ML20126J611

Supplemental Reload Licensing Submittal for Browns Ferry Unit 1 Reload 4(Cycle 5)

ML20126J611
Person / Time
Site:	Browns Ferry
Issue date:	03/31/1981
From:	Engel R, Hilf C GENERAL ELECTRIC CO.
To:
Shared Package
ML20126J605	List: ML20126J600 ML20126J607 ML20126J611 ML20126J615
References
Y1003J01A19, Y1003J01A19-R0, Y1003J1A19, Y1003J1A19-R, NUDOCS 8105010364
Download: ML20126J611 (28)

Text

.'O . v1oo3aot^19 Rev. O Class I March 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BROWNS FERRY NUCLEAR PLANT UNIT 1 RELOAD NO. 4 (CYCLE 5)

Prepared by: [

C. L. hilf I

Approved by:  !

R. E. Engelf Manager '  !

Reload Fuel Licensing l

NUCLE AR POWER SYSTEMS DIVISION e GENER AL E LECTRIC COMPANY SAN JOSE, CALIFORNI A 95125

%Io50/034y GENER AL h ELECTRIC J

Y1003J01A19 Rev. O hI f

ItiPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending TVA's operating license of the Browns Ferry Nuclear Unit

1. The information contained in this report is believed by General Electric to be an accurate and true representation 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 information in this document are contained in the contract between the Tennessee Valley Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Browns Ferry Nuclear Plant Units 1 and 2, dated June 17,1966, 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 unauthorized use, neither 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.

11 I _

Y1003J01A19 Rev. 0

1. PLANT-UNIQUE ITEMS (1.0)

• Plant Parameter Differences - Appendix A Safety Valves l Safety / Relief Valves GETAB Initial Conditions Initial MCPR Fuel Loading Error LHGR ,

I l

ODYN Code for Transient Analyses - Appendix B I New Bundle Loading Error Event Analyses Procedure - Appendix C Densification Power Spiking - Appendix D Lead Teat Assemblies (LTAs) - Appendix E

2. RELOAD FUEL BUNDLES (1.0, 2.0, 3. 3.1 AND 4.0) 1 Fuel Type Number Number Drilled Irradiated, Reload 1 8DB274H 8 8 Irradiated, Reload 1 8DB274L 112 112 Irradiated, Reload 2 8DRB265L 84 84 f

Irradiated, Reload 2 SDRB265H 68 68  !

l Irradiated, Reload 3 P8DRB284L 232 232 )

New P8DRB284L 220 220 New P8DRB265L 36 36 New GLTA-1 2 2 New GLTA-2 2 2 Total 764 764

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end 16,071 mwd /t of cycle:

• ( ) refers to areas of discussion in " General Electric Boilina Water Reactor Generic Reload Fuel Application", August 1979 OREDE-24011-P-A-1).

1

Y1003J01A19 Rev. O Minimum previous cycle core average exposure at end 16,071 mwd /t of cycle from cold shutdown considerations:

Assumed reload cycle core average exposure at end 17,846 mwd /t 1

of cycle: (

Core loading pattern: Figure 1

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1, i.e., 3.3.2.1.2)

BOC k,ff Uncontrolled 1.115 I

Fully Controlled 0.955 Strongest Control Rod Out 0.990 R, Maximum Increase in Cold Core Reactivity with 0.000 Exposure Into Cycle, Ak

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3. 3.2.1. 3) l Shutdown Margin (Ak) ppm (20"C, Xenon Free) 600 0.029

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)

EOC5 Void Coefficient N/A,* (c/% Rg) -7.17/-8.96 Void Fraction 39.79 Doppler Coefficient N/A (c/*F) -0.219/-0.208 Average Fuel Temperature (*F) 1383 Scram Worth N/A ($) -46.31/-37.05 Scram Reactivity Figure 2

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

Y1003J01A19 Rev. 0

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

EOC5 ,

8x8 8x8R P8x8R Peaking factors (local, 1.22 1.20 1.20 radial and axial) 1.40 1.56 1.54 1.40 1.40 1.40 R-Factor 1.098 1.051 1.051 Bundle Power (MWt) 5.913 6.571 6.495

, Bundle Flcs (10 lb/hr) 107.2 108.0 108,4 Initial MCPR 1.25 1.25 1.26

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

Recirculation Pump Trip 9 CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Power Core Flow h Q/A SL v na 6CPR Exposure Flant dra r.a lan t (t) (%) ,({J BR},, (t NBR}, M Q (PHIC) Jd Ox8R P6xBR lteeponse Generat<>r EOC5 104.5 100 599 Load kejection 121 1219 1230 0.18 0.18 0.20 Figure 3 witnout Bypass 1 oss of 100*F -

104.5 100 Ieedwater 124 124 1013 1069 0.15 0.15 0.15 Figure 4 He at ing Feedwater EOC5 104.5 100 Controller 367 120 1158 1189 0.15 0.15 0.16 Figure 5 Fallure 3

~ - - - . .

Y1003J01A19 Rev. 0

10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(5.2.1)

Rod

• "E ACPR LHGR** Rod Rod Block- Position Reading (Feet Withdrawn) 8x8_ 8x8R/P8x8R 8x8 8x8R/P8x8R Pattern 4

104 3. 0 0.14 0.09 11.4 13.7 Figure 6 0.11 12.3 14.2 Figure 6 105 3.5 0.16 106* 4.0 0.19 0.14 13.0 15.2 Figure 6 107 4.5 0.22 0.16 13.4 15.9 Figure 6 108 4.5 0.27 0.16 13.4 15.9 Figure 6 109 5.0 0.29 0.19 13.5 16.3 Figure 6 5.5 0.32 0.21 13.6 16.7 Figure 6 110

11. CYCLE MCPR VALUES (5.2, APPENDIX C)

BOCS to EOC5 Option A Option B Pressurization Events 8x8 8x8R P8x8R_ 8x8. 8x8R P8x8R_

Generator Load Rejection 1.30 1.30 1.33 1.22 1.22 1.23 without Dypass Feedwater Controller Failure 1.27 1.27 1.28 1.24 1.24 1.25 Non-Pressurization Events 8x8 8x8R P8x8R Loss of 100*F Feedwatcr 1.22 1.22 1.22 Heating Fuel Loading Error 1.22 1.22 1.22 Rod Withdrawal Error 1.26 1.21 1.21

• Indicates setpoint selected.

• • Includes a 2.2% peaking penalty for fuel densification-4

Y1003J01A19 Rev. 0

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

Power Core Flow s1 v Plant f%) (psig) (psig) Response Transient (%)

MSIV Closure 104.5 100 1238 1272 Figure 7 (Flux Scram)

13. STABILITY RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability Decay Ratio, x2/x0*

• Channel Hydrodynamic Performance Decay Ratio, x2 /*o 8x8 channel 0.39 8x8R/P8x8R channel 0.29

14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)

P8DRB284L, GLTA-1, and GLTA-2 Exposure MAPLHGR PCT Local Oxidation (mwd /t) (kW/ft) (*F) Fraction 200 11.2 1685 0.004

\

1000 11.3 1667 0.003 5000 11.8 1671 0.003 10,000 12.0 1647 0.003 15,000 12.0 1669 0.003 20,000 11.8 1672 0.003 25,000 11.2 1633 0.003 30,000 10.8 1596 0,002 35,000 10.2 1469 0.001 40,000 9.5 1411 0.001 5

Y1003J01A19 Rev. O P8DRB265L Exposure MAPLHGR PCT Local Oxidation

,1kW/ft) (*F) Fraction (tIWd/ t) 200 11.6 1711 0.004 1000 11.6 1700 0.004 5000 12.1 1692 0.003 10,000 12.1 1663 0.003 15,000 12.1 1683 0.003 20,000 11.9 1683 0.003 25,000 11.3 1637 0.003 30,000 10.7 1579 0.002 35,000 10.2 1526 0.002 40,000' 9.6 1463 0.001 l

15. LOADING ERROR RESULTS (5.5.4) i Limiting Event: Rotated Bundle P8DRB284L, MCPR T 1.22 See Appendix C.

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 Plant Specific Analysis Results Parameter (s) not bounded: Accident Reactivity Shape Function - Cold Resultant peak enthalpy (cal /g): 274.3 l

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1. No. indicates number of notches withdrawn out of 48. Blank is a withdraun rod.

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Figure 6. Limiting RWE Rod Pattern 12

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l Y1003J01A19 Rev. O APPENDlX A PLANT PARAMETER DIFFERENCES Safety Valves: None Safety /Rolief Valveo; 13 fnstalled 12 used in analysis i Capacity at setpoint used in analysis, 77.46%*

I GETAB Initial Conditions See the revision to Table 5-8 of NEDE-240ll-P-A (page 5-66) enclosed with the letter, J. F. Quirk (GE) to Olan D. Parr (NRC), " General Electric Co.

Licensing Topical Report NEDE-240ll-P-A, Generic Reload Fuel Apolication, Appendix D, Second Submittal," February 28, 1979.

Initial MCPR l

I The initial MCla for the P8x8R fuel was less than the operating limit MCPR based on nominal ACPRs. This is discussed on pp. B-114 and B-115 of the

" Generic Reload Fuel Application," NEDE-240ll-P-A-1.

Fuel Loading Error Results l

LHGR: 17.5 kW/f t including a 2.2% power peaking penalty due to fuel I

densification.

l

• The value formerly used was 76.25%. More recent calculations yielded the value of 77.46%.

21/22

Y1003J01A19 Rev. O APPENDIX B ODYN TRANSIENT CODE All rapid pressurization and overpressure protection events have been analyzed using the ODYN transient code as specified in Reference B-1. Code over-pressure protection analysis results are deterministic as discussed in Reference B-2. The ACPR values given for the pressurization events in Section 9 are the plant-specific deterministic values calculated by ODYN based on the initial MCPR given in Item 7 of this submittal. These ACPRs may be adjusted to reflect either Option A or Option B ACPRs by employing the con-version method described in Reference B-2. These adjustments are based on conservatism factors applied to the ratio ACPR/ICPR. The MCPR for the event is determined by adding the ACPR to the safety limit. Section 11 presents both the MCPRs for the nonpressurization events, as well as the adjusted MCPRs (Option A and Option B) for the pressurization events.

The operating limit MCPR is the naximum MCPR of the following events:

(1) turbine trip or load rejection without bypass based on ODYN; (2) feedwater controller failure event based on ODYN; (3) loss of feedwater heating event; (4) rod withdrawal error event; (5) bundle loading error accident; (6) minimum required by LOCA; and (7) minimum required by Reference B-3, Appendix C, Page C-65 where Items 3 through 7 are calculated as described in Reference B-3 but the MCPRs for the pressurization events analyzed with ODYN have been adjusted as follows:

23

Y1003J01A19 Rev. 0 (1) MCPRs are adjusted for Option B for all plants choosing to operate under Option B which meet all scram specifications given in Reference B-4.

(2) MCPRs are determined by a linear interpolation between the Option A MCPR and the Option B MCPR for all plante choosing to operate under Option B which do not meet the scram tiac specification. This interpolation is based on the teeted measured scram time and is described in Reference B-4.

REFERENCES B-1. Letter, R. P. Deniee (NRC) to G. G. Sherwood (GE), January 23, 1980.

B-2. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment Method for Determination of Operating Limits", January 19, 1981.

B-3. " Generic Reload Fuel Application", NEDE-24011 P-A-1, August 1979.

B-4. Letter (with attachment), R. H. Buchholz (GE) to P. S. Check (NRC),

" Response to NRC Request for Information on ODYN Computer Model",

September 5,1980.

24

I Y1003J01A19 Rev. O APPENDIX C

, BUNDLE LOADING ERROR EVENT ANALYSES The bundle loading error analyses, procedures, and results for the rotated bundle are presented below. The mislocated bundle loading error event analysis is no longer being reported as discussed in Reference C-2.

c.1 ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT l The rotated bundle loading error event analysis results presented in this .

l supplement are based on the new analysis procedure described and approved in Reference C-1. This new method of performing the analycis is based on a more accurate detailed analytical model.

The principal difference between the previous analysis procedure and the new 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 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 the upper regions of the limiting fuel rod. This results in the calculation 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 that the limiting event is a rotated F8DRB284L bundle resulting in a 17.5 kW/ft LHGR and a 0.15 aCPR (includes a 0.02 penalty due to variable water gap R-factor uncertainty) with a minimum CPR of >1.07. The LP.GR value includes a 2.2% power peaking penalty due to fuel densifica, tion.

25

l Y1003J01A19 Rev. 0 i REFERENCES C-1. Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, dated May 8,1978.

C-2. Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change in General l

l Electric Methods for Analysis of Mislocated Bundle Accident",

November 14, 1980.

1 1

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l l Y1003J01A19 Rev. 0 l

APPENDIX D DENSIFICATION POWER SPIKING Reference D-1 documents the NRC staff position that "... it (is) acceptable to remove the 8x8 and 8x8R spiking penalty f actor from the plant Technical Specification for those operating BWRs for which it can be shown that the predicted worst-case minimum transient LHGRs, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGRs".

The Browns Ferry-1 Reload-4 submittal contains the required information to remove the power spiking penalty from the Technical Specifications. Section 10 (Rod Withdrawal Error), Appendix A (Fuel Loading Error) and Appendix C (New Loading Error Event Analyses Procedures) include the densification effect in the calculated LHGR.

REFERENCES l

l D-1. " Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking Due to Densification Effects in BWR 8x8 Fuel Design and Performance", Reactor Safety Branch, DOR, May 1978.

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Y1003J01A19 Rev. O APPENDIX E LEAD TEST ASSEMBLIES In Spring 1981, Browns Ferry-1 will load four lead test assemblies (LTAs) which are exactly the same as the standard P8DRB284L reload bundle except for a small axial section of increased Gado11nia content in some rods. Test measurements will be performed on these bundles during Cycle 5 to benchmark the ef fect of this increased Gadolinia content. All approved thermal-mechanical and reload methods described in NEDE-24011-P-A, " General Electric Standard Application for Reload", will hold for these LTAs. Results of the reload analyses are given in Sections 2 through 16 of this report.

Since the LTAs are essentially the same as the standard P8DRB284L reload bundles except for the small axial section of increased Gadolinia, the LHGR, CPR, and MAPLilGR limits for the standard bundles will apply. For the nodal regions of increased Gadolinia, appropriate local peaking factors will be provided for the process computer.

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