ML20133A401
| ML20133A401 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/25/1985 |
| From: | Rachel Johnson, Lambert C Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20133A355 | List: |
| References | |
| PSE-SE-Z-014, PSE-SE-Z-14, NUDOCS 8510020205 | |
| Download: ML20133A401 (7) | |
Text
_
m PUBLIC SERVICE ELECTRIC AND gas COMPANY HOPE CREEK GENERATING STATION SAFETY EVALUATION NO. PSE-SE-2-014 TITLE:
SIMPLIFICATION OF POWER ASCENSION PEEDWATER SYSTEM RESPONSE TESTING - TEST NUMBER 21A DATE cen...o,ct ubME w
1.0 PUR POS E The purpose of this Safety Evaluation is to document results of an evaluation performed to ensure that the simplification of Hope Creek's Test Number 21A, Feedwater System Response, will not adversely affect reactor safety.
2.0 SCOPE The test specifications are related to the feedwater control system which is a subsystem of the feedwater system.
3.0 R$FERENCES 1.
Regulatory Guide 1.68, Revision 2, August 1978 2.
Hope Creek Final Safety Analysis Report (FSAR)
Chapter 14 3.
General Electric Startup Test Specification, Revision 0
- 23A4137, 4.
Hope Creek Generating Station Draft Technical Specifications 4.0 DISCUSSION Regulatory Guide 1.68 ( Revision 2, Augus t 1978), Appendix A, paragraph 4.u requires the demonstration of of major plant control systems during low the operability power testing; paragraph 5.s requires the calibration and performance verification of feedwater control systems during power ascension testing; and paragraph 5.v requires the verification of feedwater system operation in accordance with design requirements during power ascension testing.
Hope Creek 's Tes t Number 21A, Feedwater System Response, verifies that the feedwater system has been adjusted to PSE-SE-2-014 1 of 3 85100202O5 850930 PDR ADOCK 05000354 A
1 i~
\\
provide acceptable reactor water level control.
Nominal level setpoint changes are used to evaluate the feedwater control system settings for all power and feedwater pump modes.
Performance of this test is currently planned for heatup and Test Conditions 1 through 6.
It is proposed to delete all testing at Test Condition 4 and to delete the Automatic Load Following (ALF) tests.
Deletion of ALF testing is evaluated in Safety Evaluation PSE-SE-2-002; deletion of natural circulation testing (Test Condition 4) is evaluated in Safety Evaluation PS E-S E-Z-0 0 8.
Together these two safety evaluations adequately address the changes proposed for Test Number 21A.
5.0 CONCLUSION
The subject test can be simplified by deleting ALF testing and all testing at Test Condition 4.
All requirements of the test currently planned will be met, and the objectives of Regulatory Guide 1.68 ( Revision 2, Augus t 1978), Appendix A,
paragraphs 4.u, 5.s, and 5.v will be satisfied.
These i
changes do not adversely af fect safety systems or the safe operation of the plant and so they do not involve an-unreviewed safe,ty question.
In ad.dition,.a.Technica.l.
Specificati'ans change'is not re' quired.
The'refore, Hope Creek's Power Ascension Test Number 21A, Feedwater System Response, can be simplified by deleting ALF testing and all testi.ng at Test Condition 4.
6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS Revision to Hope Creek's FSAR and startup test procedures shall be made to reflect the simplifications of Test Number 21A, Feedwater System Response, as discussed above.
8.0 ATTACHMENTS None PSE-SE-2-014 2 of 3
8
.9.0 SIGNATURES Originator
<C Verifier L'!
'L 7
,f.c Group Head (or SSE)/
c7
/j'i
~
Systems Analysis Group H ad A.
D
,k N,L/M /
/35 Site Engineering Manager h
i
[
t Dite ~
PSE-SE-2-014 3 of 3
9 TEST NUMBER 21A - FEEDWATER SYSTEM RESPONSE TEST SIMPLIFICATION - REDUCED-NUMBER OF TESTS OBJECTIVE:
Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 4.u, 5.s, and 5.v require the demonstration of operability during low power testing, the calibration and performance verification during power ascension testing and the verification of operation in accordance with design requirements during power ascension testing of the feedwater control system.
Test Number 21A, c'eedwater System Response, verifies that the feedwater system has been adjusted to provide acceptable reactor water level control.
Nominal water level setpoint changes are used to evaluate the feedwater control system settings for all power and feedwater pump modes.
Performance of this test is currently planned for heatup and Test Conditions 1-6.
It is proposed to delete water level setpoint and manual feedwater, flow tests at Test Condition 4 and to delete Automatic Load Following (ALF) tests.
DISCUSSION:
Acceptance criteria, level 1 and level 2, define acceptable performance of the feedwater control system to testing perturbations.
Leve l 1 and 2 criteria require that the transient response of any level control system-related variable to any test input must be non-divergent.
Testing is performed at all Test Conditions.(1-6) to demonstrate compliance to these criteria.
In addition, testing is performed with the Recirculation Flow Control System in both the manual and ALF modes of operation.
The ALF mode testing and can be deleted since the ALF mode will be disabled.
Current plant Technical Specifications do not allow continued operation at natural circulation flow.
Test Condition 4 defines the region bounded by natural circulation flow near the 100% rod line.
Since this is not an intended mode of operation, testing at Test Condition 4 can be-deleted.
The objectives of Regulatory Guide 1.68, paragraphs 4.u, 5.s, and 5.v are still satisfied with the remaining testing.
1
A CONCLUSION:
The Automatic Load Following mode of operation will be disabled so not testing the Feedwater Control System in the ALF mode does not adversely affect any safety related systems or the safe operation of the plant and as such does not involve an unreviewed safety question.
Plant Technical Specifications do not allow continued operation at natural circulation conditions and therefore Test Condition 4 does not affect any. safety systems or the safe operation of the plant and as such does not involve an unreviewed safety question.
The proposed testing satisfies the objectives of Regulatory Guide 1.68, Appendix A, paragraphs 4.u, 5.s, and 5.v, as well as the requirements of Tes t Number 21A.
Therefore, Test Number 21A, Feedwater System Response, can be simplified by deleting the ALF testing and by deleting testing at Test Condition 4.
2
9 ATTACHMENT 2
?
I l
l l
F i
l j -.
i i
6
..-_-,-.._._.m..--_
,,,,-.-_ _ _ ~,
L e
l l
(
j i
i l
1 1
I i
i i
i i
l (1) Tut conditione refer to plant cordations lTusTI lorsu lasATl l
l l
l 1
l l
on rig we 14.2-4 l
l Mo.l TusT esLns lV:sss2.l UP l 1 l 2 [ 3 l 4 ] 5 l 6 twRRRANTY1 ln2)I l
l j
f l
l l
l (2) perfom Test 5, timing of 4 slowest contro1 1
i l
l 1
i l
l l
rode, in conjunction with expected m: rams 1 1 I cheatcal ans Radiom enical l z 1m lx lx 13 g3
[
j l 2 i Radiation naamurement lx lx lx l
lx l
l lx l
[
(3) Dynamic Systen Test Case to be completed i
j 3 i Fuel Iceling l3 l
l l
l 1-l l
l l
between test condition. I and 3 l
l 4 j Full Core Shutdown Margin lx l
l l
l l
l l
l l
l 5 l Control mod Drive jx lx [1(2) lx(2) g g(2) g g
l3(2) l l
(4) Af ter recirculation pump tripe (natural l 6 l SRM Performance lx l
l l
l l
l l
l
[
circulation)
I e l 1Rn performance j
ix ja l
l l
l l
l l
l 9 i LPIDI Calibration j
l3 ja l
is l
l lI l
l (5) Between 80 and 90 percent thermal power, l 10 i Arius Calibration l
lx lx 3x lx l
lx lx l
l anit near 100 preent core flow l 11 [ Process C=puter lz lx l x( 3) I lx l
tu l
l l
l 12 l RCIC l
lx lx l
l l
l l
l l
(6) elas N Runaut capability & Rectro Pump l 13 l urci l
lx l l
lx l
l l
l l
Runback must have already been cumpleted l 14 i selected proces Temp I
I3 I l
la lx(4) I lx(4) l l
l 14 l water Inv 1 Ret z., Temp l
lx l l
lx l
[
lx l
l (7) Reactor power between e0 and 90 percent i 15 l system expansion lx lx lx j
lx
]
l lx l
l l 16 i Tir thcartainty l
l l
l lx l
l
[x l
l (s) Reactor power between 45 and 65 sorcent i 17 l Core serformance l
l lx la la la lx lx l x i
l is l steam production l
l l
l l
l l
t I :
(9) Reactor power between 75 and 90 percent i 19 I Core ewr-void node Reasoneet l
l l
l t
lx l
l l
l 20 l Prenewe Regulator l
l
[x lx.
lx lx lx l
l (10) At maalmum power that will not cause scram j ai i rems sys-setpoint Chanses l lx lx-lx tu la lx l
.1 l
l zi i re.s sys-toes rw neating l
l l
l l
l 1
lx(5) l l
(11) Perform between test conditione 1 and 3 I
i 21 l readwater Pump Trip g
l l
1 l
l l
lx(6) g<
g l
l 21 l nas Fw amout Capability l
l l
l l
l l
( x(7 )
l (12) Reactor power between 40 and 55 percent l 22 l Turbine valve surveillance l l
l l
lst e) g gg(9) g g( 10 ) l g
g l 23 l MSIV Functional Test l
lx lx(11)lx(12)l l
[I(13)l
[
l (13) Reactor Iower between 60 and 85 percent i
l 23 l MSIV Full Isolation l
l l
l l
1 l
jx l
[
l 24 l Relief volves l
lx lx(20)gs lg(20)g g
lg(20)l l
(14) netween test conditions,2 and 3 l 25 i Turbine Trip a tomt l
l
]
lxt 15) g g(16) g g
g gg gy3 g g
j l Rejection l
l l
l
[
j l
l l
l (15) Cenerat or toed reject ion, within bypass l 26 l shutdown outside CRC l
l l
lx l
l l
l l
l valve capacity 1 27 l Recirculation Flow Control l l
l jz(14)l l
l x( 183 ]
l l
l 28 i Recirc-one Pump Trip l
l l
]
lx l
l 13 l
l (16) Reactor power between 60 and 80 5ercent l 2e l Rpr Trip-ho Pump.
l l
[
j lx(19)l
[
j l
l at core flow t 95 garcent - turtene trip i 28 l Recirc System perfomance l l
l lx
$3 lx l
ja l
l l 28 l Recirc Pump Runback l
l l
lx l
l
[
l l
(178 Load rejection l 2s j Recirc sys cavitation l
l l
l 13 l
l l
l l
l 30 l Loss of of fsite Pwr l
l 13 l
l l
l l
[
]
(191 Detween test consitions 5 and 6 1 31 l Pipe vibration l
lI (I lx
]x l
l lx l
l l 29 l Rectro Flow Calibration
[
[
l l
lx l
j ja l
l (19) >50% power and _)95 core flow, and performed l 32 l RwCU l
tx l l
l l
l l
l l
before Turbine Trip e Iced Rejection l 33 l RMR l
l l
lx l
l l
lI(21)l l
l 34 l Drywell & Steam Tarel l
ix [x l
lx l
l lx l
l (20) Check SRV set pointe during major scram j
l Cooling l
l l
l l
l
[
j j
l teste worsCasEK I 35 l Gaseous Re$ waste l
l lx l
[x l~
l
[x l
g GE NE RATING 5T ATION l 38 l SACS Performance l
l l
l lx i
l lx
]
[
(21) Performed during cooldown from test FINAL SAFETY ANALYS45 REPORT l 40 t Confirmatory I n-P l an t Test l l
l lx l
i I
l l
]
condition 6 rsAR 3/7 (22) The test nuasser correlates to FSAR Section TLST SCHEDULE AND CONDITIONS 14.2.12.3.x d.ere x le the indicated test number.
FIGURE 14 2 6 Anundenent 10. Ob/Bb x
hTE.D TEST PoiQ T