ML20097C008
| ML20097C008 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 08/30/1984 |
| From: | Chandler J, Kayser W, Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML14192A487 | List: |
| References | |
| XN-NF-84-72-S02, XN-NF-84-72-S2, NUDOCS 8409140231 | |
| Download: ML20097C008 (41) | |
Text
-
X N - N F 72 E SUPPLEMENT 2 l
H. B. ROBINSON UNIT 2 l
LARGE BREAK LOCA-ECCS AN ALYSIS g
WITH INCREASED ENTH ALPY RISE FACTOR:
K JZ' CURVE l
I AUGUST 1984 l
l RICHLAND WA 99352 s
ERON NUCLEAR COMPANY,INC.
ph:88m:8g;3
XN-NF-84-72 Supplement 2 Issue Date: 8/30/84
\
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H. B. ROBINSON UNIT 2 LARGE BREAK LOCA-ECCS ANALYSIS WITH INCREASED ENTHALPY RISE FACTOR: K(Z) CURVE Prepared by: 77 8/29 /8f T. Tahvili PWR Safety Analysis Reviewed by: Mkm W. V. K4yser, Manager F/19 // Y PWR Safety Analysis Concur: . ff 27!$
J. C. Chandler, Lead Engineer Reload Fuel Licensing
( bA .f Approve:
IV.
. If 1 B.' 3 tout, Manager d ht/6 [9' i Licensing & Safety Engineering 1 /
Approve:
/ ['h ce Av44'9 ue e r ng echnical Services L gf j ERON NUCLEAR COMPANY,INC.
l
NUCLEAR REGULATORY COMMIS$10N DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was denved 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 wlis utilize Exxon Nuclear fabricated reload' fuel or other technical 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 cussomers of Exxon Nudeer in their demonstration of compliance with the USN RC's regulations.
Without derogating from the foregoirg neither Exxon Nuclear nor any person acting on its behalf:
1 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 I any information, apparatus, method, or process disclosed in this document will not infnnge privately owned rights, or B. Assumes any liabilities with respect to the use of, or for derrages reeJtting from the use of, any information, ap-paratus, method, or process disclosed in this document. .
(
XN. NF. F00, 766 l
I n
i- XN-NF-84-72 Supplement 2 TABLE OF CONTENTS i
.Section- Page
(_
1 1.0 INTRCDUCTION ....................................... 1 2.0
SUMMARY
............................................ 2 3.0 K(Z) DETERMINATION ................................. 5 3.1 RESULTS OF'K(Z) DETERMINATION ................. 6
4.0 REFERENCES
......................................... 33 I
l
___-__m____._________ - _ _ - . _ - _ ..______ __ ___ . _ _ _
t 1
ii XN-NF-84-72 Supplement 2 l
LIST OF TABLES
. Table' Page 2.1-' H.B. Robinson Unit 2 K(Z) Determination Results ..... 3 3.1 H. B. Robinson Unit 2 System Data ................... 7 3.2 - Fuel. Design Parametiers .............................. 8 L
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H iii XN-NF-84-72 Supplement 2
( LIST OF FIGURES
=
Figure Page 2.1 Ngrmalized Axial Dependence Factor for 4 -
Ff=2.32versusElevationforFAH = 1.65 ..........
3.1 Axial Peaking Factor vs. Rod Length, 0.8 DECLG Break, Peaked at X/L=0.50, 0-EOL Exposure .......... 9 e
3.2 Axial Peaking Factor vs. Rod Length, 0.8 DECLG Break, Peaked at X/L=0.75, 0-9 MWD /kgU ............. 10 l 3.3 Axial Peaking Factor vs. Rod Length, 0.8 DECLG Break, Peaked. at X/L=0.75, 9-E0L MWD /kgU ........... 11 l
3.4 Heat Transfer Coefficient During Blowdown Period 3 at PCT Node, 0.8 DECLG Break, X/L=0.50, ,
0-E0L Exposure ..................................... 12 3 3.5 Clad Surface Temperature during Blowdown Period PCT Node, 0.8 DECLG Break, X/L=0.50, 0-E0L Exposure ..................................... 13 3 3.6 Depth of Metal-Water Reaction During Blowdown Period at PCT Node, 0.8 DECLG Break, X/L=0.50, -
0-E0L Exposure ..................................... 14 3.7 Average Fuel Temperature During Blowdown Period j
at PCT Node, 0.8 DECLG Break, X/L=0.50, j
0-E0L Exposure ..................................... 15 3.8 Hot Channel Average Quality, Center Volume, i 0.8 DECLG Break, X/L=0.50, 0-E0L Exposure .......... 16 3.9 Normalized Power, 0.8 DECLG Break, X/L=0.50, 17 0-E0L Exposure .....................................
h 3.10 T00DEE2 Cladding Temperature vs. Time, 4*
0.8 DECLG Break, X/L=0.50, 0-E0L Exposure .......... 18 l
N 3.11 Heat Transfer Coefficient During Blowdown Period
- at PCT Node, 0.8 DECLG Break, X/L=0.75, 0-9 MWD /kgU ........................................ 19 i -m d .
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'(;.i 3.12 Clad Surface Temperature During Blowdown Period . UQ PCT Node, 0.8 DECLG Break, X/L=0.75, 0-9 MWD /kgU ... 20 ./- - (. ,"
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3.13 Depth of Metal-Water Reaction During Blowdown Period at PCT Node, 0.8 DECLG Break, X/L=0.75,
? T . , ' + .:
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s/ 3.14 Average Fuel Temperature During Blowdown Period W..L' ,
- at PCT Node, 0.8 DECLG Break, X/L=0.75, l'
-I- 0-9 MWD /kgU ........................................ 22 4. V .e
..w, -
N ~'; :
s'1 3.15 Hot Channel Average Quality, Center Volume, .; ." ' '
- 0.8 DECLG Break, X/L=0.75, 0-9 MWD /kgU ............. 23 ,a . ". - .
3.16 Normalized Power, 0.8 DECLG Break, X/L=0.75, ~ i.J c.e
' T,; .
0-9 MWD /kgU ........................................ 24 (A
- p. , , .
- 3.17 T00DEE2 Cladding Temperature vs. Time, 0.8 DECLG
$ Break, X/L=0.75, 0-9 MWD /kgU ....................... 25 -h.ty[!
p[
yn ^
3.18 Heat Transfer Coefficient During Blowdown Period f V:b
- s. _ > J at PCT Node, 0.8 DECLG Break, X/L=0.75, .
c.
9-E0L MWD /kgU ...................................... 26 ' ;l J n a: .:.2
$~- 3.19 Clad Surface Temperature During Blowdown Period 7.d".
- e y PCT Node, 0.8 DECLG Break, X/L=0.75, 74.
if. 9-EOL MWD /kgU ...................................... 27 llW;.:
p 3.20 Depth of Metal-Water Reaction During Blowdown i '. P -
f '.? Period at PCT Node, 0.8 DECLG Break, X/L=0.75, . ,g
@ 9-E0L MWD /kgU ...................................... 28 :.:; '- M
^' -
& 3.21 Average Fuel Temperature During Blowdown Period at j . ". $
PCT Node, 0.8 DECLG Break, X/L=0.75, V ' . .. >
9-E0L MWD /kgu ...................................... 29 f ,' ~
R ' { f') .
- 1 .
3.22 Hot Channel Average Quality, Centar Volume, .f j
, :. - 0.8 DECLG Break, X/L=0.75, 9-E0L MWD /kgU ........... 30 .. &
. e A. . ,
N@
3.23 Normalized Power, 0.8 DECLG Break, X/L=0.75, 9-E0L MWD /kgU ...................................... 31 RP
.eg?, y: 4.a-E 3.24 T000EE2 Cladding Temperature vs. Time, 0.8 DECLG
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' 1 XN-NF-84-72 Supplement 2
1.0 INTRODUCTION
7 LOCA ECCS analyses were completed by Exxon NuCear Company (tNC) and documented in XN-NF-84-72(1) in July 1984 and XN-NF-84-72, Supplement 1(2) in 3
August 1984 for H.B. Robinson Unit 2. These analyses were performed as a result of the decision by Carolina Power & Light (CP&L) in 1983 to: (1) replace the H.B. Robinson Unit 2 steam generators, (2) implement a low radial leakage
, fuel management scheme in order to reduce vessel fluences and thereby allevaite concerns about thermal shock, and (3) to increase the peak assembly e discharge exposure for H.B. Robinson fuel to 44 MWD /kgU (49 MWD /kGU peak
- pellet). To implement the low radial leakage fuel management scheme, the tota nuclear enthdpy rise (F{H) was increased to 1.65 from 1.55. The analysis was performed with a linear heat generation rate (LHGR), including i the 1.^2 f actor for power uncertainty, of 14.16 kW/f t, corresponding to a
[ total power peaking of 2.32 (Fh). The analysis supported operation of the H'.B. Robinson Unit 2 reactor at 100% power (2300 MWt) with up to 6% steam tube
? plugging. The limiting break was identified to be a guillotine Dreak with a discharge coefficient of 0.8 (0.8 DECLG).
This supplement documents results of large break LOCA ECCS calculations
}
j performed to determine the permissible limits in linear heat generation rate
=
^
versus axial location, i.e., the K(Z) curve. The small break portion of the K(Z) curve is based upon results of analyses performed by ENC in 1975.(3) The i resuits of the large break calculations are presented in Section 3.0.
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-7
XN-NF-84-72 2
Supplement 2 2.0
SUMMARY
The calculational basis and results of the K(Z) determination are presented in Table 2.1. The results of the analysis satisfy the Acceptance Criteria as presented in 10 CFR 50.46, and therefore support operation of H.B.
Robinson Unit 2 with a radial peaking corresponding to FAH of 1.65. Two K(Z)
I curves were determined: the first curve is applicable for fuel rod exposures less than 9 MWD /kgU, and the second curve is applicable for fuel rod exposures greater than 9 MWD /kgU. The reduction in fuel average temperatures at high exposures results in higher allowed LHGRs at the higher fuel rod exposures.
The K(Z) curves are given in Figure 2.1. The small break portion of the curve is drawn to provide an equivalent LHGR at the top of the core to that j previously analyzed and documented in Reference 3.
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Table 2.1- H.B. Robinson Unit 2 K(Z) Determination Results-Calculational Basis License Core Power, MWt 2300
- Power Used for Analysis, MWt** 2346 Break Size, DECLG 0.8
' EnthalpyRise, Nuclear,FfH 1.65 Steam Generator Tube Plugging, % -6.00 Peaked Peaked Peaked X/L = 0.50' X/L = 0.75 X/t = 0.75
- w Hot Rod Exposure Range, MWD /kgu 0 - E0L 0-9 EOL Peak Linear Heat Generation Rate (LHGR)** -14.16 12.39 12.57' TotalPeakingFactor,Fh. 2.32 2.03 2.06 Peak Cladding Temperature, OF 2042 2197 2183 Peak Temperature Location, ft 6.0 10.75 10.75 Local Zr/H 2O Location, ft 6.0 10.75 10.75 Local Zr/H 2O Reaction (Max.), %* 4.65 6.19 5.89
<1% <1% <1%
Total Zr/H 2O Hot Rod Burst Time, sec 39.9 49.37 51.57 Hot Rod Burst Location, ft 6.0 9.0 9.0 EE Wi
$ !n
- Computer value at 380 seconds. Ef
- Including 1.02 factor for power uncertainties.
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'3.0'_K(Z) DETERMINAfION L The K(Z) curve defines the limit-on linear heat generation rate (LHGR)
- versus' axial elevation in the core. The LHGRs are reduced at the top of the core toL offset
- the effect on peak cladding temperature (PCT) of reduced
- coolant heat transfer from (1) the short uncovery periods at the top of the f fcore during small break LOCAs, and (2) reduced coolant heat capacity at the top of the core during the reflood period of the large break LOCAs.
The current analysis used the EXEM/PWR LOCA ECCS :nodels(4) with operating conditions and fuel' parameters given in Tables 3.1 and 3.2. The
-operating conditions and fuel parameters are identical to those used in the XN-NF-84-72 analysis. Boundary condition input from the XN-NF-84-72 blow-
. down,. refill and reflood analysis was used to drive tne RELAP4/ hot channel and T00DEE2/ hot rod heatup calculations. .
-The analyses were performed with a center peaked chopped cosine axial power distribution and two skewed power distributions peaked at three-quarters (X/L=0.75) of- core height. The axial power shapes used in the analysis ar'e given in Figures 3.1 to 3.3. The axial power profiles were used to define the large break portion of the K(Z) curve (Figure 2.1). The top
- - ' peaked profile defined the K(Z) curve above three-quarters of core height
.(large break portion of K(Z) curve only), while the chopped cosine profile
? defined the K(Z) curve at X/L=0.5. Between X/L of 0.5 and 0.75, the allowable LHGR'is linearly interpreted between the values given at X/L of 0.5 and 0.75.
The K(Z) curve in the top tenth of the core is defined by the LHGRs used in the small break' analysis performed by-ENC in 1975.(3)
6 XN-NF-84-72 Supplement 2 3.1 RESULTS OF K(Z) DETERMINATION The results of the K(Z) determination are summarized in Table 2.1 for the . l'imiting break, the 0.8 DECLG break. RELAP4/ hot channel and T000EE2/ hot rod" heatup calculations were performed. The results are presented in Figures 3.4 to 3.10 for the center peaked profile. The results are applicable for all exposures. Results for the skewed profile are .
presented in Figures 3.11 to 3.17'and 3.18 to 3.24 for the less than 9 MWD /kgU and greater than 9 MWD /kgU peak rod exposure cases, respectively. Time zero 1 on all plots correspond to the time of break initiation.
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7 XN-NF-84-72 Suppleaent 2 L
- Table 3.1 H. B. Robinson Unit 2 System Data L
- Primary Heat Output, MWt 2346*
. Primary Coolant Flow, lbm/hr 100.3 x 106 Primary Coolant Volume, ft3 9768**
Operating Pressure, psia 2,250 Inlet Coolant Temperature, OF_ 546.2 Reactor Vessel Volume, ft3 3660 Pressurizer Volume, rotal, ft3 1300 Pressurizer Volume, Liquid, ft3 780 h
Accumulator Volume, Total, ft3 (each of three) 1200 Accumulator Volume, Liquid, ft3 825 Accumulator Trip Point Pressure, psia- 615
~
i Steam Generator Heat Transfer Area, ft2 (one) 40,859**
' Steam Generator Secondary Flow, Ibm /hr (one) 3.37 x 106 j- Steam Generator Secondary. Pressure, psia ._
800
-Reactor Coolant Pump Head, ft 264 Reactor Coolant Pump Speed, rpm 1180-Moment of Inertia, lbm-ft / 2 rad 70,000
! Cold Leg Pipe, I.D., in _
27.5 Hot Leg Pipe, I.D., in 29.0 Pump Suction, Pipe, I.D., in 31.09 n
- Primary Heat Output used in RELAP4-EM Model = 1.02 x 2300 = 2346 MWt.
- Includes 6% SG tube plugging.
F-l
^8 XN-NF-84-72 Supplement 2 Table 3.2 Fuel Design Parameters Parameter ENC Fuel Cladding, 0.D., in' O.424 Cladding,'I.D., in 0.364 Cladding Thickness, in 0.030 Pellet 0.D.,-in 0.3565 Diametral Gap, in 0.0075
. Pellet Density,.% TD 94.0
' Active Fuel Length, in 144 Enriched U0 2. in 132 Upper Blanket, in 6.0 Lower Blanket, in 6.0 Cell Water / Fuel Ratio 1.76 Rod Pitch 0.563
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4.0 REFERENCES
-,.*.:.i-<(;:.-
,, ..s y
{ (1) "H.B. Robinson Unit 2 Large Break LOCA-ECCS Analysis with Increased ' %.7 , 9. .
l Enthalpy Rise Factor," XN-NF-84-20, Exxon Nuclear Company, Rich- gn :
land, WA 99352, July 1984. f.Z, . 9
- t. : - ;. - [ '
(2) "H.B. Robinson Unit 2 Large Break LOCA-ECCS Analysis with Increased -f+ #.q Enthalpy Rise Factor: Break Spectrum," XN-NF-84-20, Supplement 1, 1 7. 1. +,
Exxon Nuclear Company, Richland, WA 99352, August 1984. .-.Ed.
. n.7, . ~p>
s;.=
(3) " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation 3 2.@i -l 2;.: .
Model: Volume III Small Break Model," XN-75-41, Volume III, i.
, Revision 2, Exxon Nuclear Company, Richland, WA 99352, August 1975. 2 ![=:- 4 '
2 i Q.hs,k (4) " Exxon Nuclear . Company Evaluation Model EXEM/PWR ECCS Model Up- %. W. ~
dates," XN-NF-82-20(P), Exxon Nuclear Company, Richland, WA 99352, f. ., .:
February 1982. 7. L:,6;:. %,
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