ML20040G315
| ML20040G315 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/02/1982 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20040G316 | List: |
| References | |
| NUDOCS 8202120155 | |
| Download: ML20040G315 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION w
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SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. OPR-32 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendr.ent by Virginia Electric and Power Company (the licensee) dated November 5, 1981, complies with the standards and requirements of the Atonic Energy Act of 1954, as anended (the Act) and the Comission'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 Comission; 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 Comission's regulations; D.
The issuance of this arendment 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 ac'cordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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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. OPR-32 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 74, 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
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Operating Reactors Qnch #1 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 2,1982
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. DPR-37 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated November 5,1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 Comission; 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 he conducted in compliance with the Comission'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 a'ccordance with 10 CFR Part 51 of the Comission'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. OPR-37 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 75, 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 aw, sa YtevenA.Varga,Cs Nerating Reactors' ranch #1 Division of Licensi g
Attachment:
Changes to the Technical Specifications Date of Issuanct:
February 2,1982
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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 74 TO FACILITY OPERATING LICENSE NO. OPR-32 AMENDMENT NO. 75 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages 3.12-15 3.12-15 3.12-16 3.12-16 I
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TS 3.12-15 It should be noted that the enthalpy rise factors are based on intergrals 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 an upper bound envelope of 2.18 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 appropriate allowance for measurement error for a full core map (238 thimbles, including a minimum of 2 detectors per core quandrant, monitored) taken with I
the movable incore detector flux mapping system, three percent is the appropriate allowance for manufacturing tolerances, and five percent is the appropriate allowance for rod bow.
These uncertainties are statistically combined and result in a net increase of 1.Q8 that is applied to the measured value of F.
q N
InthespecifiedlimitofFhH there is an eight percent allowance for uncer-tainties, which means that normal operation of the core is expected to result N
in F
$ 1.55 (1+0.2 (1-P))/1.08. The logic behind the larger uncertainty aH in this case is that (a) normal perturbations in the radial power shape N
(e.g., rod misalignment) affect FAH, in m st 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 F nd (c) an error in the predictions for radial power shape, which H,
may be detected during startup physics tests and which may influence F, can q
A!!E.'!CriErl! NOS. 74 a 75
'TS 3.12-16 be compensated for by tighter axial control. Four percent is the appropriate allowance for measurement uncertainty for F btained from a full core map AH
(>38. thimbles, including a minimum of 2 detectors per core quandrant, monitored) taken with the movable incore detector flux mapping system.
i Measurement of the hot channel factors are required as part.of startup physics tests, during each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following core loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify opera-tional anomalies which would, otherwise, affect these bases.
For normal operation, it has been determined that, provided certain condi-tions are observed, the enthalpy rise hot channel factor FhH N
II"I' "ill be met.
These conditions are as follows:
1.
Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand position. An indicated misalignment limit of 13 steps precludes a rod misalignment no greater than 15 inches with consideration of maximum instrumentation error.
2.
Control rod banks are sequenced with overlapping banks as shown in TS Figures 3.12-1A, 3.12-1B, and 3.12-2.
3.
The full length control bank insertion limits are not violated.
4.
Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits are observed. Flux difference refers to the difference AMEfiDME!iT NOS. 74 & 75