ML20078F567

Amends 89 & 90 to Licenses DPR-37 & DPR-32,respectively, Changing Tech Specs Re Fractional Power Limit to 0.3 Multiplier Instead of 0.2 Multiplier & Restoring Rod Insertion Limits to pre-Cycle 7 Values for Unit 1

ML20078F567
Person / Time
Site:	Surry
Issue date:	09/22/1983
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20078F572	List: ML20078F567 ML20078F589
References
NUDOCS 8310110081
Download: ML20078F567 (16)

Text

{{#Wiki_filter:' cury% UNITED STATES ,7 g g NUCLEAR REGULATORY COMMISSION \\.....lj t wAsnisoros. o. c. rosss VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 90 License No. DPR-32 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application fo'r amendment by Virginia Electric and Power Company (the licensee) dated May 2, 1983, 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 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, l

t E. The issuance of this amendment is in accordance with 10 CFR Part l 51 of the Commission's regulations and all applicable requirements have been satisfied. l l I l l 8310110001 830922 PDR ADOCK 05000 P

.. t 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: B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 90, 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 R GULATORY COMMISSION b{ (e' gk, v A. Operating Reactors B anth #1 Division of Licensin

Attachment:

Changes to the Technical 3pecifications Date of Issuance: September 22, 1983 l l I e

e utu [s 8" o,% UNITED STATES NUCLEAR REGULATORY COMMISSION s a nsHtNG TON, D. C. 20555 j s VIRGINIA'ELECTHC_ANDPOWERCOMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT N0. 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE e Amendment No. 89 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 2, 1983, 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 (t) that the activities authorized by tnis 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 i 51 of the Conunission's regulations and all applicable requirements l have been satisfied. l

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-37 is hereby anended to read as follows: B. Technical Specifications The Technical Specifications contained'in Appendix A, as revised through Amendment No. 89 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 )f ey g6, 8 Operating Reactors r nch #1 Division of Licens A,ttachment: Changes to the Technical Specifications Date of Issuance: September 22, 1983 l i { I I

4 ATTACHMENT'TO LICENSE AMENDMENTS AMENDMENT NO. 90 TO FACILITY OPERATING LICENSE N0. DPR-32 l AMENDMENT NO. 89 TO FACILITY OPERATING LICENSE N0. DPR-37 i j DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows: Remove Pages Insert Pages 2.1-3 2.1-3 2.1-4 2.1-4 2.1-6 2.1-6 + TS Figure 2.1-1 TS Figure 2.1-1 2.3-2 2.3-2 2.3-3 2.3-3 TS Figure 2.3-1 TS Figure 2.3-1 3.12-3 3.12-3 4 3.12-15 3.12-15 3.12-12 3.12-12 l TS Figure 3.12-1A TS Figure 3.12-1A r O l 1 i e l

t 1 TS 2.1-3 uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to-the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNB ratio (DNBR) during steady state operation, normal operational transients and anticipated transients, is limited to 1.30. A DNBR of 1.30 corresponds to a 95% a 95% confidence level that DNB will not occur and is probability at chosen as an appropriate margin to DNB for all operating conditions.( } The curves of TS Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent limits equal to, or more conservative than, the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which the DNB ratio is equal to 1.30 or the average enthalpy at the exit of the vessel is l equal to the saturation value. The area where clad integrity is assured is below these lines. The temperature limits are considerably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30 alone but are such that the plant conditions required to violate the limits are precluded by the self-actuated safety valves on the steam generators. The three loop operation safety limit curve has been revised to allow for heat flux peaking effects due to fuel densification ghd to apply to 100% of design flow. The effects of rod bowing are also considered in the DNBR analyses. The curves of TS Figures 2.1-2 and 2.1-3 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation), represent limits equal to, or more Amendment Nos. 90 & 89

TS 2.1-4 conservative, than the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which either the DNB ratio is equal to 1.30 or the average enthalpy at the exit of the core is equal to the saturation value. At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the DNB ratio reaches 1.30 and, thus, this arbitrary limit is conservative with respect to maintaining clad integrity. The plant conditions required 'to violate these limits are precluded by the protection system and the self-actuated safety valves on the steam generator. Upper limits of 70% power for loop stop valves open and 75% with loop stop valves closed are shown to completely bound the area where clad integrity is assured. These latter limits are arbitrary but cannot be reached due to the Permissive 8 protection system setpoint which will trip the reactor on h.igh nuclear flux when only two reactor coolant pumps are in service. Operation with natural circulation or with only one loop in service is not ' ~ allowed since the plant is not designed for continuous operation with less than two loops in service. TS Figures 2.1-1 through 2.1-3 are based on a F f 1.55, a 1.55' cosine axial H flux shape and a DNB analysis procedure including the fuel densification power I spiking (4) as part of the generic margin to accommodate rod bowing (5)(6) TS Figure 2.1-1 is also valid for the following limit of the enthalpy rise hot 1.55 (1 + 0.3 (1-P)) where P is the fraction of rated channel factor: F k H [ power. TS Figures 2.1-2 and 2.1-3 include a 0.2 rather than 0.3 part power l multiplier for the enthalpy rise hot channel factor. These hot channel factors re higher than those calculated at full power over ( the range between that of all control rod assemblies fully withdrawn to l l Amendment Nos. 90 & 89

TS 2.1-6 to this limiting: criterion. Additional peaking factors to account for local peaking due to fuel rod axial ga'ps and reduction in fuel pellet stack length have been included in the calculation of this limit. \\' g s References .1) FSAR Section 3.4

2) FSAR Section 3.3.

3) FSAR Section 14.2

4) WCAP-8012, " fuel Densification-Surry Power Station", December 1972 '

Section 4.3

5) Westinghouse (C. Eiche1dinger) to NRC (V. Stello) letter dated August 13, 1976, Serial No. NS-CE-1163

--s

6) NRC (A. Schwencer) to Vepco (W. L. Proffitt) l'tter dated July 27, 1979 e

3- = ~. ( g s Anendment Hos. 90 & 89

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3__ 2. c 540 0 20 40 60 80 100 120 Power (Percent of Rated) FIGURE 2.1-1 Reactor Core Thermal & Hydraulic Safety Limits Three Loop Operation, 100% Flow Amendment flos. 90 & 89

i TS 2.3-2 (b) High pressurizer pressure - s 2385 psig. 4 (c) Low pressurizer pressure - 2 1860 psig. (d) Overtemperature T o 1 ~ ( ) (T - T') +3 ( ~ ') ~ (01)I AT, = Indicated AT at rated thermal power, "F r T = Average coolant temperature. *F T'= 574.4*F P = Pressurizer pressure, psig P' = 2235 p'sig K = 1.135 g ) K = 0.01072 2 i' K = 0.000566 for 3-loop operation 3-Kg.= 0.951 K = 0.01012 for 2-loop operation with loop stop 2 K = 0.000554 valves open in inoperable loop 3 K = 1.026 y K = 0.01012 for 2-loop operation with loop stop 2 K = 0.000554 valves closed in inoperable loop 3 AI = q - q, where q and qb are *;he percent power in the top and b t botton. halves of the core respectively, and q

• 9 is total t

b core power in percent of rated power f(AI) = function of AI, percent of rated core power as shown in Figure 2.3-1 t = 25 seconds y i t = 3 seconds 2 i (e) Overpower AT -KI ) ~ 6 (f - T') - f (AI)] ATsAT,[K4 5 1+T53 Amendment flos. 90 & 89 f

TS 2.3-3 where AT = Indicated AT at rated thermal power, "F T = Average coolant temperature, *F T' = Average coolant temperature measured at nominal conditions and rated power, *F K = A constant = 1.089 4 K = 0 for decreasing average temperature 5 A constant, for increasing average temperature 0.02/*F K = 0 for TsT' 6 = 0.001086 for T >T' f(AI) as defined in (d) above, 10 seconds t 3 (f). Low reactor coolant loop flow - 290% of normal indicated loop flow as measured at elbow taps in each loop (g) Low reactor coolant pump motor frequency - 257.5 Hz e. (h) Reactor coolant pump under voltage - 270% of normal voltage 3. Other reactor trip settings (a) High pressurizer water level - s92% of span (b) Low-low steam generator water level - 25% of narrow range instrument span (c) Low steam generator water level - 215% of narrow range instrument span in coincidence with steam /feedwater 0 mismatch flow - 51.0x10 IN/br (d) Turbine trip (e) Safety injection - Trip settings for Safety Injection are detailed in TS Section 3.7. Amendment Nos. 90 & 89

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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 2.18/P x K(Z) for P > 0.5 F (Z) s 4.36 x K(Z) for P s 0.5 s 1.55 (1+0.3(1-P)) for three loop operation H s 1.55 (1+0.2(1-P)) for two loop operation where P is the fra.ction 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 distri-bution.. maps using the movable detector system shall be made to confirm that the hot channel factor limits of this specification are satis-fled. For the purpose of this confirmation:

a. The measurement of total peaking factor s$all be increased

( by eight percent te account for manufacturing tolerances, measure-ment error and the effects of rod bow. The measurement of enthalpy f rise hot channel factor F shall be increased by four percent to j AH i ( account for measurement error. If any measured hot channel factor exceeds its limit specified under Specification 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced i until the limits under Specification 3.12.B.1 are met. If the hot channel factors cannot be brought to within the limits of F (Z) q N 2.18 x K(Z) and F s 1.55 within 24 hours, the Overpower AT and AH Overtemperature AT trip setpoints shall be similarly reduced. Amendment Nos. 90 & 89 l

TS 3.12-12 on the maximum inserted rod wortfi in the tinlikely event of a hypothetical assembly ejection and provide for acceptable nuclear peaking factors. The limit may be determined on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and still assure compliance with the shutdown requirement. The maximum shutdown margin requirement occurs at end of core life and is based on the value used in the analysis of the hypothetical steam break accident. The rod insertion limits are based on end of core life conditions. The shutdown margin for the entire cycle length is established at 1.77% reactivity. All other accident analysis wi,th the exception of the chemical and volume control system malfunction analysis are based on 1% reactivity shutdown margin. Relative positions of control rod banks are determined by a specified control rod bank overlap. This overlap is based on the consideration of axial power shape control. The specified control rod insertion limits have been establish-ed to limit the' potential ejected rod worth in order to account for the effects of fuel densification. The various control rod assemblies (shutdown l banks, control banks A, B, C, and D) are each to be moved as a bank; that is, i with all assemblies in the bank within one step (5/8 inch) of the bank . position. Ppsition indication is provided by two methods: a digital count of actuating pulses which shows the demand position of the banks, and a linear position indicator, Linear Variable Differential Transformer, which indicates the actual assembly position. The position indication accuracy of the Linear Differential Transformer is approximately +5% of span ( 12 steps) under steady state conditions. The relative accuracy of the linear position indicator has been considered in establishing the maximum allowable deviation of a control rod assembly from its indicated group step demand position. In the event that the linear position indicator is not Amendment No. 90 & 89

TS 3.12-15 It should be noted that the enthal'py 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, 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 thimbles per core quandrant, monitored) taken with the movable inedre 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 l combined and result in a net increase of 1.08 that is applied to the measured l value of F., q Inthespecifiedlimitof(Hthere is an eight percent allowance for uncer-tainties, which means that normal operation of the core is expected to result ( E' "#E*# "" # "'" I '" in FAH ' this case is that (a) normal perturbations in the radial power shape (e.g., rod misalignment) affect H, in m st cases without necessarily affecting F, q p (b) the operator has a direct influence on F through movement of rods and can q N limit it to the desired value; he has no direct control over FAH, "" (#} "" errot in the predictions for radial power shape, which may be detected during i startup physics tests and which may influence F, can q Amendment Nos. 90 8 89

3 C C::1I 3.12-1.A f 0.0 -~ t SANE C 0.2 n 3 S g = 0.4 g g a m. w 0.g -8 IA.T D "E ^ s. w E 0.8 a 1.0 0.0 O.2 0.4 0.6 0.8 1.0 ~ MON OF EA':2:D P C U EF. FIGUIZ 3.12-1A CCIC2cL 1ANK INKIZ ION L2C:3 7tlE 3-100F 3CIMAL OFI1A::~ DIN-U:r:: 1 Amendment Nos. 90 & 89 o e ,,vy..w,--- ,,--n7 .,,,.w.,. ,,,w.._,.-ww...--..r-e.w,..-- _we.-r,. .--m v. cr--c-~~..,--w.---,---m-+.-.w.w.- .,-om , e.}}