ML20215J044
| ML20215J044 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/15/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NLR-N86145, NUDOCS 8610240270 | |
| Download: ML20215J044 (5) | |
Text
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Pubhc Service Electric and Gas Company Cctbin A. McNeill, Jr.
Public Serv:ce Electric and Gas Company P.O. Box 236.Hancocks Bridge NJ 08038 609 339-4800 Vice Pre 9 dent -
Nuclear October 15, 1986 NLR-N86145 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 CH ANGES HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In accordance with license cendition 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 Saf ety Analysis Report (FSAR).
Attached is a description, justification, summary of the 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 on file provide the basis for this conclusion.
If you have any questions in regard to this matter, please do not' hesitate to contact us.
Sincerely,
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1 Attachment J~
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4 Dr. Thomas E.
Murley 2
10-15-86 C
Mr. D.
H. Wagner USNRC Licensing Project Manager Mr.
R. W.
Borchardt USNRC Senior Resident Inspector s
Mr.J. M. Taylor Director - Inspection and Enforcement l
ATTACHMENT Description of Change This change to FSAR Figure 14.2-5, entitled, " Test Schedule and Conditions," pertains to Startup Test #40, the " Conf irmatory In-Plant Test."
This test was scheduled to be conducted in Test Condition 2.
However, it was not performed in this test condition due to instrumentation malfunctions.
The test will be performed in either Test Conditions 3, 4,
5, or 6.
Bechtel specification, 10855-C-408(O), " Test Plan for In Plant Safety Relief Valve (SRV) Discharge Test," requires that the Confirmatory In-Plant Test be performed during the test progran when reactor power is greater than or equal to 25%.
Therefore, FSAR Figure 14.2-5 has been revised by addina "x"'s to test condition columns 3, 4,
5, and 6 under Test #40 alona with note (27), which states that the test "May be performed anytime reactor power is greater than or equal to 25% during the Startup Program."
These changes are encircled on the attached mark up of FSAR Figure 14.2-5.
Reason for Change The Confirmatory In-Plant' Test (Test #40) was not completed in Test Condition 2 due to instrumentation malfunctions.
These malfunctions were identified in the instrumentation shakedown portion of. Test #40 which is described in FSAR Section 14.2.12.3.40.
10 CFR 50.59 Safety Evaluation Summary The following is a summary of the safety evaluation conducted pursuant to 10 CFR 50.59 for this change.
Paragraph (a)(2) of 10 CFR 50.59 requires that the following 3 questions be responded to in order to determin9 if an unreviewed safety question is involved.
1.
Does the proposed change increase the probability of occurrence or the consequences of an accident or. malfunction of equipment related to safety, as previously evaluated in the PSAR?
No.
The deferral of the Confirmatory In-Plant Test does not alter the original test methodology or requirements.
In addition, the deferral of this test does not result in a challenge to any untested safety equipment.
The evaluation of the Hope Creek primary containment system for the effects of Safety Relief Valve (SRV) discharge loads is documented in the Plant Unique Analysis Report (PUAR), Number BPC-01-300, Revision 0, dated January 1984.
This report was reviewed and approved by the NRC and concludes that sufficient design margins exist in the primary containment system for all design conditions.
Fu rthe rmore, the
analytical methods used in calculating the expected SRV discharge loads have been shown, when used at other similar plants, to predict loads which have been conservatively bounded by the test results at other plants.
Deferring the Confirmatory In-Plant Test dill not alter measured results as the test methodology and requirements do not change.
Therefore, this change does not increase the probability or consequences of an accident as previously evaluated in the FSAR.
2.
Does the proposed change create a possibility for an accident or malf unction of a dif ferent type than any evaluated previously in the FSAR?
No. By deferring the SRV In-Plant Confirmatory Test from Test Condition 2 (TC-2) to a later test condition, the test requirements and/or plant operations remains unchanged.
There fore, this change deferring Test #40 and adding note (27), only clarifies that Test #40 can be performed any time reactor power is greater than or equal to 25% during the Startup Program.
3.
Does the proposed action reduce the margin of safety as defined in the basis for any Technical Specifications?
No.
The proposed change is not considered in the Technical Specifications.
Furthermore, tests performed at other similar plants have confirmed that the design parameters have bor..ded the measured test results.
Therefore, in consideration of these plant similarities, the expected results at Hope Creek will show that the design margin and safety of the plant will be maintained.
Since the response to these three questions is no, the change does not involve an unreviewed safety question.
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- Y TEST CPEN HEAT H) Tm sw. tens We e plant (2H PMormed dunne N N
TEST NAME VESSEL UP 4
5 6
Mmm%eW4 M w h* 6 (2) Perfonn Tem 5. teneng of 4 (22) The test number correlates to 1
Chemicalatuf Radiochemical X
X X
X X
selected control rods. m FSA R S*C8Sn I4 212 3 8 2
Radiaten Measurement X
X X
consuncten enth espected wm a e W WM w g
j 3
F uel toading X
scrams suamber 4
Full Core Shutdown Margm X
(3) Dynam.c Sywsm Tea Cme to (23) May be performed any tane test 5
Control Rod Denne X
X X(2)
Xm Xm i
8 SRM Performance X
g be coneleted between test condetsons peint cond. tons t and 3 8
IRM Performance x
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X 9
LPRM Catbeaten X
X X
X Cu'81*n pu'"O tr*8 (24) RCtC teswg if ret prevously (natural creutation) performed 10 APRM Calibration X
X X
X X
X 11 Process Computer X
XI31 X
X (5) Baw.an EC anc 90 p csat 12 RCIC \\
The c ell s&.fisw.
X X'288 thenna8 po*er. and near 100 13 HPCI l
X X
pecent core no*
14 Selected Process Temp Je a*a s f rstseat Ee j
X X'*8 X
e.. j 14 Water Level Ref Leg Temp X
X X
(61 Mn Fe Rm Wu W pb d g"#-'I l
15 System Espanseon X
X X
X X
have siready tren cocpleted C om ir e nleni, gh 4f,,,,
Recac Pump Runback must 17 Core Performance X
X X
18 Steam Producten X
X X
X (7) Reactor power between 80 and be *= f es 4 co-d.'tsen 90 percent 20 Pressure Regulator X
X X
X X
X I E EsY h oe IF.
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21 Feed Sys - Setonint Changes X
X X
X X
X X
(8) Reactor w between 45 and 21 Feed Sys - Loss FW Heating 65 percent and 75 and 90 percent X'S' Ud 21 Feedwater Pump Trip Xfs (9) Deleted erfoe**J us.M Gen #'E
- 21 Mau FW Runout Capabdity X(7' 22 Turbme Vafve Survesilance X
X"08 gy g
pbhd g e;J d 23 MSIV Functional Test X
XU 'l 23 MSIV Fullisolation oye8EieI*If 2 8 *le X
(11) Perform between test conditions 24 Relief Va'ves X, X 3 XM XM 1 and 3 M*
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25 Turbene Trip and Load Rejection XM56 -
Xf176 26 Shutdown Outside CRC
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28 Recire - One Pump Trip
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X X
28 F;PT Trip - Two Pumps X 0 91 9 et a +e r Mas oc (pal, ' '
28 Recirc system Performance (14) Between test conditons 2 and 3 g g,l 4 X
X X
X 28 Recire FYS. CR."ITAT!CN j
X Q) hg (15) Turbane tre, withie. bypass valve capae t t;y ej eta.
,I 30 Loss of Offsite Par X
31 Pipe Vibraten (16) Deleted X
X X
X X
X 29 Reenc F fon Cal bration i
X X
(17) Load re;ecten 32 RWCU D '23t 33 RHR X(2h Xmi (181 Between test conditens 5 and 6 34 Drywell and Steam Tunnel Coolmg X
X X
X X
35 Gaseous Radweste (19) >50% power and >95 core flow X
X w
X 38 SACS Performance 40 Confermatory in-Plant Test
.X X
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