ML20036B394
| ML20036B394 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/14/1993 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20036B395 | List: |
| References | |
| NUDOCS 9305190414 | |
| Download: ML20036B394 (8) | |
Text
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NUCLEAR REGULATORY COMMISSION W ASHINGT ON. D. C. 20555 1.,...../
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No. DPR-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated May 6, 1993, complies 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 Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common 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.
A l
9305190414 930514 PDR ADDCK 05000281 l
P PUR
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-37 is hereby amended to read as follows:
(B)
Technical Specifications The Technical Specifications contaired in Appendix A, as revised through Amendment No. 177, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
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3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION g
4 Her ert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 14, 1993 I
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 177 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Revise Appendix A as follows:
Remove Paaes Insert Paaes 2.1-5 2.1-5 2.2-2 2.2-2 2.3-2 2.3-2 3.3-9 3.3-9 3.12-11 3.12-11 1
TS 2.1-5 fully withdrawn to maximum a!!owable 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 3.12-1B (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 l
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 j
l For Unit 2 Cycle 12, Reactor Coolant System nominal operating pressure may be reduced to 2135 psig.
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Amendment No.177 i
_A
TS 2.2-2 The nominal settings of the power-operated relief valves at 2335* psig, 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 1
2235 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
s 2310 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
Amendment No.177
TS 2.3-2 (b)
High pressurizer pressure - s 2385* psig.
l (c)
Low pressurizer pressure - 21860 psig.
(d)
Overtemperature T 1+ t1 S AT s AT [K - K (1 + T2 S) (T-T) + 3K (P - P') - f(AI)]
o i
2 where AT = Indicated AT at rated thermal power, *F o
T = Average coolant temperature, 'F T = 574.4 F P = Pressurizer pressure, psig P' = 2235 psig K = 1.135 3
K = 0.01072 2
K = 0.000566 for 3-loop operation 3
K = 0.951 3
K = 0.01012 for 2-loop operation with loop stop 2
K = 0.000554 valves open in inoperable loop 3
K = 1.026 3
K = 0.01012 for 2-loop operation with loop stop 2
K = 0.000554 valves closed in inoperable loop 3
Al = qt " 9, where qt and qb are the percent power in the top and bottom halves of the j
b core respectively, and qt + A is total core power in percent of rated power b
f(AI) = function of AI, percent of rated core power as shown in Figure 2.3-1 t3 = 25 seconds 2 = 3 seconds T
(e) Overpower AT I
T3S AT s AT (K - K (1 *,+3 )T-Ks (T-T)- f(AI)]
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s 2310 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
Amendment No.177
TS 3.3-9 The accumulators (one for each loop) oischarge into the cold leg of the reactor coolant piping when Reactor Coolant Systern 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 l
coolant pressure exceeds 1000 psig. The operating pressure of the Reactor l
Coolant System is 2235* psig and safety injection is initiated when this I 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 enterion.
Total system uncollected leakage is contro!!ed to limit offsite doses resulting from system leakage after a Loss-of-Coolant Accident.
l For Unit 2 Cycle 12, Reactor Coolan't System nominal operating pressure may be reduced to 2135 psig.
Amendment No 177
a TS 3.12-1 t In hot, intermediate and cold shutdown conditions, the step c.
demand counters shall be operable and capable of d6Crmining 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 channelis out of service, then:
For operation above 50% of rated power, the position of the a.
RCC shall be checked indirectly using the movable incore l
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.
i 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.
1 F.
DNB PARAMETERS 1.
The following DNB related parameters shall be maintained within their limits during power operation:
Reactor Coolant System Tavg s 578.4*F Pressurizer Pressure 2 2205' psig l
Reactor Coolant System Total Flow Rate 2 273,000 gpm The Reactor Coolant System Tavg and Pressurizer a.
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 />.
2 2105 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.
Amendment No.177
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