ML19259A598

Application for Amend to License DPR-26 Modifying Tech Specs Re Accumulator Water Volume & New Limit for Total Nuclear Peaking Factor

ML19259A598
Person / Time
Site:	Indian Point
Issue date:	01/03/1979
From:	William Cahill CONSOLIDATED EDISON CO. OF NEW YORK, INC.
To:
Shared Package
ML100200238	List: ML100200235 ML19259A597 ML19259A598 ML19259A599
References
NUDOCS 7901080184
Download: ML19259A598 (12)

Text

UNITED STATES OF AMERICA NUCLEAR REGUIA'IORY COMISSION In the tutter of )

)

CONSOLIDATED EDISON C W PANY ) Docket No. 50-247 OF NEW YORK, 11C. )

(Indian Point Station, )

Unit tb. 2) )

APPLICATION EDR AMH DMENT

'IO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulatory Camtission (NRC), Consolidated Edison Cmpany of New York, Inc. (" Consolidated Edison"), as Irlder of Facility Operating License No. DPR-26, hereby applies for amendment of the Technical Specifications contained in Appendix A of that license.

Specifically, we request that technical specification 3.3 be modified to reflect a change in the required accurulator water volume and that technical specification 3.10 be modified to reflect the new limit for the total nuclear peaking factor (F ) g. The new values for accumulator water volume and FQ are a result of the recently cmpleted Indian Point Unit No. 2 IICS Reanalysis which was performed as required by the Camission's Order for Modification of License, dated April 27, 1978.

The specific proposed Technical Specification revisions are set forth in Attactment A to this Application. A Safety Evaluation of the proposed changes is set forth in Attactment B to this Application. This evaluation dcmonstrates that proposcxl changes do not represent a significant hazards consideration and will not cause any change in the types or an increase 7901080iF4

in amounts of effluents or any change in the autirrizal power level of the facility.

CONSOLIDe.TED EDISON COMPANY OF NW YORK, INC.

By: -

N / a' /,

William J. Cahill, Jr. &//

Vice P/esident Subscribed and sworn to before me this 34/ day of January, 1979.

LhxLt e Nothry Public t

Af1CELA RODERTI flotary Public, St:te of New York flo. 41 E593813 Q10!ified in Queens County Ccnunts.on Expices t,;27c3 39, 7g33

ATTACINEtTT A Technical Specification Page Revisions Consolidatcd Edison Carpany of New York, Inc.

Indian Point Unit tb. 2 Docket No. 50-247 January, 1979

3.3 ENGINEERED SAFETY FEATURES Applicability Applies to the operating status of the Engineered Safety Features.

Obj ec t ive To define those limiting conditions for operation that are necessary: (1) to remove decay heat from the core in emergency or normal shutdown situations, (2) to remove heat from containment in normal operating and emergency situations, (3) to remove airborne iodine from the containment atmosphere following a Design Basis Accident, (4) to minimize containment leakage to the environment subsequent to a Design Basis Accident.

Specification The following specifications apply except during low temperature physics tests.

A. Safety Inj ec t ion and Residuni Heat Removal Systems

1. The reactor shall not be made critical, except for low temperature physics tests, unless the following conditions are met:

a. The refueling water storage tank contains not less than 345,000 gallons of water with a boron concentration of at least 2000 ppm.

b. The boron injection tank contains not less than 1000 gallons of a 11 1/2% to 13% by weight (20,000 ppm to 22,500 ppm of boron) boric acid solution at a temperature of at least 145 F. Two channels of heat tracing shall be availabic for the flow path. Valves 1821 and 1831 shall be open and valves 1822A and 1822B shall be closed, except during short periods of time when they can be cycled to demonstrate their operability.

c. The four accumulators are pressurized to at least g00psig and each contains a minimum of M 716 ft and a maximum of 7 31 f t 3 of water g with a boron concentration of at least 2000 ppm. None of these four accumulators may be isolated.

d. Three safety injection pumps together with their associated piping and valves are operable.

Amendment No. 3.3-1

References (1) FSAR Section 9 (2) FSAR Section 6.2 (3) FSAR Section 6.2 (4) FSAR Section 6.3 (5) FSAR Section 14.3.5 (6) FSAR Section 1.2 (7) FSAR Section 8.2 (8) FSAR Section 9.6.1 (9) FSAR Section 14.3 (10) Indian Point Unit No. 2 " Analysis of the Emergency Core Cooling System in Accordance with the Acceptance Criteria of 10CFR50.46 and Appendix K of 10CFR50", December 1978 (11) Letter from William J. Cahill, Jr. of Consolidated Edison Company of New York, to Robert W. Reid of the Nuclear Regulatory Commission, dated July 13, 1976. Indian Point Unit No. 2 Small Break LOCA

,nalysis.

(12) Indian Point Unit No. 3 FSAR Sections 6.2 and 6.3 and the Safety Evaluation accompanying " Application for Amendment to Operating License" sworn to by Mr. William J. Cahill, Jr. on March 28, 1977.

Amendment No. 3.3-15

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability:

Applies to the limits on core fission power distributions and to the limits on control rod operations.

Objectives:

To ensure:

1. Core suberiticality af ter reactor trip,

2. Acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation and transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and

3. Limit potential reactivity insertions caused by hypothetical contrcl rod ejection.

Specifications:

3.10.1 Shutdown Reactivity The shutdown margin shall be at least as great as shown in Figure 3.10-1.

3.10.2 Power Distribution Limits 3.10.2.1 At all times, except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:

F (Z) < (2. 31/P) x K(Z) for P > .5 F (Z) < (14. 6 2 ) x K(Z) for P < .5 F < l.55 [1 + 0.2 (1-P)]

where P is the fraction of full power at which the core is operating.

K(Z) is the fraction given in Figure 3.10-2 and Z is the core height location of F q.

3.10-1 Amendment No.

g,NuclearEnthalpyRiseHot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

It should be noted that is based on an integral and is used as such in the DNB calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus the horizontal power shape at the point of N

maximum heat flux is not necessarily directly related to F .

bH An upper bound envelope of 2.31 times the normalized peaking factor axial dependence of Figure 3.10-2 has been detennincd frcxn extensive analyses considering all operating nuneuvers consistent with the technical specifications on power distribution control as given in Section 3.10.

The results of the loss of ecolant accident analyses bascd on 2.31 times the nornalizcd envelope of Figure 3.10-2 irdicate a peak clad tmperature of 2172.50F for the double-ended cold leg guillotine break with CD

=0.6,theworg)casebreak.

2200 F limit. This corresponds to a 27.5 F nargin to the When an F measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system and three percent is the appropriate allowance for manufaccuring tolerance.

In the specified limit of F there is a 8 percent allowance for uncertainties which AH n means that normal operation of the core is expected to result in F" < 1.55/1.08.

AH ~

The logic behind the larger uncertainty in this case is that (a) normal perturbations N

in the radial power shape (e.g. rod misalignment) affect F , in most cases without AH necessarily affecting F , (b) the operator has a direct influence on F through q

movement of rods, and can limit it to the desired value, he has no direct contral over {Nand (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in F by tighter axial q

control, but compensation for [ is less readily available. khen a measurement of AH F" is taken, experimental error must be allowed for and 4 percent is the appropriate All allowance for a full core map taken with the moveable incore detector flux mapping system.

Amendment No. 3.10-9

to limit the difference between the current value of Flux Dif ference (al) and a reference value which corresponds to the full power equilibrium value of Axial Of f-set (Axial Offset = AI/ fractional power). The reference value of flux difference varies with power level and burnup but expressed as axial offset it varies only with b urnup.

The technical specifications on power distribution control assure that Fq upper bound envelope of 2. 31 times Figure 3.10-2 is not exceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux dif ference is then within the limits specified by the procedure.

The target (or reference) value of flux difference is determined as follows. At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the control rod bank more than 190 steps withdrawn (i.e. normal full power operating position appropriate for the time in life, usually withdrawn farther as burnup proceeds). This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux dif ference.

Values for all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of 15 percent AI are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every month.

For this reason, the specification provides two methods for updating the target flux difference. Figure 3.10-5 shows a typical construction of the target flux difference b.md at BOL and Figure 3.10-6 shows the typical variation of the full power value with burnup.

Strict control of the flux difference (and rod position) is not as necessary during part power operation. This is because xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at part power.

Strict control of the flux difference is not possible during certain physics tests or during required, periodic, excore calibrations which require larger flux Amendment No' 3.10-11

accident for an isolated fully inserted rod will be worse if the residence time of the rod is long enough to cause significant non-uniform fuel depletion. The 4 week period is short compared with the time interval required to achieve a significant non-uniform fuel depletion The required drop time to dashpot entry is consistent with safety analysis.

REFERENCE

1. Indian Point Unit No. 2, " Analysis of the Dmrgency Core Cooling Systan in Accordance with the Acceptance Criteria of 10 CFR 50.46 and Appendix K of 10 CFR 50.

Amendment No. 3.10-16

Figure 3.10-2 IIOT CIIANNEL FACTOR NORMALIZED OPERATING ENVELOPE

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Amendment No.

ATEN3NE2TT B Safety Evaluation Consolidated Edison Ccrnpany of New York, Inc.

Indian Point Unit No. 2 Docket No. 50-247 January, 1979

Safety Evaluation The proposed changes, contained in Attachnent A of this Application, ara intended to conform the Irdian Point Unit No. 2 'Itchnical Specification requircments to those parameters used in the most recent evaluation of Dnergency Core Cooling Systen (ECCS) perfornance. This ECCS Reanalysis (Deconber,1978) was requiral by the Cmmission's April 27, 1978 Order for bbdification of License and is being filed concurrently with the present Application. The reanalysis, which was performed using the latest approved February,1978 Westinghouse'EOCS Evaluation bbdel, dcnonstrates that with lower required accumulator water volumes and a reduction in total nuclear peaking factor (Fg) to 2.31, the Indian Point Unit No. 2 DOCS meets the acceptance criteria established by 10 CFR 50.46 and Appendix K to 10 CFR 50. Accordingly, the technical specification limits must be nodified to incorporate the assumptions of the FOCS reanalysis.

The proposed changes have been reviewed by both the Station Nuclear Safety Cmmittee and the Consolidated Edison Nuclear Facilities Safety Cm mittee. Both Camittees concur that the proposed changes do not represent a significant hazards consideration and will not cause any change in the types or an increase in the arounts of effluents or any change in the authorized pomr level of 'he facility.