ML20065L871
| ML20065L871 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/31/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20065L868 | List: |
| References | |
| 82NEDO80, NEDO-22192, NUDOCS 8210190785 | |
| Download: ML20065L871 (31) | |
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GENERAL ELECTRIC BOILING s WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR :
QUAD CITIES NUCLEAR POWER STATION UNIT 1 CYCLE 7 AND UNIT 2 CYCLE 6 l
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NEDO-22192 DRF L12-00540 82NED080 Class I July 1982 GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR QUAD CITIES NUCLEAR POWER STATIN UNIT 1 CYCLE 7 AND UNIT 2 CYCLE 6 i
I NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CAUFCRNI A 951 5 GENER AL h ELECTRIC 1
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1 NE30-22192 DfPORTANT NOTICE REGARDCZ CONTENTS OF THIS REPORT
- PLEASE READ CAREFULLY This report was prepared by General Electric solely for Commonwealth Edison Company (CECO) for CECO's use in supporting the operation of Quad Cities Nuclear Power Station. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obcained or provided to General Electric at the time this report was preparsd.
1 '
The only undertakings of the General Electric Company respecting information in this document are contained in the General Electric Company Load Line Limit Analysis Proposal No. 424-TY590-EEO, October 29, 1981. The use of this infor-mation 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 unauth-orized 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 suel, use of such information may not infringe privately owned rights; nor do they aasume any responsibility for liability or damage of any kind which may result from such use of such information.
CONTENTS faga,
- 1.
SUMMARY
1_t Z. INTRODUCTION 2-1
- 3. DISCUSSION 3-1 3.1 Background 3-1 3.2 Analytical Basis 3-1 3.3 Aaalysis and Results 3-3 3.3.1 Stability 3-3 3.3.2 Loss-of-coolant Accident . 3-4 3.3.3 Pressurization Transientr' 3-4 3.3.4 Rod Withdrawal Error 3-6 4 REFERENCES 4_t I
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TABLES ,
Table Title Page ..
3-1 Plant Characteristics 3-7 --
3-2a Transient Input Data and Operating Conditions for License Basis Point 3-8 .
i 3-2b Transient Input Data and Operating Conditions for 100%
Intercept Point 3-9 I 3-3a GETAB Analysis Initial Conditions for License Basis Point 3-10
=
3-3b GETAB Analysis Initial Conditions for 100% Intercept Point 3-11 -l 3-4 ASME Pressure Vessel Code Compliance: MSIV Closure, Flux 3 cram 3-12 3-5 Transient Sunnary--Turbine Trip Without Bypass 3-14 3-6 Transient Summary--Load Rejection Without B ypass 3-15 3-7 Transient Summary--Loss of Feedwater Heating 3-16 3-8 Transient Summary--Feedwater Controller Failure 3-17 3-9 Transient Summary--High Pressure Coolant Injection 3-18 R
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ILLUSTRATIONS Figure Title Page
- 1-1 Quad Cities Proposed Operating Power / Flow Map 1-2 3-1 Quad Cities Operating Power / Flow Map as Shown in FSAR 3-19 3-2 Axial Power Shape for 100% and 87% Core Flow. Plant H 3-20 3-3 Void Reactivity versus Delta Void for LR w/o BP at 100/100 and 100/87, Plant H 3-21 L
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1 NEDO-22192 l
- 1.
SUMMARY
1 This report justifies the expansion of the operating region of the power / flow 4
. map for Quad Cities Nuclear Power Station Unit 1 Cycle 7 and Unit 2 Cycle 6.
The underlying technical analysis is referred to as the Extended Load Line Limit Analysis (ELLLA).
Previous analyses of this type, the Load Line Limit Analysis (LLLA) for BWR/3's did not include analyses for raced power reluced flow operation, and for BWR/4's routinely included analyses at rated power and mini =um flows of 91 to 94% of rated. Iti early 1981, an ELLLa was performed for a typical BWR/3 to support operation at rated power with flow as low as 87%. This work draws on the previous analyses to develop a set of restricted generic conclusions regarding l
applicability of the license basis safety analyses to operation within this expanded domain (Figure 1-1). It is further shown that the Quad Cities Nuclear Power Stations Unit 1 Cycle 7 and Unit 2 Cycle 6 for the current GE fuel type meets the conditions of validity of the generic conclusions. The consequences of events initiated from within the extended domain are bounded by the conse- ->
' quences of the same events initiated frem the license basis condition.
Recent analyses (Reference 1) justify the modification of the operating envelope defined by the power / flow curve while remaining within previously established operating limits and the Preconditioning Interim Operating Management Recom-t, mendations (PCIOMRs). The operating envelope is modified to include the extended operating region bounded by the 108% APRM rod block line, the rated power line, and the rated load iine.
The discussion and analyses presented show that all safety bases nor= ally
, applied to Quad Cities Nuclear Power Station Units 1 and 2 are satisfied throughout Unit 1 Cycle 7 and Unit 2 Cycle 6 for operation within this envelope.
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NED0-22192 i
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t40 100% POWER UNE
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100% INTERCEPT POINT (100/87)
APRM ROO ELOCK UNE (193/1001 (PROPOSEDI G3.SSN + 50%I (100/1001 TYPtCA L POWER ASCENSION PATH ANALYSl5 NEEDED TO C QPERATE IN THIS REGtON I TYPtCAL 100% POWERI 100% PLOW LOAO UNE
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MLNtMUM PUMP SPEED UNE I
40 1 NATURAL CincutATiON k
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50 60 70 30 90 500 110 to :0 20 40 ConEPLcw<%
Figure 1-1. Quad Cities Proposed Operating Power / Flow Map 1-2
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NEDO-22192 . . .
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I 2.
INTRODUCTION
' . . {.
Two factors which restrict the flexibility of a BWR during power ascension in .
proceeding from the low-power / low-core-flow condition to the high-power /high- t .
core-flow condition are: (1) the FSAR power / flow curve, and (2) PCIOMRs. a 4 ; $. :.
If the rated load line control rod pattern is maintained as core flow is ..
j increased, changing equilibrium xenon concentrations will result in less than I; -
rated power at rated core flow. In addition, fuel pellet-cladding interaction ,, ] . - .c considerations inhibit withdrawal of control rods at high power levels. The %=.P .
combination of these two factors can result in the inability to attain rated $.),4 core power directly. p i .. '
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This report provides the analytical basis for Quad Cities Nuclear Power y,f, Station Units 1 and 2 operation during Cycle 7 Unit 1 and Cycle 6 Unit 2 under .'3 . . .
a modified operating envelope to permit improved power ascension capability to full power within the desin bases previously applied. '?'.
h The operating envelope is nodified to include the extended operating region >-
j'i [d -g bounded by the 108% APRM rod block line, the rated power line, and the rated .
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- 3. DISCUSSION
3.1 BACKGROUND
Operation of the Quad Cities Nuclear Power Station Unit 1 Cycle 7 and Unit 2 Cycle 6 utilizing the power / flow map is described in Chapter 3 of the FSAR (Referrete 2). This section of the FSAR describes the basic operating envelope (Figure 3.2.3) within which normal reactor opeations are conducted and provides
( the basic philosophy behind the power / flow curve. FSAR Figure 3.2.3 is repro-duced as Figure 3-1 of this document.
This analysis expands the operating domain along the 108% APRM rod block
- line to 100% power at 87% flow. Rated power operation at any flow between 87% and
) 100% is acceptable within the constraints of the rod block monitor system.
Figure 1-1 shows the proposed operating map.
Certain terminology from the previous Load Line Limit Analyses is retained herein:
Rod Block Intercept Point - 85% power /61% flow.
100% Intercept Point - lowest flow point at which rated power operation i is acceptable. (87% flow for Quad Cities Nuclear Power Station.)
Rod Intercept Line - a straight line between the Rod Block Intercept Point and the 100% Intercept Point. Because the latter point lies on the APRM Rod Block line, no Rod Intercept Line exists for Quad Cities Nuclear Power Section.
3.2 ANALYTICAL BASIS
( To provide relief from the operating restrictions inherently imposed during ascension to power by the existing power / flow curve and ?CIOMRs, a modified power / flow curve has been derived. In deriving this operating curve, five design basis objectives were specified:
- RB = 0.58 W+50% where W is recirculation flow in percent of rated.
3-1
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- 1. For those transients and accidents that are sensitive to variations [ _
in power and flow, the 100% power /100% flow (licensing basis for BWR/2 g and 3's) point and the 105 power /100% flow (licensing basis for BWR/4) I point must be shown to be a more limiting condition than any condition }
within the expanded operating region (i.e. , the shaded region of Figure 1-1). __
'^ '
- 2. In no instance shall the ratio of power to flow intentionally exceed the ratio defined by the APRM rod block line. .
- 3. The slope of the APRM rod block line must be such that flow increases are capable of compensating for renon buildup while increasing reactor [ R power. =_- 3 n
- 4. The consequences of all accidents and transients analyzed in the FSAR and subsequent amendments and the license submittals must remain -
within the limits normally specified for such events. I _
- 5. Reactor power ascension from minimum recirculation pump speed to full power shall be directly attainable through combined control rod move-ment and recirculation flow increase without violation of either the _
power / flow line or PCIOMits.
i _
To meet these objectives, analyses were performed for Quad Cities Nuclear - -
Power Station and other typical BWRs. From these analyses conclusions were drawn concerning the safety conser,uences of operation in the extended operating --
region (shaded area of Figure 1-1). It was shown by specific analyses of the $
current GE fuel type that these conclusions were applicable to Quad Cities Nuclear Power Station, Unit 1 Cycle 7 and Unit 2 Cycle 6 (QCNPS-lC7, 2C6). _
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3.3 ANALYSIS AND RESULTS .-
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3.3.1 _S tability. .-is
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3.3.1.1 Channel Hydrodynamic Conformance to the Ultimate Performance .'..
Criterion . ,- .,
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The channel performance calculation for Quad Cities Nuclear Power Station, g .t . .
Unit 1 Cycle 7 and Unit 2 Cycle 6 (QCNPS-lC7, 2C6) is presented in References 3 3 (?.;, . -
3' and 4. The decay ratios are reproduced below: - ~ ' "
Extrapolated Rod Block Line* ^
- {_ .C . . ; . . ,
Channel Hydrodynamic Performance Natural Circulation Line A%4 -
Unit Cycle Channel Type "I *2 o F-1 7 P8x8R Channel 0.18 ( .( ,' '
8x8 Channel 0.28 .
2 6 P8x8R Channel 0.17
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8x8 Channel 0.27 1 1.-
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At this most responsive condition, the most responsive channels are clearly e -
within the bounds of the ultimate performance criteria of 11.0 decay ratio at 7 ,
all attainable operating conditions.
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3.3.1.2 Reactor Conformance to Ultimate Performance Criterion .M --+
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The decay ratios determined from the limiting reactor core stability conditions , L 't . .
are presented in References 3 and 4. The most responsive case fo- this analysis [ ,.
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is the extrapolated rod block line* - natural circulation conditions. -
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Unit Cycle Extrapolated Rod Block Line* y, 4 ,
Natural Circulation Power -
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Core Decay Ratio, x2 /*o # 'j D ' '. .
1 7 0.57 , ? 1.l .- t-2 6 0.53 ITY? '.
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NEDO-22192 These calculations show the QCNPS reactors to be in compliance with the ultimate performance criteria, including the most responsive condition.
3.3.2 Loss-of-Coolant Accident A discussion of low-flow effects on LOCA analyses for all operating plants (Reference 5) has been presented to and was approved by the NRC (Reference 6).
The LOCA analysis for QCNPS-lC7, 2C6 (contained in Reference 7) is applicable in the power flow domain discussed in this report.
3.3.3 Pressurization Transients As shown in Reference 2, the most limiting transient for QCNPS-lC7, 2C6 is the Load Rejection without bypass. The results of numerous transient evaluations (Tables 3-1 through 3-9) at various power / flow conditions demon-strate that transients originated from within the extended operating domain are less severe then the limiting transient at the license basis condition.
This trend was specifically demonstrated for QCNPS-lC7, 2C6 by analyzing the Load Rejection w/o Bypass, Feedwater Controller Failure, and MSIV Closure with Flux Scram events at the limiting point in the extended region (100/S7).
and comparing the results to those for the licensing basis point (100/100).
Those comparisons are shown in Tables 3-4, 3-6, and 3-8, and show that the (100/87) point results are bounded by the licensing basis results.
3.3.3.1 Changes in Nuclear Characteristics The end-of-cycle (EOC) conditions for the various plants and power / flow con-dicions were calculated in dif ferent ways depending on the plant cycle operat-ing plan. For Plant H (see Tables 3-1 through 3-9), the 100/100 EOC point was determined by assuming rated operation (100/100), and by a Haling power shape throughout the cycle (normal practice). The reduced flow points were deter-mined by using the same exposure point and simply reducing the flow. In this case, the exposures for all three points (100/100, 100/92, and 100/87) were identical, only the power shapa chan6ed. For other plants, different 3-4 his----- m
l l NEDO-22192 I
i combinations of Haling " burns" were assumed resulting in unique exposures for l each power / flow combication.
f i From a transient viewpoint, the important nuclear characteristics which are affected when changing from a high to low flow condition (100/100 to 100/87 or 100/111 to 100/100, etc.) are the scram and void reactivities.
The scram response improves (more negative reactivity) when the flow is reduced. This results because as the flow is reduced, the boiling boundary moves lower in the core, thus causing the axial power shape to peak more toward the bottom (Figure 3-2). This, in turn, results in a stronger scram response because the control rods become " effective" earlier during insertion.
r The impact on void reactivity, of changing between high and low flow condi-tions, is primarily affected by exposure. Since the high and low flow f conditions represent only a slight change in exposure, it is expected that the void reactivity characteristics should be very similar. This trend can be observed by comparing the graphs of Figure 3-3.
3.3.3.2 Evaluation of Transient Results This section provides transient result comparisons between high and low flow j inicial conditions for various plants, and justification for extending the conclusions reached to QCNPS-lC7, 2C6.
The transient results of primary importance for this study are ACPR and peak vessel pressure. Either of these have the potential to impact operation.
To ensure that the reduced flow condition (100/87) is bounded by the reference j licensing condition (100/100), it is necessary to consider dCPR and the peak vessel pressures.
It was established that the reduced flow condition has an improved scram characteristic and similar void reactivity. Therefore transients originated k from the reduced flow condition should exhibit a marked improvement. The results for QCNPS-lC7, 2C6 demonstrated this trend as shown in Tables 3-4
) through 3-9.
3-5
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?E NEDO-22192
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55 The Plant H results for 100/100, 100/92, and 100/87 also show a clear trend E-EE --
of decreasing ACPR with decreasing flow for both LR w/o BP and FWCF. The peak vessel pressure for the MSIV flux scram event was unchanged between L
100/100 and 100/92 (the 100/87 condition was not evaluated). __
ILc __
Similar trends exist for other typical BWRs as shown in the above tables. _
3.3.4 Rod Withdrawal Error ? !
The effective RBM setpoint is a function of power and flow. Above the rated -
rod line, the rod block will occur with less rod withdrawal. Thus, the evalua- d_
tion at rated is conservative for operation above the rated load line. : __
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Table 3-1 PLANT CIIARACTERISTICS Flaat A B t! D E F G H K L M u ocups.3g3 "r est Fuel S.w=llem 560 764 560 764 36 8 240 560 484 548 560 St,0 764 724 m teJ TI.ernal Paneur (HWL) 24 % jl9 3 24 % 3293 1593 997 24 % 1670 2381 24 % 24 % 3293 2581 sat =J Core Flama (Mib/hr) 77.0 102.5 77.0 102.5 49.0 29.7 77.0 57.6 73.5 78.5 77.0 100.0 93.0 kellet Valve 5e t y.sint 8090 !!OS 1805 1805 1990 1065 1090 llos 1080 1080 1090 !!!0 1830 (peig)
Relief Valve Cap.s t a y 11/85.7 II/66.0 11/87.4 11/66.0 6/72.0 4/?9.0 88/89.6 7/83.0 7/57.5 !!/85.7 11/85.7 16/99.0 4/22.5 (No./1Nau)
Saf et y Valve S -s pulut -
1250 - 1230 1240 1280 - -
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g simetrol koJ prive bis 675 675 675 675 678 6 MST* 678 675 678 675 678 678 675 @
Sper ii i rat lame N i N~87e'mer t t s.ne) = 0.175 (5). 0.7% (20). 1.57 (50). 2.75 (90).
- 200 mace laterrogation Jelay.
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Table 3-2b TRANSIENT INPUT DATA AND OPERATING CONDITIONS FOR 100% INTERCEPT POINT Plant A B s' ga F F 6: M* K L M OCNeS-l 998rs-2 T'eersaal Puent (HWs/I) 24 h/100 3293/100 24 h/lon 1299/800 159$/100 N/A 24 M/l(ut 1670/100 2181/100 2436/300 3291/100 2511/100 2511/100 Sea.ee pluw (Mib7bs/I) 1u.47/300 13.42/300 lo.47/ loo 13.1c/100 6.84/300 _
80.47/300 6.77/luo 9.$7/100 10.01/100 13.46/300 9.8/300 9.8/100 there Fluw (Mab/br/I) 72.4/94 91.1/94 72.4/94 98.1/94 4%.2/92.2 72.3/94 $1.0/92 71.?/94 73.8/94 87.0/87 4%,1/87 85.3/87 em =c Perms.are (pmes) aus2 Ball 1034 lull 1014 1021 1022 1084 108) a003 1003 1005 insbene Frea.neere (psig) 9%7 978 9%9 958 960 967 977 959 954 948 950 950 falar had Caell es Arms -9.16 -6.9% -W.6% -8.90 -9.98 -7.6% --
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3-11 1
Table 3-4 ASME PRESSUng VESSEL CODE COMPLIANCE: MSIV CLOSURE, FLUX SCPAM Peak Peak .
Steamline Vessel ..
Puk Pressure Pressure -
h eron Peak hat si v Initial Flux 9 Flux Q/A
(% initial)_ (psig) (psig) . . r-Plant 'Poder/ Flow (% initial) 127 1217 1264 (104P ,100F)* 849 '-
228 1211 1254 A (100P, 94F) 1005 126 1186 1211 (85P, 61F) 834 122 1242 1277 (104P, 100F) 491 122 1226 1260 3 (100P, 947) 521 120 1192 1217 (85P, 61F) 504 _
1263 (104P, 100F) 741 125 1218 131 1211 1252 C (100P, 94F) 860 '
706 131 1189 1214 (85P, 61F) -:
783 125 1266 1295 (104P, 100F) 797 130 1251 1280 D (100P, 91F) 405 124 1193 1217 (85P, 617) 770 126 1245 1287 (104P, 100F) 126 1247 1271 E (100P, 92.2F) 939
~
124 1196 1217 (85P, 61F) 897 -;
617 129 1260 1303 F (100P, 111F) 677* 122 1203 1234 (104P, 100F) 632* 122 1201 1231 -
(100P, 94F) 507* 123 1180 1205 G (91P, 757) 405* 123 1182 1199 (85P, 61F) 702* 122 1202 1235 (104P, 1057) ,
~
- The 105% steam-flow licensing basis point corresponds to approximately -
104% power.
3-12
I NEDO-22192 Table 3-4 ASME PRESSURE VESSEL CODE COMPLIANCE: MSIV CLOSURE, FLUX SCRAM l
(Continued)
Peak Peak p Steamline Vessel
( Neutron #***"#* #* "#*
Peak Heat Initial Flux 9 Flux Q/A si v Plant- Power / Flow (% initial) (% initial) (psig)
I (psig)
(100P, 100F) 658* 128 1222 1244 (100P, 92F) 662* 127 1223 1243
> H (100P, 87F) 635* 127 (92P, 757) 525* 128 1207 1228 (85P, 61F) 444* 128 1187 1207 (104P, 100F) 446* 124 1243 1270 (100P, 94F) 440* 122 1234 1261 I
(104P, 100F) 568* 120 7.199 1232 g (100P, 94F) 538* 122 1192 1224 (91P, 75F) 421* 120 1169 1195
- (85P, 61F) 348* 120 1164 1183 (104P, 105F) 576* 120 1199 1233 (104P, 100F) 693* 124 1236 1275 (100P, 94F) 616* 124 1229 1266 N (105P, 100F) 602 124 1246 1276 (100P, 87F) 523 123 1239 1266 QCNPS- (100P, 100F) 442 126 1303 1321 (100P, 87F) 425 123 1301 1319 QCNPS- (100P, 100F) 436 125 1326 1343
)
(100P, 87F) 422 122 1322 1338
- "ominal Rated 3-13 l ll> b l l A ' ' ' '
NED0-22192 Table 3-5 TRANSIENI SWC4ARY-TURBINE TRIP WITHOUT BYPASS Initial Initial . .
F Fy aCPR Power Flow $ Q/A al Analyste (2 NSR) (! NBR) (% initial) (2 initial) (peig) (peig) 7x7 8x8 8x8R F8x8R float R 104 100 353 114 1177 1220 0.22 0.30 0.29 -
4 R 100 94 154 114 1172 1215 0.22 0.29 0.29 -
A R 85 61 247 106 1158 1182 0.09 0.13 0.13 -
A R 104 100 183 102 1198 1225 - 0.12 0.12 -
8 R 100 94 173 102 1185 1211 - 0.12 0.12 -
8 R 85 61 15C 100 1160 1180 - 0.06 0.06 -
3 R 104 100 260 110 1171 1217 0.16 0.23 0.22 -
C R 100 94 254 113 1167 1209 0.15 0.21 0.21 -
C R W5 61 169 105 1155 .78 0.04 0.07 0.07 -
C R 104 100 249 109 1186 .228 0.13 0.18 0.18 --
0 R 100 91 249 108 1176 1214 0.12 0.17 0.17 -
0 D R 85 61 162 105 1152 1175 0.01 0.02 0.03 -
R R 104 100 333 115 1207 -- 0.20 0.28 - --
R R 100 92 325 113 118' - 0.18 0.26 - -
t R 85 , 61 181 101 1149 - 0.04 0.06 - -
0.25 0.29 Th .c 889 1110 1139 -
d 100 100 121 -
0.28 0.32 Tb .c 924 1109 1143 - -
9 100 111 121 ra.b.c 0 100 111 849 120 1109 1143 - - 0.26 0.30 F
$ 8 6 100 111 924 121 1108 1141 - -- 0.28 0.32 TD ' 8 100 100 628 116 1103 1129 - -
Fb .d.e @ 100 111 723 118 1113 1134 - - 0.22 0.26
@ 100 100 509 119 1290 1301 - 0.27 - 0.29 QCN?$-LC7 4 100 87 443 117 1273 1287 - 0.24 - 0.26 QCNFS-1C7 QCN75-1C7 3 91 75 314 115 1226 1243 - 0.20 - 0.22 QCNFS-2C6 3 100 100 -
6 100 87 358 116 1231 1304 - - -- -
QCMPS-2C6 9 91 75 258 114 1244 1262 - - -- -
QCNPS-2C6 QCN75-2C6 4 85 61 196 113 1227 1241 - - - -
"Feedwattr temperature reduction 1/2 by9ame failure
% position seras kaasuredscrastime
- Position scram with 200 asec delay R-REDY. J-0DYN 3-14
l l
NEDO-22192 Table 3-6 TRANSIENT
SUMMARY
--LOAD REJECTION WITHOUT BYPASS L
!sitta! !attial Peuer Flow 8
Q/A j si jy .cy, hg (! MBR) (t NBA) . (! initial) (2 initial) M (pets) 7x? E Rn8R PSm8R A R 104 100 376 115 1178 1225 0.23 0.31 0.30 --
A 8 100 94 360 115 1173 1216 0.22 0.29 0.29 -
A R 45 61 251 0.09 107 1157 1182 0.13 0.13 --
l 8 E 104 100 201 104 1203 1229 -
0.14 0.14 -
8- t 100 to 111 103 1189 1215 --
0.14 0.14 --
8 2 65 41 167 102 1162 1183 - 0.09 0.09 --
C R 104 100 302 til 1172 1219 0.18 0.25 0.25 -
C R 100 94 284 114 1168 1210 0.16 0.22 0.22 --
C t 85 41 168 106 1154 1177 0.04 0.07 0.07 -
} O R 104 100 ' 277 111 1189 1233 0.15 0.21 0.21 -
D R 100 91 267 113 1180 1219 0.14 0.19 0.19 -
D E 85 61 176 107 1153 1177 0.03 0.05 0.06 --
8 R 104 100 116 36 7 1:09 - 0.22 0.30 -- -
E R 100 92 344 114 1188 -
0.19 0.27 -- -
E R 83 6 '. 179 101 1149 - 0.04 0.07 -- --
F NOT Am4LY2ED G 8 104 100 507 114 1186 1208 -- -
0.17 0.17 G 8 b 100 94 489 114 1180 1202 - - - -
G 8 91 75 424 113 1175 1194 -- - - -
C 8 85 61 332 111 1165 1183 - -- - -
G e 104 105 501* 113 1184 1207 - -
0.17 0.18
/ C" 8 105 100 503 115 1178 1200 - - e.18 0.18 G* 8 105 105 481* 114 1184 1:06 - -
0.18 0.18 p 5 8 100 b
. 100 679 124 1206 1230 0.35 0.35 0.39 M 8 100 92 631 122 1206 1228 -
0.31 0.31 0.34 5 8 92 75 396* 120 '.183 1202 -
0.25 0.25 0.28 8 8 85 61 329 122 1195 1208 - 0.23 0.24 0.26 R 8 b 100 87 576 121 1205 1227 -
0.30 0.30 0.33 i E 8 104 100 502 117 1179 1213 0.14 0.19 0.19 0.19 K 8 100 94 469 117 1174 1:06 0.13 0.17 0.17 0.19 L 8 104 100 338* 108 1166 1189 -- - - --
L 8 100 94 320 108 1160 1182 - -- - -
L 8 91 75 267* 108 1145 1168 - - - -
L 8 85 61 216 106 1145 1160 - -- -- -
L 8 104 105 333 108 1167 1191 0.07 - 0.11 0.11 L* J 105 105 336 108 1165 0.08 115A -
0.11 0.11
. L' 8 105 100 346 109 1166 '.08 13 -
0.11 0.11
.- 't 8 104 100 453 120 1:08 12 -
0.22 0.22 0.24 M 8 100 94 596 120 1197 113. -
C.20 0.20 0.23 3 8 105 100 447 118 t 1189 1218 - - --
0.19 3 8 100 87 453 117 1183 1204 - -- -
0.18 QCNPS-1C7 8 100 100 558 1:1 1303 1315 -
0.29 -- 0.31 J QCNPS-LC7 8 100 37 446 118 1279 1293 --
3.25 -
0.27 i QCNPS-LC7 8 91 75 360 116 1:31 1250 -
0.22 - 0.24 QCNPs 2C6 8 100 100 497 119 1307 1322 -- 0.27 -- 0.29
-QCNPS-2C6 8 100 87 400 117 1307 1322 - 0.22 -
0.25 QCNPS-2C6 8 91 75 294 116 1:47 1267 -
0.19 - 0.21 QCNPS. C6 8 A; 61 216 114 1231 1246 -- 0.13 -- 0.18
- Feedwater temperature reduction 1 .i-1,-ee 3-15
NED0-22192 F Table 3-7 -
TRANSIElff
SUMMARY
-LOSS OF FEEDWATER HEATING '[-
W F
F F
laitial lattial . . j s1 *
'T g
g-Feuer riow 6 Q/A y
(! WSS) (1 taitial) (2 initial) g g 7mf g SuSE P8x8s hM (3 WSR) 0.13 0.13 -
=
u6 114 1018 1068 0.11 A E 1G4 100 116 He 1012 1057 0.11 0.13 0.13 - _-
A E 130 94 117 994 1020 0.13 0.15 0.13 -
A R -
SS 61 117 0.13 0.13 -
a 1 104 100 116 113 1004 1064 -
{
10$3 - 0.13 0.13 - "_
94 116 116 1002 8 R 100 988 1019 - 0.19 0.19 - b 83 61 121 121 5 t 116 1019 1068 0.11 0.13 0.13 -
c R 104 100 128
- ~
He 1013 1057 0.12 0.14 0.14 -
C a 100 94 116 0.15 0.13 - t 61 111 111 992 1017 0.13 C a 85 F 117 117 1012 1048 0.15 0.16 0.16 -
0 1 104 100 118 117 1004 1053 0.13 0.14 0.14 -
D 1 100 91 _
0 t 85 61 123 128 990 1022 0.18 0.19 0.20 - [
104 10ft 121 119 1023 - 0.14 0.16 - -
{
8 E 1016 - 0.14 0.17 - - E t a 100 92 121 120 w 61 123 124 999 - 0.19 0.21 - -
}
a R 85 112 ul 1002 1043 - - 0.14 0.14 _-
r a 100 100 116 113 1041 1083 - - 0.14 0.14 y a 100 til 0.14 0.14
{.
ra a t00 ut u0 uo 1023 1073 - - -
=
I
?
MST -u t
l 7_
h M
T i
3-16
3 C
NED0-22192 l
i y: Table 3-8
> TRANSIENT SI29fARY-FEEDWATER CONTROLLER FAILURE
)- Intstal Initial . .
Power - Flow d 0/4 .1 v^
p Plant Analvets it 481 (2 488) (T (pittal) it initial) (pots) fpstal ?n7 A8 t =8e Pessa
-A R 104 100 242 114 - 1152 1200 h.14 0.25 0.25 --
A R 100 94 ~ 241 115 1150 1193 0.19 0.26 0.2n --
h- ~ A. R 85 61 184 '111 113d IL60 0.13 0.17 0.17 --
8 A- ' 104 100 144 106 1153 1187 -- 0.09 0.09 --
8 R 100 94 - 134 107 1148 1179 - 0.09 0.10 -
)' 8 a 85 61 136 LO9 1137 1156 - 0.13 0.13 --
C R 104 100 109 105 1028 1076 0.05 0.06 0.06 -
C R 100 to 112 110 1022 1067 0.06 0.07 0.04 --
.C E 85 61 117 111 996 1021 0.09 0.10 0.11 --
D R 104 100 185 111 1147 1193 0.11 0.16 0.16 -
D E 100 91 174 -115 1142 1181 0.09 0.13 0.13 -
D R 65 61 176 116 1127 'L149 0.10 0.12 0.12 --
E R 104 100 211 til L146 - 0.14 0.21 - --
8" 1 100 92 188 106 1112 -. 0.11 0.17 -- -
t R' ' 85 61 133 109 1127 - 0.09 0.11 -- -.
F d 100 100 214 105 1031 1062 -- -- - --
F A 100 - 111 181 104 LO21 1062 - - -- --
8 F 9 100 Lil 180 105 1021 1063 - - - -
!' O d 104 100 293 b
114 1151 1182 - -- 0.15 0.16 C a 94 b 100 ~ 274 114 0.17
{~ b 1139 1172 - - 0.15 f C e 91 75 256 114 tt13 1160 - - 0.15 0.16
-c d 85 b 61 207 til 1128 1145 - - 0.13 0.14 O d 104 5 105 236 114 1143 !!77 -- -- 0.16 0.17 8 e b G 105 100 317 gg, gg33 gg77 _ _ 3,g7 g,gg a b G d 105 LO5 125 112 1111 1134 - -- 0.08 0.08 8 b G e 95 LOS 311 121 1129 1158 - - 0.21 0.23
- 5. J D 100 100 513 124 1169 12L . -- 0.33 0.34 0.37 J h m 100 92 488 123 1169 1197 -- 0.30 0.30 0.33 J 100 b N 87 465 122 1170 1194 - 0.29 0. 30 0.32 N J b 92 75 345 118 1166 1184 -- 0.23 0.24 0.26 rl N e 65 48 23Y 116 1147 1170 0. ! 7 0.19 0.21
!- K 4 104 '100 314' 114 1135 1172 0.09 0.13 0.14 0.L6
'E e 100 94 282 116 till 1195 0.11 0.L7 0.17 0.19 L J 104 100 L4: 108 1837 1161 0.08 - 0.11 0.11 D
L J 100 94 L91 110 1129 1158 0.07 - 0.11 0.!2 L- d 91 75 188 113 1124 0.07 0.1t L144 0.12 I' L d 85 61 146 109 1119 1138 0.0e -- -- --
L- d 104 105 198* 110 L136 116e 0.07 -- 0.11 0.12 a
L 4 105 105 '234I 115 1131 1161 0.11 - 0.15 0.t6 1.8 J LOS 100 126" 111 1137 1123 0.36 - 0.08 0.09 M e 104 100 362* 119 1173 1216 - 0.17 0.17 0.19 M J 100 94 332* 118 1170 1210 - 0.16 0.17 0.18 3 J LOS 100 264 115 1159 1188 - - - 0.16 3 J 100 87 266 115 1152 1172 -- - - 0.16
' QC3PS LC7 e 100 100 267 113 1136 1172 - 0.15 0.16
' QC3FS-LC7 J 100 97 260 115 1140 1171 -. 0.13 .- 0.14 j.. . QC3PS-1C7 . J 91 75 202 115 1111 1138 - 3.12' - 0.13
, QC5PS 2C6 '4 100 100 192 115 1150 ,1189 - 0.14 - 0.15 l, ' QC3F$-2C6,. 'J- 100 If ' 253 114 1152 1155 '-' O.14 0.15
- QC3PS-2C6 @ 91 75 193 114 1126 1151 - 0.13 - 0.14 lI OC3PS-2C6 J 15 61 166 115 1107 1127 - -- -
{ *Feedwater temperature reduttaa
- g% nominal rated 3-17 V
l.. '
1
m NED0-22192 E Table 3 9 E N
SUMMARY
-HIGH PRESSURE COOLANT INJECTION k
lattial lattial . .
h hw SCPR
- Flow $ Q/A s1 Power Plant Analysts (t MSR) (T M3t) (t initial) (t initial) Qs,tjl i (psig) 1 Si M P9x8R _
0.12 0.12 120 113 1C17 1068 0.10 -
A E 104 100 "
114 1012 1058 0.11 0.14 0.14 -
A 1 100 94 123 118 995 1021 0.10 0.12 0.12 -
A E 85 61 119 0.10 0.10 1C63 -
113 109 1007 8 1 104 100 til 1002 1053 - 0.09 0.09 -
S t 100 94 115 115 987 1018 - 0.13 0.13 - i-B E 85 61 117 113 1018 1068 0.11 0.14 0.14 - _
C E 104 100 122 117 1012 1057 0.12 0.15 0.15 -
{-
C E 100 94 123 _
til 993 1018 0.14 0.16 0.16 -
C E 85 61 120 100 113 111 1010 1065 0.10 0.12 .0.12 -
104 D L -
111 1003 1052 0.09 0.09 0.10 -
0 R 100 91 115 120 987 1019 0.12 0.13 0.13 -
D R 85 61 117
- - - 0.12 0.14 - -
g a 104 100 - _
- - - - 0.12 0.14 - -
a g a 100 92
- - - - 0.16 0.18 - -
g a 85 61
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NED0-22192
- 4. REFERENCES
' 1
- l. " Load Line Limit Analysis for Dresden Units 2 and 3 and Quad Cities Units 1 ~
and 2. General Electric Company, December 1978 (NEDO-24167).
L 2. " Final Safety Analysis Report, Quad Cities Nuclear Power Station."
- 3. " Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power f
Station Unit 1, Reload 6 (Cycle 7)," General Electric Company, May 1982 I
(Y1003J01A43).
- 4. " Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 2 Reload 5 (Cycle 6)," General Electric Company, June 1981 (Y1003J01A43).
b
- 5. R. L. Gridley (GE), letter to D. G. Eisenhut (NRC), " Review of Low-Core Flow Effects on LOCA Analysis for Operating BWRs," May 8, 1978.
f 6. D. G. Eisenhut (NRC), letter to R. L. Gridley, enclosing " Safety Evaluation l Report Revision of Previously imposed MAPLHGR (ECCS-LOCA) Restrictions for BWRs at Less Than Rated Flow," May 19, 1978.
- 7. "Losswf-Coolant Accident Analysis Report for Dresden Units 2, 3 and Quad f
Cities Units 1, 2 Nuclear Power Station," General Electric Company, April 1979 (NEDO-24146A, Revision 1).
i 4-1/4-2
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