ML20011A774

Yankee Atomic BWR Analysis Methods:Analysis of Typical BWR/4 Turbine Trip W/O Bypass Transients.

ML20011A774
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Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	10/30/1981
From:	Cronin J, Holzer J, Sironen M VERMONT YANKEE NUCLEAR POWER CORP.
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ML20011A772	List: ML20011A771 ML20011A774
References
YAEC-1280, NUDOCS 8111030142
Download: ML20011A774 (34)

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I YANKEE ATOMIC ELECTRIC COMPANY g

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~AN KEE g l I  :

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1 8111030142 811030 DR ADOCK 05000

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YANKEE ATOMIC BOILING WATER REACTOR ANALYSIS METHODS:

ANALYSIS OF A TYPICAL BWR/4 TURBINE TRIP WITHOUT BYPASS TRANSIENT

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<R Prepared By ), [ O!8/

R T. Cronin (Date)

BWR Transient Analysis Group Prepared By 8 #er __

to 13o IEl J. M.' Holzer J (Date)

Applied Methods Development Group Prepared By '/_,t. /#!.50 f/

M. (s. Sirond( '(Dafe)

, I Reactor Physics Grcop Reviewed By [ lo/30IBl S. W S'ch ul t z , 4f ana gel (Date)

BWR Transient Ana 's Croup Approved By ,

/8!30 8/

B. C. Slifer, Manage (Datd)

Nuclear Engineering partment Yankee Atomic Electric Company Nuclear Services Division 1671 Worcester Road I Fr amingh am , Ma s sa ch us e tt s 01701

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I DISCLAIMER OF RESPONSIBILITY I This document was prepared by Yankee Atomic Electric Company on behalf o f Vermont Yankec Nuclear Power Corporation. This document is believed to be I completely true and accurate to the best of our knowledge and information, It is authorized for un specifically by Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Ccrporation and/or the appropriate subdivisions within the Nuclear F gulatory Cecnitaion only.

With regsrd to any unauthorized use whatsoever, Yankee Atomic Electric I Company, Vermont. Yankee Nuclear Pouer Corporation und their officers, directors, agents and employees assume no liability nor make any warranty or representation with respect to the contents of this document or to its accuracy or completeness.

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ABSTRACT Simulation results obtained using Yankee Atomic Electric Company's BWR analysis methods are presented along with comparison to the results of other workers for a turbine trip without bypass transient. This work was requested by the United States Nuclear Regulatory Commission to aid in its review of Yankee Atomic Electric Company's BWR analysis methods.

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TABLE OF CONTENTS Page DISCLAIMER ii ABSTRACT iii TABLE OF CONTENTS iv LIST OF FIGURES y LTST Ut TABLES vi ACKNOWLEDGEMENTS vii

1.0 INTRODUCTION

1 2.0 METHODOLOGY EMPLOYED 2 I 2.1 2.2 Steady State. Physics Transient Physics 2

3 2.3 Core Wide Transient Analysis Model 4 3.0 ANALYSit 13 3.1 Initial Conditions 13 3.2 Analysis Results 13 3.3 Comparisons to Results of Other Workers 13 l 4.0 COFCLUSIONS 25 REFERENCEF 26 l I

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I I LIST OF FIGURES Number Title Pa ste 2.1 Peach Bottom 2 Bundle Types 7 2.2 Peach Bottom 2 SIMULATE Input Data 8 I 2.3 Comparison of SIMULATE Axial Power Distribution (without added thermal absorber) to GE and BNL Distributions 9

2.4 Axial Distribution of Thermal Absorber Added to 10 SIMULATE I 2.5 Comparison of SIMULATE Axial Power Distribution (with added thermal absorber) to GE and BNL Distributions 11 2.6 Fod Worth versus Positon and Rod Position versus 12 Time after Initici Rod Movement 3.1 Neutron Power Prediction 15 3.2 Transient Reactivity Components 16 3.3 Core Average Heat Flux Prediction 17 I

3.4 Active Core Inlet Flow Prediction 18 3.5 Steam Dome Pressure Prediction 19 3.6 Core Mid-Plane Pressure Prediction 20 3.7 Comparison of Feutron Power Prediction to 21 GE and BNL Predictions 3.8 Comparison of Core Average Heat Flux 22 Prediction to GE and BNL Predictions 3.9 Comparison of Core Inlet Flow Prediction to 23 GE and BNL Predictions 3.10 Comparison of Core Mid-Plane Pressure Prediction 24 to GE and BNL Predictions I

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I I LIST OF TABLES Number Title Page 2.1 Peach Bottom Unit 2 Initial Conditions 5 2.2 Peach Bottom Unit 2 Transient Physics 6 Parameters 3.1 Summary of System Transient Model Initial 14 Conditions  !

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ACKNOWLEDGEMENTS I

ne authors vould like to take this opportunity to thank several individuals for their valuable help. Thanks are due to: A. A. F. Ansari for his performance of steady state thermal-hydraulics calculations; R. J.

Cacciapouti and F. A. Woehlke for acting in behalf of M. A. Sironen during preparation of the final report; B. G. Baharynejad and W. P. Morse for their I aid in preparing figures and performing supporting calculations; D. L. Nichols and S. M. Henchey for preparing the draf t manuscript; and F. C. Beers and the Word Processing Center for preparation of the final teenuscript.

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1.0 INTRODUCTION

An analysis of a turbine trip without bypass event is perforced for the Peach Bottom Atomic Power Station, Unit 2. This analysis was requested by the United States Nuclear Pegulatory Commission to aid in its review of Yankee Atomic Electric Company's BWR analysis methods.

I The analysis employs the lattice physics, steady state physics, transient physics, and system transient methods described in References 1-4.

The specific models used sre described in Section 2. The primary results of the analysis are transient predictions of reactor neutron power and core pressure. These results along with comparisons to the results of other workers [5] are presented in Section 3. Conclusions regarding the analysis are given in Section 4.

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I 2.0 METHODOLOGY EMPLOYED l l

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L 2.1 Steady State Physics h.

For the transient analysis, the steady state calculations included: 1) nodelling the Peach Bottom Unit 2 (PB2) core with SIMULATE [2], 2) depleting PB2 Cycles 1 and 2, and 3) simuleeing the initial conditions of the transient. This final step provided the input for the reactivity calculations and initial conditions for RETRAN.

Based on information from EPRI [6], the PB2 model was formulated.

CASMO [1] was employed to calculate the two group cross sections for three bundle types. These bundle layouts, shown in Figure 2.1, were the most cccmon fuel types in Peach Bottom during Cycles 1 and 2. The cross section data in table format were input to SIMULATE along with a quarter core Cycle 1 loading I pattern. The model of Cycle 1 was then depleted to EOC using the Haling option and was shuffled into the quarter core loading pattern of Cycle 2.

Finally, the Cycle 2 model was depleted to EOC with the Haling option. Both loading patterns are shown in Figure 2.2. The EOC2 calculations provided the transient physics base state case with its exposure distribution and void history. There were two basic inconsistencies between the plant operation and the SIMULATE model: 1) The plant did not have the quarter symmetric loading pattern as was specified in the model. 2) By using the Haling depletion l option, the model assumed that all rods were out (ARO) at EOC and that each i

• I cycle.

l The initial conditions for the transient as described in Feference 5 were input to the SIMULATE Peach Bottom model (See Table 2.1) . This case was restarted from the EOC2 case ano r.he resulting core average axial power i distribution, shown in Figure 2.3 was similar to that case's Haling shape.

1 The goal was to provide an initial condition which was consistent with GE's and BNL's calculations. The only known measure for this consistency was the core average axial power distribution. Comparing the three power

- distributions, SIMULATE reactivity production was weaker in the bottom of the core. To adjust the SIMULAM model, a negative thermal absorber which varied i

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E I axially was added to the core. Figure 2.4 shows the emount of absorber needed to obtain an axial power distribution similar to GE's and BNL's. The result e t average axial power distribution is compared to those of the other workers in Figure 2.5. This case was used as input for the reactivity calculations.

2.2 Transient Physics Reactivity and kinetics data required as input to the RETRAN model were I generated in accordance with the methods detailed in the Trepsient Core Physics Report [31. The pretransient conditions and confi:::cration were as detailed in Reference 5; the base state SIMULATE model at the pretransient conditions used in the generation of reactivity and kinetics data was created by the steady state physics analysis as presented in Section 2.1.

RETI<AN data was spt .ifically generated for a single 12 region and 12 volmne active core channel model. Feedback reactivity data as described in Peference 3, consisted of volume fuel temperature, volume moderator density l

and volume relative poderator density reactivity functions. The procedure for the generation of these reactivity functions is detailed in Figures 2.1 and 2.2 of the given reference. These generated functions are analogous to those graphically presented in Figures 3.7, 3.10 and 3.11 of said report. The scram reactivity curve was generated by the reported procedure detailed in Figure 2.3 at base seate conditions. This scram curve is provided as Figure 2.6 in this report. Kinetics parameters - effective delayed neutron fraction, precursor cecay e s tante, and prompt neutron generation time - were I

calculated at pret re.usient conditions.

In order to characterize a core reactor state, core average reactivi ty coefficients are calculatcJ. These coefficients are not intended for use in the transient analysis, but provide indices which may be used for comparative purposes. Table 2.2 provides the characterization of the Peach Bottom core at the pr3 transient conditions. In addition t> the core average reactivity coefficients, the axial shape index, the effective delayed neutron fraction, and prompt neutron generation time characterize the core.

I I 2.3 Core Wide Transient Analysis Model "Ihe model used for the licensing transient simulation is essentially the same as the model used to perform the Peach Bottom turbine trip test simulations and is described in Section 3.1.1 of Reference 4. Two minor changes were made to the Peach Bottom model to make it even more consistent with the Vermont Yankee nodel [4]. These changes are the following: 1) the bypass system piping up to the valve chest was lumped into the steam line control volume upstream of the turbine stop valve and 2) the recireciation I system junction inertias were recalculated in a manner consistent with the Vermont Yankee model. The setpoints and flow capaciths of the safety reliet and safety valves are based on the information provided in the Brookhaven report I5).

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TABLE 2.1 PEACH BOTTDM UNIT 2 INITIAL CONDITIONS i Feactor Power (Wt) 3440.4 Core Flow Rete (M1b/hr) 101.0 t

Core Pressure (psia) 1050.0 Core Inlet Subcooling (Blu/lb) 28.9 Core Average Exposure (Wd/ST) 12776.0 Control Rod Inventory 0 [All Rods Out]

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TABLE 2.2 PEACH BOTTOM 2 I .

TRANSIENT PHYSICS PARAMETERS 1

Calculated Parameter Value

• Axial Shape Index(I) -0.1990 Moderator Density Coefficient 23.5

,i (Pressurization) , 4/Au(2)

Pressure = 1055 psia Inlet Dithalpy - 525 Btu /lbm Fuel Troperature Coefficient -0.28 at 1100 f, // F I Ef fective Peleyed .005375 Neutron Fraction Prompt Neutron Generation 42.34 Titre, microseconds lB l

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!!ote s : (1) Axial Shape Index (ASI) = p p l

(2) Au = change in relctive density (percent)

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'I Bundle Iype 1 With 80 mil channels l 2 Fuel Type I - 133 w/o U235 11 112 Fuel Type 2 - 0.71 w/o U235 1222 11222 112221 1111111 I Bundle Type 2 With 80 mil channels I 4 32 315 Fuel Type 1 - 2 93 w/o U235 Fuel Type 2 - 1 94 w/o U235 Fuel Type 3 - 1.6. w/o U235 2111 Fuel Type 4 - 1.33 w/o H235 1 21115 Fuel Type 5 - 2 93 w/o U235 with 3 w/o Gd 023 215111 3211122 Bundle Type 4 With 100 mil channels 4 Fuel Type 0 - Water Rod 32 Fuel Type 1 - 3.01 w/o U235 I 211 2511 Fuel Type 2 - 2.22 v/o U235 Fuel Type 3 - 1.87 w/o U235 21110 Fuel Type 4 - 1.45 w/o U235 211111 Fuel Type 5 - 3 01 w/o U235 with 3 w/o Gd 023 2151115 32111112 I ->

I Figure 2.1 PEACH BOTTOM 2 EUNDLE ITPES I

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Loading Pattern - Cycle 1 I 221122112211212 221122112211212 222222222222222 I 222222222222222 221122112211212 221122112211212 222222222222222 I 2222222222-2222 2211221122112 2211221122222 I 22222222222 2222222222 2211221122 I 22222222 2222222 I Loading Pattern - Cycle 2 22422242'2242242 222222222222242 5 424242424242422 222222222222242 g 224242424242422 g 222222222222242 -

424242424242422 22222222222422 2242424242422 2222222224222 42424242422 I 2222222422 2242424222 44242422 I 22222'2 I

Figure 2.2 PEACH BOTTOM 2 SIMULATE INPUT DATA I

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3.0 ANALYSIS OF TURBINE TRIP W' THOUT BYPASS EVENT E

3.1 Initial Conditions r

l In initializing the model, we tried to make the initial state as consistent as possible with the conditions described in the Brookhaven report

{ 5] while still employing Yankee Atomic Electric Company methods. The core  ;

axial power distribution is based on the 3-D SIMULATE prediction (Section 2.1). The bypass flow is based on a FIBWR [7] prediction consistent with the SIMULATE power distribution. Core inlet enthalpy is set so that the amount of l carryunder from the steam separators and the quality in the liquid region outside the separators is as close to zero as possible. This is done to maximize the initial pressurization rare. A summary of the initial operating state is provlJed in Table 3.1.

3.2 Analysis Results l

The transient is initiated by a rapid closure (0.1 sec. closing time) of the turbine stop valves. It is assumed that the steam bypass valves, which normally open to relieve pressure, remain closed. A reactor protection system scram signal is generated by the trbine stop valve closure switches.

Control rod drive motion is assumed to occur 0.27 seconds af ter the start of turbine stop valve motion. Predictions of the system parameters of main interest are shown in Figures 3.1 through 3.6.

I 3.3 Compariscns to Pesults of Other Workers This section presents comparisons of predictions for the described transient between the system models of Yankee Atomic Electric Company (YAEC),

General Electric (GE) and Brookhaven National Laboratory (BNL) . These comparisons are made to aid the Nuclear Regulatory Commission in its evaluation of YAEC methods and do not constitute a critique of either worker's methodology. The GE and BNL results presented were obtained by manual scaling from figures in De ference 5. Comparisons of neutron power, core average heat B flux, core pressure, and core inlet flow are presented in Figures 3.7 through 3.lb. In general, the YAEC results indicate a more severe transient than the results of either GE or BNL.

lI TAoLE 3.1 l

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SUMMARY

OF SYSTEM TRA6SIENT cf0 DEL INITIAL CONDITIONS lI Core Thermal Power (MWth) 3441.2 Turbine Steam Flow (% NBR) 105.0 Total Core Flow (10 61bm/hr) 102.5

> Core Bypass Flow (10 16 bra /hr) 7.3 Core Inlet Enthalpy (Btu /lbe) 521.7 Steam Dome Pressure (psia) 1034.0 Turbine Inlet Pressure (psia) 983.3 Total Recirculation Flow 10 61bm/hr) 35.6 Core Plate Differential Pressure (psi) 18.7 Average Fuel Cap Cot.luctance (Btu /hr-f t2 -F) 1000.0 Ncrrow Range Water Level (in) 29.0 I

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4.0 CONCLUSION

S The methods used in analyzing the transient here are the same as those used in our recent analysis of the Vermont Yankee Nuclear Power Station ,

[8] except that no artificial adjustments to the 3-D simulator input data were made in the Vermont Yankee t aalysis. Comparison to the results of other workers showed similar tr3nds with the largest difference being in the neutron power prediction. Here, the YAEC simulation predicts a larger amount of energy release than the other two workers. This is evidenced in the YAEC simulation's prediction of the initial peak in core average heat flux, which is higher than that of the other workers. Not knowing all the input data used by the other workers, it is difficult to conclude the precise reasens for the dif ferences in the predictions.

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I REFERENCES

1. E. E. Pilat, Methods for the Analysis of Boiling Water Reactors Lattice I- Physics, YAEC-1232, December 1980.

2. D. M. VerPlanck, Methods for the Analysis of Boiling Water Reactors Steady State Core Physics, YAEC-1238, March 1981.

3. J. M. Holzer, Methods for the Analysis of Boiling Water Reactors Trsnsient Core Physics, YAEC-1239P, August 1981.

4. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiling I Water Reactors: A Systems Transiert Analysis Model (RETRAN), YJEC-1233, April 1981. (

M. S. Lu, et al. , Analysis of Licensing rasis Transients for a BWR/4, I

5.

BNL-NUREG-26684, September 1979.

6. General Electric, Core Design and Operating Data for Cycles 1 and 2 of Peach Bottom 2, EPRI NP-563, June 1978.

7. A. A. F. Ansari, Methods for the Analysis of Boiling Water Reactors:

I Steady-State Core Flow Distribution Code (FIBWR), YAEC-1234, December 1980.

8. Yankee Atomic Electric Company, Vermont Yankee Cycle 9 Core Performance Analysis, YAEC-1275, August 1981.

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