ML18038A382
| ML18038A382 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/31/1987 |
| From: | Charnley J, Lambert P, Zarbis W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML17055D460 | List: |
| References | |
| 23A5862, 23A5862-R, 23A5862-R00, NUDOCS 8712290219 | |
| Download: ML18038A382 (24) | |
Text
23A5862 Revision 0
Class I October 1987 23A5862, Rev.
0 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR NINE MILE POINT NUCLEAR POWER STATION UNIT 1, RELOAD ll Prepared:'
~ -~C" P. A. Lambert Fuel Licensing Verified:
W. A.
r s
Fuel L sing
"'87122902i9 8712l8 poR oooo'ooooooo
(
i
',P ". 'DR Approv
- arnley, Manager F
Licensing NUCLEAR ENERGY BUSINESS OPERATIONS ~ GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA95125 GENERAL I ELECTRIC 1/2
23A5862 Rev.
0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Niagara Mohawk Power Corporation (NMPC) for NMPC's use with the U.S. Nuclear Regulatory Com-,
mission (USNRC) to amend NMPC's operating license of the Nine Mile Point Nuclear Station Unit 1.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting informa-tion in this document are contained in the Contract between Niagara Mohawk Power Corporation and General Electric Company for Fuel Assembly Fabrication and Related Services for Nine Mile Point Nuclear Power Station Unit No. 1, and nothing contained in this document shall be construed as changing said con-tract.
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; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or war-ranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or 'that such use of such informa-tion may not infringe privately owned rights; nor do they assume any responsi-bility for liability or damage of any kind which may result from such use of such information.
3/4
23A5862 Rev.
0 ACKNOWLEDGEMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by E.
G. Thacker II, P.
G. Warzek; and E. 0. Electona of the Nuclear Fuel and Engineering Services Department.
23A5862 Rev.
0 1.
PLANT-UNI UE ITEMS (1.0)*
Plant Parameter Differences:
Bundle Enrichment and Gadolinia Distribution+*
Appendix A 2.
RELOAD FUEL BUNDLES (1.0 2.0 3.3.1 AND 4.0)
Fuel T e
C cle Loaded Number Irradiated P8DNB277 P8DNB277 P8DRB299 148 200 New BD321B 10 176 Total 532 3.
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle:
Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:
e Assumed reload cycle core average exposure at end of cycle:
Core loading pattern:
21~500 MWd/HT 21,172 MWd/MT 22,439 MWd/MT Figure 1
"( ) Refers to area of discussion in "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986.
A letter "S" preceding the number refers to the appropriate section in the United States Supplement, NEDE-24011-P-A-8-US, May 1986.'*See "Supplement, to Nine Mile Point Unit 1 SAFER/CORECOOL/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31446P-l, September 1987.
23A5862 Rev.
0 4.
CALCULATED CORE EFFECTIVE MULTIPLICATIONAND CONTROL SYSTEM WORTH NO VOIDS 20 C (3.3.2.1.1 AND 3.3.2.1.2 Minimum Shutdown Margin, BOC, keff
,Uncontrolled Fully Controlled Strongest Control Rod Out R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, dk 1.098 0.954 0.988 0.0 5.
STANDBY LI UID.CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) 600 Shutdown Margin (dk)
(20'C Xenon Free) 0.041 6.
RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(Cold Water Injection Events Only)
Void Fraction (X)
Average Fuel Temperature ('F)
Void Coefficient N/A" (d/X Rg)
Doppler Coefficient N/A+ (6/'F)
Scram Worth N/A+ (0)
- 36. 2 992
-5.08/-6.35
-0.236/-0.224
- N ~ Nuclear Input Data, A ~ Used in Transient Analysis
- Generic exposure independent values are used as given in "General Electric Standard Application for Reactor Fuel," NEDE-2401I-P-A-8, dated May 1986.
23A5862 Rev.
0 7.
RELOAD-UNI UE GETAB TRANSIENT ANALYSIS INITIALCONDITION PARAMETERS S.2.2 Fuel Des~in Peaking Factors Local Radial Axial R-Factor Bundle'ower Bundle Flow Initial (MWt)
(1000 1b/hr)
MCPR BOC10 to EOC10-2205 MWd/MT P8x8R 1.20 1.83 1.40 GE828EB 1.20 1.85 1.40 1.051 1.051 6.234 6.277 97.5 99.3 1.23 1.23 EOC10-2205 MWd/MT to EOC10-1102 MWd/MT P8K8R '.20 1.83 1.40 1.091 GE8x8EB 1.20 1.84 1.40 1.051 EOC10-1102 MWd/MT to EOC10 6.234 6.259 97..5 99.5 1.23 1.23 P8x8R 1.20 GE8K8EB 1.20 1.79 1.40 1.79 1.40 1.051 1.051 6.080 6.072 98.6 100.8 1.27 1.28 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization:
Recirculation Pump Trip:
Rod Withdrawal Limiter:
Thermal Power Monitor:
Improved Scram Time:
Exposure Dependent Limits:
Exposure Points Analyzed:
No No No No No Yes EOC10-2205 MWd/MT; EOC10-1102 MWd/MT; EOC10 9.
OPERATING FLEXIBILITYOPTIONS Single-Loop Operation:
Load Line Limit:
Extended Load Line Limit:
Increased Core Flow:
Flow Point Analyzed:
Feedwater Temperature Reduction:
ARTS Program.'aximum Extended Operation Domain:
No Yes
, Yes No N/A No No No
23A5862 Rev.
0 10.
CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Transient Methods Used:
GEMINI*
Flux Q/A hCPR
(% NBR)
(E NBR)
PSxSR GE8xSEB
~Pi uxe Exposure:
BOC10 to EOC10 Loss of 100'F Feedwater Heating 118 117 0.16 0.16 Exposure:
BOC10 to EOC10-2205 MWd/MT Turbine Trip Without Bypass Feedwater Controller Failure 382 108 112 0.11 0.11 108 0.07 0.07 3
4 Exposure:
EOC10-2205 MWd/MT to Turbine Trip Without Bypass EOC10-1102 444 Feedwater Controller Failure 125 Exposure:
EOC10-1102 MWd/MT to EOC10 MWd/MT 117 0.16 0.17 107 0.09 0.09 Turbine Trip Without Bypass Feedwater Controller Failure 447 144 121 0.20 0.20 108 0.08 0.08 11..
LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
S ~ 2. 2. 1 Limiting Rod Pattern:
Figure 9
ACPR MLHGR (kW/ft)
Rod Block
~Reediri 104 105 106 107 108 109 110 Rod Position (Feet Withdrawn) 9.0 9.0 9.5 10.0 10.0 12.0 12.0 P8xSR 0.30 0.30 0.30 0.30 0.30 0.30 0.30 GEBx8EB
- 0. 28 0.28 0.28 0.28 0.28 0.29 0.29 P8x8R 16.62 16.62 16.62 16.62 16.'62 16.62 16.62 GE8x8EB 17.67 17.67 17.67 17.67 17.67 17.67 17.67 Setpoint Selected:
110
- GEMINI Methods are described in "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8-US, May 1986.
10
23A5862 kev.
0 12.
CYCLE MCPR VALUES (S.2.2)
Nonpressurization Events:
P8x8R GE8x8EB Exposure Range:
BOC10 to EOC10 Loss of 100'F Feedwater Heating Fuel Loading Error Rod Withdrawal Error 1.23 1.37 1.23 1.15 1.36 Pressurization Events:
P8x8R Option A GE8x8E8 Exposure Range:
BOC10 to EOC10-2205 MWd/MT Turbine Trip Without Bypass 1.23 1.23 Feedwater Controller Failure 1.19 1 ~ 19 Exposure Range:
EOC10-2205 MWd/MT to EOC10-1102 MWd/HT Turbine Trip Without Bypass Feedwater Controller Failure 1.27 1.21 1.28 1.21 Exposure Range:
EOC10-1102 MWd/MT to EOC10 Turbine Trip Without Bypass Feedwater Controller Failure 1.33 1.21 1.34 1.21 13 ~
OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)
Transient MSIV Closure (No Scram)
'sl (psig) 1273 Pv (psig) 1313 Plant
Response
Figure 10
23A5862 Rev.
0 14 'OADING ERROR RESULTS (S.2 ' ')
Variable Water Gap Misoriented Bundle Analysis:
Yes Event Rotated Bundle Error dCPR 0.08 15.
CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
Nine Mile Point Nuclear Power Station Unit 1 is a Banked Position With-drawal Sequence
- Plant, so the Control Rod Drop A'ccident Analysis is not required.
NRC approval is documented in NEDE-24011-P-A-8-US, May-1986.
16.
STABILITY ANALYSIS RESULTS (S.2.4)
The NRC has approved a bounding stability approach that exempts Nine Mile Point Nuclear Power Station (NMPNPS) from the requirement to routinely perform 'a stability analysis.
NRC review and approval is documented in the letter from C.O.
Thomas (NRC) to H'C. Pfefferlen (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8 'Thermal Hydraulic Stability Amendment to GESTAR II'," dated April 24, 1985.
NMPNPS Cycle 10 differs from previous cycles with the introduction of GE8x8EB fuel.
During was demonstrated that performance.
The NRC stability approach is NRC review of the bounding stability approach, it GE8x8EB fuel h'as only a small impact on stability review concluded approval states that the bounding applicable to reload cores with GE8x8EB fuel.
17.
LOSS-OF-COOLANT ACCIDENT RESULT NEW FUEL (S ~ 2 ~5.2)
LOCA Method Used:
SAFER/CORECOOL/GESTR-LOCA "Nine Mile Point Unit 1 SAFER/CORECOOL/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31446P, June 1987; and.NEDC-31446P, Supple-ment 1, September 1987.
12
23A5862 52 IE+IE pe+i Ps+pe OglE Os+pe 50 IE K EIEIEIEIEIEIKIE 48 Cel~Csl Pc~pc Ccl~lcl ~lcl
<l~l Cc~llcl OB~IRI 46 IE K+El KTEIIn+El@+Col +lE E~IEl E~IK IE 08>pc pc<I~I Cel>lol Pc~Col Cc~lPc lo~lCcl DO~Pc lo~llcl Pc~Pe 42 Pe Cel Oo CclPo pc Oo Dc Po Uo Do POPO Pc Po Oe Pc Qe 40 De~Pe Dc~00 Pc~Do Pe~00 Pe~Do PB~QB 00~08 Do~Pe 00~Pc 00~
Pe~Pe
.38 pe De~pc Do~pc OO~Dc OO~
C~lpc Do~Do Dc~DO popo ~po popo pc~pe pe 36lE lE El ItHEIHHQBOo IE Qo IE E IE QB HHKHHKHHK HD EIKEIKIE 34PB+
+PC PO+
OD+
PO+
PO+
PD+QO
+QO
+QO
+QO PC+Ps PC+PC
+Ps 32K EI EI Qo ElK OB K Qe Q K Oo IE OB IIIE g] K HH KK Q
K 30+
+
Po+
Po+
Po+
Po+
Pe+Pe
+Po
+Qo
+Po
+Po
+
+
28K Qc El HO EI HH K Ig] QB Qo K po KKEl K Qo K HH KK E] gg El QB 20 Ps+Pc Pc+Ps Pc+Os IE+OD OB+IE IE+Po IE+IE Ps+Ps OD+IE IE+IE Qo+
Ps+Pc Pc+Ps 24I@El E~IEI IE~EI E~JEI E~l lE~EI E~lR E~IEO E~I8H E~IHH E~lR E~lEI QZ 22 De~Pc @ Q QOO pe~Do Qo NDB Oo~pe OO~C>l Do~pe Oo@ S 20 Ps+PC +
PD+
QO+
PO+
Ps+PC Ps+Ps PC+Ps PC+PD PC+Pe PC+Ps +
+Oe 18 De Cel 0o Pc ColK 0o Pe Col lm po K Cel po Kpo PB Col Cel 0o Ccl 0o CclK pe 16 Pe De~Pc QO,Pc QO,Dc PoiPc Po OoiDO Pcipo IPO Eel~Col Cclilol Dc~Pe PB OB~K E~IHH ~OO Oe~@ K+E8 KKIE+KIE~K H+HE) Ha~El OB+K 12 10 8
6 2
K DciK QoiDc Po~
Oo~pc Oo~po Dclpo DclPO DciPO Qeipc OB E~IEI Cc~lEI E~ICol E~IS E~IEl Co~lEI R~EI @~Eel E~IK QBOBDcpoPcpoPcpopoOOPcPOPcPBK z+Eln+inln+inln+inln+Inln+inlE+z Ue Pe,pc,pc Pc,Uc Dc,Dc Uc,PB Qe K+US K+KK+K0+K K+El I
I I
I I
I I
I I
I 1, 3 5
7 911 1315171921 2325272931 3335373941 4345474951 FUEL TYPE A
PSDNB277 (Cycle 7)
B PSDNB277 (Cycle 8)
C ss P8DRB299 D
BD321B I
Figure 1.
Reference Core Loading Pattern
23A5862 Rev.
0
)50e 0 I NEU 2 AVE 3 COR RON FLUX SINFACE NEAT FLUX IM.ET FLOV
)50, 4 I YES 2 REL 3 BYP EL PRESS RISEIPSI)
EF VALVE FLOV SS VALVE FLOV
)00m 0 s-I00o 0 54+ 0 50,0
- 0. 0
- 4. 0 Oe0
)00+0 TINE ISECONOS) 200 ~ 0 OI0 I00o 0 TINE ISECONOS) 200+0 l50. 0 I LEV 2 VES 3 TUR LIINCH-REF-SEP-SNOT)
EL STEANFLOV INE STEANFLOV I. 0 I YOI REACTIVITY 2 00'ER REACTIVIT'I 3 SCR N REACTIVITY l 00m 0
- 50. 0 5
0.0 0 -).4 De 0
- 4. 0 I 00 ~ 0 TINE ISECONOS) 200 ~ 0
- 2. 0 0.0
~
l00. 0 TINE ISEJONS) 200e 0 Figure 2.
Plant Response to Loss of 100'F Feedwater Heating (BOC10 to EOC10) 14
23A5862 Rev.
0 160 ~ 0 I NEUTRON FLU 2 AVE SURF 3 CORE INLET HEAT FLUX LOV 300 ~ 0 I VESSEL PRE RISE(PSI) 2 SAFETY VAL FLOV 3 RELIEF VALV FLOV 100. 0 IS I
30.0 200 ~ 0 144. 0
- 0. 0 OI0 2e0 TINE (SECONOS) 6e0
- 4. ~
0 ~ 0 2IO 0 ~ 0 TINE (SECOINS) 6IO 200. 0 I LCVCL(INai-2 VESSEL ST 3 tURSINE STE CF SEP-Sl(RT)
FLOV 1%'LOV 1 ~ 0 I VOID REACTI 2 DOPPLER RE 3 SCRAM REAC ITY TIVITY IVITY 100. 0 0.0
~0
- 0. 4 I
8 I
Q
-1. ~
-100. 0
- 0. 0
- 2. 0 TINE (SECONDS) 4,0 2m 0
- 0. 0
- 2. 0
- 0. 0 TINE (SECCteS)
- 8. 0 Figure 3.
Plant Response to Turbine Trip Without Bypass (BOC10 to EOC10-2205 MWd/MT) 15
23A5862 Rev.
0
)50+ 0 I NEUTRCO( FL 2 AYE S(NF kEAT FLUX 3 CORE INLET LOV
)5e. o I VESSEL PRE 2 SAFETY YAL 3 RELIEF VAL 4 BYPASS YAL RISE(PS I)
FLOV FLOV FLOV I 00. 0
à leo 0
I I(I Se. o
- 50. 0 0.0 O. 0 OI 0 20e 0 ee. o TINE (SECONOS) oos 0 Os 0
- 20. 0 TIIIE (SECOIO)S) coo 0 60+0 l50. 0 I LCYCL(IN(3h Cf CCP-SKRT) 2 VESSEL S FLOV 3 TURBINE S NFLOV l ~ 0 I YOIO REACT) 2 OOPPLER RE 3 SCRAN REAC I TY TIYITY YITY lee. 0
- 50. 0
- 0. 0 R
8 Q -I.O 0 0
- 0. 0
- 20. 0
- 40. 0 TINE (SECOHOS)
- 60. 0 2.0 O. 0
- 20. 0 ee. 0 TINE (SECOI0$ )
- 60. 0 Figure 4.
Plant Response to Feedwater Controller Failure (BOC10 to EOC10-2205 MWd/MT) 16
23A5862 Rev.
0 ISD. 0 I NEUTRON FL 2 AVE SURFAC 3 CORE INLET HEAT FLUX LOV 300.D I VESSEL PRE 2 SAFETT VAL 3 RELIEF VAL RISE(PS I )
FLOV FLOV
)00. 0 I
ht SDe0 200'
)00. D
- 0. 0 De 0 2e0 I ~ 0 TINE (SECOIOS) 6ID D. D De 0 2o0 TINE (SECONOS) 6m 0 200. 0 I LCVCL(INOI-Cf'CP SKRT) 2 VESSEL STEA FLOW 3 TURBINE ST HFLOV I.D I VOIO RCACTIVITY 2 OOPPLER RE TIVITY 3 SCRAN REAC IVITY IDD. 0
- 0. 0 0 ~ 0 I
8 I
1.0
)00o0
- 2. 0 TINE (SECONOS)
D ~ 0 6I0 2 ~ D.
0.0
- 2. 0 TINE (SECONOS)
- 4. 0
- 6. 0 Figure 5.
Plant Response to Turbine Trip Without Bypass (EOC10-2205 MWd/MT to EOC10-1102 MWd/MT) 17
23A5862 Rev.
0 ISO' I NEUZRON FL 2 AVE SURFA 3 CORE INLET feAT FLUX LOS
)50.0 I VESSEL PRE 2 SAFETY VAL 3 RELIEF VAL 4 BYPASS YAL RISE(PS I)
FLUV FLOY FLOV lDD. 0
)00 0
8
)B K
ht 50' 5D. 0 0 ~ 0 D. 0 DI 0 20' 40e0 TINE (SECONOS) 60+0 DI 0
- 20. 0 DDo 0 ZINE (SECONOS) 60'
)50. 0 I LEVCL(INCI-2 VESSEL ST 3 TURBINE ST D. SCP SKRT)
FLOY NFLOV I VOIO RCACTI ITY 2 DOPPLER RE ZIVITY 3 SCRAN REAC IVITY IDDI0 50'
~0 D ~ 0 5
S I
~0 1.0 0+ 0 0.0
- 20. 0
- 40. 0 ZINE (SECONOS) 00e 0
- 2. ~
D. 0
- 20. 0 ZINE (DECO)eS)
- 40. 0 60 ~ 0 Figure-6.
Plant Response to Feedwater Controller Failure (EOC10-2205 MWd/MT to EOC10-1102 MWd/MT) 18
23A5862 Rev.
0 ISO+ 0 I NEU'IRON FL 2 AVE SURFAC 3 CORE IM.ET tCAT FLUX LOV 300m 0 I VESSEL PRE 2 SAFETY VAL 3 RELIEF VAL RISE IPSI)
FLOV FLOV
~\\
Cl l00. 0 IX SOe 0 200 ~ 0 l 00. 0
- 0. 0 Oo0 2I 0 TINE ISECOIOS)
Ie0 6s0 0.0 0.0 2+0 0 ~ 0 TINE ISECONOS) 6I 0 200. 0 I LCVCLIIUQI-G.-SEP-SNAT) 2 VESSEL STEA FLOV 3 TURSINE STE NFLOV I ~ 0 I VOIO RCACTI ITY 2 IXV'PLER RE TIVITY 3 SCRAH REAC IVITY
)00m 0 0.0
~1 0.0 5
8 I
Q -).0 I00. 0
- 0. 0
- 2. 0
<o0
'flNE ISECONOS) 6 ~ 0
- 2. 0 0 ~ 0
- 2. 0 TINE (SECOIOS) 4 ~ 0
- 6. 0 Figure 7.
Plant Response to Turbine Trip Without Bypass (EOC10-1102 MWd/MT to EOC10) 19
23A5862 Rev.
0
)50e 0 I NEUTRON FL 2 AVE SURFA 3 CORE INLET MAT FLUX LOU 150. 0 I VESSEL PRE 2 SAFETY VAL 3 RELIEF VAL 4 BYPASS VAL RISE(PS I)
FLOV FLOV FLOV 100. 0 100. 0
)S I
- 50. 0
- 50. 0 0 ~ 0 0,0 0.0 20e 0 40I 0 TIRE (SECOHOS)
- 60. 0 0,0 20m 0 TIRE (SECONOS) 40I 0 60m 0 150. 0 I LEVEL(INC(1 2 VESSEL STE 3 TURBINE STE Cf-SCP CKRT)
FLOV
)0:LOV 1.0 I VOIO REACT) 2 OOPPLER RE 3 SCRAN REAC ITY TIVITY IVITY 100,0
- 50. 0 0.0 I
8lI I~ -1.0
- 0. 0
- 0. 0 20 ~ 0
- 00. 0 TIRE (SECOIOS) 60o 0
-2.0
- 0. 0
- 20. 0 TIRE (SECONDS)
<Oo 0
- 60. 0 Figure 8.
Plant Response to Feedwater Controller Failure (EOC10-1102 MWd/MT to EOC10) 20
23A5862 Rev.
0 2
. 6 10 14 18 22 26 30 34 38 42 46 50 47 26 6
6 6
26 43 39 35 6
31 27 6
26 16 16 16 36 36 16 36 36 16 36 36 16 16 26 23 19 6
15 26 16 36 36 36 16 26 26 16 16 16 26 NOTES:
1.
No. indicates number of notches withdrawn out of 48.
" Blank is a withdrawn rod.
2.
Error Rod's (22, 31).
Figure 9.
Limiting Rod Pattern 21
23A5862 Rev.
0 ISO' NEUTRON 2
VE SURF 3
ORE INL UX E HEAT FLUX FLOV 300 ~ 0 I VESSEL P
SS RISE(PSI) 2 SAFETY V VE FLOV 3 RELIEF V VE FLOV
)00.0
)S I
- 50. 0 200. 0
)00.0 D. 0 0 ~ 0 Sa0
'TINE (SECONOS) 0 ~ 0 0 ~ 0 Se0 TINE (SECONOS) 200 ~ 0 I LCYC.( IH" )-RCr-SCr-SNOT) 2 VESSEL ST ANFLOV 3 TUROINE EAHFLOV
).D I VOIO RCA 2 OOPPLER 3 SCRAH RE IVITY
~
ACTIVITY TIVITY 100. 0
- 0. 0 p5 0,0 IJ l
I M -).0
)00.0 5.0 TINE (DECO)OS)
-2 ~ D 0 ~ 0
- 5. 0 TINE (DECO@)S)
Figure 10.
Plant Response to MSIV Closure No Scram 22
23A5862 Rev.
0 APPENDIX A PLANT PARAMETERS DIFFERENCES GETAB and Transient Anal sis Initial Conditions To accurately reflect actual plant conditions, the values listed in Table A-1 were used in the transient analyses instead of the values reported in NEDE-24011-P-A-8-US, May 1986.
Table A-1 PLANT PARAMETERS Parameter Rated Steam Flow Turbine Pressure Relief Valve Low Setpoint Non-Fuel Power Fraction Anal sis Value 7 32x106 lb/hr 950 psig 1102 psig 0.039 NEDE-24011 Value 7.29x106 +0.2X lb/hr 956 +2 psig 1090
+1% psig 0.035 23/24 (FINAL)
/
r
(,