ML032750925

Tech Spec Pages for Amendment No. 46, Modifying TS 5.9.5 to Add Three Additional Methodologies in Support of the Westinghouse 17X17 Rf A-2 Fuel Design with Intermediate Flowmixers

ML032750925
Person / Time
Site:	Watts Bar
Issue date:	09/30/2003
From:	NRC/NRR/DLPM/LPD2
To:
References
TAC MB7746
Download: ML032750925 (5)

Text

Reporting Requirements 5.9 5.9 Reporting Requirements 5.9.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

3. WCAP-10216-P-A, Revision 1A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F(Q) SURVEILLANCE TECHNICAL SPECIFICATION, February 1994 (W Proprietary).

(Methodology for Specifications 3.2.1 - Heat Flux Hot Channel Factor (W(Z) Surveillance Requirements For F(Q) Methodology) and 3.2.3 - Axial Flux Difference (Relaxed Axial Offset Control).)

4. WCAP-12610-P-A, <<VANTAGE + FUEL ASSEMBLY REFERENCE CORE REPORT,' April 1995. (W Proprietary).

(Methodology for Specification 3.2.1 - Heat Flux Hot Channel Factor).

5. WCAP-15088-P, Rev. 1, sSafety Evaluation Supporting A More Negative EOL Moderator Temperature Coefficient Technical Specification for the Watts Bar Nuclear Plant,' July 1999, (W Proprietary), as approved by the NRC staff's Safety Evaluation accompanying the issuance of Amendment No. 20 (Methodology for Specification 3.1.4 - Moderator Temperature Coefficient.).

6. Caldon, Inc. Engineering Report-80P, *Improving Thermal Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM^> System,'

Revision 0, March 1997; and Caldon, Inc.

Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFMV",' Revision 0, May 2000; as approved by the NRC staff's Safety Evaluation accompanying the issuance of Amendment No. 31.

7. WCAP-11397-P-A, "Revised Thermal Design ProcedureN April 1989. (Methodology for Specification 3.2.2 -

Nuclear Enthalpy Rise Hot Channel Factor).

8. WCAP-15025-P-A, 'Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17 x 17 Rod Bundles with Modified LPD Mixing Vane Grids,'

April 1999. (Methodology for Specification 3.2.2

- Nuclear Enthalpy Rise Hot Channel Factor).

9. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,' October 1999. (Methodology for Specification 3.2.2 -

Nuclear Enthalpy Rise Hot Channel Factor).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis, limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

Watts Bar-Unit 1 5. 0-33 Amendment 11, 20, 31, 38,46

Reactor Core SLs B 2.1.1 BASES BACKGROUND DNB is not a directly measurable parameter during operation; (continued) therefore, THERMAL POWER, reactor coolant temperature, and pressure are related to DNB through critical heat flux (CHF) correlations. The primary DNB correlations are the WRB-1 correlation (Ref. 7) for VANTAGE 5H and VANTAGE+ fuel and the WRB-2M correlation (Ref.8) for RFA-2 fuel with IFMs.

These DNB correlations take credit for significant improvement in the accuracy of the CHF predictions. The W-3 CHF correlation (Ref. 9 and 10) is used for conditions outside the range of the WRB-1 correlation for VANTAGE 5H and VANTAGE+ fuel or the WRB-2M correlation for RFA-2 fuel with IFMs.

The proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs.

APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES 4normal operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95%

confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b. The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the following functions:

a. High pressurizer pressure trip;

b. Low pressurizer pressure trip;

c. Overtemperature AT trip; (continued)

Watts Bar-Unit 1 B 2.0-2 Amendment 46

Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS To meet the DNB design criterion, uncertainties in plant (continued) operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered. The effects of these uncertainties have been statistically combined with the correlation uncertainty to determine design limit DNBR values that satisfy the DNB design criterion. SL 2.1.1 reflects the use of the WRB-1 CHF correlation with design limit DNBR values of 1.25/1.24 (typical/thimble) for VANTAGE 5H and VANTAGE+ fuel and the WRB-2M CHF correlation with design limit DNBR values of 1.23/1.23 (typical/thimble) for RFA-2 fuel with IFMs.

Additional DNBR margin is maintained by performing the safety analyses to higher DNBR limits. This margin between the design and safety analysis limit is more than sufficient to offset known DNBR penalties (e.g., rod bow and transition core) and to provide the DNBR margin for operating and design flexibility.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation responses are applicable to the VIOLATIONS reactor core SLs.

2.2.1 If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

(continued)

Watts Bar-Unit 1 B 2.0-4 Amendment 7, 4 6

Reactor Core SLs B 2.1.1 BASES References 4. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation (continued) Methodology," July 1985.

5. Title 10, Code of Federal Regulations, Part 50.72, wImmediate Notification Requirements for Operating Nuclear Power Reactors."

6. Title 10, Code of Federal Regulations, Part 50.73, "Licensee Event Report System."

7. WCAP-8762-P-A, 'New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," July 1984.

8. WCAP-15025-P-A, -Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17 x 17 Rod Bundles with Modified LPD Mixing Vane Grids,, April 1999.

9. Tong, L. S., "Boiling Crisis and Critical Heat Flux,"

AEC Critical Review Series, TID-25887, 1972.

10. Tong, L. S., "Critical Heat Fluxes on Rod Bundles," in "Two-Phase Flow and Heat Transfer in Rod Bundles," pages 31 through 41, American Society of Mechanical Engineers, New York, 1969.

Watts Bar-Unit 1 B 2.0-6 Amendment 7, 46

FAH B 3.2.2 BASES BACKGROUND Operation outside the LCO limits may produce unacceptable (continued) consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.

APPLICABLE Limits on Fem preclude core power distributions that exceed SAFETY ANALYSES the following fuel design limits:

a. There must be at least 95% probability at the 95%

confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition;

b. During a loss of coolant accident (LOCA), the peak cladding temperature (PCT) must not exceed 22007F for small breaks, and there must be a high level of probability that the peak cladding temperature does not exceed 22001F for large breaks (Ref. 3);

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 1); and

d. Fuel design limits required by GDC 26 (Ref. 2) for the condition when control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn.

For transients that may be DNB limited, FNS is a significant core parameter. The limits on Fea ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency. The DNB design basis is met by limiting the minimum local DNB heat flux ratio to a value which satisfies the 95/95 criterion for the DNB correlation used.

Refer to the Bases for the Reactor Core Safety Limits, B 2.1.1 for a discussion of the applicable DNBR limits. The W-3 Correlation with a DNBR limit of 1.3 is applied in the heated region below the first mixing vane grid. In addition, the W-3 DNB correlation is applied in the analysis of accident conditions where the system pressure is below the range of the WRB-1 correlation for VANTAGE 5H and VANTAGE+ fuel or the WRB-2M correlation for RFA-2 fuel with IFMs. For system pressures in the range of 500 to 1000 psia, the W-3 correlation DNBR limit is 1.45 instead of 1.3.

(continued)

Watts Bar-Unit 1 B 3.2-13 Amendment 7, 21, 46