ML19350A185
| ML19350A185 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 02/28/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19350A186 | List: |
| References | |
| 80NED261, NEDO-24243, NUDOCS 8102260551 | |
| Download: ML19350A185 (45) | |
Text
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NEDO-24243 i 80NED261 CLASSI FEBRUARY 1980 l
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GENERAL ELECTRIC BOILING WATER REACTOR LOAD LINE LIMIT ANALYSIS FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT b
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GENERAL! ELECTRIC 9
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l NEDO-24243 SONED261 Class I February 1980 i'- !
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- i GENERAL ZLECTRIC BOILING WATER REACTOR l
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- LOAD LINE LIMIT ANALYf,23 FOR l l l JAMES A. FITZPATRICK NUCLEAR POWER PLANT t
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i NUCLEAR ENERGY PRCJECTS OlvisiCN
- GENERAL ELECTRIC COMPANY SAN JCSE, CALIFC ANIA 95125 G E N E R A L () E LE CTR IC I
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NEDO-24243 g)
I IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY a
This report was prepared by General Electric solely for The Power Authority of the State' of New York (The Authority) for the Authority's use. The .
information contained in this report is believed by General Electric to be I an accurate and true representation of the facts known, obtained or provided i to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the contract between The Authority and !
General Electric Company for nuclear fuel and related services for the nuclear ;
system for The James A. FitzPatrick Nuclear Power Plant. dated June 12, 1970, i and nothing contained in this document shall be construed es changing said I contract. The use of this information except as defined by said contract, or !
for any purpose other than that for which it is intended, is not authorized; i and with respect to any such unauthorized 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 i
- he information contained in this document or that such use of such information j i may not infringe privately owned rights; nor do they assume any responsibility ;
for liability or damage of any kind which may result from such use of such information.
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y NEDO-24243 TABLE OF CONTENTS i I
Page
- 1. INTRODUCTION 1-1
- 2.
SUMMARY
2-1
- 3. DISCUSSION 3-1 3.1 Back'round g 3-1 3.2 Analytical Basis 3-1 3.3 Analysis Results 3-2 3.3.1 Transients 3-3 3.3.2 ASME Prresure Vessel Code Compliance 3-3 3.3.3 Rod Withdrawl Error 3-3 3.3.4 Thermal-Hydraulic Sta lity Antilysis 3-3 3.3.5 Loss-of-Coglant Accident 3-5
- 4. APPLICATION 4-1
- 5. REFERENCES 5-1 l
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NEDO-24243 LIST OF ILLUSTRATIONS Figure Title Page 2-1 JAFNPP Power / Flow Map 2-2 3-1 Operating Map 3-15 3-2 Scram Reactivity and Control Rod Drive Specifications 3-16 for James A. FitzPatrick Nuclear Power Plant at EOC3 (104% Power, 100% Flow) 3-3 Scram Reactivity and Control Rod Drive Specifications 3-17 for ' James A. FitzPatrick Nuclear Power Plant at EOC3 (100% Power, 94% Flow) 3-4 Scram Reactivity and Control Rod Drive Specification 3-18
, for Jtmes A. Fitzpatrick Nuclear Power Plant at EOC3 (85% Power, 61% Flow) 3-5 Limiting RWE Rod Pattern (100% Power, 100% Flow) 3-19 3-6 Limiting RWE Rod Pattern (100% Power, 94% Flow) 3-20 3-7 Limiting RWE Rod Pattern (91% Power, 75% Flow) 3-21 3-8 Limiting RWE Rod Pattern (85% Power, 61% Flow) 3-22 3-9 Generator Load Rejection Without Bypass, EOC3, 3-23 104% Power, 100% flow 3-10 Generator Load Rejection Without Bypass, EOC3, 3-24 100% Power, 94% Flow 3-11 Generator Load Rejection Without Bypass, EOC3, 3-25 85% Power, 61% Flow 3-12 . Turbine Trip Without Bypass, EOC3, 104% Power, 100% Flow 3-26 3-13 Turbine Trip Without Bypass, EOC3, 100% Power, 94% Flow 3-27 3-14 Turbine Trip Without Bypass, EOC3, 85% Power, 61% Flow 3-28 3-15 Feedwater Controller Failure, EOC3, 104% Power, 100% Flow 3-29 3-16 Feedwater Controller Failure, EOC3, 1007 rower 04% Flow 3-30 3-17 Feedwater Controller Failure, EOC3, 85% fower, 61% Flow 3-31 3-18 Loss of Feedwater Heating, Cycle 3, 104% Power, 100% Flow 3-32 3-19 Loss of Feedwater Heating, Cycle 3, 100% Power, 94% Flow 3-33 3-20 Loss of Feedwater Heating, Cycle 3, 85% Power, 61% Flow 3-34 3-21 Inadvertent Startup of HPCI Pump, Cycle 3,104% Power, 3-35 100% Flow 3-22 Inadvertent Startup of HPCI Pump, Cycle 3, 100% Power, 3-36 94% Flow V
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NEDO-24243 l LIST OF ILLUSTRATIONS (Continued)
Figure Title Page 3-23 Inadvertent Startup of EPCI Pump, Cycle 3, 85% Power, 3-37 61% Flow '
3-24 MSIV Closure, Flux Scram, EOC3, 10l,% Power, 100% Flow 3-38 3-25 MSIV Closure, Flux Scram, EOC3,100% Power, 94% Flow 3-39 3-26 MSIV Closure, Flux Scram, EOC3, 85% Power, 61% Flow 3-40 ,
3-27 Reactor Core Stability Decay Ratio 3-41 4-1 Core Flow-Circulation Flow Relationshir for Jet-Pump 4-3 Plants i
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vi w ame s immi i mi
e i NEDo-25243 1
. LIST OF TABLES l
Table Title Page 3-1 Transient Input Data and Operating Conditions for 3-6 j 104/100 Power / Flow I 3-2 Transient Input Data and Operating Conditions for 3-7
- 100/94 Power / Flow f 3-3 Transient Input Data and Operating Conditions for 3-8 l 85/61 Power / Flow l 3-4 GETAB Transient Analysis Initial Condition Parameters 3-9 2
g (104% Power /100% Flow) 3-5 GETAB Transient Analysi: '..itial Condition Parameters 3-10 (100% Power /94% Flowi 3-6 CETAB Transient Analysis Initial Condition Parameters 3-11 (85% Power, 61% flow) 3-7 Transient Summary 3-12 3-8 over-Pressurization Analysis Summary 3-13
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_ v NEDO-24243
- l. INTRODUCTION The flexibility of a typical B'4R in proceedire from lcw power-low core flow to high power-high core flow is affected by many factors; two of these are (1) the power / flow ratios defined by the power flow curve and (2) Precon-ditioning Interim Operating Management Recommendations (PCIOMRs).
The power / flow curve is the locus of power from a fixed rod pattern as a function of flow from which the occurrence of certain nonstandard events (abnormal operating transients) will yield results within defined plant safety limits; PCIOMRs reduce the rate of change of power in the fuel to improve the fuel performance. The combination of these two effects can result in the inability to attain full reactor power directly.
Recent analyses (References 1 and 2) justify the modification of the operating envelope defined by the power / flow curve while remaining within previously established operating 1Lnits and the PCIONRs.
This report provides the analytical basis for JAFNPP operation under a modified power / flow line designed to permit the direct ascension to full power I within the design bases previously applied.
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- 2. SU20ttRY A method has been derived whereby the power / flow curve can be modified to i provide relief from the difficulties associated with the effects of the l existing power / flow curve and the PCIOMRs.
Analyses show that reactor ascension to full power may proceed along a modi-fled power / flow line bounded by the 108k rod block line up to the poir.t labeled 85/61, from which continued power increases may proceed along the rod block intercept line to the 100% power /94% flow point as shown in Figure 2-1.
The discussion and analyses in the following sections of this report show 4 ., that all safety bases normally applied to JAFNPP operation within this j envelope are satisfied.
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NEDO-24263 150 140 -
130 -
100% POWER LINE 120 -
' A INTERCEPT MINT (100,94)
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APRM ROD BLOCK LINE (108/100)
(PROPOSED) (0.58W + 50%) (1M,100)
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ROD INTERCEPT LINE ROD BLOCK INTERCEPT POINT (85/61) 90 -
TYPtCAL POWER l ASCENSION PATH ;
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5 E ANALYSIS NEEDED TO 2 70 -
OPERATE IN THIS REGION ,
1 TYPICAL 100% POWER /
60 -
100% FLOW LOAD LINE 1
50 -
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NATURAL CIRCULATION f ;
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g 1/ I I I I I I I I O 10 20 30 40 50 SG 70 80 90 100 110 CORE FLOW (%)
I Figure 2-1. JAFNPP Power / Flow Map 2-2
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NED0- 24243
- 3. DISCUSSION
3.1 BACKGROUND
i
! Operation of the JAFNPP utilizing the power / flow map is described in Section 3.7 of the FSAR (Reference 4). This section of the FSAR describes the basic operating envelope (Figure 3.7-1) within which normal reactor operations are conducted and provides the basic philosophy behind the power /
- j
flow curve. FSAR Figure 3.7-1 is reproduced as Figure 3-1 of this document.
3.2 ANALYTICAL BASIS I
To provide relief from the operating restrictions inherently imposed during ascension to power by the existing power / flow curve and PCIOMRs, a modified power / flow curve has been derived. In deriving this operating curve, five design basis objectives were specified:
- 1. For those transients and accidents that are sensitive to variations in power and flow, the 104% power /100% flow 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 2-1).
- 2. In no instance shall the ratio of power to flow intentionally exceed the ratio defined by the rod block line.
- 3. The slope of the rod block intercept line between the rod block l intercept point and the 100% intercept point must ce such that flow increases are capable of compensating for xenon buildup while increasing reactor power.
- 4. '1he consequences of all accidents and transients analyze n the FSAR and subsequent amendments and the license submittals .2st remain within the limits normally specified for such events.
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g NE00-24243
- 5. Reactor power ascension from natural circulation to full power shall be directly attainable through combined control rod movement and recirculation flow increase without violation of either the power /
flow line or PCIOMRs.
To meet these objectives, analyses were perfor=ed as shown below:
- 1. The scram reactivity insertion characteristics were determined for end-of-cycle conditions, at 104% power /100% flow, 85% power /61% flow, and 100% power /94% flow, and are shown in Figures 3-2, 3-3 and 3-4.
- 2. Abnormal operational transients were analyzed at several points along the proposed power flow curve to verify that the 104% power /100%
flow point defines the limiting case.
- 3. Core thermal-hydraulic characteristics were determined for operation within the proposed envelope.
- 4. Rod withdrawal error (RWE) was re-analyzed.
- 5. The effects of the changed operating range were analyzed to determine any effects on the previous ECCS analyses.
3.3 ANALYSIS RESULTS The most significant value in the analyses discussed here is the rod block ,
intercept point; this value establishes the highest power level permitted when operating on the 108% APRM rod block line. This point is sufficiently high to permit a greater power / flow ratio for power ascension but not so high as to introduce accident or transient results more severe than at the 104/100 power /
flow point. The rod block intercept point selected is 85% power /61% flow.
This value is low enough to ensure that adequate margin to the higher limits i exist to permit generic application without re-analysis for each plant cycle.
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M' NEDO-24243 P
i Similarly, the value is suf ficiently high to provide the desired operational flexibility during power ascension.
- i. 3.3.1 Transients h
hg As shown in Reference 3, the most limiting abnormal operational transients 9 for JAFNPP, Cycle 3, are load rejection-with-failure-of-the-bypass-valves g (LRW/0BP), loss of feedwater heating (LFWH), feedwater controller failure
) (FWCF), and rod withdrawal error (RWE). These transients were analyzed under the conditions listed in Tables 3-1 through 3-6. Each transient was analyzed
- s. a power / flow ratio of 104/100,100/94 and 85/61 to provide verification of transient behavior along the rod intercept line and the 100% power line to the 104/100 point. The results are shown in Table 3-7.
3.3.2 ASME Pressure Vessel Code Compliance All main steamline isolation valve (MSIV) closure with flux scram event is used to determine compliance to the ASME pressure vessel code. The ASME Boiler and Pressure Vessel code limit is 1375 psig based on the MSIV closure transient with an assumed failure of direct scram. As can be seen in Table 3-8, the peak vessel bottom pressures at both the 85/61 point and the 100/94 point are below that for the 104/100 point, and all three are below the code limit.
l 3.3.3 Rod Withdrawal Error The effective RBM setpoint is a function of power and flow. Above the rated rod liner the rod block will occur with less rod withdrawal. Thus the evaluation at rated is conservative for operation above the rated load line.
6 3.3.4 Thermal-Hydraulic Stability Analysis Thermal-hydraulic stability analyses for JAFNPP Cycle 3 were presented in Reference 3. These were re-analyzed employing the new operating envelope.
The results are given in the following paragraphs.
3-3
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NED0-24243
- 3. 3.4.1 Channel Hydrodynamic Confor=ance to the Ulti= ate Performance Criterion The channel performance calculation for JAFNPP Cycle 3 yields decay ratios as presented below:
Extrapolated Rod Block Line -
Channel Hydrodynamic Performance Natural Circulation Power Decay Ratio. X 2
/XO 8x8R Channel 0.34 8x8 Channel 0.39 7x7 Channel 0.23 At this most responsive condition, the most responsive channels are clearly within the bounds of the ultimate performance criteria of <l.0 decay ratio at all attainable operating conditions.
3.3.4.2 Reactor Conformance to Ultimate Performance Criterion The decay ratios determined from the limiting reactor core stability con-ditions are presented in Figure 3-27. The most responsive case is the extrapolated rod block line - natural circulation condition.
Extrapolated Rod Block Line -
Reactor Core Stability Natural Circulation Power Decay Ratio, X2/Xo 0.76 These calculations show the reactor to be in compliance with the ultimate performance cciteria, including the most responsive condition.
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I 3.3.5 Loss-of-Coolant Accident A discussion of low-flow effects on LOCA analyses for all operating plants (Reference 6) has been presented to and was approved by the NRC (Reference 7).
The LOCA analysis for JAFNPP (Reference 8) is applicable in the power flow domain discussed in this JAFNPP submittal.
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I EEDO-24243 ,
Table 3-1 TRANSIENT INPUT DATA AND OPERATING CONDITIONS FOR 104/100 POWER / FLOW l
Thermal Power (MWt) 2533 104%
Steam Flow (1b/hr) 10.96 x 106 105%
Rated Core Flow (lb/hr) 77.0 x 106 100:
Dome Pressure psig 1019 ;
Turbine Pressure psig 958 S/RV Lowest Setpoint psig 1090 S/RV/ Capacity (at Setpoint) No./%NBR 11 - 85.7 S/RV Time Delay (msec) 400 S/RV Stroke Time (msec) 100 Void Coefficient c/%R NDP -8.76 -
TAP -10.93
=
Doppler Coefficient C/*F NDP -0.1985 TAP -0.1886 Average Fuel Temperature *F 1519 Scram Worth $
NDP -39.09 TAP -31.27 CRD Spec 67B Scram Reactivity Figure 3-2 4
i NDP - Nuclear Dynamic Parameter TAP - Transient Analysis Parameter l
3-6
NEDO-24243 Table 3-2 TRANSIENT INPUT DATA AND OPERATING CONDITIONS FOR 100/94 POWER / FLOW Thermal Power (MWt) 2436 100:
Steam Flow (1b/hr) 10.47 x 106 100%
Rated Core Flow (ib/hr) 72.4 x 106 94:
Dome Pressure psig 1012 Turbine Pressure psig 957 Void Coeffic'ient c/%R g~
NDP -9.16 TAP -11.45 i
Doppler Coefficient c/*F ,
NDP -0.2278 TAP -0.2164 6 Average Fuel Temperature *F 1472 Scram Worth $
, NDP -39.39 TAP *
-31.51 CRD Spec 67B Scram Reactivity Figure 3-3 NDP - Nuclear Dynamic Parameter IAP. Transient Analysis Parameter 5
3-7 I
NED0-24243 l
Table 3-3 TRANSIENT INPUT DATA AND OPERATING CONDITIONS FOR 85/61 POWER /FLO'4 l
[
Thermal Power (MWt) 2071 85%
Steam Flow (1b/hr) 8.72 x 106 83.3%
Rated Core Flow (1b/hr) 46.97 x 106 61%
Dome Pressure psig 992 Turbine Pressure psig 954 Void Coefficient c/%R g NDP -10.27 TAP -12.84 l
Doppler Coefficient c/*F NDP -0.2275 i TAP ' -0.2161 f
Average Fuel Temperature *F 1303 Scram Worth S NDP -39.64 i TAP -31.71 CRD Spec 67B l
Scram Reactivity Figure 3-4 ;:
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NDP - Nuclear Dynamic Parameter TAP - Transient Analysis Parameter 3-8 l
I NEDO-24243 .
p !
Table 3-4 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (104% Power /100% Flow)
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-l Core Power, MWT 2436.0 Core Flow, M1b/hr 77.G I
t Reactor Pressure, psia 1035.0 Inlet Enthalpy, Btu /lb 526.9 f
Non-Fuel Power Fraction . 04
{
I Axial Peaking Factor 1.40 r
- , 7x7 8x8 8x8R e
[ Peaking Factors (local, radial) (1.24, 1.17) (1.22, 1.24) (1.22, 1.39) p R-Factor 1.100 1.098 1.051 8
Bundle Power, MWt 5.012 5.306 5.966 l Bundle Flow, 103 lb/hr f 127.2 118.3 118.2 l Initial MCPR 1.32 1.40 1.37 I
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Table 3-5 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (100 Power /94% Flow) t Core Power, MWt 2436.0 Core Flow, M1b/hr 72.4 Reactor Pressure, psia 1033.9 Inlet Enthalpy, Beu/lb 525.5 No'n-Fuel Power Fraction 0.04 Axial Peaking Factor 1.40 5
~
7x7 8x8 8x8R i Peaking Factors (local, radial), (1.24, 1.19) (1.22, 1.27) (1.20, 1.40)
R-Factor 1.100 1.098 1.051 '
Bundle Power, NWt 5.058 5.402 5.942 Bundle Flow, 10 3 lb/hr 119.2 110.1 110.1 '
Initial MCPR 1.29 1.36 1.36 i
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NEDO-24243 Table 3-6 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAV.ITERS (85* Power, 61% Flow)
Core Power, MWt 2070.6 Core Flow, M15/hr 47'.0 Reactor Pressure, psia 1007.6 Inlet Enchalpy, Btu /lb 513.2 i Non-Fuel Power Fraction 0.04 Axia'l Peaking Factor 1.40
- 7x7 8x8 8x8R Peaking Factors (local, radial) (1.23, 1.32) (1.22, 1.44) (1.20, 1.57)
R-Factor 1.100 1.098 1.051 Bundle Power, MWt 4.761 5.205 5.684 Bundle Flow, 103 lb/hr 75.3 68.2 68.2 Initial MCPR 1.21 1.24 1.24 q-3-11
Table 3-7 TRANSIENT
SUMMARY
Initial Initial $ Q/A p p Power Flow (%) (%) SL v ACPR Figure Event (%NBR) (%NBR) Initial Initial (psig) (psig) 7x7 8x8 8x8R Numlier LRNBP 104 100 376 115 1178 1225 0.23 0.31 0.30 3-9 LRNBP 100 94 360 115 1173 1216 0.22 0.29 0.29 3-10 LRNhP 85 61 251 107 1157 1182 0.09 0.13 0.13 3-11 TTNBP 104 100 353 114 1177 1220 0.22 0.30 0.29 3-12 TTNBP 100 94 345 114 1172 1215 0.22 0.29 0.29 3-13 TTNBP 85 61 247 106 1158 1182 0.09 0.13 0.13 3-14 E!
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U FWCF 104 100 243 114 1152 1200 0.18 0.25 0.25 3-15 [
FWCF 100 94 241 115 1150 1193 0.19 0.26 0.26 3-16 O FWCF 85 61 184 111 1138 1160 0.13 0.17 0.17 3-17 LFWit 104 100 116 114 1018 1068 0.11 0.13 0.13 3-18 LFWil 100 94 116 114 1012 1057 0.11 0.13 0.13 3-19 LFWil 85 61 117 117 994 1020 0.13 0.15 0.15 3-20
, llPCI 104 100 120 111 1017 1068 0.10 0.11 0.12 3-21 IIPC1 100 94 123 114 1012 1058 0.11 0.14 0.14 3-22 IIPCI 85 61 119 118 995 1021 0.10 0.13 0.09 3-23
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NEDO-24243 Table 3-8 OVER-PRESSURIZATION ANALYSIS SLW.ARY Power Flow Py Figure Transient (%) (%) (psig) Number MSIV Closure 104 100 1264 3-24 (Flux Scram)
MSIV Closure 100 94 1254 3-25 (Flux Scram)
MSIV Closure 85 61 1211 3-26 (Flux Scram)
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EEDO-24243 too 45 CONTROL RCD DRIVE VERSUS TIME e SCRAM REACTivlTY VERSUS TIME A - 1
- do 1
C - 678 CRD IN PERCENT so -
1 - NOMINAL SCRAM CURVE IN (-S) _
2 - SCRAM CURVE USED IN ANALYSIS l 70 - 2
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TIME (sec) i i
t Figure 3-2. Scram Reactivity and Control Rod Drive Specifications for !
James A. FitzPatrick Nuclear Pcwer Plant at EOC3 (104% !
Power, 100% Flow) I I
l 3-16 l
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NECO-24243
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100 45 g
CONTROL ROD ORIVE VERSUS TIME C SCRAM REACTIVITY VERSUS TIME 90 -
- 40 80 - C - 678 CRO IN PERCENT
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1 - NOMINAL SCRAM CURVE IN (-S) 2 - SCRAM CURVE USED IN ANALYSIS 2
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- 0 1 2 3 4 TIME (sec)
! Figure 3-3. Scra:n Reactivity and Control Rod Drive Specifications
}
for James A. FitzPatrick Nuclear Power Plant at EOC3 6 (100% Power, 94% Flow) 3-17 l
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- u NEDO-24243 C 45 CONTROL ROD DRIVE VERSUS TIME SCR AM REACTIVITY VERSUS TIME 90 - ,
42 l, C - 67B'CRD IN PERCENT 1 - NOMINAL, SCRAM CURVE IN 1-S) ;
~
2 - SCRAM CURVE USED IN ANALYSIS
- 35 1 2
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- 30 l
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- 15 30 -
4 to 20 -
5 to -
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O ' f2 l l 0 0 0 1 2 3 4 TIME (sec)
Figure 3-4. Scram Reactivity a d Control Rod Drive Specification for James A. Fitzpatrick Nuclear Power Plant at EOC3 1 (85% Power, 61% Flow) )
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02 06 10 14 18 22 26 i' 51 26 26 I
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- I 47 42 26 t
I-l 43 26 36 36 07 42 10 4 f 35 26 36 36 36
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27 26 36 36 36 t
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Notes: 1.- Rod patterns are one-quarter core mirror symmetric,
- upper left quadrant shown.
I 2. Numbers indicate notches withdrawn (48 is full out).
I Blank indicates a fully withdrawn rod.
1
[ 3. Error rod is rod (22,31).
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A Figure 3-5. - Limiting RWE Rod Pattern (100% Power,100% Flow) f.
3-19 b-
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NEDO-24243 02 06 10 14 18 22 26 51 26 26 ,
. l 47 42 26 43 26 36 36 ,
f 39 42 10 4 ;
l-35 26 36 36 36 1
I 31 26 4 0 {
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27 26- 36 36 36 ;
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Notes: 1. Rod patterns are one-quarter core mirror sy:mnetric, I upper left quadrant shown.
- 2. Numbers indicate notches withdrawn (48 is full out).
Blank indicates a fully withdrawn rod.
- 3. Error rod is rod (22,31)
Figure 3-6.. Lisiting R*4E Rod Pattern (100% Power, 94% Flow) 3-20 N
<- m NEDO-24263 02 06 10 14 18 22 26 51 16 16 47 40 16 43 16 36 36 i
6
! 39 40 26 8
! 35 16 36 36 36 6 31 16 8 0 12 7 16 36 36 36 Notes: 1. Rod patterns are one-quarter core mirror symmetric, upper left quadrant shown.
- 2. Numbers indicate notches withdrawn (48 is full out).
Blank indicates a fully withdrawn rod.
- 3. Error rod is rod (22,31).
Figure 3-7. Limiting RWE Rod Pattern (91% Power, 75*. F*.ow)
. 3-21
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. I i 27 22 34 36 36 Notes: -1. Rod patterns are one-quarter core mirror symmetric, upper left quadrant shown.
- 2. Numbers indicate notches withdrawn (48 is full out).
Blank indicctes a fully withdrawn rod.
a 3. Error rod is rod (22,31)
Figure 3-8. Limiting RWE Rod Pattern (85% Power, 61% Flow) 3-22
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IINE (SICI Figure 3-20. Loss of Feedwater llenting, Cycle 3, 85% Power, 61% Flow
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- 1.2 I
i f ULTIMATE PERFORMANCE CRITEPIA 3
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- 4. APPLICATION The analyses described in this report in support of operation along the modified power / flow line are of a bcunding type that can be applied to all SWR /3 and EWR/4 plants whose operation is guided by a power / flow curve.
The rod block intercept point of 85% power /61% flow lies along the APRM flow-biaset rod block line having a slope represented by the equation:
0.58W + 50%
{ where W = recirculation flow rate in percent of rated The relationship between core flow and recirculation flow is shown in i Figure 4-1.
Currently, most BWRs operate on the basis of a power / flow curve approximated i
by the equation:
[
0.65W + 35**
I l
with the APRd flow-biased rod block represented by the equation, l
4 0.66W + 42%*
The Peach Bottom 2/3 proposed Technical Specification ( requested that
- the NRC permit adoption of the less restrictive equation (0.58W + 50%);
final disposition of these requests is pending NRC resolution. In i anticipation of'NRC approval of this proposed rod block line, the analyse,s f for this report were performed with this line as the upper-bound of the l
i l'*
t l
Several plants vary a few percent from these values.
l l 1 I!
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i1 w . . . --- .
NEDO-24243 proposed operating envelope. Further, the analyses justify the rod block intercept line (i.e., the line between 85/61 and 100/94 points).
Operation utilizing the current JAFNPP technical specification rod block line (0.66W + 42%) can be effected in the same manner as using the proposed rod block line, except the intersection with the rod block intercept line would occur at slightly higher power and flow (Figure 2-2). This is within the analyzed envelope and, therefore, conforms with the bases and conclusion of this report..
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l l l l l I l I l 0
0 to 20 30 40 50 60 70 30 90 100 RECIRCULATION FLCW (%WR I i
Figure 4-1. Core Flow-Recirculation Flow Relationship for Jet-Pu=p Plants 4-? '4-4
- 5. REFERENCES
- 1. Millstone Unit 1, Load Line Limit Analysis, Revision 1, Nove=ber 1977 (NEDO-21285-1).
- 2. Safety Evaluation by tha Office of Nuclear Reactor Regulation Supporting i Amendment No. 52 to Provisional Operating License No. DPR-21, Northeast Nuclear Energy Co., Millstone Nuclear Power Station Unit 1, Docket No. 50-245, July 1978.
- 3. Supplemental Reload Licensing Subnittal for the James A. FitzPatrick Nuclear Power Plant for Reload No. 2, June 1978 (NEDO-24129).
- 4. Final Safety Analysis Report, James A. FitzPatrick Nuclear Power Plant, Dockett 50-333.
- 5. Letter, Bradley to Giambusso, " Scram and Rod Block Setpoints at Low-Core j Flow," Philadelphia Electric Co., Peach Bottom Atomic Power Station, f Units 2 and 3, Dockets 50-277/278, DPR-44/56, June 17,1975.
- 6. Letter, R. L. Gridley to D. G. Eisenhut (NRC), " Review of Low-Core Flow Effects on LOCA Analysis for Operating BWRs," May 8, 1978.
- 7. Letter, D. C. Eisenhut (NRC) to R. L. Cridley, enclosing " Safety Evaluation Repcrt Revision of Previously imposed MAPLEGR (ECCS-LOCA)
Restrictions for BWRs at Less Than Rated Flow," May 19, 1978.
- 8. " Loss-of-Coolant Accident Analysis Report for James A. FiczPatrick Nuclear Power Plant (Lead Plant)," July 1977 (NEDO-21662).
' 5-1/5-2 1
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J :. : _ _ , ,_ _ ;
i NUCLEAR ENERGY DIVISIONS e GENERAL ELECTRIC CCMPANY SAN JOSE, CALIFORNIA 9513 GEN ER AL h ELECTRIC i TECHNICAL INFORMATION EXCHANGE TITLE PAGE AUTHOR SUBJECT Tiggg 730 oATE January 1980 TITLE GENERAL ELECTRIC BOILING '4ATER GE CLASS I
REACTOR LOAD LINE LIMIT ANALYSIS FOR JAMES A. FITZPATRICK NUCLEAR PO'4ER GOVERNMENT CLASS
' PLANT REPRODUCIBLE COPY FILEO AT TECHNICAL NUMBER OF PAGES SUPPORT SERVICES, R&UO, SAN JOSE,
, 52
- CALIFORNI A 96125 (Mai: Code 211)
SUMMARY
I *
? This report provides the analytical basis for JAFNPP l operation under a modified power / flow line designed to permit the direct ascension to full power within the l
design bases previously applied.
f 4
1
[ BY cutting out this rectangle and folding in half, the above information can be fitted j into a standard card file.
DOCUMENT NUMBER NE W 24243 INFORMATION PREP ARE D FOR Nuclear Energy Division SECTION Nuclear Energy Projects Division BUILDING AND ROOM NUMBER K-2607 uxitcoog 682 NE0414 (6m)
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