ML19309A671
| ML19309A671 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/08/1976 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19309A668 | List: |
| References | |
| NUDOCS 8003310643 | |
| Download: ML19309A671 (6) | |
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HANCHO 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 the pressurizer and the spray fluid is greater than 410 F.
3.1.2.6 Uithin tua yctrs of power operation, figures 3.1.2-1 and 3.1.2-2 shall be updated in accordance with appropriate criteria accepted by the AEC.
3.1.2.7 Surveillance specimens will be withdrawn at approximately 170 ef fective full power days and reinserted at the end of the first fuel cycle to allow modifications to be incorporated into the specimen holder configuration. Thereafter, capsules will be with-drawn in accordance with the following schedule:
First ccysule At the time when predicted shift of C adjusted fracture energy curve y
is approximately 50 F or at one-fourth service life, whichever is carlier.
Second e.nd third capsules At approximately one-third and two-thirds of the time interval between first and fourth capsule withdrawals.
Fourth capsule Three-fourths of service life.
Fifth capsule Standby.
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Eancs All reactor coolant system componenth are de signed to withstand the ef fects of cyclic loads due to system temperature and pressure changes.(1)
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"O RANCHO SECO UNIT 1
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TECHNICA1. SPECIFICATIONS C
Limiting Conditions for-Operation' W.
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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 FSt.".
The maximum unit heatup and coo 3doun rate of 100 F per hour catisfies stress limits for cyclic opera-tion.(2) The 200 psig pressure limit for the cecondary side of the secan generator at a temperature less than 130 F satisfics stress levels for tem-pcratures below the DTT.(3) The reactor vessel plate material and wolds have
' been tested to verify conformity to specified requirencnts and a maximum NDTT value of 10 F has been determined based on Charpy V-notch tests. The maximum NDTT value obtained for the steam generator shc11 material and welds was 70 F.
Figures 3.1.2-1 and.3.1.2-2 contain the limiting reactor coolant systiem pressure-temperature relationship for operation at DTr(4) and below to assure that stress levelt are low enough to preclude brittle fracture. These stress Icvels and their bases are defined in paragraph 4.3.3 of the FSAR.
As a result of fact neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated nuclear operation. The predicted maximun NDTT increase for the 40-year exposure is shcun on figure 4.3-1 of the FSAR.(4) The actual shift in NDTT will be determined periodically during plant operation by testing of irradiated vessel material samples-located in this reactor vessel. (5) The results of the irradiated sample testing will be evalu-ated and compared to the design curve (figure 4.3-2 of FSAR) being used to pre-dict the increase in transition temperature.
The design value for fast neutron (E > 1 MeV) exposure of the reactor vessel 10 nfcm2sce at 2,772 int rated power and an integrated exposure of is 3.0 x 10 3.0 x 1019 n/cm for 40 years operation.(6; The calculated maximum values are 2.4 x 1010 n/cm2see and 2.4 x 1019 n/cm2 integrated exposure for 40 years operation at 80 percent load.(4) 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 con-servatively estimated to be 1.8 x 106 thermal megawatt days, which is equiva-lent to 655 days at 2,772 !Nt core power. The projected fast neutron exposure 18 n/cm2 which is based on of the reactgr vessel for the two years is 1.7 x 10 the 1.8 x 10 thermal megawatt days and the design value for fast neutron exposure.
The actual shift in NDTT will be established periodically during plant opera-tion by testing vessel material samples which are irradiated by securing them l
near the inside wall of the vessel in tlie core area. 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 NOTT shift and the magnitude of the thereal and pressure stresses are sen-sitive to integrated reactor power and not to instantanecus power level.
Figures 3.1.2-1 and 3.1.2-2 are applicable to reactor core ther:al ratings up to 2,772 1Mt.
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RANCl{O SECO UNIT 1.
TECliNICAL SPECIFICATIONS ag D
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Surveillance Standards c
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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 effcet upon plant safety and the requirements for repairs.
~ 4.2.6 P.ccords of each inspection shall be kept to pernit evaluation and future comparison.
4.2.7 Periodic consideration will be given to incorporation of new or improved inspection techniques into the surveillance program.
4.2.8 Surveillance specimens will be withdrawn at approximately 170 cffective full power days and reine rted at the end of the first fuel cycle to allow nodifications to be incorporated into the specimen holder configuration.
Bases Irradiction surveillance provides the capability of determining radiation induced changes in the mechanical and inpact properties in the region of the-reactor vessel surrounding the core.
Test specinens of base metal, deposited
- veld metal and the heat-affected zone are installed in capsule assemblics placed inside the vessel.
In accordance with the schedules of tabic 4.2-1 specimens will be removed; and a serics of drop weight tests, Charpy impact tests and tension. tests will be conducted.
Threshold neutron flux detectors
'and naximum 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.
Prcoperational and inservice inspections e=phasi c areas of highest stress concentration and probability of failure. The arca predominantly selected for thcce 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-affcceed none.
Both testing methods will use present state-of-the-art equipment operated by highly trained personnel qualified within the requirements of the applicable codes.
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RANCHO SECO UNIT 1 I
iCHNICAL SPECIFICATIONS
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,e TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency l.-
. Control rods Rod drop times of Each refueling shutdown all full length rods 2.
Control rod Movement of cach Every two vecks movement rod 3.
Pressuri::cr code Setpoint 1 each refueling interval safety valves 4.
Main secam safety Setpaint 2 per stean generater valvec cach refueling interval 5.
Refueling system Functional Each refueling interval interlocks
, prior to handling fuel.
6.
Turbine steam stop Movement of each Monthly valves valve T.
Reactor coolant Leakage Calculated inventory weekly.
system Leakage check daily.
8.
Charcoal and high Charcoal and HEPA Each refueling interval and
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cfficiency filters filter for iodine at any tice work en and particulate filters could alter.their removal efficien-integrity.
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Frcon test on charcoal filter units.
9.
Fire pumps and Functional Monthly power supplies 10.
Reactor Building Functiona,1 Each refueling interval isolation trip 11.
Spent fuel Functional Each refueling interval cooling systes prior to fuel handling.
12.
Turbin. Overspeed Calibration Each refueling interval
-Trips
- 13..
Internals Vent Manual Actuation Each refueling interval Valves 1ba
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RANCHO SECO UNIT N0. -1 SURVEILLANCE SpECIf1 ENS
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Introduction:==
The Sacramento Municipal Utility District has been appraised by the Babcock & Wilcox Company (B&W) that the reactor vessel material surveillance holder tubes have experienced some wear at other B&W reactors.
Rancho Seco Unit No.1 is presently shutdown for an inpsection of these surveillance specimen holder tubes. This inspection has not been perfomed at this time and we do not know the corrective action which will be utilized to allow operation for the remainder of the first fuel cycle.
It is certain, however, that we will operate Ra.::ho Seco Unit No.1 for the remainder of the first fuel cycle with the surveillance specimen capsules removed.
This report is intended to demonstrate the acceptability of this operation.
At a later date, the District will provide information relative to the inspection results and correct 1ve action to be followed for the remainder of the first fuel cycle.
Safety Evaluation:
The present shutdown occurred at 169 effective full power days (EFPD's).
The first fuel cycle will last a total of 450 EFPD's.
By removing the capsules at this time and reinserting the specimens at the end of Cycle No.1 for irradiation during Cycle No. 2, the specimens will have 169 days exposure from Cycle No.1 plus 264 days exposure from Cycle No. 2 for a total of 433 EFPD's. This is very near the originally intended 450-day.
exposure from ; 'cle No.1 only.
The Babcock & Wilccx Company's Topical Report BAW-10100A, Reactor Vessel Material Surveillance Program, dated February 1975 states that the neutron flux (E > 1 MEV) at the capsule location is 2.4 times the flux at the one quarter wall thickness location of the reactor vessel belt.line. This shows that the specimens will have an integrated neutron exposure equivalent to the vessel at 1039 days when they are removed from the vessel at the end of Cycle No. 2.
This is greater than the 714-day exposure of the reactor vessel itself.
Data is also available from other B&W plants with similar vessels and vessel material surveillance programs.
This data may be used to adjust operating pressure and temperature limitations in our Technical Specifications in accordance with Appendix G of 10 CFR 50, if necessary, for Cycle No. 2 operation. Adjustments probably will not be necessary until sometime later and can be based on the results of the specimen testing program performed on Rancho Seco Unit No.1 specimens removed at the end of Cycle No. 2.
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Conclusion:==
It is concluded that operation of Rancho Seco No.1 with the surveillance specimen capsules removed for the remainder of fuel Cycle No.
9ccoM)able and in no way affects the health and safety of the public.
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s R. S. No.1 Surve111ar..; Specimens Page 2.
The requested change to the Rancho Seco Unit No.1 Technical Specifications will not result in:
1.
An increased probability of occurrence of any accident previously analyzed, or 2.
An increase in the consequences of any accident previously analyzed, or 3.
An increased probability of malfunction of any equipment important to safety previously analyzed, or 4.
An increase in the consequences of the malfunction of any equipnent important to safety previously analyzed, or 5.
The creation of the possibility of an accident of a different type than previously analyzed, or 6.
The creation of the possibility of a malfunction of different type than previously analyzed, or 7.
A reduction in the margin of safety in the basis of any technical specification.
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