The INL is a U.S. Department of Energy National Laboratory operated by Battelle Energy Alliance
INL/EXT-15-35316
Information on the Advanced Plant Experiment (APEX) Test Facility
Curtis Smith
May 2015
Idaho NationalLaboratory
DISCLAIMER
This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not
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INL/EXT-15-35316
Information on the Advanced Plant Experiment (APEX) Test Facility
Curtis Smith
May 2015
Idaho National Laboratory Idaho Falls, Idaho 83415
http://www.inl.gov
Prepared for the U.S. Department of Energy Office of Nuclear Energy
Under DOE Idaho Operations Office Contract DE-AC07-05ID14517
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Information on the Advanced Plant Experiment (APEX) Test Facility
The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources:
Reference 1: R. Nourgaliev and et.al, "Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing," Idaho National Laboratory, 2011. Note that this is report has not been released as an external report.
Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996.
Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003.
Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999.
Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.
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R. Nourgaliev and et.al, "Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing," Idaho National Laboratory, 2011.
In 2011, the INL produce a technical report on attributes for a “next generation” safety analysis tool called R7. As part of this report, information was provided on the APEX facility since it was used to demonstrate the capabilities of the software. The background on this use is:
• Leveraging on a unique experimental facility APEX, at the Oregon State University • APEX simulate PWR (W and CE), AP600, and AP1000 • Quarter-scale integral- effect test facility, with extensive diagnostics • Cross-benchmarking with existing models and “legacy” STH codes • Developing input decks compatible with R7 graphics-user interface (GUI) • A coarse-grained 0D-1D STH-grade model of APEX • A fine-grained 3D CFD-grade model of APEX • Analyzing past APEX experiments for their (partial) relevance to validation data needs in
support of F&B scenario model calibration
General information was provided on APEX, including the following:
STEAM GENERATOR (SG)Each SG is uuMmented asel hes a tube and O. made to
um.. a V.sOnghoune Delta 75 Stearn Generator. Each
SG WISP. 133 V tubas In a preuured water react«
nue.f pewee plAnt. hog,temperotorefft5C101 content water
heated by elo reactor generates stew n rn the 510.1 genera-
CORE MAKEUP TANKS (CMTs)RroD 00S.renteehOn CUTS are called open fear:wog tran
cett. whore the normal makoup syslem 116 oladequaia Trvars lord se. bor.. water Pfl MO parallel trans ere Oe-&One., 10 lunch. 01 any reactor 4.1.1 system presswe us
np Oilly gravy, and MO IffIT,T7Pllo Aftel 11.9t111,11POPPIVAPF
from tho reactor cootan1 %Worn odd 1eg al the rnotreatng
fOleti
REACTOR COOLANT PUMPS(RCPS)
Four van...speed RCPs ham been included as part of PO
RelICIC4 COSILII4 System OrCuLorg tame volumes of reaclOr
coolant. The RCP. oenr/a. the API 000 Canned motor
Pampa and are 0110/sed lo the tower Channet head of eaeh
SG. Each pump outlet ra ccnnectert to a rung. CI..
ACCUMULATORS (ACCs)Medium pressure noel., P.C.C* are requeed for large Ions.
of-coolant accadents. rehlkne lne RPV foaavno9ant system blonOonn The AGCs contain vraserpressucaed
with nitrogen gas Two accumulators o hes paralux Pains are
sued to respond to Me amp.* severance of the Sagest re.
actor coolant system ppe. raptly reliang the RPV
HOT LEGS (HLs)lbe trot lege detner heated water Porn the reactor mssel tothe steam generators
COLD LEGS (CLs)The odd legs return Reactor Coolant System WOkfe from Me
SteYn generalOrs back 0/ the fen., vessel
DIRECT VESSEL INJECTION(DVI)
The DM ..1113 clesean os unque el that rt connects dseclN toMe reactor vessel The perm,. the 'del,/ vrater inittDon
St•Pill.o..11,1PAing Oat ACC, and I RINST Ore.y
enter the reactor neerni
ELECTRICAL HEATER RODSThe APEX RPV cenLnns 48 clear. heater MOS MeV, Ser.
Ole the Men. heat geseerals00 10.0 in the APIOCO II1JONV1.1 asseenhhas •The APT% Glary draws up to I 01.1W.
electric of energy from the power and demo full.power test.
mg procedures
AUTOMATICDEPRESSURIZATION SYSTEM
STAGES 1-3 (ADS1-3)The ADSI.3 elapeosurves lM MI.. 0411.1011 syn.. AM
anoabs b.ser pressure sla.ty asecoon water lo «M« the re.
OCIOr eels. and Poo coro. II is Wonted by a level sot pool
ther Gas. ADSI.3 is comprised of three slopes of calves
l000led above the ',resumer
— PRESSURIZER (PZR)APEX employs a fay furrchenat PZR, wen internal 1.10/41.
end • relief valve system caps.* of comm.° the Read01
00041018yStern pressure. Tho PIA d connected to 01.2
though Ise PZR surge Ilne. The PZR rnentains the Reactor
COOkst STA.. Prouctra during steeds.state opera/bon and
IMb presswe charges donna transtents
IN-CONTAINMENT REFUELINGWATER STORAGE TANK (IRWST)Tho IRWST la the he:t sok fee PIO PRIIR heat eschmger
The IRWST water volume u sufficient to absorb decay heat
It W. saves lo condense steam horn the ADSI .3 system
spaeger donna reader Clov.down
PASSIVE RESIDUAL HEATREMOVAL SYSTEM (PRHR)
Thn PRIM subsystem prOteela the plant opens, upsets to
the normal heal removal horn Po sxmary system ay Ihe
steam gcneraor feedwater and steam systems The pasuve
0110 wheyelarn ...hos MOO S NOG safety aeons !of
loss of feed... leedwuter.to Dee., end steamer,*
beaks M111 o song. fOkoe
AUTOMATICDEPRESSURIZATION SYSTEM
STAGE 4 (ADS4)ADS. iS Contreeed by OeiVeSannecled to the hol legs
ADSe reduces the ROV pros.ro lo elmosphenc. dlowingt
Pane/ evechon from Na IRWST The eventually M.., ago
a long term coolog roode Oath comainment sump weals-
...
REACTOR VESSEL (RPV)The RPV models the upper and lamp internals oT the
API000reactOr vessel. ecte Darrel. GavelCOnlee. and COre.The max.. COr0 a... ePPro,,meR,Y MW.I.Olo.which can be disinbuted en two radial power vases and Can
to programmed. sandal, Ime..dopenelent deary POwer.
The roared nolo( To core COP11130S of aft WOK Mato
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The R7 input deck (once visualized) looked like:
6 This input deck w
as based upon an older RELAP deck:
"ADS-V3
- Pritit out
44..2 rik•
ADS4-2SEP
Mr.. 1. arr.204,..,•X••—•---•••••••••,
CUT:2 ACC-2 •
r 4YYYVVVmy
Hot LEg- 2
l• •• • • • •C441 Les.2
KtLAVO AlatA Arlutm maanation uiagranOregon State Universit)
ReadorVessel
D0
0
IOS 4-1 Va%e
110S4.1SEP•
• .1.1.•VA-. a
. • • • -
_ _ HMCO 4.7
.Someday SapMany StopTL• -F4=_.‘.
1
--it 0-4441-Cr
CMIT-1 P81..
ACC-1
7
Key components in the R7 deck were identified:
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The INL also included additional details – this information was contained in the following four chapters:
Pipe2, 2m-long
Heater Core
x
z
Pipe3, 6m-long
HX
Elbow2
Pipe5, 2m-long
P i pee, 2m-long
Pipe1, 2m-long Elbow3
Pipe7, 5m-iong
Pump, lm-long
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Chapter 1
APEX Test Facility 1.1 APEX Test Facility Description
The Advanced Plant Experiment (APEX) test facility at Oregon State University is maintained and operated by the Department of Nuclear Engineering and Radiation Health Physics. It was constructed in 1994 to perform integral system tests to simulate the important thermal hydraulic behavior of the Westinghouse AP600 reactor. The facility was then modified, through a grant from the U.S. Department of Energy, to simulate the Westinghouse AP1000 for the purpose of AP1000 plant certification through assessing system code capabilities and integral system behavior.
The test facility models a complete Westinghouse AP1000 2x4 loop containing two hot legs, 4 cold legs, 2 steam generators, pressurizer, reactor pressure vessel with an electrically heated rod bundle and upper plenum internals. It is a 1:4 length scale, 1:2 time scale, 1:192 volume scale of the prototype AP1000 and is of stainless steel construction. It completely models the passive safety systems of the AP1000 containing: 2 core makeup tanks (CMTs) , 2 accumulators (ACCs), a passive residual heat removal (PRHR) heat exchanger, in-containment refueling water storage tank (IRWST), and a 4-Stage automatic depressurization system. The steady state operating conditions of the facility are with a core power at about 1 MW, steam generator shell side pressure at 2 MPa (290 psig), and pressurizer pressure at 2.55 MPa (370 psig). Past testing efforts utilizing APEX have include hot and cold leg SBLOCAs, MSLB, Inadvertent ADS, Double-Ended DVI Line Break, Station Blackout and Long Term Recirculation. Figure 1.1 show a
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Fig. 1.1 : APEX test facility at Oregon State University.
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Chapter 2
APEX R7 Simulation Model 2.1 Introduction The physical APEX test facility is transformed into working model within R7 as shown in in Figure 2.1. The following sections work to describe the systems, sub-systems (groups), and components as they are modeled in R7 GUI.
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Fig. 2.1 : Model of APEX in R7 .
7
Prim
ary
Loop
6 CHAPTER 2. APEX R7 SIMULATION MODEL
2.1.1 APEX R7 Model Groups The APEX test facility is broken up into groups for modeling purposes. Table 2.1 gives each group with a list of the major components of that group along with the fidelity utilized in the APEX R7 model.
Table 2.1 : APEX R7 Model Components.
System Group Component Name Dimension Reactor Pressure Vessel Downcomer Fluid 1-D/3-D Downcomer Inner Wall 3-D Downcomer Outer Wall 3-D Upper Plenum 0-D Lower Plenum 0-D/3-D Core Lower Core Plate 1-D Heater Rod Section 1-D Upper Core Plate 1-D Upper Internals Support Plate 1-D Core Upper Internals 1-D Upper Support Plate 1-D Cold Leg Cold Leg 1 1-D Reactor Coolant Pump 1 1-D Cold Leg 2 1-D Reactor Coolant Pump 2 1-D Cold Leg 3 1-D Reactor Coolant Pump 3 1-D Cold Leg 4 1-D Reactor Coolant Pump 4 1-D Hot Leg Hot Leg 1 1-D Hot Leg 2 1-D
...continued on next page
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Safe
ty S
yste
ms
Prim
ary
Loop
...continued...APEX R7 Model Components.
System Group Component Name Dimension Steam Generator Steam Generator 1 U-Tubes 1-D Steam Generator 1 Primary 0-D/3-D Hot Side Lower Plenum Steam Generator 1 Primary 0-D/3-D Cold Side Lower Plenum Steam Generator 2 U-Tubes 1-D Steam Generator 2 Primary 0-D/3-D Hot Side Lower Plenum Steam Generator 2 Primary 0-D/3-D Cold Side Lower Plenum
Pressurizer Pressurizer Surge Line 1-D Pressurizer 0-D ADS 1-3 ADS 1-3 Piping 1-D Oriface (ORI-653) 1-D ADS 1 Valve (RCS-601) 1-D Oriface (ORI-655) 1-D ADS 2 Valve (RCS-602) 1-D Oriface (ORI-656) 1-D ADS 3 Valve (RCS-603) 1-D Oriface (ORI-657) 1-D ADS 1-3 Seperator Tank 0-D Oriface (ORI-659) 1-D CVS CVS Piping 1-D Ball Valve (RCS-820) 1-D Check Valve (RCS-827) 1-D Air Operated Valve (RCS-808) 1-D Check Valve (RCS-809) 1-D CVS Pump 1-D PRHR PRHR Piping 1-D PRHR Oulet Valve (RCS-84) 1-D IRWST IRWST Tank 0-D
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Chapter 3
Primary Loop Inputs
3.1 Core Downcomer
3.1.1 R7 Model
Fig. 3.1 : Core Downcomer.
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3.2 Core Lower Plenum
3.2.1 R7 Model
Fig. 3.2 : Core Lower Plenum.
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3.3 Core
3.3.1 R7 Model
Fig. 3.3 : Core.
A
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UpperPlenum
UP1
UpperSupportPlate
CRD3
CRD2
CRD1
Core 3
Core 2
Core 1
LowerCorePlate
DC-1/2/3/4
LowerPlenum
Simple Tank
Fig. 3.4 : Core.
10.453"
9.7"
3.0"
13.63"
0.75"
2.58"
97 583"
38.75"
1.51"
3.5"
6.41"
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3.4 Core Upper Plenum
3.4.1 R7 Model
Fig. 3.5 : Core Upper Plenum.
3.5 Cold Leg 1
3.5.1 R7 Model
Fig. 3.6 : Cold Leg 1.
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3.6 Cold Leg 2
3.6.1 R7 Model
Fig. 3.7 : Cold Leg 2.
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3.7 Cold Leg 3
3.7.1 R7 Model
Fig. 3.8 : Cold Leg 3.
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3.8 Cold Leg 4
3.8.1 R7 Model
Fig. 3.9 : Cold Leg 4.
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3.9 Hot Leg 1
3.9.1 R7 Model
Fig. 3.10 : Hot Leg 1.
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3.10 Hot Leg 2
3.10.1 R7 Model
Fig. 3.11 : Hot Leg 2.
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3.11 Steam Generator 1 U-Tubes
3.11.1 R7 Model
Fig. 3.12 : Steam Generator U-Tubes.
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3.12 Steam Generator 1 Hot Lower Plenum
3.12.1 R7 Model
Fig. 3.13 : Steam Generator 1 Hot Lower Plenum.
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3.13 Steam Generator 1 Cold Lower Plenum
3.13.1 R7 Model
Fig. 3.14 : Steam Generator 1 Cold Lower Plenum.
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3.14 Steam Generator 2 U-Tubes
3.14.1 R7 Model
Fig. 3.15 : Steam Generator U-Tubes.
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3.15 Steam Generator 2 Hot Lower Plenum
3.15.1 R7 Model
Fig. 3.16 : Steam Generator 2 Hot Lower Plenum.
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3.16 Steam Generator 2 Cold Lower Plenum
3.16.1 R7 Model
Fig. 3.17 : Steam Generator 2 Cold Lower Plenum.
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3.17 Pressurizer Surge Line
3.17.1 R7 Model
Fig. 3.18 : Pressurizer Surge Line.
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3.18 Pressurizer
3.18.1 R7 Model
Fig. 3.19 : Pressurizer.
.Y 6(
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Chapter 4
Safety System Inputs
4.1 Automatic Depressurization System 1/2/3
4.1.1 R7 Model
Fig. 4.1 : Automatic Depressurization System 1/2/3
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Reference 2: Owen Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger
As noted in reference 2:
“The OSU Radiation Center (the location of the Oregon State University Department of Nuclear Engineering) houses a one quarter scale model of the Westinghouse Electric Corporation advanced light-water nuclear reactor design called AP600.
The AP-600 reactor design incorporates many passive safety features for reactor core cooling. In this case, passive means that the systems are capable of core cooling using only the phenomena of gravity driven flow and natural convection of heated fluids. The model of the AP-600 (APEX) was built to perform the testing necessary for design certification.
APEX operates at 2.76 MPa (400 Psia) and has been formally scaled' to simulate the important thermal hydraulic behavior of the AP-600. APEX is electrically heated and simulates the nuclear steam supply system (NSSS) and all of the AP-600 safety systems. The systems modeled include the primary system, passive safety systems, the non-safety grade chemical and volume control system, and the residual heat removal system (PRHR)..”
Additional details on the system shown above are found in the Reference 2 report.
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Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003.
Reference 3 explains some of the concepts behind the scaling approach used by the APEX researchers. While much of the technical information in Reference 3 has been redacted from the public document, some information is still presented, including:
1
0.9
0.8
0.7
O 0.6.i. ezE;,,. 0.5:0
0.4
0.3
0.2
0.1
0
• AP1000
• APEX-1000
•4.,•'.
—31'.• --•
•a.
, r T
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9
Normalized Axial Position
Figure 18 Comparison of AP1000 and APEX Axial Void Fraction Profiles for an AverageSubchannel
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1 0.9751.95C
Bellvite Disk WasherSS301: or ECeAvo.:ent
HCMaster Carr 497131486
I. 689
SrC T ION amilc
0.745
0041 tnrd8 tOtal 2 30'(except 0,90,180, 1.270")on 018 3 circle
0..500 Woe 7../te (48)
02.33C Floe Hale (7)
ostscco
01.913 Perneter Floe Hole (6)
3/4-16 te4F-23 ter..(Threoded HcteFor 5.4iie Tube)
SS3C4 or ecoleolnec
Radlks loner edge
ce Flow Holes (13) 0.25'
30 88 thru y 034t x at depth(4 total etwall spaced on ciao ercle)
"-"10 MMI: At:. MOW
0.750
1-023C
0169 ROP50 t wr en, all_
Floe Ades (13)
Figure 21 APEX Upper Core Plate Geometry
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Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999.
Reference 4 summarizes the research effort performed using APEX at Oregon State University. Some facility information is shown in the report, however the report notes:
“Three proprietary compact disks are available as a supplement to this report. They include APEX Facility Drawings, APEX Test Reports, and the APEX Database. These three compact disks include all of the information generated in support of NRC's test program at OSU.”
The information from the compact disks is not publically available. Summary of information show in this reference includes:
Photos of key parts of the system shown above are included in the report. Information related to the testing initial conditions are also included.
Layout of the APEX Test Facility
The APEX Testing Facility
Steam Generator
Core MakeupTank (CMT)
Accumulator
DVI
Lines
ADS 1-3
Pressurizer
Reactor
break Lao
Separator
IRWST
PRH R
Sparger
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Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.
This reference describes the NRC-sponsored tests that were run using the APEX facility during the 2003-2004 time-period. Limited dimensional information is provided.
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