ML20056F814
| ML20056F814 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/26/1993 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-93-095, CON-NRC-93-95 VPNPD-93-148, NUDOCS 9308310132 | |
| Download: ML20056F814 (6) | |
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Wisconsin Elecinc POATR COMPANY
/31 w Mchgm P0 b 2046. Lwo#m v 532y pu) 2202M5 VPNPD 14 8 NRC 09 5 August 26.,
1993 r
Document Control Desk U.S.
NUCLEAR REGULATORY COMMISSION Mail Station F1-137 Washington, DC 20555 i
Gentlemen:
i DOCKET 50-266 RELIEF REOUEST Ri. 1-12 POINT BEACH NUCLEAR PLANT, UNIT 1 r
By letter dated December 20, 1990 (NRC-90-126), Wisconsin Electric Power Company (WE) submitted the Third 10-Year Interval Inservice Inspection Plan and associated requests for relief for Point Beach Nuclear Plant, Unit 1.
The submittal identified 14 requests for relief from certain requirements of the American Society of Mechanical Engineers f '. 9ME) Boiler and Pressure Vessel Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components."
A safety evaluation of the Third 10-Year Interval Inservice Inspection Plan and associated requests for relief was submitted to WE Dy the Nuclear Regulatory Commission via letter dated August 27, 992.
In this evaluation, it was stated that Request for Relief xR-1-12 is listed as active, but was not provided by WE for review.
Request for Relief RR-1-12 is part of the Third 10-Year Interval Inspection Plan and is submitted separately by this letter for your review. to this letter provides the information nc ded for the NRC to complete a review and approval as required.
Sincerely, y
Bob Link Vice President Nuclear Power
'P/jg 9303310132 930826 a)V f
j Attachment PDR ADOcK 05000266
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PDR Copy to:
NRC Resident Inspector I
g NRC Regional Ac}gipAp&pptp,g,,htti,Jipp III
RR-1-12 EOMPONENT i
Regenerative Heat Exchanger - Primary Side Welds EXAM AREA 1.
RHE Head to Shell 2.
RHE Shell to Tubesheet 3.
RHE Head to Shell 4.
RHE Shell to Tubesheet 5.
RHE Head to Shell 6.
RHE Shell to Tubesheet i
7.
RHE-N1 - Inlet Nozzle to Shell i
8.
RHE-N4 - Shell to Outlet Nozzle 9.
RHE-N5 - Inlet Nozzle to Shell
- 10. RHE-N8 - Shell to Outlet Nozzle t
- 11. RHE-N9 - Inlet Nozzle to Shell
- 12. RHE-N12 - Shell to Outlet Nozzle ISOMETRIC or COMPONENT DRAWING i
Figure 1 - ISI-PRI-1107 ASME SECTION XI CATEGORY i
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ASME SECTION XI ITEM NUMBER I
B2.51 B2.80 B3.150 ASME SECTION XI EXAMINATION REOUIREMENT A volumetric examination of 100% of all three circumferential head welds, three tubesheet to shell. welds and six nozzle to shell welds during the third i
10-year interval.
't ATTACHB. DOC B-20
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i iLTERNATIVE EXAMINATION PBNP proposes to utilize the " multiple stream" concept by performing a volumetric examination of accessible portions of all circumferential head welds, tubesheet to shell welds and nozzle welds equivalent to one of the three identical sections of the regenerative heat exchanger during the third 1
10-year interval.
Specifically, PBNP propose:; to examine to the extent l
practical the circumferential head weld RHE-01, shell to tubesheet weld RHE-02, inlet nozzle to shell weld RHE-N1, and shell to outlet nozzle weld 1
RHE-N4.
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In addition, PBNP proposes to perform a VT-2 visual examination of all regenerative heat exchanger tubesheet and nozzle areas during system leakage tests and hydrostatic pressure tests in accordance with IWA-5000 and Table IWB-2500-1.
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f REASON FCR tIMITATION I
Background
The regenerative heat exchanger (RHE) provides the major single source of radiation exposure accumulated during a normal refueling outage inservice j
inspection.
The RHE is actually three shell and tube heat exchangert t
connected in series.
The RHE is designed to recover heat from the reactor i
coolant system letdown stream during normal operation.
The letdown stream flows through the shell side of the heat exchanger.
The shell side of the RHE is ISI Class I while the tube side is ISI Class 2.
To ensure adequate coverage of the ISI Class I component welds and minimize expose, the muitiple stream concept will be implemented as it is for ISI Class 2 welds.
By extending the multiple stream concept to the ISI Class I welds of the RHE, a good representative sample of the welds will be examined j
while a significant reduction in radiation exposure to tersonnel is achieved.
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The welds that PBNP proposes to examine volumek call) are all located on the i
bottom heat exchanger (see Figure I for an outline of the regenerative heat exchanger that depicts the weld locations).
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The bottom heat exchanger welds should be the ones to be examined for two l
reasons.
First, the bottom heat exchanger operates at the highest temperature of the three and is therefore the most highly stressed. Typical operating temperatures for letdown flow are 538'F into the bottom shell and 252*F out of the top shell.
Second, the bottom heat exchanger welds can 1
genen11y be more extensively examined than the other heat exchanger welds due to ease of access.
This is reflected in the tables contained in Southwest Research Institute letter 17-7472(22), dated January 9, 1985, from Rodney M. Weber (SwRI) to Steve Pullins (WE).
Not only does this letter show the best welds to be examined, it shows that I
some of the welds for which relief is being requested are limited in the amount of area that can be examined.
For example, using the terminology contained in the referenced letter, only 25% of weld RHE-N9 can be examined by OL, 45, and 50 techniques.
ATTACHB. DOC B-21 I
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I Supporting Information Radiation levels Currently, the average dose rates at the regenerative heat exchanger are:
i 1.5 R/Hr general area (at 18")
y 4.0 R/Hr insulation surface (on contact) 7.0 R/Hr shell surface (on contact under the insulation)
Total Estimated Man-Rem Exposure Involved in the Examination Considering the tasks associated with conducting an examination on a particular examination area, the following time intervals have been required in the past:
0.2 Man-Hrs for insulation removal O.1 Man-Hrs for weld cleaning and preparation 0.7 Man-Hrs for conducting the examination 0.2 Man-Hrs for insulation replacement Using the preceding dose rates and times, the following whole body and extremity exposures can be calculated per examination:
Whole Body (using general area dose rates):
0.2 Man / Hrs for insulation removal at 1.5 R/Hr
= 0.3 Mrraan 0.1 Man-Hrs for weld cleaning and preparation at 1.5 R/Hr - 0.15 Mrraan 0.7 Man-Hrs for conducting the examination at 1.5 R/Hr
- 1.C6 Maraan 0.2 Man-Hrs for insulation replacement at 1.5 R/Hr
= 0.3 Mrwaan Total Whole Body Dose per RHE Exam
- 1.8 Mrraan Extremities (hands, using contact dose rates):
0.2 Man-Hrs for insulation removal at 4.0 R/Hr
- 0.8 MrrRan 0.1 Man-Hrs for weld cleaning and preparation at 7.0 R/Hr = 0.7 MarRan 0.7 Man-Hrs for conducting the examination at 7.0 R/Hr
- 4.9 MarRan 0.2 Man-Hrs for insulation replacement at 4.0 R/Hr
- 0.8 Mrwaan Total Extremity Dose per RHE Exam
- 7.2 Mrraan The exposure savings per inspection interval, by a reduction of eight examinations, would be 14.4 Man-Rem Whole Body and 57.6 Man-Rem Extremities.
Shieldinc The general area dose rates are reduced by approximately 50% by placing lead blankets and shields over non-examination areas. However, the highest dose rates are encountered during an inservice inspection. Also, the examiner who is conducting the examination does not have the benefit of the shielding.
ATTACHB. DOC B-22
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Previous InsDection Results Simply stated, all indications recorded during inspections to this point were found to be either insignificant or geometric in nature.
An insignificant indication is either a non-relevant indication or an indication which is i
equal to or greater than the examination recording level but less than the l
evaluation level.
Consecuences of Weld Failure t
The consequences of a weld failure of one of the RHE welds has essentially been addressed in the plant's Final Safety Analysis Report (FSAR).
In the FSAR, to evaluate chemical and volume control system (CVCS) safety, failures or malfunctions were assumed concurrent with a loss of coolant accident (LOCA) and the consequences analyzed.
A LOCA and a concurrent RHE weld t
failure is included in the more general category of a rupture in the CVCS l
line inside containment.
During such an occurrence, the remote-operated valve located near the main coolant loop, upstream of the RHE, is closed on low pressurizer level to prevent supplementary loss of coolant through the letdown line. The RHE would also eventually be isolated, with leakage being confined to the containment, in the case of a weld failure without a LOCA.
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ATTACHB.00C B-23
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