ML17290A437

WNP-2,Cycle 9 Colr

ML17290A437
Person / Time
Site:	Columbia
Issue date:	06/30/1993
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17290A435	List: GO2-93-108, Forwards COLR 93-9,Rev 0, WNP-2,Cycle 9 COLR, Per TS 6.9.3.4 & 930510 Application for Amend to Stability & Power/ Flow Sections of TS ML17290A437
References
COLR-93-9, COLR-93-9-R, COLR-93-9-R00, NUDOCS 9306170334
Download: ML17290A437 (43)

Text

a 930603 l3:47 COLR 93-9 Rev. 0 Controlled Copy No.

WNP-2

'ycle 9 Core Operating Limits Report June 1993 Washington Public Power Supply System I'>aai70aa4 9SOS>,

PDR ADOCK 05000397 P

PDR

WNP-2 Cycle 9 Core Operating Limits Report 08:33, 05/l3/93 List of Effective Pages

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WNP-2 Cycle 9 Core Operating Limits Report 930520 l7:34 List of Effective Pages (cont.)

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Table of Contents Pa~e 1.0 Introduction and Summary.....,.............................

1 2.0 Average Planar Linear Heat Generation Rate (APLHGR) Limits for Use in Technical Specification 3.2.1......,..........................

2 3.0 Minimum Critical Power Ratio (MCPR) Limitfor Use in Technical Specification 3.2.3

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4.0 Linear Heat Generation Rate (LHGR) Limitfor Use in Technical Specification 3.2.4.......................................

29 5.0 References...,............... ~.....................

~... 35 Washington Nuclear-Unit 2 COLR 93-9 Rev. 0

930SIS 13:01 1.0 Introduction and Summary This report provides the Average Planar Linear Heat Generation Rate (APLHGR) limits, the Minimum Critical Power Ratio (MCPR) limits, and the Linear Heat Generation Rate (LHGR) limits for WNP-2, Cycle 9 as required by Technical Specification 6.9.3.1.

As required by Technical Specifications 6.9.3.2 and 6.9.3.3, these limits were determined using NRC-approved methodology and are established so that all applicable limits of the plant safety analysis are met.

The thermal limits for SPC fuel given in this report are documented in the Cycle 9 Plant Transient Analysis Report (Reference 5.1) and the Cycle 9 Reload Analysis Report (Reference 5.2). The thermal limits determined through the approved methodology are modified for the GE11 and SVEA-96 LFA's as discussed below.

The WNP-2 Cycle 9 reload includes four Siemens Power Corporation (SPC), four General Electric (GE), and four ABB Atom (ABB) Lead Fuel Assemblies (LFA's). The SPC LFA's were inserted during the reload for Cycle 5.

The GE and ABB LFA's were inserted at the beginning of Cycle 6 and were designed to be compatible with the reload fuel utilized in Cycle

6. The LFA's are loaded in core locations which analysis has shown to have sufficient thermal margin such that the LFA's are not expected to be the most limiting fuel assemblies on either a nodal or an assembly power basis.

The GE11 LFA is described in the GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6 (Reference 5.3).

This reference describes the design goals of the GE11 LFA's and provides support for monitoring the GE11 LFA's at thermal limits based on the SPC 8x8 reload fuel thermal limits.

The SVEA-96 LFA's is described in the Supplemental LFA Licensing Repon S VE'A-96LFA'sfor WNP-2 (Reference 5.4). The process for developing thermal limits for the SVEA-96 LFA's based upon the SPC 8x8 reload fuel thermal limits is described in References 5.4 and 5.5.

The MAPLHGRlimits for the GE11 LFA's are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/[81-7]) is applied to account for the different number of fuel pins in the two designs.

The MAPLHGR limits for the SVEA-96 LFA's are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/[100-4]) is applied to account for the different number of fuel pins in the two designs.

Furthermore, the MAPLHGR limits for the SVEA-96 LFA's are multiplied by the following constants:

(a) 1.04 to account for a different estimation of the local power in the output from POWERPLEX compared to ABB Atom methods and (b) 1.02 to account for a different estimation of exposure in the output from POWERPLEX compared to ABB Atom methods.

The MCPR limitis the maximum of (a) the applicable exposure dependent, fullpower and full flow MCPR limit, (b) the applicable exposure and power dependent MCPR limit, and (c) the flow dependent MCPR limitspecified in this report. This stipulation assures that the safety limit MCPR willnot be violated throughout the WNP-2 operating regime. Full power MCPR limits are specified to define operating limits at rated power and flow conditions from 85% to 106%

flow. For the WNP-2 core, the Load Rejection without Bypass transient is limitingfor operation at rated power and flow. Power dependent MCPR limits are specified to define operating limits at other than rated power conditions. For the WNP-2 core, feedwater-controller-failure transients from reduced power are calculated to be more severe than from fullpower conditions. A flow Washington Nuclear-Unit 2 COLR 93-9 Rev. 0

~ 930601 1S:15 dependent MCPR is specified to define operating limits at other than rated flowconditions.

The reduced flow MCPR limit provides bounding protection for the limiting recirculation flow increase transient.

The LHGR limits for the GE11 LFA's are the same as for the SPC 8x8 reload fuel, except that a ratio ([64-2]/[81-7]) is applied to account for the different number of fuel pins in the two designs. The LHGR limits for the SVEA-96 LFA's are taken directly from Reference 5.4.

The reload licensing analyses for this cycle provide operating limits for Extended Load Line (ELLLA)operation which extends the power and flow operating regime for WNP-2 up to the 109% rod line which at fullpower corresponds to 87% of rated flow. The MCPR limits defined in this report are applicable up to 100% of rated thermal power along and below the 109% rod line. The minimum flow for operation at rated power is 87% of rated flow. References 5.1, 5.2 and 5.35 through 5.43 document the analyses in support of ELLLAoperation.

Preparation, review and approval of this report were performed in accordance with applicable Supply System procedures.

The specific topical report revisions and supplements which describe the methodology utilized in this cycle specific analysis are referenced in Section 5.0 2.0 Average Planar Linear Heat Generation Rate (APLHGR) Limits for Vse in Technical Specification 3.2.1 The APLHGR's for use in Technical Specification 3.2.1 shall not exceed the limits shown in Figures 2.1, 2.2, 2.4, and 2.5 when in two-loop operation and in Figures 2.1, 2.3, 2.4, and 2.5 when in single loop operation.

The limits for each fuel type as a function of Average Planar Exposure are provided for the SPC reload fuel, the SPC LFA's, the SVEA-96 LFA's, and the GE11 LFA's.

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930S17 I?:30 3.0 Minimum Critical Power Ratio (MCPR) Limit for Use in Technical Specification 3.2.3 The MCPR limitfor use in Technical Specification 3.2.3 shall be:

Greater than or equal to the greater of the limits determined from Tables 3.1a and 3.1b and Figures 3.1 and 3.2a through 3. lib.

Washington Nuclear-Unit 2 COLR 93-9 Rev. 0

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08:53, 0$/13/93 Table 3.1a WNP-2 Cycle 9 MCPR Operating Conditions Cycle Exposures s 4500 MWd/MTU SLMCPR ~ 1.07.'"

Condition Limit NSS~u Full Power Flow Dependent Power Dependenà TSSS~n Full Power Flow Dependent Power Dependent NSSro RPT Full Power Inoperable Flow Dependent Power Dependenà SLY NSS Full Power Flow Dependent Power Dependents SL~SSS Full Power Flow Dependent Power Dependent SLY NSS RPT Full Power Inoperablc Flow Dependent Power Dependent+

SPC gx8 SPC 9x9 SPC 9x9 SVEA-96 GE11 LFA LFA 1.25+

1.25 1.27 1.39+

Figure 3.1 Fig. 3.2a Fig. 39a Fig. 3.3a Fig. 3.2a 1.27 1.27 1.33 1.42 Figure 3.1 Fig. 3.4a Fig. 3.$a Fig. 3.5a Fig. 3.4a 1.29 1.29 1.39 1.45 Figure 3.1 Fig. 3.10a Fig. 3.11a Fig. 3.11a Fig. 3.10a 186 136 1.36 1.85 None Fig. 3.2a Fig. 3.3a Fig. 3.3a Fig. 3.2a 1.$6 1.36 1.36 1.85 None Fig. 3.4a Fig. 3.5a Fig. 3.5a Fig. 3.4a 1.56 136 1.36 1.85 None Fig. 3.10a Fig. 3.11a Fig. 3.11a Fig. 3.10a W'ashington Nuclear-Unit 2 COLR 93-9 Rev. 0

930317 11:33 Table 3.1b WNP-2'ycle 9 MCPR Operating Conditions Cycle Exposures ) 4500 MWd/MTU SLMCPR ~ 1.0'P" SLMCPR ~ 1.07'n FFTR Condition Limit NSSn>>

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1.31 1.29 1.39 1.48 Figure 3.1 Fig. 3.2b Fig. 3.3b Fig. 3.3b Fig. 3.2b 1.34;,

1.33 1.43 1.52 Figure 3.1 Fig. 3.4b Fig. 3.5b Fig. 3.5b Fig. 3.4b 1.33 1.31 1.41 1.51 Figurc 3.1 Fig. 3.6 Fig. 3.7 Fig. 3.7 Fig. 3.6 1.36 1.35 1.45 1.55 Figure 3.1 Fig. 3.8 Fig. 3.9 Fig. 3.9 Fig. 3.8 RPT Full Power Inoperable Flow Dependent Power Dependent e SLO+ NSS Full Power Flow Dependent Power Dependent SLO'SSS Full Power Flow Dependent Power Dependent SLOn>> NSS RFI'ullPower Inoperable Flow Dependent Power Dependent+

129 1.38 182 1.60 Figure 3.1 Fig. 3.10b Fig. 3.11b Fig. 3.11b Fig. 3.10b 1.56 I

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93060l l4:5i Notes for Tables 3.1a and 3.1b Note 1: The scram insertion times must meet the requirements of Technical Specification 3.1.3.4. The NSS MCPR values are based on the SPC transient analysis performed using the control rod insertion times shown below (defined as normal scram speed:

NSS).

In the event that Surveillance 4.1.3.2 shows these scram insertion times have been exceeded, the MCPR limit shall be determined from the applicable Technical Specification Scram Speed (TSSS) MCPR limits in Tables 3.1a and b.

Position Inserted From Fully Withdrawn Notch 45 Notch 39 Notch 25 Notch 5

Slowest measured average control rod insertion times t

specified notches for all operable control rods for each grou of four control rods arranged in a two-by-two array (seconds) 0.380 0.720 1.600 2.950 Note 2: For Single Loop Operation (SLO), the SLMCPR increases by 0.01. The increase is included in the MCPR limits for SLO.

Note 3: For the noted fullpower MCPR limits, the control rod withdrawal error (CRWE) event is limiting. The load rejection without bypass (LRNB) event is limiting for the remaining fullpower limits. CRWE analysis was performed with a nominal rod block

'onitor (RBM) setpoint of 1.06. Use of the nominal setpoint is in accordance with the methodology described in Reference 5.11, consistent with approved industry practice.

LRNB analysis was performed at 100% power/106% flow, 104% power/106% flow and 104% power/85% flow. The more limiting results are used for fullpower limits in Tables 3.1a and b.

Note 4: Power dependent MCPR limits are provided for core thermal powers greater than or equal to 25%'f rated power at all core flows. However, the power dependent MCPR limits for core thermal powers less than or equal to 30% of rated power are subdivided by core flow. Limits are provided for core flows greater than 50% of rated flow and less than or equal to 50% of rated flow, respectively. A step change in the power dependent MCPR limits occurs at 30% of rated power because direct scram on turbine throttle valve closure is automatically bypassed per Technical Specification 3.3.1.

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08:53, 05/13/93 4.0 Linear Heat Generation Rate (LHGR) Limit for Use in Technical Specification 3.2.4 The LHGR limitfor use in Technical Specification 3.2.4 shall not exceed the values shown in Figures 4.1 through 4.5.

Washington Nuclear-Unit 2, COLR 93-9 Rev. 0

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93060l 14:49 5.0 References 5.1 EMF-93-047, WNP-2 Cycle 9 Plant Transient

Analysis, Siemens Power Corporatiori May 1993.

5.2 EMF-93-048, WNP-2 Cycle 9 Reload Analysis, Siemens Power Corporation, May 1993.

5.3 GEll Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6, General Electric Company, December 1989 (Proprietary).

5.4 UK 90-126, Supplemental Lead Fuel Assembly Licensing ReportSVEA-96 LFA's for WNP-2, ABB Atom, January 1990 (Proprietary).

5.5 ATOF-91-120, W. R. Harris, ABB Atom, to D, L, Whitcomb, Supply System, Assembly Treatment in WNP-2 Cycle 7 Core Operating Limits Report, May 1, 1991.

5.6 SPCWP-93-0077, Udell Fresk, Siemens Power Corporation, to R. A. Vopalensky, Supply System, SPC Comments on WNP-2 Cycle 9 Draft COLR Report, May 21, 1992.

5.7 JTW:93-073, J. T. Worthington, General Electric Company, to D. L. Whitcomb, Supply System, WNP-2 Cycle 9 Core Operating Limits Report, Contract No.

C-21099, GEll Lead Fuel Assemblies, May 22, 1992.

5.8 5.9 ATOF-93-059, W. R. Harris, ABB Atom, to D. L. Whitcomb, Supply System, SV&l-96Lead Fuel Assembly Treatment in WNP-2 Cycle 9 Core Operari ng Limits Report, May 15, 1992.

ANF-89-014(P)(A),.Revision 1 and Supplements 1 & 2, Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 5xx9-9X Reload Fuel, Advanced Nuclear Fuels Corporation, October 1991.

5.10 5.11 XN-NF-79-71(P), Revision 2, including Supplements 1,

2 and 3(A), Exxon Nuclear Plant Transient MethodologyforBoiling Water Reactors, Exxon Nuclear Company, Inc., November 1981.

I XN-NF-80-19(P)(A), Volume 1

and Supplements 1

and 2,

Exxon Nuclear Methodology for Boiling Water Reactors Neutronic Methods forDesign Analysis, March 1983.

Washington Nuclear-Unit 2 COLR 93-9 Rev. 0

93060 l 14:49 5.12 XN-NF-80-19(P)(A), Volume 1 Supplements 3 and 4, Advanced Nuclear Fuels Methodology for Boiling Water Reactors:

Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology, Advanced Nuclear Fuels Corporation, November 1990.

5.13 XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, Inc., January 1987.

5.14 XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application ofthe ENC Methodology to BWR Reloads, Exxon Nuclear Company, Inc., June 1986.

5.15 ANF-913(P)(A), Volume 1, Revision 1 and Volume 1, Supplements 2, 3, and 4, COTRANSA2: A Computer Program forBoiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

5.16 ANF-524(P)(A), Revision 2 and Supplements 1 and 2, Advanced Nuclear Fuels Critical Power MethodologyforBoiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.

5.17 ANF-1125(P)(A) and Supplements 1 and 2, ANFB Critical Power Correlarion, Advanced Nuclear Fuels Corporation, April 1990 5.18 Letter, R. C. Jones (NRC) to R. A. Copeland (ANF), NRC Approval ofANFB Additive Constants for 9x9-9X BWR Fuel, November 14, 1990.

5.19 Letter ENWB-86-0067, J. B. Edgar (ANF) to Supply System, Supplemental Licensing Analysis Results, April 15, 1986.

5.20 ANF-90-01, WNP-2 Cycle 6 Plant Transient Analysis, Advanced Nuclear Fuels Corporation, January 1990.

5.21 XN-NF-84-105(P)(A), Volume 1 and Supplements 1, 2, & 4, XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., February 1987.

5.22 XN-NF-81-21(P)(A), Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, Inc., September 1982, and Supplement 1, March 1985.

5,23 XN-NF-85-67(P)(A), Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, Inc., September 1986.

5.24 XN-NF-81-58(P)(A), Revision 2, RODEX2; Fuel Rod Mechanical

Response

Evaluarion Model, Exxon Nuclear Company, Inc., March 1984.

Washington Nuclear-Unit 2 COLR 93-9 Rev. 0

930603 13:47 5.25 XN-NF-87-92 and Supplement 1, WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction, Advanced Nuclear Fuels Corporation, June 1987 and May 1988.

5.26 ANF-87-119, WNP-2 Single Loop Operation Analysis, Advanced Nuclear Fuels Corporation, September 1987.

5.27 XN-NF-79-59(P)(A), Methodologyfor Calculation ofPressure Drop in BWR Fuel Assemblies, Exxon Nuclear Company, Inc., November 1983.

5.28 ANF-87-118, WNP-2 LOCA Analysis For Single Loop Operarion, Advanced Nuclear Fuels Corporation, September 1987.

5.29 Letter, R. B. Samworth, USNRC, to G. C. Sorensen, Supply System, Issuance ofAmendment No. 62 to Facility Operating License No. NPF-21-WPPSS Nuclear Project 2 PAC No. 67538), August 5, 1988.

5.30 XN-NF-85-138(P), LOCA Break Spectrum fora BWR5, Exxon Nuclear Company, Inc., December 1985.

5.31 XN-NF-85-139, WNP-2 LOCA-ECCS Analysis, MAPLHGR Results, Exxon Nuclear Company, Inc., December 1984.

5.32 ANF-CC-33(P)(A), Supplement 2, HUV: A Generalized Multirod Heatup Code with 10 CFR 50Appendix KHeatup Option, Advanced Nuclear Fuels Corporation, January 1991.

5.33 XN-NF-81-22(P)(A), Generic Statisrical Uncertainty Analysis Methodology, November 1983.

5.34 NEDE-24011-P-A-6, General Electric Standard Applicarion for Reactor Fuel, April 1983.

5.35 "Reactor Vessel Internals Evaluation Task Report for WNP-2 Power Uprate Project," GE Nuclear Energy, April 1993 (DRAFT).

5.36 "WNP-2 Power Uprate Containment Response Evaluation Input to Engineering Report," GE Nuclear Energy, January 19, 1993 (DRAFT).

5.37 GE-NE-189-69-1092, "Effects of Adjustable Speed Drive on Reactor Internal Vibration at the WNP-2 Nuclear Power Plant," GE Nuclear Energy, October 1992.

5.38 GE-NE-189-34-0392, "Jet Pump Sensing Line Vibration Test for Washington Nuclear Project 2," GE Nuclear Energy, March 1992.

Washington Nuclear-Unit 2 COLR 93-9 Rev. 0

0 930605 l3:47 5.39 NEDE-24222, "Assessment ofBWR Mitigation of ATWS, Vol. II(NUREG 0460 Alternate No, 3)," General Electric Company, December 1979.

5.40 "Washington Nuclear Project Unit 2 System Evaluation Report for Power UprateReactor Recirculation Control System," GE Nuclear Energy, February',

1993.

5.41 GE Report 22A7104, Revision 0, "Dynamic Load ReportFuel Vertical Support,"

GE Nuclear Energy, June 30, 1982.

5.42 "Fuel LiftNon-Proprietary Letter," Letter, DM Kelly (GE) to WC Wolkenhauer (SS); February 15, 1993.

5,43 93-PU-0054, "ELLLARelated Power Uprate Task Reports," Letter, DM Kelly (GE) to WC Wolkenhauer (SS), June 3, 1993.

Washington Nuclear-Unit 2 COLR 93-9 Rev. 0