ML20023C263

Amends 52 & 20 to Licenses DPR-70 & DPR-75,respectively, Changing Partial Power Multiplier from 0.2 to 0.3

ML20023C263
Person / Time
Site:	Salem
Issue date:	05/05/1983
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20023C264	List: ML20023C263 ML20023C268 ML20023C270
References
NUDOCS 8305170119
Download: ML20023C263 (13)

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Ui4tTED STATES

[ ',$. c. f. i NUCLEAR REGULATORY COMMISSION WASHINGTO N. D. C. 20555 E

Pd3LIC SERVICE ELECTRIC AND GAS COMPANY' PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY-ATLANTIC CITY ELECTRIC COMPANY

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DOCKET NO. 50-272-SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE

. Amendment No. 52 License Nt. DPR-70 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Public Service Electric and Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic ' City Electric Company (the licensees) dated October 5,1982, 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; 3.

The facility will operate in conformisy with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that' the activities authorized by this cuendment can be conducted without en: angering 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

4 have been satisfied.

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• :crdingly, the license.is amended by enanges to the Technical S ecifications as indicated in the attachment to this license amendment, and paragraph 2.Ci(2) of Facili:y 0:erating License No. DPR-70 ~is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Ac6endices A and B, as revised through Amendment No. 52, are j

hereby incorporated in the license. The licensee shall Operate the facility in accordance with the Technical Specifications.

2.

This 'icense amencment is effective as of :ne : ate Of its 4ssdance.

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Ocerating o.eac:ces 3 ranch 41 Jivisi:nOfLicensIng tachment:

"nanges :c :ne Tecnnical S.:ecificati rs 23:e of :ssuance: May 5,1983 e

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ATTACHMENT TO LICEftSE AMEf!DMENT NO. 52

' FACILITY OPERATING LICEftSE t;0. DPR-7(>

DOCKET NO. 50-272 Revise Appendix A as follows:

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' Remove Pages Insert Pages 3/4 2-9 3/4-2-9 B 2-2 B 2 2-2 2-2 O

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NUCLEAR ENkHAL?Y HOT CHANNEL FACTOR - FN

_LH LIMITING CONDITION FOR OPERATION 3rH shall be limited by the following relationship:

3.2.3 A

g EjH1 55 [1.0 + 0 ' 3 (1-P)] [1-RBP (BU)]

1 THERMAL POWER

, and where: P = RATED THE. VAL POWER

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REP (BU)= Rod Sow Penalty as a function of region average burnup as shown in.:igure 3.2-3, where a region' is defined as those assemblies with the same loading date (reloads) or enrich =en: (first core).

AFFLICABILITY: MODE 1 ACTION:

With N exceedinc its limit:

r,H a.

Reduce THEFF.AL POWER to less than 50% of RATED TdEPSAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Pwer Ran;e Neutr:n Flux-High Trip Setpoints to 1 55% of RATED TdE:yAL POWER within the next.4

hours, N

b.

Dem:nstrate thru in-core map;in; ~ hat F.; is wi-hin its limit within 24 hcurs after exceedin; the limit or recuce THE:FAL - - -

POWER to less than 5% of P,ATED THEFS.AL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and I.dentify and correct -the cause of tbe out of limit condition n

c.

prior to increasing TdEPyAL POWER above the reduced limit required by a. 'er b. above; subsequent POWER OPERATION may proceed provided that F

is demonstrated through in-core maoping: to be within ithlimit at a nominal 50% of RATED

. THERMAL-POWER prior to exceeding this THEPyAL POWER, at a.

nominal 75% of RATED TdEFFAL POWER prict to exceeding this T-iEFFAL power and within 24'heurs after attaining 95% or greater RATED THEPS.AL POWER.

.e 52 SALEu,- UNIT 1 3/4 2-9 knendment Nc. i, FJ E~

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SAFETY LIMITS 4

BASES N

The curves are based on an enthalpy hot channel factor, F and a reference cosine with a peak of 1.55 for axial power shah!., of,1.55 An g

allowance is included for an increase in F at reduced power based on aH the expression:

F g = 1.55 D+ 0.3 (1-p)]

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions a,re higher than those calculated for the rance of all control rods fully withdrawn to the maximum allowable contrcl rod insertion assuming the axial p:wer imbalance is within the limits of the f1 ( I), function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect. on the Overtemperature AT trips will reduce the setpoints to provide protection censistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE I

s The restriction of this Safety Limit protects the integrity of the Reactor Cociant System from overpressurization and thereby prevents the release cf radienuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III cf :ne ASME Code for Nuclear pcwer Plant which permi:s a maximum transien; pressure cf 110% (2735 psig) cf cesign pressure.

The Reactor Coolant System piping and fittings are designed to ANSI 5 31.1 1955 Edition wnile the valves. are designed to ANSI B 16.5,. MSS-SP-66-1964, or 8

ASME Section III-1958, which permit maximum transient oressures of up to 120t '(2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

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The entire Reactor Coolant System is hydratested at 3107 psig,125%

of design pressure, to demonstrate integrity prior to initial operation.

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SALEM - UNIT 1 B 2-2 Amendment No. 52 4'

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" REACTOR CORE SAFETY LIMIT - FOUR I.00PS IN OPERATION

.y Anendment No. 52 SALEM - UNI 1

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PUBLIC. SERVICE' ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA. POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. ' 50-311 SALEM NUCLEAR GENERhTING STATION, UNIT NO. 2 AMENDMENT'TO FACILITY OPERATING LICENSE Amendment No. 20' License No. DPR-75 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Public Service Electric and Gas Coinpany, Philadelphia Electric Company, Delmarva Power j

and Light Company and Atlantic City Electric Company (the licensees) dated October 5,1982, complies with the standards l

and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth l

in 10 CFR Chapter I; i

l B.

The facility will operate in conformity with the application, l

the provisions of the Act, and the rules and regulaticas of the Commission; l

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without encangering the health and safety of the public, and (ii) that such activities will be I

conducted ~in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common l

defense and security or to the health and safety of' the public;

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and E.

The issuance of this amendment is in 'accordance with 10 CFR Part 51 of the Commission'.s regulations and all applicable requirements d'!

- have - been'satts'fied.

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'ccordingly.the license is acended by enan;es to the Technicai

• Scecifications as indicated in the attacnmen to this license amendmen, and paragraph 2.C2(2) of Facility Operating License No. DPR.75~is hereby amended to read as follows:

(2) Technical Scecifications The Technical Specifications contained in Apdendices A and B, as revised through Amendment No. 20, are hereby incor; crated in the license. The licensee shall cperate the facility in acccrdance with -he Technical Specifications.

2.

This 'icense acenc er.: is effective as cf -he date cf i s 'ssu'ance.

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ne Tecnnical 5;eci#ica-i:ns Date of Issuance: May 5,1983 l

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ATTACHMENT TO LICENSE AMENDMEtiT 110. 20

' FACILITY OPERATING LICENSE NO. DPR-7 DOCKET NO. 50-311 Revise Appendix A as-follows:

Remove Pages Insert Pages 3/4 2-9 3/4 2-9 3/4 2-11 3/4 2-11 B 2-1 B 2-1 2 2-2 d

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POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND P, LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R, R, shall be maintained within the region of allowable operation shown on1F,igure 3.2-3 for 4 loop operation.

Where:

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Ri = 1.49 [1. 0 + s. 3 (1. 0 - P)]

R) b.

R2 * [1-RBP(BU)] '

THERMAL POWER c.

P RATED THERMAL POWER, and Fh = Measured values of Fh obtained by using the movable incore d.

detectors to obtain a power distribution map.

The measured values of F shall be used to calculate R since Figure 3.2-3 g

includes measurement uncertainties of 3.5% for flow and 4% for incoremeasurementofFh.

RSP (BU) = Rod Bow Penalty as a function of region average burnup as e.

shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrich-ment (first core).

APPLICABILITY:

MODE 1.

ACTION:

With the combination of RCS total flow rate and R), R2 utside the region of acceptable operation shown on Figure 3.2-3:

a.

Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

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Either restore the combination of RCS total flow rate and R),

R to within the above limits, or 2

J 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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l SALEM - UNIT 2

- 3/42-9 Amendment No. 20

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0.90 0.95 1.00 1.05 1.10 Ri=FyH/1.49 [1.0' O.3 (1.0 - P) ]

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SALEM - URIT 2 3/4 2-11 Amendraent No. 20

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1 2.1 SAFETY LIMITS' e

BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation whfch would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate ' boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB thr'ough the W-3 correlation.

The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cau,se DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.

This value corresponds to a 95 percent probability at a 95 percent confidence level l

that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

N a reference cosine with a peag of 1.55 for axial power shape.

The curves are based on an enthalpy hot channel factor, F allowance is i

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included for an increase in Fg Fh = 1.55 [1 + 0.3 (1-P)]

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control I

rod insertion assuming the axial power imbalance is within the limits of the When the axial power f)(delta I) function of the Overtemperature trip.

Amendment No. 20 SALEM - UNIT 2 B 2-1 v.__

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