ML20010H921
| ML20010H921 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/22/1981 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20010H915 | List: |
| References | |
| NUDOCS 8109290414 | |
| Download: ML20010H921 (27) | |
Text
.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFQUIREMENTS C.
Minimum Critical Power Ratio (MCPR)
C.
Minimum Critical Power Ratio (MCPR) 1.
During power operation MCPR shall be 1.
MCPR shall be determined. daily during h the MCPR operating limit specified reactor power operation at > 25% rated in 3.11.C.2.
If at any time during thermal power and following any change operation it is determined by normal in power icvel or distribution that surveillance that the limiting value would cause operation with a limiting for MCPR is being exceeded, action control rod pattern as described in shall be initiated within 15 minutes the bases for SpecI. cation 3.3.B.S.
to restore operation to within the prescribed limits.
If the steady 2.
The value of Tin Specification state MCPR is not returned to with-3.11.C.2, shall be equal to 1.0 in the prescribed limits within two unless determined from the result (2) hours, the reactor shall be of surveillance testing of Specifi-brought to the Cold Shutdown cation 4.3.C as follows:
condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveil-lance and corresponding action shall a) 9'is defined as continue until reactor operation is y'B within the prescribed limits.
T=
ve 1.275
- B For core flows other than rated the MCPR limits shall be the limits identified is as shown in b) The average scram time to the above times Kg where Kf Figure 3.11-8 30% insertion position is _ deter-mined as follows:
As an alternative method providing n
equivalent thermal-hydraulic protection 7 nit i
at core flows other than rated, the cal-T ave =i=1 culated MCPR may be divided by Kf, where n
Kf is as shown in Figure 3.11-8.
Ni i=1 2.
The operating limit MCPR values as a function of Tare given in Table 3.11-1 where: n = number of surveillance whereT is given by specification tests performed to date in the 4.I1.C.2.
cy cle.
205B 8109290434 810922 '
DR ADOCK 05000293 PDR.
LI911 TING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS Ng = number of active control rods measured in the ith surveillance test.
971 = average scram time to the 30%
insertion position of all rods measured in the ith surveillance test.
c.) The cd' usted analysis mean scram time (C'B) is calculated as follows:
B =A + 1.6 5 [
T i
p.
n II Ng i=1 #
Where:
,4L = meen of the distribution for average scram insertion time to the 30% position 0.945 sec Ng = total number of active control rod measured in specification 4.3.C (f'= standard deviation of the distribution for average scram insertion time to the 30% position, 0.064 sec.
4 205B-1
TABLE 3.I1-1 OPERATING LIMIT MCPR VALUES MCPR Operating Limit
[
8x8 P8x8R f_0 1.32 1.35 0 to.1 1.32 1.36
.1 to.2 1.33 1.36
.2 to.3 1.33 1.36'
.3 to.4 1.34 1.37
.4 to.5 1.34 1.37
.5 to.6 1.35 1.38
.6 to.7 1.35 1.38
.7 to.8 1.36
- 1. 39
.8 to.9 1.36 1.39
.9 to 1.0 1.37 1.40 W
BASES 3.11A Averate Flanar Linear Heat Ceneration Fate _(AP2GR)
This specifications assures that the peak cladding te=perature following the postulated design basis loss-of-coolant accident vill not exceed the limit specified in the 10 CFR So, Appendix r.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the averag'e heat generation rate of all the rods of a fuel. assembly at any axial location and is only dependent, secondarily on the rod to rod power distribution within an asse=bly. The peak clad te=perature is calculated assu=ing a 111GR for the highest powered rod which is equal to or less than the design IEGR.
This LEGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The limiting value for APLEGR is this LEGR of tLe highest powered rod divided by its local peaking factor.
The calculational procedure used to establish the AP2GR limit for each fuel type is based on a less-of-coolant accident analysis.
The e=ergency core cooling system (ECCS) evaluation models which are e= ployed to determine the ef fects of the loss of coolant accident (LOCA) in accordance with 10CFR50 and Appendix K are discussed in Reference 1.
The mode.ls are identified as LAE, SCAT, SAFE, REFLOOD, and CHASTE. 'The LAS Code calculates the short ter= blevdown response and core flow, which are input fato the SCAT code to calculate blevdown heat transfer coefficients.
The SAFE code is used to deter =ise longer term syste= response and flows from the various ECC systems. Where appropriate, the output of SAFE is used in the RETLOOD code to calculate liquid The results of these codes are used in the CHASTE code levels.
to calculate fuel clad te=peratures and maximu= average placar linear heat generation rates O'.APLHGR) for nach fuel type.
The significant plant input parameters and the MAPLHGR's for fuel types calculated by the above procedure are the present included in Reference 2.
The curves in Figures 3.11-1 through 3.11-6 were developed assuming no core spray heat transfer credit in the LOCA analysis.
205C 8 *.ew Ag*
.--,------n-~---
-,e-
- ~ -,
m RETGENCES 1.
General Electric BWR Ceneric Reload Tuel Application, NEDE-24011-P.
2.
Ioss of Coolant Accident Analysis Report for Pilgrici Nuclear Power Sta tion, NEDO-21696, August 1977 as amended.
Amendnent No.
_B ASES:
3.11C NINDClM CRIMCAL TO1.T.R RAMO (MCFR)
Operating Limit MCP_R For any abnormal operating transient analysis evaluation with the initimi condition of the reacter being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any tine during the transient assuming instru=ent trip setting given in Specification ne difference between the specifie'd' Operating Limit MCPR in Specification 3.11C and the Safety Li=1t MCPR in Specification 1.1A defines the largest reduction in critical power ratio (CFR) permitted during any anticipated abnormal operating transient.
To ensure that this reduction is not exceeded, the most limiting transients are analized for each reload and fuel type (8x8 and P!x8R) to deter =ine that transient which yields the largest value of A CPR. This value, when added to the Safety Limit MCPR must be less than the minimum operating id=it MCPR's of Specification 3.11.C.
The resalt of this evaluation is docu=ented in the
" Supple = ental Reload Licensing Subtittal" for the current reload.
The evaluation of a given transient begins with the system gut paraneters shown in Tables 5-4. 5-6 and 5-8 of NEDE-24011-P Supple =ented by reload unique inputs given in the current Supple = ental Reload Licensing Sub=ittal. These values are input to a GE core dp)a=ic behavior transient co=puter program described in NCO-10802('. The transient code used for all pressurization event's is described in NEDE-24154-P (Reference 5). The MCPR analysis for pressurization events is done in accordance with the procedures given in Reference 6.
Amendment No.
205 C-3
r-REFERENCES _
1.
General Electric Bk'R Ceneric Reload fuel Application, NEDE-24011-P.
2.
Rb 3. Linford, Analytical Methods of Plant Transient Evaluations for the CE BVR, February 1973 (NEDO-10802).
3.
General Electric Company Analytical Model for Ioss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix E NEDE-20566 (Drafc), August 1974.
4.
Letter from J. E. Howard, Boston Edison Company to D. L. Zi-USNRC, dated October 31, 1975.
5.
Qualification of the One-Dinensional Core Transient Model for Boiling Water Reactors, October 19 78 (NEDE-24154-P).
6.
Letter, R. P. Denise (NRC) to G. G. Sherwood (GE), January 23, 1980:
taendment No.
205D e
O O g
6 e
e e
u FIGURE 3.11-6 MAXIMUM AVERAGE PLAHAR UNEAR HEAT GEMERATiol! FATE VERBUS FLANAR AVERAGE EXPOSURE FUEL TYPE P8DRB 2G5 H O
T-Uf
< l-Z<
< O' h
gi-1I r
4 10.9 ic.e 4of $
13 3 I c.'?
w z 2w C
10 4 0
10 4 1
~~
C 10.7.
^
O <H 9.9 9.9
^-
5w 9.8 9.8 E
- 6 Q
s.5
<g 94 93 54 o
o y
3 8.8 8
0 5,000 10,000 15,000 20,000 25,000 30,000 PLAMAR AVERAGE EXPOSURE (mwd /1) 205 E-6 ye
e o
4 i
NUCLE AR ENERGY BUSINESS GROUP
""- 2 '
PUBLICATION NO ER uTA And ADDENDA u E. NO.
SHEET LOSS-OF-COOLANT ACCIDENT TITLE 7
ANALYSIS REPORT FOR DILCRIM A"U"*C I"I OATE NUCLEAR POWER STATION NO TE: Correct all copies of the applicable AUGUST 1977 publication as toecified below.
ISSUE DATE REFERENCES INSTRUCTIONS T ON. A pj$E ITEM (CORRECTIONS AND ADDITIONS) 9 e, p
N E) 1 Pages 111 and iv Replace with new pages 111 and iv 2
Page 3-1 Replace with new page 3-1 3
Page 4-3 Replace with new page 4-3 4
Pages 4-7 to 4-9 Replace with new pages 4-7 to 4-9 5
Pages 4-10 to 4-12 Add new pages 4-10 to 4-12 6
Appendix A Add new Appendix A (pages A-1 to A-7)
Changes are indicated by vertical bar in the right-hand tiargin.
i i
l i
I f
I l
1 PAGE L
NED0-21696 TABLE OF CONTENTS Page 1.
INTRODUCTION 1-1 2.
LEAD PLANT SELECTION 2-1 3.
INPLT TO ANALYSIS 3-1 4
LOCA ASALYSIS COMPLTER CODES 4-1 4.1 Results of the LAMB Analysis 4-1 4.2 Results of the SCAT Anaylsis 4-1 4.3 Results of the SAFE Analysis 4-1 4.4 Results of REFLOOD Analysis 4-2 4.5 Results of the CHASTE Analysis 4-3 4.6 Methods 4-4 5.
DESCRIPTION OF MODEL AND INPUT CHANGES 5-1 6.
CONCLUSIONS 6-1 7.
REFERENCES 7-1 APPENDIX A - Lass-of-Coolant Accident Analysis with A-1 No Core Spray Heat Transfer Credit 111
~
NED0-21696 LIST OF TABLES Table Title Page 1
Significant Input Parameters to the Loss-of-Coolant Accident 3-1 2
Summary of Break Spectrum Results 4-5 3
LOCA Analysis Figure Summary -
Non-Lead Plant 4-6 4A MAPLHGR Versus Average Planar Exposure 4-7 43 MAPLHCR Versus Average Planar Exposure 4-8 4C MAPLHGR Versus Average Planar Exposure 4-9 4D MAPLFGR Versus Average Planar Exposure 4-10 4E MAPLHCR Versus Average Planar Exposure 4-11j i
4F MAPLHGR Versus Average Plana-Exposure 4-12 iv
3.
INPUI TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis -is -
presented-in Table 1.
Table 1 1
SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-C0CLANT ACCIDENT ANALYSIS i
Plant Parameters:
Core Thermal Power 2037 MWt, which corresponds to 102% of rated core power 6
Vessel Steam Output 8.14 x 10 lbm/h, which corre-cronds to 102% of ra.ed core power Vessel Steam Dome Pressure 1050 psia i
Recirculation Line Break 4.34 ft2 (D3A)
Area for Large Breaks - Suction Number of Drilled Bundles 428-t Fuel Parameters :
Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking
. Power Fuel Type Geometry (kW/ft)
Factor Ratio
- A.
8DB219L' 8x8 13.4 1.5 1.24 B.
SDB219H 8x8 13.4
'.5 1.24 C.
SDB262 8x8 13.4 1.5 1.24 s
D.
P8DRB265L 8x8 13.4 1.5 1.24 E.
P8DRB282 8x8 13.4 1.5 1.24 F.
P8DRB265H 8-x 8 13.4 1.5 1.24
- To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation'is perfo.med with an MCPR of 1.22 (i.e., 1.24 divided by 1.02) for a bundle with an initial MCPR of 1.24.
3-1 o
NEDO-21696 4.5 RESULTS OF THE CHASTE ANALYSIS This code is used, with suitable inputs from the other codes, to calculate the fuel cladding heatup rate, peak cladding temperature, peak local cladding oxidation, and core-wide metal-water reaction for large breaks. The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and channel wetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entite LOCA transient at a single axial plane in a single fuel assembly. Iterative applications of CHASTE determi-' the maximum permissible planar power where required to satisfy the requirements of 10CFR50.46 acceptance criteria.
The CHASTE results presented are:
i.
i e
Peak Cladding Temperature versus time e
Peak Cladding Temperature versus Break Area Peak Cladding Temperature and Peak Local Oxidation versus Planar e
Average Exposure for the most limiting break size Maximum Average Planar Heat Generation Rate (!IAPLHGR) versus Planar e
Average Exposure for the most limiting break size A summary of the analytical results is given in Table 2.
Table 3 lists the i
figures provided for this analysis. The MAPLHGR values for each fuel type in the Pilgrim core are presented in Tables 4A through 4F.
l t
1 4-3
NED0-21696 Table 4A MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Pilgrim Fuel Type: SDB219L Average Planar Exposure MAPLHGR PCT 0xidation CGa/t)
(kW/ft)
(*F)
Fraction 200.0 11.4 2039.
0.018 1,000.0 11.5 2039.
0.017 5,000.0 11.9 2064 0.017 10,000.0 12.1 2093.
0.019 15,000.0 12.3 2126.
0.021 20,000.0 12.1 2126.
0.021 25,000.0 11.3 2013.
0.014 30,000.0 10.2 1866.
0.008 35,000.0 9.6 1787.
0.006 40,000.0 9.0 1707.
0.004 f
4-7
NEDO-21696 Table 4B
'LiPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Pilgrim Fuel Type:
SDB219H Average Planar Exposure MAPLHGR PCT 0xidation (mwd / t)
(kW/ft)
(*F)
Fraction 200.0 11.2 2038.
0.018 1,000,0 11.3 2032.
0.017 3,000.0 11.8 2056.
0.017 10,000.0 12.2 2102.
0.019 15,000.0 12.3 2131.
0.021 20,000.0 12.1 2128.
0.021 25,000.0 11.3 2015.
0.015 30,000.0 10.2 1866.
0.008 35,000.0 9.6 1787.
0.006 40,000.0 9.0 1706.
0.004 4-8
NED0-21696 Table 4C MAPLHCR VERSUS AVERAGE PLANAR EXPOSJRE Plant: Pilgrim Fuel Type: 8DB262 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)
(kW/ft)
(*F)
Fraction 200.0 11.1 2032.
0.016 1,000.0 11.3 2028.
0.015 5,000.0 11.9 2071.
0.017 10,000,0 12.1 2061.
0.016 s
15,000.0 12.2 2091.
0.018 20,000.0 12.1 2104.
0.019 25,000.0 11.6 2049.
0.016 30,000.0 10.7 1928.
0.010 35,000.0 9.8 1803.
0.006 40,000.0 9.2 1719.
0.005 4_9
Table 4D l
MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PlanL: Pilgrim Fuel Type: P8DRB265L Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)
(kW/ft)
(OF)
Fraction 200.0 11.6 2125.
0.023 i
i 1,000.0 11.6 2127.
0.023 5,000.0 12.1 2136.
0.022 10,000.0 12.1 2102.
0.020 15,000.0 12.1 2108.
0.020 20,000.0 11.9 2091.
0.019 25,000.0 11.3 2012.
0.015 30,000.0 10.7 1919.
0.010 35,000.0 10.2 1832.
0.008 l
40,000.0 9.6 1746.
0.006 l
l l
l l
l 4-10 l
L
}
Table 4E l
t MAPLHGR VERSUS NIERAGE PLANAR EXPOSURE Plant: Pilgrim Fuel Type: P8DR3282 Average Planar Exposure MAPLHGR PCT 0xidation I
(mwd /t)
(kW/ft)
(OF)
Fraction 200.0 11.2 2087.
0.020 1,000.0 11.2 2083.
0.020 5,000.0 11.8 2110.
0.021 10,000.0 12.0 2097.
0.020 15,000.0 12.1 2108.
0.020 20,000.0 11.8 2088.
0.019 25,000.0 11.3 2011.
0.015 30,000.0 11.1 1961.
0.012 35,000.0 10.4 1860.
0.008 40,000.0 9.8 1783.
0.006 1
4-11 L
NEDO-21696 Table 4F MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Pile-im Fuel Type: P8DRB265H Average Planar Exposure MAPLHGR PCT Oxidation (mwd /t)
(kW/ft)
(OF)
Fraction 200.0 11.5 2118.
0.022 1
1,000.0 11.6 2121.
0.022 5,000.0 11.9 2115.
0.021 10,000.0 12.1 2105.
0.020 15,000.0 12.1 2113.
0.020 20,000.0 11.9 2096.
0.019 l
25,000.0 11.3 2015.
0.015 f
30,000.0 10.7 1920.
0.011 35,000.0 10.2 1834.
0.008 40,000.0 9.6 1747.
0.006 I
4-12
.~
APPENDIX A LOSS-OF-COOLANT ACCIDENT ANALYSIS WITH NO CORE SPRAY HEAT TRANSFER CREDIT A.1 Introduction 4
This Appendix describes the methods by which conservative MAPLHGR multipliers were determined for application to the MAPLHGR values reported in Section 4 of this report, assuming that no credit is l
taken for core spray heat transfer.
The_ input changes to the ap-proved 10CFR50 Appendix K computer codes are described in Section.
[
A.2, the depressurization rate sensitivity is disciassed in Section A.3, the results of the analysis are given in Section A.4, and the conclusions are presented in Section A.S.
A.2 Input Changes to the LOCA Analysis The approved versions of the SAFE, REFLOOD, and CHASTE codes were applied to the Pilgrim Plant as described in Section 4 of this report, with the input changes described below. - No changes were made to the approved computer codes.
The postulated effect of cracks in the core spray spargers is to deprive the' hot assembly of adequate spray flow during a LOCA.
This effect is represented by setting the spray heat transfer coefficients in the CHASTE heatup code to zero, from their Appendix K values of 3 0, 3.5, j
2 and 1.5 BTU /(hr-ft _oF) for the corner, other outside, and inside rods, respectively.
In the standard Appendix K analysis, the non-zero spray heat transfer coefficients are applied from the time rated core spray flow is achieved.
until the time that the hot node in the hot assembly is refloodea.
4 2
Af ter the reflooding time, a heat transfer _ coefficient of 25 BTU /hr ft 0F) ~is applied to all fuel rods, which is sufficient to cause the peak cladding temperature to decrease.
A-1
NED0-21696 In the p.msent analysis, credit was taken for a heat transfer coeffi-2 cient of 25 3TU/(hr CF f t ) applied to the outside of the channel starting at the time when the water level in the bypass (space between the channels) fills to the elevation of the hot node. Normally this outside channel cooling credit is not taken because it is not needed when core spray cooling is present.
In the present analysis, credit for outside channel cooling is appropriate because the cool channel will act as a sink for the heat radiating from the uncooled rods to the channel.
The two input changes described above were made to the C. 4JTE heatup code, and no changes were made to the SAFE blowdown code, or the REFLOOD refill code.
4 A.3 Depressurization Rate Sensitivity The sensitivity of the depressurization rate (which is calculated by the SAFE code) to the global core spray heat transfer coefficient, was investigated. If it is postulated that, in the worst case, all assemblies are deprived of spray flow, then it is appropriate to set the SAFE spray heat transfer coefficient to zoro.
This was done in a comparison case for the Design Base Accident (DBA), which is later shown to still be the limiting break. The change in the cal-culated uncovery time of the hot node, and in the curve of pressure versus time, was found to be negligible.
It was therefore concluded that no input changes in the SAFE code were appropriate for thu pre-sent analysis.
A.4 Analysis Results A.4.1 Large Break Analysis The identification of the limiting break for this analysis follows the approach in the body of this report.
Table A-1 shows the total uncovered time for several breaks, taken frca computer runs used to generate Figure 6 of this report. These uncovered times are t-expected to be unchanged for this analysis, since the effect on A-2
NED0-21696 the depressurization rate was found to be negligible in the pre-vious section. The results of CHASTE calculations show that the DBA remains the limiting break.
A.4.2 Small Break Analysis Break sizes smaller than 1.0 ft2 are not limiting as shown in the comparison of the uncovered times in Table A-2.
All breaks in Table A-2 have an uncovered time which is equal to or less than the 154 second uncovered time fo,r the DBA, except for the 0.100 ft2 break. For all these breaks (except the 100 ft2, which is described later) the Peah Cladding Temperature (PCT) will be less than the 2200 0F calculated for the DBA, because the time of hot node uncovery decreases with decreasing break size, and the decay heat decreases with increasing break size, and the decay heat decreases with increasing time. Thus the 0 900 ft2 break will have a lcWer PCT than the DBA, even though both breaks have equal uncovered times, because the 0.900 f t2 break has an uncovery time of 71 seconds, which is much later than the DBA uncovery time of 20.6 seconds. The decay heat at 71 seconds is less than at 20.6 seconds, so the heatup in the uncovered period for the 0 900 ft2 break is less than the heat-up for the DBA.
For the 0.100 f t2 break, which has an uncovered time only 3 seconds more than the DDA, the PCT is calculated by the small break model to be less than 17000F.
A.5 Conclusions Using the DBA as the limiting break, a bounding analysis was performed with the CHASTE code. The results of the calculations show that for each of the fuel types, a MAPLHGR multiplier which is independent of exposure must be applied to the MAPLHCR values given in Section 4 of this report. These MAPLHGR multipliers are given in Table A-3 The use of these multipliers conservatively determines the MAPLHGR required to keep the PCT below 22000F for the DBA, with no credit assumed for core spray heat transfer.
A-3
NED0-21696 The calculations described in this Appendix were performed at the request of the Boston Edison Company (BECO) in order to support BECO's proposal to return to service taking no credit for core spray heat transfer. The technical justification for such calculations has been presented in this Appendix. However, General Electric considers the assumption of no core spray heat transfer credit to be excessively conservative, based on the calculations which support the continued structural integrity of the core spray spargers (presented in Reference A-1) and the many recognized.conservatisms in the current LOCA models.
A.6 References A-1 Supr.ement 1 to Supplemental Reload Licensing Submittal for Pilt?im Nuclear Power Station Unit 1 Reload 4, NEDO-24224-1 Supp).ement 1, March 1980.
t l
A-4
NEDO-21696 Table A-1 Pilgrita Large Break Results LPCI Injection Valve Failure; 2LPCS + HPCI + ADS Available No Core Soray Heat Transfer with Channel Cooling Break Size Uncovered Chase (ft.2)
Tima (sec)
P,CT (of)*
l 4 343 (DBA) 154 2200 l
3 474 (80% DBA) 145 21428*
2.606 (60% DBA) 133 2062**
1.000 149 20688*
- MAPLHGR = 11.55 kw/ft. for 8D262 fuel at 10,000 mwd /t
- Conservatively estimated by using temperature differences for a MAPLHGR of 11.08 kw/ft.
-A-5
NEDO-21696 Table A-2 Pilgri.', Small Break Uncovery Time Break Uncovered Failure 2
Size (f t )
Time (sec)
Assume d, 0 900 154 LPCI Injection Valve 0.800 133 0.700 131 0.600 102 0.500 95 0.400 90 0.300 75 0.200 107 0.150 109 HPCI Injectier Valve 0.100 157 0.080 15 1 0.060 155 0.040 132 A-6
[-
NEDO-21696 Table A-3 MAPLHGR Multipliers Assuming No Core Spray Heat Transfer Credit Fuel Type Core Flow 2 90% Rated Core Flow < 90% 9-ted 8DB219L 0.93 0.85 8DB219H 0.93 0.85 8DB262 0.94 0.86 P8DRB265L 0.91 0.84 P8DRB282 0.92 0.85 F8DRB265H 0.90 0.82 l
A-7
_ _ - - _ _ _ - _ - _ _ _ _