ML20082Q827

Evaluation of 9 by 9 Fuel Impact on Stability of Susquehanna-2 at End of Cycle 4

ML20082Q827
Person / Time
Site:	Susquehanna
Issue date:	08/31/1991
From:	Damiano B OAK RIDGE NATIONAL LABORATORY
To:	Huang T Office of Nuclear Reactor Regulation
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CON-FIN-L-1697 NUDOCS 9109130082
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, ENCt.050RE ORNL/NRC/LTR 91/12 Contract Program: BWR Stability (Ll697, P2)

Subject of Document: Evaluation of 9 by 9 Fuel Impact on the Stability of Susquehanna.2 a the End of Cycle 4 Type of Document: Technical Evaluation Report Authors: Brian Damiaro Jos6 March-Leuba Date of Document: August 1991 NRC Monitor: T, L. Iluang, Office of Nuclear Reactor Regulation Prepared for-U.S. Nuclear Regulatory Commission l

Office of Nuclear Reactor Regulation  !

-under '

DOE Interagency Agreement 1886-8946-5 A NRC FIN No. Ll697, Project 2  !

Prepared by Instmment and Controls Division OAK RIDGE NATIONAL LABORATORY operated by.

MARTIN MARIETTA ENERGY SYSTEMS, INC.

for the U.S. DEPARTMENT OF ENERGY under Contract No. DE-AC05-840R21400 g ._ P4

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yldji3 009 ;L

INTRODUCTION This report documents the results of our analysis of noise data recorded at the Susquehanna-2 boiling water reactor (BWR) at the end of Cycle 4. The most significant result of this analysis is an estimate of the reactor stability at the particular operating point at which the data was recorded.

During Cycle 4 the Susquehanna 2 reactor used a full core of Advanced Nuclear Fuels (ANF) 9 by 0 fuel. Previous test data have indicated that the 9 by 9 fuel has little significant effect on the reactor stability?

The present experimental program began in 1986 to detennine whether loading 9 by 9 fuel elements could significantly destabilize a reactor. The 9 by 9 fuel has a faster fuel temperature response (due to its smaller thermal inertia) and a smaller (1. e. less negative) void reactivity coefficient compared to conventional 8 by 8 fuel. A faster fuel temperature response decreases the reactor stability while a smaller void reactivity coefficient tends to stabilize the reactor. It is not obvious how these two competing effects would combine to affect the reactor stability, thus, a series of measurements were made, beginning dunng Cycle 2, to experimentally determine the effect of the 9 by 9 fuel on reactor stability.

TEST DATA AND ANALYSIS RESULTS The test data were recordd on January 4,1991 by Susquehanna personnel using the General-Electric Trarsient Analysis Recorder (GETARS) data acquisition system. The reactor was operating at approximately 64% power with a total core flow of 45.8 M lb/h. Appendix A lists in detail the reactor conditions during recording. The data, which consisted of average power range mordtor -

(APRM), core flow, and core pressure signals, were sent in digital form on magnetic tape to the Oak

P.id;; .M

:!=:! L:b :t :/ (GKt4L) for analysis. Tisc scunded signals and d cir units are listed in l Table 1. -

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1

Table 1. List of signals and their units.

Channel Desenntion Units 1 APRM A  % Nominal 2 APRhtB  % Nominal 3 APRM-C  % Nominal 4 APRM D  % Nominal 5 APRM-E  % Nominal 6 APRM-F  % Nominal 7 Core Flow Mlb/h _

8 N.R. Pressure psi The data were analyzed in the frequency damain to produce noise descripton; such as power spectral densities O'SDs), coherences, autocorrelation functions, and transfer functions. The stability was then calculated from these descriptors using the ORNL stability analysis methodology.' The main results of the analysis are summarized in Table 2.

Figuxs 1 and 2 show the PSD and the autocorrelation function calculated for the APRM A signal. The rectits shown in these figures are typical of all the APRM data used in this analysis.

DISCUSSION OF RESULTS The results of this analysis show that the Susquehanna-2 reactor was very stable (i.e. decay ratio less than 0.3) at the end of Cycle 4. The relatively stable condition is indicated by the bmad peak in the PSD at appmximately 0.4 H: and by the rapid damping shown in the autocorrelation function.

Table 3 shows results fmm previous stability analyses performed oy ORNL for the Susquehanna-2 reactor. These analyses inchide data frnm Cple3 2,2, M e in mMeh de em contained 33%,66%, and 100% 9 by 9 fuel. Similar pown and Gow conditions were used in each 2

i Table 2. Decay ratios and natural oscillation frequencies estimated from noise data. Susquehanna-2, EOC 4.

Decay Oscillation Channel Description Ratio Frequency (liz) 1 APRM A 0.2410.06 0.6810.08 2 APRM B 0.2210.05 0.6810.08 3 APRM C 0.2210.06 0.6S10.08 4 APRM D 0.2310N 0.6810.07 5 APRM E 0.2210.05 0.6810.10 6 APRM-F 0.2310.05 0.f Td9.07 test. Comparison of the results in Tables 2 and 3 shows that the Susquehanna-2 reactor had the smallest decay ratio and highest oscillation frequency at the end of Cycle 4. The local power range j monitor (LPRM) data given in Appendix A show that the reactor had a top-peaked axial power shape when the data was collected. This is the most likely explanation for the low decay ratio and high oscillation frequency. BWR stability is heavily influenced by the axial power shape; bottom peaked axial power shapes are destabilizing and top-peaked axial power shapes are stabilizing. Furthermore, the oscillation frequency is inversely proportional to resident time 'or steam bubbles in the core. Top-peaked power shapes result in the average axial position of steam bubble fonnation being shifted _

upward. Thus, the steam bubbles, on average, have a shorter distance to travel before leaving the core when the axial power shape is top-peaked. Since the total core flow (and thus the flow velocity) for this analysis is approximately the same as in previous analyses, the shoner distance traveled by the steam bubbles translates into a shorter average residence time for steam bubbles in the core and in a correspondingly higher oscillation frequency.

These results agree with our previous conclusions that the 9 by 9 fuel does not produce major changes in stability behavior compared to BWRs loaded with standard 8 by 8 fuel.

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Table 3. Reactor conditions, decay ratios, and oscillation frequencies from previous analyses of Susquehanna 2 GETARS data" 9 by 9 elements Power Flow Oscillation Freq.

Cycle loaded (%) Date (% of nominal) (Mlb/hr) Decay Ratio (Hz) 2(TLO)* 33 2 NOV 86 61 46.7 0.3310.03 0.3910.02 2(SLO)* 33 9 NOV 86 56 43.9 0.3710.02 0.3410.02 3(TLO)* 66 23 JUL 88 60 46.0 0,4810.05 0.4810.04 4(TLO)* Ift) 8 DEC 89 63 44.6 0.2710.07 0.2710.01

' Two loop operation, minimum recirculation pump speed.

• Single loop operation.

REFERENCES

1. " Grand Gulf l and Susquehanna 2 Stability Tests " L March-Leuba and D. N. Fry, Oak Ridge National Laboratory Leuer Report ORhLfNRC/LTR-87/01, April 1987.

2. "Susquehanna-2 Stability Measurements During Cycle 3," J. March-Leuba and B. Damiano.

Oak Ridge National Laboratory Lener Report ORNL/NRC/LTR 9073, January 1990.

3. " Evaluation of 9 by 9 Fuel Impact on the Stability of Susquehanna 2 durmg Cycles 2,3, and 4," B. Damiano and J. March Leuba,_ Oak Ridge National Laboratory Letter Report ORMJNRC/LTR-91/ll, August 1991.

4. " Development of a Real-Time Stability Measurement System for Boiling Water Reactors,"

J. March-Leuba and W. T. King, Trans. Am. Nucl. Soc. 54 370-371 June 1987, t

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Figure 1. Power spectral density of APRM-A.

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Figure 2. Autocorrelation function of APRM-A, 5

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APPENDIX A Test point operating conditions MAME OR E9CRIPT10M VEUE

?0!Ifi !D ( ALL R0WS IN 1848 EL TDPDtATWE1 Ill KG F)

~ _ .. -.

WT TOTAL CORE FLOW UE3 IN N.B. 45.814 at. ACTIVE / TOTE CGIE R0W FRAC 0.t0000 NFP51 E ACTOR PRESSAE (PS!A) 967.20 Ileff COE TElgW. Pep (Inff) l 2077.4 1 80HS COREIlt.ElSUIC00 LING (ITU/LI) -37.525 IOGEL COE EENT IICRDOT (leMT) 42.595 MM551 m EA01EA)-CHAN 01 (N) 63.562

-MM553 m KABIEC)-OWN 02 (M) 62.875 MM555 m EADIED-CHAN 03 (M) 63.750 MM552 - M EA015(I)-OIAll 64 (N) 63.612 MM554 APen READIEB)-CNeil 45 (N) 63.675 MM556 APW( EA01N5(F1-CHAN 06 (N) 63.062 eMRAG CTF CALC. (HIT ML.1-APRM) 0.00000 NJP51 ItEACTR CME PES. iRIP (Pfill 2.7191 EF51tK1 CRS FLOW (ESAgt) 0.31594E-41 EF515K2 CLIAIRP L0lr RDW (18/2) 0.10430 E152 CLEmer LOOP llLET TDP (EF) 526.27 NLT51 CLEAlte L0lr EXIT TDP (EE) 438.75 WL01(DC5) EACTR linter LEWL (IlOES) 34 015 WF11 ItEACT R STDul FLaf- B.2619 GilJ02 enes MIENTR PRER (IGE) ~292 50 l DFWA ~ FV Rap AelRITS M (13/18) 2.7354 i DFWB FV RW 3,13179 E (184R) 2.7165

, DFWC FV ROW Ceunift M (ESAR) 2.6099

! WJ51 RECIRC P(BP A POER (Ill) 0 20172

\

NRJ52 RECIRC PtfP I POER - (141) 0.26892 l NFF52 FEDilATR Rh A (ES/2) 2.7392 l NFF53 FEDilATER Rm I (E3/10t) 2.7154

MFF54 FEDWATR RE C (ILSAS) 2.6090 NIT 51 FW TCP 1 eMAICI A (EF) 347.45 wais? rv ms 2 ;gsjagn! g 4,nggq

s ,94 N3T53 ni Tw 1.amars a enEno 1ss,07 l Mil 54 FV TDP 2 IANOI I (EU) 344.h l NIT 55 FV TDP 1 IRNot C (E9') 345.04 l NtT56 FV TDP 2 etR210( C (Ef f) 344.37 6

- ~ .- - .- . .. .-. . -.- -

y ,

-M NRF518K3 RECIRC RWe A1 (UM) 5.0833 WF33tK3 4 CIRC FLWe A2 (UM) 4.9419

• F57tK3 RECIRC RWe 31 (UM) s.6464

• F54tK3 ECIRC RWe 32 (UM) 6.5313 WT51 RECIRC TDre A1 (KK) 495.01 2 152 RECIRC TDre A2 (KF) 511.01 2 733 RECIRC TDre 31 (KF) 506.41

• T54 RECIRC TUF, B2 (KF) 508 54 20080, GOGATR DOST INCR. Ut4E) 6.0000 at EAT RAN UtifMRE) ._ 7 1022 l + +

i t XTG INPUTS AND SCM BATA DIT FOR SUSOElWenn-2 +

+

t

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l 8WIE OR DDGIFT!W VALUE L

- POINT ID (ALL RR$ IN UAgt ALL TDFUtATimES IN KG F)

• WTSUB CORE RW FROM FlmCTION F4 l 45 810 i WJ51 WT FROR J.P. OR lirVT (UAct) 45.814 WB TOTAL RECIRC RW 11 635

• WTRA6 Catt FLOW RA6 L

2.0000 SCRS CDNTRS.R08IDSITY 0.36737E41 tutDSYM CONTRR. R09 STIOETRT Ft.AG 0.00000 7

l .'

1 z- .

+ t.PRfl READIN65 - lMCALIBRAID FOR SU$olO N M 2+

t (PROCESS COWVTD CD001mTIS) +

(1657) (2457) (3257) (4057) 14.7 17.8 18.4 17.4 20.8 26.7 26.6 26 2 19.8 3.0 26.4 25.6 12.7 20.5 17.7 16.5 t0649) (1649) (2449) (3249) (4049) (46491 16.3 -0.0 26.7 27.4 27.1 27 5 23.7 35.0 39.3 -0.1 40.6 29.8 20.8 30.8 31.2 32.2 30.4 26.5 12.0 20.7 25.9 25.4 22.2 15.7

'0641) (1641) (2441) (3241) (4041) (4841) (5641) 23 5 28.9 31.8 29.8 30.3 26.8 17.9 34.8 42.3 39.1 35.2 34.6 40.7 28.e 32.4 30.4 29.1 27.7 28.6 30.6 26..

22.4 20.7 21.1 20.0 20.6 20.2 15.8 (0933) (1633) (2433) (3233) (4033) (4833) (5633) 73 2 29.0 31.6 31.2 30.0 27.6 18.4 34.0 39.7 37.7 16.6 36 6 43.8 27.4 32.1 32.9 -0.4 27.5 29.4 33 3 27.2 22.2 22.7 18.2 21 9 21.2 25.1 16.5 (0t25) (1625) (2425) (3225) (4025) (4825) (5625) 23.2 29.4 31.7 29.9 29.8 26.5 17.9 35.4 42.1 36 5 35.9 37.1 42.3 27.2 33.0 30.6 27.3 27.0 28 5 32.0 29.8 25.3 22 1 19.6 18.3 -0 1 24.2 19.1 (0817) (1617) (2417) (3217) (4017) (4817) (5617) 20.7 26 3 28.8 21 2 27.9 23.7 14.2 31.6 39.4 40.7 36.7 38.3 36.3 21.7 29.6 33.0 30.0 29.7 28.5 31.5 20 5 17.2 23.3 23.0 22.7 20.8 19.5 11 2 (1609) (2409) (3209) (4009) (4809) 20.2 22.5 23 2 22.8 16.1 30.5 33.5 k.7 0.2 22.2

• R U.7 32.0 32.8 20.7 18.7 27.9 29.2 44.2 11.7 8

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ORNL/NRC/LTR 91/12 INTERNAL DISTRIBUTION

1. L. II. Bell 10. Central Research Library

2. N. E. Clapp 11. Y 12 Document Reference

3. B. Damiano Department

4. B. G. Eads 12. I&C IPC

5. D. N. Fry 13. Laboratory Records Department

6. J. March-Leuba 14 Laboratory Records, ORNL-RC

7. O. B. Morgan 15. ORNL Patent Section

8. P. F. McCrea (Advisor)

9. J. b. Ball (Advisor)

EXTERNAL DISTRIBUTION

16. S. Bajwa, Division of Engineer and Systems Technology, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission, WF18E23 Washington DC 20555

17. Amira Gill, Division of Engineering and Systems Tectiology, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, WF1-8E23, Washington DC 2055:i

18. T. L. Huang, Division of Engineering and Systems Technology, Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission, WF18E23, Washington DC 20555

19. L. E. Phillips, Division of Engineering and Systems Technology, Office of Nuclear Reactot Regulation, U. S. Nuclear Regulatory Commission, WF1-8E23, Washington DC 20555

20. 11. J. Richings, Division of Engineering and Systems Technology, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission WF1-8E23, Washington DC 20555

21. NRC Central File 9

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