ML20028D076
| ML20028D076 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/07/1983 |
| From: | Gucwa L GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0313, RTR-NUREG-313, TASK-A-42, TASK-OR NED-83-009, NED-83-9, TAC-46662, TAC-46663, NUDOCS 8301170042 | |
| Download: ML20028D076 (11) | |
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Pc0 OtN e h a U 25 iwt:s Go ;a UQ' Georgia Power Power Generation Departrnent January 7, 1983 Director of Nuclear Reactor Regulation Attention: Mr. John Stolz, Chief Operatirg Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EIMIN I. HA'ICH NUCLEAR PIANT UNITS 1, 2 ADDITIONAL INFORMATICN - NUREG-0313, REV.1 IMPLEMENTATION Gentlemen:
Georgia Power Company (GPC) hereby submits additional information on the implementation of NURKi-0313, Rev.1 at the Hatch units pursuant to the NRC letter dated November 1, 1982. The following information format corresponds to the item numbers in the aforementioned NRC letter:
Question la:
Please identify the modification or modifications which have made the 10-inch diameter core spray and 3-inch diameter CRD hydraulic return piping exempt from augmented ISI.
Response
Core Spray The 10" diameter Core Spray piping from the inboard injection valve lE21-F007A(B) to the reactor pressure vessel on Unit 1 was originally Type 304 stainless steel.
This piping was replaced with carbon steel material during the 1979 maintenance / refueling outage.
In addition,
/
during the same outage, the original stainless steel safe ends were replaced with low carbon stainless steel.
This modification was performed by GPC to eliminate piping materials which have a history of cracking at other plants and not as a result of any problems encountered at Unit 1.
Carbon steel material was utilized in the original construction on Unit 2 for the piping from the inboard injection valve 2E21-F007A(B) to the reactor pressure vessel. Safe end material is inconel.
8301170042 830107 PDRADOCK05000g P
Georgia Power A Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 January 7, 1983 Page Two Control Rod Drive (CRD)
The 3" diameter CRD hydraulic return piping is Type 304 stainless steel material and was rerouted to the 4" diameter reactor water cleanup (RWCU) return piping during the 1977 maintenance / refueling outage on Unit 1.
Piping was cut at the reactor pressure vessel and the vessel nozzle capped.
The CRD hydraulic return line nozzle cap material is inconel.
The rerouted piping is designated Class 2 and is exempt by ASME code from inservice inspection due to its size (i.e., 4.4" diameter nominal pipe size).
On Unit 2, the 3" diameter CRD hydraulic return pipirg is carbon steel material from check valve 2Cll-F082 to the connection at 4" diameter RWCU return.
Material upstream of check valve 2Cll-F082 is Type 304 stainless steel.
The piping was routed to the 4" diameter RWCU return in a similar fashion to that of Unit 1.
The CRD hydraulic return line nozzle is capped and utilizes inconel material.
The CRD hydraulic return piping is designated Class 2 and is exempt by ASME code from inservice inspection due to its size (i.e., ( 4" diameter nominal pipe size).
On both units, the CRD hydraulic return line connection to 4" diameter RWCU return is examined every refueling outage pursuant to the requirements of NUREG-0619.
Question lb:
If new piping was put in, please supply a copy of the manufacturer's certified chemical composition.
l
Response
Material certifications are not provided since the material utilized (i.e., carbon steel, inconel) is not addressed by NUREG-0313, Rev.1 or because it is exempt by ASME code from inservice inspection due to piping size (Class 2 piping, ( 4" diameter nominal pipe size).
j Question 2:
Identify the methods to detect and monitor unidentified leakage in the i
I pressure boundary piping of your BWR.
Some of these methods are enumerated in Regulatory Guide 1.45, Paragraph B.
Please fill out the attached table of information regarding the systems identified in the above paragraph, l
i horts
Georgia Power d Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 January 7, 1983 Page Three
Response
In response to Items 2a and 2b, please refer to FSAR Section 4.10 for Unit 1 and FSAR Sections 5.2.7 and 7.6.9 for Unit 2 for discussion on reactor coolant pressure boundary leak detection systems.
Requested information pertaining to the leakage detection systems for Units 1 and 2 are contained therein.
In addition, please refer to Unit 1 Technical Specification 3.6.G and Unit 2 Technical Specification 3/4.4.3 regarding reactor coolant leakage.
Question 3a:
Please identify the selection basis for which 50% of the RINrSA welds will be examined (e.g., stress rule index (SRI) or other bases).
Response
Due to the similar configuration of th-recirculation inlet nozzle thermal sleeve attachment (RINTSA) welt, on the "A"
and "B"
loops recirculation inlet nozzles, they will be examined on a sampling basis such that one loop will be examined during each appropriate examination interval.
However, if unacceptable flaw indications are detected in one of the RINISA welds for a given loop, then the welds in the other loop will also be examined. This selection basis is applicable to both Hatch units.
Question 3b:
If a method such as SRI or other damage indexes are used as a selection basis for examination, please identify the value of the index for every RINrSA weld in the 50% population to be examined.
Response
This item is not applicable to either Hatch unit since selection basis for examination is not based on stress rule index (SRI) or other indexes. Selection basis is as discussed in the response to Item 3a.
Question 3c:
Please identify the dates on which the RINISA welds have been inspected and to which inspection each date corresponds (e.g., first outage inspection, second successive outage inspection, first 36 + 12 month inspection, etc.)
~
k Georgia Power h Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commissica Washington, D. C.
20555 January 7, 1983 Page Four
Response
On Unit 1, the RINTSA welds were first examined in November 1982 during the Fall 1982 maintenance / refueling outage.
One hundred percent (100%)
of the RINTSA welds were examined and a
baseline examination established. As stated in our June 29, 1981 letter, the 1982 outage was considered as the first of the three 36 + 12 month examinations of service-sensitive piping and conponents.
Flfty percent (50%) of the RINTSA welds will be examined during consecutive 36 + 12 month periods using the selection basis discussed in our response to Item 3a.
In the event that no unacceptable indications are observed after three examinations of 36 i 12 month periods, examination will revert to an 80-month schedule.
The RINTSA welds were first examined on Unit 2 during January 1979. One hundred percent (100%) of the welds were examined at that time.
As stated in our June 29, 1981 letter, credit will be taken by GPC for those examinations in meeting the requirements of NURBG-0313, Rev.
1.
It was further stated that the RINTSA welds would be examined during the next two refueling outages (i.e.,1982 and 1983 refuelings). All of the RINrSA welds on the "A" loop of recirculation were examined during the 1982 maintenance / refueling outage.
The "B"
loop RINPSA welds are scheduled to be examined during the 1983 outage.
If no unacceptable indications are observed after completion of the 1983 examination, the RINrSA welds will then be examined three times on a 36 + 12 month schedule for examinations using the selection basis discussed in our response to Item 3a.
If after three consecutive examinations of 36 1 12 months there are no unacceptable indications observed, examinations will revert to an 80-month schedule.
Question 4a:
Please identify the methods for augmented ISI of the nonconforming service sensitive pipe.
Response
Nonconforming service-sensitive piping on the Hatch units is examined using volumetric (ultrasonics) and dye penetrant examinations, as appropriate, pursuant to ASME Section XI Code requirements.
Examinations of the nonconforming service-sensitive piping and canponents are currently scheduled as described in our response dated June 29, 1981 to NUREG-0313, Rev. 1 implementation.
700775 a
Georgia Power d i
Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C.
20555 January 7, 1983 Page Five Question 4b:
i Provide a copy of the specifications for the augmented ISI rethod or methods (IV.B.3 of NUREG-0313 Rev.1).
Response
Plans and procedures utilized by GPC and its contractors are not provided herein, but are available for review by NRC I&E personnel at the plant upon request.
It should be noted that ultrasonic examination procedures have been revised as a result of cracking in Recirc piping at Nine Mile Point and were successfully demonstrated at Battelle-Columbus to meet the requirements of NFC I&E Bulletin 82-03.
Question 4c:
4 l
Identify each of the augmented ISI methods used and the training and i
certification levels the individuals using those methods received.
Indicate if cracked specimens are used in your training (IV.B.3 of NURB3-01313 Rev.1).
Response
As noted in the response to Item 4a, volumetric (ultrasonic) and dye penetrant examination methods are utilized in the examination of nonconforming service-sensitive piping.
Trainincj of personnel performing nondestructive examinations is in accordance with the requirements of American Society of Nondestructive Testirx3, Recomended Practice No.
SNf 'IC-1A, 1975 Edition.
Typically, examinations are i
performed by Level II personnel with assistance (e.g., data recording) by a Level I or trainee.
The data are subject to review by Level III personnel.
The inservice inspection group of our parent company's service organization employs Ievel III personnel.
In addition, contract Level II and III personnel are also used, when appropriate. Training is conducted using cracked specimens obtained through the BWR Owners Group for IGSOC Research.
Although these specimens are retained by the service organization, they are made available to contract examination personnel performing examinations on site.
Question 4d:
Identify the proportion of the nonconforming service sensitive pipe that is being inspected (IV.B.2b of NUREG-0313 Rev.1).
70077$
Georgia Power A Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washirgton, D. C.
20555 January 7, 1983 Page Six
Response
One hundred percent (100%) of the nonconforming service-sensitive piping and conponents on both Hatch units has been inspected to date.
The service-sensitive piping and components include the 4" recirculation bypass stub tubes, recirculation inlet nozzle thermal sleeve attachment welds, 20" RHR Suction, and 24" RHR Return.
For future inspections, only circumferential welds will be examined for the augmented program because no problems have been observed in the solution annealed longitudinal welds.
Question 4e:
Identify the inspection interval of each system of the nonconforming service sensitive pipe (IV.B.2b of NUREU-0313 Rev.1).
)
Response
For Unit 1, the inspection just completed was considered the first of the three 36 + 12 month examinations.
All recirculation (i.e.,( 4" bypass stub tubes and RINISA welds) and RHR welds were examined.
Because of unacceptable indications observed in the 20" RHR and "B" loop 4
of 24" RHR Return piping, the examination interval for RHR lines only will revert to examination for three consecutive refueling outages.
If after three consecutive outages there are no additional unacceptable indications in the 20" a U." RHR piping, examinations will be performed on a 36 + 12 month basis as defined in N'JREC -0313, Rev.
1.
The 4" recirculatlon bypass stub tubes and RINESA weldt will be examined again in 36 i 12 months pursuant to NUREG-0313, Rev.1 guidarce.
As stated in our June 29, 1981 letter, the examination of the Unit 2 I
nonconforming service-sensitive lines will continue to be performed in accordance with comitments already made to NRC to examine the lines during three successive refueling outages.
The examination schedule is included in the existing Unit 2 Iong-Term Inservice Examination Plan for Class 1 Components.
Although not noted in the plan, the RINFSA welds will also be examined three times, but not successively.
As noted previously, credit will be taken for the 1979 and 1982 examinations of the RINPSA welds.
The RINTSA welds will be examined again in 1983 and 3
will complete the requirement of three " successive" examinations.
If there are no unacceptable indications observed for the nonconforming service-sensitive piping and components af ter completion of the 1983 examinations, the examination frequency will be exterried to three 361 12 month examinations, and later to an 80-month examination schedule, as appropriate.
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k Georgia Power h Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 January 7, 1983 Page Seven Question 4f:
Identify the Stress Rule Index Numbers for the welded joints in the nonconforming service sensitive pipe (IV.B.lb (6) of NUREG-0313 Rev.1).
Response
NURDG-0313, Rev.
1 provides guidance for determination of which service-sensitive arxl nonservice-sensitive welds are to be examined.
The subject NUREG does not require that SRI numbers be used for that determination.
Stress data have been provided by our A/E for the service-sensitive and nonservice-sensitive piping on both units.
Although use of the stress data is acceptable, in an effort to be uniform with others in the industry, SRIs are in the process of being calculated for the Unit 1 service-sensitive piping (RHR only).
SRIs will be calculated for the Unit 2 service-sensitive piping (RHR only) prior to the Spring 1983 maintenance / refueling outage.
In the interim, A/E-provided stress data are available to NRC for review upon request.
Question Sa:
Please identify the methods for augmented ISI of the nonconforming nonservice sensitive piping (IV.B.3 of NUREG-0313 Rev.1).
Response
i Similar examination methods, as discussed in the response to Item 4a, are employed in the examination of the nonservice-sensitive piping.
In
- addition, mechanized ultrasonic techniques are employed in the examination of selected 12" and 28" recirculation piping welds, where appropriate.
Question Sb:
i Please provide a copy of the specifications for the augmented ISI method or methods (IV.B.3 of NUREG-0313 Rev.1).
Response
Please refer to the response to Item 4b.
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k Georgia Power h Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C.
20555 January 7, 1983 Page Eight Question Sc:
Identify each of the augmented ISI methods used and the training and certification levels the individuals using those methods received.
Indicate if cracked specimens are used in your training (IV.B.3 of NUREG-0313 Rev.1).
_Rrsponse:
Please refer to the response to Item 4c.
Question 5d:
Identify the proportion of the nonconforming nonservice sensitive piping that is being inspected (IV.B.2b of NUREG-0313 Rev.1).
Response
In accordance with the guidance of article IV.B.1.b(2) of the NUREG, twenty-five percent (25%) of the circumferential welds in each piping system are to be examined.
The nonconforming, nonservice-sensitive welds are to be chosen in accordance with the guidance of the aforementioned article and examined at least once in no more than 80 months.
The scope of examinations for nonconforming, nonservice-sensitive piping is not increased in number, but performed on an increased frequency (i.e., 80 months rather than the code-required 120 months).
If no unacceptable indications are observed after the 80 month examination period, the schedule for examination of nonconforming, nonservice-sensitive welds will revert to the code-required frequency of 120 months.
For future inspections, only circumferential welds will be examined for the augmented program because no problems have been observed in the solution annealed longitudinal welds.
Question Se:
Identify the Stress Rule Index Numbers for the welded joints in the nonconforming nonservice sensitive piping (IV.B.lb (6) of NUREG-0313 Rev. 1).
700775
Georgia Power d Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 January 7, 1983 Page Nine
Response
Please refer to Attachment 1 for SRIs of Unit 1 nonservice-sensitive piping welds (recirculation only). The subject listing does not include the SRIs for the safe end-to-nozzle welds on recirculation system piping. The SRIs for the 6" RWCU pipirq were unavailable at the time of this writing and are in the process of being transmitted to GPC by the NSSS supplier.
Stress rule index numbers for Unit 2 are unavailable at this time.
As noted in the response to Item 4f, SRIs will be calculated prior to the Spring 1983 maintenance / refueling outage. The SRIs, carbon content, and NUREG-0313, Rev. 1 guidance will be used to determine which welds are to l
be examined.
l Question 5f:
Identify the proposed inspection interval for each system of nonconforming nonservice sensitive piping (IV.B.lb of NUREG-0313 Rev.1).
Response
The proposed inspection interval for those - systems on Units 1 and 2 having nonconforming, nonservice-sensitive piping is specifically addressed in our letter dated June 29, 1981, concerning implementation of NUREG-0313, Rev. 1 for the Hatch units.
As requested by the NRC letter, a copy of this response is beirg sent directly to your contractor, EG&G Idaho, Inc.
Should you have any questions in this regard, please contact this office.
Sincerely yours,
[E G%
L. T. Gucwa Chief Nuclear Engineer JAE/mb Attachment xc:
J. T. Beckham, Jr.
H. C. Nix, Jr.
Senior Resident Inspector J. P. O'Reilly (NBC Region II)
W. Roberts (EG&G Idaho, Inc.)
room
ATTACHMENT 1 STREbb EUL5 kNDEX
~
NONSERVICE SENSITIVE PIPING WELDS RECIRC SYSTEM LOOP "B" 12-BR-Al 1.114
.075
-A2 1.409
.075
-A3 1.508
.075
-A4 1.255
.075 12-BR-B1 1.063
.075
-B2 1.441
.075
-B3 1.447
.075
-B4 1.245
.075 12-BR-Cl 1.128
.075
-C2 1.527
.075
-C3 1.446
.075
-C4 1.346
.075 12-BR-D1 1.072
.060
-D2 1.510
.047
-D3 1.577
.047
-D4 1.361
.047 12-BR-El 1.079
.060
-E2 1.451
.047
-E3 1.566
.047
-E4 1.279
.047 22-BM-1 1.134*
.059
-2 1.037
.059
-3 1.043
.059
-4 1.134*
.059 22-EM-1BC-1 0.990
.060
-lEC-2 0.988
.060
-3BC-i 0.982
.060
-3BC-2 1.016 060 28B-2 1.087
.053
-3 1.484
.059
-4 1.411
.059
-5 0.956
.055
-6 0.939
.056
-7 1.415
.059
-8 1.020
.059
-9 1.041
.055
-10 1.524
.069
-11 1.004
.069
-12 1.142
.058
-13 1.056
.058
-14 1.056
.058
-15 1.496
.059
-16 1.237
.062
-17 1.035
.062
- Final SRI Pending Analysis
ATTACHMENT 1 HATCH - 1 STRESS RULE INDEX NONSERVICE SENSITIVE PIPING WELDS j
SRI CONTENT (%)
RECIRC SYSTEM LOOP "A" 12-AR-F1 1.111
.075
-F2 1.421
.075
-F3 1.534
.075
-F4 1.266
.075 12-AR-G1 1.057
.075
-G2 1.474
.075
-G3 1.557
.075
-G4 1.430
.075 12-AR-H1 1.118
.075
-H2 1.537
.075
-H3 1.519
.075
-H4 1.382
.075 12-AR-J1 1.075
.075
-J2 1.476
.075
-J3 1.473
.075
-J4 1.152
.075 12-AR-K1 1.101
.075
-K2 1.510
.075
-K3 1.559
.075
-K4 1.262
.075 22-AM-1 1.134*
.059
-2 0.921
.056
-3 1.041
.059
~4 1.134*
.059 22-AM-1BC-1 1.072
.060
-1BC-2 1.025
.060
-3BC-1 1.041
.060
-3BC-2 1.063
.060 28A-2 1.019
.055
-3 1.319
.059
-4 1.315
.059
-5 0.935
.056
-6 1.406
.059
-7 1.330
.059
-8 1.034
.055
-9 1.499
.069
-10 1.002
.069
-11 1.144
.058
-12 1.056
.058
-13 1.053
.059
-14 1.478
.059
-15 1.237
.062
-16 1.018
.062
- Final SRI Pending Analysis
_ -.