ML15293A496
| ML15293A496 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/07/2015 |
| From: | Lawrence D Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 15-492 | |
| Download: ML15293A496 (17) | |
Text
VIRGINIA. ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 October 7, 2015 U.S. Nuclear Regulatory Commission Serial No.15-492 Attention: Document Control Desk SPS-LIC/CGL R0 Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ANNUAL SUBMITTAL OF TECHNICAL SPECIFICATIONS BASES CHANGES PURSUANT TO TECHNICAL SPECIFICATION 6.4.J Pursuant to Technical Specification 6.4.J, "Technical Specifications (TS) Bases Control Program,"
Dominion hereby submits changes to the Bases of the Surry TS implemented between October 1, 2014 and September 30, 2015.
Bases changes to the TS that were not previously submitted to the NRC as part of a License Amendment Request were reviewed and approved by the Facility Safety Review Committee. It was determined that the changes did not require a revision to the TS or operating licenses, nor did the changes involve a revision to the UFSAR or Bases that required NRC prior approval pursuant to 10CFR50.59. These changes have been incorporated into the TS Bases. A summary of these changes is provided in Attachment 1.
TS Bases changes that were submitted to the NRC for information along with the associated License Amendment Request transmittals, submitted pursuant to 10OCFR50.90, were also reviewed and approved by the Facility Safety Review Committee. These changes have been implemented with the respective License Amendments. A summary of these changes is provided in .
Current TS Bases pages reflecting the changes discussed in Attachments 1 and 2 are provided in .
If you have any questions regarding this transmittal, please contact Mrs. Candee G. Lovett at (757) 365-2178.
Very your*..
D. C. Law nrce Director Station Safety and Licensing Surry Power StationI
Serial No.15-492 Docket Nos. 50-280, 50-281 Page 2 of 2 Attachments:
- 1. Summary of TS Bases Changes Not Previously Submitted to the NRC
- 2. Summary of TS Bases Changes Associated with License Amendments
- 3. Current TS Bases Pages Commitments made in this letter: None.
cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue NE Suite 1200 Atlanta, Georgia 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7 th Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Ms. K. R. Cotton-Gross NRC Project Manager - Surry U. S. Nuclear Regulatory Commission Mail Stop 08 G-9A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager - North Anna U. S. Nuclear Regulatory Commission Mail Stop 08 G-9A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector Surry Power Station
Attachment 1 Serial No.15-492 Summary of TS Bases Changes Not Previously Submitted to the NRC Surry Power Station Units I and 2 Virginia Electric and Power Company (Dominion)
Serial No.15-492 Docket Nos. 50-280, 50-281 Attachment 1
SUMMARY
OF TS BASES CHANGES NOT PREVIOUSLY SUBMITTED TO THE NRC TS 2.3. 3.1, 3.4, 3.12. and 3.17 Bases Revisions (TS Basis Paqes TS 2.3-6, TS 2.3-9, TS 3.1-19. TS 3.4-4. TS 3.12-20. and TS 3.17-8)
Revisions were made in the TS 2.3, 3.1, 3.4, 3.12, and 3.17 Bases. The TS 3.12 Basis revision corrected an inaccurate outdated statement regarding quadrant power tilt. The TS 2.3, 3.1, 3.4, and 3.17 Bases revisions deleted a duplicate reference to a UFSAR section and corrected inaccurate references to UFSAR sections and figures.
These Bases changes were approved on December 4, 2014 and implemented on January 28, 2015.
Page 1 of 1
Attachment 2 Serial No.15-492 Summary of TS Bases Changes Associated with License Amendments Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)
Serial No.15-492 Docket Nos. 50-280, 50-281 Attachment 2
SUMMARY
OF TS BASES CHANGES ASSOCIATED WITH LICENSE AMENDMENTS Relocation of TS 4.2 and TS 4.15 Augmented Inspections (except Reactor Coolant Pump Flywheel Inspection) to Technical Requirements Manual (TS Bases Page TS 4.2-1)
These amendments revised the TSs by relocating the TS 4.2, "Augmented Inspections,"
(except the Reactor Coolant Pump flywheel inspection) and TS 4.15, "Augmented Inservice Inspection Program for High Energy Lines Outside of Containment," to the Technical Requirements Manual.
The associated Basis changes were included for information in a April 11, 2014 letter (Serial No. 14-1 58) and a March 4, 2015 letter (Serial No. 14-158A). These changers were incorporated into the Basis as part of the June 25, 2015 implementation of License Amendments 284/284 issued on April 28, 2015.
Clarification of RCS Heatup and Cooldown TS Figures (TS Bases Pages TS 3.1-7, TS 3.1-9, and TS 3.1-12)
These amendments revised TS Figures 3.1-1 and 3.1-2, Surry Units 1 and 2 Reactor Coolant System Heatup Limitations and Reactor Coolant System Cooldown Limitations, respectively, for clarification and to be fully representative of the allowable operating conditions. The clarifying revisions in the figures extended the temperature and pressure axes of the graph and added the limiting boltup temperature. The pressure/temperature curves on Figures 3.1-1 and 3.1-2 were not modified.
The associated Basis changes were included for information in a June 3, 2014 letter (Serial No.14-262). These changes were incorporated into the Basis as part of the July 22, 2015 implementation of License Amendments 285/285 issued on June 26, 2015.
Page 1 of 1
Attachment 3 Serial No.15-492 Current TS Bases Pages Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)
TS 2.3-6 12 14 The low flow reactor trip protects the core against DNB in the event of a sudden loss of power to one or more reactor coolant pumps. The undervoltage reactor trip protects against a decrease in Reactor Coolant System flow caused by a loss of voltage to the reactor coolant pump busses. The underfrequency reactor trip (opens RCP supply breakers and) protects against a decrease in Reactor Coolant System flow caused by a frequency decay on the reactor coolant pump busses.
The undervoltage and underfrequency reactor trips are expected to occur prior to the low flow trip setpoint being reached for low flow events caused by undervoltage or underfrequency, respectively. The accident analysis conservatively ignores the undervoltage and underfrequency trips and assumes reactor protection is provided by the low flow trip. The undervoltage and underfrequency reactor trips are retained as backup protection.
The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. Approximately 1125 ft3 of water corresponds to 89.12% of span. The specified setpoint allows margin for instrument error( 4 ) and transient level overshoot beyond this trip setting so that the trip function prevents the water level from reaching the safety valves.
The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the Auxiliary Feedwater System.( 4 )
The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal unit operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed.
Above 11% power, an automatic reactor trip will occur if two or more reactor coolant pumps are lost. Above 37%, an automatic reactor trip will occur if any pump is lost or de-energized. This latter trip will prevent the minimum value of the DNBR from going below the applicable design as a result of the decrease of Reactor Coolant System flow associated with the loss of a single reactor coolant pump.
Although not necessary for core protection, other reactor trips provide additional protection. The steam/feedwater flow mismatch which is coincident with a low steam generator water level is designed for and provides protection from a sudden loss of the reactor's heat sink. Upon the actuation of the safety injection circuitry, the reactor is tripped to decrease the severity of the accident condition. Upon turbine trip, at greater than 11% power, the reactor is tripped to reduce the severity of the ensuing transient.
Amendment Nos. Bases
TS 2.3-9 12-04-14 Both the Trip Setpoint and the Allowable Value must be properly established in order to adequately protect the Analytical Limit.
References (1) UFSAR Section 14.2.1 (2) UFSAR Section 14.2 (3) UFSAR Section 14.5 (4) UFSAR Section 7.2 (5) UFSAR Section 3.2.2 (6) UPSAR Section 14.2.9 Amendment Nos. Bases
TS 3.1-7 06-26-15
- 3. The pressurizer heatup and cooldown rates shall not exceed 100°F/hr. and 200°F/hr.,
respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320°FE Basis 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 III, 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 3.1-1 and 3.1-2.
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.
b) Figures 3.1-1 and 3.1-2 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.
c) Vacuum-assist fill of the Reactor Coolant System loops in COLD SHUTDOWN or REFUELING SHUTDOWN is an acceptable condition since the resulting pressure/temperature combination is located in the Acceptable Operation region of Figures 3.1-1 and 3.1-2.
- 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 generator is below 70°FE Amendment Nos. 285 and 285
TS 3.1-9 06-26-15 Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 48 Effective Full Power Years (EFPY) for Units 1 and 2. The heatup and cooldown limit curves were previously calculated using the most limiting value of RTNDT (228.4°F) which occurred at the 1/4-T, 00 azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld.
Subsequently, the reactor vessel material property basis was amended based upon new data which showed that the most limiting value of RTNDT (222.5°F) at 48 EFPY occurs at the 1/4-T, 0° azimuthal location in the Unit 2 intermediate-to-lower shell circumferential weld. The revised limiting material property (i.e., Unit 2 RTNDT of 222.5°F) justified continued use of the existing heatup and cooldown limit curves (based on the Unit 1 RTNDT of 228.4°F) to 48 EFPY for Units 1 and 2. The limiting RTNDT at the l/4-T location in the core region is greater than the RTNDT of the limiting unirradiated material.
This ensures 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 RTNDT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (B greater than 1 MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 48 EFPY for Units 1 and 2 (as well as adjustments for location of the pressure sensing instrument).
Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 48 EFPY for Units 1 and 2 prior to a scheduled refueling outage.
Amendment Nos. 285 and 285
TS 3.1-12 06-26-15 The reactor boltup temperature is defined in 10 CFR 50, Appendix G as "The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload." The reactor vessel may be bolted up at a temperature greater than the initial RTNDT of the material stressed by the boltup (e.g., the vessel flange). As noted on Figures 3.1-1 and 3.1-2, the limiting boltup temperature is 100 F. An administrative minimum boltup temperature limit greater than 10°F is imposed in station procedures to ensure the Reactor Coolant System temperatures are sufficiently high to prevent damage to the reactor vessel closure head/vessel flange during the removal or installation of reactor vessel head bolts. The limiting boltup temperature and the administrative minimum boltup temperature limit are in effect when the reactor vessel head bolts are under tension.
References (1) UFSAR, Section 4.1, Design Bases Amendment Nos. 285 and 285
TS 3.1-19 12 14 The requirement that the reactor is not to be made critical when the moderator coefficient is greater than the low power limit specified in the CORE OPERATING LIMITS REPORT has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requirement is waived during LOW POWER PHYSICS TESTS to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special operation precautions will be taken. In addition, the strong negative Doppler coefficient( 1 )(2 )
and the small integrated Delta k/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.
The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below the limiting value of RTNDT + 10°F provides increased assurance that the proper relationship between Reactor Coolant System pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps.
The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below 538°F provides added assurance that the assumptions made in the safety analyses remain bounding by maintaining the moderator temperature within the range of those analyses.
If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.
(1) UFSAR Figure 3.3-6 (2) UFSAR Table 3.3-1 Amendment Nos. Bases
TS 3.4-4 12 14 In addition to supplying water to the Containment Spray System, the refueling water storage tank is also a source of water for safety injection following an accident. This water is borated to a concentration which assures reactor shutdown by approximately 5 percent Ak/k when all control rods assemblies are inserted and when the reactor is cooled down for refueling.
References (1) UFSAR Section 4 Reactor Coolant System (2) UFSAR section 5.4 Containment Design Evaluation (3) UFSAR Section 6.3.1 Spray System (4) UFSAR Section 14.5 Loss of Coolant Accident Amendment Nos. Bases
TS 3.12-20 12 14 A 2% QUADRANT POWER TILT allows that a 5% tilt might actually be present in the core because of insensitivity of the excore detectors for disturbances near the core center such as misaligned inner control rod assembly and an error allowance. The value 1.02 (2%) was selected because the purpose of the specification is to limit, or require detection of, gross changes in core power distribution between monthly incore flux maps. In addition, it is the lowest value of quadrant power tilt that can be used for an alarm without spurious actuation.
The QPTR limit must be maintained during power operation with THERMAL POWER > 50% of RATED POWER to prevent core power distributions from exceeding the design limits.
Applicability during power operation _<50% RATED POWER or when shut down is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the FNAH and FQ(Z) LCOs still apply, but allow progressively higher peaking factors at 50%
RATED POWER or lower.
The limits of the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the UFSAR assumptions and have been analytically demonstrated to be adequate to maintain a minimum DNBR which is greater than the design limit throughout each analyzed transient. Measurement uncertainties are accounted for in the DNB design margin. Therefore, measurement values are compared directly to the surveillance limits without applying instrument uncertainty.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of temperature and pressure through instrument readout is sufficient to ensure that these parameters are restored to within their limits following load changes and other expected transient operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of RCS total flow rate, by installed flow instrumentation, is sufficient to regularly assess potential degradation and to verify operation within safety analysis assumptions. Measurement of RCS total flow rate by performance of a precision calorimetric heat balance specified in TS Table 4.1 -2A allows for the installed RCS flow instrumentation to be calibrated and verifies that the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.
Amendment Nos. Bases
TS 3.17-8 12-04-14 After an initially drained ioop is filled from the Reactor Coolant System by partially opening the loop stop valves, the loop is no longer considered to be isolated. Thus, the requirements for returning an isolated and filled loop to service are not applicable and the loop stop valves may be fully opened without restriction within two hours of completing the loop fill evolution.
The initial Reactor Coolant System level requirement has been established such that, even if the three cold leg stop valves are suddenly opened and no makeup is available, the Reactor Coolant System water level will not drop below mid-nozzle level. This ensures continued adequate suction conditions for the residual heat removal pumps.
The safety analyses assume a minimum shutdown margin as an initial condition. Violation of these limiting conditions could result in the shutdown margin being reduced to less than that assumed in the safety analyses. In addition, violation of these limiting conditions could also cause a loss of shutdown decay heat removal.
Reference (1) UFSAR Section 4.2 (2) UFSAR Section 14.2.6 Amendment Nos. Bases
TS 4.2-1 04-28-15 4.2 REACTOR COOLANT PUMP FLYWHEEL INSPECTION Applicability Applies to an inservice inspection which augments that required by ASME Section XI, Objective To provide the additional assurance necessary for the continued integrity of an important component involved in safety and plant operation.
Specification A. The Reactor Coolant Pump flywheel shall be inspected once every 20 years by a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (MT and/or PT) of the exposed surfaces defined by the volume of the disassembled flywheels.
The provisions of Specification 4.0.2 are not applicable.
Basis The inspection program for ASME Section XI of the ASME Boiler and Pressure Vessel Code limits its inspection to ASME Code Class 1, 2, and 3 components and supports. The Reactor Coolant Pump (RCP) flywheel inspection was added because there is no corresponding code requirement. The added requirement provides the inspection necessary to insure the continued integrity of the RCP flywheel.
The augmented inspection requirements for the low head safety injection piping in the valve pit, the low pressure turbine blades, and sensitized stainless steel have been relocated to the TRM.
Amendment Nos. 284 and 284