AECM-87-0112, Forwards Revs to Relief Requests I-00004,I-00009 & I-00010 & Relief Requests I-00014 & I-00015.Relief Necessary Due to Exam Limitations Identified in First Refueling Outage at Unit 1.Expeditious Review & Approval Requested.Fee Paid
| ML20215C479 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/12/1987 |
| From: | Kingsley O SYSTEM ENERGY RESOURCES, INC. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| AECM-87-0112, AECM-87-112, NUDOCS 8706180136 | |
| Download: ML20215C479 (81) | |
Text
A SYSTEM ENERGY RESOLIRCES, INC.
OtrAR D KmaEr J7 VCO fWE.Wil t um omawS June 12, 1987 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk Gentlemen:
SUBJECT:
Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Ten-Year Inservice Inspection Plan Relief Request Numbers I-00004, I-00009, I-00010, I-00014 and I-00015 AECM-87/0112 The purpose of this submittal is to provide, in accordance with 10CFR50.55a(g)5(iii), revisions to Relief Requests I-00004, I-00009, and I-00010 submitted by AECM-84/0335 dated June 29, 1984. These relief requests were approved by letter from the NRC to 0. D. Kingsley, dated July 22, 1986 (MAEC-86/0236). The revisions to Relief Requests I-00004, I-00009 and I-00010 are indicated by the clouded portions of the previously approved relief requests. In addition, new Relief Requests I-00014 and I-00015 are attached for the NRC staff's review and approval.
The above relief requests are necessary due to examination limitations identified in the first refueling outage at Grand Gulf Nuclear Station, Unit 1. The impracticality of the ASME Section XI required examinations were demonstrated while performing inservice inspections.
Your expeditious review and approval of the subject relief requests is requested. System Energy Resources, Incorporated is scheduling a refueling outage for the fall of 1987 and requests NRC approval by August 14, 1987 to allow for appropriate outage planning.
If additional information is required to support your review, please contact this office.
Your- ruly, I 8706180136 870612 ,
PDR ADOCK 05000416 ,
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ODK:mbi , l} (
Attachec: 1) $150 application fee in accordance w' h R 0. 1 \
- 2) Relief Requests '
cc: (see next page) ,my y ,a .my, j, , 3.x p j<g ww. : w , us a n i 9 A c c *l Ju:
J18AECM87060101 - I
AECM-87/0112 Page 2 cc: Mr. T. H. Cloninger (w/a)
Mr. R. B. McGehee (w/a)
Mr. N. S. Reynolds (w/a)
.Mr. H. L. Thomas (w/o)
Mr. R. C. Butcher (w/a) 1 Dr. J. Nelson Grace, Regional Administrator (w/a)
U. S. Nuclear Regulatory Commission Region II 101 Marietta St., N. W., Suite 2900 Atlanta, Georgia 30323 I
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b GRAND GULF NUCLEAR STAT 10M INSERVICE INSPECTION REQUIREMENTS O INSERVICE INSPECTION SECTION 4 j
$sNe .
TEN YEAR PROGRAM RELIEF REQUESTS REVISION 2 1 PAGE PAGE 1 of 6 GRAND GULF NUCLEAR STATION UNIT 1
( REQUEST FOR RELIEF NO. I-00004
\
INSERVICE INSPECTION ]
RPV LOWER HEAD-TO-SHELL WELD A-A 1 l
1 I. Component: The lower one-half of Unit I reactor pressure vessel lower head-to-shell weld A-A.
II. Code: The Unit I reactor pressure vessel was designed and fabri-cated to ASME Section III, Class 1 requirements. Applicable inservice inspections are to be performed in accordance with the ASME Section XI, 1977 Edition with addenda through and including Summer 1979 Addenda.
III. Code requirements: The upper portion of this weld is a circumferential shell weld and is required to be volumetrically examined for essentially 100% of the weld length once during the first 10-year inservice inspection interval, in accordance with ASME Section XI, Table IWB-2500-1, Category B-A, Item Bl.11.
The lower portion of this weld is a circumferential head weld and is thus required to be volumetrically examined for essentially 100% of the veld length, once every 10-year in-service inspection interval, in accordance with ASME Section XI, Table IWB-2500-1, Category B-A, Item Bl.21.
l IV. Information to The A-A weld joins the lowest ring of circumferential shell f support the deter- plates on the reactor pressure vessel (RPV) to the RPV bottom I mination that the head and is located 80.66 inches above vessel zero. The l code requirement bottom of the core is located at 216.31 inches above vessel j is impractical: zero. The weld is approximately 135.65 inches below the bottom of the core. 'In addition, the weld is approximately l
' 91.2 inches below the center 11nes of the recirculation pump suction nozzles (N1 nozzles) and approximately 98.4 inchen below the centerlines of the recirculation pump discharge nozzle (jet pump suction nozzles-N2) .
The upper partion of the A-A weld is a typical circumferential
! shell weld. Automated ultrasonic exanination procedure and equipment have been developed which will permit the required inservice volumetric inspections to be performed remotely.
l The lower portion of the A-A weld is a circumferential head l
l weld. Due to the curved geometry of this portion of the weld, j automated means for ultrasonic examination of this portion of I the weld have not been developed; thus, this portion of the j weld must be ultrasonically examined by manual procedure.
l l
l l NWPMSP ISI I-00004
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r GRAND GULF NUCLEAR STATION INSERVICE INSPECTION REQUIREMENTS INSERVICE INSPECTION SECTION 4 EsNe .
TEN YEAR PROGRAM RELIEF REQUESTS REVISION 2 l 1
PAGE PAGE 2 of 6 '
i GRAND GULF NUCLEAR STATION UNIT 1 REQUEST FOR RELIEF MO. I-00004 (Continued)
INSERVICE INSPECTION RPV LOWER HEAD-TO-SHELL WELD A-A IV. Information to The containment design of Grand Gulf Nuclear Station Unit I support the deter- is designated Mark III. A feature of this design is that an mination that the annulus. space of approximately 30 inches width exists between code requirement the reactor vessel outer circumference and the biological is impractical: nhield wall inner circumference. The examiners must enter (continued) this annulus space to perform manual ultrasonic examination of the A-A weld. Contact radiation levels on, and area radiation levels near, the recirculation inlet and outlet nozzles, recorded at other BWR plants during the first three to six years of reactor operation have been in the range of 200 mR/hr to 2000 mR/hr. We anticipate the area radiation levels at the A-A weld to be approximately the same as those near the nozzles due to their proximity to each other, the constricted space in annulus, the proximity to the core, which is treated as an area source, and possible reflection from the metallic insulation on the inner surface of the biological shield wall. For purposes of estimating exposure, we have assumed an area radiation level at the A-A weld of 800 mR/hr at the end of the first 40 month inservice inspec-tion period. The level is expected to increase as the plant ages.
Due to the large amount of weld area required to he examined and the nature of the radiation sources in the area, shielding is not practical. Such nhielding could need to shield the entire body, would be heavy and difficult to move and would require significant exposure to erect and move.
Based on the results of the examinations performed during the preservice inspections, it in entimated that 16 man-hours will be required to perform the required manual ultrasonic examinations. This time does not include the time required for personnel to enter and exit the annulus space; however, it does include an allowance for the extra time required by personnel due to wearing protective clothing and for mapping three recordable, but not reportable indications, which were found during preservice inspections.
We estimate that the manual ultrasonic examination of the A-A weld will require approximately 12,800 millirem of personnel exposure. Entry and exit of the annulus region plus support l personnel exposure during the examinations is estimated to I
require an additional 1,700 millirem for a total estimated L exposure of 14,500 millirem.
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GRAND GULF NUCLEAR STATIOM INSERVICE INSPECTION REQUIREMENTS
. INSERVICE INSPECTION SECTION 4 c TEN YEAR PROGRAM RELIEF REQUESTS REVISION 2 PAGE GRAND GULF NUCLEAR STATION PAGE 3 of 6 UNIT-1 REQUEST FOR RELIEF No I-00004 (Continued) 4 INSERVICE INSPECTION RPV LOWER HEAD-TO-SHELL WELD A-A V. Specific relief Permission is requested to delete all ultrasonic inservice requested: inspections of the lower one-half (Category B-A, Item Bl.21 requirement) of the entire circumference of the A-A weld, except as noted under alternate examinations.
l VI. Reasons why relief Relief from the ultrasonic inservice inspections of the A-A should be granted: weld is requested for the following reasons.
- 1. The upper one-half (Category B-A, Item Bl.11) of the A-A weld was examined by remote ultrasonics as a preservice inspection in accordance with ASME Section XI and no re-cordable indications were found.
- 2. The lower one-half (Category B-A, Item Bl.21) of the A-A weld was examined by manual ultrasonics as a preservice inspection in accordance with ASME Section XI and a total of three recordable, but not reportable, indications were found. The examination report shows that the indications are outside the heat affected zone of the weld. l
- 3. The entire reactor pressure vessel was subjected to a hydrostatic pressure test in accordance with ASME Section III.
- 4. The upper one-half of the A-A weld will be examined by remote ultrasonics during the first inservice inspection interval in accordance with the requirements of ASME Section XI.
- 5. The entire reactor pressure vessel will be subjected to a i system leakage test at each refueling outage and to a system hydrostatic test each inservice inspection interval j in accordance with the requirements of ASME Section XI.
VII. Alternate Testing: Instead of examining the entire lower one-half of the A-A !
weld, we propose to perform manual ultrasonic examinations only of the section of the veld in which the three recordable indications were found (approximately a 12 inch by 12 inch se g , once p inservice inspection
~
interval. d e manual examinations wi1 N formed to the extent possible in consideration with the discussed' limitations. With the vessel support skirt being located 6 inches below the A-A seam, the maximum obtainable "W" NWPMSP ISI l-00004 4 a.- _ aa . ,,m .w .w,. , % ., - ~ ,. _.b.a _-- __ -_ .
I 1 1
GRAND GULF NUCLEAR STATIOM INSERVICE INSPECTION REQUIREMENTS ;
INSERVICE INSPECTION SECTION 4 i Ne .
TEN YEAR PROGRAM RELIEF REQUESTS REVISION 2 PAGE i
GRAND GULF NUCI.F.AP STATION PAGE 4 of 6 '
l UNIT 1 I REQUEST FOR RELIEF NO. I-On004 (Continued)
INSERVICE INSPECTION RPV LOWER-HEAD-TO-SHELL WELD A-A i 1
I VII. Alternate Testing: dimension is also 6 inches, this preverr.s both the 60' (Continued) and 45' T-Scans from examining the total weld volume (see figure 2) . The three recordable indications were i f recorded in the base material of the head utilizing th O' se n. Will monit' r lihe size of the indiEalions and, o
any of the indications appears to_be increasing in size, i we will evaluate the indications and take appropriate actions, which may include ultrasonic examinations of other sections of the veld. The anticipated exposure for performing the alternative examination would be less than i
500 millirem.
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FL L GRAND GULF NUCLEAR STATON INSERVICE INSPECTION REQUIREMENTS t O srstru turnor INSERVICE INSPECTION SECTION 4 REVISION 2 arsouvers, me. TEN YEAR PROGRAM _ RELIEF REQUESTS PAGE GRAND GULF NUCLEAR STATION PAGE 4A of 6 UNIT 1 REQUEST FOR RELIEF NO. I-00004 (Continued)
T INSERVICE INSPECTION RPV LOWER HEAD-TO-SHELL WELD A-A VIII. NRC Discussion The following statements, conclusions, recommendations, Statements etc. have been adopted by the NRC and are to be considered part of this request-for-relief's approval:
Since a 12" by 12" patch of the A-A weld can be examined with manual volumetric methods without excessive radiation exposure and since monitoring an additional 12" by 12" patch where no flaws had occurred prior to PSI would give a measure of the condition of the remainder of the weld, we recommend that an additional ;
12" by 12" patch of the A-A weld be volumetrically examined. The second or reference patch should be at least 90* from the patch containing the flaws.
Therefore, relief is recommended as requested provided:
(a) The upper portion of the A-A weld is volumetrically examined over 100% of the weld length, (b) The manual volumetric examinations, over a 12" by 12" area, of the reportable defects and a reference 'l patch in the lower nortion of the A-A weld are performed and evaluated, and (c) the Code-required nyntem prennure tests are performed.
The requested relief has been recommended based on the impracticality of conducting manual ultrasonic examinations in radiation fields estimated to exist in the examinations area. If actual r.adiation fields are lower, for exangle if the core were removed, more J extensive examinations should be cot;Jucted. g It is further recommended that improvements in autonated ultrasonic inspection technology be monitored and that the entire length of the lower portion of the A-A weld be examined if techniques become available during the 10-year interval.
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F GRAND GULF NUCLEAR STATON INSERVICE INSPECTION REQUITIEMENTS INSERVICE INSPECTION SECTION 4
"',$jjf"" TEN YEAR PROGRAM RELIEF FEQUESTS REVISION 2 I
PAGE 1 PAGE 5 0F 6 l
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l GRAND GULF UNIT ONE BOTTOM HEAD To RING #8 7.6" -
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Percentage of Code Required Volume (CRV) that will be examined 1,eing planned volumetric methods for Seam AAT Upper Portion: 100% UT by remote ultrasonics.
Lower Portion: 1.43% UT by Manual ultrasonics due to perservice of three (3) Recordable Indications.
l NWPMSP ISI I-00004 ,
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GRAND GULF NUCLEAR STATON ggggg INSERVICE INSPECTION SECTION 4
- 73Ne[Nc. TEN. YEAR PROGRAM m agers 7 l PAGE 6 0F 6 WELD A A l
60 N
45 SHELL RING #1/[r-BOTTOM HEAD 9
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WELD AA' SCANNING LIMITATIONS FROM BOTTOM HEAD SIDE l FIGURE 2 o
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c- 1 GRAND GULF NUCLEAR STATION INSERVICE INSPECTION RDQUIREMDTIS INSERVICE INSPECTION SECTION 4 jjjrg,{ [, TEN YEAR PROGRAM RELIEF RDQUEST GRAND GUIE NUC2AR STATION UNIT 1 RELIEF REQUEST NO. I-00009 PAGE 1 of 4 INSERVICE INSPECTION OF PUMP CASING AND ATTACHMDTP WRins I. Camponent: Pump casing and attachment welds located within the surrounding concrete pump support encasement for the following pumps (see attached list and sketches):
HM PUMP NO. SKEIDI NO.
Residual Heat Removal 1E12C002B RH-8-12 Im Pressure Core Spray 1E21C001 LP-9-4 l High Pressure Core Spray 1E22C001 HP-8-10 II. Code: 'Ihe three pumps listed above were designed and fabricated to the ASME Section III, class 2 requirements. Applicable Inservice Inspection is to be preformed in accordarm with ASME Section XI, 1977 Edition __
including Summer 1979 Addenda .
III. Code Recf.lirements: Pressure retaining welds, and attachment welds -
that provide a support function are required to receive a surface examination once every ten-year interval in accordance with ASME Section XI, Table IWC-2500-1 category C-C and C-G.
IV. Information to Inaccessible pump casing welds are located where support the the concrete pump support encasement only allows a deternination that 3 inch clearance between the pump casing and the the code concrete encasement wall (see figure 1 for details requirements are of the design). Due to the limited accessibility, impractical it is impractical to surface examine those portions of the welds located within the suginq concrete _ pump support encret. The 1E12C002B and 1221C001 pudips also have a support integrally welded to the bottom exterior of the pump barrel that rests against the sump floor. The !
clearance between the floor and the bottam of the barrel is approximately 1 inch preventing sufficient access to perform the surface examination of the 1/2 inch of base material on each side of the attaching weld (see figure 2).
I
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' GRAND GULF NUCLEAR STATIOre INSERVICE INSPECTION REQUIREMEtTIS INSERVICE INSPECTION SECTION 4
$8//jegye, TEN YEAR PROGRAM pnm RD2UEST GRAND GUIF NUCLEAR STATION UNIT 1 RELIEF REQUEST NO. I-00009 PAGE 2 of 4 V. Specific relief Permission is requested to exempt fram requested: inservice inspection the inaccessible .<
portions o cas welds sted o Table 1. Also pennission is requ'es to I exempt the base material associated with the
~
support attachment welds from the surface examinations as shown in figure 2.
VI. Reasons why relief Request for exemption should be granted for should be granted: the following reasons:
l
- 1. %e punp casing welds have been volumetrically examined by radiography j and passed in accordance with the ASME '
Section III, Class 2 requirements.
- 2. We attachment welds were surface examined and accepted in accordance with
,' the requirements of ASME Section III, rements.
- 3. W e accessible length of each applicable -
casing weld will be surface examined in accordance with ASME Class 2 requirements.
- 4. The entire weld volume of each support I attachment weld will be surface examined in accordance with ASME Class 2 requirements.
- 5. We failure of these welds, thus leading to failure of the pump, would have no adverse effect on plant safety, as redurrlant emergency core cooling systems are provided.
- 6. Annunciators (i.e. low suction pressure, discharge pressure abnormal, etc.) are provided in the control room, along with other system indicators, to alert the operators to abnormal operating conditions.
- 7. We systems, including the pumps, are tested at least once per 31 days per Operating License Manual (Technical {
Specification) requirements to ensure operability.
1 1
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GRAND CULF NUCLEAR STATiote INSERVICE INSPECTION REQUIREMFRIS INSERVICE INSPECTION SECTION 4 8l"fjcj$[ TEN YEAR PROGRAM RELIEF REQUEST GRAND GUIF IUCLEAR STATICN UNIT 1 RELIEF REQUEST 10. I-00009 PAGE 3 of 4 VI. Reasons why relief 8. Pumps will be subject to a system should be granted pressure test in acconlance with ASME continued: Section XI, Class 2 requirements.
- 9. Approximately 87 percent of the welds on the subject pump, which require surface examination, are accessible. Perfonnance of the required examinations on these accessible welds should ensure that generic degradation is not occurring in these pump casing welds.
VII. Alternate testing: None NOTE: A similar request for relief from preservice inspection of the pump casing welds has been accepted by the IEC in GGNS Safety Evaluation Report, Supplement No. 2.
VIII. NRC Discussion The following statements, conclusions, statements: r - ndations, etc. have been adopted by the IRC and are to be considered part of this request-for-relief's approval.
Since the surface examinations can be conducted from either the external or internal surface of the pump casing, an attempt should be made to examine the portions of the casing welds, inaccessible on the external surface, on an internal surface if the pumps are disassembled for maintenance.
Therefore, relief is r - nded as requested provided:
(a) The surface examinations are performed to the maximum extent practical, (b) the code-required system pressure tests are performed, and (c) the surface examinations are completed from the internal surface if a pump is disassembled for mainte. nance.
RRI-009R
GRAND GULF NUCLEAR STATOM ESNCE ENON WS INSERVICE INSPECTION SECTION 4 llj'"j $ TEN YEAR PROGRAM Pnt:rFF REQUEST GRAND GULF NUCLEAR STATION UNIT 1 PET.TEF RDQUEST NO. I-00009 PAGE 4 of 4 TABIE 1 LIST OF PUMP WEII6 E12 - RHR IUMP "B" CASING Welds Surfaces hat Shall Be Examined W-1 m-4 W-7 m-25 SB-5 W-2 . m-5 m-11 SB-3 SB-6 W-3 m-6 m-12 SB-4 SB-7 Welds hat Can Be Partially Examined Welds that cannot be examined SB-2 (18" accessible, 54" inaccessib SB-1 (inaccessible)
Attachment welds that can be cartially examined SB-12 (see figure 2 for details of limitation)
E21 - LPCS PUMP CASDJG -
Welds Surfaces That Shall Be Examined W-1 m-4 W-7 m-27 SB-5 W-2 m-5 m-11 SB-3 SB-6 W-3 m-6 m-12 SB-4 SB-7 Welds That can Be Partially Examined Welds hat Cannot Be Exanined' SB-2 (3" accessible, 69" inaccessible) SB-1(in e malble)
Attachment Welds hat Can Be Partially Examined SB-12 (see figure 2 for details of limitation) h-E22-HPCS PUMP CASDJG Welds Surfaces Wat Shall Be Examined m-1 m-4 m-7 m-19 SB-6 m-2 m-5 m-11 SH-28 SB-7 W-3 m-6 DK-12 SB-5 Welds hat Can Be Partially Examined Welds That Cannot Be Examined SB-4 (6S" accessible, 4" inaccessible) SB-1 (inaccessible) l
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GRAND GULF NUCLEAR STATON ggg INSERVICE INSPECTION SECTION 4 jjj"jj$[. TEN YEAR PROGRAM PFrm m I
l BOTTOM OF PUMP COLUMN W
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' QRAND GULF NUCLEAR STATION INSERVICE INSPECTION REQUIREMENTS INSEFMCE INSPECTION SECTION 4
,8"8 g $ TEN YEAR PROGRAM RELIEF REQUESTS REVISION 1
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l GRAND GULF NUCLEAR STATION UNIT 1 REQUEST FOR RELIEF NO. I-00010 INSERVICE INSPECTION OF PRESSURE RETAINING WELDS I. Component: Inaccessible portions of Class I and Class II pressure retaining piping welds located on residual heat removal (RHR, E12), reactor core isolation cooling (RCIC, E51), main steam (MS, B21) recirculation (Recirc., B33), and reactor water cleanup (RWCU, G33) and feedwater (FW,B21) systems.
(See Table 1 for details).
II. Code: These portions of the pressure retaining piping welds were designed and fabricated to the ASME Section III, Class 1 and Class 2 requirements. Applicable inservice inspections are to be performed in accoraance with the ASME Section XI, 1977 Edition through and including Summer 1979 Addenda.
III. Code requirements: Class 1 and Class 2 pressure retaining piping welds are required to be volumetrically and surface examined, essentially 100% of the weld, once every ten-year interval in accordance with ASME Section XI, Table IWB-2500-1, Category B-J, Table IWC-2500-1, Category C-F.
IV. Information to Portions of welds that were preservice examined have physical support the deter- obstructions due to design. Due to this limited accessibility, mination that the it is impractical to volumetrically examine 100% of the welds code requirements listed on Table 1.
are impractical:
V. Specific relief Permission is requested to exempt from volumetric examination requested: the inaccessible portions of the Class 1 and Class 2 welds listed on Table 1.
VI. Reasons why relief Request for an exemption should be granted for the following should be granted: reasons:
- 1. The inaccessible portions of listed welds were examined by radiography, passed in accordance with ASME Section III, Class 1 and Class 2 requirements. 1
- 2. The inaccessible portions of listed welds were surface examined (magnetic particle or liquid penetrant), passed i
in accordance with ASME III and/or XI, Class 1 and Class 2 requirements.
- 3. The inaccessible portions of listed piping welds will be subject to a system leakage test after each refueling outage for Class 1, and each inspection period for Class 2 in accordance with ASME Section XI requirements.
hWMbi' lbt 1-UUU1U
GRAND GULF NUCLEAR STATION INSERVICE INSPECTION REQUIREMENTS INSERVICE INSPECTION SECTION 4 l 758"% TEN YEAR PROGRAM RELIEF REQUESTS REVISION 1 GRAND GULF NUCLEAR STATION UNIT 1 REQUEST FOR RELIEF NO. I-00010 (Cont'inued)
INSERVICE INSPECTION l OF PRESSURE RETAINING WELDS VI. Reasons why relief 4. The inaccessible portions of listed piping welds will be should be granted: subject to a system hydrostatic test each inspection (continued) interval in accordance with ASME Section XI, Class 1 and 2 requirements.
- 5. The portions of listed welds inaccessible for volumetric examination will be surface examined each inspection interval, in accordance with ASME Section XI.
i
- 6. Accessible portions of listed welds will be volumetrically and surface examined each inspection interval in accor-dance with ASME Section XI. Should indications be found, an engineering evaluation will be made to determine if the inaccessible portions of listed welds have been affected.
- 7. Leak detection is provided, by way of leakage detection system with continuous monitoring, for the RHR, RCIC, MS RWCU Recire. and feedwater systems.
- 8. The failure of any of these welds would have no adverse effect on plant safety as there is isolation capability and/or shut down capability as part of the plant design.
VII. Alternative All the welds identified in Table I will be inspected twice ,
testing: by volumetric examination during the 10 year interval as !
discussed in NRC in GGNS Safety Evaluation Report, Supplement #2. )
i NWFMSP 151 1-UUULU
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GRAND GULF NUCLEAR STATION j INSERVICE INSPECTION SECTION 4 g8/gj,gyg TEN YEAR PROGRAM Rrr m REQUESP GRAND GULF NUCLEAR STATION UNIT 1 FRr m REQUEST NO. I-00014 PAGE 1 of 16 -
INSERVICE INsm;n0N OF REACIOR PRESSURE VESSEL l RO_ZZLE 'IO SHELL WEIDS l I. Component: Reactor pressure vessel (RPI) to nozzle welds and associated base material, see table 1 for nozzle identification.
II. Code: 'Ihe unit 1 reactor pressure vessel was designed and fabricated to ASME Section III, class 1 requirements. Applicable inservice inspections are to be performed in accordance with Regulatory Guide 1.150 Revision 1 and ASME Section XI,1977 !
Edition with Addenda through and includirg Summer 1979. Also, Relief Request No. I-00013 permits the use of ASME Section XI,1983 Edition with the -
Summer 1983 Adlanda, Figure IWB-2500-7(b) for identifying the Code required examination volume.
III. Code Requirements: Table IWB 2500-1, Examination Category B-D, Full Penetration Welds of Nozzles in Vessels, items i B3.90 and B3.100 requires a volumetric examination of the adjoining weld, base material for 1/2 thickness on each side of the weld, and inner radius (see figure 1).
IV. Information to Of the 31 nozzles requiring a volumetric support the examination of the weld ard adjoiniry base determination that material, 25 are examined usirg automated remote the code techniques, and the remainirs nozzle to vessel requirements are welds are examined manually. 'Ihe automated impractical scanning mechanism presently used at Grand Gulf utilizes two types of scanning packages:
s RRI-0014
GRAND GULF NUCLEAR STATION gg ,
INSERVICE INSPECTION SECTION 4 g8/g y "$. TEN YEAR PROGRAM prim REQUEST GRAND GUIF NUCLEAR STATION UNIT 1 FFr m REQUEST NO. I-00014 PME 2 of 16 IV. Information to 1. The "T-scan" (shear wave sound beam support the transverse to the weld axis) traneAar !
determination that package consists of a O' straight beam, the code 45' and 60* angle beams. The 45' and requirements are 60' angle beam wedges are angulated to impractical: produce a sound beam that is (continued) perpendicular to the weld centerline at the vessel inner surface. One complete revolution of the nozzle scanner, with the T-scan package will scan for parallel oriented reflectors using a 45' and 60* angle beam, and for planar and laminar reflectors using a O' straight beam.
- 2. The "P-scan" (sound beam parallel to the weld axis) transducer package consists of 45* and 60' angle beam wedges. One wedge is pointed in the clockwise direction, and the other is pointed counterclockwise. The 45'and 60* angle beam wedges are angulated to produce a sound beam that is tangent to the weld centerline at the vessel inner surface. Two complete revolutions of the nozzle scanner with the "P-scan" package are performed to scan (fram two directions) for transverse oriented reflectors.
The nozzle design of the BWR 6 does not allow for a full volume examination of the weld and associated 1/2T of base material for the following reasons: l
- 1. Due to the short distance from the weld centerline to the nozzle to shell radius, the examination volume can only be scanned frtan one side (shell side) .
- 2. Also, this short distance prevents extending the scanning arm far enough past the weld towards the nozcle to obtain full cxnerage of the required volume while scanning from the shell side.
l RRI-0014
{
i l
J
l INSERVICE INSPECTION SECI' ION 4 l g8/ g $ , TEN YEAR PROGRAM RELIEF REQUEST GPAND GULF NUCLEAR STATION UNIT 1 RELIEF REQUEST NO. I-00014 PAGE 3 of 16 IV. Information to The six nozzles that require manual examinations support the also have limitations due to nozzle geometry; the determination that code required volume is more than physically the code accessible with known manual techniques. Common to requirements are both techniques and all nozzles is the limitations impractical: due to Near Field effects. Approximately 1/4 inch (continued) of material thickness of the vessel outer surface can not be examined (see figures 2 through 6).
Table 1 provides a detailed listing of information that compares code requirements against the examinations that are achievable. In this evaluation, the Code required volume has been subdivided into the areas recognized by Regulatory Guide 1.150 as being the more critical area and additionally, the weld and heat affected zone coverage has been reported separately. Also, table 1 compares the examination coverage that can be obtained (manually) against what is obtained with automated.
Performing supplemental manual examinations of the accessible volumes would provide very limited increases of total volumes examined. 'Ihe additional volume coverage obtained by supplementing the automated examinations with manual examinations is not justified when compared to the manrem expenditure.
Radiation fields from seven nozzles (3 recirculation, 2 feedwater, 1 core spray, and 1 low pressure core injection) of a 6 year old BWR 5 have been used in evaluating the manrem required to perform supplemental ranual examinations. A comparison of the water chemistry between Grand Gulf arrl the plant where the data was obtained indicates that the radiation levels at Grand Gulf
.may be higher Due to the possibility of additional crud formation. 'Iherefore exposure of examination personnel may also be expected to be greater. The following is an estimate of radiation exposure that an inspection team would receive when performing supplenental manual examinations.
RRI-0014 wwwaew ,.6 w w , u - - -_~ . . .- . ~ . . - - ~~ _m .
i GRAND GULF NUCLEAR SET 10M INSERVICE INSPECI' ION REQUIREMENTS l lNSERVICE INSPECTION SECTION 4 g8/"@ TEN YEAR PROGRAM RELIEF REQUEST GRAND GULF NUCLEAR STATION ,
UNIT 1 l RELIEF REQUEST NO. I-00014 l PAGE 4 of 16 IV. Information to Estimated time spent in radiation area to support the perform a supplemental manual examination determination that is approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per nozzle, per the code team. Radiation dose rate is expected to requirements are average approximately 400 mr/ hour. Thus impractical: to examine seven nozzles (not including (continued) mainsteam) during an outage, one member of the examination team would receive 2,800 mr and the other member 700 mr. The total personnel exposure would be 3.5 man rem for the examination team.
To date, there has been no reported occurrence of cracking at nozzle / vessel weld locations and adjacent areas. It has been generally accepted that the nozzle / vessel welds are not the limiting location with respect to structural integrity.
Cracking has been found at the feedwater nozzle blend radii and bore region at various domestic .
and foreign plants. 'Ihe first cracking reported !
was discovered at the nozzle blend radii as well as the feedwater spargers and brackets and nozzle bore area. Cladding was present on the cracked nozzle, whereas there is no cladding on the Grand Gulf nozzles. 'Ihe presence of the cracking was attributed primarily to rapid cycling of hot and cold water, arxi the presence of cladding. During !
i the inspection which discovered the cracking,-
which was a liluid l penetrant examination, no cracking of the nozzle / weld was observed.This indicated that the feedwater nozzle blend radii ;
was limiting. Since Grand Gulf does not have cladding on the feedwater nozzles, has feedwater flow controller in operation, and triple thermal sleeve, cracking is not expected to occur at the feedwater nozzle blend radii. 'Iherefore, since.the nozzle blend radii is more limiting than the nozzle / vessel weld, cracking at the weld is also less likely.
RRI-0014 .
[ 4.au e.4. J A % k.M.4ebe un e s m.-;,% w.e - ..,. % - -c .. w.ec - - - - -
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GRAND GULF NUCLEAR STATKM g , ,g INSERVICE INSPECTION SECTION 4 g8/g g %, TEN YEAR PROGRAM pn.Tw REQUEST GRAND GULF NUCLEAR STATION UNIT 1 FFTM REQUEST NO. I-00014 PAGE 5 of 16 IV. Information to Since the feedwater nozzle is limiting, the !
support the discussions and conclusions to follow are also '
determination that applicable to all nozzle / vessel welds on the Grand the code Gulf RPV.
reqairements are impractical: Acolied Stress levels 4
]
(continued) '
'Ihe only significant loadings that could affect .
the feedwater nozzle and nozzle / vessel weld area {
are due to internal pressure and thermal cycling i caused by the mixing of hot and cold fluids in the l nozzle bore and blend radii regions. However, the- {
rapid thermal cycling effects have significantly i decreased near the nozzle / vessel weld, and does not produce any significant thermal cycling. This. ;
same thermal cycling was the predominant '
contributor to the observed cracking at the feedwater nozzle blend radii and bore regions of other plants. In addition, the inner cladding which ai'ded the occurrence of cracking is not present on the Grard Gulf nozzle.
Note that the nozzle / vessel weld location is far enough away from the blend such that stress magnification from the geometric discontinuity has reduced significantly.
Faticue Crack Initiation /Procacation
'Ihe occurrence of fatigue crack initiation and subsequent propagation requires the presence of cyclic loading. As stated above in Applied Stress Levels, the significant loadings come from
. internal pressure. As stated earlier, the rapid
'Ihermal mixing experienced at the feedwater nozzle bore and blerd radii has diminished significantly at the nozzle / vessel weld. 'Iherefore, fatigue crack initiation is not likely to occur since the only contributors are pressure stresses and startup/ shutdown thermal gradient stresses and these events are limited in number.
I
.RRI-0014
,w i GRAND GULF NUCLEAR STATION INSERVICE ' INSPECTION SECTION 4 "8/gg%, TEN YEAR PROGRAM pnm REQUEsr GRAND GUIE NUCIEAR STATION UNIT 1 .
RM M REQUEST NO. I-00014 PAGE 6 of 16 I IV. Infomation to his is consistent with the results of the support the feedwater nozzle blend radii Ur results at Grand detemination that Gulf. Since initiation of flaws is not expected, the code propagation is not a4 issue. However, even if a requil.=uia:uits are crack is posbtlated. the predicted crack growth impractical: fran cycling is e t1.
(continued)
Stress Corrosion Crackina Potential To this date, both experimental and field experien has shown no evidence of SCC initiation in A508 m terial. SCC is only possible when an initial significant starter crack is present.
Without this condition, SCC is not a concern in A508 material. Berefore, for the Grand Gulf ,
nozzle / vessel welds, SCC is not a plausible failure mechanism.
Radiation Embrittlement I h e presence of high radiation fluency levels could effect the Nil Ductility Transition
'1bmperature (RrNDr), and therefore cause radiation embrittlement. However, all nozzles are located sufficiently away from the core such that the fluency levels are not high enough to cause material embrittlement. 2 erefore, radiation embrittle m nt is not a concern for the Grand Gulf nozzle / vessel welds. This is consistent with i assumptions made in prior fracture mechanics evaluations performed for the nozzle blend radii cracking. '
, Consecuences of Postulated Crackina In the previous sections, a discussion of various aspects concerning the structural integrity of the nozzle / vessel weld has been presented. In this e section, additional M a'maion demonstrating i significant safety margin for this location even l if an unlikely throughwall crack were to occur.
RRI-0014
$ O
i .
l GRAND GULF NUCLEAR STATioM RS N CE INSPECTION N E ;
INSERVICE INSPECTION SECI' ION 4 37gj7"" TEN YEAR PROGRAM RELTEF REQUEST 4
GRAND GULF NUCLEAR STATION )
UNIT 1 !
RELIEF RD2UEST NO. I-00014 PAGE 7 of 16 IV. Information to As discussed before, the limiting condition with support the respect to potential to failure is the nozzle bore determination that radii and bore areas. The nozzle / vessel weld area the code is bounded by any analysis for the bore or blend requirements are radii areas. Significant analysis has been impractical: perfonned for the bore and blend radii locations.
(continued) Therefore, the discussions provided in this section are a summary of the results determined for the limiting cases. It should be emphasized that a basic assumption in the design and licensing of light water systems in the USA is j that failure of the reactor pressure vessel need '
not be postulated as a design basis event.
Allowable Crack Size The basic requirenent of Section XI of the ASME Code is that flaws greater than 10% of the q critical flaw size is not permitted. Evaluations of a worst case thermal event in combination with pressure stress have shown that the lower bound fracture toughness of the A508 material is not exceeded even for crack depths approaching the i vessel thickness. Since flaw sizes in excess of the wall thickness have no physical significance, it was conservatively assuned that a crack depth equal to the thickness was critical and thus a y Section XI allowable crack depth of 10% of wall thickness was established.
Leak Before Break Regardless of the examination and repair programs in place to ensure that nozzle flaws do not exceed Section XI allowables, it is useful to postulate one or more nozzle flaws becoming very large without being detected in order to determine whether a critical condition could exist and cause rapid crack pmpagation. As mentioned earlier, the stress intensity factor for a crack at the blend radii never exceeded the lower bound fracture toughness even for a crack equal to the thickness of the vessel.
RRI-0014 ,
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i GRAND GULF NUCLEAR STATON INSERVICE INSPECTION REQUIIEMENTS INSERVICE INSPECTION SECTION 4
$$$c5 %. TEN YEAR PROGRAM RELIEF REQUEST 1 GRAND GULF NUCLEAR STATION j UNIT 1 PELTEF REQUEST NO. I-00014 PAGE 8 of 16 1
l l
IV. Infonnation to Evaluations have been performed for an analogous support the case of cracking from a hole in a plate. The determination that stmss intensity factors as a function of through the code wall crack length have been calculated. For large requirements are cracks (crack tip far away from the nozzle), the I impractical: only contributing stress is due to pressure.
(continued) 'Ihe intersection of the fracture toughness and !
stress intensity factor prescribes the critical flaw size. 'Ihe results of these calculations have ;
shown that the critical crack size for each of the i two cracks is 29 inches (total length of 58 l inches). It is virtually inconceivable that the l leakage associated with cracking of this magnitude I could escape detection in an operating BWR. Thus a '
leak before break condition is assured for the :
vessel, even in the unlikely event that nozzle flaws grow to depths greater than the Section XI allowables, as long as the vessel is at upper shelf temperature. Similar conclusions apply to the upset and emergency conditions with postulated crack lengths of 15 inches (30 inches total) and 11 inches (22 inches total), respectively required before rapid fracture could occur under these conditions.
Although the entire ASME Code prescribed area was not ultrasonically examined, the ultrasonic data obtained can still be used to evaluate the potential for cracking. Figures 2 through 6 show typical areas of examination coverage for the' various scanning techniques (table 1 provides i
specific quantities of coverage). It can be seen i that at least one side of the weld and portions thereof are examined. Results of previous examinations revealed no rejectable indications exceeding ASME Section XI criteria. In addition, nozzle blend radii are also examined, and to date no relevant indications have been located. The results of previous Ur examinations.is consistent ,
with the discussion provided in this request for J relief.
i RRI-0014 i a%%%;ansuauM.aarwm .+ w a ~ i ,
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l GRAND GULF NUCLEAR STATION IN N CE INS N ON N '
INSERVICE INSPECTION SECTION 4
- 8'"j f %, TEN YEAR PROGRAM FD M REQUEST (q I
GRAND GUIE NUCLEAR STATION !
l UNIT 1 F PM M REQUEST NO. I-00014 PAGE 9 of 16
.IV. Information to Since no cracking has been found at a nozzle j support the _
blend radius or at the examined nozzle / vessel.
determination that area, it is unlikely that irdications would c" the code be present in the unexamined area between the requirements are examined vessel / nozzle weld and the nozzle impractical: blerd radius. There are no additional (continued). considerations at the unexamined area which would make cracking more likely than at the examined nozzle / vessel weld area. Therefore, it is concluded that examined areas of the g' nozzle / vessel weld area and the nozzle blend n
radii area are sufficient to identify any flaws that may occur.
V. ' Specific relief. Permission is requested to perform ultrasonic requested: examination of only those areas accessible with Automated techniques for those 25 nozzle to vessel u lds that are examined utilizing autatated methods. Also, permission is requested to examine only those areas that are accessible with manual techniques for the six nozzles that require exandnations manually.
VI. Reasons why relief Relief as described within should be granted should be granted: .for the followirq reasons:
- 1. The entire reactor pressure vessel was subjected to an ASME Section III hydrostatic test after fabrication.
- 2. The entire reactor pressure vessel will be subjected to system leakage test at each refueling outage and a system hydrostatic test each inspection interval in accordance with the requirements of ASME Section XI.
+ !
- 3. The subject welds were volumetrically i examined in accordance with ASME Section III durirg fabrication.
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RRI-0014 j
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GRAND GULF NUCLEAR STATION INMG INSNG N
. INSERVICE INSPECTION SECTION 4 g,Z"g % TEN-YEAR PROGRAM pnt:rvP REQUEST GRAND GUIF NUCIEAR' STATION' UNIT 1 FRr m RB20EST NO. I-00014 i
PAGE 10 of 16 l VI. . Reasons why relief 4. 'Ihere is no history of service induced !
should be granted flaws in these areas of the reactor continued: pressure vessel other than those of the feedwater nozzles discussed within this request for relief.
- 5. The areas being examined are the limiting areas of the nozzle to vessel config-uration.
- 6. The performance of additional manual examinations would require significant expenditures of personnel exposure for a small increase of examined volume.
- 7. The potential for initiation and propagation of cracking has been discussed assuming both fatigue and stress corrosion cracking mechanisms. It was concluded by the use of limiting analyses results performa1 for the feedwater nozzle blerd radii, that cracking is unlikely at Grand Gulf nozzle / vessel weld locations. In fact, even if it was hypothesized that these postulated cracks went undetected, a crack length of 58 inches was required before rapid crack growth were to occur during nortal operation. It is unlikely that cracks of this size would go undetected. Therefore, a significant leak .
before break margin exists. !
VII. Alternate testing: None
+
l RRI-0014 j
m4m_m__ m ____ m. . . _ . - _ _ . . __m _mLj
GRAND GULF NUCLEAR STATION pggggg gg INSEfNICE INSPECTION SECTIQi 4 srsrru menor TEN YEAR PROGRAM nesouwets,sw. RELIEF REQUEST GRAND GUIE NUCIZAR SIATIOT UNIT 1 RELIEF REQUEST No. I-00014 i
PAGE 11 of 16 i
D
% l !
I g 1 Il
\\ ,
\ + I n 1 ---->
toj,tn2 =
nozzle wall thickness t, a shell (or headl thickness ,
=
rf nozzle inside radius l I
l *
\
t,/2 t,/2 l
A l
N O , BC D
'd N EL di hw T T '
'[
/ [
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Cladding i r;- 1/2 fn. 4-- En 2 ---->
\ p I l g f
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\ 1
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- -I ,
E xam, vol.
- A-8-C-D-E-F-G-H Corner fles.
EXAMINATION REGION INote (1)!
EXAMINATION VOLUMEINote(2)l Shell(or head) adjoinin0 region Attachment weld region C D-E-F Nonle cylinder region B C F-G Non,e inside corner region A-B G-H M N O-P FIGURE 1 RRI-0014 i
GRAND GULF NUCLEAR STATION gg ggggg INSERVICE INSPECTION SECTION 4 srsrru cursor TEN YEAR PROGRAM prrtrEF Fig;my Mr$00ncr$, INC.
GRAND GULF NUC EAR STATION UNIT 1 Pri m REQUEST No. I-00014 PAGE 12 of 16 I
r-n L_J EXAM VOLUME .
CXAMINED AUTOMATICALLY ,
N0ZZLE f
, m , _ _
l l /
l vtsstL u -l /
'" " h hwid h w, q !
FIGURE 2 Automatic Scan Coverage - 45' Angle Beam (T-Scan)
RRI-0014 i I
~ GRAND GULF NUCLEAQ STRTION INSERVICE INSPECTION SECTI N 4 8l8'gg"" TEN YEAR PROGRAM murg pm GRAND GUIF NUC1 EAR STATION UNIT 1 REIEF REQUEST NO. I-00014 PAGE 13 of 16 s
i l
i r-
', LJ EXAH VOLUME EXAMINED AUTOMATICALLY .
N0ZZLE I
[
I l c .. -
l l /
VCS$[L
/
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FIGURE 3 Autmatic Scan Coverage - 60* Angle Beam (T-Scan) d RRI-0014 1
J
GRAND GULF NUCLEAR STATON gg , ,g INSERVICE INSPECTION SECTION 4 m/"gy", TEN YEAR PROGRAM REIIEF REQUEST GRAND GUIF NUCIEAR STATION UNIT 1 REIJEF REQUEST 10. I-00014 PAGE 14 of 16 r-i L_J EXAH VOLUME EXAMINED AUTOMATICALLY
~
NOZ2LE I l
i 8 ..
r~m 1.
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FIGURE 4 Automatic Scan Coverage - 45' Angle Beam (P-Scan)
RRI-0014
GRAND GULF NUCLEAR STATION g 7ggg INSERVICE INSPECTION SECTI N 4 srsr'" '"'#" TEN YEAR PROGRAM RE30uMCtS, INC. FFTM F6@rm GRAND GUIE NUCIEAR STATIN UNIT 1 FFr m REQUEST NO. I-00014 PAGE 15 of 16 l
7 r-I L_J EXAM VOLUME ,
EXAMINED AUTOMATICALLY '
N0ZZLE f I
I i i r " 'M8 !
l j7 I $ ,/ E
- ^/ -- L VESSCL !
y--., / l
' ' 2 i
H,q h n,q " ' ' "
l
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FIGURE 5 ;
Autmatic Scan Coverage - 60' Argle Beam (P-Scan) l RRI-0014 h
GRAND GULF NUCLEAR STATION g g INSERVICE INSPECTION SECTION 4 gy"g"E TEN YEAR PROGRAM RE'E MUJEsr GRAND GUIF NUCTAR SIATION UNIT 1 PFr.T W Fd)gur M NO. I-00014 PAGE 16 of 16 i
r-1 L.J EXAM VOLUME CXAMINCD AUTOMATICALLY I
NQZZLC f l r"" =
l yy l
l V vcsstL I
L
/ \
p_q pH J
l l
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FIGURE 6 l Autmatic Scan coveage - O' Straight beam RRI-0014 g 9 O
GRAND GULF NUCLEAR STATOW g gggg INSERVICE INSPECTION SECTION 4 "8/gg,g %, TEN YEAR PROGRAM.
pFTTEF REQUEST RELIEF REQUEST No. I 00014 TABLE 1 PAGE 1 N1(A&B) Recirculation Outlet Nortle (Automatic) 79.70 In2 total code volune 2
18.37 In inner 1/4t code volune 2
15 40 In weld volune + HAZ (1/4")
Accessible Accessible Accessible code Volune Imer 1/4t Code Volune WeAd Volune Type Area % Area % Ares %
Sean (In2) examined (in2) examined (in2) examined O' 32.78 76.8% - -
9.79 14.3%
45' T scan 59.61 87.5% '18.37 97.9% 14.97 90.2%
60* T scan - 64.74 92.7% 18.37 100.0% 15.05 95.0%
45' P scan 37.43 83.9% 9.94 85.2% 9.48 35.9%
T. 60' P scan 41.08 93.6% 1*.56 94.9% 13.04 80.5%
l N2(A M) Recirculation Inlet North (Automatic) 2 78.75 In total code volune 2
18.37 In inner 1/4t code volume 2
15.40 In weld volune + HAZ (1/4") )
Accessible Accessible Accessitale Code Volune Imer 1/4t Code Volune Weld Volune Type Area % Area % Area % l Sean (In2 ) examined (in2) examined (in2) examined O' 37.59 71.3% - -
9.81 4.1% i 45' T scan 59.60 84.5% 18.37 %.8% 15.01 87.3% l 60' T scan 65.15 91.0% 18.37 100.0% 15.18 92.2%
45' P scan 37.98 92.2% 10.37 92.6% 9.78 72.6% 1 60' P scan 40.14 100.0% 11.02 100.0% 13.04 93.6%
i Note: Accessible is the volume which can be examined by present manual ultrasonic techniques I i
l RRI 14T1 l
.._________-________-________A
GRAND GULF NUCLEAR STATION gg g_
INSERVICE INSPECTION SECTICH 4 jg gj $ ,
1 TEN YEAR PROGRAM PRT:TEF REQUEST RELIEF REQUEST Wo. I 00014 TABLE 1
- PAGE 2 j N3(A D) Mainsteam Norrie (Automatic) 82.99 In2total code volume ,
2 19.52 In inner 1/4t code volune 2
16.17 In weld volune + HAZ (1/4")
Accessible Accessible Accessible Code Volune Inner 1/4t Code Volune Weld Volune Type Area % Area % Area %
Scan (In2) examined (in2) examined (in2) examined 0* 41.43 75.3% - -
10.33 9.7%
45' T scan 63.30 86.8% 19.52 98.0% 15.70 89.8% ;
60' T scan 68.94 92.4% 19.52 100.0% 15.97 92.0% !
45' P scan 39.58 79.3% 10.65 80.5% 10.16 21.7%
60' P scan 42.53 94.3% 11.71 94.9% 12.68 78.1% I N4fA F) Feedwater No22te (Automatic) 60.16 In2 total code volune 2
14.08 In inner 1/4t code volune 12.70 In# weld volune + HAZ (1/4")
Accessible Accessible Accessible Code Volune Inner 1/4t Code Volune Weld volune Type Area % Area % Area %
Scan (In2) examined (in )2 examined (in2) examined O' 30.14 71.3% - -
8.14 6.5%
45' T-scan 45.49 84.0% 14.08 95.8% 12.24 86.6%
60' T scan 49.53 90.6% 14.08 100.0% 12.26 93.0%
45' P scan 28.08 95.7% 7.39 94.5% 7.48 82.9%
60' P scan 31.31 100.0% 8.49 100.0% 10.52 100.0%
Note: Accessible is the volume which can be examined by present manual ultrasonic techniques 1
RR!a14T1 4
l',
GRAND GULF NUCLEAR STATON ggggg INSERVICE INSPECTioid SECTION 4
. 3r$rtu men . TEN YEAR PROGRAM RC30UMCES. INC.
prrT w REQUESP i
RELIEF REQUEST NO. 1 00014 .]
J TABLE 1 PAGE 3
!LS(A&B) Core Soray No22te (Automatic) 2 60.16 In total code volune 2
14.08'In inner 1/4t code volume 2
12.70'In weld 'volune + HAZ (1/4")
Accessible Accessible Accessible Code Volume Inner 1/4t Code volune Weld volune Type Area % Area % Area %
Scan (In2) examined (in2) examined (in2) examined 0* 30.14 74.0% - -
8.95 8.9%
45' T scan 45.67 85.5% 14.08 %.7% 12.30 87.8%
60' T scan 49.53 91.6% 14.08 100.0% 12.32 94.2%
45' P scan 28.08 97.8% 7.38 %.1% 8.28 83.3%
60* P scan 31.91 100.0% 8.61 100.0% 11.42 100.0%
N6(A C) RHR/LPCf Nozzle (Automatic) 2 60.16 In total coda volume 2
14.08 In -inner 1/4t code volume 2
12.70 In weld volume + HAZ (1/4")
. Accessible Accessible Accessible Code Volune Inner 1/4t Code Volune Weld Volune j Type- Area % Area % Area %
Scan (In2) examined (in2) examined (in2) examined j i
0' 30.14 74.0% - -
8.95 8.9%
45' T scan 45.67 85.5% 14.08 %.7% 12.30 87.8%
60' T scan 49.53 91.6% 14.08 100.0% 12.32 94.2%
45' P scan 28.08 ,97.8% 7.38 96.1% 8.28 83.3%
60' P scan 31.91 100.0% 8.61 100.0% 11.42 100.0%
3 Note: Accessible is the volume which can be examined by present manual ultrasonic techniques RRI 14T1 j
GRAND GULF NUCLEAR STATION INSEIWICE INSPECTICN RDRUIREMEtfTS INSERVICE INSPECT 10N' SECTION 4 NE$c$ $. TEN YEAR PROGRAM MHEF PW RELIEF FEQUEST NO. I 00014 TABLE 1 PA0E 4 N7&8 TOP HEAD COOLING SPRAY AND SPARE N0ZZLE (MANUAL.)
2 20.80 in total code' volume 2
4.71 In inner 1/4t code volume 2
5.96 In -weld volute + HAZ (1/4")
Accessible Accessible Accessible Code Volune Inner 1/4t Code Volune Weld Volune Type Area % Area '% Area %
2 Scan (In ) examined (in2) examined (in2) examined O' 10.44 100.0% - -
5.00 100.0%
45' T scan 15.75 100.0% 4.71 100.0% 5.66 100.0%
60' T scan - 16.91 100.0% 4.71 100.0% 5.71 100.0%
45' P scan 10.44 100.0% 3.11 100.0% 4.86 100.0%
60' P scan 10.44 100.0% 3.11 100.0% 4.86 100.0%
N9(A&b) JET PUMP 1NSTRUMENT N0ZZlE (MANUAL) 78.29 In2 total code volune ~'
2 18.37 In -inner 1/4t code volume -
2 15.40 In weld volume + HAZ'(1/4")- l Accessible ' Accessible Accessible j Code Volune Inner 1/4t Code Volune Weld volume Type Area % Area % Area %
2 Sean (In ) examined (in2) examined '(in 2) examined O' 37.60 100.0% - -
9.64 100.9% i 45' T scan 59.00 100.0% 18.37 100.0% 15.06 100.0%
. 60* T scan 64.44 100.0% 18.37 100.0% 15.10 100.0%
45' P scan 37.60 100.0% 9.74 100.0% 9.64 100.0%
60* P scan 37.60 100.0% 9.74 100.0% 9.64 100.0%
Note: Accessible is the volume which can be examined by present manual ultrasonic techniques -
RRI-14T1
i GRAND GULF NUCLEAR STATION g INSERVICE INSPECTION SECTION 4 gjrgej $ TEN YEAR PROGRAM prrTEF REQUEST I
RELIEF REQUEST NO. I 00014 TA8LE 1 PAGE 5 N10 CRD RETURN N0ZZLE (MANUAL) 63.14 In# total code volune 2
14.08 In inner 1/4t code volume 2
12.70 In weld volume + HAZ (1/4")
Accessible Accessible Accessible Code Volume Inner 1/4t code Volume Weld Volune Type Area % Area % Area %
Scan (In2) examined (in2) examined (in2) examined O' 30.10 100.0% - -
8.40 100.0%
45' T scan 47.20 100.0% 14.08 100.0% 12.36 100.0%
60* T scan 51.40 100.0% 14.08 100.0% 12.40 100.0%
45' P scan 30.10 100.0% 7.68 100.0% 8.40 100.0%
60' P scan 30.10 100.0% 7.68 100.0% 8.40 100.0%
N16 VIBRATION WOZZLE (MANUAL) 84.46 In2total code volume i 2
19.52 In inner 1/4t code volume 2
16.17 In weld volume + HAZ (1/4")
Accessible Accessible Accessible Code Volune Inner 1/4t Code Volune Weld Volune Type- Area % Area % Area %
Scan (In2) examined (in2) examined (in2) examined O' ~ 40.00 100.0% - -
10.14 100.0%
45' T* scan 63.20 100.0% 19.52 100.0% 15.82 100.0%
60' T scan 68.40 100.0% 19.52 100.0% 15.86 100.0%
45' P scan 40.00 100.0% 10.36 100.0% 10.14 100.0%
60* P scan 40.00 100.0% 10.36 100.0% 10.14 100.0%
Note: Accessible is the volume which can be examined by present manual ultrasonic techniques RRI 14T1 i
__ )
' GRAND GULF NUCLEAR STATON INSERVICE INSPECfION REQUIREMEtTfS INSERVICE INSPECTION SECTION 4 fy/g $ TEN YEAR PROGRAM RELIEF REQUEST GRAND GULF NUCLEAR STATION UNIT 1 RELIEF REQUEST NO. I-00015 PAGE 1 of 3 INSERVICE INSPECTION OF REACIOR FKt:5SUnt! VESSEL WEIIE I. Component: Reactor pressure vessel (RPV) components, welds ard associated base material identified in table 1 of this relief request.
II. Code: 'Ihe unit 1 reactor pressure vessel was designed ,
ard fabricated to ASME Section III, class 1 requirements. Applicable inservice inspections are to be performed in accordance with Regulatory ;
Guide 1.150 Revision 1 and ASME Section XI, 1977 I Edition with Addenda through and including Summer 1979. Also, Relief Request No. I-00013 permits the use of ASME Section XI, 1983 Edition with the Summer 1983 Addenda, Figure IWB-2500-7(b) for identifying the Code required examination volume.
III. Code Requirements: Table IWB 2500-1, Examination Category B-D, B-A and B-F, requires specified volumes to be examined volumetrically at specified periods during the ten -
year interval. Included in this volume is varying degrees of base material adjacent to each weld that also requires examination.
IV. Information to Due to geometric configurations of the GGtG Unit 1 ,
I support the reactor, certain code required examination determination that volumes, as depicted in ASME Section XI, cannot be
(-
the code examined to the extent of obtaining full code requircments are coverage. Table 1 provides a listing of the impractical: affected components and reactor pressure vessel welds with a detailed description of the cause and degree of the limitation.
Relief Request Number I-00014 provides engineering rational addressing the limitations associated with the nozzle to vessel welds. The discussions provided prior operating plant experience to justify that no further examinations were necessary and additionally this was justified by recognizing that the feedwater nozzles are the limiting case. Although 100% of the code volumes were not examined for the nozzle to vessel welds, sufficient examination coverage was obtained to l- detect any potential cracking. ;
l . !
l ;
. GRAND GULF NUCLEAR STATION' INSERVICE INSPECI' ION REQUIREME2TfS
' INSERVICE INSPECTION SECI' ION 4
- ffifjf"", TEN YEAR PROGRAM PPTTW REQUEST GRAND GUIE NUCEAR STATION UNIT 1 PETT W REQUEST NO. I-00015 PAGE 2 of 3 l
IV. Information to A study of the welds listed in table 1 has support the shown that these weld locations are also determination that bourded by the feedwater nozzle discussions.
'the code 'Ihe stresses due to any expected loadings and requirements are conditions at these locations are bounded by impractical: those at the feedwater nozzle locations.
(continued) Supporting the concept of the bounding is the -
fact that no indications have been found at any of the partially examined locations. In addition, it should be noted that a partial examination of each weld was obtained which included either all or a portion of the reactor pressure vessel inner surface.
V. Specific relief Pemission is requested to perfom ultrasonic requested: examinations within the limitations described in table 1 of this relief request.
-VI. Reasons why relief Relief as described within should be granted should be granted: for the following reasons:
- 1. 'Ihe entire reactor pressure vessel was subjected to an ASME Section III hydrostatic test after fabrication.
- 2. 'Ihe entire reactor pressure vessel will be subjected to system leakage test at each refueling outage ard a system hydrostatic test each inspection interval in accordance with the requirements of ASME Section XI.
- 3. The subject welds were volumetrically examined in accordance with ASME Section III during fabrication.
1 1
GRAND GULF NUCLEAR STATION INSER'dCE DEMON WN ,
INSERVICE INSPECTION SEcrION 4 !
Ei'je[Nc.
j TEN YEAR PROGRAM RELTEF REQUEST I GRAND GUIF NUCLEAR STATION UNIT 1 RELTEF REQUEST NO. I-00015 PAGE 3 of 3 l
i VI. Reasons why relief 4. W ere is no history of service induced !
should be granted: flaws in these areas of the reactor i pressure vessel other than those of'the feedwater nozzles disnm W in Relief 1 Request Number I-00014.
- 5. 'Ihe areas of the reactor pressure vessel being examined are the lhniting areas.
- 6. The potential for initiation and propagation of cracking has been discussed assumirq both fatigue and ,
stress corrosion cracking mechanisms. It '
was concluded by the use of limiting analyses results performed for the feedwater nozzle blend radii, that cracking is unlikely at Grand Gulf nozzle / vessel weld locations. In fact, even if it was hypothesized that these ;
postulated cracks went undetected, a '
crack length of 58 inches was required -
l before rapid crack growth were to occur during normal operation. It'is unlikely that cracks of this size would go ,!
undetected. 'Iherefore, a significant leak i before break margin exists. i s
VII. Alternate testing: None l
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= WELD BB Il 18.6 5"
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GRAND GULF NUCLEAR STATON p g g pyg g gy g INSERVICE INSPECTION SECTION 4 g8/g y "d'. TEN YEAR PROGRAM RELIEF REQUEST 6'\ I 0a
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FIGURE 13
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GRAND GULF NUCLEAR STATION INSERVICE INSPECTION REGUDUNENTS INSERVICE INSPECTION SEOTION 4 l -
""gj$ TEN YEAR PROGRAM PFr.TEF REQUEST l
I t
l l
GRAND GUIF IUCLEAR STATION UNIT 1 RELIEF REQUEST NO. I-00015 {
l l
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RECIRWIATION CUITEI' NOZZLE TO SAFE END WEID IJMITATIONS E
1 FLOW >
45* SW N0ZZLE I SAFE END I
INCONEL BUTTER
/ A
//
, 'f NI -
l 'Ibtal code volume at a cross sectional view is .95 in2 o Automated techniques utilizing a 45' shear wave obtains two directional coverage of an area equal to .17 in l'
o Also, a 45' and 60' refracted longitudinal (RL) wave is utilized that obtains one directional coverage of the entire code volume scanning from the nozzle side.
o The cross hatched area received the one directional coverage from the nozzle side, the remaining area was examined from two directions utilizing the shear wave.
o Manual examinations will provide two directional coverage of the weld and nozzle base material, the available "W" dimension on the safe end side of the weld is insufficient to obtain full coverage of the safe end material.
_ FIGURE 14
~
.,, . m .~ . .n . a e.. -
- . . . . , ~ ~ ~ , - ~
J GRAND GULF NUCLEAQ STATION INSERVICE INSPECTION SEcrIai 4 Ej'$ j7"%, TEN YEAR PROGRAM pn m REQUEST GRAND GULF NUCIEAR STATION UNIT 1 RELIEF REQUEST 10. I-00015
-N2 RECIRCUIATIQi INIET IOZZIE 'IO SAFE DTD WEID LIMITATIONS l f
FLOW =
SAFE END NOZZLE
/
INCONEL BUTTER INCONEL BUTTER 60 RL / j
/45 RL j i
Total code volume at a cross sectional view is .846 in2 {
o Automated techniques utilized a 45' and 60' RL wave for the examination of the N2F and N2H to obtain complete code volume l coverage from one direction with both angles. I o The remaining N2 nozzles (A,B,C,D,E,G,J,K,M,N) were examined comoletely from one direction with only the 60' due to excessive noise received with the 45' transducers 1
o Manual examinations can provide two directional coverage of the i area represented by the p ss hatching, this would improye total i code coverage to .271 in or 32% for the 60' and .642 in or l 75.6% for the 45*. (
1 o If the presently available 45' transducers are improved to reduce noise levels, future examinations will include the 45* for improved coverage.
FIGURE 15
l
. GRAND GULF NUCLEAR STATON INSERVICE INSPECTION sEcTION 4 g8" g %, TEN YEAR PROGRAM per m REQUEST GRAND GUIE NUCLEAR STATION UNIT 1 FFi m RDQUEST NO. I-0001'5
)
N4 FEED WATER INIEP NOZZIE 70 SAFE END WELD IlMITATIONS
/
SAFE END N0ZZLE FLOW r
)
INCONEL BUTTER INCONEL BUTTER
'60 RL /45 RL ! !
Total code volume at a cross sectional view is .70 in2 -
l l
o Automated techniques utilizing a 45' and 60' RL wave obtained !
coverage of the code volume from one direction only (safe end side),
o Manual exaninations will provide coverage from two d y ions for the area identified above by the cross hatching, .36 in or 51.4%
of the total code volume.
FIGURE 16
GRAND GULF NUCLEAR STATON .
g ygg g gg INSERVICE INSPECTION SECTION 4 g8* jf"" TEN YEAR PROGRAM RELHT REQUEST GPAND GULF NUCLEAR STATION UNIT 1 RELIEF REQUEST 10. I-00015 N5 CORE SPRAY 'IO SAFE HO WEID IJMITATIONS
~
SAFE END NOZZLE FLOW =
, INCONEL BUTTER
//
60*RL 45 RL Total code volume at a cross sectional view is .72 IN2 Automated techniques provides a 45' shear wave frun the safe end side ining two directional coverage of an area equal to
.33 or 46.6% of the total code volume.
g Also, a 45' and 60' RL w2ae examination from the safe erd side is 3
/ performed obtaining one directional coverage of the total code volume. Area receiving examination from one direction only is 53.4%
of the code volume.
7 Manual examinations with RL wave will increase two directional s coverage by 18%.
FIGURE 17
GRAND GULF NUCLEAR STATION gggggy g INSERVICE INSPECTION SECTION 4.
jg
- g % TEN YEAR PROGRAM REIZEF REQUEST GRAND GUIF NUCIEAR STATION UNIT 1 RELIEF REQUEST NO. I-00015 f N6 RESIIXJAL HEAT REMOVAL / LOW PRESSURE CORE DUECTION j SAFE END TO NOZZLE WEID IJMITATIONS l
SAFE-END N0ZZLE FLOW >
I N l INCONEL BUTTER I l l
- 60* RL /45' R L j
Tbtal ccde volume at a cross sectional view.is .93 in2 Automated techniques provides a 45' shear wave examination from the safe end side obtaining two directional coverage of .34 in2 or 36.6% of the total code volume.
3 Also, 45' and 60' RL wave examinations performed from the safe end
-s side obtains one directional coverage of the total code volume;2 l area receiving one directional coverage only is 63.4% OR .59 D1 of the code volume.
1 tions will increase two directional coverage'by 9.7%
FIGURE 18
GRAND GULF NUCLEAR STKilON gggg INSERVICE INSPECTION SECTION 4 i g3/gj $ TEN YEAR PROGRAM pnm REQUEsr GRAND GULF NUCLEAR STATION UNIT 1 RELIEF REQUEST NO. I-00015 N9 *
' JET PUMP INSTRUMENIATION NOZZLE TO SAFE END WELD i I
i FLOW >- ~
N0ZZLE SAFE-END INCONEL BUTTER {~
\ 45* RL l
l 1
i Total code volume at a cross sectional view is .29 in2 Automated techniques provides a 45* shear wave examination from the safe end side obtaining two directional coverage of .10 in2 or 34.5% of the total code volume. I k
N Also, 45' and 60' RL wave examinations performed frun the safe end x '
sideobtainsonedirectionalcoverageofthetotalcodgvolume; area receiving one directional coverage only is .19 in or 65.5%
q of the total code volume.
minations will increase two directional coverage by 27.5%
1 FIGURE 19 i
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