ML20248C648

Application for Amends to Licenses DPR-24 & DPR-27, Consisting of Tech Spec Change Request 126,revising Heatup & Cooldown Limit Curves

ML20248C648
Person / Time
Site:	Point Beach
Issue date:	08/03/1989
From:	Britt R WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20248C652	List: ML20248C648 ML20248C670
References
CON-NRC-89-092, CON-NRC-89-92 VPNPD-89-426, NUDOCS 8908100111
Download: ML20248C648 (5)

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' 231 W. MICHIGAN,P.O. BOX 2046, MILWAUKEE Wl53201 (414) 221-2345 VPNPD-89-426 10 CFR 50.59 NRC-89-092

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August 3, 1989 U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail Station P1-137 Washington, D. C. 20555 Gentlemen:

DOCKETS 50-266 AND 50-301 TECHNICAL SPECIFICATION CHANGE REQUEST 126 REVISION TO HEATUP AND COOLDOWN LIMIT CURVES POINT EEACH NUCLEAR PLANT, UNITS 1 AND 2 In accordance witn the requirements of 10 CFR 50.59 and 50.90, Wisconsin Electric Power Company (Licensee) requests amendments to Facility. Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2, respectively. The amendments will implement heatup and cooldown limit curves in the Technical Specifications applicable to both units effective-to 18.1 EFPY and related changes.

We have enclosed herewith revised Technical Specifications pages with proposed modifications identified by a margin bar in the right-hand margin. A detailed discussion of these changes follows.

Technical Specification Figures 15.3.1--l and 15.3.1-2, Heatup and Cooldown Limitation Curves, have been revised to indicate applicability to both units. Technical Specification Figures 15.3.1-3 and 15.3.1-4, applicable to Unit 2'only, are deleted.

Present Technical Specification Figures 15.3.1-1 and 15.3.1-3, Heatup Limitation Curves, for Point Beach Units 1 and 2 respectively are identical,'as.are Figures 15.3.1-2 and 15.3.1-4, Cooldown Limitations, Therefore, combining these figures such that one set of limit curves is applicable to both units is appropriate and simplifying. Technical Specifiaacions 15.3.1.B.1, 15.3.1.B.4, and 15.3.1.F.3 have been changed to reflect the use ot onc set of limit curves for both units.

Technical Specification 15.3.1.B, Bases, have also been changed to delete the references to Figures 15.3.1-3 and 15.3.1-4.

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'NRC Document Control Desk August 3, 1989 Page 2 The heatup and cooldown limitation curves have been revised to be applicable through 18.1 effective full power years (EFPY),

or approximately January 1, 1995. Exposure of reactor vessel materials to neutron radiation throughout operating life results in a change in the reference temperature (RTN of the materials due to neutron embrittlement. This change k$)

reference temperature is calculated periodically and appropriate limits revised. To ensure the heatup and cooldown limitation curves would be applicable to both Point Beach Units 1 and 2 through 18.1 EFPY, the shift in the curves was calculated using the most limiting combination of material and fast neutron fluence. The lLaiting reactor vessel material compositions for each unit were provided in our pressurized thermal shock (PTS) submittal dated January 30, 1986, as modified March 14, 1986, and are repeated in Table 1. The controlling weld and weld materials for each unit were accepted by the NRC in a Safety Evaluation Report (SER) dated July 24, 1986. As can be seen from the information in Table 1, the Point Beach Unit 2 weld chemistry is more restrictive than that of Unit 1.

The neutron fluences applicable to these materials through 18.1 EFPY are listed in Table 2, Item 2. These fluences were derived from information contained in WCAP 10638, " Adjoint Flux Program for Point Beach Units 1 and 2", which was transmitted to you with our March 14, 1986 PTS submittal. In calculating the heatup and cooldown curvea, no credit is taken at this time for the flux reduction measures now being implemented.

The Unit 2 values for fluence and weld chemistry were then utilized with the methodology contained in Section C.1.1 of Regulatory Guide 1.99, Revision 2, "REdiation Embrittlement of Reactor Vessel Materials", to calculate the adjusted reference temperature (ART) curves. The exponential correlation in Equation 3 of Regulatory Guide 1.39, Revision 2, was utilized to calculate neutron fluence in the vessel wall. Significant parameters determined using this methodology, and Unit 2 weld I ART information used to adjust the curves to 18.1 EFPY, are summarized in Table 3, Item 6. Technical Specification 15.3.1.B has been modified to indicate the use of this methodology.  ;

The heatup and cooldown curves presented in Figures 15.3.1.1 and 15.3.1-2 were calculated, utilizing the ART's from Table 3, at the one-quarter and three-quarters depths in the vessel wall in the methodology given in Section 6 of WCAP-8738, "Heatup and Cooldown Limitation Curves for... Point Beach Nuclear Plant, Unit No. 2." Additional margins for possible instrumentation l

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inaccuracies have not been added to these curves, since neither 10.CFR 50,: Appendix G nor ASME Code Section III - Appendix G g requires;the inclusion of margins.for. instrumentation uncertainties,.in addition ^to other conservatism in the methods for calculating pressure-temperature limits. . Technical Specification 15.3.1.B, Basis, has teen revised to reflect our present practice of not including additional margins.in-the-heatup and cooldown curves for instrumentation uncertainties.

The final change requested is to correct a reference in Specification 15.3.1.B, Basis, from 10 CFR 50, Appendix G, Section IV.A.2.C to Section IV.A.3. Section IV.A.2.C no longer exists.

Point Beach Technical Specification 15.3.1.B.4 requires us to submit revised heatup and cooldown limit figures to the Commission at least sixty (60) days prior to the calculated exposure of the reactor vessel exceeding the exposure to which the figures-apply. This would necessitate a submittal in approximately November 1989. We are submitting this change in advance of this-date in accordance with the commitment made in our letter dated November 29, 1988, " Response to NRC Generic Letter 88-11 Analysis of Reactor Vessel Beltline Materials, Point Beach Nuclear Plant, Units 1 and 2."

We implemented a Low Low Leakage Pattern (L4P) core with hafnium inserts in the guide tubes of peripheral assemblies in April 1989 for Unit 1 to reduce fluence to the critical reactor vessel welds. A similar core modification will be implemented on Unit 2 in October 1989. With L4P reload cores and hafnium inserts installed, we expect to achieve the following flux reduction factors (FRF) relative to L3P cores at the critical weld locations:

FRF Achieved 0 Degrees 15 Degrees Unit 1 2.0 1.8-(Lower Shell) 1.4 (Intermediate Shell)

Unit 2 2.0 We are also using excore dosimetry to monitor fluence to the vessel. This will allow us to accurately assess the effectiveness of the flux reduction measures identified above for possible future adjustment of the heatup and cooldown limits.

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NRC Document Control Desk August 3, 1989 Page 4 Additionally, in accordance with NRC's request in thc: ooint Beach Nuclear Plant PTS Safety Evaluation dated July 24, 1986, we are providing an updated assessment of our projected RT PTS values. Table 4 provides these RT projections for the critical welds using the fluence v[I5es given in Table 2 and using the trend formulas from both 10 CFR 50.61 and Regulatory Guide 1.99, Revision 2. We conclude that, with the flux reduction measures we are presently implementing, the Point Beach reactor vessels will remain below the PTS screening I

criteria through present license duration.

We have evaluated these proposed amendments, in accordance with the requirements of 10 CFR 50.91(a), against the standards of 10 CFR 50.92 and have determined that these modifications will not result in a significant hazards consideration. A proposed amendment will not involve a significant hazards consideration if it does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

With respect to the proposed changes directly related to tho heatup and cooldown limit curves, the proposed heatup and cooldown curves were derived from heatup and cooldown curves which were identical between Unit 1 and Unit 2. The change in the curves was calculated ucing the most limiting weld and fluence information from either unit as input to acceptable methodology of Regulatory Guide 1.99, Revision 2. The consequences or probability of a previously evaluated accident will, therefore, not significantly be increased or a margin to safety reduced.

{ The underlying purpose of these curves remains unchanged. This l is te define an acceptable operating range of pressures and temperatures to prctect the reactor vessels against non-ductile failure. Therefore a new or different kind of accident cannot be created.

The remaining change requested involved correcting a reference i in the bases to the appropriate section of 10 CFR 50, Appendix G. This is a purely administrative change and, therefore, does not involve a significant hazards consideration. This determination is supported by the Commission's example of amendments not likely to involve a significant hazards consideration published at 48 FR 14864.

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-'NRC Document Control Desk August.3, 1989 Page 5-Please contact'us should you have any questions regarding this submittal.

Vedy'truly'yours,

,4( .

R. W. Britt l

Chairman of the Board &

Chief Executive Officer Attachments Copies.to NRC Regional Administrator, Region III

.NRC Resident Inspector R. S. Cullen, PSCW Subscribed and sworn to before me this FM day of Auai,d 1989. '

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Notary .Public, State %f' Wisconsin .

My Commission expires 5-2 7- 90 k---~~.----- - - - - - _ - - - - - - _ - - - . _ - - _ - - _ . . . - - _ _ _ _ _ _