ML19344B170

Supplemental Reload Licensing Submittal for VT Yankee Nuclear Power Station Reload 7

ML19344B170
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	07/30/1980
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML19344B169	List: ML19344B168 ML19344B170
References
Y1003J01A02, Y1003J1A2, NUDOCS 8008250693
Download: ML19344B170 (29)

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_

l Y1003J01A02 July 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR VERMONT YANKEE NUCLEAR POWER STATION RELOAD NO. 7 3

/

Y1003J01A02 IMPORTANT NOTICE REGARDING CONTENTS OF'THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Vermont Yankee Nuclear Power Corporation for VY's use with the U.S. Nuclear Regulatory Connission (USNRC) for amending VY's operating license of the Vemont Yankee Nuclear Pouer Station.

The infomation con-tained in this report is believed by General Electric to be an accurate and true representation of the facts knaan, obtained or i

provided to General Electric at the time this report uas prepared.

1 The only undertaking of the General Electric Company respecting information in this document are contained in the contract between Vemont Yankee Nuclear Power Corporation and General Electric Com-pany for nuclear fuel and related services for the nuclear system for Vermont Yankee Nuclear Power Station, dated July 11, 1975, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended is not authorized; and with respect to any such unauthor-ined use, neither General Electric Company nor any of the contribu-tors to this document makes any representation or carranty (e: press or in: plied) as to the ccmpleteness, accuracy or usefulness of the i

information contained in thia document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liabilitji or damage of any kind which may result frcn such use of such informati.cn.

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Y1003J01A02 s

CONTENTS fagg 1.

PLANT-UNIQUE ITEMS 1

2.

RELOAD FUEL BUNDLES 1

3.

REFERENCE CORE LOADING PATTERN 1

4.

CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20'C 1

5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY 2

6.

RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS 2

7.

RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS 2

8.

SELECTED MARGIN IMPROVEENT OPTIONS 2

9.

CORE-WIDE TRANSIENT ANALYSIS RESULTS 3

10.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUENT FAILURE)

TRANSIENT SLHMARY 3

11.

OPERATING MCPR LIMIT 3

9 12.

OVERPRESSURIZATION ANALYSIS SLMfARY 4

13.

STABILITY ANALYSIS RESULTS 4

14.

LOSS-OF-COOIANT ACCIDENT RESULTS 4

15.

LOADING ERROR RESULTS 4

16.

CONTROL ROD DROP ANALYSIS RESULTS 4

i -

APPENDICES A.

A-1 B.

MARGIN-TO-SPRING SAFETY V.d.VES B-1 I

iii/iv

~

l Y1003J01A02 ILLUSTRATIONS Figure Title Pare 1

Core Loading Pattern 1

2a Scram Reactivity and CRD Specifications, EOC8 6

2b Scram Reactivity and CRD Specifications, EOC8-1 7

2c Scram Reactivity and CRD Specifications, EOC8-2

'8 3a Generator Load Rejection, EOC8 9

1 3b Generator Load Rejection, Without Bypass, EOC8-1 GWd/t 10 3c Generator Load Rejection, Without Bypass, EOC8-2 GWd/t 11 4

Loss of 100*F Feedwater Heating 12 5

Feedwater Controller Failure 13 6

Limiting Rod Pattern 14 7

MSIV Closure, Flux Scram 15 8

Reactor Core Decay Ratio vs Power 16 l

4 i

/

r 4

v/vi

Y1003J01A02 1.

PLANT-UNIQUE ITEMS (1.0)*

Appendix A: Operating MCPR Limit Exposure-dependent limits (EOC-2 GWd/t and EOC-1 GWd/t).

Appendix B: New Bundle Loading Error Event Analysis Procedures Appendix C: Margin to Opening of Unpiped Safety Valves 2.

RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1, and 4.0)1 Fuel Type Number Number Drilled Irradiated 8DB274L 24 24 Irradiated 8DB274H 96 96 Irradiated 8DB219L 12 12 Irradiated 8DPB289 60 60 Irradiated P8DPB289 96 96 New P8DPB289 80 80 Total 368 368 3.

REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 15.75 GWd/t. Assumed reload cycle exposure:

16.80 GWd/t. Core loading pattern: Figure 1.

4.

CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)

BOC keff Uncontrolled 1.105 Fully Controlled 0.948 Strongest Control Rod Out 0.981 R, Maximum Increase in Cold Core Reactivity with 0.005 Exposure Into Cycle, Ak

• ( ) refers to areas of discussion in Reference 1.

Reference 1: " General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-240ll-P-A, August 1979, 1

~.

Y1003J01A02 5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (t.k) ppm (20*C, Xenon Free) 800 0.057 6.

RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)

EOC8 EOC8-1 EOC8-2

Void Coefficient N/A* (-c/% Rg) 8.03/10.03 8.58/10.73 8.52/10.65 Void Fraction (%)

40.13 40.13 40.13 Doppler Coefficient N/A (-c/*F) 0.728/0.217 0.224/0.213 0.218/0.207 Average Fuel Temperature (*F) 1342 1342 1342 Scram Worth N/A (-S) 36.27/29.02 35.49/28.40 34.27/27.42 Scras Reactivity versus Ti=e Figure 2a Figure 2b Figure 2c 7.

RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAv.ETERS (5.2)

EOC8 EOC8-1 GVd /t EOC8-2 GWd/t Exposure 8x8 8x8R P8x8R 8x8 8x8R P8x8R 8x8 8x8R P8x8R Peaking factors 1.22 1.20 1.20 1.22 1.20 1.20 1.22 1.20 1.20 (local, radial 1.36 1.50 1.48 1.40 1.54 1.52 1.44 1.58 1.58 axial) 1.40 1.40 1.40 1.40 1.40 1.40 1.40 1.40 1.40 R-Factor 1.098 1.052 1.052 1.098 1.052 1.052 1.098 1.052 1.052 Bundle Power 5.763 6.346 6.275 5.911 6.495 6.426 6.105 6.680 6.688 (MWt)

Bundle Flov 109.3 110.0 110.9 108.2 109.0 110.0 106.9 107.8 108.2 (103 lb/hr)

Initial MCPR 1.29 1.29 1.31 1.26 1.26 1.28 1.21 1.21 1.21 8.

SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2) j Exposure-dependent li=its: EOC8-1 GWd/t to EOC8 EOC8-2 GWd/t to EOC8-1 GWd/t BOC8 to EOC8 ' GWd/t

• N = Nuclear Input Data A = Used in Transient Analysis 2

Y1003J01A02 9.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Power Core Flow h

Q/A

' SL v

ACPR Plant Transient Exposure (t)

(2)

(t N3R)' (2 N!R) (psfR) (psig) 8x8/8x8R/P8x8R Ressense Load Rejection EOC8 104.5 100 295 119 1242 1261 0.22/0.22/0.24 Figure 3a without Bypass EOCS-1 104.5 100 265 116 1237 1254 0.19/0.19/0.21 rigure 3b CWd/t EOC8-2 104.5 100 172 109 1210 1228 0.09/0.09/0.09 Figure 3c CWJ/t i

Loss of 100*F BOC8 to 104.5 100 123 122 0.14/0.14/0.14 Figure &

W Heater EOC8 Feedwater BOC8 to 104.5 100.

113 110 1022 1065 0.05/0.05/0.05 Figure 5 Controller EOC8 Failure i

10.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1)

Rod Block Rod Position Limiting Reading (%)

(Feet Withdrawn) 8x8/8x8R/P8x3R 8x8/8x8R and P8x8R Ro.1 Pattern i

104 4.0 0.09/0.12/0.13 12.0/14.0 Figure 6 j

105*

4.0 0.09/0.12/0.13 12.0/14.0 106 4.5 0.10/0.14/0.15 12.4/14.5 107 5.0 0.11/0.15/0.16 12.8/15.1 108 6.0 0.14/0.19/0.20 13.3/16.1 109 7.0 0.18/0.22/0.23 13.6/16.7 o

i 11.

OPERATING MCPR LIMIT (5.2)

See Appendix A.

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3

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Y1003J01A02 12.

OVERPRESSURIZATION ANALYSIS SL1HMARY (5.3)

Power Core Flow si v

Plant Transient

(%)

(%)

(psig)

(psig)

Response

MSIV Closure 104.5 100 1267 1290 Figure 7 (Flux Scram) 13.

STABILITY ANALYSIS RESULTS (5.3)

Decay Ratio: Figure 8 Reactor Core Stability:

Decay Ratio, x /*

0.84 2 o (Natural Circulation - 105%

Rod Line)

Channel Hydrodynamic Performance Decay Ratio (Natural Circulation -

105% Rod Line) 8x8 channel 0.39 8x8R channel 0.29 P8x3R channel 0.29 14.

LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)

Fuel type P8DPB289 was introduced in Reload 6.

15.

LOADING ERROR RESULTS (5.5.4)

See Appendix A.

16.

CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Maximum incremental rod worth 0.76% Ak.

4

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Core Loading Pattern 5

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2.

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3.

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Limiting Rod Pattern 14

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T1003J01A02 1.2 ULTIMATE PERFORMANCE LIMIT 1.0 02

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O 20 40 60 80 100 PERCENT POWER Figure 8.

Reactor Core Decay Ratio vs Power 16

Y1003J1A02 APPENDIX A OPERATING MCPR LIMIT If, during steady-state operation, the offgas activity as measured at the SJAE's exceeds 236,000 pCi/see for fifteen (15) minutes of 1.18 Ci/sec for one (1) minute, the operating MCPR limit shall be as follows:

MCPR Operating Limit Exposure Range 8x8 8x8R P8x8R E08-1 GWd/t to EOC8 1.29 1.29 1.31 EOC8-2 GWd/t to EOC8-1 GWd/t 1.26 1.26 1.28 BOC8 to EOC8-2 GWd/t 1.21 1.26 1.26 If, during steady-state operation, the offgas activity as measured at the SJAE's is less than specified above, the operating limit shall be as follows:

MCPR Operating Limit Exposure Range 8x8 8x8R_

P8x8R EOC8-1 GWd/t to EOC8 1.29 1.29 1.31 EOC8-2 GWd/t to EOC8-1 CWd/t 1.26 1.26 1.23 BOC8 to EOC8-2 GWd/t 1.21 1.21 1.21 A-1/A-2

l l

l Y1003J01A02 APPENDIX B LOADING ERROR RESULTS (NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES)

The bundle loading error analyses results presented in Section 15 in this supplement are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error events. The use of these new analy-ces procedures is discussed below.

B.1 NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error event analysis results presented in this sup-plement are based on the new analysis procedure described and approved in Reference B-1.

This new method of performing the analysis is based on a more accurate detailed analytical model.

The principle difference between the previous analysis procedure and the new l

(

analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel t

l spacer buttons, caused by misorientation, causes the bundle to lean. The effect of the variable water gap is to reduce the power peaking and the R-factor in l

the upper regions of the limiting fuel rod. This results in the calculation L

l of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.

The results of the analysis indicate the P8DPB289 bundle a 17.7 kW/ft LHGR (includes densification spiking penalty of 2.2%) and 0.19 ACPR (includes a 0.02 penalty due to variable water gap R-factor uncertainty) with a minimum CPR of 1.07.

B-1

l Y1003J01A02

]

B.2 NEW ANALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analyses results presented in this t

supplement are based on the new analysis procedure described in Reference A-1.

This new method of performing the analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core.

The use of the statistically corrected Haling analyses procedure indicates that the LHGR is 16.6 and that minimum CPR for mislocated bundles (e.g.,

P8x8R into P8x8R) is greater than the safety limit (1.07) for all exposures throughout cycle 8.

1 o

4 l

l l

I B-2

l Y1003J01A02 REFERENCES B-1 Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, dated May 8,1978.

d B-3/B-4

Y1003J01A02 APPENDIX C MARGIN-TO-SPRING SAFETY VALVES The rationale for changing the basis for providing pressure margin to the spring safety valves is presented in Reference C-1.

This change has been accepted by the NRC (Referen'ce C-2).

On this basis the plant can operate at full power throughout the cycle.

The core response to the limiting anticipated event is given in Table C-1 and Figure C-1 Table C-1 CORE-WIDE TRANSIENT ANALYSIS RESULTS Power Flow s1 y

Plant Transient Exposure

(%)

(%)

(psig)

(psig)

Response

MSIV Closure BOC-EOC 104 100 1162 1183 Figure B-1 Trip Scram REFERENCES C-1.

J. F. Quirk (GE) letter to Olan D. Parr (NRC), " General Electric Licensing Topical Report NEDE-240ll-P-A, ' Generic Reload Fuel App 1f-cation', Appendix D, Second Submittal," dat?d February 28, 1979.

C-2.

Letter, T. A. Ippolito (NRC) to D. L. Peoples (Commonwealth Edison Co.),

enclosing Safety Evaluation Supporting Amendment No. 42 to Operating License No. DPR-25, Dresden Nuclear Power Station, Unit 3, April 16, 1980.

C-1

1 NEUTRON FLUX X VESSEL FTES RISE (PSI) 2 AVE SURFrCE HEAT FLUX 2 SArETT VfLVE FLOW 150.

3 CORE INLF T FLOW 300.

3 REL.lEF VILv(F10W 14 II 01PHS5 VI LYL f LfM_

S 5

6

'E 100.

200.

d 1

E

\\

to 50.

100.

3

~

T M

2 2

O.-[mi-i,

-Q

• I pu n f 2e G(

0.D.

10.

20.

30.

t40.'

O.

10.

20.

30.

tm.

TIME (SEC1 TIME (SEC) 8 n

M 0

l LEVEL (IMH-REF-SEP-SKIRT I V ID REF,'TIVITY O

2 VESSEL S1EAMFLOW 2

ER ' ' ACTIVITY 200*

3 ItMRINE ( ifDMFL OW 3 SCRAM RFi TIVITT g*

4 FLLONRTEF FLOW

,)( TifTfCTIC, TIVITY S

100 0

M

^

e O.

2 1

3 2 M 2 0

-1.

O

\\3 C

g m

L 100.

1-

-2.

1-I 0.

10.

20.

30.

40.

O.

0.6 1.2 1.8

2. t4 TIME (SEC)

TIME ISECl Figure C-1.

MSIV Closure, Position Scram

-