ML20237D958
| ML20237D958 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/10/1987 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20237D960 | List: |
| References | |
| NUDOCS 8712280055 | |
| Download: ML20237D958 (10) | |
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UNITED STATES y
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VIRGINIA ELECTRIC AND POWER COMPANY i
DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE 1
Amendment No. 115 License No. DPR-32 3
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1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated February 23, 1987, complies with the j
standards and requirements of the Atomic Energy Act of 1954, as amend-ed (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 j
Commission; i
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 j
l defense and security or to the health and safety of the public; i
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.
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-32 is hereby amended to read as follows:
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8712280055 871210 PDR ADOCK 05000200 P
PDR I
1
(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.115, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
F0 TH NUCLEAR REGULATORY COMMISSION
+
o r ert N. Berkow, Director
-Project Directorate II Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 10, 1987
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UNITED STATES 2
NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555 t%u"k/,"?
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VIRGINIA ELECTRIC AND POWER COMPANY l
J DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 115 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 February 23, 1987, 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.
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:
i
)
(B) Technical Specifications T6e Technical Specifications contained in Appendix A, as revised through Amendment No. 115, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION H rbert 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: December 10, 1987 l
4
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO 115 FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 115 FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages TS 3.3-1 TS 3.3-1 TS 3.3-8 TS 3.3-8 TS 3.12-3 TS 3.12-3 TS 3.12-15 TS 3.12-15
^
TS Figure 3.12-8 TS Figure 3.12-8 e
TS 3.3-1 3.3 SAFETY INJECTION SYSTEM Applicability Applies to the operating status of the Safety Injection System.
Objective To define those limiting' conditions for operation that are necessary to provide sufficient borated cooling water to remove decay heat from the core in emergency situations.
Specifications A.
A reactor shall not be made critical unless the following conditions are met:
1.
The refueling water storage tank contains not less than 387,100 gallons of borated water. The boron concentration shall be at least 2000 ppm and not greater than'2200 ppa.
2.
Each accumulator system is pressurized to at least 600 psia and 3
3 contains a minimum of 975 ft and a maximum of 1025 ft of borated water with a boron concentration of at least 1950 ppm.
Amendment Nos. 115 and 115
I TS 3.3-8 Time After Shutdown Decay Heat. % of Rated power I
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.3 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.75 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.48
)
Thus, the requirement for core cooling, in case of a postulated less-of-coolant accident while in the hot shutdown condition is reduced by orders of magnitude below the requirements for handling a postulated loss-of-coolant accidert occurring during power operation. Placing and maintaining the reactor in the hot shutdown condition significantly reduces the potential consequences of a loss-of-coolant accident, allows access to some of the Safety injection System components in order to effect repairs, and minimizes the exposure to thermal cycling.
Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to hot shutdown condition is considered indicative of unforeseen problems, i.e., possibly the need of major maintenance.
In such a case the reactor is to be put into the cold shutdown condition, l
The accumulators are able to accept leakage from the Reactor Coolant System without any effect on their availability. Allowable inleakage is
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based on the volume of water that can be added to the initial amount without exceeding the volume given in Specification 3.3.A.2.
The maximum acceptable inleakage is 50 cubic feet per tank.
Amendment Nos. 115 and 115
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TS 3.12-3 B.
Power Distribution Limits 1.
At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:
F (Z) s 2.32/P x K(Z) for P > 0.5 9
F (Z) s 4.64 x K(Z) for P s 0.5 q
(s1.55[1+0.3(1-P)]forthreeloopoperation s 1,55 [1 + 0.2 (1-P)] for two loop operation where P is the fraction of rated power at which the core is operating K(Z) is the function given in TS Figure 3.12-8, and Z is the core height location of F.
q 2.
Prior to exceeding 75% power following each core loading and during each effective full power month of operation thereafter, power distribution maps using the movable detector systen shall be made to confirm that the hot channel factor limits of this specification are satisfied. -For the purpose 4
of this confirmation:
eas a.
The measurement of total peaking factor shall be increased by eight percent to account for manufacturing tolerances, measurement error and the effects of rod bow.
The measurement of enthalpy rise hot channel factor F AH shall be increased by four percent to account for measurement error. If any measured hot channel factor j
exceeds its limit specified under Specification 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under Specification i
3.12.5.1 are met. If-the hot channel factors cannot be brought to within the limits of F (Z) s 2.32 x K(Z) and q
F s 1.55 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the overpower AT and H
Overtemperature AT trip setpoints shall be similarly reduced.
Amendment tios. 115 pnd 115
TS 3.12-15 It should be noted that the enthalpy rise factors are based on integrals and are used as such in the DNB and LOCA calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in radial (x-y) power shapes throughout the core.
Thus, the radial power shape at the point of maximum heat flux is not necessarily directly related to the enthalpy rise factors. The results of the loss of coolant accident analyses are conservative with respect to the ECCS acceptance criteria as specified in 10 CFR 50.46 using the upper bound F (Z) 9 times the hot channel factor normalized operating envelope given by TS Figure 3.12-8.
When an F measurement is taken, measurement error, manufacturing tolerances, q
and the effects of rod bow must be allowed for. Five percent is the i
appropriate allevance for measurement error for a full core map (greater than or equal to 38 thimbles, including a minimum of 2 thimbles per core quandrant, 4
monitored) taken with the movable incore detector flux mapping system, three percent is the appropriate allowance for manufacturing tolerances, and five percent is appropriate allowance for rod bow. These uncertainties are statistically combined and result in a net increase of 1.08 that is applied to the measured value of F.
q In the specified limit of FAR, there is an eight percent allowance for
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uncertainties, which means that normal operation of the core is expected to result in s 1.55 [1 + 0.3 (1-P)]/1.08. The logie behind the larger H
uncertainty in this case is that (a) normal perturbations in the radial power
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shape (e.g., rod misalignment) affect FAR, in a et cases without necessarily affecting F, (b) the operator has a direct influence on F through movement q
q of rods and can limit it to the desired value; he has no direct control over 1
H, and (c) an err r in the predictions for radial power shape, which may be detected during startup physics tests and which may influence F, can q
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I Amendment Nos.115 and 115-
TS FIGURE 3.12-8 NOT CHAN'!EL FACTOR NORMALIZED OPERATINC ENVELOPE SURRY POWER STATION
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Amendment Nos. 115 and 115 A
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