ML20217J531
| ML20217J531 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/08/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217J526 | List: |
| References | |
| NUDOCS 9710210169 | |
| Download: ML20217J531 (27) | |
Text
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p u tou g \ UNITED STATES g
2
} NUCLEAR REGULATORY COMMISSION
' WASHINGTON, D.C. 3066H001 M
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELARDING T E FIRST. SECOND. AND THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN RE0 VESTS FOR RELIEF EDB DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY BEAVER VALLEY NUCLEAR STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-334 AND 50-412
1.0 INTRODUCTION
The Technical Specifications (TSs) for Beaver Valley Power Station, Unit Hos.1 and 2 (BVPS-1 and BVPS-2), state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code and applicable addenda as required by the [qde of Federal Reaulations (CFR ,10 CFR 50.55a relief has beu granted by ').he Commission pu(g),
rsuant except to 10 where specific written CFR 50.55a(g)(6)(1).
In accordance with 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed l alternatives would provide an acceptable level of quality and safety or. ii l
compliance with the specified requirements would result in hardship or un(usu)al difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access ,
provisions and the pre-service examination requirements, 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 regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsecuent intervals comply with the requirements in the latest edition and addenca of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.
The applicable edition of Section XI of the ASME Code for the current BVPS-1 and BVPS-2, first (Unit 2) and second (Unit 1) 10-year inservice inspection (ISI) interval is the 1983 Edition. The Code of record for BVPS-1 and BVPS-2, second (Unit 2) and third (Unit 1) 10-year interval inservice inspection will be the 1989 Edition.
ENCLOSURE 9710210169 971008 PDR ADOCK 05000334 G PDR
____-_A
i Pursuant to 10 CFR 50.55a(g) with an examination requireme(5), if the licensee nt of Section XI of thedetermines ASME Code that conf'ormance-is not .
practical for its facility, information shall be submitted to the Commission in support of-that determination and a request'made for relief from the ASME Code requirement After evaluation of the determination, pursuant to 10CFR50.55a(gk6)(i),theCommissionmaygrantreliefandmayimpose
. alternative requ)irements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the-public interest,- giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
Pursuant to 10 CFR 50.55a reliefs granted to license (g)(6)(ii)(A) the Commission revoked all previous es 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 inservice inspection interval in effect on September 9,1992, the Item Bl.10 requirements (examine essentially 100% of the volume vi each shell weld) of the 1989 Edition of the ASME' Code. Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5),- licensees that determine that they are unable to satisfy completely the requirements for augmented inspection shall submit information to.the Commission to support that determination and propose an alternative that would provide an acceptable level of quality and safety.
Under 10 CFR 50.55a(g)(6)(ii)(A)(4), licensees may satisfy the augmented requirements by performing the ASME Section XI reactor vessel shell weld examinations scheduled for im)lementation during inservice inspection intervals in effect on Septem)er 8,-1992; ' As a result, the licensee is required to: submit both an alternative to 10 CFR 50.55a(g)(6)(ii)(A) and a request for relief per 10 CFR 50.55a(g)(5)(iii), or a proposed alternative per required coverag)e (essentially 100%) during the examinations.10 CFR 50.55a(3 By' letter dated March 25, 1997, Duquesne Light Company (licensee), submitted requests for relief from the ASME Section XI requirements for BVPS-1 and BVPS-2. The licensee provided additional information in its letter dated August 29, 1997.
2.0 EVALUATION The NRC staff, with technical assistance from its contr ator, the Idaho National Engineering and Environmental Laboratory (INEFL), .has evaluated the information provided by the licensee in support of its 10-year inservice inspection interval program plan requests for relief for BVPS-1 and BVPS-2.
~ Based on the results of the review, the NRC staff adopts the contractor's conclusions and recommendations-presented in the Technical Letter Report (TLR) attached.- -
1
l j- . j i
- Relief Request BVI-RV-Welds, Rev. 0: Section XI.- Table IWB-2500-1, '
L Examination Category B-A, Items B1.11 and Bl.30 (Unit 1 - Second Interval)
- . requires volumetric examination of essentially 100% of the weld length during each-10-year interval.
This request was submitted, pursuant to 10 CFR 50;55a(g)(5)(iii), since the i ASME Section XI requirements cited above, as well as those listed in
! 10 CFR !,0.55a(g)(6)(ii)(A),--Augmented examination of reactor vessel cannot'be
! met by the licensee. However, while a request under 10 CFR 50.55a(g)(5)(iii)
- is acce) table for ASME requirements deemed impractical, it is not allowed under tie augmented examination rule. Further, since all existing requests for relief were revoked by the augmented examination rule, it is not
, appropriate to address the current request prior to the augmented examination t rule being satisfied. Therefore, the evaluation of Request for Relief BVI-RV-WELDS is deferred until such time as the licensee has satisfied the
- regulations for the augmented reactor vessel examination.
Relief Request _ BVI-83.120-2, Rev. 0, Examination Category B-0, Item B3.120,
~
F Pressurizer Surge Nozzle Inner Radius Section-(Unit 1 - Second Interval) -
o requires 100% volumetric examination of all pressurizer nozzle inner radius l sections as defined by Figure IWB-2500-7(b). Pursuant to 10 CFR 50.55a examina(a)(3)(ii),
tion of pressurizer the licensee surgeproposed nozzle inner an alternative to the radius section 100% volumetrit.
RC-TK-1-RADIUS-6.
T l The licensee stated:
L "A visual examination-(VT-2) of this area will be performed in
- conjunction with the boric
- acid walkdown, performed every shutdown.
! Also, this area is included and documented in the Mode 3 walkdown of the
- RCS boundary,-performed during each startu).following refueling outages
, as' required by Item No..B15.20. Both of-tiese activities =are performed ~
i by qualified VT-2 examiners. These examinations are augmented by'the
, leakage detection methods noted above. If the-insulation is removed for
- l. maintenance or other purposes, the UT exam of the inner radius section
- . will-be performed."
The Code requires volumetric examination-of the pressurizer surge nozzle
, inside radius section. However, volumetric examination of the surge nozzle j inner radius is restricted by the insulation, heater penetrations and the
. nozzle-geometric configuration. If required to remove insulation, disconnect the heater. connections, and prepare the vessel surface for volumetric
-examination, high radiation exposures would be incurred by maintenance and
. ' examination personnel. In addition, there is a high potential that removing i the-insulation would result in damaging the fragile heater connections. If-performed, this would only result _in minimal volumetric coverage due to the geometry of the-nozzle and interference from heater penetrations, a
i The licensee has proposed to perform VT-2 visual examinations during each
- ' shutdown-and to perform the UT examination-when insulation is removed for.
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, maintenance or other purposes. Considering that the subject examination area has no history of service related failures, and that generic degradation will be detected by examinations of other Class 1 nozzle inner radii, the NRC staff concludes that the licensee's proposed alternative provides reasonable assurance of operational readiness of the subject systems. Therefore, performing the Code vt,1umetric examinations would result in a burden without a compensating increase in the level of quality and safety. Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee's proposed alternative is authorized and the ,
requested relief is granted.
Request for Relief BVI-89.31-2, Rev 0: Examination Category B-J, Item B9.31, Reactor Coolant System Branch Connection Welds (Unit 1 - Second Interval) requires 100% volumetric and surface examination of piping branch connection welds as defined in Figure IWB-2500-9. Pursuant to 10 CFR 50.55a(g)(5)(iii),
the licensee requested relief from 100% volumetric examination of piping branch connection Welds DLW-LOOPI-7-S-04 (6"), DLW-LOOPl-7-S-05 (4"), DLW-LOOP 2-1-S-03 (6"), DLW-LOOP 2-7-S-05 (12"), DLW-LOOP 3-1-S-02 (14"), DLW-LOOP 3-1-S-04 (6"), DLW-LOOP 3-7-S-06 (4").
The Code requires volumetric and surface examinations of the subject piping branch connection welds. However, the configuration and metallurgical properties preclude achieving 100% coverage of the subject pipe branch connection. The branch connections are " set-on" designs that are not condocive for examination from the main piping run. As a result, the examinations are limited to one side only; from the branch piping side.
The pipe branch connection geometry makes the Code-required 100% volumetric examination impractical. To obtain complete volumetric coverage, modification or replacement of the branch connections with connections of a design providing.for complete coverage would be required. Imposition of this f requirement would cause a considerable burden on the licensee.
The licensee proposed to examine the subject welds to the extent possible.
Based on a review of the coverage plots provided, it appears that a significant portion (>50%) of the welds is examined with a one-directional beam from the branch piping side. Based on the significant amount of the volume examined from one side, it is concluded that degradation, if present, will be detected. As a result, reasonable assurance of continued structural integrity will be provided for the subject components. Therefore, relief is granted and the alternative imposed, pursuant to 10 CFR 50.55a(g)(6)(1).
Request for Relief BV2-Pl.11-1, Rev. 0: Examination Category B-A, Item Bl.11, Reactor Pressure Vessel Head Circumferential Weld (Unit 2 - First Interval) requires volumetric examination of essentially 100% of the weld length during each 10-year interval.
This request was submitted, pursuant to 10 CFR 50.55a(g)(5)(iii), since the ASME Section XI requirements cited above, as well as those listed in 1
i
C 10 CFR 50.55a(g)(6)(ii)(A), Augmented examinatfon of reactor vessel cannot be met by the licensee. However,whilearequestunder10CFR50.55a(g)(5)(iii) is at.ceptable for ASME requirements deemed impractical, it is not allowed under the augmented examination rule. Further, since all existing requests for relief were revoked by the augmented examination rule, it is not appropriate to address the current request prior to the augmented examination rule being satisfied. Therefore, evaluation of Request for Relief BV2-Bl.11-1 is deferred until such time as the licensee has satisfied the regulations for the augmented reactor vessel examination.
Relief Request BV2-B3.110-2, Rev. 0: Examination Category B-0, Item B3.110, Pressurizer Nozzle-to-Shell Welds volumetric examination of all press (Unit 2 - First Interval) requires 100%
urizer nozzle-to-shell welds as .iefined in Figures IWB-2500-7(a) through (d), as applicable. Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination of pressurizer nozzle-to-shell Welds 2RCS* PRE 21-N-10 (6" relief nozzle), 2RCS* PRE 21-N-ll (6" relief nozzle), 2RCS* PRE 21-12 (6" relief nozzle),
2RCS* PRE 21-N-13 (6" relief nozzle), 2RCS* PRE 21-N-14 (4" spray nozzle).
The Code requires volumetric examination of the pressurizer nozzle-to-vessel welds. However, volumetric examination of the subject relief and spray nozzle-to-vessel welds is limited due to the radius of curvature in the transition area between the nozzle and the vessel shell. Therefore, the i nozzles' geometric design configuration makes the volumetric examination impractical to perform to the extent required by the Code. To meet the Code requirements, the nozzles would have to be modified to facilitate access for ultrasonic search units. Imposition of this requirement would create a considerable burden on the licensee.
The licensee examired the subject welds to the extent possible obtaining approximately 57% coverage of the required area for each weld. As a result, the staff concludes that significant patterns of degradation, if existing, would be detected, providing reasonable assurance of the structural integrity.
Therefore, the requested relief is granted and the alternative imposed pursuant to 10 CFR 50.55a(g)(6)(1).
Request for Relief BV2-D2.20-1, Rev 0: (Unit 2 - First Interval) ASME Section XI, Examination Category D-A, D-B, and D-C, Items DI.20 through D1.60, D2.20 through D2.60, and D3.20 through D3.60 require VT-3 visual examination as defined by Figure IWD-2500-1. In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Code requirements associated with the selection and examination of integrally-welded attachments. The licensee stated:
"The alternate examination is to implement the sampling criteria of Code Case N-509 to BV-2, Class 3 welded attachments for the current ten-year interval, ending November 16, 1997. This requires examination of 10% of the welded attachments. Ninety-seven (97) percent of the total number I
9 i
of Class 3 welded attachments have been examined during this pr'esent interval. Approval of this alternative will preclude the hardship associated with efforts required to gain access to perform visual examinations on the remaining 3% of welded attachments (two supports).
A best effort examination of these two supports was performed with the insulation installed. No deformation or any other anomaly was reported.'
In lieu of Code requirements for selection and examination of integral attachment welds, the licensee proposes to apply alternativas contained in Code Case N-509, Alternative Rules for the Selection and h aninntion of Class 1, 2, and 3 Integrolly Welded Attachments for C1 ass 3 integra1 attachment welds only. Code "ase N-509 has reduced the percent of areas to be examined during each interval from 100% to 10%, for all components. This Code Case has previously been found acceptable by the NRC staff, provided that licensees examine a minimum of 10% of all Code Class 1, 2, and 3 piping, pumps, and valve integrally-welded attachments. The licensee has met this provision for Code Class 3 components.
Considering that the majority of Code examination requirements are based on component sampling to assure that service-related degradation is not occurring, it is icgical to extend the sampling 3rocess to welded integral attachments. The licensee has examined 97% of tie Class 3 integral attachment welds during *nis interval. No relevant service-induced degradation was observed.
Based on the fact that the licensee has examined a large percent of Class 3 integral attachment welds, the NRC staff believes that degradation, had it occurred, would have been detected. Since the conditions for use of the alternatives contained in Code Case N-509, for Code Class 3 components, have been satisfied and are sufficient to provide reasonable assurance of weld integrity, an acceptable level of quality and safety has been provided.
Therefore, the licensee's proposed alternative for class 3 integrally-welded attachments is authorized pursuant to 10 CFR 50.55a(a)(3)(1) and the requested relief is granted.
Use of alternatives contained in Code Case N-509 are authorized for the first interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this code case for Class 3 items, the licensee should follow all provisions in Code Case N-509, with limitations issued in Regulatory Guide 1.147, if any.
Request for Relief BV3-!WA-1, Rev. 0: IWA-5242(a) Insulation Removal For VT-2 Visual Examination Of Bolting in Class 1 and 2. Borsted Systems (Unit 1 -
Third. interval, and Unit 2 - Second Interval) requires that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure-retaining bolted connections for VT-2 visual examination.
Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an alternative to 1
1 o
the ASME Section XI requirements for removing insulation from Class 1 and 2, pressure-retaining bolted connections for VT-2 visual examination during VT-2 visual examinations. The licensee stated:
"A visual, VT-2 examination will be performed during the system pressure tests of IWB-5000 and IWC-5000 with the insulation installed. In addition, insulation will be removed from all Class 1 bolted connections each refueling outage and from all Class 2 bolting connections on a frequenc outage, out y ofnot approximately 3 a.id to exceed once per 1/3 years period), (every when the other Code refueling requirement, regarding insulation removal (IWA-524E), is not met. A visual VT-2 examination shall be )erformed, with the insulation removed, for evidence of leakage ()oric acid residue) independent of the system pressure test."
"Duquesne Light requests NRC approval of the proposed alternative requirements for VT-2 visual examination of Class 1 and 2 insulated pressure retaining bolted connections in lieu of those of IWA-5242(a).
NRC approval is requested in a time frame that will support implementation of the alternative requirements for the Unit I twelfth refueling outage currently scheduled to begin in September of 1997.
Unit I twelfth refueling outage will be the first outage of the third ten year interval at B2 aver Valley Power Station Unit No. 1. The second ten year interval at Beaver Valley Power Station Unit No. 2 will begin November 17, 1997."
The C;Je requires the removal of all insulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. As an alternative, the licensee has proposed to perform the VT-2 visual examination with a minimum 4-hour hold time with the insulation in place. In addition, the insulation will be removed for direct visual examination each refueling outage for Class 1, and each period for Class 2 components.
The licensee's proposed alternative is similar to that found in Code Case N-533, Alternative Requirements for VT-2 Visual hamination of Class 1 insulated Pressure-Retaining Bolted Connections,Section XI, Division 1, except the proposed alternative was extended to address Code Class 2 bolted connections.
Code Case N-533 is currently under review by the NRC staff and has not yet been approved for general use by incorporation into Regulatory Guide 1.147, inservice inspection Code Case Acceptability; however, it has been found acceptable on a plant-specific basis with the commitment of a 4-hour hold time on insulated systems.
For Class I systems, the licensee's proposed alternative provides a thorough approach for ensuring the leak-tight integrity of systems borated for the purpose of controlling reactivity. First, the 4-hour hold time allows time for any leakage to penetrate the iraulation, thus provides a means of detecting any significant leakage with the insulation in place. Second, by
O removing the insulation each refueling outage, the licensee will be able to detect minor leakage that could occur by observing the presence of boric acid crystals or residue. This two-phased approach will provide an acceptable level of quality and safety for bolted connections in borated systems.
Therefore,pursuantto10CfR50.55a(a)(3)(1)thelicensee'sproposed alternative is authorized for use on C1:ss 1 systems and the requested relief is granted.
For Class 2 systems, the frequencies proposed for insulation removal have not been found acceptable by the NRC staff. Therefore, the licensee's proposed alternative is not authorized for Class 2 systems and the requested relief is denied.
Request for Relief BV3-lWA-2, Rev. 0: IWA-5250(a)(2 , Corrective Action Resulting from Leakage at Bolted Connections Unit 2 - Second Interval) requires that the so(Unit 1)-
urce of leakages detected during Third Interva a system pressure test be located and evaluated by the Owner for corrective action. When the leakage is at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100. Pursuant to 10 CFR 50.55a a alternative to the ASME Section XI require ts (me)n(3)(ii),
for removalthe oflicensee proposed an bolting at leaking connections for VT-3 visual examination. The licensee stated:
" Leakage on Code Class 1, 2, and 3 bolted connections shall be evaluated to determine the susceptibility of the bolting to corrosion and potential future failure. This review will considct as a minimum, location of leakage, history of leakage, bolted connection materials, visual evidence of corrosion with the connecting assembled, corrosiveness of the process fluid, history and studies of similar bolted material in a similar environment, and other com)onents in +he vicinity that may be degraded due to the leakage, if t11s review assures the integrity of the connection, no bolting will be removed for examination. If this review cannot assure the integrity of the connection, the bolt closest to the leakage source shall be removed. VT-1 examined and evaluated in accordance with IWA-3100(a . When the removed bolt has rejectable degradation, the remaining) bolts will be removed, VT-3 examined and evaluated in accordance with IWA-3100(a). If no degradation is observed on the removed bolt, no further action will be required."
In accordance with the 1989 Edition of the Code when leakage occurs at bolted connections, all bolting is required to be remov,ed for VT-3 visual examination. In lieu of the Code required removal of bolting to perform a VT-3 visual examination, the licensee has proposed to perform an evaluation of the bolted connection to determine the susceptibility of the bolting to corrosion and the potential for failure.
This alternative allows the licensee to utilize a systematic approach and sound engineering judgement; provided, as a minimum, all evaluation factors l
e
. -g-lis d in the licensee's proposed alternative are considered. Furthe'rmore, if the initial evaluation indicates the need for a more in-depth evaluation, the bolt closest to the source of leakage will be removed, VT-1 examined, and evaluatedinaccordancewithIW4-3100(a).
Removal of all bolting as part of the corrective action when leakage is observed at a bolted connection can result in a significant burden on the licensee. Based on the licensee's proposed alternative, to perform an evaluation of the bolting to determine the potential for degradation, the NRC staff concludes that the operational readiness of bolted connections will be maintained. Therefore,pursuantto10CFR50.55a(a) proposed alternative is authorized and the requested (3)(ii) the licensee's relief is granted.
3.0 CONCLUSION
S The NRC staff has reviewed the information provided by the licensee for requests for relief at BVPS-1 and BVPS-2. Based on the evaluation of Relief Recuests BVl-B3.120-2 and BV3-lWA-2, the NRC staff concluded that imposing the Coce requirements on the licensee results in a burden without a compensating increase in quality and safety. Therefore, the licensee's pr posed alternatives are authorized pursuant to 10 CFR 50.55a a In addition, pursuant to 10 CFR 50.55a(a)(3; *i1), the licensee's pr(op)o(3)(i ).
sed 1ternative contained in Relief Request BV3-lWA-1 is authorized for Class 1 bolted connections only and for Class 2 systems, the frequencies proposed for .
insulation removal have not been found acceptable by the NRC staff.
Therefore, the licensee's proposed alternative for the Class 2 systems is denied.
For Relief Request BV2-D2.20-1, the NRC staff concluded that the licensee's proposed alternative provides an acceptable level of quality and safety.
, the licensee's proposed Therefore,pursuantto10CFR50.55a(a)(3)(1)lly-weldedattachments.
alternative is authorized for Class 3 integra Use of alternatives contained in Code Case N-509 are authorized for the first interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this code case for Class 3 items, the licensee should follow all provisions in Code Case N-509, with limitations issued in Regulatory Guide 1.147, if any.
For Relief Requests BV1-B9.31-2 and BV2-B3.110-2, the NRC staff concluded that the ASME Code examination requirements are in. practical. Therefore, pursuant to 10 CFR 50.55a(g)(6)(1) the licensee's request for relief is granted and the alternative imposed. The reliefs granted are authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Requests for Relief BVl-RV-WELDS and BV2-Bl.ll-1 were submitted by the licensee to address limited A5ME Code examination coverage for reactor
f . -
10 - ,
I l pressure vessel welds. However, the licensee has yet to fulfill the 'au examinatfon requirements found in 10 CFR 50.55a(g)(6)(ii)(A) for welds gmented that were not essentially 100% examined. Therefore, the NRC staff concluded that evaluation of these requests are to be deferred until such time as the licensee has satisfied the regulations by submitting an alternative to the augmented examination for reactor pressure vessel welds where essentially 100%
volumetric coverage was not obtained.
I
Attachment:
Technical Letter Report Principal Contributor: T. McLellan j Date: October 8,1997 i
i l
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l TECHNICAL LETTER REPOHI ON THE FIRST. SECOND AND THIRD 10 yEARJNSERVICE INSPECHQNJNIERY_ ALE DUQUESNE LIGHT COMPANY .
REAVER VALLEY POWER STATION. UNITS 1 AND 2 DOCKET NUMBER: 50 334 AND 50 412
1.0 INTRODUCTION
! By letter dated Merch 25,1997, the licensee submitted requests for relief for the Beaver i
Valley Power Station, Units 1 and 2, First, Second and Third Ten Year Inservice inspection i
Programs, in a letter dated August 29,1997, the licensee submitted additional information in support of the requests for relief. The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the subject requests in the following section.
2.0 EVALUATION The Code of record for the current intervals at Beaver Valley Power Station, Units 1 and 2, is the 1983 Edition through the Summer 1983 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The Unit 1 interval ends September 20,1997, and the Unit 2 interval ends November 16,1997. The information provided by the licensee in support of the requests has been evaluated ard the bases for disposition are documented below. The licensee submitted requests for both the current and future inspection intervals for each unit. The evaluations designate the unit and interval to which each request applies.
2.1 Relief Raouest BV1 RV W9lds. Rev. O.Section XI. Table IWB 25001. Examination Cateoorv B A. Items B1.11 and B1.30 (Unit 1 - Second Intervall NOTE: This request was submitted, pursuant to 10 CFR 50.55a(g)(5)(iii), to satisfy the ASME Section XI requirements cited above, as well as those listed in Attachment
_ _w
4 2 10 CFR 50.55a(g)(6)(ll)(A), Augmented exam / nation of reactor vessel. However, while a request under 10 CFR 50.55a(g)(5)(iii) is acceptable for ASME requirements deemed impractical,it is not allowed under the augmented examination rule.
Further, since all existing requests fo. relief were revoked by the augmented rule,it is not appropriate to address the current request prior to the augmented rule being satisfied. Therefore, it is recommended that evaluation of Request for Relief l BV1 RV Welds be deferred until such time as the licensee has satisfied the regulations for the augmented reactor vessel examination.
2.2 Relief Raouest BV1 B3.120 2 Rev. O. Examination Category B D. Item B3.120.
Pressurizer Surce Norrie Inner Radius Section (Unit 1 - Second Interval)
Code Reauirement: Examination Category B D, item B3.120 requires 100%
volumetric examination of all pressurizer nozzle inrier radius sections as defined by Figure IWB 2500-7(b).
Ucensee's Pronosed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an alternative to the 100% volumetric examination of pressurizer surge nozzle inner radius section RC TK 1 RADIUS 6.
The licensee stated:
"A visual examination (VT 2) of this area will be performed in conjunction with the boric acid walkdown, performed every shutdown. Also, this area is included and documented in the Mode 3 walkdown of the RCS boundary, performed during each startup following refueling outages as required by item No. 815.20. Both of these activities are performed by qualified VT-2 examiners. These examinations are augmented by the leakage detection methods noted above. If the insulation is removed for maintenance or other purposes, the UT exam of the inner radius section will be performed."
Ucensee's Basis for the Prooosed Alternative (as stated):
"In accordance with 10 CFR 50.55ata)(3)(ii), relief a requested on the basis that compliance with the Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3
'The Beaver Valley Unit 1 pressurizer lower head and the pressurizer surge nozzle is a carbon steel casting to SA 216 GR. WCC. To perform the required UT ,
examination on the surge nozzle inner radius section, the outside surface of the lower head must be accessible. This surface is made accessible by removing the insulation surrounding the surge nozzle. The design of this insulation requires disconnection of the 78 heater cables from the immersion heaters prior to removing the insulation (see figure l'). Each cable consists of two wires, each mechanically connected to the heater (see figure 2). Great care must be taken during the disconnection to ensure the ceramic terminal blocks to which the heater pins are brazed, are not damaged. Otherwise, an unbrazing/ brazing evolution would be required to replace the blocks. The dose estimate for this exam assumes that no ceramic terminal blocks would require replacement. Another concern involved in this examination is the presence of asbestos in the cable jackets, though, radioactive l
contamination is not typleally a concern in this area, respirators would be required due to the potential asbestos exposure. Additional cover alls would be required over i
' the anti contamination clothing, causing a heat stress concern. Special monitoring and material control would also be necessary due to the presence of asbestos.
"The dose estimate included below is based on a survey obtained on 4/19/96 during 1R11. The hours estimated to perform the activities involved in disconnecting the cables are based on similar efforts performed during 1R08 when two cables were disconnected due to potential short circuits. Comprehensive dry run exercises were performed on a mock uo in preparation for the 1R08 efforts. Therefore, the estimated times used in the dose estimate are believed to be quite accurate.
"If the insulation was removed from this area, the complete code required exam could be performed. Special search units were designed to perform this specific examination. The other five pressurizer inner radius sections have been successfully examined. No recordable indications were noted on these examinations. Also, the adjacent safe-end to nozzle weld has been completely examined (1R08) with satisfactory results.
"An informal survey of Westinghouse plants found a mix of plants having approved relief requests and others that perform this examination. Those performing this examination found no unacceptable indications with one exception. One utility found very small cladding cracks. These cracks were attributed to a one time event that was caused by thermal shock when cold water was allowed to enter the empty pressurizer with the heaters on. BV Units 1 and 2 have 30 years of combined operation experience, with no problems in this area. The BV 2 surge nozzle inner radius section and the surge nozzle to vessel weld were UT examined during the current interval and found no recordable indication. The BV-1 surge nozzle inner radius section was not examined in the first interval. Re'ief was granted based on the configuration of the nozzle and the lack of an adequate UT technique.
Figures provided by the licensee are not included with this evaluation, n
1 4
"The radiation exposure for cable disconnection /reconnection, insulation removal /relnstallation, surface preparation, and the UT examination is 54,600 mR as noted in the chart below. The dose estimate is based on a survey conducted with the insulation installed. Once the insulation is removed the rates shown in the survey would increase. Shielding at this location is not practical since the source of the radiation is component surface to be examined.
"A remote visual examination from the inside of the pressurizer was considered as an altamative to the UT examination. A screen located at the surge line nozzle and r baffle plates in the lower section of the pressurizer would restrict access to the area of interest. The distance from ths manway to the surge norrie area is approximately 40 feet, making positioning adjustments of the remote camera difficult. Because of these limitations, a remote visual examination is not considered a viable alternative.
- Several methods are available to detect leakage from this area if a through wall leak occurred. Listed below are sorr.a examples:
- s. The control room operators perform Operation Surveillance Test (OST) 1.6.2
" Reactor Coolant System Water inventory Balance" overy three days when the plant is operating at steady conditions. Leakage through the subject welds would be discovered by the conduct of this OST.
- b. Containment altborne radiatien monitors continuously sample the containment atmosphere and alarm in the control room. The sensitivity of this detection method is dependent on the size of the leak, reactor coolant activity, and containment background activity.
c.- Leakage from this area would cause an increase in containment pressure, temperature and humidity which are indicated in the main control room. The containment pressure alarms in the main control room,
- d. Substantialleakage from this area would collect in the containment sump. The sump level is indicated and alarmed in the main control room.
"There are no credible failurt rrschanisms other than fatigue for this area. Corrosion degradation protection is provided by the combination of the austenitic stainless steel cladding of the surge nozzle inner radius and by the chemistry controls on the reactor coolant system. Strict chemistry standards are maintained to ensure a non-corrosive environment. Oxygen, chloride, fluoride and other contaminant concentrations are malntained below the thresholds known to be conducive to stress corrosion cracking. Since the surge nozzle is cast, the typical failure mechanisms associated with weld material do not apply to this examination. Erosion and Erosion / Corrosion degradation is not credible at this location. The austenitic stelnless steel cladding resists this mechanism. There is relatively low fluid velocity in the surge nozzle and reactor coolant chemistry minimizes the amount of particles in the fluid that could potentially cause erosion. Creep and stress relaxation are not
O 5
concerns for the surge nozzle inner radius area since the design temperature of 680'F is below the temperature where creep becomes a concern.
" Fatigue degradation is a concern in this area due to the potential thermal cycling caused by the insurge and outsurge of the reactor coolant flow. Since the surge nozzle is cast with the bottom head, there is no nozzle to vessel weld. The inner radius is believed to be less susceptible to fatigue problems than a nozzle to vessel weld. Initiation of fatigue cracking may have equal potential at the inner radius as compared to a nozzle to vessel weld, but the chances of having a pre existing flaw are less likely in the inner radius casting than at a nozzle to vessel weld due to the manufacturing process. This hypothetical flaw would then had to of been overlooked by the shop NDE (which includes surface, UT and RT examination).
Inservice f atigue crack growth for such a flaw would be very small since the pressurizer is hot during the insurges and outsurges resulting in relatively high fracture toughness of the material. A thermal sloove, installed in the surge nozzle provides a measure of protection from the affects of fluid temperature changes.
Examinations are performed on the nozzle to safe end weld, which is within 18" of the inner radius. The nozzle to safe end weld has been satisfactorily examined without limitation during the first two intervals.
"The radiation exposure associated with the preparation activities for this examination is considered a significant hardship. If the preparation activities were performed, the subsequent examination would not significantly increase the level of quality and safety due to the low probability of the presence of a flaw in this area based on the information presented above. It is therefore concluded that the intent of 10 CFR 50.55ata)(a)(3)(ii) is met."
Eyf luation: The Code requires volumetric examination of the pressurizer surge t
nozzle inside radius section. However, volumetric examination of the surge nozzle inner radius is restricted by the insulation, heater penetrations and the nozzle geometric configuration if required to remove insuletion, disconnect the heater connections, and prepare the vessel surface for volumetric examination, high radiation exposures would be incurred by maintenance and examination personnel, in addition, there is a high potential that removing the insulation would result in damaging the fragile heater connections. If perfctmod, this would only result in minimal volumetric coverage due to the geometry of the nozzle and interference from heater penetrations. Therefore, performing the volumetric examination would result in a burden without a compensating increase in the level of quality and safety.
6 The licensee has proposed to perform VT 2 visual examinations during each shutdown and to perform the UT examination when insulation is removed for maintenance or other purposes. Considering that the subject examination area has no history of service related failures, and that generic degradation will be detected by examinations of other Class 1 nozzle inner radii, the INEEL believes that the licensee's proposal provides reasonable assurance of operational readiness.
Therefore, pursuant to 10 CFR 50.55ata)(3)(ii), it is recommended that the proposed alternative be authorized.
2.3 Re, quest for Relief BV1 B9.312. Rev. O, Examination Cateoorv B J. Item B9.31.
Reactor Coolant System Branch Connection Welds (Unit 1 - Second interval)
Code Reauirement: Examination Category B J, Item B9.31 requires 100% volumetric and surface examination of piping branch connection welds as defined in Figure IWB-l 2500 9.
1 Licensee's Code Relief Reouest: Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination of piping branch connection Welds DLW LOOP 17 S 04 (6"), DLW LOOP 1-7 G 05 (4"), DLW LOOP 21 S-03 (6"),
DLW LOOP 2 7 S 05 (12"), DLW LOOP 3-1 S-02 (14"), DLW LOOP 31 S 04 (6"),
DLW LOOP 3 7 S-06 (4").
Licensee's Basis for Reouestino Relief (as stated):
"In accordance with 10 CFR 50.55alg)(6)(i), relief is requested on the basis that compliance with the Code requirement is impractical. Ultrasonic examinations were performed on the RCS branch connection welds listed above. Limitations were identified due to the configuration of the weld joints. These limitations prevented UT examination of the code defined required examination volume for flaws oriented in the axial direction (with respect to the branch pipe). The limitations also prevented examinatien of the required volume from both sides of the weld. However, the UT techniques employed were able to examine the required volume (90% or greater) from the branch connection side of the weld in one direction.
"The examination techniques employed were developed by the DLC NDE group and required fabrication of three unique calibration standards as well as specialized focused refracted longitudinal transducers to examine the welds. A 55* refracted
- )
. 7 longitudinal transducer was used for the 4" branch connections. A 60' refracted longitudinal transducer was used for larger branch connections. These tra.nsducers are designed to focus on the intier one third of each of the welds. The attached sketches identify the general weld configuration and the extent of the UT examination coverage.
"An informal survey of Westinghouse plants found that there have been no problems with the integrity of this type of weld configuration."
Licensee's Pronosed Alternative (as stated):
i "The altamative to the examination requirements is to parform the complete surface
' examination and the ultrasonic examination, using enhanced techniques employing the specialized transducers to the maximum extent possible. These enhanced
, techniques represent the most current technology available for this examination and esult in the maximum coverage allowed by the joint design and material used."
hiluation: The Code requires volumetric and surfaco examinations of the subject pipng branch connection welds. However, the configuration and metallurgical propttties preclude achieving 100% coverage of the subject pipe branch connection.
The bmnch connections are " set on" designs that are not conducive for examination trom the main piping run. As a result, the examinations are limited to one side only; from the branch piping side, The pipe branch connection geometry makes the Code-required 100% volumetric examination impractical. To obtain complete volumetric coverage, modification or replacement of the branch connections with connections of a design providing for complete coverage would be required, imposition of this requirernent would cause a
)
considerable burden on the licensee.
The licensee proposed to examine the subject welds to the exterit possible. Based on a review of the coverage plots provided, it appears that a significant portion
(> 50%) of the welds is examined with a one-directional beam from the branch piping side. Based on the significant amount of the volume examined from one side, it is reasonable to conclude that degradation, if present, will be detected. As a result, reasonable assurance of continued structuralintegrity will be provided.
l 8
Therefore,it is recommended that relief be granted, pursuant to 10 CFR 50.55a(g)(6)(i). -
2.4 Recuest for Relief BV2 B1.11 1. Rev. O. Examination Cateaorv B A Item B1.11.
Reactor Pressure Vessel Head Circumferential Weld (Unit 2 First Intervall I
NL,TE: This request was submitted, pursuant to 10 CFR 50.55a(g)(5)(iii), to satisfy 1
the ASME Section XI requirements cited above, as well as those listed in 10 CFR 50.55a(g)(6)(ii)(A), Augmented exam /nat/on of reactor vessel. However, while a request under 10 CFR BO.55alg)(5)(ill) is acceptable for ASME requirements deemed impractical, it is not allowed under the augmented examination rule.
Further, since all existing requests for relief wtre revoked by the augmented rule, it is not appropriate to address the current request prior to the augmented rule being satisfied. Therefore,it is recommended that evaluation of Request for Relief BV2 B1.11 1 be deferred until such time as the licensee has satisfied the regulations for the augmented reactor vessel examination.
2.5 Relief Raouest BV2 B3.110 2 Rev. O. Examination Cateoorv B D. Item B3.110.
Pressurizer Nozzle-to Shell Welds (Unit 2 - First intervall
.C.qde Reautrement: Examination Category B D, item B3.110 requires 100%
volumetric examination of all pressurizer nozzle to shell welds as defined in Figures IWB 2500 7(a) through (d), as applicable.
Licensee's Relief Reauest: Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from 100% volumetric examination cf pressurizar nozzle to shell Welds 2RCS* PRE 21 N 10 (6" relief nozzle), 2RCS' PRE 21-N 1i '6" relief nozzle),
2RCS' PRE 21 12 (6" relief nozzle), 2RCS' PRE 21-N 13 (6" relief nozzle),
2RCS* PRE 21 N 14 (4" spray nozzle).
l
. 9 Licennan's Ramla For Malief (as stated):
"In accordance with 10 CFR 50.55alg)(6)G), relief is requested on the basis that c6mpliance with the Code requirement is impractical. Ultrasonic examinations were pr. normed on the Pressurizer nouis to vessel upper head welds. These examinations were performed on the Pressur!rer nonle to vessel upper head welds.
These examinations were limited due to the norrie curvature. The curvature of the L outside surface caused the transducer to lose contact with the surface. Each of these welds are essentially identical in configuration. The table' below identifies the j examination coverage for the welds for each scen angle used.
"The examination volumo calculations are based on the required examination volume iilustrated in ASME Section XI, Figure IWB 2500 7(b). The limitations preventing full coverage are inherent to the design of these nozzles. There are no current NDE methods that would appreciably increase the amount of the examination coverage (see attached sketch for the nonle configuration and the scan limitations').
Licennan's Prononad Alternatlyg (as stated):
"The alternative to the examination requirement is to perform the ultrasonic examination to the maximum extent possible.- The UT tect nique used provides the maximum coverage possible, considering the nonle design and materials used in fabrication."
Evaluation: The Code requires volumetric examination of the pressurizer nozzle to-vessel welds. However, volumetric examination of the subject tellet and spray nonle to-vessel welds is limited due to the radius of curvature in the transition area between the nonle and the vessel shell. Therefore, the noules' geometric design configuration makes the volumetric examination impractical to perform to the extent required by the Code. To meet the Code requirements, the nonles would have to be modified to facilitate access for ultrasonic search units, imposition of this nequirement would create a considerable burden on the licensee.
The licensee examined the subject welds to the extent possible obtaining approximately 57% coverage of the required area for occh weld. As a result, the INEEL believes that significant patterns of degradation, if existing, would be
'8 The table provided by the licensee is not included in this evaluation.
Sketches provided by the licensee are not included with this evaluation.
- ~ ~ -
O t
{ 10 detected, providing reasonable assurance of the structuralintegrity. Therefore, pursuant to 10 CFR 50.55stg)(6)(1), it is recommended that relief be granted.
2.6 Ranuant for Rollef BV2 D2.201. Rev 0. ASME Coda Clann 3 Intaarally Waldad Attachments (Unit 2 First Interval)
Code Raouirement: ASME Section XI, Examination Category D A, D 8, and D C, Items D1.20 through D1.60, D2.20 through D2.60, and D3.20 through D3.60 requira VT 3 visual examination as defined by Figure IWD 25001.
Licennan's Prononad Alternativa: In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an altomative to the Code requirements associated with the selection and examination of integrally welded attachments. The licensee stated:
"The altemate examination is to implement the sampling criteria of Code Case N 509 to BV t Class 3 welded attachments for the current ten year interval, ending November 16,1997. This requires examination of 10% of the welded attachments.
Ninety seven (97) percent of the total number of Class 3 welded attachments have been examined during this present interval. Approval of this alternative will preclude the hardship associated with effort required to gain access to perform visual examinations on the remaining 3% of welded attachments (two supports). A best effort examination of these two supports was performed with the insulation installed. No deformation or any cther anomaly was reported."
Licennan's Baalm for the Pronomad Alternative (as stated):
"The ASME Section )rl Subcommittee has approved and published Code Case N 509.
The latest revision of Regulatory Guide 1.147 does not include this code case. For Class 3 welded attachments, the code case requires examination of a 10% sample opposed to the 100% exam requirement of the ASME XI 1983 Edition, Summer 1983 Addenda (BVPS code of record). Justification of this code case was based on the costs, in both radiation exposure and dollars, associated with these examinations versus the added benefit of the examinations.- The justification relief on industry operating experience associated with ISI of welded attachments over the last 20 years - This information was obtained through an extensive industry survey along with a Licensee Event Report search. The ASME XI Code originally established a visual examination requirement for Class 3 component supports and their welded attachments. These components were assigned a relatively low inspection priority (visual) because isolated failures were not generally considered a threat to plant
' safety. The ASME XI Code established these requirements on the welded attachments to preclude, by examination, occurrences of a stress induced flaw,
I e
11 initiating a failure of a pressure retalning component. Only one such failure has occurred in the industry. This failure was of limited significance to the safety of the plant and caused no danger to the health and safety of the public. The failure was discovered when leal < age was observed during normal operation and not by an ISI {
examination."
"The BV 2 ten year plan scheduled 428 Class 3 welded attachments for examinat;on. Four hundred and sixteen (416) welded attachments have been examined to date. All were acceptable. Thus,97% of the scheduled Class 3 welded attachments have been examined during the current interval. These examinations provide an adequate level of assurance of the integrity of this j component category. Cons'dering the past 20 years of industry wide experience concerning welded attachments, removal of the surrounding piping and conduit constitutes hardship and unusual difficulty without a compensating increase in the level of safety. The 97% completion rate far exceeds the alternate rules of Code Case N 509 for a 10% sampling requiremem "The examination results obtained on the 416 welded attachments are consistent with the industry experience compiled in support of the Code Case = 309. No defects were identified on the 416 examinations. Therefore, a 100% examination requirement is considered unwarranted due to the absence of identified failures at BV 2 and throughout the nuclear industry.
"It should be noted that all of the BV 2 Class 3 welded attachment examinations performed to date were VT 3 examinations. Case N 509 requires a VT 1 examination for Class 3 welded attachments. This difference is considered significant. The only condition specifically noted in IWA 2211 for a VT 1 examination that is not specifically noted in IWA 2213 for a VT-3, is cracks, if a crack avas observed during a VT 3 examination, it would be reported since it impacts the general structural condition of the welded attachment."
Evaluation: In lieu of Code requirements for selection and examination of integral attachment welds, the licensee proposes to apply alternatives contained in Code Case N 509, Allemative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments for Class 3 integral attachment welds only. Code Case N 509 has reduced the percent of areas to be examined during ecch interval from 100% to 10%, for all components. This Code Case has previously been found ecceptable by the staff, provided that licensee's examine a minimum of 10% of all Code Class 1,2, and 3 piping, pumps, and valve integrally welded attachments. The licensee has met this provision for Code Class 3 components.
I 12 Considering that the majorit) sf Code examination requirements are band on component sampling to assure that service related degradation is not occurring, it is logical to extend the sampling process to welded integral attachments. The licensee has examined 97% of the Class 3 integral attachment welds during this interval. No relevant service induced degradation was observed. Based on the fact that the licenses has examined a large percent of Class 3 Integral attachment welds, the l
1 lNEEL staff believes that degradation, had it occurred, would have been detected.
Since the conditions for use of the alternatives contained in Code Case N 509, for Code Class 3 components, have been satisfied, an acceptable level of quality and safety has been provided. Therefore, it is recommended that the licensee's proposed alternative for Class 3 Integrally welded attachments be authorized pursuant to 10 CFR 50.55ata)(3)(1). Use of alternatives contained in Code Case N 50g should be authorized for the first interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time,if the licensee intends to continue to implement this code case for Class 3 items, the licensee should follow all provisions in Code Case N 50g, with limitations issued in Regulatory Guide 1.147, if any.
-2.7 Raouest for Relief BV3 IWA 1. Rev. O. lWA 5242(a). Insulation Removal For VT-2 Visual Examination Of Boltina in Class 1 and 2. Borated Systems (Unit 1 - Third Interval. and Unit 2 Second Interval)
Code Raouirement: IWA 5242(a) requires that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure totaining bolted connections for VT-2 visual examination.
Licensee's Pronomad Alternative Examination: Pursuant to 10 CFR 50.55a(a)(3)(ii),
the licensee proposed an alternative to the ASME Section XI requirements for removing insulation from Class 1 and 2, pressure-rataining bolted connections for VT-2 visual examination during VT 2 visual examinations. The licensee stated:
"A visual, VT 2 examination will be performed during the system pressure tests of IWB 5000 and IWC 5000 with the insulation installed, in addition, insulation will be p- - -
1 e
13 removed from all Class 1 bolted connections each refueling outago and from all Class 2 bolted connections on a frequency of approximatoly 3 and 1/3 years (every other refueling outage, but not to exceed once per period), when the Codo requirement, regarding insulation removal (lWA 5242), is not met. A visual VT 2 examination shall be performed, with the insulation removed, for evidence of leakage (boric acid residue) Independent of the system pressure test.
I "Duquesne Light requests NRC approval of the proposed attemative requirements for VT 2 visual examination of Class 1 end 2 insulated pressure retalning bolted connections in lieu of those of IWA 5242(a). NRC approval is requested in a time frame that will support implementation of the alternative requirements for the Unit 1 twelf th refueling outage currently scheduled to begin in September of 1997. Unit 1 tweif th refueling outage will be the first outage of the third ten year interval at Beaver Valley Power Station Unit No.1. The second ten year interval at Beaver Valloy Power Station Unit 2 will begin November 17,1997.
Licensee's Basis for the Pronosed Attemative (as stated):
"In accordance with 10 CFR50.55ala)(3)(ii) relief is requested from ASME I Section XI,1989 Edition, IWA 5242(a). IWA 5242(a) requires that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure retelning bolting for visual examination VT 2. Compliance with this code requirement would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety, based on the following:
1.
Code Class 1 and 2 systems borated for the purpose of controlling reactivity are large systems covering many areas and elevations of the reactor containment building. Many of the bolted connections are located in areas difficult to access and have significant radiation levels, scaffolding is required to access the majority of the bolted connections. As required by IWB 5221, the System Leakage Test is conducted at normal operating pressure and temperature. The visual, VT 2 examination is performed prior to plant startup at mode 3. This examination is a critical path activity and normally has a short duration of approximately 4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. To delay reactor criticality due to installation of insulation and disassembly and removal of scaffolding under subatmospheric conditions is an undue burden without a compensating increase in the level of quality and safety. Outage durations and costs will be significantly increased and personnel safety will be impacted under these operating conditions.
- 2. " Leakage of borated water willleave a residue of boron. Boron residue is evidence of leakage and inconsequential to whether the system piping is under pressure or not at the time of the visual VT-2 examination."
Evaluatien: The Code requires the removal of allinsulation from pressure retaining bolted connections in systems borated for the purpose of controlling reactivity when
e
_, -14 performing VT 2 visual examinations during system pressure tests. As an alternative, the licensee has pro;>osed to perform the VT 2 visual examination with a minimum 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time with the insulation in place. in addition, the insulation will be removed for direct visual examination each refueling outage for Class 1, and each period for Class 2 components.
l The licensee's proposed alternative le similar to that found in Code Case N 533, Alternative Requirements for VT 2 VisualExamination of Class 1 Insulated Pressure.
Retalning Bolted Connections,Section X.'. Division 1, except the preposed alternative was extended to address Code Class 2 bored connections. Code Case N 533 is currently under review by the NRC staff and has not yet been approved for general use by incorporation into Regulator Guide 1.147, inservice inspection Code Case
- Acceptability; however, it has been found acceptable on a plant specific basis with the commitment of a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time on insulated systems.
For Class 1 systems, the licensee's proposed altamative provides a thorough approach for ensuring the leak tight integrity of systems borated for the purpose of controlling reactivity. First, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time allows time for any leakage to penetrate the insulation, thus provides a means of detecting any significant leakage with the insulation in place. Second, by removing the insulation each refueling outage, the licensee will be able to detect minor leakage that could occur by observing for the presence of boric acid crystals or residue. This two phased approach will provide an acceptable level of quality and safety for bolted connections in borated systems. Therefore, it is recommended that the licensee's proposed
- alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for use on Class 1 systems.
For Class 2 systems, the frequencies proposed for insulation removal have not been found acceptable by the NRC staff. Therefore, the licensee's proposed alternative should n01 be authorized for Class 2 systems at the present time.
aw a
o 15 2.8 Reauest for Relief BV3 lWA-2. Rev. O lWA 5250fa)(2). Corrective Action Resultinc
{
from Leakaoe at Bolted Connections (Unit 1 - Third interval, and Unit 2 Second l Interval)
Code Reauirement: IWA 5250(a)(2) requires that the source of leakages detected during a system pressure test be located and evaluated by the Owner for corrective action. When the leakage is at a bolted connection, the botting shall be removed, VT 3 visually examined for corrosion, and evaluated in accordance with IWA 3100.
Licensee's Procosed Alternative Examination: Pursuant to 10 CFR 50.55a(a)(3)(ii),
the licensee proposed an alternative to the ASME Section XI requirements for removal of botting at leaking connections for VT 3 visual examination. The licensee stated:
i
" Leakage on Code Class 1,2, and 3 bolted connections shall be evaluated to determine the susceptibility of the bolting to corrosion and potential future failure.
This review will consider as a minimum, location of leakage, history of leakage.
l bolted connection materials, visual evidence of corrosion with the connecting i
assembled, corrosiveness of the process fluid, history and studies of similar bolted materialin a similar environment, and other components in the vicinity that may be degraded due to the leakage. If this review assures the integrity of the connection, no bolting will be removed for examination. If this review cannot assure the integrity of the connection, the bolt closest to the leakage source shall be removed VT 1 examined and evaluated in accordance with IWA 3100(a). When the removed bolt has rejectable degradation, the remaining bolts will be removed, VT 3 examined and evaluated in accordance with IWA 3100(a). If no degradation is observed on the removed bolt, no further action will be required."
Licensee's Basis for the Prooosed Alternative (as stated):
"In accoruance with 10CFR50.55ata)(3)(ii) relief is requested from the ASME Section XI,1989 Edition, Subparagraph lWA 5250(a)(2) requirement that if leakage occurs at a bolted connection, the botting shall be removed, VT-3 visually examined for corrosion and evaluated in accordance with IWA 3100. Compliance with the Code requirements would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety. Leakage at a bolted connection may be a significant variable in the degradation mechanism of bolted connections. However ,leekage is not the only variable and in some cases may not be the degradation mechanism. Bolting materials, leaking medium, duration of the leak, and location of the leak are other variables to be considered. Removal of botting at a mechanical connection may not be the most prudent decision to make.
f 0
e 16 All variables should be considered before disassembling w bolted connection for a visual VT 3 examination. Duquesne Light proposes an alternative to the ,
requiremen s of IWA 5250(a)(2) that will provide an equivalent level of quality and safety at Class 1,2, and 3 bolted connections."
Evaluation
- In accordance with the 1989 Edition of the Code, when leakage occurs at bolted connections, all botting is required to be removed for VT 3 visual I examination, in lieu of the Code roquired removal of bolting to perform a VT 3 visual I examination, the licensee has proposed to perform an evaluation of the bolted connection to determine the susceptibility of the botting to corrosion and the potential 'or isilure.
This alternative allows the licensee to utilize a systematic approach and sound engineering judgement; provided, as a minimum, all evaluation factors listed in the licensee's proposed alternative are considered. Further nore, if the initial evaluation indicates the need for a more in depth evaluation, the bolt closest to the source of leakage will be removed, VT 1 examined, and evaluated in accordance with IWA 3100(a).
Removal of all botting as part of the corrective action when leakage is observed at a bolted connection can result in a significant burden on the licensee. Based on the licensee's proposed alternative, to perform an evaluation of the botting to determine the potential for degradation, it is reasonable to conclude that the operational readiness of bolted connections will be maintained. Therefore,it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
s
0 I
-o 17
3.0 CONCLUSION
The INEEL staff has reviewed the information providad by the licensee for requests for i
relief at Beaver Valley Power Station, Units 1 and 2. Based on the evaluation of Relief Requests BV133.120 2 and BV3 lWA 2, it has been determined that imposing the Code requirements on the licensee will result in a burden without a compensating increase in quality and safety. Therefore, the INEEL recommends that the licensee's proposed altematives be authorized pursuant to 10 CFR 50.55a(s)(3)(ii). It is also recommended, pursuant to 10 CFR 50.55a(s)(3)(ii), that the licensee's proposed attemative in Relief Request BV3 lWA 1 be authorized for Class 1 bolted connections only. For Class 2 l systems, the frequencies proposed for insulation removal have not been found acceptable by the NRC staff. Therefore, the licensee's proposed altamative should not be authorized l for Class 2 systems at the present time.
For Relief Request BV2 D2.201, it is concluded that the licensee's proposed altemative provides an acceptable level of quality and safety. Therefore, it is recommended that the proposed alternative be authorized for Class 3 Integrally welded attachments, pursuant to 10 CFR 50.55a(a)(3)(l).
For Relief Requests BV189.312 and BV2 83.110 2, it has been determined that the ASME Code examination requirements are impractical. Therefore, it is recommended that relief be granted, pursuant to 10 CFR 50.55a(g)(6)(1).
Requests for Relief BV1 RV-Welds and BV2 B1.11-1 were ,ubmitted by the licensee to address limited ASME Code examination coverage for reactor pressure vessel welds.
However, the licensee has yet to fulfill the augmented examinat/on requirements found in 10 CFR 50.55alg)(6)(ii)(A) for welds that were not essentially 100% e::= mined. Therefore,-
it is recommended that evaluation of these requests be deferred until such time as the licensee has satisfied the regulations by subr.11tting an alternative to the augmented examination for reactor pressure vessel welds where essentially 100% volumetric coverage was not obtained.