Standby i
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-Proposed 4,,
Amendment No. 43
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.2.la.
SURVEILLANCE SPECIMEN CAPSULE-IRRADIATION COMPARISON FREQUENCY Capsule Lead (EFPY)
Comparison Frequency Less than or equal to 1 Monthly Greater than I and less than or equal to 3.5 BI-Monthly 43A.I Greater than 3.5 and less than or equal to 4.5 Semi-Annual Greater than 4.5 Annual The equivalent effective full power years (EFPY) of predicted neutron fluence (E > 1 MEV) received by the Rancho Seco No. I reactor vessel at the surface location shall be subtracted from the equivalent EFPY of predicted neutron fluence (E > 1 MEV) received by the reactor vessel survelliance specimen capsules at Davis Besse No. I to determine the capsule lead factor.
Proposed 4-12b Amendment No. 43
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ADMINISTRATIVE f4NTROLS b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
c.
The change is documented, reviewed by the PRC, and approved by the Plant Superintendent within 7 days of implementation.
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 Information to be reported to the Commission, in addition to the reports required by Title 10, Code of Federal Regulations, shall be int accordance with the Regulatory Position in Revision 4 of Regulatory Guide 1.16, " Reporting of Operating Information - Appendix "A" Technical Specifications."
SPECIAL REPORTS 6.9 2 Special reports shall be submitted to the Director of the Regulatory Operations Regional Office within the time period specified for each report.
These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
A.
Startup Report A summary report of unit startup and power escalation testing and an evaluation of the results from these test programs shall be submitted within 60 days foilowing commencement of commercial power operation.
The test results shall be compared with design predictions and specifications.
B.
A Reactor Building structural integrity report shall be submitted within 00 days of completion of each of the following tests covered by Tec ilcal Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1).
.1 Annual inspection
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-Tendon stress surveillance
.3 End anchorage concrete survelliance
.4 Liner plant survelliance-C.
Inservice Inspection Program.
D.
If, subsequent to' cycle 4 the reactor vessel surveillance specimen lead time as described in Table 4.2.2 is less than 2.5 EFPY, a re-port describing the means of providing the necessary surveillance 43A.1 for the reactor vessel shall be submitted within 5 months of dis-covery of the inadequate lead time.
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$roposed Amendment No. 43
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