ML20081K823

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Errata & Addenda for Amend 6 to GE Reload Fuel Application for Oyster Creek
ML20081K823
Person / Time
Site: Oyster Creek
Issue date: 08/31/1983
From:
GENERAL ELECTRIC CO.
To:
Shared Package
ML20081K795 List:
References
79NED288, NEDO-24195-A-06, NEDO-24195-A-06-ERR, NEDO-24195-A-6, NEDO-24195-A-6-ERR, NUDOCS 8311100275
Download: ML20081K823 (23)
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Text

r NUCLEAR ENERGY BUSINESS OPER ATIONS e GENERAL ELECTRIC COMPANY SAN JOSE, CA LIFORNI A 9512S V .

GENERAL ELECTRIC APPLICABLE TO:

PUBLICAN NO NEDO-24195 79NED288 ERRATA And ADDENDA T. i. E. NO.

TITLE eneral Mectdc Reload NO 6 Fuel Application for Oyster OA E August 1983 Creek -

NOTE: Correct allcopies of the applicable ISSUE OATE August 1979 publication as spect/ied below.

REFERENCES Sect ON, INSTRUCTIONS ITE M ApNE) p n c gp L (CO A RECTIONS AND ADDITIONS) 01 Title Page Replace with new Title Page 02 Page v/vi Replace with new page v/vi 03 Page 5-39 Replace with new page 5-39 04 Page 5-41b Replace with new page 5-41b

/ 05 Page 5-59a Replace with new page 5-59a (3") 06 Page 5-59b Replace with new page 5-59b 07 Appendix C Insert new Appendix C V) 83'11100275 031028 PAGE 1 Of 1 PDR ADOCK 05000219 P PDR

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O NEDO-24195 79NED288 l

I Class I >

August 1979

Amendment 6 August 1983

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i GENERAL ELECTRIC  !

j RELOAD FUEL APPLICATION  !

l-FOR I OYSTER CREEK

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! i i NUCLEAR POWER SYSTEMS DIVISION + GENERAL ELECTRIC COMPANY

!' SAN JOSE, CALIFORNIA 95125 O GENER AL $ ELECTRIC o

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NEDO-24195 TABLE OF CONTENTS (Continued)

Py S. REACTOR LIMITS DETERMINATION 5-1 5.1 Fuel Cladding Integrity Safety Limit 5-3 5.1.1 Statistical Model 5-3 5.1.2 Bounding BWR Statistical Analysis 5-4 5.2 MCPR Operating Limit Calculational Procedure 5-5 5.2.1 Transient Descriptions 5-10 5.2.2 Exposure-Dependent Limits 5-20 5.2.3 Effect of Fuel Densification of MCPR Operating Limit 5-20 5.3 Vessel Pressure ASME Code Compliance Model 5-27 5.4 Stability Analysis Method 5-27 5.5 Accident Evaluation Methodology 5-29 5.5.1 Control Rod Drop Accident Evaluation 5-30 5.5.2 Loss-of-Coolant Accident 5-39

  • 5.5.3 Main Steamline Break Accident Analysis 5-42 5.5.4 Loading Error Accident Calculational Methods 5-42 k 5.5.5 One Recirculation Pump Seizure Accident Analysis 5-44 5.5.6 Refueling Accident Analysis 5-44 5.6 References 5-50 APPENDICES A REFERENCE CYCLE SUPPLEMENT A-1 ~

B THERMAL HYDRAULIC MODEL FOR NON-GE FUEL B-1 n C CYCLE 10 RELOAD LICENSE SUBMITTAL C-1 h O v/v1

R NEDD-24195

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( j 5.5.2 Loss-of-Coolant Accident This analysis of the Oyster Creek Nuclear Generating Station loss-of-coolant accident (LOCA) is provided to demonstrate conformance with the ECCS accept-ance criteria of 10CFR50.46. The objective of the LOCA analysis contained herein is to provide assurance that the most limiting break size, break loca-tion and single failure combination has been considered for the plant. The documentation contained in this section is intended to satisfy these require-ments.

The general description of the General Electric (GE) LOCA evaluation models is contained in Reference 5-18. Applicability and approval for pre-pressurized reload fuel are given in Reference 5-25. Model changes are described in References 5-20 and 5-21, which were approved by the USNRC (Reference 5-19).

The analysis utilizes the short-term thermal-hydraulic model (LAMB) and the transient critical power model (SCAT) in addition to the long-term thermal-hydraulic model (SAFE) and the core heat-up model (CEASTE).

(

's' This LOCA analysis differs from previous BWR/2 LOCA analyses in the following ways:

1. Core flow coastdown and core depressurization are now calculated with the LAMB and SCAT codes, as opposed to no modelling of coastdown. .

In order to use the LAMB computer code for a non-jet pump plant, LAMB inputs were developed to execute the code for a recirculation line break. These inputs reflect the geometry of the non-jet pump plant. -

The LAMB code only allows for two recirculation pump loops; therefore, the five loops were modeled such that the intact loops are combined into one loop, and the broken loop is modeled as the second loop.

2. For the large break region, dryout time is now calculated using the SCAT code, as opposed to assuming a set dryout time. l l
3. Increased fission gas release at higher exposures has been included in m I co the results. Ei

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J The Oyster Creek LOCA analysis reported here was performed as an independent, self-contained analysis similar to that performed as a lead plant analysis. I 5-39 -

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NEDO-24195

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spectrum in this region shows that the 0.1 ft break is the most limiting break size. The, limiting break in this region shifts from the previous 0.13 ft to 2

0.10 ft because of the use of the SBM. The previous worst break of 0.13 ft was' determined by.the LBM without credit for coastdown core flow. The dif-ferences in theee models cause the time between uncovery and rated core spray

~

to have a greater effect than.the higher decay heat at time of uncovery for 0.13 ft2 The time between uncovery and rated spray is approximately 15 sec-onds greater for the 0.1 ft2 break compared to the 0.13 ft2 break. This larger time between.these two events is responsible for the 0.10 ft2 break being more limiting than all other break sizes in this region.

5.5.2'10 Break Spectrum Conclusions A summary of the analytical results is given in Table 5-11. Table 5-12 lists the figures provided for this analysis. The MAPLHCR values for each General Electric fuel type available to' Oyster Creek are given in Table 5-14. Peak cladding temperatures and oxidation fractions are also given in this table.

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A ' The' MAPLHGR values at 35 and 4d GWd/t are the only ones affected by the m m

.= m Q. ~ increased fission gas release. 3 2 The results of the complete break spectria show that two break sizes may be limiting depending on fuel exposure. The exposure-dependent limiting break

. is due to the combined effects of gap conductance and rod internal pressure.

For fuel in the low exposure region ($1000 mwd /ST), the DBA break size is limiting because of the PCT limit of 2200 F. Because of a gradual improvement in gap conductance after 1000 mwd /ST, the small break (0.1 f t ) becomes limiting ,

due' to PCT at the mid-exposure range (5000 mwd /ST to 15,000 mwd /ST). At higher exposure (i20,000 mwd /ST), as fission gas release increases, the DBA break becomes limiting due to peak local oxidation fraction.

The break spectrum sununary curve is shown in Figure 5-24. This figure gives the maximum PCT _and the maximum local oxidation, over all exposures, as a function of break size.

+

5-41b

9 NEDO-24195 7-Table 5-14a U MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Oyster Creek Fuel Type: P8DRB239 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 9.5 2198 0.095 1000 9.5 2198 0.095 5000 9.5 2194 0.085 10,000 9.5 2193 0.085 15,000 9.5 2194 0.085 20,000 9.0 2092 0.169 25,000 8.9 2048 0.169 30,000 8.9 2049 0.170 35,000 8.7 2050 0.170 n co 40,000 8.3 2049 0.170 2

  1. '"*g

'v' Table 5-14b E N

. MAPLilGR VERSUS AVEIV.GE PLANAR EXPOSURE Plant: Oyster Creek Fuel Type: P8DRB265L Average Planar Exposure  !!APLilGR PCT 0xidation

(!!Wd / t ) (kW/ft) (*F) Fraction 200 9.5 2198 0.095 1000 9.5 2198 0.095 5000 9.5 2198 0.086 10,000 9.5 2198 0.086 15,000 9.5 2198 0.086 20,000 9.0 2081 0.170 25,000 8.8 2049 0.169 30,000 8.8 2045 0.170 ,

35,000 8.6 2049 0.170 g m 40,000 8.3 2050 0.170 S (w i 5-59a -

NEDO-24195 e Table 5-14c MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Oyster Creek Fuel Type: P8DRB265H Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 9.5 2198 0.095 m 1000 9.5 2198 0.095 ,f'

'5000 9.5 2198 0.086 10,000 9.5 2198 0.086 15,000 9.5 2198 0.086 20,000 8.9 2078 0.170 25,000 8.8 2050 0.170 30,000 8.8 2044 0.166

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35,000 8.6 2049 0.170 m e

40,000 8.3 2050 0.170 2

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5-59b

a mwi _m. _ _ _ __ __ .,s. y - - a = _- . - -

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NEDO-24195 O

APPENDIX C CYCLE 10 RELOAD LICENSE SUBMITTAL 4

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NEDO-24195 >

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. C.1 REFERENCE CORE LOADING PATTERN (1.0, 2.7. 3.3.1 and 4.0)

Fuel Type Group Number Irradiated Exxon Type VB. A 24 Exxon Type VB B 188 Exxon Type VB C 148 i

New- Exxon Type'VB D- 28 P8DRB239 E 112 P8DRB265H F 60 Total 560 Nominal previous cycle core average exposure at end of cycle: 17,453 Wd/t Minimum previous cycle core average exposure at end of cycle >

from cold shutdown considerations: 17,453 Wd/t O Assumed reload cycle core average exposure at end of cycle: 14,212 Wd/t Core loading pattern: Figure C-1 Sources of non-GE bundle characteristics:

Local Thermal-Hydraulic Nuclear Peaking Model (Fuel and Bundle Type _ Libraries Factor R-Factor Channels)

Exxon Type VB Developed via GE See Appendix B See Appendix B lattice methods

*( ) Refers to areas of discussion in " General Electric Relcad Fuel Application s -

for.0yster Creek", NEDO-24195, August 1979.

C-1 9 y,,r6-- ,ryeiy -y ~

ew- e .2%- -, ,,-yr'vM-+>wts' -w-r-- - - * * + - ' - - - - * - - - * - -*r-

NEDO-24195 m

(Q C.2 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - No VOIDS, 20*C (3.3.2.1.1 and 3.3.2.1.2)

BM k,gg Uncontrolled 1.108 Fully Controlled 0.944 Strongest Control Rod Out 0.984 R Maximum Increase in Cold Core Reactivity with 0.000 Exposure Into Cycle, Ak C.3 STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak) m (20*C, Xenon Free) 600 0.044 C.4 TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)

O Void Coefficient N/A* (C/% Rgo) -6.28/-7.86 Void Fraction (%) 36.2 Doppler Coefficient N/A (c/*F) -0.219/-0.208 Average Fuel Temperature (*F) 1157 Scram Worth N/A ($) -37.64/-30.11 Scram Reactivity vs Time Figure C-2 C.5 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

EOC Fuel Design P8x8R Ex8 Peaking factors (local, 1.20 1.28 radial and axial) 1.738 1.650 1.40 1.40 R-Factor 1.051 1.098 Bundle Power (MWt) 5.839 5.553 3

Bundle Flow (10 lb/hr) 91.13 90.75

'- Initial MCPR 1.32 1.29

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

NEDO-24195

() C.6 C_0RE-WIDE TRANSIENT ANALYSIS RESULTS (5.2. Q R

Exposure h 'Q/A Transient (mwd /t) (% NBR) (%) P8x8R Ex8 Figure Turbine Trip E0C 495 119 0.25 0.22 Figure C-3 without Bypass Loss of 100*F BOC to EOC 114 113 0.12 0.11 Figure C-4 Feedwater Heating Feedwater Controller EOC 305 118 0.20 0.18 Figure C-5:

Failure C.7 LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(5.2.1)*

Limiting Bod Pattern: Figure C-6 Rod Position WR EHM (kW/f t)

Reactor Power (%) (Feet Withdrawn) P8x8R Ex8 P8x8R Ex8 O' 104 6.5 0.23 0.29 15.2 15.6 105 7.5 0 25 0.31 14.7 15.4 106** 8.5 0.27 0.33 14.9 15.8 107 9.0 0.27 0.33 15.3 16.4 108 9.0 0.27 0.33 15.3 16.4 109 9.5 0.28 0.34 15.8 16.9 110 10.0 0.29 0.34 16.2 17.5 C.8 CYCLE MCPR VALUES *** (5.2)

Non-pressurization Events Exposure Range: BOC to EOC P8x8R Ex3 loss of Feedwater Heater 1.19 1.18 Fuel Loading Error 1.25 1.27 Rod Withdrawal Error

  • 1.34 1.40
  • See Alternate Rod Withdrawal Error Analysis.

f)

'- ** Indicates APRM rod block setpoint selected.

C-3

NEDO-24195

() Pressurization Events Exposure Range: BOC to EOC Option A Option B P8x8R Ex8 P8x8R Ex8 Turbine Trip w/o' Bypass l'.38 1.35 1.33 1.30 Feedwater Controller Failure 1.33 1.30 1.25 1.23 C.9 OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

P,1 P y

Transient (psig) Igsigl Plant Response MSIV Closure 1261 1298 Figure C-7 (No Scram)

C.10 ' STABILITY ANALYSIS RESULTS (5.4) '

Decay Ratio: Figure C-8 Reactor Core Stability Decay Ratio, x2 /*0: 0.61 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 P8x8R Channel: 0.15 Exxon 8x8 Channel 0.15 C.ll LOADING ERROR RESULTS (5.5.4)

Rotated Bundle: Exxon 8x8 P8x8R ICPR: 1.25 (before rotating) 1.24 MCPR: 1.07 (after rotating) 1.08 ACPR: 0.20* 0.18*

()

  • Includes 0.02 ACPR penalty.

C-4 t

. - - , . - . - , . , . . . , , - - - . - - ,e- ~,, -n ,, ., , ,- , , , s . . , .

NEDO-24195

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oMMMMMo
mE+EM M i+Es+ M ME+iM@
Mit M M NE+sM M M M M 40 I:3 Ei3 E E 3 3 2 Y 3 @1CEE312C 3a I CT.B TI EY8 ET3EB['IEIEI[Y@E+32+@3 T Y [@E 0

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M M M M M M M E4 E M E4 M M
Mi+sM!+Es+ 88Ms+EE4i+ MMM

'::54 M M M M M M E4 M M M M E4 o  :: "E+sM!+ MME+8s+ Mi+ 24i+ " .

i E+8 M M M M E 4 M E4 :4 84 .

"i+sM M M M Ms4s+i!4"
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E4E+iME+EE+8 IIIIIIIIIi 1 357 9111315171921232527293133353739414345474951 FUEL TYPE A = EXXON-VB D = EXXON-VB B = EXXON-VB E = PODRB239 C = EXXON-VB F = P8DRB265H

-> Figure C-1. Reference Core Loading Pattern for Cycle 10 C-5

NEDO-24195 O

I 45 CONTROL ROD DRIVE VS TIME SCR AM REACTIVITY VS TIME 90 -

- 40 80 -

- 35 70 -

67A CRD IN PERCENT 30 60 -

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- 25 3 8 4 2 >

9. 50 -

D 2 2 O s a

5 20 :::

40 _ NOMINAL SCR AM CURVE IN 1-83 15 30 -

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- 10 20 -

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10 - SCR AM CURVE

! USED IN ANALYSIS l

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0 1 2 3 4 5 6 l

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V EOC Scram Reactivity Figure C-2.

C-6 a

NEDO-241"5 O

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} t NEUTRON FLUK 1 YESSEL PRESh RISEiPSI) 2 AVE SUROCE HEAT FLUX 2 SAFETY VALC rLC+W 3 CORE IM.ET FLOW 132.. 3.... 3. --==ce RELIEFu?LuEer- WALvE Flow

$ s ... . - -

% 1 2....

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n w

5 .. . t....

... 2.. 4.. S. . ... 2.. 4.8 8*.

TIME (SEC0 e S3 TIME ISECO251

(~)

U lLEVELEINCH.EEF-SEP-SMRT) 1 VO!D REACT. ITY 2 VESSEL STEAFLOW 2 00PPLER REAC VITY

2. .. .  ? M_E_1_5'PZ'o' i..

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=

i.... .

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A O m A_a -

r v. v-- g

... T de. d._

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5

-i. .

-i....

-2..

... 2. . ... ... ... 4.. s..

2..

TIME (SECO W S) TIME (SECO2SI Figure C-3. Plant Response to Turbine Trip Without Cypass C-7

NEDO-24195

%d 1 NEufRON FLUX 1 VES'5EL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 REL!EF VALVE Flow 3 CORE INLET FLOW

  • 3 BYPNSS VALVE FLOW 13s.0 '=r

'i=_" of 100.0

. . O  ; I a ,

W 108.6 B

] 50.0 W

r 54.8

0. 0  : ;f - ; j
e. 0 0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECONOS) TIME (SECON053 O l LEV [L(INCH-REF.SEP. SERT 3 1 v01 3 REACTIVITY 2 VES iEL STEAMFLOW 2 00F'LER REACTIVITY 3 TURIINE eer y tra STEAPFLOW 3 SCRAM REACTIV!TV 15 9. o er_~ 3. 8 ' ver o_ near-'w'vv 2

g . . . .

-- :,: -: 5 .. . E -. -

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= -

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56.0 W . .e W ,

2.e 0.0 300.0 2 0s. O e. s 100.0 200.0 fire tsEcch011 TIME (SECCNCS) b V Figure C-4. Plant Response te Loss of 100*F Feedwater Heating C-8

NEDO-24195 3

(a 150.0 INEurRON FLUX  :

] 1 VEs5EL PRESS RISEtPS!!

2 ave SURFACE HEAT F1, x 2 5AF ETY VALVE FLOW 3 CDRE INLET FLOW -

3 REL IEF VALVE FLOW r

u. .. ' r aa Mrv ee I 4 BYP LS$ VALVE FLOW r

100.0

$ . . y'

183 8 , 3.~. .

l as b

- 50.0 b  %,

30.0

P '

88 '

- .. - .~ .2

0. 0
0. 0 20.0 00.0 0. 0 20.0 40.0 TIME (SECONOSI f!ME (SECCNOSI O

V I LEY ELtINCH-REF-SEP-SMRT) t Vol ) REACTIv!TY 2 VESSEL STEAMFLOW 7 DOP)LERREACTIVITYl 3 SCR LM REACTIVITY 150.0 3m TUR3!N.E erem ven enSTEAMFLOW

~ 1.0 m vnt u_ menttiviev i

  • ^ ^

lT.

100.0 ' ' "

22 2 2 0 5 g

80 . - . - . .-

- , .- . .~ ~ .

b G

E S t. e W -1. 0 W .

G8

/ ,

-2.0

0. 0 20.0 to.O 0. 0 20.0 40.0 TIME (SECOW S) TIME (SECONOS)

)

Figure C-5. Plant Response to Feedwater Controller Failure C-9

u NEDO-24195 f

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2 6 10 14 18 22 26 51 34 34 ,.

i 47 4 0

,I 43 34 38 30 39 4 2 18 L

35 34 30 42 i

0 0 31 18 27 34 38 42

-NOTES: 1. Rod pattern is-1/4 core mirror symmetric.

! 2. No. indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.

~3. -Error rod is (22,.1).

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l Figure C-6. Limiting RWE Rod Pattern i

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.._--____.____2._..__.._...__...-_.___ . . _ . , - _ . - _ , _ , . - _ _ - . _ . .

NEDO-24195 G

V 1 UTR0r4 F.UX 1 CSSEL PRESS RISECPSI) 2 A E SURF A:E P( AT FLUX 4 $AFETY VALVE FLOW 3 C ,RE IM.ET FLOW 3 RELIEF VALVE FLOW 13 0. t 30s.0 'e =*ee u* _or ete-M E

!! lee.e --

y 2ss.s 3 7 8

U W

r - -

30.6  % J see.s C~ N-

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8. 0 5.0 1. 0 5'. 0 TIME ISECONOSI TIME ISECONDS)

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ILEVELCINCd-REF.SEP.SnRT) 1 VOID REACH!VITY 2 VESSEL STEAMFLOW 2 DOPPLER RfACTIVITY TUR91p 233.3 3. c r e ne .EveSTE,AMFLOW n e = 3. vgvit SCRAMneire,uetv REACTIVITY

3. s 3

Is a. s -- N ~%- E g s.e _ .- .

A u

% - C A,... <. .

V

. N. .

5.,.,

, N,, . .

388.0

,y,,

8. 0 5.0 9.8 5. 9 TIME ISECONOSI TIME (SECONDS) b(/ Figure C-7. Plant Response to MSIV Closure, No Scram I

l C-ll

NEDO-24195 O

1.25 -

A - NATURAL CIRCULATION B - 100 PERCENT ROD LINE C - ULTIMATE PERFORMANCE LIMIT C

1m B

5 x

50.7s -

p O i E

0.50 - A a

0.25 -

a l I I i 1 0 20 40 80 80 100 120 PERCENT POWER O Figure C-8. Reactor Core Decay Ratio C-12

A NEDO-24195 O

ALTERNATE ROD WITHDRAWAL ERROR ANALYSIS WITH LESS THAN LIMITING INSTRUMENT FAILURES O

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C-13

NEDO-24195 LOCAL ROD WITHDRAWAL ERROR (WITH LESS THAN LIMITING INSTRUMENT FAILURES)

Oyster Creek calculates ACPR for the RWE based on the status of instrument availability. The ACPR for the limiting instrument failure (APRM Status 1) is provided in Subsection C.7. The ACPRs for the RWE for other than the limit-ing APRM instrument conditions are provided below.

Aprm Status 2 (If any LPRM' input to the APRM system at the B, C, or D level is failed or

, bypassed or any APRM channel is inoperable or bypassed)

Limiting Rod Pattern: Figure C-6 0 ( E)

Reactor Power Rod Position -

(%) (Feet Withdrawn) P8x8R Ex8 P8x8R Ex8 104 3.5 0.'12 0.11 13.7 13.5 4.0 105 0.15 0.14 14.7 14.8 106* 4.5 0.16 0.17 15.0 14.9 107 4.5 0.16 0.17 15.0 14.9 Cycle MCPR Values P8x8R Ex8 Rod Withdrawal Error 1.23 1.24 n.

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  • Indicates APRM rod block setpoint selected.

C-14

NEDO-24195

'APRM Status 3

.(All B, C, and D LPRM inputs to the APRM System are operating and no APRM channels are inoperable or bypassed.)

Limiting Rod Pattern: Figure C-6 A t)

Reactor Power Rod Position .

(%) (Feet Withdrawn) . P8x8R Ex8 P8x8R Ex8 104. 2.0 0.05 0.04 13.4 11.3 105 2.5 0.07 0.07 13.4 11.3 106* 3.0 0.10 0.09 13.5 12.3 107 3.5 0.12 0.11 13.7 13.5 Cycle MCPR Values P8x8R Ex8 Rod Withdrawal Error 1.17 1.16

r O

4 1

e

  • Indicates APRM rod block setpoint selected.

C-15

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