ML20215N220

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GE BWR Extended Load Line Limit Analysis for Nine Mile Point 1 Cycle 9
ML20215N220
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 02/28/1986
From: Gridley R, Sozzi G
GENERAL ELECTRIC CO.
To:
Shared Package
ML17055C615 List:
References
NEDC-31126, NUDOCS 8611040464
Download: ML20215N220 (20)
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NEDC-31126 Class II February 1986 GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR NINE MILE POINT 1 CYCLE 9 W

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Approved:

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Approved: j G. L. Sofzl, (fidnager R. L. Grid 1'ey, Manager Application Analysis Services Safety and Fuel Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

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NEDC-31126 Class II February 1986 GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR NINE MILE POINT 1 CYCLE 9 Approved: /d h

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Approved:

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G. L. Sofzl, (Pldnager R. L. Grid 1%y, Manager Application Analysis Services Safety and Fuel Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

cr NEDC-31126 IMPORTANT NOTICE REGARDING CONTENTS 0F THIS REPORT PLEASE READ CAREFULLY

' This report was prepared by General Electric solely for Niagara Mohawk Power Corporation (NMPC) for NMPC's use in supporting the operation of Nine Mile Point Nuclear Power Station Unit 1.

The information contained in this report is believed by General Electric to be an accurate and true representa-j tion of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting informa-tion in this document are contained in the contract governing Niagara Mohawk Power Corporation Purchase Order No. 31179 dated September 30, 1985. The use of this information except as defined by said contract, or. for any purpose f

other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither the General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document, or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

NEDC-31126 CONTENTS 2.ag 1.

SUMMARY

1-1 2

INTRODUCTION 2-1 3.

DISCUSSION 3-1

3.1 Background

3-1 3.2. Analysis and Results 3-1 3.2.1 Stability 3-1 3.2.2 Loss-of-Coolant Accident 3-2 3.2.3 Containment Response 3-4 3.2.4 Transients 3-4 3.2.5 ASME Pressure Vessel Code Compliance 3-5 3.2.6 Rod Withdrawal Error 3-5 3.3 conclusion 3-5 4.

REFERENCES 4-1 iii/iv

NEDC-31126 TABLES Table Title P3 3-1 Stability Results 3-7 3-2 Transient Results 3-8 3-3 Transient Input Data and Operating Conditions 3-9 3-4 GETAB Analysis Initial Conditions 3-10 3-5 ASME Pressure Vessel Code Compliance: MSIV Closure (No Scram) 3-11 ILLUSTRATIONS Figure Title Pajge 1-1 NMP-1 Operating Power / Flow Map 1-2 v/vi

NEDC-31126 1.

SUMMARY

This report justifies the expansion of the operating region of the power /

flow map for the Nine Mile Point Nuclear Power Station Unit 1 (NMP-1), Cycle 9.

The operating envelope is modified to include the extended operating region bounded by the 108% average power range monitor (APRM) rod block line, the rated power line, and the rated load line, as shown in Figure 1-1.

In this report, rated power is defined as 1850 MWt.

The technical analysis contained in this report is referred to as the extended load line limit analysis (ELLLA) and the entire shaded area in Figure 1-1 is referred to as the ELLLA region.

The discussion and analyses presented show that the consequences of most events initiated from within the ELLLA. region are bounded by the consequences of the same events initiated from the licensing basis condition for NMP-1, Cycle 9.

The Loss of Feedwater Heater (LFWH) transient initiated from within the ELLLA region is the only event that is not bounded by the same event initiated from.the licensing basis condition. However, the consequences of this transient will not affect the minimum critical power ratio (MCPR) operat-ing limit for NMP-1, Cycle 9.

Therefore, it is shown that all safety bases normally applied to NMP-1 are satisfied throughout Cycle 9 for operation within the ELLLA region.

1-1

NEDC-31126 120 RATED POWER = 1860 MWt RATED F LOW = 67.5 heWhr (100,85)100%

INTERCEPT POINT (10041) 100 -

100% APRM ROD BLOCK Lp65 (85M1)

RATED 86 powgg UNE e0 RATED LOAD LWE

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REGION EXPANDED BY EXTENDED g

LOAD UNE UMIT ANALYS48 o

DESCRISED IN THl3 REPORT

... REGION EXPANDED BY EXTENDED Si$ LOAD LWE UMIT ANALYSIS 40 DESCRIGED m REFERENCE 2 INITIAL REG 40N BASED ON NATURAL CIRCUuTIO*4

/ LOAD LINE UMIT ANALYSIS DESCRISED M REFERENCE 1

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CAVITATION PROTECTION

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20 40 00 30 1%

CORE FLOW (% oirsted)

Figure 1-1.

NMP-1 Operating Power / Flow Map 1-2

NEDC-31126 2.

INTRODUCTION The flexibility of a boiling water reactor (BWR) during power ascension in proceeding from the low power / low-core-flow condition to the high power /

high-core-flow condition is limited by two factors. First, if the rated load line control rod pattern is maintained as core flow is increased, changing equilibrium xenon concentrations will result in less than rated power at rated core flow. Second, fuel pellet-cladding interaction considerations inhibit withdrawal of control rods at high power levels. The combination of these two factors can result in the inability to attain rated core power directly.

In this report the analytical bases are provided to overcome these limitations. This is accomplished by allowing operation with a rod pattern that requires few adjustments when ascending to full power. This requires an expansion of the power / flow map to allow 100% power operation at 85% flow (ELLLA). The operating envelope is modified as shown in Figure 1-1.

Future reload submittals will incorporate the use of this extended load line in the analysis.

2-1/2-2

NEDC-31126 3.

DISCUSSION

3.1 BACKGROUND

Previous analyses (References 1 and 2) provided the analytical bases for NMP-1 operation under a modified power / flow line designed to enable direct ascension to full power within the design bases previously applied. The load line limit analysis (LLLA), described in Reference 1 and illustrated in Figure 1-1, enabled the reactor to ascend to full power along a modified power / flow line. This line was bounded by the 108% rod block line up to the 85% power /61% flow point and from there proceeded along the rod block intercept line to the 100% power /100% flow point. An ELLLA, described in Reference 2 and also illustrated in Figure 1-1, provided analyses justifying operation up to 100% of rated power at 91% rated flow. The analysis described in this report extends the ELLLA region to now be entirely bounded by the 108% rod block line up to the rated power line. This allows operation of the plant at 100% rated power with flow as low as 85% rated.

3.2 ANALYSIS AND RESULTS The. modified power / flow curve shown in Figure 1-1 has been derived to provide relief from the operating restrictions inherently imposed during ascension to power utilizing the standard power / flow curve. Effects on the analyzed transients and accidents, which could possibly be impacted by operation in the ELLLA region, are discussed in the following sections.

3.2.1 Stability A stability analysis was performed at the proposed extended APRM rod block line power and the natural circulation flow (65% power /22% flow). The channel hydrodynamic performance and the reactor core stability decay ratio are given in Table 3-1.

1 3-1

I NEDC-31126 The results show that at this least stable condition, the channels and the reactor core decay ratio are within the ul:

4te performance criteria (1.0 decay ratio at all attainable conditions). No technical specifications for stability are required for the NMP-1 reactor.

3.2.2 Loss-of-Coolant Accident The extended load line limit option will allow operation at rated power down to 85% core flow. The effects of this reduced core flow on the conse-quences of a postulated LOCA are as follows:

a.

The lower initial core flow can affect the coastdown response and may yield an earlier boiling transition time.

b.

The higher initial voiding in the bundle due to reduced core flow may result in a slightly earlier dryout time.

A discussion of low-flow effects on LOCA analyses for all operating plants (Reference 3) was presented to and approved by the Nuclear Regulatory Commission (Reference 4). The effects of reduced initial core flow as this i

applies to NMP-1 are addressed further in the following sections. These sections address the effects throughout the break spectrum and include considerations for 3, 4 or 5 recirculation loop operation. Note that full j

power cannot be achieved during 3-loop operation.

3.2.2.1 Small Breaks There will be no significant effect on small break severity due to the lower initial core flow. The peak clad temperature (PCT) is sensitive to the vessel inventory, not the initial core flow. By the time the break uncovers, i

the core flow coastdown will have been completed. The length of time prior to l

core uncovery will also eliminate the effect of higher initial core voiding.

Slight differences in dryout time (if any) will be insignificant compared to the time required to uncover the fuel. Current Maximum Average Planar Linear Heat Generation (MAPLHCR) limits set by the small break will remain unchanged 3-2

NEDC-31126 under both normal operating condition and operation with.one and two recirculation loops out-of-service.

3.2.2.2 Design Basis Accident (Large Breaks)

For the Large Break Design Basis Accidents (DBA) breaks, credit is not taken for coastdown flow due to the rapid decrease in core inlet flow. The lower coastdown flow for ELLLA operation will not affect ECCS calculations for this break size. Dryout time will be slightly faster because of the higher initial voiding, but the effect on the MAPLHCR will be negligible as shown in Reference 3.

The initial voiding in the core is not dependent on the number of recirculation loops operating, but on the initial flow only; therefore, these results are applicable to 3-and 4-loop operation.

3.2.2.3 Intermediate Breaks The lower initial core flow can affect the PCT for the intermediate breaks because the onset of transition boiling and the uncovery of the high power axial plane occurs somewhat earlier. Calculation of the LOCA results using the same conservative LOCA models which were used for the original NMP-1 analysis (Reference 5) would predict higher PCTs for the intermediate breaks during ELLLA operation at 100% power /85% core flow. This could cause the intermediate break to be limiting for certain exposures. However, incorpora-tion of the NRC approved modified Bromley film boiling correlation (Reference

6) into the intermediate break analysis, with the estimated consequences of reduced fuel gap conductance due to increased fission gas release, results in a more realistic intermediate break FCT which is well below the 2200*F limit and approximately 70 degrees below the large break (DBA) PCT. Therefore the small break will still remain limiting at lower exposures and the large (DBA) break limiting at higher exposures. Note that PCT constraints set MAPLHCR limits for the small break and maximum oxidation fraction constraints set MAPLHGR limits for the large break.

3-3

NEDC-31126 3.2.2.4 LOCA Summary Based on the preceding discussion, the standard Cycle 9 MAPLHGR values listed in Reference 5 are applicable for RT.TJ.A operation in 3, 4-and 5-loop operation.

3.2.3 Containment Response The impact of plant operation in the proposed domain for NMP-1 has been evaluated for the containment LOCA response. The operating condition was 102%

of rated power and 85% flow. The results show no impact on the containment LOCA response. The maximum drywell pressurization rate observed is less than the value used in plant-unique testing for defining LOCA-related pool swell loads.

3.2.4 Transients As shown in Reference 7, the most limiting transient event for NMP-1 Cycle 9 is the Turbine Trip Without Bypass event.

For the ELLLA, the following transient events were analyzed at the 100%

power intercept point (100% power /85% flow): Turbine Trip Without Bypass, Feedwater Controller Failure and Loss of Feedwater Hester. These analyses were performed using the nuclear parameters resulting from the end of cycle (EOC) and E0C-1000 mwd /ST target exposure shapes, consistent with rated operation.

The results for both the licensing basis case,and the reduced core flow case are shown in Table 3-2.

Comparison of the initial conditions for these cases is presented in Tables 3-3 and 3-4.

As shown in Table 3-2, the (100%

power /85% flow) transient results are bounded by the licensing basis case (100% power / 100% flow) for the Turbine Trip Without Bypass and.the Feedwater Controller Failure transients. The Loss of Feedwater Heating (LFWH) transient, which is an exposure independent event, was not bounded by the licensing basis 3-4

NEDC-31126 case. However, the increased ACPR result from this event is still bounded by the Cycle 9 licensing basis limiting MCPR result.

3.2.5 ASME Pressure Vessel Code Ccapliance The Main Steam Isolation Valve (MSIV) Closure With No Scram event is used to determine compliance with the ASME Pressure Vessel Code. This event was analyzed at the 100% power intercept point (100% power /85% flow) using the nuclear parameters resulting from the EOC target exposure shape. The result-ing peak vessel pressure is shown.in Table 3-5 and is still below the design pressure of 1375 psig.

3.2.6 Rod Withdrawal Error A Rod Withdrawal Error analysis is not required for the 100% Power and 85% Flow Extended Load Line because the APRM rod block line slope (0.55) does l

not change. The initial 100% power and 85% flow state's proximity to the APRM rod block line, as compared to previously analyzed states, results in a reduc-tion of rod motion prior to a rod block. This reduced rod motion prior to a rod block results in a reduction of the ACPR at each rod block setpoint.

Therefore, the aCPRs of the 100% power /100% flow state will bound the 100%

power and 85% flow state.

3.3 CONCLUSION

The results of most of the transiants for the 100% power intercept point j

(100% power /85% flow) are bounded by the same transients for the licensing basis point (100% power /100% flow). The only exception'is the LFWH transient; however, this event does not affect the current Cycle 9 licensing basis MCPR

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operating limit. The containment LOCA response is unaffected at the operating condition of 102% power and 85% flow. The overpressurizatioa protection analysis results are within ASME Pressure Vessel Code allowable for the 100%

power /85% flow point. The stability results are within the bounds of the ultimate performance criteria, 11.0 decay ratio, and the MAPLHGR results are unchanged by the extended operating region.

3-5

p-NEDC-31126

-l Therefore, it is concluded that all safety bases normally applied to NMP-1 are satisfied throughout Cycle 9 for operation within the ET.T.T.A region.

3-6

NEDC-31126 Table 3-1 STABILITY RESULTS Rod Line Analyzed:

Extrapolated Rod Block Line Natural Circulation Power Reactor Core Stability 0.74 Decay Ratio X /Xo:

2 Channel Hydrodynamic 0.74 Performance Decay Ratio X /X : (P8x8R) 2 0 l

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Table 3-2 TRANSIENT RESULTS Initial Peak Peak Peak Peak Power /

Neutron Heat Steam Line Vessel Exposure Flow Flux Flux Pressure Pressure A CPR (mwd /ST)

(% NBR)

(1 NBR)

(% NBR)

(psig)

(psig)

P8x8R Turbine Trip EOC-1000 100/100 686.0 121.3 1270 1285 0.270 without Bypass EOC-1000 100/85 646.0 120.5 1269 1284 0.233 Turbine Trip EOC 100/100 695.8 125.9 1292 1306 0.340 without Bypass EOC 100/85 665.3 122.0 1282 1297 0.272 Feedwater Con-EOC-1000 100/100 157.4 109.4 1139 1176 0.078 b

[

troller Failure EOC-1000 100/85 148.2 108.8 1140 1173 0.073 b

Feedwater Con-EOC 100/100 205.9 114.9 1145 1184

.0.156 troller Failure EOC 100/85 181.3 112.1 1143 1177 0.098 100/100 115.9 115.4 1017 1073 0.142 Loss of Feedwater Heater" 100/85 118.5 117.7 1019 1075 0.158 aExposure independent event

NEDC-31126 Table 3-3 TRANSIENT INPUT DATA AND OPERATING CONDITIONS Licensing Basis Point Intercept Point (100% power /100% flow)

(100% power /85% flow)

Thermal Power (MWt/%)

1850/100 1850/100 Steam Flow (M1b/hr/%)

7.32/100 7.33/100 Core Flow (M1b/hr/%)

67.5/100 57.4/85 Dome Pressure (paig) 1030 1029 Turbine Pressure (psig) 950 949 Relief Valves (No./% NBR) 6/45.7 6/45.7 Low Setpoint (psig)a 1102 1102 Spring Safety Valves 16/140.8 16/140.8 (No./% NBR)

Low Setpoint (psig)a 1230 1230

" Nominal setpoint + 1%

1 3-9

NEDC-31126 Table 3-4 GETAB ANALYSIS INITIAL CONDITIONS Licensing Basis Point Intercept Point (100% power /100% flow)

(100% power /85% flow)

Core Power (MWt) 1850 1850 Core Flow (M1b/hr) 67.5 57.4 Core Pressure (psig).

1045 1042 Inlet Enthalpy (Btu /lb)'

526.6 522.3 Nonfuel Power Fraction 0.04 0.04 Axial Peaking Factor 1.40 1.40 P8x8R Fuel Local Peaking Factor 1.20 1.20 Radial Peaking Factor 1.71a/1.63b 1.698/1.64b R-Factor 1.051 1.051 Bundle Power (qt) 5.806a/5.542b 5.725a/5.572b Bundle Flow (10 lb/hr) 102.14 a/104.09b 86.05a/87.03b aEOC - 1000 mwd /ST bEOC 3-10

NEDC-31126 Table 3-5 ASME PRESSURE VESSEL CODE COMPLIANCE: MSIV CLOSURE (N0 SCRAM)

Peak Initial Peak Peak Steam Line Peak Vessel Exposure Power / Flow Neutron Flux Heat Flux Pressure Pressure (mwd /ST)

(% NBR)

(% NBR)

(% NBR)

(psig)

(psig)

EOC 100/100 572 147 1289 1328 E0C 100/85 574 145 1290 1329 3-11/3-12

NEBC-31126 4

REFERENCES 1.

"Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis License Amendment Submittal," General Electric Company, May 1977 (NEDO-24012).

2.

J. V. Woodford, "Nine Mile Point Nuclear Power Station Unit 1 Extended Load Line Limit Analysis License Amendment Submittal (Cycle 6)," General Electric Comptny, April 1979 (NEDO-24185).

3.

Letter from R. L. Gridley (GE) to D. G. Eisenhut (NRC), " Review of Low Core Flow Effects on LOCA Analysis for Operating BWRs," May 8,1978.

4.

Letter from D. G. Eisenhut (NRC) to R. L. Gridley (GE), enclosing " Safety Evaluation Report Revision of Previously Imposed MAPLHGR (ECCS-LOCA)

Restrictions for BWRs at Less Than Rated Flow," May 19, 1978.

5.

" Loss-of-Coolant Accident Analysis Report for Nine Mile Point Unit One Nuclear Power Station," General Electric Company, August 1981 (NEDO-24348, as amended).

6.

J. E. Leonard, et al., " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K-Amendment No.1 - Calculation of Low Flow Film Boiling Heat Transfer for BWR LOCA Analysis," General Electric Company, October 1982 (NED0-20566-1-A, Rev.

1).

7

" Supplemental Reload Licensing Submittal for Nine Mile Point Nuclear Power Station Unit 1, Reload 10," General Electric Company, October 1985 (23A4717).

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