ML20003J325
| ML20003J325 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 05/01/1981 |
| From: | Groce R Maine Yankee |
| To: | Carter J Office of Nuclear Reactor Regulation |
| References | |
| FMY-81-71, NUDOCS 8105110268 | |
| Download: ML20003J325 (52) | |
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ENGINEERING OFFICE FRAMINGH AM M ASSACHUSETTS 017o1 j
May 1, 1981 siv.st a-stoo g
2.C.2.1 81-71 g
4' United States Nuclear Regulatory Commission
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Washington, D. C.
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Attention: Office of Nuclear Reactor Regulation h
Division of Systems Integration
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Reactor Systems Branch M
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Mr. James Carter, Mail Stop P1130 L
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References:
(a) License No. DPR-36 (Docket No. 50-309)
_2 (b) MYAPC letter to USNRC (WMY 80-165), dated December 23, 1980.
(c) MYAPC letter to USNRC (FMY 81-21), dated Februa ry 24, 1981.
(d) MYAPC letter to USNRC (WMY 80-67), " Revised Steam Line Break Analysis," dated April 17, 1980.
(e) YAEC-1104, " Maine Yankee Plant Accident Analysis
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Model Using FLASH-4," W. J. Szymezak, dated November 1976.
Dear Sir:
Subject : Maine Yankee RETRAN Benchmark Analysis In support of our application of RETRAN, References (b) and (c), we are submitting a benchmark comparison of the RETRAN and FLASH steam line break models.
The FLASH model was documented in References (d) and (e).
A revised preliminary RETRAN model description for the Cycle 6 analysis accompanies this letter as Attachment A.
Attachment B contains the benchmark analysis results. A revised hard copy of the updated RETRAN input deck to be used for the Cycle 6 analysis will be submitted by Mid-May.
Minor modelling changes may be made to the model described in Attachment A during the actual cycle 6 analysis. These will be reported fully in the report to be submitted documenting the analysis in mid-June.
Should you desire additional information regarding the analysis, please contact Mr. Philip Guimond at Ext. 2123.
Respectfully, MAINE YANKEE ATOMIC POWER COMPANY d
W Robert H. Groce Senior Engineer - Licensing PJC/ nam 8105110 Mi f
Enclosu res
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9 ATTACIDfENT A MYAPS SLB RETRAN MODEL DESCRIPTION l
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c-b TABLE OF CONTENTS P,,a,gg TABLE OF C0NTENTS..........................................
11 LIST OF TABLES.............................................
iii LIST OF FIGURES............................................
iv 1.0 GENERAL....................................................
1 2
2.0 C0RE.......................................................
3.0 REACTOR VESSEL.............................................
5 4.0 R EACTOR CO O LANT L00 PS......................................
7 5.0 PRESSURIZER AND CVCS.......................................
9 6.0 CO NDE NS ATE SY S T EM..........................................
11 7.0 IIEATER DRAIN SYSTEM........................................
13 8.0 FEEDWATER SYSTEM...........................................
14 9.0 S TE AM DR IVE N F E E DWATER P UMP................................
16 10.0 FEEDWATER CONTROL SYSTEM AND FEEDWATER ISOLATION SYSTEM....
17 11.0 AUXI LIA RY F EEDWATER SYSTEM.................................
18 12.0 S TEAM C E NE RAT 0 RS...........................................
19 13.0 MAI N STE AM SY S TEM..........................................
22 14.0 BREAK MODELLING............................................
23 15.0 R EACTOR PROTECT IVE SYSTEM..................................
24 16.0 EME RGE NCY CORE COOLING SYSTEM..............................
25 17.0 LOW POWE R H0 DE LS...........................................
26 28 18.0 R E F E RE NC E S.................................................
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LIST OF TABLES Number Title Page 47 1.0 RPS Setpoints 1
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LIST OF FIGURES Number Title Pg 1.0 Reactor Vessel and Core Modelling 29 2.0 Core Heat Slab Nodalization 30 3.0 Upper Head Heat Slab Nodalization 31-32 4.0 Intact SCs RC Loop Piping 33 i
5.0 Faelted SC RC Loop Piping 34 6.0 Faulted SG Constant UA Modelling 35 7.0 Pressurizer CVCS and ECCS Modelling 36 8.0 Condensate System and Reheater Drain System Modelling 37 9.0 Main Feedwater and Auxiliary FW System Modelling 38 10.0 Steam Generator Modelling 39 11.0 Main Steam System and Break Modelling 40 12.0 Normalized Scram Reactivity Worth vs. Time Af ter Trip 41 13.0 Plant Startup on AFW Model Nodalization 42 14.0 Plant Startup on AFW Thru ist-PT FW Heaters l
Model Nodalization 43 15.0 Low Power Operation Model Nodalization 44 j
16.0 Co asate and Feedwater Pump Recirculation Flow l
Control Modelling 45 l
17.0 Pressurizer Level Control System Modelling 46 I
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1.0 CENERAL The Maine Yankee Atomic Power Station Nuclear Steam Supply System (NYAPS NSSS) consists of a pressurized water reactor with three parallel heat transfer loops. Each loop contains piping, a single Reactor Coolant Pump (RCP), and a U-Tube Heat Exchanger - Steam Generator. A pressurizer is connected to the hot leg of one of these loops via a surge line pipe.
References [1] and [2] describe in detail, the systems and components that comprise the NYAPS NSSS.
Together with the steam generators, the feed train, steam lines, turbine, and condenser form a second heat transfer loop. The Steam Line Break (SLB) accident postulates a rupture in the steam line portion of this loop.
I The following sections describe the modelling with RETRAN of each system either operating or required to function during a SLB transient.
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2.0 CORE The MYAPS reactor core consists of 217, 14x14 fuel assemblies.
The central 136.7 inches, in the axial direction, of each fuel pin contains 2 fuel pellets where the fissioning process takes place.
the enriched UO This region is modelled as the Active Core Volume, Volume 1 - see Figure 1.
Heating of the reactor coolant fluid takes place by conduction of the heat generated in the fuel pellets across the fuel pellet to fuel cladding gap space and through the clad to the cladding surface. Reactor coolant water flows along the heated clad surface and is heated by convection. A small fraction, 2.5 percent, of the energy generated by the fissioning process is deposited directly in the passing reactor coolant fluid by gamma radiation.
The approximately 38,000 fuel pins are lumped into a single average core heat slab, Heat Slab No.1, shown in Figure 2.
This slab is divided into three regions, the first containing 95% TD UO, the second representing 2
the fuel to clad gap space, the third the Zircaloy clad. Appropriate thermal conductivity and volumetric heat capacity data tables are provided for each region.
i Power generated within the core heat slab is determined by solution of a point kinetics reactivity model.
Six delayed neutron precursor groups are modelled. Moderator and fuel temperature reactivity feedback effects are included via input tables of reactivity defect. The direct deposition of 2.5% of the core power in the moderator by gamma radiation is modelled explicitly. J
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CEA scram reactivity is represented by a tabular function of reactivity worth versus time after trip. The reactivity effect of the horic acid injected by the ECCS is precalculated manually and input as a tabular function versus time after initiation of Safety Injectioa flow.
Decay heat power levels used are nominal va'ues from Reference 3.
The "enthalpy transport" option is activated in the core volume in order to obtain correct core exit fluid temperatures in Junction 2.
The remaining upper and lower portions of each fuel assembly are heated only by axial conduction and gamma deposition. Both effects are neglected in this model. The lower inactive core region is lumped together with the reactor vessel lower plenum as the Inlet Plenum Volume, Volume 9.
Flow enters the core from this volume via Junction 1.
The upper unheated portion of the core is lumped together with the region below the fuel alignment plate as the Upper Inactive Core Volume, Volume 20.
Flow enters this volume from the core volume via Junction 2.
Several pathways exist - such as through the CEA guide tubes and core shroud annulus - which allow the reactar coolant fluid to traverse the core region of the reactor vessel without contacting the heated fuel clad surface. Consequently, very little heating of this fluid takes place and it contributes very little to the cooling of the fuel pins. These portions of the reactor vessel and core region are represented in the model by a lumped Core Eypass Volume, Volume 18.
Reactor coolant fluid enters this region through Junction 16 from the inlet plenum and exits via Junction 18 to the Upper Inactive Core Volume. Hydraulic resistance in these pathe j
is set to result in the nominal steady-state value of core bypass flow, l l
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2.86 percent of total RCS loop flow during three pump operation.
The heated core exit flow and unheated core bypass flow mix in the Upper Inactive Core Volume. Most of this flow continues into the Outlet Plenum Volume, via Junction 28.
A small fraction of the total flow enters the CEA Guide Tube Shrouds, Volume 21, via Junction 23, and flows into the reactor vessel upper head, Volume 19, via Junction 30.
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3.0 REACTOR VESSEL The remaining portions of the reactor vessel are modelled as follows:
Separate volumes are used to model the downcomer and lower - inlet -
plenum.
Under idle loop or reverse flow conditions, flow entering the downcomer from one loop may circumferential1y traverse the downcomer and exit into another loop, without having to pass through the lower plenum.
This requires the two volumes to be modelled individually. This also results in a closer approximation to the actual transport delay for changes in fluid enthalpy without having to use the RETRAN transport delay option.
Downcomer A single volume, No. 8, is used to model the downcomer region.
Fluw normally enters the downcomer from the RCS loop cold leg piping, Junctions 8 and 14, flows through the volume and into the lower or inlet i
plenum, Junction 15.
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Inlet Plenum A single volume is used to model the lumped reactor vessel and lower inactive core regions. Flow enters this volume, Volume 9, from the downcomer via Junction 15 and exits via Junctions 1 and 16 to the Core and Core Bypass l
Volumes.
Outlet Plenum i
The reactor vessel outlet plenum is divided into two volumes, the l
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Outlet Plenum, Volume 2, and the volume enclosed by the CEA Guide Tube Shrouds, Volume 21.
This is required to obtain an accurate representation of the transport lag associated with fluid entering the Upper Head Region of the vessel from the Upper Inactive Core. Junction 30 connects the guide tubes to the Upper Head and Junction 19 the Upper Head and Outlet Plenum.
Upper Head The Upper Head Region is represented by Volume 19 and heat slabs 3 and 4.
During the SLB accident, the temperature of the fluid in this region may exceed local saturation pressure and void formation may result-This arises from the combination of RCS fluid shrink, lowered RCS pressure, low fluid flow rates through the region, and sensible heat storage in the vessel closure head metal. Accurate deternination of the extent of this voiding requires that the RV closure head and upper guide structure sensible heat be modelled. These are represented in the model as heat slabs 3 and 4.
See Figure 3.
With the exception of these two structures and the core fuel mass itself, no other sensible heat storage in RV valls or internals is modelled.
This is a significant conservatism for the SLB analysis, resulting in lower heat input to the RCS fluid than would be the case in reality. --
4.0 REACTOR COOLANT LOOPS The behavior of the NSSS during a SLB is asymmetric. The loops whose SGs have been isolated from the break by the EFCVs or NRVs behave dif ferently from the loop whose SG continues to blowdown. The two loops with isolated SGs behave similacly. These two RCS loops are lumped in the model. Minor dif ferences between the two loops would occur during the portion of the transient prior to the EFCV or NRV closure due to their unequal steam line lengths. These differences are unimportant and are neglected here.
The lumped loop representing the intact SGs and the loop representing the faulted SG are each divided into five volumes - see Figures 4.0 and 5.0.
The hot leg piping and SG inlet plenums are lumped together as one volume in each loop, Volumes 3 and 10.
Flow enters these volumes from the RV outlet plenum via Junctions 3 and 9, and enters the SG tube bundles through Junctions 4 and 10.
The heat transfer between the RCS and the secondary system takes place in the SG tube bundles. Heat is convected from the RCS fluid flowing l
inside the tubes, conducted through the tube metal, and convected again to the SG feedwater inventory. The average density in the inlet (inlet to U-bend apex) and exit (U-bend apex to tubesheet outlet) sides of the i
tube bundles is different. This af fects the calculation of natural l
circulation flow. Thus, the RCS side of the SG tube bundles are divided into two equal volumes per SG, the split occurring at the apex of the U-bends. Th's also allows proper representation of elevation and density effects within the SG under low flow or stagnant flow conditions. The I
"enthalpy transport" option is used in the SG tube bundle volumes.
The SG outlet plenum, cold leg suction piping and RCP volume are all lumped into a single volume in each loop. The RCP generates frictional heating of the RCS fluid which is accounted t:r by RETRAN. These volumes are, therefore, " heated" volumes and the "enthalpy transport" option is activated to calculate correct exit junction fluid temperatures.
The pressurizer, surge line, charging, letdown and ECCS are discussed els ewhe re.
5.0 PRESSURIZER AND CVCS The pressurizer and surge line are modelled as separate volumes.
This maintains the fidelity of the pressurizer level calculation and allows the temperature of the fluid in the surge line to be colder than the saturated pressurizer liquid.
As shown in Figure 7.0, Junction 35 connects the pressuritsr to the surge line and Junction 17, the surge line to the RCS loop with the unisolated SG.
The transient is not expected to be sensitive to pressurizer location Reference (5).
An equilibrium pressurizer model is used. During the SLB to be analyzed, the pressurizer will empty rapidly. Comparative analyses, using the RETRAN non-equilibrium and equilibrium pressurizer models show that predicted response for outsurge events is similar. However, the non-equilibrium model behaves erratically if the pressurizer empties.
The SLB results in the depressurization of the RCS. The pressurizer spray and relief valves are modelled, but are not expected to be actuated during the transient.
l The pressurizer heaters are represented in the model and controlled l
by pressurizer pressure. The low pressurizer level cutoff of the heaters to prevent burnout of the heaters should they become uncovered is modelled.
This setpoint is reached very early during the SLB - less than 25 seconds.
f Thus, the heat added by these heaters is negligible.
l The pressurizer level controller determines the normal charging and letdown flow rares from the RCS. The charging flow is modelled as a i
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positive fill junction, 37, into the lumped loop cold leg volumes. The letdown flow is modelled via Junction 38 as a negative fill from the single loop cold leg volume, 14.
This coincides with the physical plant layout where the charging lines are connected to RCS loops 2 and 3 and letdown to loop 1 (the pressurizer is also connected to loop 1).
The pressurizer level signal and controller are appropriately modelled by control blocks -
Figure 17.
Care has been taken to account for actual plant calibration and compensation. The effect of two phase conditions in the pressurizer liquid on calculated pressurizer level are accounted for by monitoring LIQL rather than MlXL in the local signal calculation.
The CVCS cnarging and letdown flows are isolated by an SIAS signal.
SIAS occurs in the first twenty seconds of a major SLB. The decrease in RCS level during these twenty seconds will result in some cold water being added to the RCS by the charging pumps. This is modelled, although the effect on RCS cooldown is small.
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d 6.0 CONDENSATE SYSTEM The condenser hotwell, condensate pumps, and 2nd through 7th point feedwater heaters comprise the Condensate System. Steam exhausted by the low pressure turbine is condensed on the tubes in the condenser and falls into the condenser hotwell. Two parallel feedwater heating trains begin at the condenser. - The condensate pumps take their suction flow from the condenser hotwell and pump this water through the feedwater heaters, where it is heated by condensing steam extracted from the turbine. This system is modelled as shown in Figure 8.0.
The condensed turbine exhaust steam is modelled as a positive fill, Junction 225, to the condenser hotwell. The condenser hotwell is modelled as a single volume, 201. The two feed-heater trains are lumped and divided into 5 volumes. The condensate puny.uction piping and condensate pumps (2) are modelled as two separate volumes, 228 and 202. Heat is added to the fluid by the condensate pumps so the "enthalpy transport" option is used in Volume 202 and 230. Volume 230 represents the standby condensate pump.
This pump will start if the MFW pump suction header pressure falls below a set value (420 psig for operation at low power levels with one condensate pump normally operating; 280 psig for operation at high power levels with i
two condensate pumps normally operating).
The 5th, 6th, and 7th point feedwater heaters and connecting piping is all lumped into Vclume 203, as are the Steam Jet Air Ejectors and Gland Seal Steam Condenser. A non-conducting heat exchanger model is used to represent the heat added to the fluid in all these components by the turbine extraction steam. The output of this heat exchanger is controlled as a function of time and plant status. l l
In a similar fashion, the 3rd and 4th point feedwater heaters and connecting piping are lumped into Volume 204. Extraction steam heating is again represented by a non-conducting heat exchanger.
Finally, the 2nd point heaters and piping are modelled as Volume 205, with another non-conducting heat exchanger representing the extraction heating.
Following a reactor-turbine trip the turbine extraction steam diminishes rapidly and feedwater heating is lost. For SLB analyses where the feedtrain pumps continue to operate it is conservative to shut off this feedwater heating source at the time of the break. This would result in cooler feedwater flow to the SGs later in the transient.
For cases where the MFWRV is assumed to fail open and the feedtrain pumps are tripped, the amount of feedwater flowing to the unisolated SG will depend upon the expansion of the hot feedwater lef t in the feedtrain forcing water into the SG.
Continued feedwater heating will increase the amount of expansion and is more conservative for this case. Feedwater heating will be assumed to continue for a conservatively long length of I
time for the case where MFWRV failure is assumed.
i "Enthalpy transport" is used in Volumes 203, 204 and 205. Junction l
205 connects the condensate system to the main feedwater pump suction header.
Junction 241 represents the condensate pump recirculation flow control valve. This valve monitors total condensate flow and is set to l
maintain a minimum of 4000 GPM total flow through the pumps. This logic f
is modelled using RETRAN control blocks (Figure 16.0).
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7.0 HEATER DRAIN SYSTEM The heater drain tank collects the steam condensed in the moisture-separator reheaters (MSRs), as well as the condensed extraction steam from the 1st point feedwater heaters. This fluid is pumped from the drain tank into the MFW pump suction header. The drain tank is modelled as Volume 207 - see Figure 8.0.
Junction 225 is a positive fill junction and represents the combined MSRs and 1st point.FW heater drain flows into the d rain tank. This flow is stopped following a reactor-turbine trip. The reheater drain pump is modelled separately in Volume 208. This pump is tripped of f when the level in the drain tank, Volume 207, falls below the s e t po int. This is represented in the model by ramping the pump speed to zero.
(RETRAN-01 has a problem with tripped pump frictional torque when use of the problem restart option is made. Erroneous results would occur, therefore, tripped pumps will be modelled by controlling pump speed directly to allow the use of the problem restart option.)
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8.0 FEEDWATER SYSTEM The Feedwater System extends from the feedpump suction header to the SG FW header ring nozzles. This includes the Main Feedwater Pumps (MFWP),1st point FW Heaters, MFWRVs, MFWRV Bypass Valves, and MFWP Recirculation Valves. Figure 9.0 shows the model nodalization.
The two electric MFWPs are lumped along with various piping into Volume 209. The 1st point FW Heaters and piping to the final diverging header is lumped into a single volume, Volume 210. A non-conducting heat exchanger represents the turbine extraction steam heating to these heaters.
"Enthalpy transport" is used in this volume and in the MFWP volume where pump heat is added to the fluid.
The two electrically driven MFWPs have minimum flow protection in the form of recirculation flow to the condenser. The MFWP recirculation flow control valve is modelled as a negative fill junction, Junction 237, taking flow from Volume 210. This fill flow is controlled by the value of FW flow in the MFWP suction, Junction 208 (Figure 16.0).
When the FW flow per pump falls below 1000 GPM, the recirculation flow control valve opens. An orifice in the recirculation path to the condenser limits this l
recirculation flow to 1000 GPM per pump. This is accounted for in the model by tripping the fill on when flow in Junction 208 decreases below 2000 GPM, and limiting the negative fill flow to 2000 GPM. When flow in the pump suction lines increases above 2600 GPM per pump, the recirculation flow l
control valves would close. A reset trip is used to perform this function in the model. Representation of this system is desirable to obtain accurate I
pressure distriF itions in the feedtrain under low flow conditions.
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r Upstream from the diverging header, the asymmetric SG conditions during the SLB are not " felt" by the FW system because of the crossover piping and headers interconnecting them. Thus, the dual FW trains have been lumped together up to the last cross-connecting header, termed the diverging header here and in Figure 9.0.
Downstream of this header each FW line " sees" its own SG conditions.
Therefore, a minimum of two trains must be represented; the line to the f aulted SG, and the lumped FW lines to the two intact SGs. The piping in these lines is further divided into two volumes each. The volume between the diverging header and the MFWRVs and MFWRV Bypass Valves, Volumes 212 and 213. The volume between the control valves and the SG FW header ring, Volumes 214 and 215.
The MFWRVs and MFWRV Bypass Valves are modelled as independent junctions in each train by tables of normalized equivalent junction flow area versus valve position. These tables were derived from valve Cy versus stroke data and the full open "K" loss factor input for each junction.
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9 '. 0 STEAM DRIVEN FEEDWATER PCMP A single steam driven FW pump capable of providing full FW flow at 100% power will be installed for Cycle 6.
This pump will be used in place of the two electrically driven pumps during operation between 50%
and 100% power. Volume 209 is modified to reflect this pump and associated piping for analyses where it is assumed to be operating. Pump and turbine speed is controlled to maintain a constant discharge header pressure. A transfer function supplied by Transamerica DeLa. val Company, the pump and turbine manufacturer, will be used to supply input " motor" torque to the pump model. This will be documented in more detail in the final Cycle 6 SLR analysis report.
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10.0 FEEDWATER CONTROL SYSTEM AND FEEDWATER ISOLATION SYS TEM During normal operation, the SG level control system positions the MFWRV and MFWRV Bypass Valves. A 3-element controller positions the valves to maintain a level setpoint. The three inputs to this system are SG sensed level, FW flow rate, and main steam flow rate. The simplified SG modelling used in the analysis is not capable of accurately determining sensed SG level during a SLR. Thus, the output of the 3-element SG level controller to the MFWRV or bypass valve must be conservatively determined prior to the analysis. The resulting valve position is input to RETRAN as a function of time only.
The Feedwater Isolation System will be modelled in RETRAN to perform the functions described below for the systen as modified during Cycle 6.
1)
Each EFCV low SG pressure trip signal will cause the MFWRV and MFWRV bypass valve to the associated SG to be shut.
2)
Any EFCV low SG pressure trip signal in conjunction with a SIAS will cause the main feedwater, heater drain, and condensate pumps to be tripped.
11.0 AUXILIARY FEEDWATER SYSTEM The Auxiliary Feedwater System (AFWS) provides auxiliary feedwater to the SGs through junctions with the MFW piping slightly upstream of the SG FW inlet nozzles. Separate piping runs from an auxiliary feedwater header to the individual MFW lines. Check valves exist in the AFWS piping upstream from the junction " tees".
These check valves close when pressure in the MFW piping is higher than in the AFWS piping.
The system is modelled - Figure 9.0 - by a single header volume, Volume 229, and three junctions. Two junctions represent the AFWS piping.
One, Junction 240, and a check valve represent the piping between the AFW header and the lumped intact SGs. The other, Junction 239, represents the AFWS piping to the faulted loop SG.
A positive fill junction, 238, represents the AFW pumps via a flow versus header backpressure curve.
Activation of the system will be modelled to reflect the Cycle 6 modifications described below:
1)
Any low SG level signal will autsmatically start two auxiliary feedwater pumps without any timer delay.
2)
Any Excess Flow Check Valve Low SG Pressure - EFCV - closure signal will trip both pumps if they are running and start an adjustable timer which will prohibit the starting or resterting of the pumps until the set elapsed time is reached.
3)
The three auxiliary feedwater control valves individually trip closed on receipt of an EFCV trip signal from their associated SG.
These valves will automatically re-open when the associated SG pressure returns above the EFCV trip setpoint...
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12.0 STEAM GENERATORS The primary (RCS) side of the SGs has already been discussed in Section 4.0.
Modelling of the secondary (steam-side) of the SGs is discussed below. Figure 10.0 shows the SG nodalization. As with the feedwater lines, the two intact SGs are lumped together in the model.
The liquid in the downcomer region of the SG during power operation is a mixture of " cold" FW and saturated recirculating fluid falling down f rom the steam separator deck. This mixture is subcooled and denser than the saturated water covering the tube bundle. The SG inventory modelled is equal to the total tube bundle and downcomer fluid mass accounting for the differences in density. An additional amount of mass is added to compensate for the dif ference in initial enthalpy of the actual downcomer region fluid and the conditions assumed in the model.
4 As in the FLASH (4) modelling of the SLB transient, a semi-infinite bubble rise velocity is used in the steam generators, Volumes 219 and 220.
This results in pure steam blowdown from the faulted SG.
The overall SG heat transfer coefficient, UAsc, was matched to a value calculated fres plant measured data at 97% power.
An integral assumption of past SLR analyses has been the use of a constant overall SG UA.
Flssh (4) was modified to maintain the UA in the SGs constant. Special modelling is required to accomplish this with RETRAN. Three nonconducting heat exchangers are used to model each SG as shown in Figure 6.0.
Control blocks are used to calculate the rate of heat l
transfer between each of the primary side SG volumes and the SG secondary 1
si'de volume as follows.
4=(UA)3g(t=0)*(TPRi (t) - Tsec (t))
where k
= Instantaneous heat transfer rate - MWt (UA)gg (t=0) = Overall primary to secondary heat transfer coefficient at 100% power initial condition (MWt/0F)
TPRi (t)
= Average temperature in SG primary side volume at time t.0F Tsec (t)
= Average temperature in SG secondary side volume at time t,0F Heat is removed from each SG primary side volume by the nonconducting heat exchanger at a rate, 6, determined by the calculation outlined above.
Thetotal4removedfromtheSGprimarysidevolumesisaddedtotheSG secondary side volume using another nonconducting heat exchanger. This calculation maintains the SG UA constant until the liquid level in the SG secondary falls below 9 feet (approximately one third of tube bundle height).
The SG UA is linearly ramped to zero at a mixture level of 0.0.
A comparison j
is made between the linearly ramped UA and a value of UA determined by assuming the forced convection heat transfer to saturated steam (using the Dittus Boelter correlation) over the entire SG tube heat transfer area.
The maximum of these UA values is selected to calculate Q for the non-l conducting heat exchangers.
Modelling of constant SG UA in the intact loop SGs is similar with l
one exception. The intact loop SGs are isolated by the closure of the EFCVs early in the transient. Following EFCV closure, the secondary side heat -
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transfer coefficient falls to a low value as reverse heat flow from intact SG secondary to RCS fluid begins. This is more conservative than the assumption of co'nstant SG UA in this reverse heat transfer mode since it retards the addition of stored energy from the SGs to the RCS. As a result, the RCS will cool down slightly faster, which is conservative for return-to power calculations.
The primary to secondary /1,T is monitored to determine when reverse heat transfer begins. The SG UA is then calculated based on a conduction solution from the SG inventory through the SG tube metal to the primary water.
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e 13.0 MAIN STEAM SYSTEM The steam line piping from the S0 outlet nozzles to the turbine stop valves comprise the Main Steam System (MSS). The SG ASME Code safety valves, steam bypass valves to condenser, the EFCVs and NRVs are included as part of the MSS. The steam line piping is broken down into six volumes -
see Pigure 11.0.
Two volumes, Volumes 222 and 223, model the portion of piping between the SG nozzles and EFCVs from the faulted SG and lumped intact SGs.
The AS:lE Code safety valves are modelled by negative fill junctions, Junctions 20 and 21, from these volumes. The EFCVs and NRVs are modelled in Junctions 229 and 230, depending upon whether they are assumed to function during a particular SLB transient. The steam line piping downstream of the NRVs to the steam header is modelled as one volume in each loop - faulted and intact - Volumes 224 and 225. The steam line header is represented as Volume 226 and connects the steam lines from the faulted and intact SGs.
The remainder of the steam line piping, the four lines to the turbine inlet bowl, are lumped into Volume 227. The turbine inlet stop valves are modelled in Junction 22, controlling the negative fill flow representing the steam flow to the turbine. These valves are closed oa reactor-turbine trips, but are assumed to remain open until then.
The steam dump and bypass systems are not activated during a SLB and are not currently modelled.
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14.0 BREAK MODELLING The areak modelled here is a double-ended guillotine break at the SG nozzle. Two negative fill junctions are used, each representing one side of the break: Junction 234 the SG nozzle side, and Junction 235 the.
steam line side. Both break junctions use a Moody [4] critical flow table to determine break flow versus upstream pressure. The use of a large bubble rise velocity in the SG modelling results in single phase steam flow from the faulted SG.
This maximizes the energy removal rate from the RCS.
RCS cooldown rate is also maximized. Junction 228 is modelled containing a valve which is closed when the two break fill junctions are turned on.
(
15.0 REACTOR PROTECTIVE SYSTEM The Reactor Protective System (RPS) monitors a numbe-* of plant process variables. 'When preset limits are reached, a reactor trip signal.
is generated that releases the CEAs into the core. Table 1 provides a list of the trips modelled along with the nominal trip setpoints, time delays, setpoints (modified by applicable uncertainties) and signal source as modelled in RETRAN.
Inserted CEA reactivity worth is a function of CEA position in the core. CEA position is a function of time after trip signal generation.
Conservative values for both functions were used to generate the reactivity worth versus time af ter trip table (shown in Figure 12.0) and input in the RETRAN scram reactivity table.
I t
16.9 EMERGENCY CORE COOLING SYSTEM The ECCS system consists of the High Pressure Safety Injection (HPSI) pumps, the Low Pressure Safety Injection (LPSI) pumps, and the Safety Injection Tanks (SITS). During a SLR, the system is activated by the SIAS as the pressurizer empties.
RCS pressure remains high enough throughout the transient that only the HPSI pumps actually inject fluid into the core.
Reliance is placed on boron injected by the HPSIs to maintain margin to criticality af ter shutdown. This is accounted for in a separate manual calculation of boron reactivity worth versus time.
The effect of the cold SI water on the RCS cooldown is modelled by a positive fill table and two fill junctions, as described in the RCS piping section. The fill table is modified to reflect the number of HPSI pumps assumed to be operable for each transient.
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1 17.0 LOW POWER MODELS 1
During plant startup and low power operation, the configuration of the feedwater system varies. Each configuration is described below, along with any changes required to the RETRAN model nodalizations.
Plant Startup to Approximately 5% Power The feedwater system is usually aligned in one of two ways during plant startup.
First, the auxiliary FW pumps may be used to supply FW directly to the SGs via the auxiliary FW piping. The MFWRVs and MRFRV Bypass Valves would be closed in this configuration. The RETRAN model for this condition is shown in Figure 13.0.
Second, the auxiliary FW pumps may be aligned to feed flow through the 19t point FW heaters and MFW system piping to the SCs. The MFW pump discharge valves, upstream from the lat point FW heaters, would be closed.
Auxiliary feedwater flows into the Main Feedwater System downstream of these valves, through the 1st point heaters and MFWRV Bypass Valves into the SGs.
Figure 14.0 shows the corresponding RETRAN nodalization.
Low Power Operation During plant operation at power levels from approximately 6% to 50% power, a single electric MFW pump is operating. The characteristics of Volume 209 are adjusted accordingly. The model nodalization remains unchanged from the full power model, Figure 9.0, with the exception of the addition of Volume 231, Figure 15.0, representing the standby MFW pump. - _ -
This pump will start automatically if pressure in the MFWP discharge header falls below 1075 psig.
1 l
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18.0 REFERENCES
1.
Maine Yankee " Final Safety Analysis Report", Rev. 6, 1971.
2.
YAEC-1101, " Maine Yankee Plant Analysis Model Using CEMINI-II",
P. A. Bergeron, June 1976.
3.
NUREG-75-087, Branch Technical Position ASB 9-2, Rev.1, November 24, 1975.
4.
YAEC-1104, " Maine Yankee Plant Accident Analysis Model Using FLASH-4",
W. J. Szymczak, November 1976.
5.
BNL-NUREG-25781, Informal Report, Maine Yankee Steam Line Break Analysis Using RELAP-3B, W. C. Shier, March 1979, Thermal Reactor Safety Division, Brookhaven National Laboratory.
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Signal Source U
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ATTACHMENT B MODEL BENCHMARK ANALYSES INTRODUCTION The benchmark analyses consisted of two transients.
The first was a 3 RCP flow coastdown. Startup test data [1] was used for comparison. This coastdown was extended in time until stable natural circulation -N/C-conditions were obtained which could be compared with N/C data from [1].
The second transient repeated the analysis of Case No. 3 from Reference [2], which had used the FLASH IV model from Reference [3].
FLOW COASTDOWN AND NATURAL CIRCULATION ANALYSIS The RCP flow coastdown transient used a truncated vereton of the RETRAN model described in Attachment A.
The feedtrain nodalization upstream l
f rom the downcomer Volumes, 218 and 221, was deleted. Feedwater was supplied by two negative fill junctions.
Since the flow coastdown tests [1] were performed prior to initial core startup, no nuclear core power existed, only pump hest from the RCPs'.
For convenience, the RETRAN model assumed a core power of 11.3 MJ.
This is equal to the core power measured during the N/C test conducted sometime later [1]. This small power level does not ef fect the coastdown results and allowed the modelling of the N/C test by continuation of the coastdown tradmient to stable N/C conditions. --
- -- ~
The SG Bypass System was modelled and maintained SG pressure at 905 psia, its nominal setpoint at hot standby.
The SG heat transfer area, modelled by the SG heat slabs, was increased to produce an overall UAsc equal to the measured plant full power value. These values will be used in the SLB analysis and were therefore modelled for this analysis.
The pump frictional torque coefficient was adjusted to produce good agreement betwee-the test data and RETRAN. Figure 1 illustrates normalized core flow vs. time. Figure 2 compares steam generator flow vs. time. Figure 3 shows pump speed vs. time.
The ratio of normalized power to normclized flow was determined by extending the coastdown until N/C flow was well established. The measured plant value [1], normalized to 2630 MWt, and 393,000 CPM is approximately 0.20.
RETRAN predicts a value of 0.17.
This value is within the measurement uncertainty of the startup test value.
RETRAN predicts slightly higher N/0 flows than measured data. This is conservative for SL3 analysis since higher heat transfer and faster cooldowns should result.
FLASH BENCHMARK ANALYSIS The hot full power SLR transient reported as Case No. 3 in Reference l
[2] was reanalyzed with RETRAN.
1 I
Several modifications were made to the base RETRAN model described 9k in Attachment A.
1.
The MFW and condensate pump recirculation flows were not allowed to function as they had not been modelled i; FLASH.
2.
The steam line modelling was changed to exactly match that used in the FLASH modelling, 3.
For convenience, reactor power vs. time was input directly as a table of values from the FLASH run.
4.
The MFWRV, MFWRV bypass valve and EFCV operating times were assumed to be the same as in the FLASH analysis.
5.
The SG UA was maintained constant in both SGs. FLASH assumed the SG UA to instantaneously drop from 100% to 07 at mixture levels below 0.25 feet.
Figures 4 through 15 provide a comparison of the results of this transient with the FLASH results. Excellent agreement was obtained.
A notable exception being the RCS pressure response. Separate modelling of the pressurizer and upper head allows the relatively stagnant upper head water to flash, creating a second " pressurizer" which tends to keep system pressure higher. In the FLASH nodalization the pressurizer was connected to the Upper Head and therefore no stagnant upper head existed and no upper head flashing could occur. _ _ _ -_
' REFERENCES 1.
Maine Yankee Atomic Power Station, Startup Test Report, J. 'D. LeBlanc, July 1973.
2.
MYAPC Letter to USNRC, WMY 80-67, Revised Steam Line Break Analysis, April 17, 1980.
3.
YAEC-1104, " Maine Yankee Plant Accident Analysis Model Using FLASH-4", W. J. Szymczak, November 1976.
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