ML20244B413
| ML20244B413 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/31/1987 |
| From: | Kim H, Robare D, Rogers A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20244B402 | List: |
| References | |
| DRF-A-3119, DRF-A00-03119, NEDC-31515, NUDOCS 8906130104 | |
| Download: ML20244B413 (58) | |
Text
. - - _ - _
NEDC-31515 DRF A00 03119
~ Class II bECEMBER 1987 IAS 108-1187 MAXIMUM EXTENDED 1 DAD LINE LIMIT I
AND FFEDUATER HEATER OUT OF SERVICE ANALYSIS FOR i
ENRICO FERMI ATOMIC POWER FIANT UNIT 2 I
H.T. Kim .
Technical Project Engineer ,
\
l Approved: Approved: e /AEdu /'
A. E. Rogers, (inager [D. J. Robare, Manager Plant Performance Engineering Plant Licensing Services
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IEDCo31515 IMPORTANT NOTICE REGARDING
~
CONTENTS OF THIS REPORT l .
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Flosse Raad Carefully i
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The only undertakings of General Electric Company respecting information in this document are contained in the contract between Detroit Edison Company (Deco) and General Electric Company, as identified in the purchase order for this report and nothing contained in this document aball be construed as f changing the contract. The use of this information by anyone other than Dn.Co or for any purpose other than that for which it is intended, is not author-
- imed; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the complete.
mass, accuracy, or us,efulness of the information contained in this document.
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, NEDC 31515 l
.. l 1
J CONTENTS
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- 1. INTRODUCTION AND
SUMMARY
tel 1.1 Introduction 11 1.2 Summary 13
- 2. TRANSIENT EVENT ANALYSIS 2.1 Limiting Transients 2-1 {
21 2.2 Slow Tiow Runout Event 2-2 j 2.3 Overpressurization Analysis 2-2 '
1 2.4 Rod Withdrawal Analysis 23 2.5 Operating Limit MCPR 23
- 3. THERMAL-MYDRAULIC STABILITY 3-1 4.
ECCS PERFORMANCE 4.1 TJHOS 4-1 4.2 MELLIA 4-1 4.3 Conclusion 42 i 43 )
- 5. CONTAINMENT ANALYSIS 5-1
- 6. TEEDWATER N0ZZ12 AND PEEDVATER SPARCER TATICUE USACE 6-1 6.1 Method and Assumptions 61 6.2 Teedvater Nozzle Tatigue 6-2 6.3 Teedvater Sparger Tatigue 6-3
- 7. ACOUSTIC AND TIDW INDUCED IAADS IMPACT 7-1
- 8. REACTOR PROTECTION SYSTEM IhW POWER SETPOIhT 81
- 9. APRM INSTRUMEh7ATION SETPOINT POR MELLIA 9-1
- 10. OPERATIONAL LIMITATION 10-1
- 11. REFERENCES 11-1
-111-
'3
NEDC 31515 l TABLES zaus nu. .
rui 21 Initial Conditions for ME11L and FVHOS Transient '
Analysis 2-4 2-2 Key Input Parameters for Transient Analysis 25 23 Summary of Transient Peak Values Results 26 24 Required Operathg Lielt McFR Values 2-7 25 iinerpressurisation Analysia Results {
MSIV Closure (Flux Sr. ras), 1021 Fewer 28 {
61 Feedwater Nozzle Fatigue Usage for a 32-Year Seal .
Refurbishment . Period 64 62 Feedwater sparger Fatigue Usage for a 32-Year Se~al Refurbishment Period 65 O
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, 2122fSTRC10NS 21pra 21112 *
. ZAst 1-1 NE12.L Operating Region 15 l 21 Feedwater Controller Failure, Maximum Dengnd .
)- 1021 Power /1001 Core n ow, Rated Tg* 370 F 29 2-2 Feedwater Controller Failure, Maximum Degand 1028 Fower/1001 Cere n ow, Rated 7,9 320 F 2 10 23 Feedwater Controller Failure, Maximum Degand 1021 Power /1001 Core Flow Rated 7 ,270 F 2 11 2-4 FoodwaterCentro11erFailure,MaximumDgnand 1021 Power /751 Core n ow, Rated 7,y 270 F 2-12 25 Turbine /GeneratorTripwithSypassFailvge 1021 Power /1001 Core Mew, Rated 7, 270 F --
2-13 26 Turbine / Generator Trip with Bypass Failgre 1021 Power /751 Core Flow, Rated 7,y 270 F 2 1f. j 27 Feedwater Controller Failure, Maxian Degand w/o Sypass 1021 Power /1001 Core M ow, Rated 7,9 370 F 2-15 28 Feedwater controller Failure, Maximum Degand w/o Bypass 1 1021 Power /1001 Core Flow, Rated Tyy 320 F 2-16 29 Feedwater Controller Failure, Maximum Degand w/o Sypass 1021 Power /1001 Core n ow, Rated T,9 270 F 2 17 2-10 Feedwater Controoler Failure, Maximum Dgnand w/o Bypass 1021 Power /751' Core n ow, Rated 7,y 270 F 2 18 61 Feedwater Nozzle Fatigue Damage Due to PUH05 as
, a Funetton of Time 66 6-2 Feedvater Sparger Fatigue Damage Due to FWHOS as a Function of Time 67 91 APRM Setpoint Configuration for NELLL Domain 9-2 10 1 parrower Extended Operating Region 10 2 ofeedwater temperature
- V.
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NEDCo31515 t> . .
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- 1. INTEDDUCTION AND SIROERY
1.1 INTRODUCTION
This report presents the results of a safety and impact evaluation to support (1) operation of the Enrico Termi Atomic Power plant Unit 2 (Fermi 2)
Cycle 1 in an extended power / flow operating envelope called the Maxim a Extended lead Line Limit (ME11L) domain and (ii) continued operation of the plant during the operating cycle with feedvater heater (s) out of service (FVH05) in the normal operating power / flow map as well as in the MELLL domain.
1.1.1 Marf eue Emended had Line t<=f t analyst rwrit te Two factors.which restrict the flexibility of a boiling water reactor (SVR) during power ascension in proceeding from the low-power / low core Gow condition to the high power /high core n ow condition are: (1) the normal operation power / flow map (Reference 1) and (2) Preconditioning Interim Operating Management Recommendation (PCIDEs).
If the rated load line control rod pattern is maintained as core flow is increased, changing equilibrium zenon concentrations will result in Iss than rated power at rated core flow. In addition, fuel pellet cladding interaction considerations inhibit withdrawal of control rods at high power levels. The combination of these two factors can result in the inability to attain rated cora power directly.
An extended operating envelope above rated load line permits improved power ascension capability to full power and provides additional flow range at rated power to compensate for reactivity reduction due to exposure during an operating cycle. Moreover, ability to operate the plant at rated power for a wider core now range enhances fuel cycle economics if the now control spectral shift mode of operation is implemented.
The expanded operating domain offered by the MELLIA can be utilized to improve operating flexibility and capacity factor for the Termi 2 plant. The MEL11 domain is defined and illustrated in Figure 1-1. This region is bounded 1-1 b
_ _ _ _ _ _ _ _ _ - _ - _ _ _ _ )
. WEDC 31515 by the rated power for flow rarse 751 to 1001 of rated, its corresponding power / flow constant red line, and 1001 rod line. .
I .
1.1.2 Taedwater Heater (s) our of service trunos)
Design evaluations reported in the Fermi-2 Final Safety Analysis Report l (FSAR) justify operation with full feedwater heating which corresponds to a rated feedwater temperature * (Tg) of 420*F. Operation with reduced feedwater temperature occurs in the event that certain stage (s) or string (s) of indi-vidual heaters become inoperable during the fuel cycle or due to the inability-of the feedwater heaters to achieve the design feedwater heating +.
Loss of feedwater heating from the last two stages (highest pressure) heaters would result in the h1 5hest temperature reduction. 1ess of heating from the early stages of low pressure heaters would result in only a slight reduction of feedwater temperature. It is estimated that feedvater temperature reduction during an operating cycle due to inoperable out of service heater stages is less than 100*F.
Safety and impact evaluations for a 150*F reduction of rated feedwater temperature from normal 420*F were performed and the results are presented in this report. The 150'T reduction was chosen to cover any partial feedwater i heating conditions beyond the estimated 100*F reduction such as the one which has occurred during the cycle 1 startup at Fermi-2 (Reference 2).
- To simplify discussion en WHOS operation, the term " rated feedwater tempera-ture' is used in this report to mean 'feedwater camperature at 1001 core thermal power *. The feedwater at any giver. core power and flow is dependent upon the combination of operable heater (s) in each of the two strings of heaters and the core power level.
4Both of these two partial f.edsAter heating modes are denoted as WHOS in this report.
1-2
't l
. NEDC 31515 1
l 1.2
SUMMARY
The limiting abnormal operating transtants reported in the Updated Tinal Safety Analysis Report (UFSAR) were reevaluated at and of cycle 1 for (102P,75F)* and (102P,100F) conditions with rated feedwater temperature reduction of 150 F. The purpose was to evaluate effects on the cycle 1 Operating Limit MCPR (OLMCPR) of Fermi-2 plant operation in the MELLL domain and the FWHOS operation in the normal operating power / flow map as well as in the MELLL domain. The transient analysis was also performed to support the Fermi 2 Technical Specification OLMCPR for the Main Turbine Bypass Out of Service (MTBOS) option. The results are given in Tabla 2-3 and 2-4 The results show that the MCPR value for the FWCF event with MTB05 under FVHOS for rated feedwater temperature below 370'F exceeds the Fermi 2 Technical f
Specifications OLMCPR values. The MCPR values for all other transients are bounded by the present Fermi-2 Technical Specification OLMCPR's. Therefore, no change in OLMCPR is required for operation in the MELLL domain and for FVHOS operation with a 150 F rated feedwater temperature reduction in the normal operating power / flow map as well as in the MELLL domain However, OLMCPR for the MTBOS case should be increased to the approprate values for FVHOS operation below 370 F rated feedwater temperature.
The analysis results also show that performance in the MELLL domain with the FWHOS is within allcwable design limit for overpressure protection, thermal hydraulic stability, ECCS performance, and teactor internal loads.
Evaluation of containment LOCA response in the KELLL domain with FWHOS operation indicates that all the containment parameters of concern are below the design limits. An increase in the vent system load was calculated, which has the potential to lead to vent system stresses above the allowable stresses. Evaluation of this increased load is to be performed to determine acceptability.
+ This denotes 102% power and 75I cote flew condition.
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IIEDC 31515 Evaluation of the FWHOS effect en the feedwater messle and feedwater sparger fatigue was performed and the thermal sleeve seal refurbishment interval acceptable for the FWHOS operation was defined.
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j The recommended scram bypass setpoints for TCV/TSV fast closure are defined for the range of feedwater temperature reduction to maintain the 301 rated power for bypassing this scras function.
rod block scras configuration are redefined to accomodate operation in the i MELLL domain. The same protection margin to the operating map as thL current configuration is maintained with the new setpoint. j
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- All core wide transients described in FSAR Chapter 15 were examined for operation in the ME11L region and FWOS operation in the normal operating power / flow map as well as in the ME11L region. E ree limiting abnormal operating transients were reevaluated in detail. They are:
- a. Turbine / Generator Trip with Bypass Failure (T/CINBF) b.
Feedwater Flow Controller Failure (W CF)
- c. 100*F less of Teodwater Heating (LFW)
The reevaluation was performed at (102F,75F) and (102F.100F) for a rated feedwater temperature reduction of 150*F. The evaluation was performed at the and of cycle 1. Plant best balance, core coolant hydraulics and nuclear transient parameter data were developed and used in the transient analysis.
The initial conditions for the transtant analysis are presented in Table 2-1.
Key input data used in the transient analysis are listed in Table 2 2. De computer model described in Reference 4 was used to simulate both the WCT and T/CINEF events. De transient peak values and critical power ratio results for the two cases analyzed are summarized in Table 2-3 and 2-4 respectively with the licensing values for comparison. The transient responses are presented in Figures 2-1 to 2-10.
The results show that for the case of main turbine bypass in service, the current Termi-2 Technical Specification 01MCFR's (Option A - 1.25 and Option B - 1.24) bound the DIRCFR's required for operation in MEILL region and FWOS operation with a 150'F reduction of rated feedwater temperature in the normal operating power / flow map as well es in the MEL11 domain.
The Fsrai-2 plant specific analysis for the 100*F LNH transient was performed for the MELLL and FWOS operation using the General Electric Eree-Dimensional SVR Core Simulator (Reference 5). ne MCPR results are given in Table 2-4.
The results show that the 100*F L NH ACFR at (100P,75F) condition is smaller than tha value at (100P,100F) condition indicating that 2-1
~
the LFWH ACFR in the MELLL region is bounded by the FSAR value. Also, the results show that the ACFR for the 100*F loss initiated free 270',F is bounded by the 420'T initiation case indicating that the 100*F LFUM has less effect on cold feedwater than on the normal feedwater temperature. Pharthermore, it is
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1ess likely that a 100*F loss would occur at an initial feedwater temperature of 350*F or 270*F during FWHOS operation. brefore, the loss of feedwater 1 heating for the MELLL and FVHOS operation is adequately bounded by the 420 F normal feedwater heating CPR results.
For the MTSOS option the limiting transient, i.e.. FWCF event was analyzed for the ME11L and the FWHOS operation. 2he results are included in Tables 2-3 and 2 4 for use in modifying the 01MCFR for the NTROS option as contained in the current Ferai 2 Technical Specifications.
2.2 SIM F1M RUNDUT EVENT The Kg factor is designed to maintain cote thermal margins in the event of a slow flow runout event. That is, K gdefines a set of Mr?R limits as a function of core flow such that a slow flow runout event 1stitiated from any given power / flow point will result in a minimus CFR no less than the safety limit MCFR.
An evaluation of the slow flow runout event initiated from the liatting point in the MELLL region up to 102.55 maximum core flow verified that the existing K g curves are applicable for the extended operating region.
2.3 OVERPRESSURIZATION ANALYSIS 14 war initial operating pressure and steam flow rate provide better overpressure protection for the most limiting Main Steaaline Isolation Valve (MS1V) closure event (flux scras) during MELLL and FWHOS operation. This event was analyzed at (102F,75F) for 150'T rated feedwater temperature reduction and compared with the FSAR case. Table 2-3 also indicates that the peak pressures for the T/CTNBF and FWCF events analyzed are below the ASME code value of 1375 psig. Hence, it is concluded that the pressure barrier integrity is maintained under MELLL and FWHOS operation conditions. .
2-2
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l
, NEDC-31515 2.4 ROD WITHDRAWAL ERROR -
The current rod block monitor (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 licensing basis evalua, ion at the rated condition is conservative for operation in the MELLL region above the rated load line.
l A generic rod withdrawal error (RVE) analysis was performed to examine the effect of the initial feedwater temperature. An initial condition of 250*F was used to bound all FWHOS operations. The analysis indicated that the initial steady state feedwater temperature has negligible effect on ACPR.
Therefore, it is included that OLMCPR does not need to be increased due to RVE for FWHOS operation.
2.5 OPERATING LIMIT MCPR The analysis results show that (1) the FWCT vithout bypass (3CPR for rated feedwater temperature below 370 F was found to exceed the current licensing basis and (ii) the g5CPR's for all the other transients analyzed for the MELLIA and the TVHOS conditions, and in combination, are bounded by the current licensing basis values. Based on these results, it is concluded that the current technical specification OLMCPR's are adequate for operation in the normal as well as the MELLL operating region with or without the FWHOS except the following TWHOS operating conditions for the MT50S option;
- 1. For operation between rated feedwater temperature of 370 F and 320 F the required Option A and Option B OLMCPR's are 1.33 and 1.26 respectively.
- 2. For operation between rated feedwater temperature of 270 F and 320 T the required Option A and Option B OLMCPR's are 1.34 and 1.27
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respectively.
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2-3 /3 I l
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NEDC 31515 Table 2-1 .,
I
. INITIAL CONDITIONS FOIL MEL11 AND FVHOS MANSIENT ARALT815 (102P.100F) , (102P 75F)
Rated T_. Rated T_.
420*F 370'T 320'T 270*F 270'F Thermal Yewar (NWt) 3358 3358 3358 3358 3358 Steam Flow (M1b/hr) 14.478 13.505 12.716 11.796 14.478 Core Flow (N1b/hr) 100 100 100 100 75 Vessel Dose Pressure (psta) 1035 1015 1010 1000 1000 i CoreCoolantInlitf$thalpy 526.1 518.8 515.7 505.3 495.7 (Stu/lb) l e
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Table 2 2 -
i REY INFUT PARANETERS FOR TRANSIENT ANALT318 l
- 1. Turbine Sypass capacity, 5 N&R 26.0 1
- 2. Fuel lattice C (F8m8R)
! 3. MCFR Safety I.iait 1.06
- 4. Control Rod Drive Speed, Position Versus Time Figure 15.0 1 (FSAR)
Nuclear Characteristics Used in ODYN Analysis ROC 1 i
- 6. Safety / Relief Valve Capacity I N&R At 1121 psig .
90.85 Manufacturer *
- Target Rock i
Quantity Installed 15
- 7. Relief Manction Delay,' seconds 0.4
- 8. Relief Funetton Stroke Time, seconds 0.15
- 9. Set Points for Safety / Relief Valves Safety / Relief Manction, psig 1121, 1131, 1141
- 10. Migh Flux Trip.
Analysis set Point (121 a 1.02), E NAR 123.4
- 11. Righ Pressure Scram Set Point, psig 1101.0
- 12. Vessel 1Avel Trips. Feet Above Separator Skirt Sottoa
. Level 8 -(18), feet 5.917 14 vel 3 - (L3), feet 1.75 14 vel 2 - (L2), feet (-) 4.71
- 13. Migh Pressure Recirculation Pump Trip Pressure set Fotnt, psig 1135.0 Delay Time, seconds 0.3
- 14. Total Stomaline Volume, eu ft 4658.0
- 15. Rehaater Sypass Flow E NBR 10 4
- 16. Turbine Control Valve Full Stroke Closure Time, seconds 0.20 2-5 of
.' MEDC 31515 Table 2-3 St20WLY OF TRANSIINT PEAK val #ES RESULTS -
- Rated Max. Max. Max. Max.
Core W Temp. Neutron Surface Dome Vessel Power Flow Rodgetion Flux Meat F1ur Fressure Pressure Transient 11 RBP) M _ f F) ft W1R) (I Initial) fesfri_ _fenir)_
FWCF(FSAR) 102 100 0 209 110.8 1150 1187 n
FUCF 102 '100 50 197 111.6 1141 1177 F
FWCF 102 100 100 194 113.4 1133 1167
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FUCF 102 100 150 193 115.0 1116 1149 FWCF '102 75 150 152 107.f* 1120 1143 T/CTNBP(FSAR) 102 100 0 367 115.3 1179 1216 T/CTNBP 102 100 150 327 111.8 1165 1200 T/CTNBP 102 75 150 256 109.2 1166 1191 FWCF w/o SP 102 100 0 386 120.3 1182 1222 0%AR) **
FWCF w/o SP 102 100 50 366 121.1 1172 1212 FWCF w/o BP 102 100 100 364 123.0 1166 1203 I Is FUCF w/o EP 102 100 150 351 124.6 1160 1196
. FWCF w/o SP 102 75 150 306 117.2 1161
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- 1187 B
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NEDC.31515 Table 2-4 REQUIRED OPERATING LIMIT MCPR VALUES .
Rated Power Core Flow Tg Egduction M CPR Transient 111ER) (1 MRR) _
f F) Ontinn A Detion B-
- WCF(TSAR) 102 100 0 1.21 1.14 FWCF 102 100 50 1.22 1.15 FVCF 102 100 100 1.24 1.16 FWCF 102 100 150 1.25 1.18 FWCF 102 75 150 1.20 1.13 T/CTNBP(FSAR) 102 100 0 1.25" 1.15 T/CTNBP 102 100 150 <1.25 <1.15 T/GTNBP 102 75 150 <1.25 <1.15 b b W CT w/o BP(TSAR) 102 100 0 1.31 1.24 FWCT w/o BP 102 100 50 1.31 1.24 FWCF w/o BP 102 100 100 1.33 1.26 FWCF w/o BP 102 100 150 1.34 1.27 FWCF w/o BP 102 75 150 1.26 1.19 LNH (FSAR)* 102 100 0 1.20 LNH 100 100 0 1.15 IEWH 100 75 0 1.14 LWHF 100 75 70 1.13 ,
LNHF 100 75 150 1.13 RWE(FSAR) 102 100 0 1.24**D a current Ferai-2 technical specification 01MCPR.
l b Current Termi-2 technical specification 01RCPR for MTBOS option. ,
1 c Based on GE Transient Analysis Code REDY (Reference 6).
2-7 ,7 J _. _
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- NEDco31515 Table 2 5 OVIRPRIS$UILIEATION ANALYS!$ RESULTS
- MSIV C10$t1LE (FIDX SCRAM) 1021 POWER Feak Peak Steam Line Vessel Initial Core Flow RatgdTyg hessue hesme (I NRR) f F) Imain) (enfe) 100 420 1222 1248 75 270 1187 1223 t
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- 3. THERMAL NTDRAULIC STABILITY .
N General Electric Company has established stability erste$ta c.
demonstrate sempliance to requirements set forth in 10CFRSO dypendia A, General Design Criteria (CDC) 12. These stability sorgliance criteria con.
sider potential limit cycle response within the limits of safety systes and/or operator intervention and assure that for GE SUR fuel designs this operating mode does not result in specified seceptable fuel design limits being exceeded. Furthermore, the enset of power oscillations for which corrective actions are necessary is reliably and readily detected and suppressed by operator actions and/or metamatic system functions. b stability sempliance of all licensed CE SE fuel designs includir.g these fuels sentained in the teneral Electric Standard dyp11 cation for Raseter Phel (GRSTAR II, Esfer-ence 7) is demonstrated on a generic ' basis in Reference 8.(for operation in the normal as well as the extended operating desata with or without Fumos).
N RC has reviewed and approved ta' is in Reference 7 and therefore a specific analysis for each cycle is met required. m Fermi 2 Cycle 1 sore contains licensed CE SWR initial core fuel and hence, the generic evaluation in Reference 8 is applicable to Fermi-2.
For operation in the MEll.L operatimig demain the stability margin (defined by the core decay ratio) is reduced at high powers for a given eers flow. In general, reduced feedvater temperature aise results in reduced stability margin when operating in the high power / low core flow regten of the operating demain.
Nowever, the fuel integrity analyses in Reference S are independent l
, of the stability margin, since the reactor is already assumed to be in limit cycle oscillations (no stability margin). Esference 8 also demonstrates that even if neutron flux limit cycle oscillations did occur just below the neutron flux scram setpoint, fuel design 11atts are not exceeded for all iteensed CE SUR fuel designs including these fuels contained in SESTAR II (Reference 7).
h se evaluations demonstrate that substantial thermal / mechanical margin is available for the CE SVR fuel designs even in the unlikely event of very large ese111ations.
To provide assurance that acceptable plant performance is achieved during operation in the least stable region of the power / flow asp, as well as during 31
NEDC.31515 all plant maneuvering and operating states, a generic set of operating reces.
mondations has been developed as est forth in Reference 9 am8 esmunteated to ,
all CE SUR utilities. These reseemendattens, which are betag taylemented in the Forst 2 Technical Specifications, Amstruct the operater en how to reliably detect and suppress limit cycle neutron nun oscillations should they occur.
The recommendations were developed to conservatively bound the sapected performance of all eurrent product lines and are applicable to Forsi.2 opera.
tien in the MEL11 regten and with the FWHOS.
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, ItEDC - 31515 4 ECC$ PERFORMANCE .
ECCS performance impact was evaluated for Forst-2 plant operatien in the MEL11 region with or without FWHOS. The evaluation shows that these modes of operation result in improved Peak Cladding Temperature (PCT) in the event of a LOCA as compared to the FSAR basis geparate discussions for the two conditions follow, giving the basis for concluding that the calculated PCT is lower in sech case.
4.1 FWHOS It has been concluded from several previous studies that operating at reduced feedvater temperature reduces calculated PCT in the event of a 1DCA.
Reduction of feedwater temperature results in increased subcooling in the vessel thus increasing the total system mass and the mass flow rate of a LDCA break for a given vessel pressure. Novever, at the limiting conditions for the ECCS performance evaluation of 106.21 power /1001 now, the initial vessel pressure is lower due to reduced steam production. This results in a net decrease in the calculated break now rate at most times during the event using the Moody Slip now model required by 10CFE50 Appendix K. Also the ine'reased total system mass delays the time of lower planum nashing. The increased system mass and the decreased break now set together to result in later jet pump, break, and core uncovery times.
The delay in jet pump uncovery means a delay in time of dryout. Thus, there is more time to remove stored energy from the fuel. Time of dryout is .
the most algnificant parameter in terms of calculated PCT. The time delay between time of dryout and lower plenum nashing also increases s11 5 htly. This is stro somewhat beneficial because it allers the bundle to beat up more before nashing occurs and the higher temperature between fuel and coolant improves heat transfer leading to less stored energy in the fuel at the end of '
the' lower plenum flashing period. Core uncovery also occura later due to the increase in total systes mass. This also allows more energy to be renoved 4-1 ?/
i NEDC - 31515 g prior to the start of the final eere bestup phase. Datsybaatkilbe slightly fewer eartug the sore uncevery perised, roeuttig ta a slower core bestup.
As a result of these seebimed effects, b peak cladding temperature for operation with reduced feedvater temperatures is lower than een operating at rated feedvater temperatures. This conclusten applies for operatten in the normal as well as in the NEu L operating resten.
l 4.2 NE1MA 1
l hre is one major eencern in the MERL region with regard to SCCS l
performance and PCT r'esponse, h higher re.d line will permit a higher power (higher initial stored energy in the fuel) at a given flow. This Leeresses the probability of losing smelsate boiling at the highest power antal mode prior to the time of jet pump uncevery. This phenomenon is sailed early boiling transitten (87) and could affect the calculated FCT.
There are two parameters *1ch play a sujer role la determinig the i salculated PCT that are affected by the higher eere power at lower more new:
(1) the time of boiling transitten at the high power antal mode of the limiting fuel assambly and (2) the saleulated sore imeevery Asratien. garly l boiling transitten result's in a less efficient removal of the initial stored energy free the fuel, Wich tends to inersase the salculated PCT. Lever l- . initial core new tends to decrease the calculated PCT. This occurs because i l increased subescling in the downsener and lower plenue at lever core flers increases the initial system inventory. N increased systen Leventory leads to a shorter core uncovery 6sratten and lower calculated PCT.
l The variatten of the bundle inlet flew during a 1DCA event is determined by a munher of parameters, including the break size, the water 1syventory in the reacter at the start of the event, and the steady state sore power and flow conditgens, N first of these is accounted for in the standard SAR IDCA 1
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analysis. m effect of variations in the remaining two was assounted for by performing an analysis at the limiting power flow point aleeg the M112, boundary. The assumed initial MCFR is also an important parameter in determining whehr or not early boiling transition will occur at & high power axial mode. Credit was taken for the technical specification requirement on MCFR versus core flow (K g ) with an additional 25 conservatism 1
added to satisfy 10CF150 Appendix K.
A 1DCA analysis was performed for Fermi-2 at the 102E power and 751 core flow condition in the MELLL region. This condition was selected because it is l the lowest core flow at which Fermi-2 san operate with a rated McFR limit.
The results of this calculation show that 6arly BT does not occur. Thus, the calculated PCT is less than that of the FSAR sase. b refore, there is no impact on the MAP 1JtCR and the present limits remain applicable in & MELU, region.
4.3 CONCWSION It is concluded that for the Elli and the FUNOS eperating conditions, and in combination, chers is no impact on the MAPIJICR limits resulting from ECCS performance analysis.
I l
4-3 33 l 1
o NEDC-31515 I
- 5. CONTAIIDENT ANA1.YS15 .
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l '
he impact of operation in the MELLL region and the FUHOS operation on the containment IDCA response was evaluated.
s Operation in the MELLL region and operation with FUHOS both lead to higher IDCA blowdown flowrates, compared with those presented in the containment analysis section of the FSAR, for certain time periods following the double. ended guillotine break of a recirculation suction line. This is a result of the increased degree of subcooling of the initial blowdown flow using the Moody Monogeneous Equilibrium Model (HDt). This trend differs from that observed in the ECCS performance analysis because the MDt is such more i sensitive to changes in the subcooling than the Slip Flow model. De limiting operating condition for peak drywell pressure is the power / flow condition with
~
highest subcooling in the MELLL region under FWHOS. The results of the analysis performed under this operating condition show an increase in the peak drywell pressure over the value reported in the FSAR, but the value is still below the design limit of 62 psig. Other containment parameters of concern (peak suppression chamber pressure, peak drywell temperature, peak suppression pool camperature) under MELLL and/or FUHOS operation are bounded by the results reported in the FfAR.
An evaluation was also made of the impact of MELLL with FUHOS operation on the lhCA containment hydrodynamic loads. Pool swell, condensation
. oscillation and chugging loads under MELLL with FVHOS operation are bounded by the corresponding design loads. For the case of MELLL with FWHOS operation, an increase in the vent system thrust loads is calculated. This increase has l the potential to lead to vent systes stresses above the allowable stresses. An evaluation of this increase in vent system thrust loads is to be performed to determine acceptability.
5-1 n
, . . EDC-33515
- 6. FEEDWAT R 90ZZLE AND FEEDUATER SPARGER FATICUE USACE 5.1 METHOD AND AS$13tFTION$ , , ,
The fatigue experienced by the feedwater nossle and foodwater sparger results from two phenomena; system cycling and rapid cycling. System cycling is caused by major temperature changes associated with system transients. J These transients are identified on thermal cycle diagrams. Thermal stresses due to these transients are calculated by determining inner and outer metal surface temperatures using conventional heat transfer and stress ar.alysis methods. Fatigue usage is determined by dividing the number of design cycles for each transient by the number of allowable cycles for st.h stress cal- }
sulated. Cumulative system fatigue usage is determined by summing all of the respective transient fatigue usage fa'ctors.
Rapid cycling is caused by small, high frequency temperature fluctuations due to mixing of relatively solder mostle annulus water with the reactor coolant. The colder water impinging the nozzle bore originates from leakage past the thermal sleeve secondary seal and from the boundary layer of colder water formed by heat transfer through the thermal sleeve. The mixing region extends from the feedwater mozzle surface region to the feedwater sparger surface; therefore, rapid cycling applies to both of these compont.nts. Once thermal stress due to rapid cycling is determined, fatigue usage is calculated and the results are added to the cumulative system cycling usage factor to obtain the total usage factor.
Operation with FWHOS will cause a change in calculated rapid cycling fatigue only. This is because the system transient is very mild (small temperature change and relatively long duration) and is bounded by the origi-mal design basis thermal stress analysis. General Electric has developed l
standardized rapid cycling duty asps for each BUR plant that cover the design basis rapid cycles in the same manner that thermal cycle diagrams cover the design basis thermal transients (system cycling). The methodology used to 6-1 r
. WEDC - 31515 develop the duty maps is based on the results of estensive testipg of feed-water nossles by General Electric. MOS is analysed by modifying the design cycles in order to gauge its effect en fatigue usage. Redmoed feedwater temperature will tend to increase fatigue usage due to an inernase in thermal stress, while reduced leakage flev (due to a decrease in fee 6 water flow) will tend to decrease fatigue usage because of reduced leakage to the sparger and mostle surfaces.
An evaluation of the affect of M05 en the feedvater aozzle and feedvater sparger fatigue was performed for the following condition: '
A 150'T reduction in rated foodwater temperature was assumed for various lengths of time during an 18 month fuel pycle, se that a relationship could be determined for incremental fatigue damage as a function of time spent with mos.
6.2 FEEDUATElt N0ZZ12 FATICUE The analysis done for normal design duty indicated that refurbishment of the thermal sleeve seals after 32 years would be necessary to keep the 40 year total fatigue usage (systou cycling plus rapid cycling) below a value of 1.0.
For a refurbishment schedule of 32 years, the 40 year total facigue usage was calculated as shown in Table 6-1. The fatigue damage per cycle for M 05
. . operation is conservatively estimated by taking the difference between the M os fatigue and the normal operation fatigue and dividing that quantity by the number of cycles in 40 years. Incremental fatigue damage as a function of time spent with M 05 per cycle is graphically shown in Figure 6 1.
If M OS for the assumed times were used for every cycle, fatigue usage less than 1.0 could be achieved by reducing the refurbishment interval as moted in Table 6 1. Although the refurbishment interval is , impacted by 7 or 8 years assuming MOS for every cycle, only one refurbishment is required, as is the case of normal operation.
6-2
)G
l EEDC - 31515
. In eenclusten, the impleneatation of FUNOS at Fermi-2 will beve a rapid eyeling fatigue impact en the feedwater maaste. S e incremental fatigue usage due to FWHos for its per year is 0.258 per eycls. h is estimate is senserva.
tlw since the analysis assumes apper bound asal leakage ratas and design basis corresten rates for the materials involwd. ,
6.3 FEEDUATER spARCER FATICUE Wince the feedwater sparger is not an ASME Seiler & Pressure Vessel Code eenponent, a fatigue analysis was not originally performed. To gauge the effect of FUROS on the sparger, the fatigue usage la abo sparger for the .
original thermal duty eycles and rapid cycle duty map was calculated. The fatigue usage for systes cycling is 0.26 for ths 40 year life of the plant.
The calculated fatigue usage for rapid eyeling is 0.45, gbfeb makes a total 40 year usage facter of 0.71. This result shows that the usage facter will not exceed 1.0 for the 40. year life of the plant.
m results for M OS operation are summarised in Table 6-2 along with the above results for normal operation. N results aber that implementation of M OS at Ferai 2 will aise have a rapid cycling fatigue impact en the feedwater sparger. Incremental fatigue damage as a function of time spent with M OS per cycle is graphically shown in Figure 6-2. Note that these fatigue results are based en the assumed senservative leakage rates and ope' rating modes described by the design basis duty map. The actual impact of M OS operacion en sparger fatigue usage will depend upon actual leakage rates, actual time spent with M OS, and thermal sleeve seal refurbishment schedule.
6-3 1 *7
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EEDC - 31515 l
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Table 6 1 .
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20ZILE FATICUE USACE FOR A 32 YEAR SEAL REFURRISMNENT PERIOD j I
18 Month Cycle 18 Month Cycle FWHOS Operation FWHOS Operation for 657 Mours for 1577 Nours Normal Each Cycle Each Cycle Omaration f3.01 mar waar) (12.01 mar wear) 40 Year Total Fatigue Usage 0.9815 3.8517 7.8700 Additional Usage Due to FWHos --
2.8702 6.8885 i
Additional Usage per Cycle -- 0.1076 0.2583 Note: The total 40 year usage factor for FWHOS operation after every cycle can be kept to below 1.0 by refurbishing the seals after 25 years for )
the 657 hours0.0076 days <br />0.183 hours <br />0.00109 weeks <br />2.499885e-4 months <br /> / cycle case, and by refurbishing the seals after 24 years j for the 1577 hours0.0183 days <br />0.438 hours <br />0.00261 weeks <br />6.000485e-4 months <br />'/ cycle case.
l 6-4 St
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- WEDC - 31515 Table 6-3 -
SPARCER FATICUE USAGE FOR A 32 YEAR SEAL REFUERISWENT FSRIOD 18 Nonth Cyc1s 18 Nonth Cycle FUH05 Operation FUHOS Operation for 657 Neurs for 1577 Mours Normal Each cycle Each Cycle OnaIalin f5.02 mar waar) (12.01 mar waar) 40 Year Total Fatigue Usage 0.7125 1.8856 3.8919 Additional Usage * .
Due to FVHOS -- 1.4331 --
3.4395 Additional Usage per Cycle -- 0.0537 0.1290 Note: The total 40 year usage factor for FVHOS riparation after every cycle can be kept to below 1.0 by refurbishing'the seals per the schedule defineci in Table 61.
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- 7. ACOUSTIC AllD FIM IIIDUCED 1mDS IIEPACT, ,
he acoustic loads are lateral loads en the vessel internals that result from propagation of the decompression wave created by a sudden recirculation suction line break, h neoustic leading en the vessel intavaals is propor-tional to the total pressure wave amplitude in the vessel rneirculation outlet restle. he total pressure amplitude is the se of the initial pressure subcooling plus the experimentally determined pressure undershoot below saturation pressure. An increased downconer subcooling, as is seen in the ate 11L region with Mos, results in a lower saturatise pressure, thereby having a larger total pressure amplitude and resulting in larger acoustic loads.
m high velocity now patterns in the downeoner resulting free a recir-culation suction line break create lateral loads en the shroud and the jet pump. Rose loads are proportional to the square of the critical mass flux out of the break. D e additional subcooling in the downconer resulting from operating in the stELLL region with MOS leads to an increase in critical now and, therefore, in n ow induced loads, he reactor internals most impacted by acoustic and now induced loads are the shroud, shroud support, and jet pump. H e impact en these components was evaluated throughout the ate 11L region with M05. D e analyses concluded that these components have enough design margin to handle the increased load.
. ing during operation in the stEILL region with the MOS.
7-1 L2
- 8. REACTOR FROTECTION SYSTEM lau POW R SETPCINT At reactor power levels, where significant amounts of steam ar's being generated, the fast closure of turbine stop or control valves will result in rapid reactor vessel pressurization. When pressure increases, power increases, especially if the bypass valves fail to open. For this reason, scram occurs on turbine stop valve position and control valve fast closure to provide marsin to the core thermal-hydraulic safety limit.
The required lower bound for stop valve position and control valve fast closure scram is 30% of rated thermal power. Turbine first-stage pressure is the parameter used to initiate the turbine stop valve closure scram bypass functions. At normal feedwater heating operating conditions, this 301 power is equivalent to approximately turbine first stage pressure of 154 psig.
Selow 301 power, the turbine stop valve or control valve scram functions are disabled. At these low power levels, high neutron flux scram and vessel pressure scram and other normal seras functions are sufficient to provide the safety limit margin even with stop valve or control valve sudden closures.
During the WHOS condition, as feedvater temperature is reduced, steam fiev decreases. A given core power based on full feedwater heating (i.e.,
rated 420*F feedwater temperature) will not produce the same steam flow as the same core power based on reduced rated feedwater temperature. Thus, it is necessary to resdjust turbine stop and control valve scram bypass setpoints for WHOS operation to maintain the required 301 lower bound power for stop and control valve fast closure scram. However, a study has been performed which shows that the current serpoint at 420oF (normal feedwater heating at rated power) is applicable down to 370oF. The recommended scram bypass setpoints are <154 psig for 370'F < rated Tg < 420 % , 4 32 psig for 3204 (
rated T g < 370*F, and (123 psig for 270*F < rated Tg < 320*F.
81 C
EDC-31515
- 9. APRM INSTR 13tENTATION SETFOINT FOR MELLIA
( .
In order to allow operation in the MELLL operating demain, the current Average Power Range Monitor (APRM) Simulated Thermal Power Monitor (STFM) scram and rod block configuration and setpoints are modified to acceanodate this region. his consists of:
(1) Raising the current APRM rod block and STP scram line to higher power serpoints.
(2) Clipping the AFAM rod block at high core flow.
l De proposed APRM red block and STFM scram setpoints for the MELLIA are illustrated in Figure 9-1 and tabulated below.
Meninal Eatooint Allowable Yalue AFRM Rod Block .66W + 581 .66W + 611 1081 Max. 1101 Max.
AFRM STP .66W + 641 .66W + 671 113.51 Max. 115.51 Max.
S e new setpoints maintain the same slope (0.66), same clip serpoint (1n.51 max. for STFM scram and 1081 for rod block) at rated pvver condition and same margin (61) between the STFM scram and rod block setpoints as the current 2001 rod line Technical specification. D erefore, no loss in rod block warning or scram protection axists due to this serpoint change when operating in the MELLL domain.
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- 10. OPERATIONAL 1. IMITATION g
Plant operation in the entire MELLL operating domain is not feasible because of RAM control red withdrawal restriction caused by the present RBM astpoint at Fermi 2 (0.66 V, + 401). Actual operating domain is reduced to somewhat narrower extended operating envelope, i.e., approximately to the region which is illustrated in Figure 10-1. This region is bounded by the 1081 AFAM rod block line, (0.58W, + 501)*, rated power line, and the rated core flow line.
The analysis results and impact evaluation for the MEU.L and PVHOS operation presented in this report are conservatively applicable to operation of Termi-2 plant in this marrower extended operating domain with the PVHOS.
Nowever, the concern on the increased vent system thrust loads stated in Section 5 still exists for operation in this region with FVHOS.
~
l
- AFRM Rod Block < 0.58 VD + 501, where VD I" ***II""I'*I'" ##I II'" I" percent of rated. This less restrictive equation was approved by the .
USNRC in Reference 10.
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~ ~ . . .
EDC - 31515 2 . .
900 MS =
w, secucutmoef 440P SIWE 760w 989 =
Oletets ftM tes)
- 88878 900 = * (800/100)
.. )
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/
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I ., ,, moe me= =.epper e S.Se g e(Otl 7 . ..
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f
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, , $WhfP SfMD carus 4L y
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%I8 SSE Pb8 win Figura 10-1. Marroser Exterded Operattig Region 10-2 n
,- ,_ , , NEDC - 31515
- 11. REFERENCES I'. 'FERNI 2 DFSAR*, March 1987 (Docket No. 50 341) 2.
Eatter, F.E. Agosti to DSNRC, 'Resueed Foodwater Beating *,'NRC 87 0073, May 19, 1987.
- 3. ' FERMI 2 Technical Specifications", appendix 'A' to License No. NFF 43, July 1985 (Docket Bo. 50-351).
4.
'Qualificaties of the One-Dimensional core Transient Model for Soiling Water Reactors". Oct. 1978 (NEDO 24154).
.. I S. 'Three Dimensional SUR Core Simulator,' January 1977 (NEDO 20953 A).
- 6. 1.B. Linford, ' Analytical Methods of Plant Transient Revaluations for the !
General Electric Boil'ing Vater Reactor". NED0-10802, General Electric.
April 1973.
- 7. ' General Electric Standard Application for Rosetor bl (Supplement for United States),' May 1986 (NEDE 24011-F A 8 DS, as amended).
- 8. ' Compliance of the General Electric Boiling Water Reactor N1 Designs to Stability Licensing criteria,' October 1984 (NEDE 22277 F 1).
- 9. '3UR Core Thermal Nydraulic Stability ' SIL No. 380 Revision 1 February 10, 1984.
- 10. safety Evaluation by the Office of Duclear Raaetor Regulation Supportins i Amendment No.59 to Provisional Operating License No. DPR 19. Amendment No. 52 to Facility Operating License No. DPR 25, Amendment No. 70 to l
Facility Operating License No. DPR 20, and Amendment so. 64 to Facility Operating License Bo. DPR 30, consonwealth Edison Company and '
Iowa Illinois Gas and Electric Company Dres6en Station Unit pos. 2 and
- 3. Quad cities station Unit Nos. I and 2. Docket pos. $0 237, 50 249, 50 254, and 50 265.
21-1
m.__ . _ _
- y. - . . _ _ _ _ _ _ _
i e 1 =
w Enclosure 2 Letter from C. E. Johnson (GE Nuclear Energy) to S. E.
Kremer/ Attn. M. K. Deora (Detroit Edison),
Subject:
Fermi-2 Vent System Thrust Loads Analysis -- Maximum RPV Downcomer '
Region Subcooling, TDEC-5829, May 5, 1988.
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l REC'D MAY 91988 l MC 391 *(408)925-2588 i
-May 5, 1988 ,
TDEC-5829 Mr. 5. E. Kremer. Supervisor Nuclear Procurement i j
Enrico Femi-2,108 EF2 NOC The Detroit Edison Company 6400 N. Dixie Highway Newport. MI 48166 Attention: M. K. Deora (280 EF2 TAC) i
SUBJECT:
Femi-2 Vent System Thrust Loads Analysis'-- ;
Maximum RPV Downcomer Region Subcooling :
Gentlemen:
At the present time Detroit Edison, at Femi-2 is controlling operations so as to remain under 45 Btu /lb subcooling in the RPV.
Some time ago you requested that we increase the subcooling value to one that would allow a clean MELLLA operating domain and allow an incremental feedwater temperature reduction from rated power and temperature of 150'F.
The subcooling value consistent with the above guidelines is 60 Btu /lb. The attachedengineeringreport(C.R.ParkertoC.M.Johnsondated5/3/88.same subject) shows the resulting containment vent system thrust loads that would result from a LOCA occurring while Fermi-2 was operating under the above asseptions.
Since these loads exceed the current vent header capability as reported in Rev. 3 of your Plant Unique Load Definition Report dated April 1983. Detroit Edison should re-evaluate Femi-2 vent header load carrying capability to see if the current design can, in fact, accommodate these higher loads. If not.
then further iteration needs to be done to select a maximum value of subcooling which will match the maximum vent header capability.
The goal, of course, is to increase plant operating flexibility by utilizing as much of the MELLLA map as possible. At the present 45 Stu/lb controlling limit, you would be able to use less than one-third of the possible domain. ,
/
4 9 . ,
b
. 5..E. Kremer May 5. 1988 TDEC-5829 - Page 2 Please keep us informed of the results of your on-going investigation in this matter.
Kindly let us know if you have any questions relative to these results.
Very truly yours.
C. M. nson Proje Manager Enri o Fermi-2 Project Attach. - Five Page Report CMJ:mrk
, cc: S. G. Catola. Deco
8 (CMJLTR-440)
.A
_ _ _ - - _ _ _ _ _ _ - _ . _ _ _ - . _ _ _ 1
0 NUCLEAR FUEL & ENGINEERING SERVICES DEPARTMENT May 3; 1988 cc: J.E. Torbeck DRF T23 00648 To: C.M. Johnson ;
i From: C.R. Parker
Subject:
Fermi-2 Vent System Thrust Loads Analysis - Maximum RPV Downtomer i
Region Subcociing
References:
- 1. NED0 24568, ' Mark I Containment Program Plant Unique Load Definition - Enrico Femi Atomic Power Plant: Unit 2",
rev. 3. Apr.1983.
- 2. NED0-21888, ' Mark I Containment Program Load Definition Report". rev. 2, Nov. 1981.
- 3. DET-04-028-3, 'Enrico Fermi Atoute Power Plant Unit 2 Plant Unique Analysis Report *, Vol. 3, rev. O, April 1982.
This letter transults the results of the vent system thrust loads analysis performed for Femi 2 for the reactor operating condition that has the maximum
~
RPV downconer region subcooling. This operating condition, considering Maximus Extended Lead Line Limit (MELLL) and Foodwater Heater Out of service (FWHOS) operation, is 705 thermal power / 425 core flow with rated feedwater temperature reduction of 150 'F. The RPV downconer region subcooling for this ,
.?
page 2 e coadition is 60 Stu/1ba (based on the RPV done pressure). The vent system thrust loads corresponding to this operating condition represent an increase of 14-325 over the thrust loads presented in the PULD (ref.1). W increase in the peak drywell pressure for this operating condition over the value presented in the pWLD is 4.3 psi. -
The analysis was performed using the engineering computer program DDCPT05 to
~
calculate the containment pressurization and the vent system flowrates. The blowdown flowrates were determined using the engineering computer program LAM 8 08. The usual procedure is to use MCPT for modeling of the reactor vessel, but LAM 8 provides a more detailed model of the vessel than that provided by MCPT. It was decided to use LAM 8 for vessel modeling, instead of MCPT, because it provides more realistic blowdown flowrates. The blowdown flowrates calculated by LAR were then used as input to MCPT. It should be noted that the Moody Homogeneous Equilibrium Model (NEM) was the critical flow model used to calculate the blowdown flowrates and that critical mass fluxes specified by MEM were adjusted for some non-conservatism that occur in the model at stagnation qualities near zero and for orifice-type flows. The vent system flowrates and containment pressures from BCPT were then used to calculate the vent system thrust loads. The vent system thrust loads were calculated using the procedure given in the Mark I Containment Lead Definition Report (ref. 2). This is the procedure used to calculate the thrust loads in the pWLD.
Considering MELLL and FWHOS operation, the reactor operating conditien analyzed (705 power / 425 core flow with rated feedwater temperature reduction of 150 'F) is the limiting operating condition for vent system thrust loads.
Performing the analysis at this condition results in increases in the vent thrust leads and the peak drywell pressure ever those presented in the PULO.
The maximum vent system thrust leads, for both this analysis and the ppLD, are presented in Table 1. W maximum thrust leads for the current analysis occur at 4.398 sec. post LDCA. h, definition of tLue forces (location and direction) is contained in the fULD (ref. 2). Comparing the went thrust loads from the current analysis with those Presented in the PULD, the fecrease in the maximum thrust loads is 24 325. The containment system stress analysis in the plant Unique Analysis Report (puAR) for Fermi-2 (ref. 3), to which the v
- m a,_ .
May 3. 1988
,- page 3 n
. 'vetit system thrust leads are an input, looks at the leading history in I .- intervals. The maximum value of each lead cesponent (e.g.. vent system thrust, seismic, condensation estillation, etc.) during a particular interval is determined and then the maximum values of all the leads for that interval are copbined. The intervals of interest for this vent system thhist loads analys'is are: 0.-l.50 sec.,1.50 5.00 sec., and 5.00-35.0 sec. following a LOCA. Table I presents the maximum vent thrust leads for the current analysit for each of these three time intervals. The maximum values for these time intervals occur at 1.50 sec., 4.398 sec., and 5.00 sec., respectively. The calculated peak drywell pressure is 52.3 psig and occurs at 4.398 sec.
j l post-LOCA. This compares with the value of 48.0 psig for peak drywell l pressure in the pULD (and also in the puAR). The loads presented in Table 2 are to be used in the containment system stress analysis and are to be l combined with the other load components to obtain the total load. The higher drywell pressures must also be considered in the containment system stress analysis.
Detsils of the analysis and evidence of verification may be found in DRF T23 00648.
C.* .1'sh~ '
Ada "
C.R. parker Verified by: 5. Mintz plant Anal. Services plant Anal. Services WC769, Ext.5-1791 WC769. Ext.5-2025
8
- Table 1 Fermi-2 ..
Maximum Vent System Thrust Loads ,
PULD 705 P/ 425 F
{ggg fkins) ikins)
FIV1 -54 -87 F1H1 -115 -143 F2V 44 57 F2H 18 23 F3V 0.95 1.2 l.
I F3H -3.4 -4.5, l 1
. I F1VT -432 -532 )
i F2VT 704 912 F3YT 76 96 C
- - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ i
r.
- ~
. T$1e t Forst 2 ..
- Went System Thrust Loads (70%P/425F)
TIME 0.-1.50 see, 1.50-5.00 sec. 5.00-15.0 sec.
{prgg
-45 -67 * -64 FIV)
-97 -143 -139 F3H1 57 55 F2V 38 13 12 F2H 15 1.2 1.1
- - F3V 0.80
-3.0 -4.5 -4.3 F3M 532 -516 F1YT -360 912 876 FtyT 608 .
96 92 F3YT 64 Glote: All forces are in kips.
7 L-_-__--______-- _ _ _ _ -
..f' <=
e 1
i Enclosure 3 Letter from R. J. Howard (GE Nuclear Energy) to S. E.
Kremer/ Attn: D. R. Hoskins (Detroit Edison) ,
Subject:
ELLLA Safety Evaluation, TDEC-5927, November 1, 1988.
l l
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- s. a , t GE Nuclear E:ergy H n in : :: w.
~::.- e i.r .e k .- , :: w ;
NC391,(408)925-3506 November 1. 1988 TDEC-5927 Mr. S. E. Kremer, Supervisor. Nuclear Procurement Enrico Fermi-2,108 EF2 NOC The Detroit Edison Company 6400 N. Dixie Highway Newport. MI 48166 Attention: Mr. Dale Hoskins (210 EF2 AIB)
SUBJECT:
ELLLA Safety Evaluation
REFERENCE:
NF-88-0227 Gentlemen:
This will serve to confim that Fermi-2 operation in the ELLLA operating domain with a maximum feedwater temperature reduction of 50*F t;ill not result in a peak drywell pressure under the limiting condition that will exceed the UFSAR reported value of 56.5.
Very truly youty.
I E, J6 rd A i ng Project Manager En co Fermi-2 Project RJH:mrk cc: S. G. Catola. Deco
_ _ _ - _ _ _ _ _ _ - _ - - _ _ _ _ _