ML20086M469
| ML20086M469 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/12/1991 |
| From: | Broughton T GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-91-2129, NUDOCS 9112170473 | |
| Download: ML20086M469 (10) | |
Text
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o GPU Nuclear Corporation
-,1 Nuclear
- er.rareo o
Middletown, Pennsylvania 17o57 o191 717 944 7621 TELEX 84 2386 Writer's Direct Dial fiumber:
(717) 948-8005 Decent >cr 12, 1991 C311-91-2129 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
.Three Mile Island Nuclear Station, Unit 1 (THI-1)
Operating License No. DPR-50
-Docket No. 50-289 Response to NRC Request for Additional Informatiori Regarding the Second Ten Year Inservice Inspection (ISI) Interval The GPU Nuclear Inservice Inspection (ISI) Program Plan for The second ten-year interval was submitted to the NRC on April 19, 1991. Attachment 1 provides the GPU Nuclear response to the NRC request for additional information or clarification dated September 19,-1991 which was required by the NRC Staff in order to complete the Program Plan review.
The ASME Section XI Code allows exemption of certain piping sections from the Code test requirements where testing cannot be performed due to design limitations. Attachment 2 documents an interpretation of the Code regarding a test performed during the 9R Outage where Code requirements aie not entirely clear.
Sincerely, h
T. G. Brod hton Vice President and Director, TMI-l MRX/mkk Attachments cc:
Region 1 Administrator TMI-1 Senior Project Manager TMI Senior Resident Inspector EG&G 91121704'/3 9112i2 _
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.k C311-91-2129 Page 1 ATTACHMENT 1 GPU Nuclear Response to NRC's Request for Additional Information Regarding THI-l's Second Ten-Year Inservice Inspection (ISI) Interval GPU Nuclear responses to the specific information requests are presented below following a reprint of the NRC request:
NRC Reauest A:
Paragraph 10 CFR 50.55a(b)(2)(iv) requires that ASME Code Class 2 piping welds in the Residual Heat Removal (RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) systems shall be examined.
These systems should not be completely exempted from inservice volumetric examination based on Section XI-exclusion criteria. The staff has previously determined that a 7.5%
augmented volumetric sample constitutes an acceptable resolution at similar plants.
It appears that the Reactor Building Spray (RBS) System is completely excluded from Class 2 piping weld volumetric examinations based on pipe wall thickness.
The ISI Program Plan should be revised to include volumetric examination of a representative sample of welds for the RBS System. This weld sample should be taken from the discharge of the RBS pumps to the first weld downstream of the last normally closed valve (BS-VIA/B). Verify that at least a 7.5% sampling of the Class 2 piping welds in the RBS System will be performed.
GPU Nuclear Response to NRC Reauest A:
Paragraph 10 CFR 50.55a(b)(2)(iv) states:
~
" Pressure-retaining welds in ASME Code Class 2 piping (applies to Tables IWC-2520 or IWC-2520-1, Category C-F).
(A) Appropriate Code Class 2 pipe welds in Residual Heat Removal Systems, Emergency Core Cooling Systems and Containment !! eat Removal Systems, shall be examined. When applying editions and addenda up to the 1983 Edition through the Summer 1983 Addenda of Section XI of the ASME Code, the extent of examination for these systems shall be determined by the requirements of paragraph IWC-1220, Table IWC-2520 Category C-F and C-G, and paragraph IWC-2411 in the 1974 Edition and Addenda through the Summer 1975 Addenda."
Because the 1986 Edition of the Section XI Code is not specifically referenced by the foregoing paragraph (i.e., up to the... Summer 1983 Addenda of Section XI...), GPU Nuclear interprets this to mean that the 1986 Edition is endorsed by the NRC for compliance without the exceptions stated in the regulations.
The 1986 Edition does not allow the exemption of certain Class 2 RHR, ECC, or CHR piping. The fact that the 1986 Code does not allow credit for examination of many of these welds is the basis for the GPU Nuclear Relief Request No. 4.
In addition to the examinations described in Relief Request 4 of the April 19, 1991 submittal, additional Class 2 piping identified in the Program
Page 2 Plan as " Class 2 where the piping wall thickness does not meet the minimum requirements for examination in Table IWC-2500-1, Category C-F-1" will be examined as follows:
S.y.1.tm prawing Pemarks Decay Peat ID-ISI-FD-005 Expanded to include the 6" and 8" lines Removal off 011-T1 (Drawing Coordinate G-4) and the 6" and 10" lines upstream of MU-V14A/B (Drawing Coordinate C-3).
The areas identified in the previous submittal remain in effect.
Reactor ID-ISI-FD-012 All piping Building Spray Makeup and ID-ISI-FD-017 All piping Purification Spent Fuel ID-ISI-FD-018 All piping Cooling At least 7.5% of the non-exempt piping welds described above will be examined during the interval.
NRC Recuest Q:
Augmented examinations have been established by the NRC when added assurance of structural reliability is deemed necessary.
Examples of documents that may require augmented examination are:
(1)
Branch Technical Position MEB 3-1, "High Energy fluid Systems, Protection Against Postulated Piping failures in Fluid Systems Outside Containment;"
(2)
Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations;" and (3)
Regulatory Guide 1.14, " Reactor Coolant Pump flywheel Integrity."
Address these and any other augmented examination requirements that may have been incorporated in the Three Mile Island, Unit 1 Second Ten-Year Inservice Ir.spection Program Plan.
C311-91-2129
.Page 3 EU Nucitg_ Relia'11g_to NRC Request B:
In addition to ASME Code Section XI inspections, the following augmented
-inspections are included in the THI-l ISI Program. A separate schedule (TMI-ISl-003) which has been prepared for these inspections is available at the site for NRC review.
(1)
GPU Nuclear will comply with the requirements of Regulatory Guide 1.150, Revision 1 for THI-l Reactor Vessel Welds until such time as a technique is available which has been fully qualified in accordance with ASME Section XI, Appendix VIII. GPU Nuclear intends to utilize an Appendix VIII qualified examination technique for reactor vessel welds when such a technique is commercially available.
(2)
Examinations of the reactor coolant pump motor flywheel assembly are performed in ucordance with Technical Specification 4.2,4 which states:
"The accessible portions of one reactor coolant pump motor flywheel assembly will be ultrasonically inspected within 3% years, two within 6% years, and all four by the end of the 10 yaar inspection interval.
However the U,T.
procedu-
- developmental and will be used only to the extent nat it is shown to be meaningful.
The extent of coverage will be limited to those areas of the flywheel which are accessible without motor disassembly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports will be used."
(3)
Examinations of certain main steam system welds outside of containment are performed in accordance with TMI-l Technical Specification 4.15.
This requires examination of welds MS-0001, MS-0002,- MS-0003 and MS-0004L at 3h year intervals or the nearest refueling outage.
(4)
GPU Nuclear commitments in regard to NUREG 0612 " Control of Heavy Loads,"
include inspections to ensure that all load bearing welds will be
- examined over a normal inservice inspection interval of 10 years using standard ISI techniques for periodic inspection of the head and internals handling fixture (Tripod).
A visual examination of the tripod will be performed each refueling outage prior to use.
Magnetic particle examination of the accessible welds will be performed on approximately 33% of the welds during each inspection period such that 100% of the l
accessible welds will be-examined during the ten year interval.
(5)
Main feedwater welds FW-0034 through FW-0039 are scheduled for examination once every interval because of a postulated break analysis performed for these weld locations. These examinations assure the section of pipe is in compliance with the breair exclusion requirements as specified in Paragraph B.I.b (7) of Branch Technical Position MEB 3-1.
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C311-91-2129
.Page 4 NRC Reouest C - Reouest for Relief No. 1:
Relief is requested from performing the Code-required surface examination of dissimilar metal and terminal end welds CF-0001, CF-0020, RC-0001, RC-0033, RC-0052, RC-0106, RC-0054, RC-0087, RV-0009BM, and RV-00108M because they are located between the RPV and reactor vessel primary shield wall and are inaccessible from the 0.0. surface.
The Licensee proposes a volumetric examination from the I.D. surface in lieu of the Code-required surface examination.
This proposal could be considered acceptable if the follcwing conditions were met:
(1)
The remote volumetric examination includes the entire weld volume and heat affected zone instead of only the inner one third of the weld as required by the Code.
(2)
The ultrasonic testing instrumentation and procedures are demonstrated to be capable of detecting 0.D. surface-connected defects in the circumferential orientation in a laboratory test block.
The laboratory test blocks should contain crack-like defects and not machined notches.
GPU Nuclear Response to NRC Reouest C:
(l; GPU Nuclear will examine the full weld volume and adjacent base metal P on each side of the weld.
(2)
GPU Nuclear has contracted with Babcock and Wilcox Huclear Service (BWNS) to perform the first period examinations (welds RC-0033 and RC-0054). We will either qualify the BWNS technique on crack-like defects through the B&W Owners Group (BWOG) or as a GPU Nuclear sponsored project if necessary.
GPU Nuclear has not iJsued the contract for third period examinations; however, that examination technique will be qualified on crack-like flaws. The results of these qualifications will be available at the site for NRC review when complete.
NRC Reouest D - Reouest for Relief No. 5:
Relief is requested from performing the Code-required volumetric examination of welds OH-03958, DH-0397B, OH-04018,-and DH-04038 on Decay Heat Removal Coolers DH-C-1A and DH-C-1B because of geometric and material property considerations (304 series stainless steel).
Provide an estimate of the percentage of the Code-required examination that can be completed on each of the nozzle-to-vessel welds and nozzle inner radii for which relief is requested along with sketches showing the configuration of the subject areas. Has GPU Nuclear considered augmenting the limited ultrasonic examination with an I.D. surface examination if the coolers are disassembled for maintenance? Please confirm that the correct Examination Category is C-B and not C-D as submitted.
C311-91-2129 Page 5 GPU Nuclear ResDonse to NR0_.Rp.na1LD:
GPU Nuclear agrees that the Excmination Category for this weld is C-B.
This was a typographical error in the original submittal.
The nozzle design is shown in detail "B" of drawing B-255.33-A previously submitted.
The most likely ultrasonic examination technique would utilize a longitudinal and/or surface wave technique from the shell inside surface.
A straight beam (0*) and surface wave examination would also b3 required from the nozzle inside surface.
This approach would theoretically provide 100% examination if the weld material does not greatly redirect or disperse the sound beam and search unit skew angles are precisely controlled. The extent of sound beam redirection cannot be determined until the examination is in process.
These effects could be overcome with an automated data acquisition and analysis system including a remote operated manipulator.
However, GPU Nuclear does not consider this technology to be reasonably available for such a specific nozzle design.
Significant lead time would be required for technique development and personnel training.
It is expected that there would be insufficient notification time, in the event that cooler maintenance were to becom necessary, and this technology would not be practical for use.
The coolers have not been disassembled since the start of plant operatior. so no previous dose rate history is available.
GPU Nuclear estimates a dose rate of 16R/hr would be encountered in the cooler interior.
A manual ultrasonic examination is expected to require about two hours of scanning time inside the cooler. A liquid penetrant examination of the area of interest would require about ten minutes of time in the cooler.
Liquid penetrant examinations would provide.a more reliable and sensitive examination for inside surface initiated flaws than would the ultrasonic examination. Any ultrasonic examination indications in this area wnuld likely be confirmed with either a liquid penetrant or eddy current examination. GPU Nuclear believes the overall level of plant safety will be increased by performing the proposed alternate liquid penetrant examination in lieu of the code required ultrasonic examination.
NRC Recuest E - Reauest for Relief No. Q:
Relief is requested from performing 100% of the Code-required veiun t-ic examination of welds DH-0399 and DH-004 on Decay Heat Removal C. A 7H-C-1A-and DH-C-1B because of restricted access to the weld due to a sliding witing fl ange.
Provide an estimate of the percentage of the Code-required e:. d nation that could be completed on each of the welds for which relief is requestJ, Olscuss the possibility of an ultrasonic examination using multiple V-paths.
Please provide a sketch of the relative weld location with the sliding flange l
bolted into position.
i
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C311-91-2129 Page 6 i
i Ofj(L Nuclear Response to NRC Re_ quest E:
The configuration design is in detail "A" of drawing B-25533-A (enclosed with GPU Nuclear's April 19, 1991 submittal).
Figure 1 (Attached) is a 1 to 1 scale of the weld and sliding flange in both the installed and disassembled conditions.
The figure shows that there is no scan su-face available for ultrasonic examination (UT) while the cooler is in the assembled condition.
The figure also shows the sliding collar in a position for maximum outside surface access with the coolar disassembled.
Even with the cooler disassembled, insufficient Outer Diameter (0D) scan surface is available to perform a UT examination.
If the cooler is required to be disassembled for maintenance, UT of the weld from the end surfacc or inside surface may be possible.
GPU Nuclear estimates the dose rates to be approximately 16R/hr at the inside surface of the cooler.
GPU Nuclear estimates that a liquid penetrant examination of the inside surface would require only 10 to 15 minutes at the weld as compared with an estimated 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of personnel time at the weld for UT.
Liquid penetrant examination of the OD surface would result in minimal personnel exposure compared to examinations on-the inside or end surfaces.
The cooler is designed for the maximum allowable working pressure and temperature (505 psig at 250af).
The weld was 100%'radiographed during fabrication.
Therefore, unacceptable fabrication related flaws should not exist. The borated water storage tank (BWST) provides a constant 40 psig static head to the cooler while in the standby mode. Throughwall leakage, if not detected during the scheduled VT-2 examinations, would be detected by an increase in activity along with an increase in auxiliary building sump level or a decrease in BWST level.
Inservice induced flaws would originate from the inside or outside surface of the weld.
Liquid penetrant examinations would provide a more reliable and sensitive examination for surface propagating flaws than would UT. If the l
cooler is disassembled, a liquid penetrant examination of the weld will be performed on both the inside and outside surface. A VT-2 examination, with emphasis on this area, will be performed each period whether the cooler is disassembled or not.
Use of the alternative exami:ation would not result in a decrease in the overali level of plant safety for the following reasons.
Disassembly of the cooler only for inspection purposes would increase the likelihood.of leakage from the bolted connections.
Sufficient monitoring is in place to detect throughwall leakage if it were to occur.
If the cooler is disassembled for i
maintenance, liquid penetrant would be performed and would provide adequate sensitivity to detect service induced flaws.
L l
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C311-91-2129
.Page 7 HR[ lqquest F,Reauest for Relief Nomild:
Relief is requested from performing the Code-required hydrostatic pressure test on piping between valves RC-V23 and RC-V4. Discuss the alternative examination to be performed on this section of piping between (committed to during a conference call with NRC, INEL and IMI-l on 6/27/91).
GPU Nuclear Response to NRC Reouest F:
Check valve RC-V23 is on the Auxiliary Spray Line, the first valve from the Reactor Coolant System (RCS).
The piping between RC-V23 and motor operated valve (MOV) RC-V4 (approximately 4' of 1 1/2" diameter piping) does not contain a pressure tap connection. As stated in our relief request No 16.4, THI-l will perfonn a Vf-2 examination of the piping between RC-V23 and RC-V4 with the RCS at 22200 psig' (1.02 times the normal operating pressure of the RCS). The VT-2 i
examination will be performed during each plant startup at hot shutdown conditions in accordance with SP 1303-8.1, " Reactor Coolant System."
If RC-V23 and RC-V4 both have some quantity of seat leakage and RC-V4 leaks less than RC-V23, the piping between the two valves will be pressurized by the RCS.
If RC-V23 has zero seat leakage, the piping between RC-V23 and RC-V4 will not be pressurized by the RCS. However, in addition to SP 1303-8.1, a system l
operational leakage test along with a VT-2 examination will be performed on the auxiliary presstrizer spray line including the piping from RC-V23 to RC-V4.
This VT-2 examination will be performed in accordance with ASME Section XI Table IWC-2500, Item No. C7.30, during. plant shutdown when auxiliary pressurizer spray is in ope,ation.
The system operational leakage test will be l
performed when RCS pressure and the total dynamic head of the inservice Low Pressure Injection (LPI) pump provides 2300 psig at the discharge pressure gauge of the operating LPI pump. This pressure should be-sufficient to observe any _ leakage from the piping between RC-V23 and RC-V4.
l l
l
' Code Case N-498, if approved by the NRC, will allo $r this examination to be performed at nonnal operating pressure (approximately 2155 psig).
1
lhe ASME Section XI Code allows exemption of certain piping sections from the Code test requirements where testing cannot be performed due to design limitations. The following is provided in order to document an interpretation or clarification for a situation where the Code is not entirely clear.
System:
((uclear Se" vices Riter Water _ Sy1Les Affected Piping: Nuclear Service Heat Exchangers discharge and Intermediate Service Coolers discharge to NR-V18 and NR-Vl9.
Specifically, the 24" and 12" piping from fR-V12A/B/C/D, NR-V16A/B/C/0, NR-VISA /B and NR-V14A/B to NR-V18 and NR-V19 ISI Class:
3 ISI Drawing Number:
10-ISI-fD-002 Design Prers:
80 psig Code Required Test Pressu e:
88 psig Code Requirement: The 198E Edition of ASME Section XI Table IWD-2500-1, item No. 02.10 requires a hydrostatic test every ten (10) years.
IWD-5223(d) states, "For open-ended portions of discharge lines beyond the last shutoff valve in non-closed systems (e.g., service water systems), confirmation of adequate flow during system operation shall be acceptable in lieu of system hydrostatic test."
Issue: During the NR hydrostatic test on October 30, 1991 it was not possible to reach the required test pressure of 88 psig because of seat leakage in NR-718 and/or NR-V19.
Basis for acceptability: NR-V18 and NR-Vl9 are 24" butterfly valves that maintain back pressure on the NR Pumps. By procedure back pressure is maintained > 30 psig to prevent runout of the NR Pum.cs. During normal operation,-two NR Pumps operate continuously.
NR-V18 and NR-V19 do not have a closed operational or clcsed safety function. They are required to be throttled open to reintain flow from the heat exchangers.
NR-V18 and NR-Vl9 are adjusted infrequently to provide NR water from the heat exchangers to the Mechanical Draft Cooling Towers and/or to the intake and Pump Screas House for deice operation in-the winter.
It is noteworthy that the piping from NR-V12A/B/C/D, NR-V16A/8/C/D, NR-VIGA/B, und NR-Vi4A/B t a l'R-V18 and NR-vis will receive a VT-2 leakage examinction each inspection period (three times within the interval) during normal o;eratf on in accordance with ASME Section XI, Table IWD-1500-1, item No. 22. 0.
The purpose of this interpretation /elarificLtlon is to document our interpretation-of the Code in considering NR-V12A/B/C/D, NR-V16A/B/C/0, NR-VISA /B, and NR-V14A/8 as the last shutoff valve referenced in IWD-5223(d)..
Because NR-V18 and NR-Vl9 have no closed safety tr operational closed function they would not normally be censidered shatoff valves.
11 accordance with A5ME Section XI, IWD-5223(d) confirmation of tvequate flow during system operation has been verified. This clarification /interpretativa is sinilar to the IIR-V5 and RR-V6 clarification included in section 4.5 of Bas.is Document IMI-1-ISI-Basis-1 (Page 31.0) from our April 19, 1991 submittal of TMi-1 plens and schedules for the secor.d 120 menth ISI interval wh'ch began or.
April 20, 1991.
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