ML20137W911

Amend 45 to License NPF-12,modifying Tech Specs to Reflect 1.9% Reduction in Thermal Design Flow

ML20137W911
Person / Time
Site:	Summer
Issue date:	09/25/1985
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20137W916	List: ML20137W911 ML20137W918
References
NUDOCS 8510040418
Download: ML20137W911 (12)

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{{#Wiki_filter:. [o %yjo, UNITED STATES y 3m j NUCLEAR REGULATORY COMMISSION r, t WASHINGTON, D. C. 20555 \\ } SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING l! CENSE Amendment No.45 License No. NPF-12 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment to the Virgil C. Sumer Fuclear Station. Unit No.1 (the facility) Facility Operating license No. NPF-12 filed by the South Carolina Electric & Gas Company acting for itself and South Carolina Public Service Authority (the licensees), dated March 6,1985, and supplemented April 30 and August 9,1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission; C. There is reasonable assurance: (1) 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 set forth in 10 CFR Chapter I; D. The issuance of this license amendment will not be inimical to the comon 2 defense and security or to the health and safety of the public; E. The issuance of this license amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is hereby amended by page changes to tha Technical Specifications as indicated in the attachments to this license anendment and paragraph 2.C(2) of Facility Operating license No. NPF-12 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 45, are hereby incorporated into this Itcense. South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. l 0510040410 050925 ADOCK 05 % y 5 l DR

. 3. This ifcense amendment is effective seven days after its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION ff lk % \\, Elinor G. Adensam, Chief licensing Branch No. 4 Division of Licensing l

Enclosure:

l Technical Specification Changes l l Date cf Issuance: september 25, 1985 I l l l l l l

4 ATTACHMENT TO LICENSE AMENDMENT NO. 45 FACILITY OPERATING LICENSE NO. NPF-12 i i DOCKET NO. 50-395 I Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf page is also provided to maintain document completeness. i Amended Overleaf ?.*Lt E*S* 2-2 2-5 3/4 2-8 3/4 2-7 3/4 2-9 l 3/4 2-10 3/4 2-11 i B 3/4 2-4 B 3/4 2-5 i o i t f E i i i i i }. t 1 i

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• When operating in the reduced RTP region of Technical Specification 3.2.3 l

(Figure 3.2-3), the restricted power level must be considered 100% RTP for this ffgure. Figure 2.1-1 Reactor Core Safety Limit - Three Loops in Operation SUMMER - UNIT 1 2-2 Amendment No. 45

e. I" TABLE 2.2-1 = REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS L e Total Functional Unit Allowance (TA) Z_ S Trip Setpoint Allowable Value 1. Manual Reactor Trip Not Applicable NA NA NA NA ( 2. Power Range, Neutron Flux 7.5 4.56 0 1109% of RTP 1111.2% of RTP High Setpoint Low Setpoint 8.3 4.56 0 125% of RTP $27.2% of RTP I 3. Power Range, Neutron Flux

1. 6 0.5 0

15% of RTP with 16.3*.' of RTP with High Positive Rate, a time constant a time constant 12 seconds 12 seconds i 4. Power Range, Neutron Flux

1. 6 0.5 0

15% of RTP with 16.3% of RTP with High Negative Rate a time constant a time constant u J. 12 seconds 12 seconds ( 5. Intermediate Range, 17.0 8.4 0 125% of RTP 131% of RTP 1 Neutron Flux 6. Source Range, Neutron Flux 17.0 10.0 0 1105 cps 11.4 x 105 cps l 7. Overtemperature AT 7.1 2.94 -1. 8 See note 1 See note 2 8. Overpower AT 4.5 1.4

1. 2 See note 3 See note 4 9.

Pressurizer Pressure-Low 3.1 0.71 1.5 11870 psig 11859 psig l 10. Pressurizer Pressure-High 3.1 0.71 1.5 12380 psig 12391 psig l 's 11. Pressurizer Water Level-High 5.0 2.18 1.5 192% of instrument 193.8% of instrument l 5 span span 5 12. Loss of Flow 2.5 1.0 1.5 >90% cf loop >89.2% of loop x? design flow

• 3esign flow
• l Loop design flow = %,200 gpa RTP = RATED THERMAL POWER l

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I. L _ _.4... i i l i. I 1 l 1 i i ._l. g e e i I O.0 1 0 2 4 6 8 10 12 2 BOTTOM TOP OF CORE HEIGHT, FT. OF L l FUEL FUEL i FIGURE 3.2 2 1 1 l t K(Z). NORMALIZED Fo(Z) AS A FUNCTION OF CORE HEIGHT 1 i I I Summet UNIT 1 3/427 i

PCWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combinati'on of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown j on Figure 3.2-3 for 3 loop operation. Where: N ^N a* R = 1.49 [1.0 + 0.2 (1.0 - P)] ' b. P _ THERMAL POWER RATED THERMAL POWER Fh=MeasuredvaluesofF"powerdistributionmap N c. obtained by using the movable incore detectorstgobtaink The measured values of F shall be used to calculate R since Figure 3.2-3 includes mekYurement ungertainties of 3.5% for flow and 4% for incore measurement of F3g. APPLICABILITY: MODE 1. ACTION: With the combination of RCS total flow rate and R outside the region of l i acceptabic operation shown on Figure 3.2-3: a. Within 2 hours either: / L Restore the combination of RCS total flow rate and R to within the above limits, or r 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within l the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours. Identify and correct the cause of the out-of-limit condition prior c. to increasing THCRMAL POWER above the reduced THERHAL POWER limit required by ACTION items a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated l RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels: SU MER - UNIT 1 3/4 2-8 Amendment No. 45

POWER DISTRIBUTION LIMITS ACTION: (Continued) 4 ] 1. A nominal 50% of RATED THERMAL POWER, j 2. A nominal 75% of RATED THERMAL POWER, and i 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS j 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. j 4.2.3.2 The combination of indicated RCS total flow rate and R shall be l ] determined to be within the region of acceptable operation of Figure 3.2-3: a. Prior to operation above 75% of RATED THERMAL POWER after each fuel } loading, and b. At least once per 31 Effective Full Power Days. i 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the 4 I region of acceptable operation of Figure 3.2-3 at least once per 12 hours when I the most recently obtained value of R obtained per Specification 4.2.3.2, is assumed to exist. 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL t CALIBRATION at least once per 18 months. l 4.2.3.5 The RCS total flow rate shall be determined by measurement at least 4 i once per 18. months. i, J 1 1 ( i i 1, i T i I SUMMER - UNIT 1 3/4 2-9 Amendment No. 45

MEASUREMENT UNCERTAINTIES OF 3.5% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF F N ARE INCLUDED IN THIS FIGURE 38 36 ACCEPTABLE UNACCEPTABLE OPERATION REGION OPERATION REGION 34 b 2 2 U o 'e 5 9 q 32 m 3 9 u. k>o 30 (1.00; 29.87) ' [ 98% RTP (1.00, 29.57) o 96% RTP (1.00, 29.27) I E 94% RTP 97, > SEE NOTE

• 20 NI [q 92% RTP II*

90% RTP 11.00,28.38) j 28 26 24 0.90 0.95 1.00 1.05 1.10 R = Fj/1.49 [1.0 + 0.2(1.0 P] FIGURE 3.2 3 RCS TOTAL FLOW RATE VS. R THREE LOOP OPERATION NOTE: When operating in this reg 6on. the restricted power levele shell be considered to be 1C0% of rated thermal power (RTPI for Figure 2.11. SUMMER - UNIT 1 3/4 2-10 Amendment No. 45

This page deleted. .e SUnt4ER - UNIT 1 3/4 2-11 Amendment No. 45

POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained. d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. Fhwillbemaintainedwithinitslimitsprovidedconditionsa.through

d. above are maintained.

As noted on Figure 3.2-3, RCS flow rate and l Fh may be " traded off" against one another (f.e., a low measured RCS flow rateisacceptableifthemeasuredFhisalsolow)toensurethatthe calculated DNBR will not be below the design DN8R value. The relaxation of F as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. l R,ascalculatedin3.2.3andusedinFigure3.2.3,accountsforFh less than or equal to 1.49. This value is used in the various accident analyses where Fh influences parameters other than DN8R, e.g., peak clad temperature and thus is the maximun "as measured" value allowed. Fuel rod bowing reduces the value of DN8 ratio. Credit is available to offset this reduction in the generic margin. The generic design margins, totaling 9.1% DN8R, completely offset any rod bow penalties.* This margin includes the following:

1) Des'ign limit DNBR of 1.30 vs. 1.28

2) Grid Spacing (K ) of 0.046 vs. 0.059

3) Thermal Diffusi8n Coefficient of 0.038 vs. 0.059

4) DN8R Multiplier of 0.86 vs. 0.88

5) Pitch reduction The applicable value of rod bow penalties is referenced in the,FSAR.

When an F measurement is taken, an allowance for both experimental error q and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. The radial peaking factor F,y(Z) is measured periodically to provide assurance that the hot channel factor, F (Z), remains within its Itait. The 0

• The generic margins also offset the penalty associated with the thermal design flow reduction included in Amendment 45 to the Technical Specifications.

SupMER - UNIT 1 B 3/4 2-4 Amendment No. 45

i POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) F limit for Rated Thermal Power (FRTP) as provided in the Radial Peaking xy x Factor Limit Report per specification 6.9.1.11 was determined from expected power control maneuvers over the full range of burnup conditions in the core. I When RCS flow rate and F are measured, no additional allowances are g necessary p*for to comparison with the limits of Figure 3.2-3. Measurement l errors of 3.5% for RCS total flow rate and 4% for FN have been allowed for in determining the limits of Figure 3.2-3. l . The 12 hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3. 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted. The limit of 1.02 was selected to provide an allowance for q the uncertainty associated with the indicated power tilt. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned control rod. In the event such action does not correct thIe tilt, the margin for uncertainty on F is reinstated by q reducing th~e maximum allowed power by 3 percent for each percent of tilt in excess of 1.0. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of 4 symme.tric thimbels. These locations are C-8, E-5, E-11, H-3, H-13, L-5, 2-11, N-8. 3/4.2.5 DN8 PARAMETERS The limits on 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 initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimus DNBR of 1.30 throughout each analyzed transient. The 12 hcur periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. SUPMER - UNIT 1 B 3/4 2-5 Amendment No. 45 ._.}}