ML20090E917
| ML20090E917 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/28/1984 |
| From: | Chandler J, Kayser W, Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20090E885 | List: |
| References | |
| TAC-55445, TAC-55446, TAC-55816, TAC-55817, XN-NF-84-03, XN-NF-84-3, NUDOCS 8407200097 | |
| Download: ML20090E917 (67) | |
Text
_._
e EXHIBIT D
.!l XN-NF-84-03 u
Issue Date: 2/28/84 PRAIRIE ISLAND UNITS 1 AND 2 LIMITING BREAX LOCA-ECCS ANALYSIS WITH INCREASED ENTHALPY RISE FACTOR
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Prepared by:
7 %M'
)/18 /1939 T. Tahvili, Project Manager PWR Safety Analysis Concur:
194 4
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3 W. V. Kayter, Manager
, PWR Safety Analysis Concur:
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J. C. Chandler
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Reload Fuel Licensing Approve:
J f 3 Ph[V R. 3. Stout, Manager Licensing & Safety Engineering i
Approve:
v G. A. Sofer, Manager Fuel Engineering & Technical Services i
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gf E(ON NUCLEAR COMPANY,Inc.
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j-i XN-NF-84-03 J
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.1 TABLE OF CONTENTS
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Section Page
1.0 INTRODUCTION
AND
SUMMARY
1 3
2.0 LIMITING BREAK'LOCA ANALYSES.......................
4
.p 2.1 LOCA ANALYSIS MODEL...........................
4 2.2 RECOMMENDAJIONFOROPERATIONWITH VARIABLE F y..................................
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2.3 RESULTS.......................................
7
3.0 CONCLUSION
58 Ss
4.0 REFERENCES
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a; LIST OF TABLES y.,
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'A Table Page l
2.1 Prairie Island Units 1 and 2 System Data.............
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')j 2.2 Fuel Des ign P arame te rs...............................
10 F7 2.3 Prairie Island Units 1 and 2 TOPROD LOCA-ECCS i
Analys is Resul t s, Event T imes........................
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111 XN-NF-84-03 I
o LIST OF FIGURES r
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Figure Page 2.1 RELAP4/EM Blowdown System Nodalization for Prairie Island Unit 1 and 2....................
12 t
2.2 Axial Peaking Factor versus Rod Length, 0.4 DECLG Break....................................
13 n
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2.3 Hot Channel Factor Normalized Operating Envelope for Fg=2.28 w ith FaH=1.65.................
14 i
2.4 Enthalpy Rise Dependent Total Peaking Factor.......
15 U
2.5 Hot Channel Factor Normalized Operating
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Envelope for Fg=2.32 w ith Fay 1 1. 55...............
16 2.6 Hot Channel Factor Normalized Operating Envelope for Fg with Fag i 1.65....................
17 l >,
2.7 Downcomer Flow Rate During Blowdown Period, 0.4 DECLG Break....................................
18 1
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2.8 Upper Plenum Pressure during Blowdown i
Period, 0.4 DECLG Break............................
19 e
2.9 Average Core Inlet Flow during Blowdown
. Period, 0.4 DECLG Break............................
20 2.10 Average Core Outlet Flow during Blowdown Period, 0.4 OECLG Break............................
21 2.11 Total Break Flow during Blowdown Period, i
0.4 DECLG Break....................................
22 2.12 Break Flow Enthalpy during Blowdown, 1
0.4 DECLG Break....................................
23 J
2.13 Flow from Intact Loop Accumulator during i
Blowdown Per iod, 0.4 DECLG Break...................
24 1
2.14 Flow from Broken loop Accumulator during
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Blowdown Per iod, 0.4 DECLG Break...................
25
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P iv XN-NF-84-03 LIST OF FIGURES (Cont.)
Figure Page 2.15 Pressurizer Surge Line Flow during Blowdown Pe r iod, 0. 4 DECLG B re ak..,........................
26 2.16 Heat Transfer Coefficient during Blowdown Period at PCT Node, J.4 DECLG. Break, 0-15,000 MWO/MTM Case.............................
27 2.17 Clad Surface Temperature during Blowdown Period at PCT Node, 0.4 DECLG Break, 0-15,000 MWO/MTM Case.............................
29 2.18 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 0.4 DECLG Break, 0-15,000 MWO/MTM Case.............................
29 2.19 Average Fuel Temperature during Blowdown Period at PCT Location. 0.4 OECLG Break, 0-15,000 MWO/MTM Case.............................
30 2.20 Hot Assembly Inlet Flow during Blowdown Period, 0.4 OECLG Break, 0-15,000 MWD /MTM Case....
31 2.21 Hot Assembly Inlet Flow during Blowdown Period, 0.4 DECLG Break,15,000 MWD /MTM to EOL Case.......................................
32 2.22 Heat Transfer Coefficient during Blowdown Period at PCT Node. 0.4 CECLG 3reak, 15,000 MWD /MTM to EOL Case........................
33 2.23 Clad Surface Temperature during Blowdown Period at PCT Node. 0.4 OECLG Break, 15,000 MWD /MTM to EOL Case........................
34 2.24 Depth of Metal-Water Reaction during Blowdown Period at PCT Node, 0.4 DECLG Break, 15,000 MWO/MTM to EOL Case........................
35 2.25 Average Fuel Temperature during Blowdown Period at PCT Location, 0.a CECLG Break, 15,000 MWO/MTM to EOL Case........................
36
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,j LIST OF FIGURES (Cont.)
a F igure Page 2.26 Hot Assembly Outlet Flow during Blowdown Period, 0.4 DECLG Break, 0-15,000 MWD /MTM Case............
37 2.27 Hot Assembly Outlet Flow during Blowdown Period, 0.4 DECLG Break, 15,000 MWD /MTM to E0L Case.......
38 7
"l 2.28 Accumulator Flow during Refill and Reflood Periods, Broken Loop, 0.4 OECLG Break.............
39 2.29 Accumulator Flow during Refill and Reflood Periods, Intact Loop, 0.4 DECLG Break.............
40
[T 2.30 HPS! Flow during Refill and Reflood Periods, Broken Loop, 0.4 DECLG Break......................
41 2.31 HPSI Flow during Refill and Reflood Periods,
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Intact Loop, 0.4 0ECLG Break......................
42 2.32 LPSI Flow during Refill and Reflood Periods, j
Broken Loop, 0.4 OECLG Break......................
43 U
2.33 LPSI Flow during Refill and Reflood Periods, Intact Loop, 0.4 OECLG Break......................
44 2.34 Containment Back Pressure, 0.4 DECLG Break........
45 2.35 Normalized Power, 0.4 DECLG Break, 0-15,000 MWD /MTM Case.............................
46 2.36 Normalized Power, 0.4 CECLG Bresk,
_j 15,000 MWD /MTM to EOL Case........................
47 2.37 Reflood Core Mixture Level, 0.4 DECLG Break, 0-15,000 MWD /MTM Case.............................
48
.J 2.38 Reflood Downcomer Mixture Level, j
0.4 DECLG Break, 0-15,000 MWD /MTM Case............
49
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2.39 Reflood Uppe* Plenum Pressure.
0.4 DECLG Break, 0-15,000 MWO/MTM Case............
50 i
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vi XN-NF-84-03 LIST OF FIGURES (Cont.)
s Figure Page 2.40 Core Flooding Rate, 0.4 OECLG Break, 0-15,000 MWO/MTM Case.............................
51 2.41 Reflood Core Mixture Level. 0.4 DECLG Break, 15,000 MWO/MTM to EOL Case........................
52 2.42 Reflood Downcomer Mixture Level, 0.4 OECLG Break,15,000 MWO/MTM to EOL C ase....... -
53 2.43 Reflood Upper Plenum Pressure, 0.4 DECLG Break, 15,000 MWO/MTM to EOL Case.......
54 2.44 Core Flooding Rate. 0.4 OECLG Break, 15,000 MWO/MTM to EOL Case........................
55 2.45 7000EE2 Cladding Temperature versus Time.
0.4 DECLG Break,.,0-15,000 PWO/MTM Case............
56 2.46 T000EE2 Cladding Temperature versus Time, 0.4 DECLG Sreak, 15,000 MWO/MTM to EOL Case.......
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XN-NF-84-03
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1.0 INTRODUCTION
AND
SUMMARY
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This document presents analytical results for a postulated large break u
loss-of-coolant accident (LOCA), oerformed for the Prairie Island Units 1 and m
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[j 2 nuclear reactors. The analyses assume a reactor operating power of 1683 MWt (includes 25 power uncertainty), and use of Exxon Nuclear Company's (ENC's) l TOPR00 fuel.
The calculations were made for the double-endeo cold leg guillotine break, with a discharge coefficient of 0.4 (0.4 DECLG) identified in the previous analyses as the most limiting break.(1.2.3.4) l The analyses were performed using the EXEM/PWR ECCS evaluation model(5),
,3 with the MOD 7 computer model for evaluating the rod stored energy and l
fission gas release (6). The EXEM/PWR ECCS evaluation model includcs the NRC f
fuel swelling and flow blockage model, NUREG 0630.(15)
The analyses are i
4 4' applicable up to a five percent (5%) steam generator (SG) tube plugging, and j
' maximum peak pellet exposure limit of 55,000 MWO/MTM. The allowable linear heat generation rate, including the 1.02 factor for power uncertainty, was, i
14.76 kW/ft, corresponding to a total power peaking factor of 2.23 (Fq ), and i
total enthalpy rise of 1.65 (F[g). The peaking limits are appitcab.le over the l
entire exposure, i
The analyses were performed assuming an entico core with TOPR00 fuel.
Wtth respect to a LOCA, the TOPR00 fuel design is more 1imiting than prevtous ENC XN 1 and XN 2 reload fuel designs in Prairie Island Units 1 and 2.
This is due to the increased core flow area which reduces core reflood rates in the i
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LOCA analysis for TOPR00 fuel and results in higher PCTs. This analysis is j
i therefore appilcable to the Xfi 1 and XN 2 fuel designs for peak pellet burnups i
less than 55,000 MWO/MTM.
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XN-NF-84-03
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-I i-i The calculational basis and results are summarized in Table 1.1. The maximum calculated peak cladding temperature (PCT) is 2199CF, oc' curring at j
191 seconds into the accident at a location 9.37 feet from the bottom of the active core, with a total metal water reaction less than one percent. The i
~I 21990F PCT includes a 10F temperature addition due to the use of NRC interim
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upper plenum injection (UP!) modc1(7) as modified by Westinghouse (8).
The f
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results of the analyses show that within the limits established, the Prairie.
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!sland Nuclear Reactors operating at the stated power level, and with steam generator tube plugging up to 5%, satisfy the criteria specified by 10 CFR I
50.46( 9).
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l Table 1.1 Prairie Island Units 1 and 2 TOP 900 LOCA-ECCS Analysis Results 0 - 15000 paid /MTM 15000 - 55000 MWD /MIM Analys is Results Pet Pellet Exposure Pe d Pellet Exposure Pe A Clad Temperature (PCT), of*
2199 2060
8 Yim of PCT, sec. 191 249 pea Cled Temperaturt Location, f t.
9.37 9.37 Local Zr/110 Reaction (max.),1" 6.38.
4.37 2
Local Ir/tl 0 Location, f t. frtan bottos 9.37 9.37 2
Total it? Gexration,1 of total Zr Reacted
< l.0
< l.0 m
Hot Pal Eurst Time, sec.
30.20 38.80 Hot Rod Earst Location, f t.
6.0 6.25 Calculational Basis License Core Pomer, Wt 1650 Power Used for Analysis, WL"*
1683 pea L inear Power for Analysis, LW/f t"*
14.76 i
Total Peding Factor, fg 2.28 T
Enthalpy Rise, Eclear, F gi 1.65 g
g Steam Generator Tube Plugging (%)
5.00 g
Computer value at 330 secorals.
"* Inclading 1.02 f actor for power uncertaintles.
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4 XN-NF-84-03 m0 u
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,s 2.0 LIMITING BREAK LOCA ANALYSIS
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This report provides LOCA-ECCS analyses performed for Praiiie Island
....a Units 1 and 2 with a steam generator tube plugging up to 5"..
The analytical L
techniques used are in compliance with Appendix K of 10 CFR 50, and are q
described in the ENC WREM models(10), and the Emergency Core Cooling System Evaluation Model Updates: WREM-II(18), WREM-IIA (14) and EXEM/PWR(5),
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A LOCA break spectrum analysis was performed and reported in XN-NF a 46(1).
The limiting LOCA break was determined to be a large double-ended guillotine break of the cold leg, with a discharge coefficient of 0.4 (0.4 g
DECLG). The analyses performed and reported herein for the 0.4 DECLG break U
consider:
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(1) A revised stored energy model R00EX2(6) in place of the previously applied GAPEX(ll) model.
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(2)- The NRC upper plenum injection (UPI) interim model, developed by the NRC Staff (7) and modified by Westinghouse (8),
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(3) Updates to the latest Prairie Island Units 1 and 2 application to reflect all model revisions and documented in XN-NF-82-20(P), Rev.ision 1(5),
(4) The FLECHT/ ENC 2 WREM-II(18) heat transfer coefficient multipliers.
2.1 LOCA ANALYSIS MODEL
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The Exxon Nuclear Company EXEM/PWR ECCS evaluation model(5) was
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used to perform the analyses. This model consists of the following computer codes: RODEX2(6) code for initial rod stored energy and internal fuel rod gas
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inventory; RELAP4-EM(12) for the system blowdown and hot channel blowdown calculations; CONTEMPT-LT/22 as mooified in CSB 6-1(17) for computation of q
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5 XN-NF-84-03 containment backpressure; REFLEy(5,15) for computation of system reflood; and T000EE2(5,15,16) for the calculation of final fue] rod heatup.
The Prairie Island nuclear reactor is a two-loop Westinghouse l'
pressurized water reactor with an upper plenum injecticn and dry containment.
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The reactor coolant system ic nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow-paths or " junctions" as described in XN-NF-77-25( A)(17).
The system nadalization is depicted in
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F igur.e 2.1.
The pump performance characteristic curves are supplied by the NSSS vendor.
Five percent of the steam generatcr tubes are assumed to be plugged in each generator.
The transient behavior was determined from the governing conservation equations for mass, energy, and momentum.
Energy transport, flow rates, and heat transfer are determined from appropriate correl ations.
System input parameters are given in Table 2.1.
The reactor core is modeled with heat generation rates determined from reactor kinetics equations with r,eactivity feedback and with decay.
heating as required by Appendix K of 10 CFR 50. The chooped cosine axial power profile used for' the analyses is shown in Figure 2.2, with a maximum axial peaking f actor of 1.342, corresponding to a total peaking f actor of 2.25, and T
FIHof1.65. The Fg determined'with this axial profile in combination.with the_ current K(Z) function developed originally by the NSSS vendor is used to T
define the envelope for Fq, where the K(Z) curve is limited by large break LOCAs. Where small break LOCAs are limited, the K(Z) curve was modified such
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that Linear Heat Generation Rates (LHGRs) were determined by the NSSS vendor analyses. The K(Z) curve is represented in Figure 2.3.
The analysis of the loss-of-coolant accident is performed at 102 percent of rated power. The fuel design parameters are shown in Table 2.2.
~1.
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6 XN-NF-84-03 L
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Two cases of LOCA-ECCS calculations were performed with input which I
bounds the fuel history up to 55,000 MWD /MTM peak pellet exposure. The most J
limiting fuel conditions from beginning-of-life to 15,000 MWD /MTM (first mj case), and from 15,000 MWD /MTM to end-of-life (second case) were determined and used in each calculation.
Decay power, internal rod pressure and the i
d fission gas releases are highest at E0L, while the stored energy is calculated Q
to be highest at lower exposure (~2 MWD /KgU).
The combination of highest u
stored energy, rod pressure, and decay power was used to bound the LOCA-ECCS w
analysis over the exposure ranges shown.
The small rod diameter for ENC TOPR00 fuel, as compared to other fuel designs in the Prairie Island reactors, results in a larger core flow r?
area.
The larger core flow area decreases the core flooding rates, which y
results in higher PCTs.
Furthermore, the 55,000 MWO/MTM exposure limit P)-
considered in this analysis encompasses the exposure limits expected for the L
previous ENC XN-1 and XN-2 fuel designs operating in Prairie Island units.
Therefore, the LOCA-ECCS analyses reported in this document bound the previous Prairie Island ENC fuel designs.
2.2 RECOMMENDATIONS FOR OPERATION WITH VARIABLE F H The analysis reported in Reference 4 had analyzed the Prairie IslandLOCA-ECCSforatotalpeakingfactorof2.32(Fg)withFfsof1.55.
T T
The current analysis is performed with a total peaking f actor of 2.28 and FaH Since the change in Fg as a function of FfH between the previous and T
of 1.65.
1 the current analysis is small, it is recommended that a linear interpolation be used between the two total enthalpy rises (FfH) of 1.55 and 1.65 to evaluate thetotalpeakinglimit(F{);seeFigure2.4. Since the LOCA analysis is si r?
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7 XN-WF-84-03 performed to bound expected operation within the Fg andFfH limits, the analysis supports operation of the Prairie Island reactors if all rods or bundles are within these limits, irrespective of the number of rods or bundles at these 1imits.
To assure compliance to Appendix K criteria, K(Z) curves have been developed which define the total peaking limits as a function of the location of _ the axial peaking.
In this analysis, a K(Z) curve has been developed (Figure 2.3) which is applicable when the total enthalpy rise (F[g) is 1.65 i
and Fg is 2.28.
The current Prairie Island Units 1 and 2 Technical Specification K(Z) was developed for F[H of 1.55 with Fg of 2.32 (Figure T
T 2.5). ' The two K(Z), curves differ slightly. To simplify the monitoring of Fg as'a function of axial locations for F H less than or equal to 1.65, it is recomended that a new K(Z) curve (Figure 2.6) be incorporated in the Prairie Island Technical Specification.
The new curve has 'been developed to conservatively bound the values shown in Figures 2.3 and 2.5, and is identical to Figure 2.5.
2.3 RESULTS Table 2.3 presents the timing and sequer.ce of events as determined for the large guillotine break w ith a discharge coefficient of 0.4.
Comparison of these results with the previous LOCA-ECCS analysis for a TOPR00 fuel shows very slight change in the event times. Figures 2.7 through 2.15 present plotted results for system bicwdown analysis. Unless otherwise noted on. the figures, time zero corresoonds to the time of break initiations.
Figures 2.16 through 2.27 present results for the hot channel blowdown calculations. Figures 28 through 33 present the accumulators, HPSI and LPSI
7 J
8 XN-NF-84-03 O
flows during the refill and reflood periods of LOCA transient. Figure 2.34 presents calculated containment backpressure time history. Figure's 2.34 and 2.35 show the normalized power calculation results. The reflood calculation
..j results are shown in Figures 2.37 through 2.44.
q The maximum peak cladding temperature (PCT) calculated for the 0.4 1 i DECLG break is 21990F (Figure 2.45) and is calculated to occur at fuel rod
'l exposures less than 15,000 MWD /MTM.
This value includes a 10F temperature addition associated with the use of the NRC interim upper plenum injection n
(UPI) model as modified by Westinghouse. The maximum linear heat generation T
. rate is 14.76 kW/ft (Fg =2.28) 1or ENC TOPROD fuel. The maximum local metal-J.
water reaction in this case is 6.38% at 380 seccnds, and the total core metal-water reaction is less than 1%. The PCT location is at an elevation of 9.37
.1 [
feet from the bottom of active core. At fuel rod exposures between 15,000 and 55,000 MWD /MTM (EOL), the maximum PCT is calculated to be 20800F (Figure 2.46) including an 80F for UPI effect, occurring at 9.37 feet elevation relative to.
I' the bottom of the ' active core. The local metal-water reaction is 4.37% at 380 seconds, with a total metal-water. reaction of less than 1%.
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9 XN-NF-84-03 Table 2.1 Prairie Island Units 1 and 2 System Data Primary Heat Output, MWt 1650*
Primary Coolant Flow, lbm/hr 6.82 x 107 Primary Coolant Volt.me, ft3 10,247.** t Operating Pressure, psia 2,250.
Inlet Coolant Temperature, OF 530.
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Reactor Vessel Volume, ft3 2364.
Pressurizer Volume, Total, ft3 1000.
Pressurizer Volume, Liquid,.ft3 600.
Accumulator Volume, Total, ft3 (each of two) 2000.
Accumulator Volume, Liquid, ft3 1250.
Accumulator Trip Point Pressure, psia 714.7 Steam Generator Heat Transfer Area, ft2 48,925.i Steam Generator Secondary Flow, lbm/hr 3.54 x 106 Steam Generator Secondary Pressure, psia 724.7 Reactor Coolant Pump Head, ft 277.,
Reactor Coolant Pump Speed, rpm 1190.
2 Moment of Inertia, ibm-ft / rad 78,000.
Cola Leg Pipe, I.D.,
in 27.5
-Hot Leg Pipe, I.D.,
in 29.0 Pump Suction Pipe, I.D.,
in 31.0
- Primary Heat Output used in RELAP4-EM Model = 1.02 x 1650 = 1683 MWt.
- Includes total accumulator and pressurizer volumes, t Includes 5% SG tube plugging.
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10 XN-NF-84-03
'fd Table 2.2 Fuel Design Parameters
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7' Parameter ENC Standard TOPR00 3.
'l Cladding, 0.0., in.
0.426 0.417
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- Cladding, I.D., in.
0.364 0.358 I
Cladding Thickness, in.
0.031 0.0295 Pellet 0.D., in.
0.3565 0.3505 Diametral Gap, in.
0.0075 0.0075 Pellet Density, % TD 94.0 94.0
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Active Fuel Length, in.
144.0 144.0 Enriched UO, in.
144.0 132.0 2
Upper Blanket, in.
6.0 r,
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Lower Blanket, in.
6.0 Cell Water / Fuel Ratio 1.67 1.79 Rod Pitch 0.556 0.556
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J 11 XN-NF-84-03 Table 2.3 Prairie Island Units 1 and 2 TOPROD LOCA-ECCS Analysis Results, Event Times Event Time (sec.)
Start 0.00 Break Initiation 0.05 Safety Injection Signal 0.65 Accumulator Injection, Broken Loop
' 4.80 Accumulator Injection, Intact Loop 8.70 End-of-Bypass 21.05 Safety Injection Flow 25.60 Start of Reflood 36.80 Accumulator Empties, Broken Loop 39.95 Accumulator Empties, Intact Loop 43.95 Peak Clad Temperature Reached, 0-15,000 MWD /MTM Case 191.0 Peak Clad Temperature Reached, t
15,000 MWO/MTM to EOL Case 249.0 r
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o PRAIRIE ISLAND 152 TOPROD FUEL 5 XSG FQ=2 28 COD =. 417 F2=1.342 3
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RELATIVE HEIGHT Figure 2.2 Axial Peak ing Factor versus Hott Length, 0.4 DECLG !!reak G
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j Figure 2.3 Hot Channel Factor Normalized Operating Envelope for Fg=2.28 with FAH = 1.65 t
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2.24 2.22 1.45 1.50 1.55 1.60 1.65 1.70 1.75 T
Total Enthalpy Rise (FaH) i.
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Figure 2.4 Enthalpy Rise Dependent Total Peaking Factor 9
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Figure 2.31 IIPSI Flow during Refill and Reflood Periods, T
Intact Loop, 0.4 DECLG Break 6
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M figure 2.33 LPSI flow during Refill and Reflood Periods, h
Intact Loop, 0.4 DECLG Break 6
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TIME AFTER BREAK ( SEC 1 8
Figure 2.37 Reflood Core Mixture Level, 0.4 DECLG Break, 0-15,000 MWD /MTM Case
e PRAIRIE ISLAND HIP 2 REFLEX ( 110.40ECLG 5x SG TOPRC9 FUEL i
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F igure 2.39 Reflooti Upper Plenum Pressure, 0.4 DECLG Break, 0-15,000 MWD /MIM Case
PRRIRIE ISLAND HIP 2 REFLEXf11 0.4DECLG 5X SC TOPROD g
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8 F igure 2.42 Rel locul llowntomer Mixture Level, 0.4 DECLG Break,15,000 MWI)/MIM to E01. Case b
t.-
t.
3 2
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PRAIRIE ISLAND HIP 2 REFLEX ( 110.4DECLG Sx SG TOPROD FUEL i
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TIME AFTER BREAK ( SEC )
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1 PRAIRIE ISLAND HIP 2 REFLEX ( 1 ) 0. 4 0ECLG 5 2. SG TOPROD FUEL a
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TIME OFTER BREAK ( SEC )
8 f igure 2.44 Core I looiling Rate, 0.4 DErl.r. !!retak,15,000 MWD /MTM to EOL Case e
h I
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. TIME - SECONOS 8
Figure 2.45 T00DEE2 Cladding Temperature vs T ime, 0.4 DECLG Break, 0-15,000 MWD /MIM Case i
a 57 XN-NF-84-03 d
Ne2 4
95 M
9 5
i o
6 m
j g
w=
cw w
d w
im 4
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8 7
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+
58 XN-NF-84-03 3.0 CONCt.USION For breaks up to and including the double-ended severance of.a reactor coolant pipe, the Emergency Core Cooling System for both Prsfrie Island units will meet the Acceptance Criteria as presented in 10 CFR 50.46, with the 2.23 Fh and 1.65 Ffg limits.
The criteria are as follows:
(1)
The calculated peak fuel element clad temperature does not exceed the 22000F limit.
(2)
The amount of fuel element cladding that reacts chemically with water or steam docs not exceed i percent of the total amount of zirestoy in the reactor.
(3) The cladding temperature transient is terminated at a time when the core gecmetry is still amenable to cooling.
The hot fuel rod cladding oxidation limits of 17". are not exceeded during or af ter quenching.
(4)
The core temperature is reduced and decay heat is removed for an extended period of time, as requ ired by the' long 11ved rad ioact iv ity.
remaining in the core.
e
l l
59 XN-NF-84-03
(
r 4'
l t
I
4.0 REFERENCES
j 1.
ECCS Large Break Soectrum Analysis for Prstrie Island Unit 1 using ENG WREM-IIA FWR Evaluation Model, XN-NF-76-46, Novemoer 1978.
2.
Prairie Island Unit 2 Nuclear Plant Cycle 5 Safety Reoort, XN-NF-79-67, August 1979.
3.
LOCA ECCS Analysis for Prairie Island Unit 1 and 2 with ENC TOPR00 Fuel, XN-NF-50-49, Novemoer 12, 1950.
t 4.
Prairie Island Units 1 and 2 Limiting Break LUCA-ECCS Analysis Us ing E AEM/PWR, XN-NF-aJ-35, May 1953.
j 5.
Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Vodates, xN-NF-sz-zG(P), Revis ton 1. August 1382; supplement 1 March 1952; and Supplement 2, March 1982.
i 6.
R00EX2: Fuel Rod Thermal-Mechanical Resoonse Evaluation Model, XN-NF-51 55(P), Reviston 2, Feoruary 1983.
[
7.
U. S. Nut' lear Regulatory Ccmmission, " Safety Evaluation Reoort on
[
Interim ECCS Evaluation Model for Westingneuse Tuc-Leon Plants,"
Analysis Branch,' Division of System Safety, Office of Nuclear i
Reactor Regulation, November 1977.
8.
Letter, L. O. May(er to Director of Nuclear Reactor Regulation.
February 24, 1978 Oceket No. 50-232 and 50-306).
~
9.
" Acceptance Criteria for Emergency Core Cooling Systems for Light l
Water Cooled Nuclear Power Reactors," 10 CFR 50.46 tod Accendit K of l
10 CFR 50; Federst Register, Volume 39, Numcer 3. janu1ry 4,1974 10.
Exxon Nuclear Cemonny % REM-8ased Genecie 2WR ECCS Evibt:" 1: n '.'e c e l.
43-75-41, July 1975, ano supplements anc Aevtsions :nere:o.
11, g X: A Comouter Program for Predicting Pellet to-Cladding Hest e ransrer coefficients. XN-73-25, August IJ,1973.
i 12.
U.S. Nuclear Regulatory Commission Letter, T. A. Ippolito (NRC) to W. S. Nechodom LENC), "SER for ENC RELAP4 EM Update," March 1979.
j 13.
U.S. Nuclear Regulatory Commission, " Minimum Contsinment Pressure i
Model for PWR ECCS Performance Evaluation," Branch Tecnnical Position CSB 61.
14 E.txon Nuclear Ccmoany WREM-Based Generic PWR ECCS Evaluation Mocel t
Update enc WREM IIA, AN=NF 76-30( A), May 1979.
i t
1 0
~
e:
I o
j 60 XN-NF-84-03 7
I 15.
Exxon Nuclear Comoany ECCS Cladding Swelling and Rupture Model, XN-NF-62-U/(F), Kev is ion 1, August 1962.
.j 16.
G.
N.
L auben, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," NRC Report NUREG-75/057, May 1975.
q 17.
Exxon Nuclear comoany ECCS Evaluation of a 2-Loco Westinghouse PWR with Dry Containment Using the ENC WREM-II ECCS Model - Large Breu Example Problem, XN-t4F-77-25( A), Septemoer 1978.
.A 18.
Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-II, Xti-76-27, July 1976, XN-76-27, Supplement 1, Septemoer 1976, and XN-76-27, Supplement 2, November 1976.
"1 e'
-M e
m se.e q
=nm
%ep J e-R u