ML20045B656

Change Request 160 to Licenses DPR-24 & DPR-27,revising TS 15.3.1.G.3 to Reduce RCS Raw Measured Total Flow Rate Limit for Unit 2 by 2,600 Gpm & TS 15.2.3 Re Limiting Safety Sys Settings,Protective Instrumentation

ML20045B656
Person / Time
Site:	Point Beach
Issue date:	06/11/1993
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20045B657	List: ML20045B656 ML20045B666
References
CON-NRC-93-072, CON-NRC-93-72 VPNPD-93-113, NUDOCS 9306180232
Download: ML20045B656 (8)

Text

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Wisconsin fElectnc -

POWER COMPANY 231 W MicNgan Po Bor 2046. Mdwovkoa WI 53201 (414)2202345

_VPNPD-93-113 10 CFR 50.4 NRC 07 2 10 CFR 50.90 June 11, 1993 Document Control Deck U.S. NUCLEAR REGULATORY COMMISSION Mail Station P1-137 '

Washington, DC 20555 Gentlemen:

DOCKETS 50-266 AND 50-301 i TECHNICAL SPECIFICATIONS CHANGE REOUEST 160 -

MODIFICATION TO TS 15.3.1.G.3-REACTOR COOLANT  !

SYSTEM RAW MEASURED TOTAL FLOW RATE AND TS 15.2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION .

POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 / 1

'In accordance with the requirements of'10 CFR:50.4 and 50.90, Wisconsin Electric Power-Company _(Licensee).hereby requests amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear' Plant, Units.1 and 2 respectively, to' ,

incorporate _ changes toLthe plant Technical-Specifications.

The proposed revisions will modify Technical Specification Section 15.3.1.G, " Operation Limitations," Specification 3, to I reduce the reactor' coolant system raw measured total flow rate limit by 2,600 gallons per minute -(gpm), change the overtemperature and overpower setpoints, and change the Reactor Core Safety Limits for Unit 2. Marked-up Technical Specifications pages, a safety. ,

evaluation, and the no significant hazards consideration are enclosed.

This change is being requested because, as more tubes are plugged in the Unit 2 steam generators, the RCS flow' rate decreases.

The' J RCS flow rate limit is the Technical Specification that must be  ;

changed to' accommodate this situation.  ;

DESCRIPTION OF CURRENT LICENSE CONDITION-  ;

Technical-Specification 15.2.1, " Safety Limits, Reactor Core,"

specifies the reactor core safety limits that-aretused to maintain the integrity of the. fuel cladding. The specification states- D that-the combination _of thermal power level, coolant pressure,  :

and' coolant temperature shall'not. exceed the limits shown in-  !

Figure'15.2.1-1. This specification currently applies to both Units 1 and.2.

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1. I Document Control Desk June 11, 1993 Page 2 Section 15.2.3, " Limiting Safety System Settings, Protective Instrumentation," Specification 15.2.3.1.B(4) is the over-temperature AT core limit protection setpoint function.

Specification 15. 2. 3.1. B (5) is the overpower AT core-limit protection setpoint function. These functions provide setpoints that prevent exceeding the reactor core safety limits shown in Figure 15.2.1-1. This specification currently applies to both Units 1 and 2.

Specification 15.3.1.G, " Operational Limitations," specifies the Reactor Coolant System (RCS) operational limitations for DNB (Departure from Nucleate Boiling)-related parameters.

Specification 15.3.1.G.3 specifies that reactor coolant system raw measured total flow rate must be 2181,800 gpm. This specification currently applies to both Units 1 and 2.

DESCRIPTION OF PROPOSED CHANGES This Technical Specification Change Request proposes to revise the reactor core safety limits figure, overtemperature AT setpoint, overpower AT setpoint, and the minimum raw measured total flow rate limit for Unit 2.

A new figure is being added to Specification 15.2.1, " Safety Limit, Reactor Core," which is applicable to Unit 2. The title of the existing figure is being modified to indicate it is applicable to only PBNP Unit 1. Specification 15.2.1 is being modified to read:

"1. The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 for Unit 1 and Figure 15.2.1-2 for Unit 2."

The associated basis is also being changed to reflect the revision to Specification 15.2.1. The proposed basis revision is as follows:

"The curves of Figure 15.2.1-1 and 15.2.1-2 are applicable for 14 x 14 OFA. The curves also apply to the reinsertion of previously-depleted 14 x 14 standard fuel assemblies into an OFA core."

Specifications 15.2.3.1.B(4) and (5) is being modified as follows:

"T' s 573.9*F (Unit 1)

T' s 570.0*F (Unit 2)"

The associated basis is also being changed to reflect the revision to Specification 15.2.3.1.B(4) and (5). .The proposed basis revision is as follows:

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Docume.nt Control Desk June 11, 1993 Page 3 "With normal axial power distribution, the reactor trip limit, with allowance for errorsm, is always below the core safety limit as shown on Figure 15.2.1-1 for Unit 1 and.

Figure 15.2.1-2 for Unit 2. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the overtemperature AT setpoint is automatically reduced *M.

The overpower, overtemperature, and pressurizer pressure system setpoints include the effect of reduced cystem pressure operation (including the effects of fuel densification). The setpoints will not exceed the core safety limits as shown in Figure 15.2.1-1 for Unit 1 and Figure 15.2.1-2 for Unit 2."

Reference (2) changes to "FSAR, Page 14-5."

Specification Section 15.3.1.G, " Operational Limitations," is being modified to provide Reactor Coolant System flow limits specific to each unit as follows:

"3. Reactor Coolant System raw measured Total Flow Rate (See Basis):

a. 2 181,800 gpm (Unit 1)

b. 2 179,200 gpm (Unit 2)"

The associated basis is also being changed to reflect the revision to TS 15.3.1.G.3. The proposed basis revision is as follows:

"The reactor coolant system total flow rate for Unit 1 of 181,800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm).

The reactor coolant system total flow. rate for Unit 2 of 179,200 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (175,400 gpm). . The raw measured flow is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimeter at the beginning of each cycle."

BASIS AND JUSTIFICATION The 2,600 gpm flow reduction in RCS raw measured total flow rate limit for Unit 2 requires a change to the Reactor Core Safety Limits graph which in turn causes the Overtemperature and Overpower AT setpoints to be changed. These changes have been determined to be acceptable based on evaluations performed by Westinghouse and i Wisconsin Electric. Unit 2 raw measured total flow rates of up to 2,600 gpm less than the existing Technical Specification limit do j not pose a safety concern. The results of these evaluations are discussed in the-attached safety evaluation. l

r-Docume.nt Control Desk' June 11, 1993 Page 4 It has been determined that the proposed amendments do n-t involve a significant hazards consideration, authorize a significant change in the types or total amounts of any effluent release, or result in any significant increase in individual or cumulative occupational exposure. Therefore, we conclude that the proposed amendments meet the requirements of 10 CFR 51.22(c) (9) and that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared.

This change is required to support start-up and operation of PBNP Unit 2 following its annual maintenance and refueling outage presently scheduled for September.25 through November 13, 1993. As such, we would appreciate processing of this change request by no later than September 30, 1993.

Please contact us if there are any questions.

Sincerely,

,f;-

e d'/m Bob Link Vice President Nuclear Power CAC/jg Enclosures cc: NRC Regional Administrator, Region III NRC Resident Inspector Public Service Commission of Wisconsin l

Subscribed and sworn before me on this ll'h day of Juar 1993.

J>Ja hu 6%L U l otprfjPublic, State of Wisconsin 1

My commission expires io-29 90 .

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r TECHNICAL SPECIFICATIONS CHANGE REOUEST 160 SAFETY EVALUATION INTRODUCTION Wisconsin Electric Power Company (Licensce) has applied for amendments to Facility Operating Licenseu DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2. The amendments propose to revise Technical Specification 15.3.1.G.3, namely reducing the RCS raw measured total flow rate limit for Unit 2 by 2,600 gpm.

This revision also proposes to change TS 15.2.1, " Safety Limit, Reactor Core" and TS 15.2.3, " Limiting Safety System Settings, Protective Instrumentation."

EVALUATIOH The evaluation performed by Westinghouse covers the Non-Loss of Coolant Accident (Non-LOCA) Transient Analyses, Loss of Coolant Accident (LOCA) analysis, and Steam Generator Tube Rupture (SGTR) analysis. Additionally, the affects of reduced RCS flow were assessed for the systems and component integrity evaluations.

The evaluation performed by Westinghouse concludes that the existing accident analyses and system and component integrity evaluations accommodate a reduction in the RCS total flow rate limit of 2,600 gpm.

Non-LOCA Transient Analyses Evaluation The accident analyses for Point Beach Nuclear Plant are performed by Westinghouse, using Westinghouse's NRC-approved methodologies.

The methodology includes Departure from Nucleate Boiling Ratio '

(DNBR) design margin. As part of the Westinghouse. evaluation, 1%

of the DNBR design margin was allocated to offset the reduction in DNBR margin that would occur at reduced RCS flow rates. This allocation of DNBR design margin justifies up to a 2,600 gpm reduction in the RCS total flow rate limit. The use of DNBR margin in this manner is d2 scribed in the Point Beach FSAR, which states on page 3.2.2-15, "The design margin is applied to account for a rod bow penalty of less than 3% (50,51,52) and to provide for flexibility in the design and plant operation."

RCS flow is one of the factors used to generate the Reactor Core Safety Limits. This reduction in the RCS flow limit requires a change to the Reactor Core Safety. Limits. The Overtemperature and Overpower AT setpoints provide automatic protection to prevent exceeding these safety limits. The change to the. Reactor Core Safety Limits requires a change to the Overtemperature and Overpower AT setpoints. The T' term of these setpoint functions will be reduced from 573.9*F to 570.0*F. Reducing these setpoints provides appropriate protection against the departure

1 from nucleate boiling (DNB) in the core, such that the core thermal limits will. remain satisfied for all of the licensing basis accidents' described in the FSAR for Point Beach Unit 2.

The FSAR Section 14.1.8, " Loss of Reactor Coolant Flow,"

transient analyses were reanalyzed for the lower RCS flow condition with acceptable DNB results.

• An evaluation of the Point Beach FSAR non-LOCA accident analyses that contain non-DNB acceptance criteria was also performed.

All acceptance criteria continue to be met with lower RCS flow.

The reanalysis of the locked rotor event as part of FSAR Section 14.1.8, " Loss of Reactor Coolant Flow," was performed.

A normal operating pressure of 2000 psia was used in the analysis in order to satisfy the RCS pressure limit criteria for this l transient. Therefore, the 2,000 psia operating pressure will be ,

maintained and the 2,250 psia operating pressure will not be used for Unit 2. ,

LOCA Evaluation The Small Break LOCA analysis in the Point Beach FSAR Section 14.3.1 was performed as part of the increased peaking factor change in 1988 (Unit 2 Amendment No. 123). This Small Break LOCA analysis supports a RCS flow rate limit as low as 174,000 gpm. This is less than the RCS flow rate limit of 179,200 gpm being proposed in this Technical Specification Change t Request. Therefore, the Small Break LOCA analysis ~is not affected by this reduction of the RCS flow limit.

4 The Large Break LOCA analysis is described in the Point Beach FSAR Section 14.3.2. In the evaluation performed by Westinghouse  ;

it states that this approxinately 1.5% reduction in RCS flow limit is well within the allowed variance for this parameter and there would be little if any impact on the transient results.

The evaluation by Westinghouse concludes that the Large Break LOCA results do not change and that compliance with 10 CFR 50.46 is maintained.

SGTR Evaluation The Steam Generator Tube Rupture analysis is described in the Point Beach FSAR Section 14.2.4. This analysis is not affected by the RCS flow rate limit reduction, because the flow rate used in the analysis is lower than the proposed RCS flow rate limit.

Other parameters that could affect this analysis are RCS pressure and temperature. This analysis is based on an RCS pressure of 2,250 psia and average temperature 573.9'F. Unit 2 is operated at 2,000 psia and 570*F. The lower pressure'would result in a slightly lower mass release and the lower temperature would result in a slightly higher mass release in this analysis. In the evaluation performed by Westinghouse it was determined that

the pressure effect is greater than the temperature effect and that the off-site radiation doses for the FSAR Section 14.2.4 SGTR analysis remain applicable for Unit 2.

System and Component Intecrity Evaluations ~  :

The evaluation performed.by Westinghouse addresses the number of cycles and the magnitude of the temperature changes. The design specifications for Point Beach assumed a large number of transient cycles. To date, Point Beach Unit 2 has had fewer fatigue cycles than the original design assumed and it is not expected that Point Beach Unit 2 will experience more than a small fraction of the allocated fatigue cycles during the next several years. With regard to the magnitude of the temperature changes, Westinghouse previously assessed the operation of Point Beach at 570*F RCS average temperature and concluded that "

compliance with design standards is maintained. Westinghouse has determined that this conclusion remains valid for this decrease in the RCS flow rate limit.

CONCLUSION The effect of a 2,600 gpm reduction in the RCS total flow rate limit was assessed for each of the PBNP accident analyses. The acceptance criteria of all the accident analyses are still met at this lower flow rate limit. A 2,600 gpm reduction in RCS total flow rate limit has been determined to be acceptable.

Additionally, this reduction in the reactor coolant system raw measured total flow rate limit will not cause any safety limits to be exceeded and the margins of safety for Point Beach Nuclear Plant Unit 2 are not reduced.

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TECHNICAL SPECIFICATION CHANGE REOUEST 160 "NO SIGNIFICANT HAZARDS CONSIDERATION" In accordance with the requirements of 10 CFR 50.91(a), Wisconsin Electric Power Company (Licensee) has evaluated the proposed changes against the standards of 10-CFR 50.92 and has determined that the operation of Point Beach Nuclear Plant, Units 1 and 2 in accordance with the proposed amendments does not present a significant hazards consideration. The analysis of the require-ments of 10 CFR 50.92 and the basis for this conclusion are as follows:

1. Operation of this facility under the proposed Technical Specifications will not create a significant increase in the probability or consequences of an accident previously evaluated. This proposed change reduces the Unit 2 Reactor Coolant System raw measured total flow rate limit by 2,600 gpm. Evaluations performed by Westinghouse and Wisconsin Electric have determined that all the safety analysis requirements are still met at the reduced flow rate limit without increased consequences. A reduction of the RCS flow limit does not affect any parameters that could affect the probability of an accident. Therefore, there is no increase in the probability or consequences of an accident previously evaluated.

2. Operation of this facility under the proposed Technical Specifications change will not create the possibility of a new or different kind of accident from any accident previously evaluated. This proposed change reduces the Unit 2 Reactor Coolant System raw measured total flow rate limit by 2,600 gpm. Evaluations performed by Westinghouse and Wisconsin Electric have determined that all the safety analysis requirements are still met at the reduced flow rate limit and this change does not create the possibility of a new or different kind of accident. There is no physical change to the facility, its systems, or its operation.

Thus, a new or different kind of accident cannot occur.

3. Operation of this facility under the proposed Technical Specifications change will not create a significant reduction in a margin of safety. This proposed change reduces the Unit 2 Reactor Coolant System raw measured total flow rate limit by 2,600 gpm. Evaluations performed by Westinghouse and Wisconsin Electric have determined that all the safety analysis requirements are still met at the reduced flow rate limit. The DNBR margin used for this change in the RCS flow limit is margin in excess of the margin of safety for DNBR. The reduction of the overtemperature and overpower AT setpoints prevent the possibility of exceeding the core safety limits. Therefore, this reduction in RCS total flow rate limit does not reduce any existing margin of safety.