ML18153D333
| ML18153D333 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/06/1993 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML18152A448 | List: |
| References | |
| NUDOCS 9305170178 | |
| Download: ML18153D333 (9) | |
Text
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9305170178 930506 PDR ADOCK 05000281 P
PDR Technical Specifications Changes 1 I
TS 2.1-5 fully withdrawn to maximum allowable control rod assembly insertion. The control rod assembly insertion limits are covered by Specification 3.12.
Adverse power distribution factors could occur at lower power levels because additional control rod assemblies are in the core; however, the control rod assembly insertion limits dictated by TS Figures 3.12-1 A (Unit 1) and 3.12--1 B (Unit 2) ensure that the DNBR is always greater at partial power than at full power.
The Reactor Control and Protection System is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System temperature, pressure and thermal power level that would result in a DNBR less than the design DNBR limit(3) based on steady state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less than or equal to 574.4°F and a steady state nominal operating pressure of 2235* psig. For deterministic DNBR I analysis, allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power, +4°F in Reactor Coolant System average temperature and +/-30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions. The steady state nominal operating parameters and allowances for steady state errors given above are also applicable for two loop operation except that the steady state nominal operating power level is less than or equal to 60%.
For statistical DNBR analyses, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability that the minimum DNBR for the limiting rod is greater than or equal to the statistical DNBR limit. The uncertainties in the plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a statistical DNBR limit which must be met in plant safety analyses using values of input parameters without uncertainties. The statistical DNBR limit also For Unit 2 Cycle 12, Reactor Coolant System nominal operating pressure may I be reduced to 2135 psig.
Amendment Nos.
TS 2.2-2 The nominal settings of the power-operated relief valves at 2335* psig, I the reactor high pressure trip at 2385** psig and the safety valves at 2485 psig are established to assure never reaching the Reactor Coolant System pressure safety limit.
The initial hydrostatic test has been conducted at 3107 psig to assure the integrity of the Reactor Coolant System.
1 )
FSAR Section 4
- 2)
FSAR Section 4.3 2235 psig for Unit 2 Cycle 12 operation at Reactor Coolant-System nominal operating pressure of 2135 psig.
~ 231 O psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
Amendment Nos.
TS 2.3-2 (b)
High pressurizer pressure - 5. 2385* psig.
(c)
Low pressurizer pressure - ~ 1860 psig.
(d)
Overtemperature T where LlT0 = Indicated i:lT at rated thermal power, °F T = Average coolant temperature, °F T' = 574.4°F P = Pressurizer pressure, psig P' = 2235 psig K1 = 1.135 K2 = 0.01072 K3 = 0.000566 for 3-loop operation K1 = 0.951 K2 = 0.01012 for 2-loop operation with loop stop K3 = 0.000554 valves open in inoperable loop K1 = 1.026 K2 = 0.01012 for 2-loop operation with loop stop K3 = 0.000554 valves closed in inoperable loop ill = qt - qb, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt+ qb is total core power in percent of rated power f(L'.ll) = function of ill, percent of rated core power as shown in Figure 2.3-1 t 1 = 25 seconds t 2 = 3 seconds (e) Overpower 8T t3S LlT::;; 8T0 [K4 - Ks (
)T - K6 (T - T') - f(L'.ll)]
1 + t3S
- 2310 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
Amendment Nos.
TS 3.3-9 The accumulators (one for each loop) discharge into the cold leg of the reactor coolant piping when Reactor Coolant System pressure decreases below accumulator pressure, thus assuring rapid core cooling for large breaks. The line from each accumulator is provided with a motorized valve to isolate the accumulator during reactor start-up and shutdown to preclude the discharge of the contents of the accumulator when not required. These valves receive a signal to open when safety injection is initiated.
To assure that the accumulator valves satisfy the single failure criterion, they will be blocked open by de-energizing the valve motor operators when the reactor coolant pressure exceeds 1000 psig. The operating pressure of the Reactor Coolant System is 2235* psig and safety injection is initiated when this pressure drops to 600 psig.
De-energizing the motor operator when the pressure exceeds 1000 psig allows sufficient time during normal startup operation to perform the actions required to de-energize the valve.
This procedure will assure that there is an operable flow path from each accumulator to the Reactor Coolant System during power operation and that safety injection can be accomplished.
The removal of power from the valves listed in the specification will assure that the systems of which they are a part satisfy the single failure criterion.
Total system uncollected leakage is controlled to limit offsite doses resulting from system leakage after a Loss-of-Coolant Accident.
For Unit 2 Cycle 12, Reactor Coolant System nominal operating pressure may 1 be reduced to 2135 psig.
Amendment Nos.
F.
TS 3.12-11
- c.
In hot, intermediate and cold shutdown conditions, the step demand counters shall be operable and capable of determining the group demand positions to within +/-2 steps.
The rod position indicators shall be available to verify rod movement upon demand.
- 2.
If a rod position indicator channel is out of service, then:
- a.
For operation above 50% of rated power, the position of the RCC shall be checked indirectly using the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the non-indicating rod exceeding 24 steps, or
- b.
Reduce power to less than 50% of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of rated power, no special monitoring is required.
- 3.
If more than one rod position (RPI) indicator channel per group or two RPI channels per bank are inoperable during control bank motion to achieve criticality or power operations, then the requirements of Specification 3.0.1 will be followed.
PNB PARAMETERS
- 1.
The following DNB related parameters shall be maintained within their limits during power operation:
Reactor Coolant System T avg~ 578.4°F Pressurizer Pressure ;:::: 2205* psig Reactor Coolant System Total Flow Rate;:::: 273,000 gpm
- a.
The Reactor Coolant System T avg and Pressurizer Pressure shall be verified to be within their limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 2105 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
Amendment Nos.
Significant Hazards Consideration
Significant Hazards Consideration Virginia Electric and Power Company has reviewed the proposed changes against the criteria of 1 O CFR 50.92 and has concluded that the changes as proposed do not pose a significant hazards consideration. Specifically, operation of the Surry Power Station in accordance with the proposed Technical Specification changes will not:
- 1.
Involve a significant increase in the probability or consequences of an accident previously evaluated. The probability of any accident previously analyzed is not increased. Pressurizer safety valves continue to be operable and capable of performing their intended function. Operating at a reduced RCS pressure does not affect the frequency of accident initiating events. Although the Reactor Coolant System pressure is reduced for one cycle of operation, the departure from nucleate boiling ratio (DNBR) margin is maintained for accidents that challenge the DNBR limits. The loss of load analysis with the modified high pressurizer pressure reactor trip setpoint continues to meet the overpressure design limit. Therefore, there is no increase in the consequences of any accident previously evaluated which is created by operation of Surry Unit 2 at reduced pressure.
Furthermore, the proposed changes are being made to provide increased margin between operational pressure and the onset of safety valve leakage.
This reduces the potential for valve seat damage and any consequential plant transient that may result from increased RCS leakage.
- 2.
Create the possibility of a new or different kind of accident from any previously evaluated. There are no new failure modes or mechanisms associated with operating Surry Unit 2 at reduced pressure for up to one cycle. Furthermore, the proposed changes are being made to provide increased margin between operational pressure and the onset of safety valve leakage. This reduces the potential for valve seat damage and any consequential plant transient that may result from increased RCS leakage. Therefore, there are no new or different kind of accidents created by operation of Surry Unit 2 at reduced pressure.
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- 3.
Involve a significant reduction in the margin of safety. The limiting DNB analyses continue to meet the DNBR acceptance criteria at reduced pressure operation.
The applicable overpressure safety analyses acceptance criteria continue to be met with the high pressurizer pressure reactor trip setpoint reduced to less than or equal to 231 O psig. Therefore, the existing margin of safety is not reduced by operation of Surry Unit 2 at reduced pressure.
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