ML20210L271
| ML20210L271 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/30/1986 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20210L276 | List: |
| References | |
| CON-NRC-03-82-096, CON-NRC-3-82-96 SAIC-86-1628, NUDOCS 8604290277 | |
| Download: ML20210L271 (56) | |
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SAIC-86/1628
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TECHNICAL EVALUATION REPORT SECOND INTERVAL INSERVICE INSPECTION PROGRAM O
EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 O
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Submitted to U.S. Nuclear Regulatory Commission Contract No. 03-82-096
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Submitted by Science Applications International Corporation Idaho Falls, Idaho 83402 lO P
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.. . CONTENTS I N TR O DUCT IO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 q- 4 I. CLASS 1 COMPONENTS ......................
A.- Reactor Vessel ........................ 4
- 1. Relief Request ISI-2.1.1, Inaccessible Reactor o Pressure Vessel Welds, Category B-A, Items 81.11
" 4 and 81.12 ........................
- 2. Relief Request ISI-2.1.3, Full Penetration Welds of Nozzles in Vessels, Category B-D, Items 83.90 a n d 83.10 0 . . . . . . . . . . . . . . . . . . . . . . . . 7
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- 3. Relief Request 151-2.1.6, Reactor Vessel Nozzle-to-Safe End Welds (Nominal Pipe Size Less than 4 Inches), Category B-F, Item B5.20 ........... 9
- 4. Relief Request ISI-2.1.5, Reactor Pressure Vessel Support sk irt Weld, Category B-H, Item B8.10 . . . . . . . 11 O
l S. Relief Request ISI-2.1.9, Pressure Retaining Welds l in Control Rod Drive Housings, Category B-0, Item B14.10 ....................... 13 19 B. Pressurizer (not applicable to BWRs) l%/
C. Heat Exchangers (no relief requests) l D. Piping Pressure Boundary ................... 15 q 1. Relief Request ISI-2.1.7, Circumferential Weld j" in Containment Penetration Assedly, Category B-J, I t em B9 .1 1 . . . . . . . . . . . . . . . . . . . . . . . . 15 E. Ptnp and Valve Pressure Boundary ............... 19 1
l0 1. Relief Request 151-2.1.8, Pump Casings and Valve Bodies, Category B-L-2, Item B12.20, and Category B-M-2, Item B12.50 ............... 19 II. CLASS 2 COMP 0NENTS ................. 23 9 .
A. Pressure Vessels and Heat Exchangers ............. 23
- 1. Relief Request 1S1-3.1.1, Pressure Retaining Welds in Pressure Vessels, Category C-A, Items Cl .10, C1.20, and C1.30 .............. 23 O
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( B. Piping ............................ 26 l 1. Relief Request 151-3.1.2, Integrally Welded Piping Attachments, RHR, Core Spray, HPCI, and RCIC Suction Lines from Torus, Category C-C, Item C3.20 . . . . . . . . . . . . . . . . . . . . . . . . 26 i C. P um p s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
) Relief Request ISI-3.1.3, Surface Examination l 1.
of Pressure Retaining Welds in Class 2 Pumps.
C ategory C-G, Item C 6.10 . . . . . . . . . . . . . . . . . . 28
) D. Valves (no relief requests)
III. CLASS 3 COMP 0NENTS (no relief requests) ,
3- IV. PRESSURE TESTS ........................ 30 A. Class 1 Components (no relief requests)
- 8. Class 2 Components (no relief requests) h C. Class 3 Components . . . . . . . . . . . . . . . . . . . . . . . 30 l 1. Relief - Request ISI-4.1.1, Class 3 Piping 2 Inches and Smaller, Categories 0-A, 0-8, and 0-C, Items 01.10, 02.10, and 03.10 . . . . . . . . . . 30 3 2. Relief Request ISI-4.1.3, Hydrostatic Test of Plant Service System, Categories D-A, D-8, and 0-C; Items 01.10, D2.10, and 03.10 . . . . . . . . . . 32 V. GENERAL , . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 3
- 1. Relief Request ISI-8.1.1, Ultrasonic C al i bra ti on Blo ck s . . . . . . . . . . . . . . . . . . . . 34
- 2. Relief Request ISI-2.1.4, Ultrasonic Examination Requirements of Austenitic and Dissimilar Metal 3 Piping Welds . . . . . . . . . . . . . . . . . . . . . . . 36 l 3. Relief Request 151-8.1.2, Change Hatch 2
! ISI Second Ten-Year Interval to Begin on January 1,1986 ..................... 39 O
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- 4. Relief Request ISI-8.1.3, Reporting Requirements of Procedural Changes Affecting Inservice Inspection Programs .............. 41 3
- 5. Relief Request ISI-8.1.4, Requirements of Secti on XI, Subsecti on IWE . . . . . . . . . . . . . . . . 43 6'
. Relief Request ISI-8.1.5, Due Date for Owner's
, Data Report for Inservice Inspection, Form NIS-1. . . . . 45 J l VI. COMPONENT SUPPORTS ...................... 47
- 1. Relief Request 151-5.1.1, Acceptance Criteria for Springs and Snubbers, Category F-C, Item F3.50 . . . . 47 3 i,
- 2. Relief Request 151-5 .1.2, Snubber Testing :
Program IWF-5300 and IWF-5400 .............. 49 !
- 4. REFERENCES .......................... 52
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TECHNICAL EVALUATION REPORT SECOND INTERVAL INSERVICE INSPECTION PROGRAM j Hatch Nuclear Station Units 1 and 2 INTRODUCTION This report evaluates requests for relief from Section XI of the
) American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code
- by the licensee, Georgia Power Company (GPC), for Hatch Nuclear Station, Units 1 and 2 (HNS-1 and HNS-2). The relief requests cover the j second 120-month inspection interval starting January 1,1986. The I requests are based upon the 1980 Edition of Section XI, with Addenda through the Winter of 1981, as specified in the applicable revision of
The rest of this introduction summarizes (a) the scope of this report, (b) the previous review of rg national Corporation (SAIC),l]jef requests 1 and by Science (c) the history of HNS-1 Applications and HNS-2 Inter-
) since the earlier review.
The current revision to 10 CFR 50.55a requires that Inservice Inspection (ISI) programs be updated each 120 months to meet the require-ments of newer editions of Section XI. Specifically, each program is to meet the requirements (to the extent practical) of the edition and addenda
) of the Code incorporated in the regulation by reference in paragraph (b) 12 months prior to the start of the current 120-month interval.
Hatch-l's construction permit was issued in September 1969, prior to the effective date of implementation for ASME Section XI, and thus the plant was not designed to meet the requirements of inservice inspection.
NRC regulation (10 CFR 50.55a(g)(1)) requires that the plant meet the
} inservice inspection requirements to the extent practical, except design and access provisions.
t Hatch-2's construction permit was issued in December 1972. At that time, the 1971 Edition of the ASME Code,Section XI, was in effect. The first edition of Section XI had not oeen published until 1970, and exami-l nation requirements for Class 2 components were not included until the 1974 l Edition. Since the Hatch 2 plant system design and ordering of long lead j time components were well under way by the time the Section XI rules became
- effective, full compliance with the access and inspectability requirements i was not always practical. However, a review of the applicable (1) Code
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- Hereinafter referred to asSection XI or Code.
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D l sections (from IS-141 in the 1974 Edition to IWA-1500 in subsequent ;
editions) and (2) NRC regulations (10 CFR 50.55a(g)(2) and (g)(4)) indi- l cates that the intent should be to continue striving to address both access and inspectability requirements with the best available instrumentation and procedures.
D The regulation recognizes that the requirements of the later editions l and addenda of the Code might not be practical to implement at facilities because of limitations of design, geometry, and materials of construction of components and systems. It, therefore, permits exceptions to impracti-cal examination or testing requirements to be evaluated. Relief from these D requirements can be granted provided the health and safety of the public are not endangered, giving due consideration to the burden placed on the licensee if the requirements were imposed. This report only evaluates' requests for relief dealing with inservice examinations of components and with system pressure tests. Inservice test programs for pumps and valves (IST programs) are being evaluated separately.
Finally,Section XI of the Code provides for certain components and systems to be exempted from its requirements. In some instances, these exemptions are not acceptable to the U.S. Nuclear Regulatory Commission (NRC) or are only acceptable with restrictions. As appropriate, these instances are also discussed in this report.
O In its previous report dated Septenter 2,1982,(l) SAIC evaluated relief requests for HNS-1 covering the last 80 months of the first 120-month interval that ended May 5,1986. The report also evaluated relief requests for HNS-2 covering tne entire first 10-year inspection interval (September 5,1979, to September 5,1989). These requests were based on
. the 1974 Edition of the Code with addenda through Summer 1975. The applic-able Code and interval were in accordance with the revision of 10 CFR 50.55a in effect at that ime. On July 29, 1983, the NRC issued its formal Safety Evaluation Report, 2) which included SAIC's report as an appendix.
On August 12, 1983,(3) GPC submitted a revised ISI program for the
_) first 120-month interval for Hatch Units 1 and 2. The revised program was for an 80-month period beginning January 1,1984, for both units.
The revised program was verbally disapproved by the NRC in January 1985.
By letter dated July 18, 1985,(4) GPC stated its intention to complete the examinations for the Hatch Unit 1 first ten-year interval on J Decemb er 31, 1985, and to an inspection program which was written to the 1980 Edition of the ASME Section XI Code with Addenda through Winter 1980 where practical. The subject letter also requested relief from certain of l these requirements.
The NRC issued its Safety Evaluation Report by letter dated November 7, 19 85.( 5) A meeting between the NRC and GPC was held November 19,1985, to discuss the safety evaluation. At this meeting, the NRC requested that GPC provide a write-up concerning how the 1980 Edition through Winter 1980 Addenaa inservice has been applied inspection. at Hatch Unit The information was1supplied since January 1,1984, Novemoer relativg)to 27, 1985. b 9
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On June 25, 1985, GPC submitted an } program for Units 1 and 2 covering the second 120-month interval.t{? The relief requests and program were based upon the 1980 Edition of Section XI of the Code with addenda through Winter 1981. The Code edition and inspection intervals (interval O start date was modified by a relief request) were in accordance with the revision of 10 CFR 50.55a applicable at the time.
NRC requested additional inform on, required to evaluate the revised GPC ISI plan, on December 23,1985.ppi The licensee responded to(the t
request by submitting additional information on February 7, 1986. 9) In O this response, relief request ISI-2.1.2 was withdrawn, and relief request ISI-4.1.2 and a portion of relief request 151-2.1.1 were withheld pending review of actual conditions encountered during inspection efforts.
The 20 relief requests contained in Reference 7 and modified by Reference 9 are evaluated in this report. The material included in the O paragraphs titled Code Relief Request, Proposed Alternative Examination, and Licensee's Basis for Requesting Relief is quoteo oirectly from tne relief requests in References 7 ano 9 except for minor editorial changes such as removing references to figures and tables not included in this report. l 0
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I. CLASS 1 COMPONENTS A. Reactor Vessel
) 1. Relief Request ISI-2.1.1, Inaccessible Reactor Pressure Vessel Welds, Category B-A, Items 81.11 and 81.12 (Item Bl .12 applies to Unit 1 only)
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Code Requirement One circumferential and one longitudinal pressure-retaining weld in the beltline region of the reactor pressure vessel must be
) volumetrically examined in accordance with Figures IWB-2500-1 and
-2 over essentially 100% of the weld length in the second and suc-cessive inspection intervals. Examinations may be performed at or l near the end of the interval.
i Code Relief Request Relief is requested from volumetric examination over 100% of the length of one circumferential and one longitudinal weld in the beltline region.
h Proposed Alternative Examination For future examinations of RPV circumferential, longitudinal, and meridional welds, the examinations will be performed to the 3 extent possible.
Licensee's Basis for Requesting Relief D The HNS-1 and -2 construction permits were issued before the effective date of implementation for ASME Section XI, and thus the plant was not designed to meet the requirements of inservice in-spection; therefore,100% compliance is not feasible or practical.
j Currentl y, it is not feasible to perform the required volumetric examinations on these welds. Accessibility for weld inspection was not provided for in the original plant design which
, occurred prior to the issuance of Section XI inservice inspection requirements.
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Examination from the reactor vessel outer surface is pre-cluded due to the close proximity to the biological shield wall and obstruction by the vessel insulation. The mirror type insulation consists of interlocking panels which were not decigned
) to be easily removable at the weld locations. Furthermore, the annular dimensions between the shield wall and the insulation are not sufficient to allow direct access to personnel. Access through the biological shield wall is only provided at reactor vessel nozzle locations; however, there are no nozzle penetrations in the beltline region.
Examination of the beltline region welds from inside the vessel is impeded by vessel internal design features. The core shroud, jet pumps, and various brackets welded to the vessel wall are not designed to be removable.
Evaluation Portions of the concrete biological shield and the perma-nently installed insulation would have to be removed to perform
) the required examination of the welds from the vessel exterior.
The vessel internals, shroud, and jet pumps preclude volumetric examination of almost all the beltline weld volume from the vessel interior. Hence, impos,ition of Code requirements would be impr actical .
} For Hatch Unit 1, there are two circumferential welds (C-3 and C-4) in the beltline region. Weld C-4 has three access doors through the concrete shield, and removable RPV insulation in these areas was provided during the design. These three access ports allow the manual examination of approximately 15% of the weld.
Weld C-3 has two usable access doors allowing approximately 10%
coverage; therefore, an equivalent length of 25% of one beltline region circunferential weld can be examined. Portions of circum-ferential welds outside the beltline region will be examined in order that the total equivalent length being examined equals the length of one beltline circumferential weld.
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Using the same access doors as discussed above, a total of between 50 and 75% equivalent length of one beltline longitutinal
' weld can be manually examined. This percentage is split between the three longitudinal welds C-3-A, C-3-B, and C-3-C. Sufficient lengths of longitudinal welds outside of the beltline region will be examined in ceder that the total length being examined equals the length of one beltline longitudinal weld.
The Hatch Unit 2 RPV is equipped with permanently installed tracks. Approximately 17% of each circumferential beltline weld (welds 2C-3 and 2C-4) can be examined using a mechanized system.
Therefore, an equivalent length of approximately 34% of one
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beltline region circumferential weld can be examined. Portions of circunferential welds outside the beltline region will be examined in order that the total equivalent length being examined equals the length of one beltline circumferential weld. For Hatch Unit 2, 3 the mechanized system allows examination of 100% of the actual length of one beltline region longitudinal weld and relief is not required for these welds.
The licensee has com:aitted (in Reference 9) to examining enough welds of the same type within and outside of the beltline O region such that the equivalent examination length (volumetric examination of 100% of one beltline weld) is met using multiple welds.
O Conclusions and Recommendations Based on the above evaluation, it is concluded that for the welds discussed above, the Code requirements are impractical. It is further concluded that the alternative examinations discussed in the evaluation will provide necessary added assurance of
,0 structural reliability. Therefore, it is recommended that relief be granted from the volumetric examination of the identified welds, provided additional welds under each of the Code Item numbers within the beltline region and of similar type outside the beltline region are examined to the extent that the length of examined welds equals the length of the weld requiring i
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It is furtner recommended that the licensee should, as originally proposed, keep abreast of improvements in state-of-the-art NDE techniques. If a means becomes available for
- _ examining the Code-required beltline welds, these welds should V be examined in lieu of the examinations recommended above.
References References 7 and 9.
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- 2. Relief Request ISI-2.1.3, Full Penetration Welds of Nozzles in Vessels, Category B-D, Items 83.90 and B3.100 O- Code Requirement The nozzle-to-vessel weld and the nozzle inside radius section, including adjacent areas of the nozzle and vessel, of all reactor vessel nozzles shall be volumetrically examined in accor-
. dance with the applicable portion of Figure IWB-2500-7.
At least 25% but not mcre than 50% (credited) of the nozzles shall be examined by the end of the first inspection period and the remainder by the end of the third inspection period. If examinations are conducted from inside the component and the nozzle weld is examined by straight beam ultrasonic method from 0 the nozzle bore, the remaining examinations required to be conducted from the shell may be performed at or near the end of each inspection interval.
O Code Relief Request Relief from volumetric examination of 100% of the Code required volume is requested for the below listed nozzles on the reactor pressure vessel.
O Hatch Unit 1 Nozzle Identification Limited Examinations N2A N to V O N2B N to V N2C N to V N2D N to V
. N2E N to V N2F N to V N2G N to V
.O N2H N to V N2J N to V N2K N to V N48 N to V; IRS N40 N to V; IRS
'O Hatch Unit 2 2N4A N to V; IRS 2N4C N to V; IRS O
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Proposed Alternative Examination At a minimum, 85% of the Code required examination volume, of the above listed nozzles, shall bm ultrasonically examined.
D Licensee's Basis for Requesting Relief Hatch was designed prior to issuance of Section XI inservice inspection requirements. Consequently, physical limitations
'3 prevent sound beams from passing through the entire examination volume as shown by Code Figure IWB-2500-7(a) through (d).
Evaluation 3
The licensee estimates that a minimum of 85% of the Code required volume will be examined for the nozzles for which relief is requested. In addition, the nozzles will be visually examined during periodic system leakage tests and hydrostatic tests g required under Category B-P.
The licensee has attempted to minimize the limitations of volumetric examination of the RPV nozzles by addressing a variety of scan heads and scan angles. The licensee has determined that scan limitations to these nozzles exist regardless of transducer g* size. A minimum of 85% coverage of the RPV nozzle Code required volume has been achieved.
Conclusions and Recommendations Based on the above evaluation, it is concluded that for the
- nozzle inside radius sections and welds discussed above, adherence l to the Code requirements is impractical. It is further concluded that the proposed examinations will provide necessary assurance of structural reliability during this interval. Therefore, relief is g recomended as requested provided:
(a) the volumetric examinations are performed to the maximum extent practical, and l (b) the Code required system pressure tests are performed in O accordance with IWB-5000.
References g References 7 and 9.
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- 3. - Relief Request 151-2.1.6, Reactor Vessel Nozzle-to-Safe End Welds (Nominal Pipe Size Less Than 4 Inches), Category B-F, Item B5.20 C
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Code Requirements 3 All dissimilar metal nozzle-to-safe end welds in the reactor vessel shall be surface examined in accordance with Figure IWB-2500-8 during each inspection interval. The examinations may be performed coincident with the vessel nozzle examinations required by Examination Category B-D. Dissimilar metal welds between com-binations of (a) carbon or low allcy steels to high alloy steels, (b) carbon or low alloy steels to high nickel alloys, and (c) high
} alloy steel to high nickel alloys are included.
Code Relief Request J Relief is requested from surface examining 100% of the required area of each of the following reactor vessel nozzle-to-safe end welds:
N10 N12A N168 2NilB 2N16A 3 NilA N128 2N10 2N12A 2N16B NilB N16A 2 Nil A 2N128 Proposed Alternative Examination
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The nozzle-to-safe end welds listed above will receive a remote visual examination. In addition, these nozzles will be l
pressure tested per IWB-5000 of ASME Section XI since they are located within the hydrostatic test boundary of the Nuclear Steam j Supply System.
Licensee's Basis for Requesting Relief l
9 The nozzle-to-safe- end welds for the nozzles listed above are physically inaccessible for surface examination.
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! These 2-inch instrument nozzles have very limited access due to the design of the concrete shield. Each nozzle has small doors that can be opened allowing 12 to 18 inches of access. However, due to the distance the RPV wall is recessed from the outside of the shield wall (e.g., insulation thickness, air gap, and shield thickness), the' weld cannot be physically reached. As an alternate, the weld will be examined using remote visual means (e.g. , fiber optics, boroscope, etc.).
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The licensee has committed to perform remote visual examina-ion of the subject welds. The visual examination in conjunction with the Code required pressure tests will provide adequate
! information as to the integrity of the nozzle-to-safe end welds.
3 Conclusions and Recommendations Based on the above evaluation, it is concluded that for the
] welds discussed above, adherence to the Code requirements is impractical. It is further concluded that the proposed examina-tions will provide necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested provided:
(a)- the remote visual examinations are performed, and Q
(b) the pressure tests required by IWB-5000 are performed.
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References References 7 and 9.
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- 4. Relief Request ISI-2.1.5, Reactor Pressure Vessel Support Skirt Weld, Category 8-H, Item B8.10 3
I Code Requirement Integrally welded attachments to the reactor pressure vessel shall be volumetrically or surface examined over 100% of the length of the weld to the vessel. The examination volume will cover the area specified in Figures IWB-2500-13, -14, or -15, as applicable.
j Code Relief Request Relief is requested from performing 100% surface examination of the RPV support skirt weld as defined in Figure IWB-2500-13.
O Proposed Alternative Examination l-Examination area A-B of Figure IW8-2500-13 will be 100%
surface examined. Also, a limited ultrasonic examination will be performed to the extent practical to provide as much coverage as g possible of the weld.
Licensee's Basis for Requesting Relief At both Hatch units, examination area C-D of Figure
- O IWB-2500-13 is nat accessible for meaningful examination because of location and geometric configuration of welded areas. Physical access by the examiner is also restricted because of high radi-ation and obstruction due to CRD housings and its support system.
The combination of these factors prevents these welds from being g examined from inside the support skirt.
Evaluation
- O The reactor vessel support skirt-to-vessel weld is impractical to surface examine from the inside surface considering limited access for examination personnel, insulation removal re-quirements, and personnel radiation exposure.
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l Hatch Unit 1 has no access through the support skirt; there-fore, the inside surface of the reactor vessel support sk1rt is totally inaccessible. For Hatch Unit 2, the support skirt near the skirt-to-vessel weld is very limited. Health physics
- ) indicates that the dose rate in this area during the last Hatch Unit 2 outage was approximately 180 mr/hr; hcwever, it is very l difficult to quantify the total exposure for an examination.
l Magnetic particle techniques cannot be used due to the space restrictions. The use of dye penetrant would require a very i
thorough cleaning of the weld and adjacent base material to remove
[] rust and scale. The preparation of the weld would potentially have to be performed using techniques such as wire brushes since power tools may not fit into the limitec area.
I Examination of the outside surface along with an ultrasonic examination to the extent practical will provide adequate
[) information about the structural integrity of the subject weld.
Conclusions and Recommendations
- ) Based on the above evaluation, it is concluded that for the weld discussed above, the Code requirement is impractical. It is further concluded that the proposed examinations will provide necessary assurance of structural reliability during this interval.
Therefore, it is recommended that relief from the requirement of a full volumetric or surface examination of the reactor vessel
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support skirt weld be granted provided the licensee performs a partial surface examination and an ultrasonic examination to the extent practical as proposed as an alternative examination.
References
, References 7 and 9.
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5.. Relief Request-ISI-2.1.9, Pressure Retaining Welds in Control Rod Drive Housings, Category B-0, Item 814.10 Code Requirements Volumetric or surface weld examinations shall be performed '
during each -inspection interval and shall include 100% of the O welds in 10% of the peripheral control rod drive (CRD) housings in accordance with Figure IWB-2500-18. The examinations may be performed at or near the end of the inspection interval.
O Code Relief Request Relief is requested from the volumetric or surface examination of the peripheral CR0 housing welds.
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Proposed Alternative Examination All peripheral CRD housing welds shall be visually examined during the system hydrostatic pressure tests in accordance with
- IW8-5000.
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Licensee's Basis for Requesting Relief.
The examinations of these welds are limited because of the lO ~ location and design of the housings. Physical accessibility by I an examiner is extremely limited by the close proximity of the housings to each other and by the support arrangement. Also, the insert and withdraw lines to the Control Red Drive system are connected at the top of the housing flange and limit access to '
much of the lower weld. The combination of these factors limit O the examination of these welds.
Evaluation
- O Because of the physical inaccessibility of the lower welds and the high radiation fields to which examining personnel would be exposed, examination of the peripheral CR0 housing welds is impractical . ,
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IW8-1220(a) allows an exemption for those components that are connected to the reactor coolant system and are part of the reactor coolant pressure boundary and that are of such a size and shape so that upon postulated rupture, the resulting flow of coolant from
,, the reactor coolant system under normal plant operating conditions
- J is within the capacity of makeup systems which are operable from on-site emergency power.
The licensee has shown (see Section 4.2 of the FSAR),that the makeup system has sufficient capacity to shut down and cool the
,s reactor in an orderly manner should a complete failure of a CRD LJ housing weld occur. Figure 4.2-8 of the FSAR shows that diere are 28 peripheral CRD housings. Each housing has an attachment weld to the reactor vessel and a weld joining the housing to the flange.
Section 4.2 of the FSAR shows that the failure of a CRD housing weld will produce a maximum leakage rate of 840 gal / min. The available makeup systems are RCIC-400 gal / min, CRD-160 gal / min, w>- and the transfer system to feedwater-1000 gal / min. Theref ore, the reactor can be shut down and cooled down in an orderly manner using makeup sy? ' ms. supplied by onsite power, as required by IW B-1220. Since loss of coolant is assumed to occur during normal operation,_ it is the licensee's interpretation that the service transformer is the scurce of onsite power.
Thus, the exemption requirements of paragraph IWB-1220(a) l are satisfied and the examinations required for Code-exempted components would be performed by the licensee. Both of these bases contain sufficient justification for Code relief.
lO Conclusions and Recommendations l Based on the above evaluation, it is concluded that the welds
, discussed above meet the exemption requirements of IWB-1220(a) and H ,s relief is not required.
l j_) References t-References 7 and 9.
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i-8.- Pressurizer
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Not applicable to BWRs.
C. Heat :;. changers No relief requests.
] D. Piping Pressure Boundary
- 1. Relief Request ISI-2.1.7, Circumferential Weld in Containment Penetration Assembly, Category 8-J, Item 89.11 (applies to Hatch 9 Unit 1 only)
Code Requirement For circumferential welds with nominal pipe size 4 in. and 0- greater, surface plus volumetric examinations in accordance with Figure IW8-2500-8 shall be performed during each inspection interval on essentially 100% of the weld. The examination shall include the following:
(a) All terminal ends in each pipe or branch run connected to
) vessels.
(b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed the following limits under loads associated with specific seismic events and operational conditions:
(1) primary plus secondary stress intensity range of 2.4Sm for ferritic steel and austenitic steel, and (2) cumulative usage factor U of 0.4.
(c) All dissimilar metal welds between combinations of:
(1) carbon or low alloy steels to high alloy steels, (2) carbon or low alloy steels to high nickel alloys, and (3) high alloy steels to high nickel alloys.
(d) Additional piping welds so that the total equals 25% of the O circumferential joints in the reactor coolant piping system."
This does not include welds excluded by IWB-1220. These additional welds may be located in one loop (one loop is currently defined for both PWR and BWR plants in the 1980 Edi tion) .
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b,-
For welds in carbon or low alloy steels, only those welds i showing reportable preservice transverse indications need be examined for transverse reflectors.
Code Relief Request Relief is requested from volumetric and surface examination of pressure-retaining piping welds in primary containment penetra-
) tion assemblies as follows:
Weld Identification No. Penetration No.
1821-lFW-18A-7A X-9A 1821-lFW-188-6A X-98
) lE51-lRCIC-4-D-20A 1E41-lHPCI-10-0-15A X-10 X-ll lEl l-lRHR-208-D-13A X-12 lEll-lRHR-24A-R-3A X-13A l El l-2RHR-248-R-38 X-138 1G31-lRWCU-6-D-158 X-14
) 1G31-lRWCU-6-D-15C lE21-lCS-10A-3A X-14 X-16A l E21-lCS-108-4A X-16B lEll-lRHR-4-HS-6A X-17
} Proposed Alternative Examination A UT baseline was performed on the two new welds in the Reactor Water Cleanup (RWCU) system while they were accessible during repair to ensure a high quality weld.
) In accordance with IW8-5221 of ASME Section XI, a system leakage test is to be performed on all 12 welds prior to startup following each reactor refueling outage.
All pipe-to-penetration (flued head) welds outside containment will be examined volumetrically. In addition, a surface examina-
) tion will be performed on the accessible weld (s) of the flued head penetration assembly.
Licensee's Basis for Requesting Relief J
These welds are inaccessible for examination due to the design of the flued head. All 12 circumferential butt welds, except the two located in the Reactor Water Cleanup (RWCU)
)
l 16 D .
)
Evaluation The identified welds are completely inaccessible for volu-
)-
metric or surface examination because the welds are located inside a containment penetration. Each primary containment penetration assembly, due to its design, leaves one pressure-retaining piping weld inaccessible for examination by either surface or volumetric means (the RWCU penetration has two inaccessible welds). The welds can only be examined by inspecting for evidence of leakage
) during system hydrostatic pressure tests.
The two stainless steel welds that are located in the RWCU penetration were made to replace a Type 304 SS pipe that had-undergone IGSCC. The welds involved are a flued head with a Type 308L corrosion resistant clad on the inside surface to a Type 304L
) solution annealed pipe (<.035% carbon), and a Type 304L pipe-to-pipe weld. These welds were made in accordance with the guidelines of NUREG-0313 to minimize susceptibility to IGSCC.
The initial design of the assemblies did not provide for accessibility for inservice examinations. If, however, the work-
) manship and quality assurance of the welding as well as the preservice examinations are assumed adequate, then an examination of the first pressure boundary weld either upstream or downstream of the inaccessible weld should reflect service-induced failures for that particular piping section. Thus, the first pressure 3 boundary weld adjacent to the inaccessible weld on each of these
/- proces; pipes should be volumetrically examined, where practical, over 100% of its length during each inspection interval. Such an examination would maintain sample size. The licensee should also conduct visual examinations at these penetrations, as proposed, which would indicate any cracks through the metal.
)
l l Conclusions and Recommendations l
1 y Based on the above evaluation, it is concluded that for the <
welds discussed above, the Code requirements are impractical. It l
! is further concluded that the alternative examination discussed l above will provide the necessary added assurance of structural j reliability. Therefore, it is recommended that relief be granted
! from the volumetric examination of the identified welds, with the following provisions:
t 17 h -
t
(a) The first accessible pressure. boundary weld either up-stream or downstream of the inaccessible weld on each of these process pipes should be examined by volumetric and J surface methods, where practical, over 100% of its length during each inspection interval.
(b) The proposed visual examinations should be performed on the containment penetration assemblies when leakage and hydrostatic tests are conducted in accordance with 3 IWB-5000.
References~
Q Reference 7.
3 D
D O
O C) .
18 O .
i
E. Pump and Valve Pressure Boundary
! 1. Relief Request ISI-2.1.8, Pump Casings and Valve Bodies, Category B-L-2, Item B12.20, and Category B-M-2, Item 812.'50 ' '
Code Requirements A visual examination (VT-3) of the internal surfaces must be y performed on at least one pump in each group of pumps performing similar functions in the system. The visual examination may be .
performed on the same pump selected for volumetric examination of the welds.
A visual (VT-3) examination of the internal surfaces must be j perforrred on at least one valve in each group of valves exceeding 4 in. nominal pipe size which are of the same constructional design, such as globe, gate, or check valves and manufacturing method, and that are performing similar functions in the system such as containment isolation and system overpressure protection.
The visual examination may be performed on the same valve selected 3 for volumetric examination of valve body welds.
Code Relief Request
] Relief is requested fron visual examination of the internal surfaces of Class 1 pumps anJ valves in piping greater than 4 in, nominal pipe size.
] Proposed Alternative Examination Class 1 pumps and Class I valves exceeding 4 in. nominal pipe size are subject to visual examination of the internal surfaces when disassembled for maintenance. The coverage provided by exami-nations during routine maintenance coupled with periodic leak
] tests and hydrostatic tests will provide adequate assurance of the structural integrity of the Class 1 pumps and valves, while keeping exposure to radiation and contamination as low as reasonably achiev able.
J Licensee's Basis for Requesting Relief Disassembly of these valves and pumps for the visual examination during the inspection interval, in the absence of other required maintenance, represents an unnecessary exposure 19 3 .
D w to radiation and contamination. Valves on the Reactor Recir-culation (RC) system and the Residual Heat Removal (RHR) system suction lines would require off-loading the fuel elements and draining the RPV prior to disassenbly. Work on the RC system pump g discharge valves and the RHR system injection valves would require the installation of plugs in the jet pump risers. Preparatory work of this scope is considered impractical for the sole purpose of conducting a visual examination.
Contamination levels in the valves and pumps associated with g the RC system loops are particularly high due to the physical location at the bottom of the system. During routine maintenance, the valve body and the pump casing internal surfaces are visually examined. Many of the valves, particularly the containment isola-tion valves are disassembled for maintenance of leak-tightness.
Disassembly of other Class 1 valves and the pumps solely for g internal examination is counter to the ALARA guidelines to keep the occupational dose rates as low as reasonably achievable. In view of the cost in man-rem and in view of the minimal benefits obtained, we conclude that this Code requirement does not provide sufficient benefits to justify the exposure.
g A preliminary review of the Hatch Unit 1 records show that the following valves have been disassembled at least one time during the first ten-year interval:
IB21-F010A,8 18" Check-Feedwater IB21-F032A 18" Check-Feedwater g lEll-F015B 24" M0 Gate-RHR lEll-F017A 24" M0 Globe-RHR lEll-F030B 24" Check-RHR lE21-F006A 10" Check-Core Spray lE41-F002 10" M0 Gate-HPCI lE41-F003 10" M0 Gate-HPCI O l-MSIV 28" A0 Globe-Main Steam The Class 1 valves greater than 4-inch diameter (that were not examined on at least one loop) are:
O IB31-F023A,8 28" M0 Gate-Recirculation IV31-F031A,8 28" M0 Gate-Recirculation lE21-F003A,B 10" M0 Gate-Core Spray 1G31-F001 6" M0 Gate-RWCU IG31-F004 6" M0 Gate-RWCU O lEll-F008 20" M0 Gate-RHR lEll-F009 20" M0 Gate-RHR lE41-F006 14" M0 Gate-HPCI l E21-F007A,8 10" Manual-Core Spray 9
20 0
o 1G31-F027 6" Manual-RWCU 1821-F0llA,8 18" Manual-Feedwater lEl 1-F060A, B 24" Manual-RHR lEll-F067 20" Manual-RHR
^O As a precautionary feature, the core would normally be off-loaded if valves 1B31-F023A,B or 1831-F031 A,8 were to be disassembled.
Also, if the Recirculation Pumps were disassembled, it would be desirable to off-load the core for safety reasons. (Note: the
- O Recirculation Pumps are the only Class 1 pumps.) The actual expo-sure involved to disassemble a valve, examine it, and return it to service cannot be easily quantified. However, since so many valves are normally disassembled during the required 10-year interval, it is not justifiable to increase the exposure.
O Evaluation The visual examination is to determine whether unanticipated severe degradation of the casing is occurring due to phenomena O such as erosion, corrosion, or cracking. However, previous ex-perience during examinations of pumps at other plants has .not shown any significant degradation of casings.
Disassembly of large valves to the degree necessary to examine the internal pressure-retaining surfaces is a major effort, which
'O may involve large personnel exposures. Many of the required valves were disassembled for maintenance during the first interval, and the Code examinations performed. To do this disassembly solely to perform a visual examination of the internal body is impractical.
The licensee has committed to the concept of visual exami-O nation if the valve is disassembled for maintenance. The visual examination specified is to determine whether anticipated severe degradation of the body is occurring due to phenomena such as erosion or corrosion.
The alternate tests proposed by the licensee, along with O visual examination for leakage during system pressure tests under Category B-P and periodic testing of valves in accordance with IWV, will provide an adequate level of safety.
O Conclusions and Recommendations Based on the above evaluation, it is concluded that for the examinations discussed above, adherence to the Code requirements is impractical. It is further concluded that the proposed exami-nations will provide necessary assurance of structural reliability
,0 21 O .
?. , . .,
during this interval. Therefore, relief is recommended as
! requested provided:
(a) visual examination of the pump casing for leakage is
.] conducted in conjunction with system leakage and hydro-static tests under Category 8-P, (b) the required visual examinations are conducted under Category 8-L-2 if a reactor coolant pump is disasserribled for maintenance, 3
(c) periodic inservice testing of the valves is conducted in accordance with IWV, (d) visual examination of the valves for leakage is conducted in conjunction with system leakage and hydro-O static tests under Category 8-P, and (e) the required visual examinations are conducted under Category 8-M-2 if a valve is disassembled for main tenance.
O References References 7 and 9.
O
.O O
lO
'O.
l 22 lO
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3, l II. Ct. ASS 2 COMPONENTS i
A. Pressure Vessels and Heat Exchangers D
- 1. Relief Request ISI-3.1.1, Pressure Retaining Welds in Pressure Vessels, Category C-A, Items C1.10, C1.20, and C1.30
)
Code Requirement Shell circumferential welds at gross structural discon-tinuities (Cl.10), head circumferential welds (C1.20), and 9 tubesheet-to-shell welds (Cl.30) in Class 2 pressure vessels must be volumetrically examined in accordance with Figure IWC-2500-1 or IWC-2500-2 over essentially 100% of the weld length.
Gross structural discontinuity is defined in N8-3213.2.
Examples are junctions between shells of different thicknesses, 9 and shell (or head)-
cylindrical to-flange welds shell-to-conical shell welds.
and head-to-shell junctions, In he t case of multiple vessels of similar design, size,' and service (such as steam generators and heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels.
D Code Relief Request Relief is requested from volumetric examination of 100% of the Code required volume for Residual Heat Removal (RHR) system O heat exchanger circumferential welds.
Proposed Alternative Examination O The ultrasonic examination of the shell and head circumferen-tial welds will be supplemented by a surface examination. The tubesheet-to-shell weld cannot be properly prepared for surface examination nor can the examination be performed due to the tube-sheet studs and nuts adjacent to the weld. In addition to the examinations described above, system pressure tests per Article O IWC-5000 of ASME Section XI will be performed on these welds.
l l
O 23 O .
. - - - - ~ ,,
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1 .
Licensee's Basis for Requesting Relief The shell and head circumferential weld examinations are limited by vessel supports adjacent to these welds. In addition, the ultrasonic examination of the head circumferential weld from
) the head side cannot be performed due to configuration. The examination volume as required by Figure IWC-2500-2 for the tube-sheet-to-shell weld cannot fully be met due to configuration also.
- For Hatch Unit 1, There are three Category C-A circumferen-tial welds in each of the two RHR heat exchangers. These welds and their UT limitations are given below.
IEll - 2Hx-A(B)-1: Shell Head to Upper Shell Ring. These welds y cannot be examined from the shell head side due to the J curvature of the head. Only about 65 in. of a total cir-cumference of 179 in. (approximately 36%) can be examined from the Upper Shell Ring side due to support interference.
1 Ell - 2Hx-A(B)-2: Shell Upper Shell Ring to Lower Shell Ring.
Complete coverage is obtained from the Upper Shell Ring
) side and 0% coverage from the Lower Shell Ring side due to support interference.
lEll - 2Hx-A(B)-3: Lower Shell Ring to Flange. Complete coverage is obtained from the Lower Shell Ring side. Examination from the flange side cannot be performed due to the
) geometry.
For Hatch Unit 2, there are three Category C-A circumfer-ential welds in each of the two RHR heat exchangers. These welds and their UT limitations are given below.
2 Ell - 2Hx-A(B)-1: Shell Head to Upper Shell Ring. These welds cannot be examined from the shell head side due to the curvature of the head. Only about 65 in. of a total circumference of 179 in. (approximately 36%) can be examined from the Upper Shell Ring side due to support interference.
2 Ell - 2Hx-A(B)-2: Upper Shell Ring to Lower Shell Ring.
l Complete coverage is obtained from the Upper Shell Ring side and approximately 35% coverage from the Lower Shell Ring side due to support interference.
)
2 Ell - 2Hx-A(B)-3: Lower Shell Ring to Flange. Complete coverage is obtained from the Lower Shell Ring side. Examination I from the flange side cannot be performed due to the geome try.
)
l l 24
) .
3, ...
Evaluation Relief is requested from 100% coverage of the Code-required volume for Class 2 RHR heat exchanger welds.
O The licensee has committed to performing partial volumetric examinations using current ultrasonic technology. In addition, surface examinations,. to the extent practical, will be performed.
However, the technology of ultrasonic testing is changing g rapidly and significant improvements can be expected during this 10-year interval. It is clearly incumbent on the licensee to keep up with and use volumetric examination tools that are among the most up-to-date comercially available to maximize the quality of examination results, p" The licensee should ensure that the applicable system pressure tests specified in Article IWC-5000 are performed.
Conclusions and Recommendations
- O Based on the above evaluation, it is concluded that for the -
welds discussed above, adherence to the Code requirements is
- impractical. It is further concluded that the proposed examina-tions will provide necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested provided:
O (a) the volumetric and surface examinations are performed to the maximum extent practical, (b) the Code-required system pressure tests are performed.
O Every effort should be made by the licensee to assure that the equipment and procedures used to perform these examinations are among the most up-to-date that are comercially available at the time the examinations are performed.
lO References References 7 and 9.
O.
-o 25 0 .
rri _ _ _ _ _ _ _ _ _ _ _ _
h, .
B. Piping
- 1. Relief Request ISI-3.1.2, Integrally Welded Piping Attachments,
) RHR, Core Spray, HPCI, and RCIC Suction Lines -from Torus, Category C-C, Item C3.20 Code Requirement The surfaces of 100% of each welded attachments of piping required to be examined by Category C-F and C-G are to be examined over the area defined in Figure IWC-2500-5. The welds subject to examination are those welded attachments whose base material design thickness is 3/4 in. or greater.
Code Relief Request Relief is requested from performing the Cooe-required surface
)
examination of the welded attachments on RHR, Core Spray, HPCI, and RCIC suction lines to the torus.
Proposed Alternative Examination Visual examination (VT-1) in accordance with IWA-2211 will be performed to ensure the integrity of these attachments.
) Licensee's Basis for Requesting Relief For Hatch Unit 1, the suction piping is surrounded by re-inforcing ribs which may limit access on one or more sides of the pipe, in particular, when using magnetic particle techniques.
This method is preferred since the torus has a heavy coating of
) paint, and removing the paint and cleaning the surface to perform penetrant examinations would be extremely difficult with the space l imitations. Approximately 80-100% of the RCIC (lE51) and HPCI (lE41) welds can be examined, approximately 50-75% of the Core Spray (lE21) welds, and approximately 25% or less of the RHR (IEll) welds can be examined.
For Hatch Unit 2, the welds are totally inaccessible to i perform surface examinations; therefore, a visual examination will L need to be performed in lieu of the Code requirements.
)
t
) .
26 ,
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Evaluation The reinforcing ribs surrounding the suction piping-to-torus attachment welds interfere with the performance of surface exami-
) nations on these welds. A heavy coating of paint also interferes with examinations.
The licensee is proposing to perform a visual examination (VT-1) of the subject welds at both units. The paint on these welds not only precludes surface examination, but also visual
) examinati on. The licensee has not given sufficient justification (man-hours and radiation exposure) that removal of the paint and performance of a dye penetrant examination is impractical. For the Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) attachment welds at Hatch 1, a magnetic particle examination to the extent practical will provide nec-essary assurance of structural integrity for these two welds.
. However, for the remainder of the attachment welds at Hatch 1 l and all of the attachment welds at Hatch 2, relief cannot be l justified at this time without further information regarding expenditure of man-hours and radiation exposure to perform the required examinations.
)
Conclusions and Recommendations Based on the above evaluation, it is concluded that:
) (a) For the attachment welds in the RCIC (lE51) and HPCI (lE41) systems at Hatch 1, a magnetic particle examination to the extent practical will provide necessary assurance of structural reliability. Therefore, relief should be granted for these two welds at Hatch 1.
) (b) For the attachment welds in the Core Spray (1E21) and RHR (IEll) systems at Hatch 1, and all of the torus-to-piping welds at Hatch 2, the licensee has not provided sufficient justification that the Code requirements are impractical.
Therefore, relief from examination of these welds should not
) be granted at this time.
References
) References 7 and 9.
27
)
n
.. m
~C. Pumps
- 1. Relief Request 15I-3.1.3, Surface Examination of Pressure Retaining
) Welds in Class 2 Pumps, Category C-G, Item C6.10 ( Applies to Hatch Unit 2 only)
O Code Requirement A 100% surface examination of the pump casing welds from one Class 2 planp in each group of pumps that are of similar design, size, function, and service .in a system. The surface examination shall cover the area shown in Figure IWC-2500-8 and may be per-O formed from either the inside or outside surface of the component.
Code Relief Request d Relief from the requirement of performing a surface examination on Class 2 ptsnp casing welds is requested.
Proposed Alternative Examination The Class 2 pump casing welds will be surface examined when disassembled for maintenance.
l
.O Licensee's Basis for Requesting Relief Disassembly of these pumps for the surface examination during the inspection interval, in the absence of other required mainten-ance, represents an unnecessary exposure to radiation and contami-
. nati on. In view of the cost in man-rem and in view of the minimal
.O benefits obtained, it is concluded that this Code requirement does not provide a corresponding benefit in reliability.
Evaluation lO -
This relief request applies to the Core Spray pumps and RHR Pumps on Hatch Unit 2 only. The Hatch Unit 1 pumps have a differ-ent design and do not contain pressure-retaining welds. The pressure-retaining welds in the Core Spray and RHR pumps at Hatch Unit 2 are completely encased in the suction casing and can be O .
28
- 1 -
. ,e nc - - -.n. , _ _ . - . - - _ . , - . . - _ , , , _ , , __ , , _ _ _ _ _ _ , , _ _ _
b
) . ,
! accessed only when the pump is completely disassembled. At least one of the six pumps has been disassembled for maintenance and the welds examined.
)- The licensee has committed to the concept of surface examina-tion if the pump is disassembled for maintenance. As an added assurance of weld integrity, significant leaks from the pump casing to the suction casing would be indicated by drops in both discharge pressure and flow from the pumps during testing required by IWP. The licensee's proposed alternative examination is
}-- practical and should be accepted.
Conclusions and Recomendations Based on the above evaluation, it is concluded that for the pump welds discussed above, adherence to the Code requirements is impr actical . It is further concluded that the proposed examina-tions will provide necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested provided:
}
(a) the required surface examinations are conducted under Category C-G if a Class 2 Core Spray and RHR pump is disassembled for maintenance, and 3 (b) periodic inservice testing of the pumps is conducted in accordance with IWP.
References 3
References 7 and 9.
D D
(
P 29
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g--- -
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h :,: s D. Valves No relief requests.
III. CLASS 3 COMPONENTS No relief requests.
IV. PRESSURE TESTS A. Class 1 Components No relief requests.
)
- 8. Class 2 Components No relief requests.
)
C. Class 3 Components i 1. Relief Request ISI-4.1.1, Class 3 Piping 2 Inches and Smaller, Categories D-A,'D-8, and D-C, Items 01.10, D2.10,-and D3.10 Code Requirement The pressure-retaining components within each system boundary
] shall be subject to a pressure test each inspection period and a hydrostatic test each inspection interval. There are no exemptions or exclusions from these requirements.
IWD-5223 states that the nominal operating pressure of the system or component functional test shall be acceptable as the system test pressure. The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with design temperature of 2000F (930C) or less, and at least 1.25 times the system pressure Psv for systems with design temperature above 2000F (9300). The system pressure Psy shall be the lowest pressure setting among the number of safety or relief valves
] provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure l Pd shall be substituted for Psv-g l
30 3 .
Code Relief Request Relief from performing the required system pressure tests on Class 3 lines 2 in. and smaller unless:
) (a) they are connected to larger lines which will be pressure tested; (b) isolation valves are not provided so that these smaller lines may be isolated in case of leakage.
)
Proposed Alternative Examination Accessible piping 2 in, and smaller will be visually examined under normal operating pressure.
Licensee's Basis for Requesting Relief y These smaller lines have wall thicknesses in excess of what ASME Section III required for retaining internal pressure. Using heavier walled piping in these small lines essentially means they
> are over-designed for the pressure they are retaining, and are not susceptible to the type leakages found during hydrostatic testing.
Evaluation The licensee has requested relief from performing the required pressure tests based on the fact that the subject lines
! are over-designed for the pressure they are retaining.- Require-ments of Section XI should not be mitigated by the over-design conditions of components. Consequently, in the absence of any i practical reason why the required pressure tests cannot be
! performed, relief should not be granted.
i 1
) Conclusions and Recomendations i Based on the above evaluation, it is concluded that the Code-required examinations are not impractical. Therefore, it is concluded that relief from performing the required pressure tests on Class 3 piping less than 2 in, should not be granted.
l I
References Reference 7.
I t
31
) .
) .
- 2. Relief Request ISI-4.1.3, Hydrostatic Test of Plant Service System, Categorles D-A, D-8, and D-C; Items 01.10,- 02.10, and D3.10
)
Code Requirement i
The pressure-retaining components within each system boundary l shall be subject to a pressure test each inspection period and a hydrostatic test each inspection interval.
)
IWD-5223 states that the nominal operating pressure of the system or component functional test shall be acceptable as the system test pressure. The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with design temperature of 2000F (930C) or less, and at least 1.25
) times the system pressure Psv for systems with design temperature above 2000F (930C). The system pressure Psv shall be the lowest pressure setting among the nunber of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure
) Pd shall be substituted for Psy.
Code Relief Requesc
) IWD-5223 of ASME Section XI requires that the Plant Service Water System be tested at a pressure of 1.10 times the design
, pressure. Relief is requested from testing those portions where I
it is necessary to use a butterfly valve 10 in. in diameter or greater as a hydrostatic test boundary valve.
)
Proposed Alternative Examination f
A hydrostatic test will be perforned on those portions of the Plant Service Water System which have a butterfly valve 10 in. in diameter or greater at the normal operating pressure.
l Licensee's Basis for Requesting Relief
)
i Butterfly valves are basically flow control valves and are not intended to be block valves. The normal leakage through these large valves makes it impr actical to attain and maintain the hydrostatic test pressure.
)
j 32
(
) . .
Evaluation The only way to test the subject piping sections to Code requirements is to install hydrostatic pump fittings. Based on j- the fact that the plant would be promptly shut down to repair the seals if they leak, the pressure test should provida adequate assurance of system integrity without compromising public safety.
The relief requested and the proposed alternative examination should be accepted.
I Conclusions and Recomendations Based on the above evaluation, it is concluded that for the examinations discussed above, the Code requirements are imprac-tical. It is further concluded that the alternative examinations
). discussed above will provide necessary added assurance of structural reliability. Therefore, the following are recommended:
(a) Relief should be granted from the Code requirements to hydrostatically test the subject piping sections.
) (b) The alternative tests proposed ;y the licensee should be performed in lieu of the Code testing, and (c) The systems where it is necessary to use a 10-inch butterfly valve as a hydrostatic test boundary valve should receive a visual examination (VT-2) during normal
} operating pressure.
j References References 7 and 9.
h
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33
) .
l
'V. GENERAL
- 1. . Relief Request ISI-8.1.1, U1trasonic Calibration B1ocks
)
Code Requirement Calibration blocks for ASME Code Section XI examinations are required to meet Section XI Appendix III,- or Section V require-j ments, as specified in Section XI, IWA-2232.
E Code Relief Request j Relief is requested to use existing pipe weld calibration blocks that are not certified to the above current requirement.
Proposed Alternative Examination
) The majority of existing Hatch basic calibration blocks used for pipe weld examinations fabricated with diameters, thicknesses, and side-drilled holes in accordance with the 1974 Code Edition will be used for the second 10-year ISI program.
D Licensee's Basis for Requesting Relief For the two primary reasons listed below, these same basic calibration blocks will be used to. provide the most meaningful and thorough examinations possible:
3 (1) Side-drilled holes as calibration reflectors result in a more sensitive ultrasonic examination than one using notches.
, (2) Correlation of ultrasonic data with previous examinations as 3 required by Subarticle IWA-1400 of Se::cion XI makes it neces-sary that these basic calibration blocks be used so future examination results can be correlated with past results.
] Evaluation Technical evaluation of the proposed calibration blocks prior l
to the preservice and first-interval examination indicated that l examination effectiveness would not be reduced by use of the pro- .
- posed calibration blocks. An important feature of the overall ISI
) ,
1 34 h .
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). .
program is that past inspections serve as a baseline by which inservice examination results are evaluated. Accordingly, it is appropriate to use methods during inservice inspection which are consistent with those used previously, provided the previous examination methods were technically acceptable.
Conclusions and Recommendations Based on the above evaluation, it is concluded that for the methods discussed above, adherence to the code requirements is impr actical . It is further concluded that the prorosed methods will provide necessary assurance of structural reliability during this interval. Therefore, relief is recommended as requested.
)
References j Reference 7.
E
)
)
l J
35
)
- 2. Relief Request ISI-2.1.4, Ultrasonic Examination Requirements of I
Austenitic and Dissimilar Metal Pipi,ng Welds
.- Code Requirements IWA-2232 states that:
l (a) Ultrasonic examination of Class 1 and Class 2 vessel welds in ferritic material greater than 2 in. (51 nin) in thickness .
}
shall be conducted in accordance with Article 4 of Section V, amended as follows:
(1) The requirements of T-431, Instrument Calibration, shall be verified at th.e-beginning and end of the weld exami-nation performed on a vessel during one outage.
}l (b) Ultrasonic examination of Class 1 and Class 2 ferritic steel piping systems shall be conducted in accordance with Appendix III, amended as follows:
(1) For examination of welds, reflectors that produce a 3 response greater than 50% of the reference level shall be recorded.
(c) If the requirements of (a) and (b) above are not applicable, the ultrasonic examination shall be conducted in accordance
] with the applicable requirements of Article 5 of Sectica V, amended as follows:
(1) For examination of welds, reflectors that produce a response greater than 50% of the reference level shall be recorded.
] (2) For examination of welds, all reflectors which produce a response greater than 100% of the reference level shall be investigated to the extent that the operator can determine the shape, identity, and location of all such reflectors in terms of the acceptance-rejection standards
] of IWA-3000.
(3) The size of reflectors shall be measured between points which give amplitudes equal to 100% of the reference l evel .
3 (4) The material from which the basic calibration block is f abricated shall be one of the following:
(a) nozzle dropout from the component; l (b) a component prolongation; or
- (c) material of the same material specification, product
'. form, and heat treatment as one of the materials being joined.
l 3 .
36 l
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) .. .
(5) Ultrascnic examination shall be performed in accordance with a written procedure. Each procedure shall include at least the information required by T-523 of Article 5 of Section V.
)
Code Relief Request Relief is requested from the requirements of austenitic and dissimilar metal piping welds to be examined in accordance with
) Article 5 of ASME Section V.
Proposed Alternative Examination Austenitic and dissimilar piping welds will be examined in accordance with Appendix III.
Licensee's Basis for Requesting Relief
)
Article 5 of ASME Section V does not provide the detailed guidance necessary to examine austenitic and dissimilar metal piping weld with the exception of austenitic piping welds which have been repaired by weld overlay. These clad overlaid piping j welds will be examined in accordance with Article 5 of Section V and Appendix III of Section XI.
Since ferritic piping uelds will be examined per Appendix III i of Section XI and to provide consistency, austenitic and dissimi-l lar metal piping welds will also be examined in accordance with Appendix III.
Evaluation Ultrasonic exa.mination of austenitic and high nickel alloy
)
welds is usually more difficult than in ferritic materials, because of the wide variations that may occur in the acoustic properties of austenitic and high nickel alloy welds, even those in alloys of j the same composition, product form, and heat treatment. It may, t therefore, be necessary to modify and/or supplement the provisions of Article 5 of ASME Section V as allowed in accordance with para-graph T-ll0(c) of Article 1.
l Appendix III defines the ultrasonic examination methods, equipment, and requirements applicable to Class 1 and 2 ferritic steel piping systems. However, Article III-1000 refers to 37
) .
5
~
Supplement 7 of Appendix III for dissimilar metal welds and austenitic steels. Supplement 7 states:
y (a) Because of the inherent coarse-grained structure, ultrasonic examination of the following welds may be subject to marked variations in attenuation, velocity, reflection, and refrac-tion at grain boundaries:
(1) high alloy steels;
- 2) hich nickel alloys;
- 3) dissimilar metal welds between combinations of (1) and (2) above and carbon or low alloy steels.
(b) However, these rules may be used, with the following reconmended modifications, when ultrasonic examination of these welds is performed.
)-
(1) III-4410 Beam Angle - Add: other angles may be used where metallurgical characteristics impede effective use of 45-degree angle beam.
(2) Table III-3430-I Calibration Notches - Substitute: depth 10% of t.
) (3) Figure III-3230 may not apply to austenitic materials.
Article 5 of Section V allows documented modification of ultrasonic examination requirements. Appendix III, Supplement 7 gives rules for austenitic steels and dissimilar metal welds.
) Therefore, .the licensee's request to perform austenitic steel and dissimilar piping weld ultrasonic examinations in accordance with Appendix III is in conformance with the requirements of both Article 5 of Section V and Appendix III. Therefore, relief is not l necessary provided that the licensee performs the subject examinations in accordance with Supplement 7 of Appendix III.
Conclusions and Recommendations Based on the above evaluation, it is concluded that the
} relief request is not necessary and, therefore, should not be granted.
) References Reference 7.
l 38 3 .
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- 3. Relief Request ISI-8.1.2, Change-Hatch 2 ISI Second Ten-Year Interval to Segin on Jaauary 1,1986 Y
Code Requirement IWA-2420 Inspection Procram 8: The inspection -intervals g .- shall comply with the following except as modified by IWA-2400(c):
1st Inspection Interval - 10 years following initial start of power unit commercial operation 2nd Inspection Interval - 10 years following the 1st inspection 3} interval 3rd Inspection Interval - 10 years following the 2nd inspection interval 4th Inspection Interval - 10 years following the 3rd inspection
- g interval.
IWA 2400(c) states that each inspection interval may be decreased or extended (but not cumulatively) by as much as one year. For power units that are out of service continuously for g six months or more, the inspection interval during which the outage occurred may be extended for a period equivalent to the ou tage.
g Code Relief Request
! It is requested to begin the Hatch 2 second 10-year interval "
on January 1,1986, rather than September 1989.
O Proposed Alternative Examination l The Hatch 2 ISI plan for the first interval will be moved ahead by 40 months, to allow common interval dates for Hatch Units 1 and 2. The Hatch 2 RPV welds scheduled for examination during
- O ~ the final 40 months of the first interval will be scheduled for l the first 40 months of the second interval.
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- a. , _ _ _ _ . _ . - _ _ _ _ . _,
I Licensee's Basis for Requesting Relief Starting the Hatch 2 inservice inspection second 10-year interval 40 months early would result in compliance with a later,
} more stringent code, while not reducing the number of required ;
examinations for the first interval. Additionally, compliance l with the same code for both units would enhance detection of l generic problems and would also reduce costs of maintaining two separate programs.
D Evaluation A move to start the second 10-year interval for Hatch 2 on January 1,1986, would require compliance with the 1980 Edition of the ASME Code Section XI, Winter 1981 Addenda, instead of the
} Winter 1980 Addenda during what would have been the last 40 months of the first interval. The newer Code is more stringent, and a more comprehensive inspection would be achieved. In addition, compliance with the same Code for both units will enhance the possibility of detecting generic problems and would also reduce a costs involved in maintaining two separate programs. The RPV welds will also be examined at the same point in time as was planned in the earlier ' program (i .e., those welds that were scheduled during the third 40-month period will be examined during the first 40-month period of the new 10-year interval).
O Conclusions and Recormiendations Based on the above evaluation, it is concluded that changing the start date of the second 10-year interval for Hatch 2 will not J have an adverse effect on providing necessary assurance of struc-tural reliability during this interval. Therefore, relief should be granted provided that the licensee's proposed alternative examination is performed.
O References References 7 and 9.
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- 4. . Relief Request ISI-8.ll3,2 Reporting Requirements of Procedural
. Changes Affecting Inservice' Inspection-Programs Code Requirement According to 10 CFR 50.59, the licensee should prepare plans and schedules for inservice inspection and submit them to the enforcement and regulatory authorities having jurisdiction at the plant site.
Code Relief Request
- j. Relief is requested from ISI plan updating when procedural changes in ISI plans are made, in accordance with Plant Technical Specifications and the Safety Analysis Report, provided that they l
do not involve any unreviewed safety questions.
D 7
Proposed Alternative Examination L
l Records of all changes will be maintained. Changes affecting L ASME Section XI requirements shall be complied with. Changes will l be incorporated 'into the ISI Program Plan (s), whenever a need for
} their update is warranted.
I i
Licensee's Basis for Requesting Relief j 10 CFR Part 50.59 allows changes to a nuclear facility and procedural changes in accordance with Plant Technical Specifica-tions and the safety analysis report, without prior- approval from the NRC, provided there are no unreviewed safety questions. The facility's Inservice Inspection Program / Plan should be revised to include these changes.
J Changes such as valve operator, stroke time, pump performance, welds, etc., made in compliance with 10 CFR 50.59, which affects the ASME Section XI requirements, may not be shown on the existing revision of the ISI program plan document.
Records of these changes will be maintained and any changes
] in ASME Section XI requirements pertaining to these changes will be complied with. However, a revision to the ISI Program Plan (s) for every minor change is unrealistic, costly, and a time consuming task. Any delay in revising the ISI Program Plan (s) will not endanger public health and safety.
)
l j 41
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} .- . .
Evaluation For changes made to welds during the interval, such as y replacement or. repair welds, IWA-7530 requires that prior to return of the plant to service, a preservice inspection shall be made in accordance with IWB-2200, IWC-2200, IWD-2200, or IWF-2200. Federal Register 10 CFR 50.55a(g)(5)(iv) states that where an examination or test requirement by the Code or addenda is determined by the licensee to be impractical and is not included in the revised inservice inspection program as permitted by 10 CFR 3- 50.55a(g)(4), the basis for this determination shall be demon-strated to the satisfaction of the NRC not later than 12 months after the expiration of the initial 120-month period of operation from the start of facility comercial operation and each sub-sequent 120-month of operation during which the examination or test is determined to be impractical.
}
If the above requirements are met, the licensee is not required to submit a revision of the ISI program for each weld repair and replacement.
)
Conclusions and Recomendations Based on the above evaluation, it is concluded that the relief request to delay revising the ISI program plan for minor
] changes is not required, provided 10 CFR 50.55a(g)(5)(iv) and IWA-7530 are complied wi th.
i References Reference 7.
)
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D 42 J .
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- 5. Relief Request ISI-8.1.4,-Requirements of Section XI, Subsection IWE
)
Code Requirement There are no requirements in 10 CFR 50.55a to examine the containment in accordance with Subsection IWE.
)
Code Relief Request Relief is requested from the requirements of Subsection IWE.
)
Proposed Alternative Examination None.
)
Licensee's Basis for Requesting Relief
, Federal Register, Vol. 48, No. 2.8/ Monday, February 7,1983,
} Page 5532 relative to 10 CFR Part 50 Codes and Standards for Nuclear Power Plants; Item No. 4, indicates that Subsection IWE,
" Requirements for Class MC Components of Light Water Cooled Power Plants," was added to Section XI by Winter 1981 Addenda. However, 10 CFR Paragraph 50.55a presently only incorporates those portions of Section XI that address the ISI requirements for Class 1, 2, D and 3 components and their supports. The regulations do not currently address the ISI of containments. Since this amendment is only intended to update current regulatory requirements to include the latest Code addenda, the requirements of Subsection IWE are not imposed upon licensees by the NRC as a result of this amendment.
D Evaluation The requirements of Section XI Subsection IWE have not been
] referenced in 10 CFR Paragraph 50.55a. Therefore, relief from the requirements of IWE is not required.
l i.
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i
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Conelusions and-Recommendations g- Based on the above evaluation, it is concluded that relief from the requirements of IWE is not required and therefore should not be granted.
D References Reference 7.
D D
D J
J e
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- 6. Relief Request ISI-8.1.5, Due Date for 0wner's Data Report for j Inservice Inspection, Form NIS-1 Code Requirement Paragraph IWA-6230 of Section XI requires that the Owner's Data Report for Inservice Inspection, Form NIS-1, shall be filed with the enforcement and regulatory authorities within 90 days of
] the completion of the inservice inspection.
Code Relief Request-
)
- Relief from the ISI completion date relative to the reporting requirement and the due date is requested.
Proposed Alternative Examination
]
Form NIS-1 will be submitted to the NRC within 120 days after completion of the outage in which the examinations were performed.
J Licensee's Basis for Requesting Rt. lief The 90-day due date from the completion of the inservice inspection is unrealistic to prepare the NIS-1 Form and to have the multiple reviews required. The preparation of the NIS-1 Form
)- itself requires almost 90 days with at least another 30 days needed for the review by site personnel and the Inspector.
Evaluation D The licensee has requested relief from submitting the NIS-1 Form in the required 90 days from completion of the inservice in spection. Other licensees have found that 90 days is an adequate period of time to file the NIS-1 Form. Thus, the licensee't, basis for the request is not consistent with industry-wide practice.
] Therefore, relief from the 90-day time period for submittal of the NIS-1 Form is not recommended.
t 45
- Conclusions and Recommendations Based- dn the above evaluation, it is concluded that relief j
i from the 90-day submittal requirement of the NIS-1 Form is not warranted and should not be granted.
) References Reference 7.
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l_ VI. COMPONENT SUPPORTS f -1. Relief Request ISI-5.1.1, Acceptance Criteria for Springs- and
[ Snubbers, Category F-C, Item F3.50 J
I Code Requirements Acceptance standards for component support structural j integrity given in IWF-3410 are as follows:
(a) Component support conditions which are unacceptable for con-tinued service shall include the following:
(1) deformations or structural degradations of fasteners,
, springs, clamps, or other support items; (2) missing, detached, or loosened support items; (3) arc strikes, weld spatter, paint, scoring, roughness, or general corrosion on close tolerance machined or sliding surfaces; (4) fluid loss beyond specified limits or lack of fluid
^)
indication (hydraulic snubbers only);
(5) improper hot or cold positions (snubbers and spring supports).
O Code Relief Request Relief is requested from the requirement that snubbers and spring supports be in a proper hot or cold position.
O Proposed Alternative Examination The visual examination will verify that the indicator falls within the operational limits of the spring or snubber.
O Licensee's Basis for Requesting Relief There are no exact design positions on the scales for spring supports or snubbers, but an operational range where the indicator
,O should be located. Also, the temperature used in the analysis for the hot and cold positions may not correspond with the temperature the system is at during the inspection.
i
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),, .-
Evaluation .
The licensee is proposing to verify that the spring support is within the operable range plus acceptable tolerances, i.e.,
) within analyzed hot and cold load settings. It is assumed that an analyzed hot and cold setting can be determined for each support.
The licensee states that the analyzed hot and cold load positions-may have been calculated using conservative temperatures, and that the system may not be at the analyzed temperature during the inspection.
It is recognized that spring support-analyzed hot and cold positions may have been calculated using conservative tempera-tures. However, the licensee should verify that the observed position of the spring supports is consistent with the temperature of the system during the inspection.
)
t Conclusions and Recommendations Based on the above evaluation, it is concluded that for the
] spring supports discussed above, the Code requirements are im-practical. Therefore, relief should be granted provided the
-licensee verifies that the position of the spring supports is consistent with the temperature of the system during the inspection.
)
References References 7 and 9.
)
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- 2. Relief Request 151-5.1.2, Snubbrr Testing Program IWF-5300 and IWF-5400 Code Requirements IWF-5300: Inservice tests for snubbers 50 kips (22,680 kg) or greater:
In the course of preparation.
IWF-5400: Inservice tests for snubbers less than 50 kips (22,680 kg):
(a) Inservice tests shall be performed during normal system
) operation or plant outages.
(b) A representative sample
- of 10% of the total number of non-exempt snubbers whose rating is less than 50 kips shall be tested each inspection period. Each representative sample shall consist of previously untested snubbers. After all nonexempt snubbers in the plant have been tested, the test shall be repeated taking the same snubbers (or their replace-ments) in the same sequence as in the original tests. These tests shall verify that:
(1) during low velocity displacement, the specified maximum
> drag or free movement force will initiate motion of the snubber rod in both tension and compression; (2) activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression; (3) snubber bleed, or release rate where required, is within the specified range in compression or tension. For units specifically required not to displace under con-tinuous load, the ability of the snubber to withstand load without displacement shall be demonstrated.
(c) Snubbers that fail the inservice tests of (b) shall be repaired in accordance with IWF-4000 and retested. An additional sample of 10% of the total nunter of snubbers shall also be tested at that time. Additional semple testing shall be continued until all units within the sample have met J the requirements of (b).
- A representative sample shall include snubbers from various J locations taking into consideration service and environment.
89 49
h....
(d) Components whose supports fail the test requirement shall be evaluated to ensure that the supported component has not been impaired.
Inspection and test results shall be recorded for each
) (e) snubber.
Code-Relief Request GPC requests relief from the Section XI inservice testing requirements in IWF-5300 and IWF-5400.
I Proposed Alternative Examination
)
The functional testing requirements of Technical Specifica-tion 3/4.6.L for Unit 1 and 3/4.7.4 for Unit 2 will be implemented in lieu of the Code requirements of IWF-5300 and IWF-5400. VT-4 examinations will also be performed in accordance with the Technical Specifications.
)
Licensee's Basis for Requesting Relief The snubber inservice testing requirements in IWF-5300 and in
] IWF-5400, Table IWF-2500-1 of Section XI are not complete and Hatch has already implemented a comprehensive snubber testing program.
Hatch's snubber testing program is defined in the Plant Technical Specifications, Section 3/4.6.L for Unit 1, and 3/4.7.4 for Unit 2.
)
Evaluation A review of the Plant Technical Specifications cited indicated that:
) (1) snubbers found faulty or inoperable are to be repaired or replaced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of discovery, (2) a specific schedule varying from 31 days to 18 months has been established for the examination of each type of snubber, l (3) at least 10% of the total of each type of snubber will be j functionally tested either in place or in a bench test. For each snubber type that does not meet the functional test acceptance criteria, an additional 10% of that type of snubber will be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested or sampling statistics establish 95% reliability,
)
50 1
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(4) - visual inspection criteria, transient event inspection, and functional tests have been established, I. (5) acceptance criteria have been established, (6) failure analyses criteria have been established, (7) functional testing of replaced or repaired snubbers have been established, and (8) a snubber service life replacement program has been es tablished.
p The Plant Technical Specifications make r.o distinction between the testing of snubbers with a capacity of 50 kips or
! greater and those with capacities of less than 50 kips. The examination procedures presented cover all snubbers, not just those with ratings less than 50 kips. The Code requires that 10%
of all snubbers with capacities less than 50 kips be tested. If, y' in the testing of snubbers, the licensee were to ensure that at least .10% of all snubbers with ratings less than 50 kips are examined, the program would be more definitive than that required by Code. Visual and functional tests are more specific than that required by Code. The functional testing of repaired / replaced snubbers is conducted in addition to the examination of a random j
selected sample. If a sample of snubbers proves the unaccept-ability of the class of snubbers tested, then an engineering i
evaluation of that group is conducted. Since there is no reason l the licensee should not be able to meet the Code's criterion,
( relief is not necessary.
)
Conclusions and Recommendations Based on the above evaluation, there is no reason the
!. licensee's snubber examination should not be able to comply with
) Code, and no relief is required. The licensee should ensure that l
at least 10% of all snubbers with ratings less than 50 kips are l examined.
l 1
References Reference 7.
l l
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- 4. REFERENCES
)
- 1. Science Applications, Incorporated, Edwin I. Hatch Nuclear Plant Units 1 and 2, Inservice Inspection Program, Technical Evaluation Report, SAI Report No. 186-028-25, Septemoer 2, 1982.
- 2. J. F. Stolz (NRC) to J. T. Beckham, Jr. (GPC), July 29, 1983; Safety
} Evaluation Report on first interval inservice inspection plan.
- 3. J. T. Neckham, Jr. (GPC) to J. F. Stolz (NRC), August 12, 1983; revised first interval inservice inspection program.
- 4. L. T. Gucwa (GPC) to J' F. Stolz (NRC), July 18, 1985; updated
} .
inservice inspection program for the first interval for Hatch Unit 1.
- 5. H. R. Denton (NRC) to J. T. Beckham, Jr. (GPC), November 7,1985; Safety Evaluation Report on updated inservice inspection program for Hatch-1 first interval.
) 6. L. T. Gucwa (GPC) to D. Muller (NRC), November 27, 1985; extent of examinations conducted to the 1980 Edition with Addenda through Winter i
1980 for the Hatch-1 first interval.
- 7. L. T. Gucwa (GPC) to J. F. Stolz (NRC), June 25, 1985; second interval p inservice inspection program for Hatch Units 1 and 2.
l 8. Communication, NRC to GPC, Decenber 23, 1985; request for additio,nal information on inservice inspection program.
1
- 9. L. T. Gucwa (GPC) to 0. Muller (NRC), February 7,1986; response to
)
l request for additional information.
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