ML20237B280
| ML20237B280 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/31/1998 |
| From: | Uyeda G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20237B272 | List: |
| References | |
| EMF-98-036, EMF-98-036-R01, EMF-98-36, EMF-98-36-R1, NUDOCS 9808180157 | |
| Download: ML20237B280 (56) | |
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! SIEMENS EMF-98-O'36 Revision 1 i
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1 j Post-Scram Main Steam Line Break Analysis l for Millstone Unit 2 !
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July 1998 EMil ,
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n Siemens Power Corporation p Nuclear Division 9808100157 980812 DR ADOCK 050003 6 l
ISSUEDIN SPC ND ON UNE Siemens Power Corporation - Nuclear Division DE /o j gi j - --
EMF-98-b36 Revision 1 issue Date:
Post Scram Main Steam Line Break Analysis for Millstone Unit 2 Prepared: ygm Sky gg' RS nalysis 4
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Customer Disclaimer important Notice Regarding Contents and Use of This Document i l
Messe Read CarefuMy Siemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the agreement between Simens Power Corporation and the Customer pursuant to which this document is issued.
Accordingly, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on its behalf:
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< until so authorized in writing by Siemens Power Corporation or until af ter six (6) months following termination or expiration of
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I expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document. f j
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i EMF-98-036 Post-Scrabi M:In Sterm Linn Brack Antlysis for5Millstona Unit 2 Rsvision 1 pag, ; 4 Nature of Changes .
4 Paragraph I Ittm or Page(s) Description and Justification )
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- 1. 6-1 Reference 2 was updated to show that this report adheres to Siemens Power Corporation's most recent Main Steam Line Break f
methodology.
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Po:} Scrim M:in Stsam Line Brrk An: lysis EMF-98-036
. _ftr Millstona Unit 2 Rsvision 1 Page li Contents -
1.
I n t ro d u c ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2.
Summary......................................................................................................2-1 3.
Met h od And Mo del Description ... . . . . ....... ... .. . . . . . .. . .. . . .. . . . . .. .... . . . . . . . . . ...
3.1 Meth od Description .... ....... ... . ...... .. . .. . ... . ... . . . . . . . .
3.2 M o d e i D e scriptio n . . . . . . . . . . . . . . . . . . . ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.2.1 3.2.2 A NF- R E LA P M ode l .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.2.3 X TG PWR Mod e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......
X COBRA-Ill C Mode t . . . .. . ... . . . . . . . .. . .. . ... ... ... ... ... 3. -4. . . . . . . . . . .
4.
An 4.1 a l y s i s R e s u lt s . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.2 E ve n t D e s criptio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .
S c e n a rio s Analy zed . . . . . . . . . . . . . ... ... . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .
4.3 A N F-R E L A P Re su lt s. . . . . . . . . . . . . . . . . ........~.......4-3 4.3.1 . . . . .. . . ... .. . . . . . . .
Li mitin g D N B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.3.2 Li miti n g F C M . . . . . . . . . . . . . . . . . . . . . . . ........ ...............................
4.4 4.5 X TG PW R Re su lt s . . . . . . . . . . . . . . . . . . . ....................4-7 .....................
XCO BRA-IllC a n d Fuel Failure Results ............... . ........... .. . . . .. .... .
5.
Conclusions...................................................................................................
6.
References...................................................................................................
Tables 3.1 Millstone Unit 2 Post-Scram MSLB System and Neutronics Parameters............... 3-6 3.2 Millstone Unit 2 HPSI Delivery Curve................... . ......... ... ...... .. . .. .. ....... ........... 3-7 3.3 Actuation Signals and Delays for Reactor Scram, MSIV, SIS and Feedwater SatetyActions............................................................
4.1 Post-Scram Main Steam Line Break' Analysis Summary ..................................... 4-8 4.2 Sequence of Events - HFP Loss of Offsite Power - Outside vs. Inside C o n t a i n m en t B re a k . . . . . . . . . . . . . . . . . . . . . . . . . ...............4-9 ......................
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4.3 Sequence of Events - HFP Offsite Pow !
Containment Break............................er Available - Outside vs. Inside
.............................................................*4-9 4.4 ,
Sequence of Events - HZP Loss of Offsite Containment Break................................. Power - Outside vs. Inside
.....................................................4-10 4.5 Sequence of Events - HZP Offsite Power Ava
- Outside vs. Inside Containment Brea k......................................ilable ................................................4-10 4.6 Comparison of ANF-RELAP and XTGPWR Reactivity ............................ .
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i 4.1 t HFP Loss of Offsite Power (Steam Generators' Secondary Pressures)......... ...... 4-12 4.2 HFP Loss of Off site Power (Break Flows) ................................. .. 4-13 4.3 HFP Loss of Offsite Power (MFW and AFW Flows) ...................................., . . ... 4-14 4.4 HFP Loss of Of fsite Power (Pressurizer Pressure) ............... ..
4.5 HFP Loss of Of f site Power (Pressurizer Level) ..................... .
4.6 HFP Loss of Of fsit.s Power (Total HPSI Flow) ............................. .... 4-17 4.7-HFP Loss of Of fsite Power (Reactor Power) .......................... .
4.8 HFP Loss of Offsite Power (Power Fractions by Region and Core Average)... .
- 4.9 HFP Loss of Of fsite Power (SG Secondary Mass) ...........................................
4.10 HFP Loss of Offsite Power (Core Inlet Temperatures)......................
. . . . . . . . . . . . . . . . 4- 2 1 4.11 HFP Loss of Of fsite Power (Reactivity Components) ..........................
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. . . . . . . . . . . . . 4- 2 2 4.12 HFP Loss of Offsite Power (Core inlet Flow Rates)................................ ..........423 4.13 HFP Loss of Of fsite Power (Core Exit Pressure) ............................. .... 4-24 4.14 HFP Offsite Power Available (Steam Generators' Secondary Pressures) .....
, ....... 4 25 4.15 HFP Offsite Power Available (Break Flows) .................................... ..... 4-26 4.16 HFP Of fsite Power Available (MFW and AFW Flows) ... ... 4-2 .......................
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l 4.17 HFP Offsite Power Available (Pressurizer Pressure) .......................... -
4.18 HFP Offsite Power Available (PressurizerLevel)..............................................
4-29 4.19 HFP Offsite Power Available (Total HPSI Flow)............................................... 4-30 1
4.20 HFP Offsite Power Available (Reactor 1 Power)................................................. 4-31 l 4.21 HFP Offsite Power Available (Power Fractions by Region and Core Average)..... 4-32 1
i 4.22 HFP Offsite Power Available { Steam Generators' Secondary Mass) .................. 4-33 4.23 HFP Offsite Power Available (Core inlet Temperatures) ................................... 4-34 I 4.24 HFP Offsite Power Available (Reactivity Components)..................................... 4-35 i
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4.25 HFP Offsite Power Available (Core inlet Flow Rates) ............................
4.26 HFP Offsite Power Available (Core Exit Pressure)........................................
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Siemens Power Corporation - Nuclear Division
Post-Scrcm Miln Stum Lins Brrak An: lysis EMF-98-036
' for Millstons Unit 2 R: vision 1 Pagev Nomenclature ,
AFW Auxiliary Feedwater CHF r itical Heat Flux DNB Departure from Nucleats Boiling DNBR Departure from Nucleate Boiling Ratic ECCS Emergency Core Cooling System EOC End-of-Cycle FCM Fuel Centerline Melt HFP Hot Full Power HPSI .
High Pressure Safety injection HZP Hot Zero Power LHGR Linear Heat Generation Rate LPSI Low P:sssure Safety injection MDNBR i Minimum Departure from Nucleate Boiling Ratio MFW Main Feedwater MSlS Main Steam isolation Signal MSIV Main Steam isolation Valve MSLB Main Steam Line Break MTC Moderator Temperature Coefficient PWR Pressurized Water Reactor RCP Reactor Coolant Pump RCS Reactor Coolant System RWST Refueling Water Storage Tank SG Steam Generator SlAS Safety injection Actuation Signal SIS Safety injection System ,
SPC Slemens Power Corporation '
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' Pss4Scrzm Mrin Stsim Lins Brsak Antlysis EMF-98-036 for Millstona Unit 2 RIvision 1 Page 1-1 I
- 1. Introduction i
_ i This report documents an analysis of the MSLB event for the Millstone Unit 2 Nuclear Power Plant. This analysis addresses the asymmetric thermal-hydraulic and neutronic characteristics resulting from a MSLB.
' This analysis replaces the Cycle 10 MSLB analysis reported in Reference . For the 1
Reference 1 analysis, the limiting case was identified as the HZP with loss of offsite scenario and subsequent turbine generator assisted primary coolant pump coastdown. The Reference 1 analysis assumed that the break in the main steam line occurred inside containment. - The purpose for the re-analysis of the MSLB event is listed below:
a to use increased trip setpoint uncertainties for the harsh containment conditions associated with breaks inside containment. -'
to upgrade the ANF-RELAP and the XCOBRA-IllC models to be consistent with Cycle 13 plant configuration (integral SG flow restrictors, etc.) and operating conditions.
Section 2 of this report presents a summary of the MSLB analysis. Section 3 contains descriptions of the analytical models. The analysis results are presented in Section 4.
Conclusions and references are presented in Section 5 and 6, respectively, I
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- 2. Summary .
Fuel responses were evaluated against the DNB criterion using the modified Barnett CHF correlation. All cases produced MDNBR values well above the 1.158 limit (a correlation limit of 1.135 plus a 2% mixed core penalty). The HFP scenario with loss of offsite power was determined to be the most limiting from a MDNBR standpoint, with a MDNBR of 1.71.
With offsite power maintained for operation of the primary coolant pumps, both the HFP and the HZP scenarios resulted in a return to higher power levels than the scenarios where offsite power is lost. However, these scenarios provide substantially greater margin to the MDNBR limit because of the higher coolant flow rates. None of the scenarios evaluated predicted fuel failure to occur as a result of violating DNBR limits.
Fuel responses were also evaluated against the FCM criterion, in the HFP and HZP scenarios with offsite power maintained for operation of the primary coolant pump $, the tractor returned to higher power levels than scenarios where offsite power is lost. The HFP scenario with offsite power available had the highest LHGR value, 24.3 kW/ft, it is essumed for the purposes of this analysis that fuel fails above a conservative FCM volume of 21 kW/ft. One full assembly,0.5% of the core, is predicted to fai! due te FCM.
The analysis in this report supports the operation of Millstone Unit 2 for Cycle 13 and for til future cycles bounded by the analysis conditions. '
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, Pcst,-Scr:m M:In Stsrm Line Br :k An: lysis EMF-98-036 4
, fcr Millstrns Unit 2 Rsvisiin 1 Page 3-1 i
- 3. Method And Model Description A brief description of the MSLB analysis methodology is presented in Section 3.1. The model descriptions are presented in Section 3.2.
3.1 Method Description i The SPC methodology for the analyzing MSLB events (Reference 2) consists of three computer codes:
ANF-RELAP, XTGPWR, and XCOBRA-IllC. The ANF-RELAP code is used to calculate the general system themial-hydraulic responses during a MSLB. The XTGPWR code is used to calculate the axial and the radial power distributions and the reactivity at the time of peak post-scram power. The XCOBRA-IllC code is used to calculate detailed core flow and enthalpy distributions.
Fuel failures are based on DNB and FCM criteria. The MDNBR is determined using $he modified Barnett CHF correlation. The FCM criterion is compared to the maximum post-scram LHGR.
4 The use of the above three computer codes provides a conservative method for calcul the system and core responses during a MSLB.
3.2 ModelDescription A description of the ANF-RELAP modelis presented in Section 3.2.1. Some important assumptions and conservatism used in developi.r. the model are also discussed. Section 3.2.2 contains a description of the XTGPWR model, tcllowed by a description of the XCOBRA-IllC model in Section 3.2.3.
3.2.1 ANF-RELAP Model Millstone Unit 2 is a Combustion Engineering designed two-by-four loop PWR with two hot 1:gs, four cold legs with one RCP in each cold leg, and two inverted U-tube steam g;nerators. In the ANF-RELAP model, the cold legs in each loop are lumped together.
3.2.1.1 Nodalization
- The reactor coolant system of the plant is nodalized into control volumes interconnected by flowpaths or " junctions". The primary system contains the reactor vessel, the Siemens Power Corporation - Nuclear Division C
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Post-Scram Main Steam Line Break Analysis EMF-98-036 fo,r Millstone Unit 2 Revision 1 Page 3 2 pressurizer, the hot leg piping, the tube side of the steam generators, the primary coolan i
pumps, and the cold leg piping. The HPSI system is attached to the . Thecold leg piping secondary system contains the shell side of the SGs, the MFW, the AFW, the main ste piping, and the MSIVs.
The ECCS consists of two trains, each containing a HPSI . Theand singlea LPSI pump failure criterion is satisfied by assuming the loss of a diesel .generator The for the anal primary effect of this failure is the loss of one HPSI pump .and The one LPSI pump remaining HPSI pump is modeled with flow rates given as a function of the primary pressure. The LPSI system is not modeled because the RCS pressure remains above the LPSI pump shutoff head. The plant configuration and operating conditions for are listed in Table 3.1 tlirough Table 3.3.
3.2.1.2 Break Location, Size, and Flow Model A double ended guillotine break is assumed to occur in the Loop 2 steam line doivn of the integral SG flow restrictor and upstream of the MSIV. This break location wil in the largest cross sectional flow area and will produce the most and the rapid cooldown highest return to power.
The flow is choked at the integrar SG flow restrictor, which has .
an area of 3 51 ft 2 Only i steam is allowed to flow out of the break. The break flows are calculat Moody critical flow model. On the SG side of the break, steam flows out of the break throughout the entire transient. On the MSIV side of the break ,
break flow termi nates after the MSIVs are fully closed. The MSIVs are fully closed 7 seconds after rec isolation signal. As an added conservatism, the main steam check in this analysis. e valves are not 3.2.1.3 Boron injection .
Boron injection into the primary system acts to mitigate the. Injection return to of power boron is modeled from the HPSI system. The HPGl system is conservatively modeled take suction from the RWST at 35 F with a boron concentration of 1720 ppm
. Initially, the line volume between the check valves isolating the system pumps eg and the cold l injection location,183 ft*, is assumed to be filled with unborated water 'The time r to flush this unborated water from the safety injection lines is included in the ANF REL -
model. The characteristics of the HPSI system are shown in Table 3.2.
Siemens Power Corporation - Nuclear Division
Post-Scr:m Miin Sts m Lins Brnk An: lysis EMF-98-036
- for tdifistons Unit 2 R:vislan 1 Page 3-3 3.2.1.4 Single Failure Assumption t
i The single failure assumed in this analysis is the Ioas of one diesel generator, which results in the disabling of one of the two HPSI pumps required to be in service during normal operation. In addition to the single failure, there is no credit taken for the charging pump system. This assumption results in an additional delay in the time required for boron to rcach the core. The delay is amplified when combined with the assumption of a stagnant upper head which serves to maintain the primary system pressure due to flashing of the hot fluid in the upper head.
- 3.2.1.5 Trips and Delays Trips for the HPS1 system, the MFW valves, and the MSIVs are given in Table 3.3. Biases, j
to account for uncertainties, are included in the trip setpoints. For the steam and feedwater valves, the delay times given are between the time the trip setpoint is reached end the time full valve closure is reached. For the HPSI system, the delay time give'n is from the time the setpoint is reached until the pumps have accelerated to rated speed.
The delay time required to sweep the unborated water from the injection lines is accounted for by setting the initial boron concentration of the injected flow to zero until the injection line volume has been cleared.
3.2.1.6 Neutronics The reactor kinetics in ANF-RELAP are calculated using a point kinetics model. The reactivity feedback effects due to changes in boron concentration, changes in moderator f
dtnsity, and Doppler effects are calculated. The EOC values for the required control rod shutdown worth, the reactivity feedback curves, delayed neutron fraction, delayed neutron fraction distribution and related time constants, and prompt neutron generation time are used in this analysis. The ANF-RELAP default fission product and actinide decay constants * '
cre also utilized. The technical specification EOC MTC limit is used. Other neutronics '
pirameters, listed in Table 3.1, are based on nominal Cycle 13 EOC values.
The core is assumed to have the most reactive control rod stuck out of the core so that the power distribution is highly peaked in the stuck rod region.
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- Post Scrrm Miin Ststm Line Break Antlysis EMF-98-036 for'Millstona Unit 2 Rsvision 1 Page 3-4 3.2.1.7 Feedwater For the HFP scenarios, normal MFW flow is assumed to be delivered to both SGs. The MFW flow increases as the secondary pressure decreases at the lowest possible fluid temperature until the feedwater regulator valve closes: Fluid temperature is determined assuming heating of the feedwater ceases at the same time the break is initiated. The MFW flow will be terminated 14 seconds after receiving the isolation signal.
For the HFP scenarios, the AFW flow is assumed to be zero at break initiation. After 180 seconds, AFW is delivered at the maximum capacity of the AFW system with flow restrictors installed on the AFW delivery lines. For the HZP scenarios, the AFW flow is increased to the maximum capacity immediately at break initiation. For all scenarios all of ,
the AFW flow is directed to the affected SG to maximize the cooldown rate. The operator is assumed to terminate the AFW flow at 600 seconds.
3.2.1.8 Decay Heat ,
For the HFP scenarios, the initial decay heat is calculated on the basis of infinite irradiation time at a power of 2700 MW prior to transient initiation. For the HZP scenarios, the initial decay heat is calculated on the basis of infinite irradiation time at a power of 1 W prior to transient initiation. For both scenarios, decay heat generated from return to power is calculated.
3.2.2 XTGPWR Model XTGPWR is a three-dimensional, two-group diffusion theory computer code. XTGPWR provides three contributions in the MSLB analysis, all of which are closely related to the ANF-RELAP and XCOBRA-Ilt calculations. One, XTGPWR is used to generate the
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neutronics input paramete.s for ANF-RELAP as summarized in Table 3.1. Two, XTGPWR
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is used to verify the rearaivity calculated by ANF-RELAP is conservative. Three, XTGPWR is used to generate the radial and axial power distributions for use in the XCOBRA-IllC calculations.
3.2.3 XCOBRA-lllC Model Based on the overall core conditions calculated by ANF-RELAP at the time of peak power, the XCOBRA-lllC fuel assembly thermal hydraulic code is used to calculate th'e flow and the enthalpy distributions for the entire core. The limiting assembly DNBR calculations I l
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were performed using the modified Barnett correlation, which is an assembly-based correlation,
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, Post crcm Mzin Stzim Lins Brstk Anitysis EMF-98-036
~ . , for Millstone Unit 2 RIvision 1 Page 3-6 Table 3.1 Millstone Unit 2 Post Scram MSLB System -
and Neutronics Parameters -
System Parameter HFP Value' HZP Value' Core Power (MW) 2700 1 E-6 Primary Pressure (psia) 2250 2250 Pressurizer Level (%) 65 40 Cold Leg Temperature (*F) 549 532 Primary Flow Rate Per Loop (Ibm /sec) 18,820 19,241 Secondary Pressure (psia) 880 892 Total SG Mass (ibm) per SG 167,237 253,989 .
Total Steam Flow (Ibm /sec) per SG 1634 4 Tc.tal MFW Flow (Ibm /sec) per SG 1634 4 MFW Temperature ('F) 432 432 Total AFW Flow (Ibm /sec) 184 184 AFW Temperature (*F) {
32 32 I Neutronics Parameter Value Delayed Neutron Fraction, p 0.0054 MTC (pcm/ F) 1
-28 l HFP Scram Worth (pcm) 6438 Shutdown Mart,in Requirement (pcm) 3600 Initial Boron Concentration of HPSI Line (ppm) O !
RWST Boron Concentration (ppm) 1720 These values are from the applicable steady state ANF-RELAP calculation.
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l Table 3.2 Millstone Unit 2 HPSI Delivery Curve HPSI Flow Rate Versus RCS Pressure
- Pressure (psial Total HPSI Flow (ibm /sec) 14.7 77.03 50 75.88 100 74.23 150 72.54 200 70.68 300 66.67 500 58.39 700 48.50 900 36.78 1000 27.65 -
1020 25.25 1050 21.16 1075 17.25 1120.6 0.00 1121 0.00 l
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P:st-Scr:m M:in Stsam Line Breck AnIlysis EMF-98-036
', for Ahlistons Unit 2 RIvisi:n 1 Page 3-8 Table 3.3 Actuation Signals and Delays for Reactor Scram,
1 Actuation Signal
_Inside Containment Outside Containment Setpoint Setpoint Low SG Pressure Trip 550 psia 658 psia Low SG Pressure MSIS 370 psia 478 psia SIAS (Low Pressurizer Pressure) 1500 psia 1578 psia l,
MSIV Closure Required Actuation Signal Law SG Pressure MSIS {
Delay - 6.9 seconds -
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HPSI Actuation Required Actuation Signal SIAS Delay - 25.0 seconds Main Feedwater Valve Closure Required Actuation Signal Low SG Pressure MSIS Delay - 14.0 seconds l
Reactor Scram Low SG Pressure Trip Delay - 0.9 seconds 3 seconds to fullinsertion "
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'4. Analysis Results -
4.1 Event DescrQtion A MSLB event is initiated by a rupture in a steam line. At break initiation, liquid in the SG secondary sides begins to flash, forcing steam out of the break. After a short amount of time, the intact SG is isolated by the closure of the MSIVs. The release of steam results in a large transfer of energy from the primary to the secondary side of the steam generators.
The large transfer of energy causes a rapid cooldown of the primary system. The cooling of the primary system, combined with a negative MTC, leads to an insertion of positive reactivity and hence, may cause a return to power. The return to power may cause fuel failures by violating limits for either MDNBR or FCM.
In general, the power increase for the MSLB' event is terminated by the dryout of the affected SG and by the Doppler feedback in the fuel. The event itself is terminated ultimately when the power begins monotonically decreasing as a result of boron injection or reheating of the primary and when the operator secures the AFW flow, which is credited to occur 10 minutes after initiation of the event. The consequences of this event are governed by the steam flow rate out of the break, the primary pump operating conditions, the magnitude of the MTC, and the initial primary operating conditions.
A MSLB is a cooldown event and is classified as a Condition IV event. Condition IV events are postulated accidents which are not expected to occur but are evaluated to demonstrate the adequacy of the design.
The consequences for the event are bounded by analyzing both HFP and HZP conditions.
At HFP conditions, the stored energy in the primary coolant is maximized, the available
- thermal margin is minimized, and the pre-trip power level is maxiniized. These conditions result in the g.eatest potential for cooldown. Initiating this event from rated power also has the potential for the highest post trip power since it maximizes the delayed neutron concentration, thus providing for the greatest power rise for a given positive reactivity insertion. Thus this event, initiated from full power conditions, will bound all other cases initiated from at power operation modes or power levels.
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. Post Scram Main Stearn Line Brock Analysis 98 6
, _for Millstone Unit 2 Rsvision 1 Page 4-2 For the zero power, and subcritical plant statete, there is also a potential for a return to power. The most limiting steam line break event at zero power is one which is initiated at the highest temperature and pressure, thereby providing the greatest capacity for cooldown. The most limiting conditions will occur when the core is critical. This~ cond will maximize the available positive reactivity, producing the largest return to power, Therefore, a steam line break occurring trom critical conditions will bound the results of ,
the event initiated from suberitical conditions.
4.2 Scenarios Analyzed
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The SPC methodology for analyzing MSLB events is documented in Reference 2.: In th i
cnalysis of the MSLB event, eight scenarios are considered. The reactor is initially' -
I operating at either HFP or HZP. From either of these initial conditions, the transient is assumed to occur either with or without offsite power and either inside or outside l )
j containment. With offsite power available, primary coolant pumps remain operatio$al j i throughout the entire transient. With loss of offsite power, the pumps are tripped at transient initiation. For inside containment breaks, harsh containment environment setpoints must be used for the steam generator low pressure trips and for the low '
pressurizer pressure trip (see Table 3.3).
i For all analyses, the most reactive control rod is assumed to be stuck out of the core .
This analysis considers only the response of the primary and secondary systems and do not include containment response for structural or equipment qualification purposes.
Two criteria are used in determining the most limiting scenario from a fuel failure standpoint:
one, a DNB criterion based on the modified Barnett CHF correlation; two, a FCM LHGR limit which precludes centerline melt.
i A summary of the calculated MDNBR, peak LHGR, and fuel failure results is presented ! in Table 4.1. ;
)
l The results of the ANF-RELAP calculation:s are presented in Section 4.3. The results fro I the XTGPWR calculations are presented in Section 4.4 and the results from the XCOBRA-tilC calculations are presented in Section 4.5.
- SiImens Power Corporation Nuclear Division
l . Pist-Scrarn M in Stsim Lina Brack Antlysis
,f:r Mihst:na Unit 2 sion 1 Page 4-3 4.3 ANF-RELAP Resuks The limiting DNB case was identified as the HFP outside containment scenario with loss of offsite power. The limiting FCM case was identified as the HFP outside containment '
scenario with offsite power available. ,
4.3.1 Umiting DNB i
The sequence of events for the DNB limiting case is summarized in Table 4.2. In addition, ksy systern parameters describing the transient is illustrated in Figure 4.1 through Figure 4.13.
l At time zero, a double ended guillotine break outside containment was assumed to occur in the Loop 2 steam line. The SG pressure responses are shown in Figure 4.1. The SG in Loop 2 is termed the "affected" SG and the SG in Loop 1 is termed the " unaffected" SG.
After break initiation, the pressure in the affected SG decreased immediately and then etabilized at atmospheric pressure around 220 seconds.
The break flow rates are shown in Figure 4.2. The break flow from the SG side of the break was governed by the affected SG pressure. The break flow from the unaffected SG was terminated due to MSIV closure at 16 seconds.
Figure 4.3 chows the MFW flow to the SGs increased initially in response to falling SG s:condary side pressures and was terminated at 23 seconds due to MFW valve closure.
AIW flow began at 1~80 seconds.
The pressurizer pressure and liquid level are shown in Figure 4.4 and Figure 4.5, respectively. The pressure decreased rapidly due to the contraction of the primary coolant during the initial phase of cooldown. The pressurizer emptied at approximately 60 seconds '
cnd system piessure was thereafter established by the saturation temperature of the primary coolant.
l The HPSI flow is shown in Figure 4.6. Ninety seconds after break initiation, the RCS pressure dropped below the shutoff head of the HPSI system and HPSI flow to the RCS began. .
=
The core power is shown in Figure 4.7. The reactor power declined to a decay heat level during the first 150 seconds of the transient. A peak power level of 7.7% of rated power Siemens Power Corporation - Nuclear Division u ----- _ _ - - - - - - - --
- Post-Scr:m M:in Sts:m Line Bre:k Antlysis EMF 98-036
, for Millstons Unit 2 R: vision 1 Page 4 4
[207 MW) occurred at 488 seconds with most of the core's power produced by the stuck
)
rod region, as shown in Figure 4.8. Figure 4.8 shows the core average, power in terms of percent rated power and the fraction of the core power generated in the stuck rod region, the affected core region (excluding the stuck rod region) and the unaffected core region.
The SG mass inventories are shown in Figure 4.9. After MFW flow was terminated, the affected SG secondary side mass inventory declined. With all of the AFW flow being directed to the affected SG, the affected SG's mass inventory began to increase around 250 seconds.
Core inlet temperatures are shown in Figure 4.10. This figure illustrates the temperature difference between the affected and unaffected regions of the core. The inlet temperature for the affected side of the core declined steadily during the first 400 seconds then leveled out.
in Figure 4.11, the reactivity feedback effects due to boron, moderator density, and Doppler feedback are plotted individually, along with total reactivity. The total core reactivity, initially at 0.00$, decreased instantly due to the scram worth at reactor trip, but then steadily increased due to moderator and Doppler feedback associated with the primary system cooldown. The reactor was soon brought into a quasi-steady-state with the Doppler and the moderator reactivities balancing the scram reactivity.
The core inlet flows are shown in Figure 4.12. The core flow decreased due to the RCP trip. Natural circulation flow was well-established by 250 seconds.
The transient was terminated at 600 seconds, at which time the reactor was in a quasi-stable-state. AFW flow was assumed to be terminated by the operator. Termination of AFW would cause the affected SG to dry out and an increase in the priri.ary system
- temperature. The increase in primary temperature, will drive the reactor subcritical and restore shutdown.
The ANF-RELAP boundary conditions required by XTGPWR and XCOBRA-IllC include core reactor power, core inlet temperatures, inlet mass flow rates, and core exit pressure.
! Figure 4.7, Figure 4.10, Figure 4.12, and Figure 4.13 show the reactor' power, the core inlet temperatures, the core inlet mass flow rates, and the core exit pressure, respectively .
I Peak LHGR and MDNBR were evaluated at the point of peak power.
Siemens Power Corporation - Nuclear Division l
L__ - - - - - - - - -
Post Scr:m Miln Ststm Lina Brezk Anilysis EMF-98 036
, * ,_fsr idt!!stons Unit 2 thvisi:n 1 Page 4-5 4.3.2 Limiting FCM The sequence of events for the limiting FCM case is summarized in Table 4.3. In addition, key systern parameters describing the transient is illustrated in Figure 4.14 through Figure 4.26.
At time zero, a double ended guillotine break outside containment was assumed to occur in the Loop 2 steam line. The SG pressure responses are shown in Figure 4.14. The SG in Loop 2 is termed the "affected" SG and the SG in Loop 1 is termed the " unaffected" SG.
After break initiation, the pressure in the affected SG decreased immediately and then stabilized by approximately 180 seconds.
The break flow rates are shown in Figure 4.15. The break flow from the SG side of the break was governed by the affected SG pressure. The sharp decrease in break flow at 490 seconds was caused by the affected SG drying out.
Figure 4.16 shows the MFW flow to the SGs increased initially in response to falling SG secondary side pressures and was terminated at 25 seconds due to MFW valve closure.
AFW flow began at 180 seconds.
I The pressurizer pressure and liquid level are shown in Figure 4.17 and Figure 4.18, respectively. The pressure decreased rapidly due to the contraction of the primary coolant during the initial phase of cooldown. The pressurizer emptied at approximately 60 seconds and system pressure was thereafter established by the saturation temperature of the primary coolant.
The HPSI flow is shown in Figure 4.19. Fifty-five seconds after break initiation, the RCS pressure dropped below the shutoff head of the HPSI system and HPSI flow to the RCS began. , !
The core power is shown in Figure 4.20. The reactor power initially declines due to the insertion of the control rods. The severe cooldown results in power increasing after 52 s:conds. A quasi steady state reactor po'wer level (approximately 370 MW) was established by 260 seconds. A peak power level of 14% of rated power [Q78 MW1 j
occurred at 462 seconds with most of the core's power produced by the stuck rod region es shown in Figure 4.21. Figure 4.21 shows the core average power in terms of percent Siemens Power Corporation - Nuclear Division i .
Post-Scram Mrin Stasm Lina Brock An: lysis EMF-98-036 for Millstona Unit 2 Ravision 1 Page 4-6 rated power and the fraction of the core power generated in the stuck rod region, the affected core region (excluding the stuck rod region) and the unaffected core regio The SG mass inventories are shown in Figure 4.22. The mass inventory in the affe 1
SG decreased throughout the transient. The unaffected SG mass inventory was essentially constant through the transient, with the exception of a slight decrease at th beginning of the transient. The relatively high teactor power level caused the affec to dry out by 490 seconds.
Core inlet temperatures are shown in Figure 4.23. This figure illustrates the temper difference between the affected and unaffected regions of the core. The intet tem for the affected side of the core declined steadily during the first 400 seconds then out. ;
f When the primary-to-secondary heat transfer deteriorated as the affected SG dried energy flow into the secondary system no longer balanced the flow of energy out of tne break. As a result, the RCS coolant temperature rose (Figure 4.23), the secondar 1
pressure decreased (Figure 4.14), and, since the break flow is determined by the secondary system pressure, the break flow also declined (Figure 4.15),
in Figure 4.24, the reactivity feedback effects due to boron, moderator density, and Doppler feedback are plotted individually, along with total reactivity. The total core reactivity, initially at 0.00$, decreased instantly due to the scram worth at reactor trip i then steadily increased due to moderator and Doppler feedback associated with the primary system cooldown. The reactor was soon brought into a quasi-steady-state with the Doppler and the moderator reactivities balancing the scram reactivity.
Because the pumps remain running, the coolant temperature used in the moderator feedback calculation is lower than it is for the loss of offsite power scenario.
Consequently the reactivity insertion due to moderator feedback is greater. The increased moderator feedback causes a larger increase in the core power than in the loss of offsite power scenario (compare Figure 4.11 with 4.24, and Figure 4.7 with 4.20).
The transient was terminated at 600 seconds, at which time the reactor is in a quasi-stable-state. AFW was then assumed to be terminated by the operator. Termination of AFW coupled with the affected SG dry out would increase the primary system Siemens Power Corporation - Nuclear Division '
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n i
, Pist-Scr:m MIin Strm Line Br;;k Anllysis EMF-98-036 frr Millstona Unit 2 R;visiin 1 Page 4 ~
1 1
i temperature. The increase in primary temperature will drive the reactor subcritical and l restore shutdown. j l
The ANF-RELAP boundary conditions required by XTGPWR and XCOBRA-IllC include core l reactor power, core inlet temperatures, inlet mass flow rates, and core exit pressure.
Figure 4.20, Figure 4.23, Figure 4.25, and Figure 4.26 show the reactor power, the core inlet temperatures, the core inlet mass flow rates, and the core exit pressure, respective .
Peak LHGR and MDNBR were evaluated at the point of peak power.
4.4 XTGPWR Results The reactivities calculated by XTGPWR were compared with the reactivities calculated by ANF-RELAP to ensure that the ANF-RELAP treatment of reactivity was conservative. For cach case, the comparison of ANF RELAP and XTGPWR reactivities was performed at the same initial and final conditions. A comparison of the ANF-RELAP and the XTGPWR reactivity changes at the time of peak post-scram reactor power is provided in Table 4.6.
The comparison shows ample conservatism for the ANF-RELAP reactivity change compared to the XTGPWR reactivity change. '
4.5 XCOBRA-lllC and Fuel Failure Results Two criteria were used to determine if any fuel failures result from the MSLB event:one, a DNB criterion based on the modified Barnett t.orrelation; two, a FCM LHGR limit which precludes centerline melt.
The DNB criterion was evaluated using tbs XCOBRA-IllC computer code and the modified B rnett CHF correlation. The calculated MDNBR values are listed in Table 4.1. All case ,
producec MDNBR values well above the correlation lirnit. Therefore, no fuel failures were '
predicted to occur as a result of violating DNBR limits.
Fuel responses were slso evaluated against the fuel centerline melt criterion. The calculated peak LHGR values -
Ested in Table 4.1. The HFP with offsite power scenario hid the highest LHGR valen "< 0 .
kW/ft, exceeding the 21.0 kW/ft ' limit. As a result, one full assembly,0.5% of the cr.w, is predicted to fait due to FCM.
- Siemens Power Corporation . tuclear Division
- Port Scrzm Mrin Stsam Line Break Anclysis EMF-98-036 for Millstone Unit 2 Rsvision 1 Page 4-8 Table 4.1 Post Scram Main Steam Line Break Analysis Summary ~
Initial Power Offsite Power Break Location Peak Post.
Level Availability MDNBR Peak LHGR Fuel Failure Scram Power (kW/ft) (% of Core)
(MW)
HFP off outside 207.5 1.71 17.96 containment 0.0 HFP off inside 207.2 bounded
- containment bounded' bounded" HFP on outside 378.0 2.28 24.27 0.5 containment HFP on inside 376.8 bounded' bounded" containment bounded
- HZP off outside 182.9 1.89 15.76 0.0 containment HZP off inside 182.4 bounded
- containrnent bounded" bounded" HZP on outside 343.5 2.37 23,47 containment 0.3 HZP on inside 344.1 bounded
- bounded" d containment bounded Note:
When ANF-RELAP results indicated a case was essentia4y equivalent or bounded by an MDNBR and peak LHGR calculations were not oone.
I l
- Bounded by HFP outside containment case without offsite power available (Table 4.2)
- Bounded by HFP outside containment case with offsite power available (Table 4.3)
- Bounded by HZP outside containment case without offsite power available (Table 4.4)
Bounded by HZP outside containment case with offsite power available (Table 4.5)
Siemens Power Corporation - Nuclear Division
PostScr:m M:in Ststm Lina Bruk Antlysis
, for M'illstona Unit 2 Rsvision 1 Page 4-9 Table 4.2 Sequence of Events - HFP Loss of Offsite Power - Outside '
vs. Inside Containment Break '
Outside Inside Tinie(s) Time (s) Event O. O.
Reactor at HFP O. + 0. +
4.1 Double ended gulliotine break. Loss of offsite power.
6.8 Low SG Pres. trip, Reactor trip B.6 14.5 MSIV and MFW valves closure trip signal 15.5 21.4 MSIVs closed 18.2 20.6 Si signal 22.6 28.5 MFW valves closed 43.3 45.7 SI pumps at rated speed (25 s delay) 180.1 180.1 AFW starts 488. 534.
N/A 559.0 Peak post-scram power reached (207.47 MW outside, 207.17 MW inside)
Si lines cleared. Boron begins to enter primary system 600. 600.
Calculation terminated. Power decreasing.
Table 4.3 Sequence of Events - HFP Offsite Power Available -
Outside vs. Inside Containment Break j i
Outside Inside Time (s) Time (s) Event {
O. O. Reactor at HFP
- 0. + 0. +
Double ended auillotine break. ,
4.2 7.5 Low SG Pres. trip, Reactor trip _
10.5 18.5 -
MSIV and MFW valves closure trip sianal {
15.9 19.7 SI signal I 17.4 25.4
{
MSIVs closed j 24.5 32.5 MFW valves closed ,
41.0 44.7 Si pumps at rated speed (25 s delev) 180.1 180.1 AFW starts {
462. 356.
Peak post-scram power reached (378.03 MW outside, 376.79 MW inside) l N/A 355.4 Silines cleared. Boron benins to enter primary system {
490. N/A SG dry out j 600. 600.
Calculation terminated. Power decreasina.
i l
Siemens Power Corporation - Nuclear Division l-
Pbst Ecrim Main Stsam Lina Break Anilysis EMF-98-036
,for Millstona Unit 2 Rzvision 1 Page 4-10 Table 4.4 Sequence of Events - HZP Loss of Offsite Power - Outside vs. Inside Containment Break
- Outside Inside '
Time (s) Time (s) Event O. O. Reactor at HZP O. + 0. + Double ended guillotine break.
Loss of offsite power. Shutdown reactivity inserted. Full AFW flow started, all directed to the affected SG.
7.4 11.9 MSIV closure trip sianal 14.3 18.8 MSIVs closed 21.8 21.5 SI sianal 46.8 46.6 S; oumos at rated speed (25 s delav) 301.7 296.3 Si lines cleared. Boron beains to enter primary system 320. 314.
600. Peak post-scram power reached (182.91 MW outside,182.41 MW inside 1 600.
Calculation terminated. Power decreasina.
(
Table 4.5 Sequence of Events - HZP Offsite Power Available -
Outside vs. Inside Containment Break Outside inside .
Time (s) Time (s) Event
__ O. O. Reactor at HZP O. + 0. +
Double ended guillotine break. N-1 rods inserted to simulate shutdown 7.7 13.1 condition. AFW increased to maximum flow and all directed to affected SG.
MSIV closure trio sianal 14.6 20.0 MStVs cloced 17.2 17.6 Si sianal 42.2 42.7 Si pumps at rated speed (25 s delavl 302.1 293.7 St Lines cleared. Boron beains to enter primary system 302. 294. .
Peak post-scram power reached (343.50 MW outside,344.09 MW inside) 600. 600.
Calculation terminated. Power decreasina.
l l
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i Siemens Power Corporation - Nuclear Division l a_-_________ _-. 1
, Post-Scram MIin Sts:m Lins Braak An:lytis EMF 98-036 for Millstone Unit 2 Revision 1 Page 4-11 I'
Table 4.6 Comparison of ANF-RELAP and XTGPWR Reactivity
- Initial Offsite Break ANF-RELAP XTGPWR Power Power Conservatism in Net Location Reactivity Reactivity input Parameters Level Availability Change Conservatism in Change (MTC, Doppler, and ANF-RELAP
($) ($) Scram Worth Bias) model HFP off (5) ($)
outside + 0.00 -6.30 + 5.30 + 1.00 containment HFP off inside bounded' bounded
- bounded
- containment bounded" HFP on outsido + 0.00 -5.87 + 4.86 + 1.01 containment HFP on inside bounded" bounded" containment bounded" bounded" HZP off outside + 6.69 + 3.00 + 2.72 containment +0.97 HZP off inside bounded
- bounded
- contairiment bounded
- bounded
- HZP on outside + 6.68 + 3.43 + 2.34 +0.91 containment I HZP on inside bounded
- bounded" d containment bounded bounded
- l I
1
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- Bounded by HFP outside containment case without offsite power available (Table 4.2)
- Bounded by HFP outside containment case with offsite power available (Table 4.3)
- Bounded by HZP outside containment case without offsite power available (Table 4.4)
Bounded by HZP outside containment case with offsite power available (Table 4.5)
Siemens Power Corporation - Nuclear Division
______--___-__-----_----q
, #3st-Scr:m M:in St2 m Lins Bruk An: lysis EMF-98-036 4
, _f:r M111st:nz Unit 2 R vision 1 Page 4-12 ,
1000.0 .
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Figure 4.1 HFP Loss of Offsite Power (Steam Generators' Secondary Pressures) '
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I l-Siemens Power Corporation - Nuclear Division l
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, P'ost- cr m Miin Sts:m Lina Brock An:1ysis EMF-98-036 I
! . , for Maist:n3 Unit 2 RIvision 1 !
Page 4-13 l
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Figure 4.2 HFP Loss of Offsite Fower *
(Break Flows) l l
l Siemens Power Corporation Nuclear Division
_ _ _ _ _ ,_ _ _ , , _ _ - - - - - - - - - _ _ _ _ _ . - _ _ _ _ - _ _ _ - - - . _ _ _ _ _ - _ _ _ _ . _ _ _ - -_-- _ - - - - - - - - - - -- - ---- ----- - - ~
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. P:st2 Scrim M:In Stum Lina Bre~k Analysis EMF-98-036
, f:r Millst::n2 Unit 2 Rsvision 1 l Page 414 i
l 4000.0 Millstone Cycle 13 MSLB HFP Pumps Off
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, l 3500.0 Unoffected SG MFW j -----
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l Figure 4.3 HFP Loss of Offsite Power . i (MFW and AFW Flows) )
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i Siemens Power Corpor'ation - Nuclear Division
_ _ _ _ _ , _ _ _ _ _ _ - _ . - - - - - - - - -' _______---A -
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,, for Millstona Unit 2 R: vision 1 Page 4-15 1
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1 Figure 4.4 HFP Loss of Offsite Power .
(Pressurizer Pressure) i i
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l Siemens Power Corporation - Nuclear Division l
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P:st Scrzm Mrin Stum Line Brnk Antlysis EMF-98-036
, int Millstona Unit 2 Rsvision 1 Page 4-16 70.0 ~ Millstone Cycle 13 MSLB HFP Pumps Off 60.0 -
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Fgure 4.5 HFP Loss of Offsite Power (Pressurizer Level)
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i Siemens Power Corporation - Nuclear Division
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Post-Sct:m M in Stnm Lina Brrk An lysis ~
' f:r 74tistona Unit 2 R; vision 1 l
Page 4-17 l
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Figure 4.6 HFP Loss of Offsite Power (Total HPSI Row)
I Siemens Power Corporation - Nuclear Division
_ _ _ _ _ _ _ _ _ l
, P:st S,crtm Mrin Strcm Line Br:ak An: lysin EMF 98-036 of:r Millst:n) Unit 2 Rsvisi:n 1 Page 4-18 500.0 , ,
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I Figure 4.7 HFP Loss of Offsite Power i (Reactor Power) l i
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. , f>:st, Scr m M:in Starm Lina Brrk Antlysis EMF-98-036
, for Millstona Unit 2 RIvlsl:n 1 Page 4-20 1
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c, Millstone Cycle 13 MSLB HFP Pumps Off 7 25.0 ,
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Figure 4.S HFP Loss of Offsite Power (SG Secondary Mass)
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Siemens Power Corporation Nuclear Division
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-. E:st Scr:m M;in St=m Lina Bruk Anilysis EMF-98-036
. far PJillst:na Unit 2 R1visiin 1 Page 4-21 600.0 , ,
Millstone Cycle 13 MSLB HFP Pumps Off Affected Region :
550.0 -----
Unoffected Region 2 500.0 s
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(Core inlet Temperatures) i l
!. Siemens Power Corporation - Nuclear Division E_ _ _ ._ _ _ _ _ _ _ - - _ _ _ _ _ _ _ . _ _ - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ---
-- - - - - - - - . _ _ _ _ _ _ _ _ _ _ _ .---_____-_-_------J
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Figure 4.11 HFP Loss of Offsite Power .
(Reactivity Components) f l
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Siemens Power Corporation - Nuclear Division 1
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, Po'st-Scr m Main St=m Line Br:1k An lysis EMF-98-036 i
. f::r Mil;ston2 Unit 2 R:: vision 1 i Page 4-23 l
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Figure 4.12 HFP Loss of Offsite Power -
(Core inlet Flow Rates) l 1
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23 , Millstone Cycle 13 MSLB HFP Pumps Off
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Figure 4.13 HFP Loss of Offsite Power *
(Core Exit Pressure) i e
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,P$t S.cr:m M in Sti:m Lina Br:ak Anity Is EMF-98-036 'i E f:r Mills' t:n2 Unit 2 R:visi:n 1 I I Page 4-25 '
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Figure 4.14 HFP Offsite Power Available *
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,P:st-S,crrm MIin Statm Line Bruk Anitysis EMF-90-036 j f:r Milletona Unit 2 RIvision 1 ;
Page 4-26 3000.0 Millstone Cycle 13 MSLB HFP Pumps On y ,i Affected SG -
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Figure 4.16 HFP Offsite Power Available *
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(Pressurizer Pressure)
I Siemens Power Corporation - Nuclear Division '
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EMF-98-036
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l Siemens Power Corporation - Nuclear Division l
. . - _ _ _ _ - _ _ l
, fo"qt-Qcram Main Starm Lina Bruk Analysis EMF-98-036 on R: vision 1 (or'.Milfstons Unit 2 i Page 5-1 )
f
- 5. Conclusions Th3 SPC methodology for analyzing steam line break events provides a conservative m thod of calculating the system and core responses during a MSLB.
Fu I responses were evaluated against the DNB criterion using the modified Barnett CHF correlation. The HFP with loss of offsite power scenario was determined to be the most limiting from a MDNBR standpoint, with a MDNBR of 1.71. All cases produced MDNBR vtlues well above the correlation limit, therefore, no fuel failures were predicted to occur cs a result of violating DNBR limits.
Fuel responses were also evaluated against the fuel centerline rnelt criterion. The c !culated peak LHGR values are listed in Table 4.1. The HFP with offsite power scenario h d the highest LHGR value,24.3 kW/ft, exceeding the 21.0 kW/ft limit. As a result, one full assembly,0.5% of the core, is predicted to fail due to FCM.
The analysis in this report supports the operation of Millstone Unit 2 Cycle 13 and for all future cycles bounded by the assumptions made in this analysis.
l Siemens Power Corporation - Nuclear Division t ..
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- N fo,gt-@cr:m Mrin2 St Im Line Br k AnrJysis i Millstonn Unit Rsvision 1 Page 6-1
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l 6. References
- 1. Millstone Unit 2 Steam Line Break Analysis, ANF-88-127, Advanced Nuclear Fueis Corporation, October 1988.
l
- 2. Steamline Break Methodology for PWRs, EMF-84-093(P), Revision 1, Siemens Power Corporation, June 1998.
- 3. Letter, D. E. Uhl to R. l. Wescott, "MP-2 MSLB Analysis information Request," NE l SAB-203, Revision 1, Janusry 13,1998. (This letter transmitted NUSCO document 4
M2-EV-98-0021, Rev sion 0, "TechnicalEvaluation for MSLB Analysis Information !
Request MP2," January 1998.)
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