ML20138G505
| ML20138G505 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/15/1985 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20138G467 | List: |
| References | |
| PSE-SE-Z-023, PSE-SE-Z-23, NUDOCS 8510250490 | |
| Download: ML20138G505 (19) | |
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT i
SAFETY EVALUATION No. PSE-SE-2-023 !
TITLE: SIMPLIFICATION OF TEST NUMBER 32, REACTOR WATER CLEANUP SYSTEM j Date: OCT 151985 !
1 1.0 PUR POS E l
4 The purpose of this Safety Evaluation is to determine if the proposed deletion of the RWCU non-regenerative heat exchanger flow test in the blowdown mode, the bottom head flow rate calibration, and pump available NPSH test from !
the power ascension test program impacts safety systems or involves any unreviewed safety questions.
2 . 0' SCOPE The scope of this Safety Evaluation is to determine the adequacy of the power ascension program with respect to the proposed RWCU test deletions. -
3.0 REFERENCES
- 1. NRC Regulatory Guide 1.68, Revision 2, August 1978
- 2. General Electric Startup Test Specification 23A4137, 4
Revision 0
- 3. FSAR Chapter 14, Section 14.2.12.3.32 4.0 DISCUSSION Regulatory Guide 1.68, Revision 2, Appendix A, paragraph 4.r requires the demonstration of the operability of reactor coolant systen purification and cleanup systems during low power testing. Test Number 32 demonstrates the operation of the RWCU system. Process variables are recorded with the reactor at rated temperature and pressure during steady state operation of the RWCU system -
in three modes: hot standby, normal, and blowdown. In addition, the RWCU pump available net positive suction head (NPSH) is determined and the bottom head drain flow calibia' tion is verified during the hot standby mode of operation. It is proposed to delete the RWCU non-re' generative heat exchanger (NRHX) flow test in the PSE-SE-2-023 1 of 4
, n 8510250490 851017 PDR ADOCK 05000354 A , PDR
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blowdown mode, as well as, the bottom head flow indicator calibration from the power ascension testing program, and to perform the RWCU pump NPSH test under cold conditions during preoperational testing. It is also proposed to perform the NRHX flow test in the normal mode during Test Condition 1 instead of during Test Condition Heatup.
The present test procedure specifies that process variables will be measured and compared to acceptance criteria which define required system perfo mance. These criteria are provided in GE Test Specifica; ion 23A4137, Revision 0. They require that the temperature at the tube side outlet of the NRHX not exceed specified limits when the RWCU is in the blowdown or normal mode of
- operation. Also, the outlet temperature and cooling water supply of the NRHX shall be within specified limits. Further, the RWCU pump vibration in any mode and available NPSH during the hot standby mode shall be within specified limits. Finally, bottom head and RWCU flow indications shall agree within specified limits.
The RWCU flow test for the NRHX provides information concerning cooling water flow that will be useful during operation of the system. However, performance of this test in the blowdown mode is not necessary to prove that the heat exchange capabilities of the NRHX meet design.
During the normal operating mode of the RWCU system, temperature and flow measurements will be obtained to demonstrate th*e heat exchange capability of the units.
From these data, the performance of the heat exchangers in the blowdown mode of operation will be shown to be within the limits of flow and temperature imposed by the design. Also, the RWCU flow test for the NRHX will be performed at Test Condition 1 rather than at Test Condition Heatup. Test Condition 1 is a more representative operating condition since feedwater heaters will be in service.
The determination of the RWCU pump available NPSH is currently being performed as part of the system's preoperational test. However, since the system j preoperational test is being performed under cold i conditions, calculational methods will be employed to determine the NPSH at the limiting condition (hot standby) using this cold test data as the basis for extrapolation.
PSE-SE-2-023 2 of 4
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The bottom head flow rate calibration is not required because differences between botton head and RWCU flow indications do not result in any problems in the operating modes of the RWCU system. Because this test and the resulting information do not impact upon any safety conditions, it can be performed at any time following the power ascension test program. Flow rate dif ferences exceeding 25 gpm will be recorded and used as a basis for later calibration.
5.0 CONCLUSION
Performance of the NRHX is adequately demonstrated by a flow test in the normal operating mode during Test Condition 1. Testing of the RWCU pump available NPSH at cold conditions and performing calculations to determine the NPSH at limiting conditions based on the cold test 3
data satisfies one of the objectives of Regulatory Guide 1.68, Appendix A. The dif ference between the bottom head flow and the RWCU flow indication does not impact upon
, any safety conditions. Therefore, the proposed changes and deletions will not adversely af fect any safety systems or the safe operation of the plant and as such do not involve an unreviewed safety question. Regulatory Guide 1.68, Appendix A, paragraph 4.r objectives are still met with the remaining RWCU testing. FSAR Section 14.2.12.3.32 will require revision to reflect the deletion of the bottom head flow calibration.
6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS j Revision to Hope Creek's FSAR and startup test procedures
! shall be made to reflect the simplification of the RWCU l test as described above.
8.0 ATTACHMENTS None PSE-SE-Z-023 3 of 4
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l 9.0 SIGNATURES Originator t 4,2ja? .,/ I /0 [/(,/ 8(
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l Systems Analysis Group Head /1,lb. 1% C M, e IS[$6 Site Engineering Manager ((l! h vk,. 'i hD[" _ / g[ l Dhte '
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? PSE-SE-2-023 4 of 4
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0 ATTACIIMENT 2 l
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i l l l l l l l l l l l l (1) Test conditions refer to plant conditione IT stl lorna lasATI l l l l l l l on ragwe 14.2-4 l mo.l Tas7 ian IvssssLl ur l t 2 l 2 l 4 l 5 l 6 ImamantTl l(223t ! I a l I l l l tal Perform Test 5, timing of 4 elowest control l l l l l l l l l rode, in con)mction with espected scrane 1 I me=1 cal and mediodianical I x x la n lx l lx [x q l
l 2 l andiatton no ar-ent lx lx la l lx l l lx l l ( 3) Dynamic syst. Test case to be completed l 3 l met loadtne lx l l l l l l l l l between test conditions 1 and 3 l 4 l ruit Core Shutdown marsto l l l l l l l l l l l 5 l Control and Draw. l3 lx l1(2) lx(2) lg(2) l l [x(2) l l (4) Af ter recirculatson pump tripe (natural l 6 I siu( rerlormance 1x l l l l l l l l l ctreulation) l 8 l 1831 Performance l l [x l l l l l l l l 9 l LPius Calibration l lx lx l la l l la l l (5) setween so and 9o percent thermal power, i" - ^-
l 11 l Process Computer
- : r M M l- : :: lr l l aus near 100 sorcent core flow lx lx lx(38 l la l lx l l l l 12 l acIC l l3 lI l l l l l l l (6) atan rw munaut Capability 6 pacirc Pump i 13 l MPCI l l l l la l l l l l ambeck must have already been camp 1sted l 14 l Selected Process Tap g lx l l lx lx(4I l lx(4) l l l 14 l Water Invol Ref lag Temp l lx l l lx l l lx l l (7) Reactor power between So and 90 percent l 15 l System supansion lx lx la l lI l l lx l l l 16 l TIP tancertainty l l l l lx l l [x l l (9) Reactor peer between 45 and 65 percent l 17 l Core performance l l lx la lx la lx lx l x l 1 t3 l cteam Production l l l l l l l l l l (9) a actor power between 75 and so perc.nt l 19 l Core Pur-void node nossonaal l l l l lx la l l l l 20 l Prenew magulator l l la lx la la la la l l (to) At maximum sewer that will not casse scram i at l read Sye-set.otat Change. l lx lx lx lx lx la lx l l l 21 l read Sye-Ices FM Meating l l l l l l l lx(5) l l (11) Perform between test conditions 1 and 3 l 21 l Feedwater Pump Trip l l l l l l l lx(6) l l l 21 l Man FW Ihmiout capability l l l l l l l l3(73 l l '"'"'^r---'a " * * - - - "
l 22 l Turbine Velve Swwe111ance l l l l lx(8) l lx(9) lx(1olg g l 23 l MSIV Functional Test l l lxIIII M l l fed 884 l l l
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l 23 l nsIv rull reolation l l l l la l l l 24 l melief wives l 25 l Turbine Trip 6 toat l
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l l x(20) > lx(20)l lx( 15 3 lK( 16)lg l
l lx(2o3l lgg gg g g g
g (14) setween teet condistone 2 and 3 l l Reject k a l l l l l l l (15) Generator load rejection, within bypass l l l l 26 l Chutdown outelde CRC l l l lx l l l l l l valve capacity l 27 l Recirculation Flow Control l l l lx(14Il l lx( 18ll l l l 28 l Recirc=Cne Pump Trip l l l l lx l l lx l l (16) Reactor power between 60 and 80 percent l 28 l RPT Trip-Two Puepe l l l l lK(193l l l l l at core flow 195 sorcent - turtene trip l Je l Recirc systen Performance l l l la lx lx l lx l l l 2B l Recirc Pump Runback l l l l lx l l l l l (17) Load rejection l 2e I sectre sys Cavatation l l l l lx l l l l l l 30 l Loss of Offette Pwr l l [x l l l l l l l (le) setween test com11tione 5 and 6 ,
l 31 l Pipe Vibration l lx lx lx lI l l lx l l l 29 l aectre riov Calibration l l l l lx l l la l l (19) >50s so or and _>95 core flow, and performed l 32 I snecu l l lK l l l l l l l before Turbine Trap 6 to.d majection i 33 l man l l l lx l l l lx(2881 l l 34 l Drywelt 6 steen Tunnel l lx lx l lx l l lx l (2o) Check SRV set pointe during major scram l
l l Cooling l l l l l l l l l l tests ope ensta l 25 i Geseous madwast. l l lx l la l l lx l l GE'er nATimo st Arsons l 39 l SACS Performance l l l l lx l l [x l l (21) Performed during cooldown from test plNAL SAFETV ANALY$a$ REPORT l 40 l Confirmatory In-Plant Test l l l j r. l l l l l l condition 6 FSAn 3/7 (22) The test saaeber correlates to rsaa section TEST SCHEDULE AND CONDITIONS 14.2.12.3.e there a to the indicated test number. .
T rieune is a 6 A e to,essee
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- p. Appendix A, Paragraph 2.e - Compliance with Regulatory Guide 1.56, Maintenance of Water Purity in Boiling l Water Reactors, is addressed in Section 1.8.1.56. j 1
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- q. Appendix A, Paragraph 4.m - Following fuel load, there is no planned startup test of the MSIV leak control system. The preoperational test demonstrates the operability of the system at design conditions.
Testing following fuel load does not contribute any additional meaningful data.
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- r. ApN$i'$f DaF.FrW @?Mn%f!.c*k1NiarW" scram testing is not performed because the plant does not have this design feature.
- s. Appendix A, Paragraph 5.n - Although there will be no startup test procedure designated loose parts monitoring, additional data to supplement the preoperational program on loose parts monitoring will be taken as stated in Section 14.2.10.
- t. Appendix A, Paragraph 5.q - There are no startup tests of the failed fuel detection systems. Preoperational testing and periodic surveillance testing after fuel <
load ensure the proper operation of radiation monitoring systems used for isolation signals in case of gross fission product release. Data is recorded from these systems and used as baseline data.
- u. Appendix A, Paragraph 5.s - Although there will be no startup test procedere designated hotwell level control, operation of the hotwell level control system.
will be verified uLing station operating procedures and i monitoring hotwell level durina Phase III stactup ,
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remono Appeg1ix us a n + mrenmeo so "Sune 7;" RM*b A" /kh*'Y C'"*"T*-l A, Paragraph 5.dd - Compliance with Regulatory Guide 1.68.2, Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants, is addressed in Section 1.8.1.68.2.
l 1.8-42 Amendment 3 1
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- w. Appendix A, Paragraph 5.gg - The ATWS subsystems are thoroughly checked out logically and functionally during the preoperational test program, as described in Sections 14.2.12.1.2.c.6, 14.2.12.1.3.c.3, 14.2.12.1.4.c.4, 14.2.12.1.8.c.9, 14.2.12.1.9.c.7, and 14.2.12.1.10.c.4. Portions of ATWS governed by Technical Specifications will be functionally checked just prior to fuel load using station surveillance and calibration procedures. Additionally, the recirculation pump trips (RPT), which are ATWS related, are accomplished during Phase III testing, as discussed in Section 14.2.12.3.28.c.
- x. Appendix A, Paragraph 5.LL - Hope Creek design does not incorporate the recirculation flow control valve; however, the runback of the reactor recirculation pumps for cavitation protection and loss of feedwater. pump is accomplished during phase III testing, as discussed in Section 14.2.12.3.28.c.
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' $ rivasoeur, rf kerr rsace, A4y *3*- sue:rnurso fen Nsanoes<wo- op rair 1.8.1.68.1 Conformance to Regulatory Guide 1.68.1, r+< r e r /oo t ;bo **.
Revision 1, January 1977: Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants HCGS complies with the intent of Regulatory Guide 1.68.1. For further discussion of the initial test program, see Section 14.
1.8.1.68.2 Conformance to Regulatory Guide 1.68.2, Revision 1, July 1978: Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants HCGS complies with the intent of Regulatory Guide 1.68.2 For further discussion of the initial test program, see Chapter 14.
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1.8-42a Amendment 3 1
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HCGS FSAR 3.9.2.1.2 Piping Vibration 3.9.2.1.2.1 Preoperational and Startup Vibration Testing of Recirculation Piping :
The purpose of the preoperational vibration test phase is to verify that operating vibrations in the recirculation piping and residual heat removal (RHR) suction piping are within acceptable limits. This phase of the test uses visual observation to supplement remote measurements. If, during steady-state operation, visual observation indicates that vibration Ts significant, measurements are made with a hand-held vibrograph.
s Visual observation and manual and remote measurements are made during the following steady-state conditions:
- a. Minimum flow
- b. 50% of rated flow
- c. 75% of rated flow
- d. 100% of rated flow
- e. RHR suction piping at 100% of rated flow in the shutdown cooling mode.
3.9.2.1.2.2 Preoperational Vibration Testing of Small Attached Piping During visual observation of each of the above test conditions,
- a. through e., attention is given to small attached piping and instrument connections to ensure that they are not in resonance I with the recirculation pump motors or flow-induced vibrations.
If the operating vibration acceptance criteria are not met, corrective action such as modification of supports is taken.
3.9.2.1.2.3 Operating Transient Loads on Main Steam-and-Occirculatica Piping l
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HCGS FSAR The purpose of the operating transient test phase is to verify I that pipe stresses are within ASME B&PV Code limits. The !
amplitude of displacements and number of cycles per transient of the main steam and recirculatier piping are measured and ",
displacements compared with acceptance criteria. The deflections are correlated with stresses to verify that'the pipe stresses l remain within ASME B&PV Code limits. Remote vibration and deflection measurements are taken during the following transients:
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- c. Occirculation p;;p Otcrt ,
- b. R00ircul:ti0n ;;;; tri; Ot 1001 Of r;t;d fl0W
{g. Turbine main stop valve closure at 100% power
[. Manual discharge of each safety / relief valve (SRV) at 1000 psig and at planned transient tests that result in SRV discharge.
- T ,
3.9.2.1.3 Dynamic Effects Testing of Main Steam and Recirculation Piping Systeras To verify that snubbers are adequately performing their intended function during the plant operation, a program for dynamic testing as a part of the initial startup operation testing is conducted. The main purpose of this program is to ensure the I
following:
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- a. The vibration levels from the various dynamic loadings during transient and steady-state conditions are below the predetermined acceptable limits.
- b. Long-term fatigue failure does not occur due to underestimating the dynamic effects caused by cyclic loading during plant transient operations.
The purpose of dynamic testing is to account for the acoustic !
wave due to the SRV lift (RV1), SRV loads resulting from air clearing (RV2), and turbine main stop valve closure loads (TSVC).
The maximum stress developed in the piping from the RVI, RV2, and TSVC transients is used as a basis for establishing criteria that 3.9-30
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- 14. .12.3.10 Average Power Range Monitor Calibration a Objective he test objective is to calibrate the APRM.
! b. Pre quisite '
l The cor is in a steady-state condition at the desired power le 1 and core flow rate. Instrumentation used
- to determi e core thermal power has been calibrated.
- c. Test Method A heat balance is aken at selected power levels. Each APRM channel readin is adjusted to agree with the core thermal power as det emined from the heat balance. In addition, the APRM ch nels are calibrated at the frequency required by e Technical Specifications.
- d. Acceptance Criteria Level 1: l 1
- 1. The APRM channels must be librated to read equal to or greater than the actu core thermal power.
- 3. In the startup mode, all APRM chan els must produce a scram at less than or equ to the thermal power setpoint required by t hnical specification.
Level 2:
l With the above criteria met, the APRMs are consi ered accurate if they agree with the heat balance or t e minimum value required based on TPF, MLHGR, and fraction of rated power to within.the limits specif d i in the GE startup test specifications. '
i 14.2-162 Amendment 8
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14.2.12.3.22 -Turbine Valve Surveillance
- a. Objective The test objective is to demonstrate the methods to be used and the maximum power level for routine surveillance testing of the main stop, control, and bypass valves.
- b. Prerequisite The plant has been stabilized at the required power -
level.
- c. Test Method Individual main stop, control., and bypass valves are manually closed and reset at selected power levels.
The response of the reactor is monitored and the maximum power level conditions for the performance of this test are determined. The rate of valve stroking and timing of the closed-open seque.1ce are chosen to minimize the disturbance introduced.
d.. Acceptance Criteria Level 2:
Peak heat flux, vessel pressure, and steam flow shall remain below scram or isolation trip settings by a margin consistent with the GE startup test specification.
14.2.12.3.23 Main Steam Isolation Valves
- a. Objectives
- 1. To functionally check the MSIVs at selected power levels end determine the marieur pe::Or level the-/
ce= be tested :t individ :ll-j 14.2-179 Amendment 8
- 2. To determine isolation valves' closure times.
- b. Prerequisites i The plant has b,een stabilized at the required power level.
- c. Test Method
- 1. Individual closure of each MSIV is performed at selected power levels to verify functional -
performance and to determine closure times. 4*wL
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- 2. A test of the simultaneous full closure of e -
MSIVs is performed at about 100% power. Ope acion of the RCIC and HPCI systems and the relief valves is demonstrated. Reactor parameters are monitored to determine transient behavior of the system during the simultaneous full closure test. The reactor will immediately scram due to the actuation of the MSIV position switches.
Recirculation pumps will trip if Level 2 in the RPV is reached. The feedwater control system will prevent the RPV water level from reaching the steam lines.
- d. Acceptance Criteria Level 1:
- 2. Following the full closure of all MSIVs, vessel pressure and heat flux level shall be as specified in the GE startup test specification.
14.2-180 Amendment 10
- 3. The reactor must immediately scram and the feedwater control system must prevent the water from reaching the main steam lines following full closure of MSIVs from high power.
Level 2:
- 1. Peak neutron flux, vessel pressure, and steam flo.w shall remain below scram or isolation trip settings by a margin consistent with design requirements when individually testing the MSIVs.
- 2. The RCIC and HPCI systems shall function in accordance with the GE startup test specification following the MSIV closure from high power.
14.2.12.3.24 Relief Valves
- a. Objectives
- 1. To demonstrate proper operation of the main steam relief valves and determine their capacity
- 2. To demonstrate their leaktightness following operation.
- b. Prerequisites The reactor is on pressure control with adequate bypass or main steam flow.
- c. Test Method Strae t a so an o 20*/. of R A r+O THeerut.- PO M C- .
A functional test of each safety relief valve (SRV) shall be made er arly la th- etart".;* program *wees i ,-w m,.,*<rmi mw4. 4. -m...ii.. *we << *4-m EEEEhEEbb0hEih TNEEEEE'iE"EhEh'EEp55EEd at rEted reactor pressure, Bypass valves (BPV) response be OR l ;; nit;r d during the 104 prc Cure t00t :nd th:
l electrical output response is monitored during the rated pressure test. The test duration will Le about 1
14.2-181 Amendment 10 :
. l HCGS FSAR 10/84 10 seconds to allow turbine valves and tailpipe sensors to: reach a steady state.
The tailpipe sensor responses will be used to detect '
the opening and subsequent closure of each SRV. The ,
BPV and MWe responseg will be analyzed for anomalies ind Leating a restricq on in an SRV tailpipe.
' O R- gbf
- Valve capacity will be based on certification by ASME code stamp and the applicable documentation being available in the onsite records. Note that the nameplate capacity / pressure rating assumes that the flow is sonic. This will be true if the back pressure is not excessive. A major blockage of the line would -
not necessarily be offset and it should be determined that none exists through the BPV response signatures.
OR. twc, s
) Vendor bench test data of the SRV opening responses will be available onsite for comparison with Section 5.2.2. The acoustic monitoring subsystem will be monitored during the relief valve test program to determine that the setpoints do reflect valve open/ valve closed conditions.
. SRV opening and reclosure setpoint data will be obtained and evaluated during each high power trip test at which an SRV actuation is anticipated,
- d. Acceptance Criteria Level 1:
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- 1. There should be positive indication of steam discharge during the manual actuation of each valve.
- 2. The vendor bench data for SRV capacity is greater than or equal to the values stated in
! Section 5.2.2 and the accident analysis.
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! 14.2-182 Amendment 8
1.' Decay ratio for pressure control variables is as specified in the GE startup test specification.
- 2. The temperature measured by thermocouples on the discharge side of the valves should return to the temperature recorded before the valve was open as required in the GE startup test specification.
The acoustic monitors shall indicate the valve is closed after valve closure.
- 3. During the re+2ced end rated pressure functional tests, steam flow through each relief valve as _
compared to average relief valve flow is as' specified in the GE startup test specification.
., 14.2.12.3.25 Turbine Trip and Generator Load Rejection -
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- a. Objective The test. objective is to demonstrate the proper response of the reactor and its control systems following trips of the turbine and generator.
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- b. Prerequisites Power testing has been completed to the extent necessary for performing this test. The plant is stabilized at the required power level.
- c. Test Method l
! This test is performed at three different power levels l l in the power ascension program. For the turbine trip, i the main generator remains loaded for a time so there is no rise in turbine generator speed, whereas, in the l generator trip, the main generator output breakers open l l
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14.2-183 Amendment 10
- b. Prerequisites The system piping to be tested is supported and restrained properly. Instrumentation for monitoring vibration has been installed and calibrated, where applicable.
- c. Test Method This test is an extension of the preoperational test program. During steady state operation, designated pipes as delineated in Section 3.9.2 will be monitored for vibration. Dynamic vibration measurements will be _
made on applicable piping following various plant and system transients as specified in Sections 3.9.2.1.2.3, 3.9.2.1.3, and 3.9.2.2.4.
- d. Acceptance Criteria Level 1: l The piping displacements at the established locations shall not exceed the limits specified by the piping designer, which are based on not exceeding ASME '
Section III Code stress values or ANSI B31.1 values.
These acceptable vibration levels will be used as acceptance criteria in the appropriate piping vibration startup test procedures.
14.2.12.3.32 Reactor Water Cleanup System l a. Objective j i The test objective is to demonstrate the operation of ,
the RWCU system.
- b. Prerequisites The reactor has been operated at a near rated temperature and pressure long enough to achieve a steady-state condition.
14.2-194 Amendment 8 l
- c. Test Method ano Poa+e. astucra S Aa0 401; OF RATE 0 With the reactor at rated temperature and pressure 4TH *" " N 'As process variables are recorded during steady-state operation in threes cedee of operation of the RWCU
- het-Standby, 2nd Orcel. The 5:tt :
system; blerd_li,ndicater rill be calibrated by *aktag-head drain fic i c drain enly :nd usin; the R4CU-fler fre= the eystee inlet flew bedI?diceter as a etandard te cc ysre 6 \ T HG Noam n u moo &
- d. Acceptance Criteria .
Level 2:
Alop m A L.
'-6-d mode
- 1. The data indicating operation in the l shallbeacceptableasspecifiedbytheGEstartu/p-
- test specification.
-2. Ree:libr:t bette h::d fler indic ter 2;2in:t RSCU flee indicater if the deviatien is greater th:n CE et:: tup t :t Op::ific:ti:::.
f.2. Pump vibration as measured on the bearing housing and coupling end shall be less than or equal to GE startup test specifications. ,
14.2.12.3.33 Residual Heat Removal System
- a. Objectives
- 1. To demonstrate the ability of the RHR system to ,
remove residual and decay heat from the nuclear system, so that refueling and nuclear syst.em servicing can be performed l !
- 2. To demonstrate the capability of the RHR system to [
reduce the suppression pool temperature below the ;
established limit immediately following a blowdown. t i
14.2-195 Amendment 10 l
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.i ATTACIIMENT 3
.