ML20215F452
| ML20215F452 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/02/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NLR-N86140, NUDOCS 8610160211 | |
| Download: ML20215F452 (6) | |
Text
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- 3.,. -*
Public Service Electric and Gas Company Corbin A. McNeill, Jr.
Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge,NJ 08038 609339-4800 Vice President -
NuclTar October 2, 1986 NLR-N86140 United States Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406 Attention:
Dr. Thomas E.
- Murley, Regional Administrator Gentlemen:
INITIAL START UP TEST PROGRAM CHANGES HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In accordance with license condition 2.c.10. of Operating License NPF-57 and the provisions of 10 CFR 50.59, Public Service Electric and Gas Company (PSE&G) is submitting 39 copies of the changes made to the Hope Creek Initial Start-up Test Program.
~ This program is described in Chapter 14 of the Final Safety Analysis Report (FSAR).
Attached is a description, justification, 10 CFR 50.59 safety evaluation and an associated marked up FSAR page _ for each change.
Per the requirements of 10 CFR 50.59, paragraph (a)(2), none of these changes involve an unreviewed safety question.
The 10 CFR 50.59 safety evaluations provide the basis for this conclusion.
If you have any questions in regard to this matter, please do not hesitate to conta,ct us.
Sincerely L
8610160211 861002 DR ADOCK 0500 4
i Attachment (g/ 3 e'b L l 4
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-Dr. Thomas-E. Murley 2
10-2-86 C
Mr.
D.
H. Wagner USNRC Licensing Project Manager Mr.'R. W.
Borchardt USNRC Senior Resident Inspector 4
Mr. J.
M. Taylor Director-- Inspection and Enforcement 4
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4 DESCRIPTION OF CHANGE This change to FSAR Figure 14.2-5 pertains to the Shutdown Outside the Control Room Complex Startup Test (Test #26).
An "x" was added to Test Condition column 1.
This indicates that che hot standby demonstration portion of Test #26 was completed in_ Test Condition #1.
REASON FOR CHANGE The hot standby demonstration portion of Test #26 was performed in Test Condition #1 at approximately 20% reactor power rather than Test Condition #2 as Figure 14.2-5 currently indicates.
This change was made for administrative reasons and did not alter the original test intent, conditions, or methodology as proscribed in Regulatory Guide 1.68.2, Revision 1.
10 CFR 50.59 SAFETY EVALUATION Pursuant to 10 CFR 50.59, paragraph (a)(2), the following three questions are responded to in order to determine if an unreviewed safety question is involved in this change.
1.
Does the probability of occurrence, or the consequences of an accident or malfunction of equipment, important to safety previously evaluated in the FSAR increase?
No.
This change does not alter the original test methodology or performance requirements and is in agreement with the requiremen,ts of Regulatory Guide 1.68.2, Revision 1.
There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR..
2.
Is the probability for an accident or malfunction of a different type than any previously evaluated in the FSAR created?
No.
This change does not alter the test methodology.
The change reduces the power level at which the test is initiated by 5% of rated power and is within the 10% to 25%
power range required by Regulatory Guide 1.68.2, Revision 1.
The Startup Test #26 is enveloped by the FSAR Chapter 15 accident analyses.
The probability for an accident or malfunction of a different type than any previously evaluated in the PSAR is not created.
3.
Is the margin of safety as defined in the basis for any Technical Specification reduced?
No.
This change is not related to Technical Specification requirements.
The margin of safety as defined in the basis for any Technical Specification is not reduced.
Since the response to these questions is no, the change does not involve any unreviewed safety questions.
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g. i TEST I2 U M0""*d 8""9 "***
OPEN HEAT TEST NAME 0
VESSEL UP (2) Perform Tem 5. tenmg of 4 (22) The tem number arrelates to l
I -
1 Chemical and Radechemical X
X X
X X
selected control rods. m FSAR Section14212.3 2
Radiation Measurement X
X X
con uncton enh expected whom a a W mW tm s
{
3 Fuel Loading X
4 Full Core Shutdown Margm h
X C) Dynamic System Test Cme to (23) May be perfonnedany time test I
5 Corarol Rod Drive X
X XUI i]4 k
6 SRM Performance X
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conditens 1 and 3 C
XI XCl be completed betwoon iem condatens perma 3
1RM Performance X
X (24) RCIC tesu.4 W utpreviously
- I 9
LPRM Calibratiore X
X X
X tw wcuken pump tnps pwfemed
"'#3*
10 APPM Calibration X
X X
X X
X 11 Proces Computer X
XW X
X
- 4) Between 80 and 90 percent Tke c*Il sidle.em 12 RCIC \\
X XG41 thermal power, and near 100 13 H6C1 X
X percent core flo*
de =*n s f rdin s=ig Ef i
14 Selected Process Terno X
X'il X
(6) Mau FW Runaut Capability and eeIsr = gj g, -
14 Water Level Raf Leg Temp X
X X
Recirc Pump Runback must convenM g9 "
15 System E xpansion X
X X
X X
have siready been completed Jf,,,
17 Core Performance X
X X
X X
X 18 Steam Prcafucten X
(7) Reactor power between 80 and fee s.
fes f cedlSee 90 percent
$ A )aN h oe b.
I 20 Pressure Regulator X
X X
X X
X J J g
(8) Pm m between 45 and
),
21 F eed Sys - Setpoint Changes X
X X
X X
X X
65 percent and 75 and 90 percent i'
21 Feed Sys - Loss FW Heating X s)
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21 Feedwater Pump Trip Xtel (9) Deleted federmeJ 6,M Gen WD I '
21 Max F W Runout Capability Xm h
- WJ at Y N*dej k 7 22 Turbine Valve Surveillance X (81
- Xno,
- 00) At rnadnum poww tMt *M not ji 23 MSIV Funct,onal Test X
XHH ogesKim&I8lf J#%
i 23 MSIV Fult isolation X
(11) Perform between test conditions 24 Relief Valves X, X G01 XIM XQOf l and 3 Naa46e [W.
1 25 Turbine Trip and Load Rejection XHgt XH76
!l 26 Shutdown Outside CRC 3:
X
- 02) Delmed i
27 Recirculaten Fion Control XH81 O
X s1 (13) Deleted 28 Recire - One Pump Tnp X
X 28 RPT Trip - Two Pumps X1191 (14) Between test conditions 2 and 3 28 Recirc System Performance X
X X
X 28 Recirc SYS CAVITATIct; g
X (15) Turbine Enr, withir. bypass valve capaelty 30 Loss of Offsite Pwe y
x (16) Deleted 0
31 Per Vibraten X
X X
X X
X 29 Recirc Flow Calibration X
X (17) Load rejection 32 RWCU y t23) 33 RHR X(E XQU (18) Between test conditions 5 and 6 34 DryweII and Steam Tunnel Cooling X
X X
X X
(19) >5(r% power and >95 core flow 35 Gaseous Radweste X
X X
38 SACS Performance X
X (20) Check SRV operability during 40 Confirmatory In-Plant Test X
X maror Krarn tests g g g,H,0,',E,,N, sy T ion FSAR 3/7 f tN AL SAFETV Astysis nEPORT i
T EST SCHEDULE 8dtD CONDITIO*e5 i
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FIGURE 14 2 5 Amendmem is es T6 e
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DESCRIPTION OF CHANGE This change to FSAR Figure 14.2-5 clarifies when the instrumentation shakedown associated with the Confirmatory In-Plant Test (Test #40) occurred.
This change deletes the "x "
from the Test Condition 1 column for Test #40.
-REASON FOR CHANGE The "x" in the Test Condition 1 (TCl) column for Test #40 was placed there to indicate that the shakedown of ins trume ntation associated with Test #40 would occur in TCl.
The Confirmatory In-Plant Test itself was not_ scheduled during TCl.
The shakedown, used to verify the test setup, was deferred to TC2 due to continuing instrumentation and calibration activity.
10 CFR 50.59 SAFETY EVALUATION
^
Pursuant to 10 CFR 50.59, paragraph (a)(2), the following three questions are responded to in order to determine if an unreviewed safety question is involved in this change.
1.
Does the probability of occurrence, or the consequences of an accident or malfunction of equipment, important to safety previously evaluated in the FSAR increase?
No.
This change does not alter the test methodology or the severity of the reactor transient.
The probability of so, occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR does not increase.
2.
Is the probability for an accident or malfunction of a different type than any previously evaluated in the FSAR created?
No.
The test methodology is not altered.
The shakedown test is less challenging to the plant than t.he confirmatory test that follows.
The probability for an accident or malfunction of a different type than any previously evaluated in the FSAR is not created.
3.
Is the margin of safety as defined in the basis for any Technical Specification reduced?
No.
This change is not related to Technical Specification requirements.
The margin of safety as defined in the basis for any Technical Specification is not reduced.
Since the response to these questions is no, the change does not involve any unreviewed safety questions.
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con 6 tens rve m pimit (211 Performed durmennoidown g. j TEST OPEN HEAT N
TEST NAME 1
2 3
4 5
6 Cond' tens on F @re 14.2 4 from test condetsen 6 VESSR UP (2) Perform Test 5. taming of 4 (22) The test number arrowes to 1
Chemicaland Radiochemical X
X X
X X
selected control rods, in FSAR Secten14.2.12.3;s Wncton with expoeted where s is the inomated test M'
2 Rad.eten 4..asurement X
X X
3 Fuel Loading X
4 Full Core Shutdown Um gin X
(3) Dynamic System Test case to (231 May be performedany tene test h]
5 Control Rod Drive X
XW XW XW be completed beween test
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6 SRM Performance X
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conditons I and 3 '
nndstens perma 8
BRM Performance X
X (24) RCIC tMirg if as previousty A'* C CUI't' n DU"D @8 performed
'r 9
LFRM Calbraten X
X X
X P
(namral circulaton)
,i 10 APRM Calbration X
X X
X X
X 11 Process Computer X
XW X
X (5) Between 80 and 90 percent IA8)r&e e e !J g&.tfe.
12 RCfC \\
X XG4) thermal power, and near 100 Jem.asf rat;n p g Ee 13 HPC3
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X X")
X X
X percent core flo*
14 Selected Process Temp (6) Man FW Runout Capability and ceIer n,gg 4 '
l 14 Water Level Ref Leg Temp
- X X
X Recarc Pump RunDock must 9 "
i 15 System Expansion X
X X
X X
have already been completed c e n e e nien f. 4(f en
,i 17 Core Performance X
X X
X X
X 18 Steam Production X
(7) Reactor power between 60 and from tes e cmd.'Me=
ll, 90 percerit IZhA;her b.
I g 20 Pressure Regulator X
X X
X X
X (8) Reactor power between 45 aM 21 Feed Sys - Setpoint Changes X
X X
X X
X X
65 percent and 75 and 90 percent
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21 Feed Sys - Loss FW Heating X58 Ud 21 Feedwater Pump Trip Xtsi (9) De'eted fedecael su,M Cr ea f'#
j 21 Max FW Runout Capability X 3 gg g
4,,[gj g r;J d 22 Turbme Valve Surveillance X (5 x
oye8 Kip.af ely le *;.
23 MSIV Functonal Test X
X0 13 23 MSiv Full isolaton X
(11) Perform between test conditions 24 Reisef Valves X, XM Xco)
XQ08 1 and 3 ICE 8 b' [8 #
- 25 Turbme Trip end Load Rejection Xng XH 78 II2I O'I'I'd 26 Shutdown Outside CRC
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X 27 Recirculation Flow Controt XU41 XU83 (13) Deleted 28 Recire - One Pump Trip X
X 28 RPT Trip - Two Pumps X119)
(14) Between test conditions 2 and 3 28 Recirc System Penormance X
X X
X 28 Recirc SYS. CN/ITATICN gg X
(15) Turbine tnp, withir. byp As s valve caeacity x *g l
30 Loss of 0ffsite Par Y
(16) De!eted 31 Pee Vibraten X
X X
X X
X 29 Recire Flow Cahbration X
X (17) Load rejection 32 RWCU X123)
,/
33 8t H R X(23)
X'21)
(IS) Between test conditens 5 and 6 34 Drywell and Steam Tunnel Cooling X
X X
X X
(19) >5% mr and >95 core fw 35 Gaseous Radwaste X
X X
38 SACS Performance X
X (20) Checit SRV operability during 40 Confirmatory tr> Plant Test (x-y rna;or scram tests go,, g,,,,
CENER A11NGsf ATION FINAL SAFETY Asiatysis mEPOAT i
TEST SCHEDULE AND CONDITIO*.S a
FIGURE 14 2 5 Ame dmea 15 cle6 I
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