ML20100G511
| ML20100G511 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/04/1985 |
| From: | Holm J, Stout R, Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17321A517 | List: |
| References | |
| XN-NF-84-21-S01, XN-NF-84-21-S01-R02, XN-NF-84-21-S1, XN-NF-84-21-S1-R2, NUDOCS 8504080273 | |
| Download: ML20100G511 (42) | |
Text
"
XN-NF-84-21 I REVISION 2
[
SUPPLEMENT 1 E
DONALD C. COOK UNIT 2 CYCLE 5
[ 50/0 STEAM GENERATOR TUBE PLUGGING
~
[ LIMITING BREAK LOCA/ECCS ANALYSIS- )
KLZ? CURVE APRIL 1985 1
RICHLAND,WA 99352 l ERON NUCLEAR COMPANY,INC.
l pr188= =88;j3 1
) XN-NF-84-21 Revision 2 Supplement 1 Issue Date: 4/4/85 L
DONALD C. COOK UNIT 2 CYCLE 5
( 5% STEAM GENERATOR TUBE PLUGGING LIMITING BREAK LOCA-ECCS ANALYSIS: K(Z) CURVE Contributor: M. J. Ades Prepared by: 9. '2p f 3 /30/f[
T. Tahvili PWR Safety Analysis Reviewed by: w J ha f((
=
7 'S. Holm, Manager R Safety Analysisj" Approve: (( u K.B. Stout,Janager aL- 5/fo//J' Licensing & Safety Engineering Approve: //W A GE M .4/mpf
- G. F.'0wsley/ Mandger Reload Licensing Liaison Approve: M)
J.
A Mo~rgan, Manager V/9/>;s-Cu er Services Engineering Approve: [h 4 <//v/g
~
G. L. Ritter, Manager /
Fuel Engineering & Technical Services l
gf '
ERON \ UCLEAR COMPANY, NC.
NUCLEAR REGULATORY COMMISSION DISCLAIMER j IMPORTANT NOTICE REGAROING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was clerived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-fabricated reload fuel or other techn~ cal services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxen Nuclear in their demonstration of comoliance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor i
any person acting on its behalf:
A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-motion contained in this document, or that the use of any information, apparatus, method, or process disclosed in this docurnent will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for - -
darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
XN- N F. F00, 766 1
i O
t .
L ,
i XN-NF-84-21 Revision 2 Supplement 1
(
l- TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
.......................................... 1 2.0~
SUMMARY
................................................ 2 3.0 ANAL Y S IS RE S UL TS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
'3.1 LOCA ANALYSIS MODEL'.............................. 5 3.2 K(Z) DETERMINATION ............................... 6 4.0- REFERENCES ............................................ 33 I
f-s i-4 i-
N , .
o l
Y, l ii- XN-NF-84-21 f.. -
Revision 2 Supplement 1 LIST OF TABLES-Table. Page
~
2.1 -D.C. Cook Unit 2 K(Z) Determination Results .............. 3 3.1- Donald C. Cook' Unit 2 System Input Parameters ............
-9~
-3.2 1.0 DECLG Break. Analysis Pa'rameters ...................... 10 i P
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v iii XN-NF-84-21 Revision 2 Supplement 1 LIST OF FIGURES e
L Figure Page h 2.1 0.C. Cook Unit 2; K(Z) - Normalized Fl0(Z).
as a Function of Core Height for ENC Fuel................. 4
( 3.1 D.C.CookUnit2Fh(Z)DistributionsUsedin L LOCA-ECCS Analysis ....................................... 11 3.2 D.C.CookUnit2Fh(Z)-LOCA-ECCSAnalysisLimit........ 12
{
3.3 Heat Transfer Coefficient During Blowdown Period at PCT Node,1.0 DECLG Break, B0C, Axial Offset of +9.0% ................................................. 13 3.4 Clad Surface Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, B0C, Axial Offset of +9.0% ................................................. 14 3.5 Depth of Metal-Water Reaction during Blowdown Period at PCT Node,1.0 DECLG Break, BOC, Axial Offset of +9.0% .......................................... 15 3.6 Average Fuel Temperature during Blowdown Period at PCT Node, 1.0 DECLG Break, B0C, Axial Offset of +9.0% ................................................. 16 3.7 Hot Channel Average Quality, Center Volume,.
1.0 DECLG Break, B0C, Axial Offset of +9.0% .............. 17 3.8 Reflood Core Mixture Level,
{ 1.0 DECLG Break, B0C, Axial Offset of +9.0% .............. 18 3.9 Reflood Downcomer Mixture Level, 1.0 DECLG Break, B0C, Axial Offset of +9.0% ........... .. 19 3.10 Reflood Upper Plenum Pressure, 1.0 DECLG Break, 80C, Axial Offset ~of +9.0% .............. '20 3.11 Core Reflooding Rate, 1.0 DECLG Break, B0C, Axial Offset of +9.0% ............................... 21 3.12 T000EE2 Cladding Temperature versus Time, 1.0 DECLG Break, BOC, Axial Offset of +9.0% .............. 22'
- l. 3.13. Heat Transfer Coefficient During Blowdown Period
{ at PCT Node, 1.0 DECLG Break, E0C, Axial Offset of +3.2% ................................................. 23
d iv XN-NF-84-21 Revision 2 Supplement 1 LIST OF FIGURES (Cont.)
Figure Page
)
3.14 Clad Surface Temperature during Blowdown Period at PCT Node,1.0 DECLG Break, EOC, Axial Offset of +3.2% ................................................. 24 3.15 Depth of Metal-Water Reaction'during Blowdown
, Period at PCT Node,1.0 DECLG Break, E0C, Axial,
- j. Offset'of +3.2% .......................................... 25
]
l 3.16 . Average Fuel Temperature'during Blowdown' Period at PCT Node, 1.0 DECLG Break,.E0C, Axial Offset of +3.2% ................................................. 26 3.17 Hot Channel Average Quality, Center Volume, 1.0 02CLG Break, E0C, Axial; Offset of-+3.2% .............. 27 3.18' Reflood Core Mixture Level,-
1.0 DECLG Break, E0C, Axial Offset of +3.2% .............. 28 3.19 Reflood Downcomer Mixture Level,
- 1.0 DECLG Break, E0C, Axial Offset of +3.2% .............. 29' 3:20 Reflood Upper Plenum Pressure, 1.0 DECLG Break, E0C, Axial Offset of +3.2% ..............- 30 3.21 Core Reflooding. Rate, 1.0 DECLG Break, E0C, Axial 0ffset of +3.2% ............................... 31 )
3.22 T000EE2 Cladding Temperature versus Time,-
1.0 DECLG. Break, E0C, Axial Offset of +3.2% .............. 32 .,
1 7
. 1 f
/
a 1 XN-NF-84-21 Revision 2 j Supplement 1
1.0 INTRODUCTION
- This report documents the results of a large break LOCA-ECCS analysis L
performed to determine the axial dependence of permissible limits on power
(- - peaking, i.e., the K(Z) curve.
A large break LOCA-ECCS analysis for the Cycle 5 operation of the D.C.
Cook Unit 2 reactor at 3425 MWt with ENC fuel was performed and reported in July 1984.(1) This previous analysis addressed only cosine axial power dis-k tributions. The results presented in Reference 1 were performed for the
_p reviously identified limiting break using the EXEM/PWR(2) ECCS models. An
{
L earlier report (3) presented analytical results for a spectrum of postulated
( large break'LOCAs. The limiting break was identified as the 1.0 Double Ended Cold Leg Guillotine (DECLG) break. <
~The calculations presented in this report are with upskew axial power
. distributions. They used the EXEM/PWR(2) .ECCS models ' as discussed in Reference 1 with one exception: the revised FLECHT-based heat transfer correlations were used as reported in Reference 4.
(
[
k
2 XN-NF-84-21' K Revision 2 Supplement 1
/
2.0
SUMMARY
c' A large break LOCA-ECCS analysis has been performed to determine the L
axial dependence of permissible limits on power peaking, K(Z). The axial
( dependence of the power peaking limit is denoted K(Z) and is defined as K(Z) =
FQ(Z)/ max Fg(Z), where FQ (Z) is the maximum peaking allowed at any elevation Z. 'The upper portion of this K(Z) curve conservatively bounds the previously reported small break portion, and therefore the small break portion of the
[
K(Z) has been eliminated.
The determination of the limits on power peaking versus axial location from the large break LOCA-ECCS analysis is based on three power distributions:
a center peaked chopped cosine power distribution; a power distribution
. representative of the largest axial offset (+9%) anticipated at BOC con-ditions; and finally a power distribution representative of the largest axial offset anticipated (+3.2%) at E0C conditions. Axial offset is defined as the power in the top of the core minus the power in the bottom of the core divided by the total core power. The power distributions are analyzed at the exposure
{
at which they are expected to occur. A summary of these results is presented in Table 2.1.
Lines are constructed tangent to these power distributions. These lines
' define the K(Z) limit shown in Figure 2.1 on core power peaking for ENC fuel in D.C. Cook Unit 2. Since margin remains to the limit on peak clad temperature, it is planned to perform further analysis to support a K(z) curve with values closer to 1.0 in the upper core upper elevation.
i
Table 2.1 D. C. Cook Unit 2 K(Z) Determination Results Calculational Basis License Core Power, MWt* 3411
-Break Size, DECLG 1.0 Steam Generator Tube Plugging, % 5 Peaked ** Peaked Peaked l X/L = 0.50 X/L = 0.63 X/L = 0.79 Exposure, MWD /kg 10.0 2.0 20.0 Axial Offset 0% +9.0% +3.2% w Peak Linear Heat Generation Rate 11.61 11.21 9.62 (LHGR)*
T Total Peaking Factor, Fg 2.04 1.97 1.69 Peak Cladding Temperature, OF 2014- 1806 1820 Peak Temperature. Location, ft 9.63 10.88 11.13 Local Zr/H 2O Location, f t 9.63 10.63 11.13 Local:Zr/H 2O Reaction (Max.), % 4.7 2.18 2.35 Time for Max. Local Zr/H 2O Reaction, sec 400 .450 450 Total Zr/H 2O <1% <1% <1%
T5'E.
u<
Hot Rod Burst Time,.sec 70.51 ~100.26 106.76 "J 28 Hot Rod Burst Location, ft. 7.0 8.88 10.63 -
e n> y
- 2% power uncertainty is' added to this value in the LOCA analysis.
- This case was analyzed in Reference 1.
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3 5 XN-NF-84-21
- Revision 2 Supplement 1 F
L -
3.0 ANALYSIS RESULTS
/ The K(Z) curve is the variation of the limit on power peaking with axial l
elevation in the core. .The allowed power peaking is reduced at the top of the core to offset the effect on peak cladding temperature (PCT) of reduced
{_
coolant heat transfer from (1) the short uncovery periods at the top of the
( core during small break LOCAs, and (2) reduced coolant heat capacity at the top of the core durirg the reflood period of the large break LOCAs. The analysis model and the results of the analysis are described below.
3.1 LOCA ANALYSIS MODEL The Exxon Nuclear Company EXEM/PWR-ECCS evaluation model was used to perform the analysis required. This model(2) consists of the following computer codes: RODEX2(6) code for initial stored energy; RELAP4-EM(7) for h the system blowdown and hot channel blowdown calculations; ICECON(8) for the computation of the ice condenser containment backpressure; REFLEX (2,9) for computation of system reflood; and T00DEE2(2,10,11) for the calculation of final fuel rod heatup. T000EE2 calculations performed for this analysis are similar to those reported in Reference 1 in that the heat transfer coefficient multiplier for the presence of mixing vanes was set to 1.0, and the' heat transfer- coefficient multiplier for _ local peaking was adjusted as per Reference 5. The T000EE2 and REFLEX calculations are also similar to those in Reference 1 in that the FLECHT-based heat- transfer correlations have been modified to' account 'for the difference in axial power distribution between those used in the D.C.. Cook Unit 2 analysis and that used in the FLECHT tests:
upon which the correlations were developed. The FLECHT-based heat transfer
6 XN-NF-84-21 Revision 2 Supplement 1 k'
correlations that were used in the analj sis presented 'in this report are discussed in Reference 4.
3.2 K(Z) DETERMINATION The limits on axial power peaking are defined by drawing tangent lines to power distributions for which the acceptance criteria in 10 CFR 50.46 for LOCA-ECCS analysis are met. For the purpose of this analysis, three power distributiens have been considered. These are a center peaked chopped cosine i.
axial power distribution and two top skewed power distributions peaked at 7.5 ft (X/L = .63) and peaked at 9.5 ft (X/L = .79).
The cosine axial power distribution has been analyzed previously in Reference 1. The analysis in Reference 1 did not use the heat transfer correlation . presented in Reference 4 but used a more conservative heat transfer correlation. The analysis for the cosine axial power distribution considered three exposures: 2.0,10.0 and 47.0 MWD /kg. For'the Cycle 5 core, the resulting peak clad temperatures from this previous analysis were 20070F, 20140F, and 19930F, respectively. The analysis thus indicated no significant exposure dependency for the peak clad temperatures.
The analysis also indicated no significant exposure dependency for the maximum local Zr-H 2O reaction and the total Zr-H 2O reaction.
Previous calculations have indicated that the higher-in the core at
-which a particular-value of power peaking occurs, the larger the calculated peak clad temperature will be. For this reason, two power shapes repre-sentative of the maximum power peaking in the core top have been chosen to be analyzed for the determination -of the limit on power peaking versus core height. These power distributions were^ calculated for D.C. Cook Unit 2 at ,
7 XN-NF-84-21 Revision 2 Supplement 1
' ?
L_ BOC5 and E0C5 conditions. The two power distributions represent the most top
~
skewed shapes anticipated to occur at the cycle conditions and with the power distribution control procedures (PDC-II)(13) in effect at the plant. Because
} ,
[ there is no significant exposure dependence in the calculated peak clad
-[-- temperatures, each power shape is analyzed at the exposure conditions at which 6 g-
.j,i, ..
it is anticipated to occur. j {. .y l
+C.(
The axial power distributions used in the analysis are normalized W : ~ =, :
a to 1.0. They thus represent a relative axial power distribution FZ (Z). The }.
3:. .e_.
~ . . ..
total power distribution Fg(Z) is determined by multiplying the relative .; [
L 3 /. A 3- axial power distribution by a constant value for Fag, Fg(Z) = FAH*FZ (Z). The i'
[ y n.,.....
b, method for determining the limiting FQ (Z) entails increasing FAH until the g Jf 7 e . n. .
= peak clad temperature for the resulting F (Z) Q approaches 22000F. This ?.f g:=-
i analysis resulted in considerable margin to the 226. 3F limit, and thus further ' t .y 'z i
analysis could be performed using larger values of FAH to ac7ieve higher I
.3 l.
1
_ limits. py?' q
=
3 The power distributions utilized in the analysis are plotted in Figure 3.1. As shown in Table 2.1, these axial power distributions result in v.(
f5.9 hD. ~.
conformance to the criteria of 10 CFR 50.46 with significant margin remaining i 1. r t,
, .y ,;.
to the criteria. The power distributions plus the tangent lines ccnstructed N,". 7. . ,3
- L <,0..'
j to define the Fg(Z) limits are shown in Figure 3.2. The lines tangent to the ;4g.
, ,g ,
-i power distributions represent the limit resulting from the large break LOCA- .Md,,
- ns;: -
ECCS analysis. /JN j
Wg '.
? The operating conditions on fuel parameters for the current
$)'&.:.'
i analysis are given in Tables 3.1 and 3.2. The operating conditions and fuel J Sf
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I 8 XN-NF-84-21 Revision 2 Supplement 1 parameters are identical to those used in the XN-NF-84-21(P), Rev. 2 analysis.
Boundary condition input from the XN-NF-84-21(P), Rev.1(13) blowdown was used to drive the RELAP4/ hot channel analysis. Use of the results for previous containment pressures, safety inje ction system transients, and refill reported in XN-NF-84-21(P), Rev. 1 analysis was made to calculate REFLEX /reflood and T00DEE2/ hot rod heatup.
As indicated previously, the results of the K(Z) determination are summarized in Table 2.1. For the limiting break, the 1.0 DECLG break, RELAP4/ hot channel, REFLEX /reflood, and T00DEE2/ hot rod heatup calculations were performed. The results for the center peaked axial power distribution are reported in Reference 1. The results for the skewed power distributions are presented in Figures 3.3 through 3.12 for B0C, and in Figures 3.13 through 3.22 for E0C, respectively. Time zero on all plots corresponds to the time of break initiation.
- 9 XN-NF-84-21 Revision 2 Supplement 1 Table 3.1 Donald C. Cook Unit 2 System Input Parameters Thermal Power, MWt* 3425 Core, MWt 3411 Pump, MWt 14 Primary Coolant Flow, Mlbm/hr 143.1 Primary Coolant Volume, ft3 11,768 Operating Pressure, psia 2250 Inlet Coolant Temperature, OF 542 Reactor Vessel Volume, ft3 4945 Pressurizer Volume, Total, ft3 1800 Pressurizer Volume, Liquid, ft3 1080 Accumulator Volume, Total, ft3 (each of four) 1350 Accumulator Volume, Liquid, ft3 (each of four) 950 Accumulator Pressure, psia 636 Steam Generator Heat Transfer Area, ft2_
SG1, SG2, SG3, SG4 11,588,3(12,446)
Steam Generator Secondary Flow, Ibm /hr - 6 SG1, SG2, SG3, SG4 3.505 3(3.764x x10 10,6)
Steam Generator Secondary Pressure, psia 799 Reactor Coolant Pump Head, ft 277 Reactor Coolant Pump Speed, rpm 1189 Moment of Inertia, lbm-ft2 82,000 Cold Leg Pipe, I.D. in. 27.5 Hot Leg Pipe, I.D. in. 29.0 Pump Suction Pipe, I.D. in. 31.0 Fuel Assembly Rod Diameter, in. 0.360 Fuel Assembly Rod Pitch, in. 0.496 Fuel Assembly Pitch, in. 8.466 Fueled (Core) Height, in. 144.0 Fuel Heat Transfer Area, ft2 ** 57,327 Fuel Total Flow Area, Bare Rod, ft2 ** 53.703 Refueling Water Storage Tank Temperature, OF 80 Accumulator Water Temperature, OF 120
- Primary Heat Output used in RELAP4-EM Model = 1.02 x 3425 = 3493.5 MWt
- ENC Fuel Parameters.
10 XN-NF-84-21 Revision 2 Supolement 1 ,,
Table 3.2 1.0 DECLG Break Analysis Parameters Peak Rod Average Burnup (MWD /kg) 10.0 2.0 20.0 (B0C) (BOC) (E0C)
Axial Offset (%) 0. +9.0 +3.2 Total Core Power (MWt)* 3411 3411 3411 Total Peaking (F ) 2.04 1.97 1.69 Fraction Energy Deposited in Fuel
. Fully Moderated Core 0.974 0.974 0.974
. Voided Core 0.954 0.954 0.954 C_ycle 5 (85% ENC Fuel)
Peaking ,
. Axial 1.316 1.229 1.129
.EnthalpyRise(FIH) 1.55 1.60 1.50
- 2% power uncertainty is added to this value in the LOCA analysis.
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s Figure 3.1 0.C. Cook Unit 2 Fr)(Z) Distributions Used in LOCA-ECCS Analysis Y
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18 XN-NF-84-21 Revision 2 Supplement 1
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4.0 REFERENCES
(1) XN-NF-84-21(P), Rev. 2, " Donald C. Cook Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break LOCA/ECCS Analysis," Exxon Nuclear Company, Inc., Richland, WA 99352, July 1984.
(2) XN-NF-82-20(P), Rev. 1, August 1982; and Supplement 4, July 1984,
" Exxon Nuclear. Company Evaluation Model EXEM/PWR ECCS Model Up-dates," Exxon Nuclear Company, Inc., Richland, WA 99352.
(3) XN-NF-82-35, Supplement 1, " Donald C. Cook Unit 2 Cycle 4 Limiting Break LOCA-ECCS Analysis Using EXEM/PWR," Exxon Nuclear Company, Inc., Richland, WA 99352, November 1982.
(4) XN-NF-85-20, "A Modification of the FLECHT Based Reflood Quench and Heat Transfer Correlations," Exxon Nuclear Company, Inc., Richland, WA 99352 (Draft).
(5) Letter, H.R. Denton (NRC) from J.C. Chandler (ENC), " Supporting Documentation for Unit 2 Technical Specification Changes for Cycle 5 Reflood," dated May 7, 1984 (JCC:076:84).
(6) XN-NF-81-58(A), Rev. 2, "RODEX2: Fuel Rod Thermal-Mechanical Re-sponse Evaluation Model," Exxon Nuclear Company, .Inc., Richland, WA 99352, February 1983.
(7) U.S. Nuclear Regulatory Coninission letter, T.A. Ippolito (NRC) to W.S. Nechodom (ENC), "SER for ENC RELAP4-EM Update," March 1979.
(8) XN-CC-39, Rev. 1, "ICECON: A Computer Program Used to Calculate Containment Backpressure for LOCA Analysis (Including Ice Condenser Plants)," Exxon Nuclear Company, Inc., Richland, WA 99352, November
-1977.
(9) XN-NF-78-30(A), " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA," Exxon Nuclear Company, Inc.,
Richland, WA 99352, May 1979.
(10) XN-NF-82-07(A), Rev. 1, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., Richland, WA 99352, March 1982.
(11) G.N. Lauben, NRC Report NUREG-75/057, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," May 1975.
(12) XN-NF-77-57(P)(A) and its Supplements 1, 2, 2 Addendum 1, " Exxon Nuclear Powcr Distribution Control for Pressurized Water Reactors, Phase II," Exxon Nuclear Company, Inc.. Richland, WA 99352.
O e 1 8
, l 34 XN-NF-84-21 1 Revision 2 Supplement 1
-k"; -
(13) XN-NF-84-21(P), Rev.1, " Donald C. Cook Unit 2 Cycle 5 5% Steam i ' Generator Tube Plugging Limiting Break LOCA-ECCS Analysis," Exxon
_* Nuclear Company, Inc., Richland,. WA 99352, May 1984, i 1
l
- l. -
.G i.
7 i ~
f 1
f 4
I i
l.
I 1
I 1
?
f I
1-
--e p , as ~
q en. g m ,.,- er- .g - m - - . , p.- -m--w-- y y - yn 9 y
l h.
),
l -
XN-NF-84-21 Revision 2 Supplement 1 Issue Date: 4/4/85 DONALD C. COOK UNIT 2 CYCLE 5 ST, STEAM GENERATOR TUBE PLUGGING LIMITING BREAK LOCA-ECCS ANALYSIS: K(Z) CURVE Distribution M.J. Ades D.J. Braun J.C. Chandler
.R.A. Copeland N.F. Fausz J.S. Holm S.E. Jensen J.E. Krajicek G.F. Owsley G.L. Ritter H.G. Shaw R.B. Stout T. Tahvili AEP/H.G. Shaw (10)
USNRC/J.C. Chandler (41)
Document Control (5)
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