ML17264A962

7 to Hydro Test of Reactor Coolant System.

ML17264A962
Person / Time
Site:	Ginna
Issue date:	05/22/1989
From:	Marlow T ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A963	List: ML17264A962 ML17309A619
References
PT-7, NUDOCS 9707310158
Download: ML17264A962 (52)

Text

(I vP 41 ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION P-/7 7 7 CONTROLLED COPY NUMBER /~yeah'(

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~Q H/W~4 PROCEDURE NOo REV. ZO. ~7 Cg, GP Pc.U '29M ll53 PORC REVIEW DATE Pc.v Pl T'l55 pc. w ir'l T',l 'i 5 7 J=

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P,c.& R'9~ 1'l 4k PLANT SUPERINTENDENT qcu 'Fj ~ ll 45' EPFECTIVE DATE 9707310iSB'970725 PDR ADQCK 05000244 P PDR NON-QA CATEGORY 1.0 aZVXZWED >Y:

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@ +~w- '7-P'T C~ay y-r A VT-2 Visual Examination of the Reactor Coolant zypfzz and items listed herein shall be completed at the pressure indicated. No leakage shall be allowed ~+augh pipe, vessels, or valve bodies Lea3cage through packing, flange gasketing, etc., shall be noted and gmmtified, evaluation by operations or Plant Supervision. If if possible, for test performance is not fox. 10 year hydro, mark N/A in 2352 psig inspection completed column.

INSPECTION 929XZZK 2352 ps'.

RCS Loop A Hot Leg to Steam Ganarator 2 ~ RCS Loop A Hot Leg Safety In]ection Line to V878H 3 ~ RCS Znop A Hat Leg RHR Line to KV 701 4 ~ RCS Inop A Hat Leg Sample Line to V998 5~ RCS Loop B Hot Leg to Steam Generator 6~ -RCS Loop B Hat Leg Safety Infection Line to V878P 7~ RCS Loop B Hot Leg Sample Line to AOV 955 8 ~ RCS Znop B Hot Leg Ctmxging Line to V9315 9~ RCS Loop A Crossaver Pipe

10. Loop A Crossover Pipe Instrumentation Lines

~tt RCS

11. RCS Loop A Crossaver Pipe Drain to V523 and V541

12. RCS Loop A Crossaver Pipe Excess Letdown Line to HCV 123 1 Inop B Crossover Pipe 3 C tp Lt RCS Loop B Crossover Pipe Drain to V540 RCS Loop=8 Crossaver PiPe Letdown Line to AOV 200A, B, 202 17 RCS XDOP A Cold Leg to Reactor Coolant Pump

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RCS Loop A Cold Leg Safety Xn)ection Line to MV 8408, V842 878B

19. RCS Loop A Cold Leg Alternate Charging Line to MV 392B

20. RCS Loop B Cold Leg to Reactor Coolant Puxp Cold Leg Safety In)ection Line to MV 839B, V842A 878D

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21. RCS Loop B 22 RCS Loop B Cold Leg RHR Line to MV 720

'3 RCS Loop B Cold Leg Charging Line to V9314

'4 Pressurizer

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26. Pressurizer Surge Line to RCS Zoop B Hot Leg

27. Pressurizer Spray Line to RCS Loop A Cold Leg

28. Pressurizer Spray Line to RCS Loop B Cold Leg

29. Pressurizer Mini Spray Line araund PCV 431A

30. Pressurizer Hixd. Spray Line around PCV 431B 31.

32 ~

Pressurizer Pressurizer

~liary Spray Line to V9313 Stoaa Space Sample Line to AOV 951 33 ~ Pressurizer Liquid Space Sample Line to MV 953 34 Reactor Vessel to NK 852A

'5, Reactor Vessel to KN 852B

36. Reactor Vessel Head Vent Line to SV 590, 591 37 Reactor Vessel Level Instrumentation Line to V599 ctor Vessel Head Closure Seal Lealcoff Connections

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pd. y+'7 X8 A VT-2 Visual Examinatj.on of the Reactor Coolant, and items 1 isted herein, shal 1 be completed at the pressure indicated No S'r pac~,

e shall be all~ through pipe, vessels, or valve bodies. Leakage through flange gasketing, etc., shall be noted and quantified, if possible, for urther evaluation by Operations or Plant S~~ision. If not for 10 year hydro, mark N/A in 2352 psig inspection completed column.

test performance is INSPECTION QSHIZe 2352 QSSQ

39. Accessible surfaces on the Reactor Vessel Head

40. Reactor Vessel Incore Penetrations in Sump A,

41. Check all accessible ma)or components for leakage (i.e. Reactor Vessel, Steam Generators, etc.) within the Reactor Coolant System Boonhey, which wer~eot mentioned in preceding steps.

REMARKS:

Inspecteti By: @ Date: 5 P I~

2, Level I Inspector Personnel 'perations Date: o 8 /2 Materials Engineering Supervision S'ate:

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Enclosure 3 ISI UT/PT Inspection on Bi-Metalic Wleds 1979 = Penetration 26, 27, 34 1988 = Penetration 16, 32

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ROCHESTER GAS AND EL ~IC GINNA STATION gv w 1979 INSERVICE NONDESTRUC AMINATIONS SUGARY REPORT t Ppl Examination Area Page Examination Procedure Examination Calibration Indications Identification Number Method Number/Revision Sheet Number Sheet Number No Recordable Insignificant Geometric Other Remarks 826-27 & 34 104 UT SWRI 800-57/0 45'-1012 01293 SS Side 45 -1011 01292 45'T-1011 Inc. Side 0010 0005 X 0011 OLS60 0012 Attn PL-FW-XI-PR 105 NDE 100-1 20061 RC-1000-LSW-1-PH 106 VT NDE 100-1 20066 RC-1000k-PR 107 NDE 100-1 20064 AS-1001-23-PR 108 VT NDE 100-1 20067 RC-1000-HSW-1-PR 109 NDE 100-1 20065 PL-FW-XII-PR 110 NDE 100>>1 20060 RC-1001&-PH 111 NDE 100-1 340136 X Pipehanger Half Installed per Drawing QCN $ 14 AS-1001-13-PR 112 NDE 100-1 340168 AS-1001-9-PR-1 113 NDE 100-1 340177 RC-FWW-PH-I 114 NDE 100-1 340147 X Pipehanger not in location shown on Drawing QCNE13 SI-100SKSW-I-PH 115 NDE 100-1 340134 X Pipehanger not built per Drawing QCN815

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Enclosure 4 Ginna Strategic Plan The purpose of the Ginna Strategic plan was to provide crack initiation and growth predictions for Ginna CRDM nozzles and combine the results with a an economic analyses which could. be used as the basis for addressing the CRDM Alloy 600 issue at Ginna Station; The Strategic plan u'ses a combination of Finite element modeling (ANSYS 5.0), Statistical Analysis and a computer program which evaluates alternative strategies to address the CRDM issue, CIRSE (CRDM Nozzle Inspection and Repair Strategic Evaluation).

The finite elements analysis utilizes Ginna specific data supplied by BEW Nuclear Technologies, in 1994. This included Ginna specific list of materials, material certifications, list of welds, welds data-sheets; procedure qualifications, weld wire certifications, applicable fabrication drawings, as-built drawings and contract variations.

The statistical analysis using a Weibull distribution approach is performed using the industry inspection data provided by past inspection experience, including the sample plant inspection which included the same heats of material from the Ginna vessel head. Differences in material susceptibility, stress and temperature are addressed. Note that the analysis conservatively assumes a crack occurs in the sample plant inspected (which contained the same material heat numbers as Ginna) immediately following the inspection.

The CIRSE model utilizes the information developed in the finite element model, the statistical analysis and,others to evaluate the alternative scenarios for responding to the issue of CRDM nozzle cracking. The program input and output are processed using EXCEL. The analysis approach consists of: establishing the input parameters, predicting crack initiation and growth at each planned refueling outage, assessing the total number of nozzles requiring corrective action at each refueling outage, calculating the cost of any planned inspections and remedial measures plus any required repairs at each refueling outage. Present worth of the future costs are also tabulated.

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