ML20197G004
| ML20197G004 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/11/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20197F988 | List: |
| References | |
| NUDOCS 9712300353 | |
| Download: ML20197G004 (28) | |
Text
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L p \- UNITSO STATES '
- NUCLEAR REGULATORY COMMISSION j I wasHWeToN, D.C. 30seHept r
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RAFETY EVALUATION SY THE OFFICE OF NUCLEAR REACTOR REGULATION FIRS'l 10 YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN. REVISION 4 AND ASSOCIATED REQUESTS FOR RELIEF i COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION. UNITS 1 AND 2 l
, DOCKET NOS. STN 50-456 AND STN 50457 l ,
1.0 INTRODUCTION
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l Th6 Technical Specifications (TS) for Braidwood Station, Units 1 and 2, state that the inservice inspection of the American Society of Mechanical Engineers (A8ME) Code Class 1,2, and 3 i components shall be performed in accordance with Section XI of the ASME Boiler and Pressure :
1 Vessel (B&PV) Code (Code) and applicable addenda as required by 10 CFR 50.55a(g), except :
whors specific written relief has been granted by the Commission pursuant to 10 CFR ,
s 50.55a(g)(6)(i). In 10 CFR 50.55a(s)(3) it states that altematives to the requirements of j paragraph (g) may be used, when authortrod by the NRC, if (i) the proposed altomotives would ;
p, ovide an acceptable level of quality and safety or (ii) compliance with the specified l' requirements would result in hardship or unusual difficulty without a compensating increase in
, the level of quality and safety.
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Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supporis) shall meet the requirements, except the design and access provisions and the preservice examination requiremnts, set forth in the ASME Code,Section XI, " Rules for inservice inspection of Nuclear Power Plant Components," to the extent practical within the !
limitations of design, geometry, and materials of construction of the components. The l regulations require that inservice examination of components and system pressure tests
- conducted during the first 10 year interval and subsequer.1 intervals comply with the i requirements in the latest edition er,d addenda of Section XI of the ASME Corte incorporated by ;
reference in 10 CFR 50.b5a(b) 12 months prior to the start of the 120-month interval, subject to i
- _ the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the Braidwood Station,' Units 1 and 2, first 10 year inservios inspection (181) interval is the 1983 Edition through the Summer 1983 Addenda.
Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an ,
examinalien requirement of Section XI of the ASME Code is not practical fodts facihty. -
_ information shall be submitted to the Commission in support of that determination and a ;
F request mode for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose allemative requirements that are determined to be authorized by law, will not endanger
- life, proper 1y, or the common defense and secunty, and are otherwise in the public hiterest, 9712300353 971211
- PDR ADOCK 05000456 ,
i O PDR - .i
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2-giving due consideration to the burden upon th6 licensee that could result if the requir6ments were imposed.
By letter dated October 8,1996, the Commonwealth Edison Company (Comed, or the licensee), submitted to the NRC its First 10 Year Inservice inspection Interval Program Plan, Revision 4, and associated requests for relief for Braidwood Station, Units 1 and 2. The licensee provided supplemental information by letters dated March 28,1997, and May 19,1997.
The October 8,1990 submittal contained a relief request, NR 29, related to the methods used to examine the reactor pressure vessel (RPV) welds. By letter dated February 25,1997. Comed requested an expedited review of NR 29. The rtaff approved NR 29 by letter dated May 13, 1997.
2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its First 10-Year Inservice Inspection Interval Program Plan, Revision 4, and associated requests for relief for Braidwood Station, Units 1 and 2. Based on the information cubmitted, the staff adopts the contractor's conclusior,s and recommendations presented in the Technical Letter Report (TLR) attached.
As a result of Resistance Temperature Detewon modifications, the population of welds in several ASMC Class systems were changed. These changes, and corresponding inspection sampling, have been incorporated in the ISI Plan Summary tables of Revision 4. No deviations from ASME ISI requirements were noted as a result of the changes in weld populations. Other non-technical changes, including the incorporation of previously approved requests for relief and those originating from typographical errors, are not evaluated in this Safety Evaluation. The inforrnation provided by the licensee in support of the new requests for relief has been evaluated and the bases for disposition are documented below.
Relief Reauest NR 25: Section XI, Table IWB 2500-1, Examination Category B H, item B8.20, requires 100% surface examination of integral attachment welds to the pressurizer as defined by Figure IWB 2500-15. Pursuant to 10 CFR 50.55a(g)(5)(iii), the licerisee requested relief from the Code-required 100% surface examination of the following pressurizer integral attachment welds: (Unit 1) 1PZR-01 PSL-01,1PZR 01 PSL 02,1PZR 01 PSL 03, and 1PZR 01 PSL-04; (Unit 2) 2PZR-01 PSL 01,2PZR 01 PSL 02,2PZR 01 PSL-03, and 2PZR 01 PSL-04, in lieu of the Code requirements, the licensee has proposed to perform a VT 1 of the upper surfaces of the two accessible lugs when the insulation panels (removable type) are removed.
The licensee estimated that 5.74" per accessible lug (4" of the top and .87" of each side) will be achievable when the insulation panels are removed. This is approximately 28.7% of the total exam area for one lug. Also, a best effort surface inspection (liquid penetrant) will be performed on those portions of the lug that are inspectable when the insulation panels are removed. In conjunction with the above proposed altemative technique, the periodic VT 2 examinations in accordance with the requirements of ASME Section XI, Table IWB-25001, Examination
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i Category B P and applicable reactor coolant system monitoring requirements spooined in the ,
TS will provide reasonable assurance of continued structuralintegrity of the pressuriser shell.
l The Code requires 100% surface examination of the subject pressurtser inugral attachment wolds. However, seismic restraints sarrounding the attachment welds provide minimal
- clearances for access to perform the required examinations. The restraints are embedded in i concrete and would require major plant mooifications to allow periodic removal for access to the pressuriser integral shachment wolds, insulation on the lower shell limits necess to the wold 3
' attachment area. This insulation was not designed to facilitate hs removal, and significant exposure to plant personnel would insult if the licensee is required to inove it to perform the !
required surface examinations. Further, it is expected that if this insulation were to be removed, 1
full access to the attachment wolds would be limited due to the seismic restraint configuration +
stated above. For these rcasons, the staff has determined that the surface examinations are -
impractical to' perform to the extent required by the Code. ,
! ' By reinoving panels of insulation on the upper shell, the licensee obtained approximately 28.7% l coverage of each integral attachment weld, in addition, the licensee proposed to perform a
_ VT 1 visual examination of the remaining accessible areas. Based on the percent of Code-reouired surface examination coverage of tained in combination with the VT 1 visual i examination, and exrminations performed on other Class 1 integral attachment welds, it is conclJded thst degradation, if present, would be detected, providing reasonable assurance of structuralintegrtty. Therefore, relief muy be granted pursuant to 10 CFR 50.55a(g)(6)(l). ,
Relief Reneemt NR q: Section XI, Table IWC 25001 Examination Category C-C, item C3.20, requires 100% surface examination of in egral attachment welds as defined by Figure IWS-2500 5. In accordance with 10 CFR M 55a(g)(5)(iii), the licensee requested relief, from the extent of surface examinations of the following examint%n areas as required by the Code:
(Unit 1) 1FW-06 33,1FW 10 2g,1FW 112g,1FW 12 2g, RH 03 378,1RH-04 73A, 1RH 05 21B,1RH-07 25A,1RH 08-02A,1RH-09 45A,1SI 04 02A,1SI 12 23A, and 18126 09A, (Unit 2) 2FW 06-01,2FW 10 24,2FW 1125,2FW-12 25,2RH-04-03,2RH-05 35, 2RH-06-03,2RH-09 28,2Sl44-03,2SI 121gA, and 2SI 26-03.
As an altemative examination in lieu of the Code requirements, the licensee proposed that when ,
a weld is scheduled tor insper'!on, a surface examination of the scoessible weld on the exposed outside surface of the penetration w.ll be performed !n conjunction with the above proposed altnmative technique, ti,e periodic VD2 examinations in accordance with the requirements of ASME Section XI, Table IWC-25001, Examination Category C H will provide reasonable assurance of continued structuralintegri'y of the piping systems.
The Code requires that the subject Class 2 integral attachment welds receive 100% surface
. examination. However, a portion of each welded attachment is located inside a flued head containment penetration, restricting accessibility such that performing the Code-required i- surface examinations is impractical. To perform these surface examinations, modifications or "
replacement of the components with those of a design providing for complete access would be
. necessary, imposition of this requirement would cause a considerable burden on the licensee.
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4 The licensee proposes to perform the Code-required surface examinations to the extent predical, i.e., on the accessible portion of each attachment weld. The accessible areas ,
comprise approximately 50% of the welds for each integral attachment weld. Based on the number of integral attaements that will be examined, and the coverage obtained on each, it is concluded that generic degradation, if present, will be detected. As a result, reasonable assurance of structural integrity will be provided. Therefore, pursuar ! to 10 CFR 50.55a(g)(6)(l),
the licensee's request for releef may be granted.
Relief Maquest NR 27. Revlalon 1 (Unit 1 Only): Section XI, Table IWB 25001, Examination Category B A, item 83.90, requires 100% volumetric examination of reactor pressure vessel noule to shell wolds as defined by Figure IWB 2500 7.
Nett: The onipinal request also included nonle welds on Unit 2, however, the licensee revised the request in a letter dated May 19,1997, to include only the welds on Unit 1.
A separete request will be submitted upon completion of the RPV examinations Ibr Unit 2.
Pursuant to 10 CFR 50.55a(g)(5)(iii) the licensee requested relief from the Code required 100%
coverage of the following reactor pressure vessel noule to shell welds: 1 RV-01-006, 1RV 01009,1RV 01010, and 1RV 010 013. As an attemative examination in lieu of the Code requirements, the licensee proposed to examine to the fullest extent practical using the available underwater volumetric inspection techniques of the Reactor Vessel outlet (tot leg) noule welds.
The code requires 100% volumetric examination of the subject RPV noule to vessel walds.
However, complete examination is restricted by their geometric configuration which makes the ,
100% volumetric examination impreciical to perform for these areas. To gain access for examination, the RPV noules would require design modifications, imposition of this requirement would create an undue burden on the licensee.
The licensee has examined these welds to the extent practical, obtaining significant (approximately 84%) coverage of each noule to vessel weld. In addition, noule-to vessel welds on various other Class 1 components are being examined as required by the Code.
Therefore, any existing pattems of degradation would have been detected by the examinations that were completed and reasonable assurance of the structuralintegrity has been provided.
Based on the impracticality of meeting the Code coverage requirements for the subject noule-to-vessel wolds, and the reasonable assurance provided by the examinations that were completed on these and other Class 1 noules, the licensee's request for relief may be granted pursuant to 10 CFR 50.55a(g)(6)(l).
J Relief Reauest NR 28. Revision 1 (Unit 1 Only): Section XI, Table IWB-2500-1, Examination Category B D, item B3.100, requires 100% volumetric examination of reactor pressure vessel )
noule inside radius sections as defined by Figure IWB-2500 7. i
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3 5 Nein: The orspinal toquest also induded nouie inner redius sections on Unit 2; ,
]. however, the licensee reviseo' the request in a letter deted May it,1997, to indude only ;
- the areas on Unit 1. A separate request will be submitted upon completion of the RPV l l examinadons tbr unit 2. ,
Pursuant la 10 CFR 50.55a(g)(5)(iii), the licensee requested relief for the C+t :Whd 100% l coverage of the b"cds reactor pressure vessel noule inner redeus sections: 1RV-01015, ;
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- 1RV-01016,1RV-01-01g, and 1RV-0120. As an altomative examination in lieu of the Code requirements, the licensee has proposed to use the reactor vessel inlet (cold log) noule inner i rad 6us section examined to the fullest extent practical using the available underwater volumetric !
- 4. Inspection techniques, in oor(unction with the partial volumetric examination, a supplemental ;
- VT 1 of the noule inner radius area was conducted from the interior of the Reactor Vessel i L using underwater camera equipment.
The Code requires 100% volumetric examination of the subject RPV noule inside radius sections. However, complete examination is restricted by their geometric configuration which makes the 100% volumetric examination impractical to perform for these areas. To gain access ,
4 for examination, the RPV noules would require design modifications, imposition of this
] requirement would create an undue burrian on the licensee.
The licensee has examined these welds to the extent practical, obtaining sigreificant !
4 (approximately 82%) coverage for each noule inside radius section. In addition, other Class 1 .
- noules inside radii are being examined as required by the Code. Therefore, any existing pattoms of degradation wouid have been detected by the examinotions that were completed and reasonable assurance of the structuralintegrity has been provided.
Based on the impracticality of meeting the Code coverage requirements for the subject noule j inside radius sections, and the reasonable assurance provided by the examinations that were >
- completed on these and other Class 1 noule in$lde rodil, relief may be granted pursuant to
Raquest for Relief No. NR-30. Revision 1: Section Y.1, Table IWC 25001, Examination
, Category C H, items C7.30, C7.40, C7.70 and C7 CO, require a system leakage test at i
operating pressure for pressure retalning piping and valves. As an altemative in lieu of the
, Code, the licensee proposed a perform 10 CFR 50, Appendix J leakage tests as an optional
- attemative to the Section XI required pressure test on the subject primary reactor containment ,
i-penetration piping and associated valves. When implementing the Appendix J tests and i invoking this relief request, peak design pressure and procedures for the detection and location of through-wall flaws will be used. Pursuant to 10 CFR 50.55a(s)(3)(i), relief is requested on the L basis that the proposed altemative would provide an acceptable level of quality and safety. ;
- The Code requires that a VT 2 visual examination be performed during system pressure testing i f for Class 2 pressure-rotalning piping. As an altemative, the licensee proposes to implement the
requirements of 10 CFR 50, Appendix J, for the subject containment penetration piping. This
' altemative is contained in ASME Cooe Case N-522, Pressure Tes6ng of Containment .
PWnetration Piping. The subject piping is classified as Class 2 becausa it penetrates primary -!
reactor containment and is considered an extension of the containment vessel, Slik:e the piping l
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l on olther side of these penetrations is non class, the requirements of Appendix J are more !
appropriate than those of Examination Catego#y C H. Appendix J pressure tests vertfy the leak- !
tight integrity of the primary reactor containment and of systems and components that penetrate t containment by local leak rate and integrated leak rate tests, in Appendix J pressure tes's, ;
cordainment isolation valves and connecting pipe segments must withstand the peak calculated containment intomal pressure related to the maximum design pressure. . In addition, Appendix J !
lest frequencies provide assurance that the cords'nment pressure boundary is being maintained ,
at an acceptable level while monitoring for deterioration of seals, valves, and pip!ng. l The licensee has committed to perform the Appendix J testing at no less than the peak
. calculated containment pressure and will use procedures to detect and locate through-wall !
flaws. The staff concludes that the licensee's proposed attemative provides an coceptab;e level !
of quality and safety since it will test the subject penetrations for their intended function. ;
Therefore, the licensee's proposed altemative pressure test may be authorized, pursuant to 10 CFR 50.55a(s)(3)(1). ;
i
. Raquest for Rallef No. NR 31. Revision 1 (Unit 1 Only): Section XI, Table IWB 25001, !
Examination Category B A, item 81.11 requires 100% volumetric examination, as defined by ;
Figure IWB 25001, for all circumferential shell welds, i
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80Lt: The original mquest also included RPV shell welds on Unit 2; however, the <
l licensee revised the request in a letter dated May 19,1997, to Mude only one shell .
weM on Unit 1. A separate request willbe submitted upon completion of the RPV examinations for Unit 2. t i
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the volumetric examination, to the extent required by the Code, for RPV circumferential shell Wold 1RV-02 002, The Code requires 100% volumetric examination of the subject RPV weld.
H wvever, complete examination is restricted by the geometric configuration of the core barrel locating lugs which make the 100% volumetric examination impractical to perform for this weld, l To gain access for examination, the RPV would require design modifications. Imposition of this requirement would create an undue burden on the licensee.
The licenses has examined Wold 1RV-02 002 to the extent practical, obtaining approximately i 81% volumetric coverage, in addition, other RPV shell welds are being examined to the extent required by the Code Therefore, any existing pattoms of oogradation would have been L detected by the examinations that were completed and reasonable assurance of structural ,
integrity has been provided. Based on the irnpracticalsty of meeting the Code coverage m requirements for the subject welds, and the reasonable assurance provided by the significant i volumetric coverage that was completed on this and other RPV shell welds, the licensee's request for relief may be granted pursuant to 10 CFR 50.55a(g)(6)(i),
i N918: Request for Relief NR 31 included weM 1RV-02-002 that was evaluated above -
under 10 CFR 50.55e(g)(6)(I) for the ASME Sectron XI examination requirements. As ,
permitted by 10 CFR 50.55a(g)(6)(ii)(A)(4), the licensee substituted the augmented RPV j examination for the Section XI RPV examination. Sirke the licensee completed and
, credded the augmented examination for the Section XI examinations, and failed to
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achieve %ssen6ety 100W coverage on Ne wekt, the noensee proposed an aNemaHn k we requirement. me neman is evoluokd wow u noqunt kr new NMM.
RagumauerjellafiledHL-M (Unit 1 Only): 10 CFR 50.55a(g)(6)(ii)(A), Augmented Examite of Reactor Vessel. Regulatory Requirement: in accordance with 10 CFR 50.55a(Gh4. 'A). alllicensees must imp'ement once, as part of the inservice inspection interval in etiaat on September 8, igg 2, an augmented volumetric examination of the RPV welds spoolfled in hem 81.10 of Examination Category B A of the ig6g Ed6 tion of the ASME Code,Section XI. Examination Category B A, Hems 81.11 and 81.12 require volumetric exerrination of essentially 100% of the RPV circumferential and longitudinal shell welds, as defined by Figures lWL 25001 and 2, respectively. Essentally 100%, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90% of the examination volume of each wold.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensee has proposed an attemative to the coverage requirements of the augmented RPV examination required by the regulations. The essentially 100% coverage requirement could not be met for Wold 1RV-02-002. Essentially 100% of all other Examination Category B. A, item B1.10 wolds have been examined. To -
comply with the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examho essentially 100% of each of the item 81.10 shell wolds. In accordance with the regulations, essentially 100% is defined as greater than 90% of the examination volume of each weld. As an altemative to the greater than go% coverage requirement of the regulations, the licensee proposes that the examination coverage obtained be considered to provide an acceptable level of quality and safety for the RPV Wold 1RV 02-002.
At Braidwood, Unit 1, the augmented coverage requirements can not be met for the subject shell wold due to physical restrictions caused by co o barrel locating lugs that limit scan coverage. For Wold 1RV 02-002, the physical obstructions limited coverage to approximately 81% of the required volume. To achieve complete coverage for the subject welds, design modifications would be required to increase access from the inside surface (ID).
As a result of the augmented volumetric examination rule, licensees must make a reasonable ,
effort to maximize examination coverage of their reactor vessels. In cases where examination coverage from the 10 is inadequate, examination from the outside surface (OD) using manual J inspection techniques is a potential option. However, at Braidwood, Unit 1, the design of the reactor building prevents access for equipment and personnel from the OD. Therefore, the staff concluded that the licensee has made a reasonable effort to maximize examination coverage.
The licensee has examined greater than g0% of the Code required volume of all other RPV shell welds in combination with examining a large percentage (81%) of the sub}ect shall weld.
Further, the licensee has performed visual examinations of the vessel interior as required by the Code. The level of RPV shell weld examination coverage is significant and should have detected inservios degradation, if present. Therefore, the staff concluded that the licensee's .
proposed attemative provides an acceptable level of quality and safety, and the licensee's proposed altemative may be authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(s)(3)(i).
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3.0 CONCLUSION
S The staff has reviewed the licensee's submittal and condudes that as a resuH of Resistance Temperature Detection modifications, the population of welds in several ASME Class systems was changed. These changes, and corresponding inspection sampling, have been incorporated in the 181 Plan Summary tables of Revision 4. No deviations from ASME 181 tequirements were noted as a !ssuH of the changes in wold populations. Other non-technical changes includmg the incorporation of previously approved requests for relief, and those originating from typographical errors, are not evaluated in this Safety Evaluation.
The staff has concluded for Requests for Relief NR 25; NR 26; NR 27 Revision 1; NR 28, Revision 1; and NR 31 Revision 1, that the Code examination coverage requirements are impractical for the subject welds listed in the above requests for relief. Further, the staff concludes that the licensee's ahomate examinations that were performed provide reasonable assurance of the structuralintegrity of the subject components and as authorized by law, will not endanger life or property or the common defense and securtty and is otherwise in the public intotest giving due consideration to the burden upon the licensee that could resuH if the requirements were imposed. Therefore, for these requests, relief may be granted pursuant to 10 CFR 50.55a(g)(6)(l).
The staff also concludes that the licensee has maximized examination coverage for the reactor vessel welds and that service-induced degradation, if present, would have been detected. j Thus, for Request for Rolief NR 34, the licensee's proposed attemative provides an acceptable level of quality and safety. Therefore, the licensee's proposed attemative may be authorized pursuant to 10 CFR 50.55alg)(6)(ll)(A)(5) and 10 CFR 50.55a(a)(3)(i).
Additionally, the licensee's proposal to imp;ement Appendix J testing as an aHemative to Code required pressure tests, for Class 2 containment penetration piping connected to non-class systems, provided an acceptable level of quality and safety. Therefore, the staff concludes that Request for Relief NR 30 may be authorized, pursuant to 10 CFR 50.55a(3)(i).
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Attachment:
Technical Letter Report Principal Contributor: T.K. McLellan Dated: December 11, 1997
TECHNICAL LETTER REPQBI flRST 10 YEAR INTERVAL INSERVICE INSPECTION PROGRAM REVIS10N .4 COMMONWEALTH EDISON COMPANY RBAIDWOQD NUCLEAR POWER STATION. UNITS 1 AND 2 QQCKET NUMBER! 50-456 AND 50-457 1
1.0 INTRODUCTION
By letter dated October 8,1990, the licensse submitted Revision 4 to the Braidwood Nuclear Power Station, Units 1 and 2, first ten year interval inservice inspection program, included in Revision 4 were changes to reflect recent plant modifications, corrections to typogthphical errors, incorporation of existing approved relief requests, and several new requests. This Technical Letter Report provides an evaluation of the technical changes incorporated by Revision 4, with the exceptbn of Relief Request NR 29, that was previously reviewed to expedite its evaluation, in letters dated March 28,1997 and May 19,1997, the licensee submitted additionalinformation in support of the new requests for relief. The Idaho National Engincoring and Environmental Laboratory (INEEL) staff has evaluated the subject revision in the following section.
Pursuant to 10 CFR 50.55alg)(6)(ii)(A) the Commission revoked previous reliefs gra.nted to licensees for the extent of volumetric examinations of reactor vessel shell welds, as specified in Section XI, Division 1 of the ASME Boiler and Pressure Vessel Code. The Commission further required that all licensees augment their reactor vessel examination by implementing once, as part of the intervico inspection interval in effect on September 8, 1992, the item B1.10 requirements (exa 11ne essentially 100% of the volume of each shell weld) of the 1989 Edition of the ASME Code.
Under 10 CFR 50.55a(g)(6)(li)(A)(4), licensees may satisfy the augmented requirements by perform lng the ASME Section XI reactor vessel shell weld examinations scheduled for implementation during the inservice inspection intervai in effect on September 8,1992. As a result, the licensee is required to submit both an alternative to 10 CFR 50.55alg)l6)(ii)(A) and a request for relief per 10 CFR 50.55a(g)(5)(iii), or a ATTACHMENT
2 proposed alternative per 10 CFR 50.55a(3), for the same welds when the licensee obtains less than the required coverage (essentially 100%) during the examinations. The licensee has submitted Requests for Rolief NR 31, Revision 1 and NR 34 to satisfy each of these requiremmts.
2.0 EVALUATION The Code of record for the Braidwood Nuclear Power Station, Units 1 and 2, first 10 year intervalis the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI,1983 Edition with the Summer 1983 Addenda.
As a result of Resistance Temperature Detection modifications, the population of welds in several ASME Class systems were changed. These changes, and corresponding inspection sampling, have been incorporated in the ISI Plan Summary tables of Revision 4. No deviations from ASME ISI requirements ' vere noted as a result of the changes in weld popubtions. Other non technical changes including the incorporation of previously approved requests for relief, and those originating from typographical errors, are not evaluated in this Technical Letter Report. The information provided by the licensee in support of the new requests for relief has been evaluated and the bases for disposition are documented below.
2.1 flnlief_Reguest NR 25.Section XI. Table IWB-2500-1. Examination Cateaorv B H. Item B8.20. Examination of Pressurizer Intearal Attachment Welds Code Reauirement: Section XI, Table IWB 25001, Examination Category B-H, item B8.20, requires 100% surf ace examination of integral attachment welds to the pressurizer as defined by Figure IWB 250015.
Lingnage's Code Relief Reaues.t: Pursuant to 10 CFR 50.55alg)(5)(iii), the licensee requested relief from the Code required 100to surface examination of the following pressurizer integral attachment welds: (Unit il 1PZR 01 PS: -01,1PZR-01 PSL 02, 1PZR-01 PSL-03, and 1PZR 01 PSL-04; (Unit 2) 2PZR 01 PSL-01, 2PZR 01-PSL 02, 2PZR r
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3 4603,-.and 2PZR-01 1PSL-04.
't leansds Baals for Reaumati ng Relief (as stated):
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" Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is' being requested on the basis that compliance with the specified code requirement has been determined to be impractical.-
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- "Braidwood Units 1 and 2 Pressurizer seismic lugs are welded to the Pressurizer shell ,
- (reference Attachment 1)'. There are 4 seismic lugs per unit, located 90 degrees apart (rsference Attachtrent 5). In order to perform examinations on the seismic lug welds, the outs!de surface of the lower vessel shell to lug ares must be accessiple. The exam
!. wurface is not accessible since it is covered by the seismic lug restraint and lower Pressurizar shell insulation (reference Attachment 3 and 4). Also, the configuration of
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- tse Pressurizer coffin limits access to the seismic lugs. The impact of removing the seismic lug restraint, altering the Pressurizer coffin and removing the lower shell- .
insulation is presented below.-
2 "The selsmic restr.3ini(Reference Attachment 1 and ), which surrounds the lug, L prohibits access needed to perform a meaningful surface exam. There are 4 restraints
.ccated at about the 428' elevation, one for each lug, which were not desi.
J for removal. The' top of the concrete floor at this locatics,is 6t 428' 3" eleva - s. This i1 floor, which is 2'6" thick, interferes with accas's to 2 of the 4 lugs (Referenes Attachment 2,3, and 5). Also, the Pressurizer coffin itself severely limits access to the remaining 2 seismic restraints (Reference Attachment 5). All of the restraints, which are embedded in the concrete, would requira ma,ar modification to the existing b Pressurizer coffin to alinw for removal and acvess. This modification would require the redesign of the seismic restraint and Pressurizer coffin to allew fo; periodic removal end access to the seismic restraints. lmplementation of this redesign would require ognifinnt engineering resources, construction resources and a significant dose to plant personnel.
"Only the upper panels were desigr4t; with clips to provide for removal. insulation on the lower shell of the Pressurizer prohibits access needed to perform a meaningful surface examinatiori of the seismic lug weld areas. The removal of the insulation covering the lower Pressurizer shell to the seismic lug area will result in high radiation exposure to plant personnel. The insulation on the pressurizer consists of ;
pand which are fastaned together. The lower panels are festened together with screws.
To provide access from below would require scaffolding from the 401' elevation grating to the 428' elevation of the seismic restraint. Also, to remove the Pressurizer shell insulation would require removal of the screw fasteners. Access to these screws is limited by the floor and pressurizer coffin (reference Attachinent 3 ar.d 5). As stated above, the insulation could be removed from the upper portions of the lugs.- This can only be accomplished for 2 of the .4 se smic i lugs, because ra:: cess is prohibited by the
) Attachments provided by the licensee are not included in this evaluation report.
, m
4 Pressurizer coffin configuration (Reference Attachment 5). The current configuration of the seismic restraint also ony allows limited access for visual examinat5n. To provide suitable access for all 4 seismic lug restraints would require major modifications and significent resources.
"Even if the non removable insulation is removed (Reference Attachment 3,4, & 5), full surface examination of the seism!c lugs would not be achieved. The Pressurizer coffin, concrete floor and seismic restraint geometry would greatly limit access to all sides.
The resulting coverage would only be a small percentage of the weld volume. The limited data obtained from these examinations do not provide a compensatory increase in qu:lity and safety to justify the hazards af personnel radiation exposure to obtain the data.
When the removable insulation panels are remov3d, it is estimated that 5.74" of surface per accessible lug will be achievable. This accessible portion of surface can be visually inspected. it is expected that only a bast effort Liquid Penetrant (PT) exam
] can be performed on the accessible exposed surfaces. Access and clearance interferences willlimit how well the surfact of the examination volume can be prepped for the FT examination. Because the examination is being performed on slightly rusted carbon steel components, which will receive a best effort surf ace prep, that a white to pinkish background will be expected after developing. Even with a pinkish background, detection of relevant indications will still be possible. Also, bleed out from the lower edge of the non removable insulation will interfere with some of the accessible exam volume.
This volume of interference will depend upon the amount of bleed out and will mask any e . relevant indication.
4 Licensee's Procosed Alternative Examination (as stated):
"A VT-1 of the upper surfaces of the 2 accessible lugs when the removable insulation panels are removed, it is estimated that 5.74" per accessible lug (4" of the top and 87" of each side) will be achievable when the removable insulation parmis are removed.
This is approximately 28.7% of the total exam volume for one lug. Also, a best effort surface inspection (Liquid Penetrant) will be performed on those portions of the lug that are inspectable when the removable insulation panels ore removed. In conjunction with the above proposed alternative technique, the periodic VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWB-2SOO-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structural ir9egrity of the Pressuriz1r shell.
Evaluation: The Code requires 100% surface examination of the subject pressurizer integral attachment welds. However, seismic restraints surrounding the attachment welds provide minimal clearances for access to perform the rer,uired examinations. The restraints are embedded in concreto and would require major plant modifications to allow 1 i
periodic removal for access to thu pressurizer integral attachment welds. Insulation on l the lower shelllimits access to the weld ittachment area. This insulation was not
[
5 designed to f acilitate its removal, and significant exposure to plant personnel would result if the licensee is required to remove it to perform the required surface examinations. Further, it is expected that if this insulation were to be removed, full access to the attachment welds would be limited due to the seismic restraint configuration stated above. For these reasons, the INEEL staff has determined that the surface examinations are impractical to perform to the extent rsautred by the Code.
By removing psnels of insule' - n on the upper shell, the licensee obtained approximately 28.7% coverage of each
- agral attachment weld, in addition, the licenses proposed to perform a VT 1 visual examination of the remaining accessible areas. Based c'i the percent r.f Code required surface examination coverage obtained in combination with the .,
VT-1 visua! examination, and examinations performed on other Class 1 integral attachment welds, it is concluded that degradation, if present, would be detected, providing reasonable assurance of structural mtegrity. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55alg)(6)(i).
2.2 Belief Raouest NR-28.Section XI. Table IWB-2500-1. Examination Cateoorv C-C. Item C3.20. Examination of Class 2 Piuino Inteoral Attachment Welds Code Raouirement: Section XI, Table IWC-2500-1, Examination Category C C, item C3.20, requires 100% surface examination of integral attachment welds as defined by Figure IWB 2500-5.
Licensee's Code Relief Reauest: ln accordance with 10 CFR 50.55alg)(5)(iii). the licensee requested relief, to the extent required by the Code, from surface examinations of the following excmination areas: (Unit 1) 1FW-06 33,1FW 10-29,1FW-1129,1FW 29,1 RH-03 378,1 RH-04-73 A,1 RH-05 21 B,1 RH-07-25 A,1 RH-08-02A,1 R.1-03 -45 A,1 SI-04-0 1SI-12-23A, and 1S;-26-09A; (Unit 2) 2FW-06-01, 2FW-10-24, 2FW 11-25, 2FW-12 25, 2RH-04-03, 2RH-05 35, 2RH-OS 03, 2RH-09-28, 2SI-04-03, 2SI 12-19A, and 2SI-26-03.
1,icensee's Basis for Reauestino Relict (as stated):
{
F
'85 -
s
- Pursuant to _10'CFR 50.55a(g)(5)(iii), relief is being requested on the basis that .l
=
. compliance with the specificJ code requiroment has been determined to be impractical.
"Some penetrations at Staidwood were originally d6 signed where one of the integral ;
s'.cachment welds is inside the flusd head penetration assembly, thus making the welds_
inaccessible for inservice inspection. Access from outside of the closed'end of the .
flued head penetretion assembly for examiners is prc:hibited by the integral attachment.
Access irom the open end of the penetration is severely restrained due to geometry and clearance. Sse Attachments8 1,2,3,4, and 5 for penetration details. The. integral attachment weld is set back some distance inside the flued head penetration assembly and the clearance between the pipe and penetration sleeve is small. See Table 1 on -
Attachment 6,8 I
, "To satisfy the Code requirement to perform a surface examination of this weld, modification to the flued head penetration assembly and/or piping to allow access would be required. Braidwood would incur significant engineering and installation costs to perform such a modification without a compensating increase in the level of quality and
- safety to justify such modifications."
Licenn%'s Pronomad Alternative Examination (as stated):
"When a weld is scheduled for inspection, a surface examination of the accessible weld on the exposed outslos 'rface of the penetration will be performed. In conjunction with i the above proposed anarnative technique, the periodic VT 2 examinations in accordance with the requirements of ASME Section XI, Table IWC 2500-1, Examination Category C-H will provide reasonable assurance of continued structural integrity of the piping systems."
, Evaluation: The Code requires that the subject Class 2 integral attachment welds receive 100% surface examination. However, r portion of each welded attachment is
[ - located inside a flued head containment penetration, restricting accessibility such that performing the Code-ruquired surface examinations is impractical. To perform these
. surface examinations, modifications or replacement of the components with those of a E desiga providing for complete access would be necessary, imposition of this _ requirement would cause a considerable burden on the licensee.
- The licensee proposes to perforn, the Code-required surface examinations to the extent practical, i.e., on the accessible portion of each attachment weld. The accessible 12 Attachments provided by the licensee are not included in this evaluation report.
4
<__ _ = _ _ _ _ ____ _ . _ ._,_____.-.,1 , , _ ,,_
l c* .
7 -
areas comprise approximataly 50% o, the welds for each integrol attachment weld. Based on the number of integral attachments that will be examined, and the coverage obtained on each, it is concluded that generic degradation, if present, will be detected. As a result, reasonable assurance of structuralintegrity will be provided. Therefore, pursuant to 10 CFR 50.5Sa(g)(C)(i), it is recommended that the licens6e's request be granted.
2.3. Relief Raouest NR 27. Revision 1. (Un/t 1 OnM.Section XI. Table IWB-2500-1.
Examination Cateoorv B-D. Item C3.90. Examination of Reactor Pressure Vessel Nozzle-to-Shell Welds Note: The originalrequest also included nozzle welds on Unit 2, however, tM !!censee
,evised the request in a letter dated May 19, l'997, to include only the welds on Unit 1.
A separate request willbe submitted upon completion of the RPV examinations for Unit 2.
Code Reauirement: Sect!on XI, Table IWB-2500-1, Examination Category B A, item B3.90, requires 100% volumetric examination of resetor pressure vessel nozzle-to shell welds as clefined by Figure IWB 2500-7.
Licensee's Code Relief Raouest: Pursuant to 10 CFR 50.55a(g)(5)(iiin t' ilicensee requested relief from the Codtrequired 100% coverage of the follow..ig reactor pressure vessel nozzle-to-shell welds: 1RV-01-006,1RV 01-009,1RV-01-010, and 1RV-010-013 c
l 8
1.10P.Due's Basis foL R ecuestino Relief (as stated):
" Pursuant to 10 CFR 50.55alg)(6)(i), relief is requested on the basis that the code requirement to examine essentially 100% of the welds' volume is impractical due to geometric interference.
"All RPV welds are examined using remotely operated underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concerns due to the high radiation levels in these areas. The outlet (Hot Leg) nozzles are constructed with an integral extension on the 1.D. surface which mates with the internal core barrel. The extension provides a flow path for the reactor coolant from the core into the hot leg nozzles. The integral extensione partially obstruct the ci cumferential scan for reflectors transverse to the weld (Reference Attachment 1).8 The integral extencion, that confines the movement of the transducer package, along with the curvature of the RPV shell, combine to limit full Code volume coverage when scanning in the direction parallel to the weld (Reference Attachment 2). This configuration limits the examination aggregate volume coverage obtained for each weld and adjacent base metal to approximately 84% instead of the Code required essentially 100% coverage.
" Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the ID of the hot leg nozzles and/or the building structure surrounding the RPV at the nozzles' elevation. Braidwood Uriit 1 RPV was designed with an RPV shield wall (Reference 3 and 4).a This wallimpedes access to the OD of the RPV shell for insulation removal, surf ace preparation and ultrasonic inspection. Modifying the nozzle ID surf ace would incur extensive radiation exposure to station personnel and could be detrimental to the component. When designing, fabricating and installing these welds, strict ASME Section lli quality controls and procedures were ut ed that minimized the introduction of fabrication defects. Additionally, the per'odie VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolaat system monitoring requirements specified in the Technical Specifications will pmvide reasonable assurance vf the continued structuralintegrity of the reactor vessel. Comed has recently performed these volumetric examinations to the fullest exter.c practical, i.e.,84%, during the A1R06 refuel outage and no recordable indications (NRI) were detected. The NRI results of the examination provide further assurance that unacceptable inservice flaws have not developed in the subject welds."
3 Attachments provided by the licensee are not included in this evaluation report.
I 1
d .
I 9
Licensee's Prooosed Alternative Examination (as stated):
"The Reactor Vessel outlet (:k,t Leg) nozzle welds will be examined to the fullest extent practical using the available underwater volumetric inspection techniques."
Evaluation: The Code requires 100% volumetric ee nination of the subject RPV nozzle to-vessel welds. However, complete examination is restricted by their geometric configuration which makes the 100% volumetric examination impractical to perform for these areas. To gain access for 9xamination, the RPV nozzles would require design modifications. Imposition of this requirement would create an undue burden on the licensee.
The licensee has examined these welds to the extent practical, obtaining significant (approximately 84%) coverage of each nozzle to-vessel weid. In addition, nozzle-to-vessel welds on various other Class 1 components are being examined as required by the Code. Therefore, any existing patterns of degradation would have been detected by the examinations that were completed and reasonable assurance of the structuralintegrity has been provided.
Based on the impracticality of meeting the Code coverage requirements for the subject nozzle-to-vessel welds, and the reasonable assurance provided by the examinations that were completed on these and other Class 1 nozzles, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
- 2. 4 Relief Reauest NR-28. Revisbn 1. (Unit 1 On/v)Section XI. Table IWB-2SOO-1.
Examination Cateaorv B-D. Item C3.100. Examination of Reactor Pressure Vessel Nozzle Inner Radius Sectinns Note: The originalrequest also included nozzle inner radius sectio ts on Unit 2, however, the licensee revised the request in a letter d.1ted May 19,1997, to include only the areas on Unit 1. A separate request willbe submitted upon completion of the RPV examinations for Unit 2.
l
.a .
10 Code Reouirement: S3ction XI. Table IWB-25001, Examination Category B-D, item B3.100, requires 100% volumetric examination of reactor pressure vessel nozzle inside radius sections as defined by Figure IWB 2500-7.
Licensee's Code Relief Reauest: Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief for the Code-required 100% coverage of the following reactor pressure b vessel nozzle inner radius sections: 1RV 01-015,1RV-01-016,1RV-01-019, and 1RV 01-20.
Licensee's Basis for Reauestina Relief (as stated):
\
" Pursuant to 10 CFR 50.55a(g)(6)(i), relief is requested on the basis that the code requirement to examine essentially 100% of each inner radius section is impractical due to the geometry.
"All of the inner radius sections of ths RPV nozzles are examined using remotely operated i underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concerns due to the high radiation levels in these
( creas. Examination of the inner radius sections of the inlet (Cold Leg) nozzles is limited by the nozzle geometry (Reference Attachment 1 and 2).' The design of +he underwater volumetric inspection equipment was unable to scan the radius area where it transitions from the shell into the nozzle bore. This geometry limits the volumetric examination aggregate coverage obtained for the nozzle inner radius section to about 82% ;
instead of the esentially 100% Code required examination volume. '
" Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the nozzles' inner radius section and/or the building structure surrounding the RPV at the nozzles' elevation. Braidwood Unit 1 RPV was designed with an RPV shield wall (Reference 3 and 4). This wallimpedes access to the OD of the RPV shell for insulation removal, surface preparation and ultrasonic inspection. Modifying the "
nozzle ID safaces would incur extensive radiation exposure to station personnel.
Based on industry experience, no operating issues to date have been identified in PWR nozzle inner radius sections. If an inservice flaw ne.e to develop in this region, the flaw would be expected to initiate at the ID surface. A visual inspection (VT-1) of the
__ entire ID inner radius surface provides reasonable assurance that unallowable inservice flaws have not developed in the subject area. Additionally, the periodic VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements
' Attachments provided by the licensee are not included in this evaluation report.
___ __ _ _ _ _ _ _ _ ______________.___.__m. _ _ _ . - - - _ _ _
e * .
I I 11 specified in the Technical Specifications will provide reasonable assurance of the continued structuralintegrity of the reactor vessel. Comed has recently performed these volumetric examinations to the fullest extent practical, i.e.,82%, during the A1R06 refuel outage and no recordable indications (NRI) were detected. The NRI results of the examination provide further assuranca that unacceptable inservice flaws have not developed in the subject welds."
1)censee's Prt,cosed Afternative Examination (as stated):
"The Reactor Vesselinlet (Cold Leg) nozzle inner radius section was examined to the ]
fullest extent pNtical using the available underwater volumetric inspection techniques. In conjunction with the partial volumetric examination, a supplemental VT-1 of the nozzle inner radius area was conducted from the interior of the Reactor Vessel using underwater camera equipment."
Evaluation: The Code requires 100% volumetric examination of the subject RPV nozzle inside radius sections. However, complete examination is restricted by their geometric configuration which makes the 100% volumetric examination impractical to perform for these areas. To gain access for examination, the RPV nozzM would require design modifications, imposiCon of this requirement would create an undue burden on the licensee.
The licensee has examined these welds to the extent practical, obtaining significant (approximately 82%) coverage for each nozzio inside radius section, in addition, other Class 1 nozzle inside radi! are being examined as required by the Code. Therefore, any existing patterns of degradation would have been detected by the examinatioiis that were completed and reasonable aesurance of the structuralintegrity has been provided.
Based on the impracticality of meeting the Code coverage requirements for the subject nozzle inside radius sections, and the reasonable assurance provided by the examinations that were completed on these and other Class 1 nozzle inside radii, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
.: o ,.
12
= 2;'.i? Reauant for Relief No. NR 30. Revision 1. Examination Catsoorv C-H. Items C7.30. C7.40.
C7.70 and C7.BO. Praamura Testina of Containment Panatration Pinino Attached to liSD- ;
class Pining l.
Code Requirement Section XI, Tat.le IWC 25001, Examination Category C-H, items C7.30, C7.40, C7.70 and C?.80,' require a system leakage test at operating pressure for pressure 4
retaining piping and valves.
i Licanmaa Pronosed Alternative (as stated)
"Braidwood Station will perform 10 CFR 50, Appendix J leakage tests as an optional alternative to the Section XI required pressure test on the subject primary reactor containment penettstion pipints and associated valves. When implementing the Appendix J tests and invoking this relief request, peak design pressure and procedures for the detection and location of thrcugh wall flaws will be used. Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the propossd alternatis ; would provide an acceptable level of quality and safety."
, Licennan's Basis for Alternative (as stated)
"Specifically, Braldwood Station requests relief to perform 10 CFR 50 Appendix J leakage testing in lieu of the pressure test required by ASME Section XI, Table IWC 2500-1,
, Examination Category C H on the Code Class 2 Containment Penetration piping with attached nonclassed piping.
"The applicable components are piping lines and valves which are portions of non-safety related systems that penetrate the primary reactor containment. At esch containment g penetration, the process pipe is classified Code Class 2 and provided with isolation i valves that are.either locked shut during normal operation, capable of automatic closure, or capable of remote closure to support the containment safety function. The -
balance of piping outside the ie.:dation vanes is non-code and, therefore, outside the
- scope of the ASME Boiler & Pressure Vessel Code,Section XI. These components perform no _
other safety function. The only reason that the penetration piping is classified as Class 2 is because of its function as part of the containment pressure boundary.: The remaining portion of the system is nonnuclear related and the integrity of the system in relation to its primary function is not within the scope of Section XI. Since-containment integrity is the only safety related function, it is logical to test the Class 2 penetration portion of the system to the Appendix'J criteria.
- "The primary reactor containment integrity, including all containment penetrations, is
,'m._._-
13-
- periodically verified by performing leakage tests in accordance with a 10 CFR 50,
- Appendix J. The Appendix J test frequency provides assurances that the containment pressure boundary is being maintained at an acceptable level while monitoring for deterioration of sealsivalves and piping. If a pipe existed with a through wall flaw, the isolation valves located on both sides of the containment well would prevent any release outside containment. Multiple throughawall flaws or leakage oaths occurring simultaneously inside and outside uf containment between the isolation valves in a pipe segment is unlikely. Each of the Code Class 2 lines and their associated isolation ~
. valves are tested during an Appendix J leakage test at a pressure not less than 44.4 psig (Peak calculated containment pressure). The Appendix J leakage test is performed at
~ intervals in accordance with the requirements of the Braidwood Technical Specifications.
" Performance of these Appendix J leak tests will verify the integrity of the subject Code Class 2 lines and valves at the Containment Penetrations. _ The performance of ASME Section Xt, Examination Category C-H pressure tests on these same lines will proV.de ilttle, if any, additional verification of_ primary reactor containment integrity and impose a burden of duplicate testing.- Duplicate testing results in a significant increase in total amcunt of work force and radiological exposure without a compensating increase in the level of quality or safety.
)_
, "Per the preceding information, Braidwood Station requests relief to use Appendix J test as an optional alternative to the ASME Section XI requirements for pressure testing these Code Class 2 containment penetration components on the basis that the proposed alternate provisions provide an acceptable level of qua!!ty and safety. The proposed alternative is consistent with the requirements of Code Case N-522."
i Evaluation: The Code requires that a VT-2 visual examination be oerformed during system
-pressure testing for Class 2 pressure-ret'ining piping. _ As an a'ternative, the licensee proposes to implement the requirements of 10 CFR 50, Appendix J, for the subject containment penetration piping. This alternative is contained in ASME Cods Case N 522, Pressure Testing of Containment Penetration Piping.
The subject piping is classified as Class 2 because it penetrates primary reactor o containment and is considered an extension of the containment vessel. Since the piping on either side of these penetrations is non-class, the requirements of Appendix J are more appropriate than those of Examination Category C-H. ' Appendix J pressure tests
- verify the leak-tight integrity of the primary reactor containment arid of systems and components that penetrete containment by local leak rate and in*egrated leak rate tests.
In Appendix J pressure tests, containment ! solation valves and connecting pipe segments
_ m- .. , _ . _ . _ . . - _.__ __ _ _ _ _ . . _ , . _ _ _ - _ _ _ _ . .
4m ..
14:
l must withstand the peak calculated containment internal pressure related to the maximum design pressure.' in addition, Appendix J test frequencies provide assurance that the I
containment pressure boundary is being maintained at an acceptable level while' monitoring for deterioration of seals, valves, and piping. -
' The licensee has committed to perform ther Appendix J testing at no less than she peak calculated containment pressure and will use procedures to' detect and loc 6te through-wall flaws. The INEEL staff believes that an acceptable level of quality and safety will s be provided by the licensee's proposed alternative since it will tes': the subject
. : penetrations for their intended function. Therefore, it is recommended that the licensee's proposed alternative pressure test be authorized, pursuant to l
-10 CFR 50.55a(a)(3)(i). l
\
.l 22.6 Raouant for Relief Noc NR 31. Revision 1. (Unit f OnM Examination Cataaorv B.A. ! tem l
4 B1- 11. Reactor Prennu o Vennel Circumferential Shall Walda
.l Norn: The originalrequest also included RPV shell welds on Unit 2, however, the }
licensee revised the request in a letter dated May 19,1997, to include only one snell. '
weld on Unit 1. A separate request willbe submitted upon completion of the RPV examinations for Unit 2.
Code Raouirement-Section XI, Table IWB 2500-1, Exa nination Category B-A, item B1.11 requires 100% volumetric examination, as defined by Figure IWB-2500-1, for all-circumferential shell welds.
Licennea's Code Relief Reauest Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the volumetric examination, to the extent required by the Code, for RPV circumferentia! shell Weld 1RV-02-OO2.
<~..
15 Licensee's Basis for Reauestina Relief (as stated):
" Pursuant to 10 CFR 50.55a(g)(6)(i), relief is requested on the basis that the code requirement to examine essentially 100% of the weld volume is impractical due to geometric interf erence.
Examination of the subject RPV shell weld was conducted on Braidwood Unit 1 during A1R06 refuel outage (Spring 1997). During this exam at BralJwood Unit 1, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of the required volume. The examination of the lower shell course-to Dutchman weld,1RV 002, is restricted by six core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (See Attachment 1).s These lugs obstruct the automated UT examination tool from examining the Code required volume of the weld and base material under and below each lug in both the circtimferential and perpendicular scan directions (156' total for all 6 lugs, See Attachment 2,3, and 4).5 All w;ld and base material can be exaained between the lugs (204' total length between all 6 lugs). The 6 lug interferences limit the examination aggregate voNme coverage obtained for the weld and adjacent base material to approximately 81% of the Code required volume.
Licensee's Prooosed Afternative Examination (as stated)
"The ultrasonic examination of the Braidwood Unit 1 RPV shell weld,1RV-02-002, was performed to the maximum extent practical in conjunction with tha partial ultrasonic examination, a supplemental VT-1 of the RPV shell weld was conducted from the interior of :
the RPV using underwater camera equipment."
Evaluation: The Code requires 100% volumetric examination of the subject RPV weld.
However, complete armination is restricted by the geometric configuration of the core barrel locating lugs which makes the 100% volumetric examination impractical to perform for this weld. To gain access for examination, the RPV would require design modifications, imposition of this requirement would create an undue burden on the licensee.
The licensee has examMed Weld 1RV-02-002 to the extent practical, obtaining approximately 81% volumetric coverage, in addition, other RPV shell welds are being examined to the extent required by the Code. Therefore, any existing patterns 5
Attachments provided by the licensee are not included in this evaluaticn report.
y+ .
i 16-
. of degradation would have been detected by tho' examinations that were completed and -
- reasonable assurance of structuraiintegrity has been provided, l . Based on the impracticality of meeting the Code coverage requirements for the subject welds, and the reasonable assurance provided by the significant volumetric coverage that was completed on this and other RPV shell welds, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(l).
1 Netgu Request for Relief NR 31 included weld 1RV-02-002 that was evaluated above under 10 CFR 50.55alg)(6)(i) for the ASME Section XI examination requirements. As permitted
. examination for the Section XI RPV examination. 'Sincs the licensee completed and i - creditsd the augmented examination for the Section X! examinations, and failed to achieve ' essentially 1009," coverage en this weld, the licensee proposed an attemative to this requirement. This attemative is evaluated below as Request for Relief NR 34.
i.
l 2.7 Reauest for Relief No. NR 34. (Un!t f On/v) 10 CFR 50.55alolf 6)lillfA). Auamented Examination of Reactor Vennel !
D Reaulatorv Reauirement: In accordance with 10 CFR 50.55a(g)(6)(ii)(A), all licensees must implement once, as part of the inservice inspection interval in effect on September 8,1992, an augmented volumetric examination of the RPV welds specified in item B1.10 of Examination Category B-A of the 1989 Edition of the ASME Code,Section XI.
Examination Category B A, items B1.11 and B1.12 require volumetric examination of
- essentially 100% of the RPV circumferential and longitudinal shell welds, as defined by Figures IWB-2500-1 and -2, respectively. Essentially 100%, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is gren than 90% of the examination volume c,f each weld, .
Licensen's Proooned Alternative: Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licenses has proposed an alternative to the coverage requirements of the augmented RPV
' examination required by the regulations. The essentially 100% coverage requirement
., ,- . . . . . . ~_ - -. - -. - - - . . - . - - - - - --
17 could not be met for Wold 1RV 02 002.' Essentially 100% of all other Examination Category +
B-A, item P.1.10 welds has been examined. _The licensee stated:
" Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5); relief is requested from the requirement to
- examine more then 90%'of the examination volume of the RPV circumferential shell weld, 1RV-02-002, on the basis that the alternative to the examination requirements would provide an acceptable level of quality and safety."
- - Licensaa's Ramin for Alternative (as statadh
- 10 CFR 60.55a(g)(S)(ii)(A)(1) revokes all relief requests with respect to volumetric examination coverage for RPV shell welds specified in item B1.10. Portions of a previously granted First Interval relief request, NR-9, addressed limited exam coverage l on the Braidwood RPV shell welds.
L " Augmented examination of the subject RPV shell weld was conducted on Braidwood Unit 1 during A1R06 refuel outage (Spring 1997). During this exam at Braidwood Unit 1, physical
- obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of.the requireJ volumei The examination of the lower shell course-to-Dutchman weld,1RV - 002, is restricted by six core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (See Attachment-1).'
'Examinction of the Code required examination volume was completed to the maximum extent practical using alternate UT techniques qualified to the highest standard available.
RPV oxaminations were conducted from the I.D. of the vessel. Access to allow examination from the O.D. (Shell side) of these welds is restricted due to the structural concrete !
surrounding the vessel. The examination techniques employed heve been demonstrated and
- . qualified to the Performance Demonstration initletive (PDl) Program which meets the
- intent of the rules of Appendix Vlli of the ASME Code,Section XI,1992 Edition with 1993 Addenda._ Although the techniques have been qualified at PDI for single direction scanning, examinations were performed from both sides of the weld on the same surface, where feasible."
Evaluation: To comply with the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100% of each of the item B1.10 shell welds. In accordance with the regulations, essentially 100% is defined as greater than 90% of the examination volume of each weld. As an alternative to the greater than 90% coverage requirement of the regulations,'the licensee proposes that the examination coverage obtained be considered to provide an acceptable level of quality and safety for the RPV Weld 1RV-02-002. ,
' Attachments provided by the licensee are not included in this evaluation report, i
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c4 e 18 At Braidwood, Unit 1, the augmented coverage requirements cannot be met for the subject shell weld due to physical restrictions caused by core barrel locating lugs that limit scan coverage. For Weld 1RV-02-002, the physical obt.tructions limited coverage to approximately 81 % of the required volume. To achieve complete coverage for the subject welds, design modifications would be required to increase access from the inside surface (1D).
As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels. In cases-where examination coverage from the ID is inadequate, examination from the outside surface (OD) using manual inspection techniques is a potential option. However, at Braidwood, Unit 1, the design of the reactor building prevents access for equipment and personnel from the OD. Therefore, it is concluded that the licensee has made a reasonable effort to maximize examination coverage.
The licensee has examined greater than 90% of the Code-required volume of all other RPV shell welds in combination with examining a large percer tage (01 %) of the subject shell weld. Further, the licensee has performed visual examinations of the vesselinterior as required by the Code. The level of RFV shell weld examination coverage is significant and should have detected inservice degradation, if present. Therefore, the licensee's proposed alternative provides an acceptable level of quality and safety, and it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(g)(6)(lii(A)(5) and 10 CFR 50.55ata)(3)(i).
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19 3.0 ' CONCLUS1QN The INEEL staff has reviewed the licensee's submittal and concludes that the Code examination coverage requirements are impractical for the subject welds listed in Requests for Relief NR- ;
25, NR 26, NR 27, Rev.1, NR 28, Rev.1, and NR 31, Rev.1. Further, reasonable assurance of the structuralintegrity of the subject comps nents has been provided by the examinations that were performed. Therefore, for these requests, it is recommended that relief be granted pursuant to 10 CFR 50.55alg)(6)(i).
-The INEEL staff also concludes that the licensee has maximizea examination coversge for the reactor vessel welds and that service-induced degradation, if present, would have been detected. Thus, for Request for Relief NR 34, the licensee's proposed siternative provides an act:eptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(gs(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i).
Additionally, the licensee's proposal to implement Appendix J testin0 as an alternative to Code required pressure tests, for Class 2 containment penetration piping connected to non-class systems, provides an acceptable level of quality and safety. Therefore, it is recommended that Request for Relief NR-30 be authorized, pursuant to 10 CFR 50.55a(3)(i).
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