ML20206H249
| ML20206H249 | |
| Person / Time | |
|---|---|
| Issue date: | 05/07/1999 |
| From: | Charemagne Grimes NRC (Affiliation Not Assigned) |
| To: | Walters D NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| PROJECT-690 NUDOCS 9905110117 | |
| Download: ML20206H249 (13) | |
Text
p May 7,.1999
- Mr. Douglas J. Walters i
Nuclear Energy Institute 1776 l Street, NW., Suite 400
' Washington, DC 20006-3708
SUBJECT:
' LICENSE RENEWAL ISSUE No. 98-0052, " LICENSE RENEWAL"
Dear Mr. Walters:
By letter dated May 8,1998, the Nuclear Energy Institute (NEI) provided the NRC staff with comments on the Working Draft Standard Review Plan for License Renewal (SRP-LR). Several comments were related to existing ASME Code Section XI, Subsection IWE and Subsection IWL
. programs for inservice inspection of PWR and BWR containment structures. By letter dated i
November 4,1998, NEl provided supplemental comments to the IWE/lWL issues that were raised initially in the May 8,1998, letter. The staff reviewed the NEl comments provided by both the May 8, and the November 4,1998, letters and drafted a consolidated description of license renewalissues on aging management of containments. On January 8,1999, the staff transmitted a facsimile of the consolidated description of license renewal issues on aging management of containments to NEl for their comment.
On January 25,1999, the staff (Fred Bower, DRIP /PDLR; and Sam Lee, DRIP /PDLR) had a conference call with NEl (Doug Walters, NEl; John Carey, EPRl; and Bob Nickell, EPRI) to further clarify the issues documented in the facsimile. Some issue were clarified, however, some issues required additional clarification (see conference call summary dated January 28,1999).
- On March 16,1999, the staff (Fred Bower, DRIP /PDLR; and Sam Lee, DRIP /PDLR) had an additional conference call with NEl (Doug Walters) to further clarify the issues documented in the j
facsimile. The consolidated description of license renewal issues on aging management of
. containments was revised to incorporate the results of the conference calls. The revised list is provided in Enclosure 1.
i The March 16,1999, conference call was also used to clerify other selected priority 1 license renewal issues. The clarification of these issues is included in Enclosure 2.
4 The staff is proceeding to evaluate priority 1 license renewal issues as clarified in the enclosures.
If there are any industry comments on the description of the license renewal issues in the enclosures, we request that you document those comments within 14 days following your receipt of this letter. If you have any questions regarding this matter, please contact Sam Lee at 301-415-3109 or Fred Bower at 301-415-1488.
i Sincerely, original signed by:
Christopher I. Grimes, Director License Renewal Project Directorate Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation Project No. 690
/
Enclosures:
As Stated
. cc: See next page
{9{Cil.
gy 7
A[
Distribution: See next page gn p '//O 3 DOCUMENT NAME:G:\\ BOWER \\PRIORTYtWPD '
MU OFFICE LA /j PDLR PDLR:SQ g PDLR:D hiton FLBower PTKuo [ 4-CIGrim@
NAME DATE 3 ///99 '
3 /2[/99 M99 g / Q /99.
h 9
OFFICIAL RECORD COPY PDR REVOP ERONUMRC
//
1' 9905110117 990507 4
f j
NUCLEAR ENERGY. INSTITUTE L
(Ucense Renewal Steering Committee)
Project No. 690 cc:
Mr. Dennis Harrison Mr. Robert Gill U.S. Department of Energy
- Duke Energy Corporation NE-42 Mail Stop EC-12R Washington, D.C. 20585 P.O. Box 1006 Charlotte, NC 28201-1006 Mr. Ricard P. Sedano, Commissioner Mr. Charles R. Pierce State Ualson Officer Southern Nuclear Operating Co.
Department of Public Service 40 inverness Center Parkway 112 State Street BIN B064 j
Drawer 20 Birmingham, AL 35242 Montipelier, Vermont 05620-2601 Mr. Douglas J. Walters Mr. Barth Doroshe Nuclear Energy institute Baltimore Gas & Ebetric Company 1776 i Street, N.W.
1650 Calvert Cliffs Farkway
- Washington, DC 20006
. Lusby, Maryland 20657-47027
. DJW@NEl.ORG National Whistleblower Center Chattooga River Wateished Coalition 3233 P Street, N.W.
P. O. Box 2006 Washington, DC 20007 Clayton, GA 30525 Mr. William H. Mackay Entergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 GSB-2E Russellville, Arkansas 72802 4
L h
p:--
e.
i I
Distribution:
(
Hard coov.
PUBLIC Docket File -
~
PDLR RF
[
.N. Dudley, ACRS - T2E26 E-mail:
. R. Zimmerman
)
W. Kane.
1 l-D. Matthews S. Newberry C. Grimes F. Akstulewicz-l J. Strosnider R. Wessman G. Bagchi l
H. Brammer T. Hiltz G. Holahan C. Gratton R. Correia R. Latta J. Moore R. Weisman M. Zobler G. Mizuno F. Cherny -
E. Hackett i.
~ A. Murphy _
D. Martin W. McDowell S. Droggitis PDLR Staff G. Tracy L
A. Thadani L
110050 l
l'
n
=,-
i i
DESCRIPTION OF LICENSE RENEWAL ISSUES ON AGING MANAGEMENT OF CONTAINMENTS (SUBSECTIONS IWE AND IWL OF ASME SECTION XI)
License Renewal Issue 08-0048. " Elevated Temperature of Prestressino Tendons"
-i l
Priority 1 Backaround l-Pages 18 and 38 of NUREG-1611 indicate that elevated temperatures may increase the i
prestress loss in containment tendons. Thus, an applicant should augment the tendon.
' surveillance program to include additional tendons selected based on their sun exposure or proximity to hot penetrations. This information is incorporated into the draft standard review plan for license renewal on pages 3.3-1.2 and 3.4-15.
NEl Comments The selection of the 4% prestressing tendon sample, in accordance with IWL-2521, is supposed to be random. The NRC st'aff has the option of modifying Regulatory Guide 1,35 to provide guidance on the selection of tendons to include one tendon from such regions of the containment potentially exposed to higher temperature. Selecting additional tendons -
beyond the 4% sample size need not be the only option. In addition, if the argument that a relatively slight increase in prestressing tendon service temperature (from 20C to 32C) caused by exposure to sunlight could double the loss of prestress is valid, the NRC staff should notify the appropriate ASME Code Section XI bodies to consider adding a paragraph to IWL-2521 to permit tendon selection to be based, in part, on environmental conditions.
This should not be solely a license renewal consideration.
License Renewal issue 98-0049. "Aaina Manaaement of Inaccessible Areas of Containments" Priority 1 Backaround NUREG-1611 indicates that an applicant should manage the potential aging effects of containment structures in inaccessible areas when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas. This information is incorporated into the draft standard review plan for license renewal on pages 3.3-10 and 3.4-19 NEl Comments For (1)~one-time inspection for license renewal, (2) leaching of calcium hydroxide, (3) aggressive chemical attack, (4) corrosion of structural steel and liners, and (5) corrosion of
c
. embedded reinforcing steel, NUREG-1611 and the working draft standard review plan for license renewal go well beyond the additional requirements of the 10 CFR 50.55a rulemaking, recommending augmented inspection of inaccessible areas even when no evidence exists in an accessible area that potential degradation has taken place.
The PWR Containment Industry Report identified corrosion of below-grade concrete and steel structures caused by aggressive ground water as a significant aging effect for which current periodic inservice inspection programs could not be shown to be adequate. A similar conclusion was reached in the BWR Containment Industry Report. Both Industry Reports suggested plant-specific programs to address the issue, starting with a determination of ground water chemistry, in particular the thresholds for aggressive acidity (pH<5.5), chloride content (>500 ppm), and sulfate content (>1500 ppm). For below-grade concrete and steel containment structures that are not in direct contact with aggressive ground water, no further action was deemed to be required. When aggressive ground water is in direct contact with below-grade concrete or steel containment structures, a
. number of alternatives were suggested, including the installation of a barrier system (e.g., a water-proof membrane), reducing the water table to prevent direct contact, or sampled inspections of inaccessible regions below the water table. In no case did either of the Industry Reports suggest augmented inspection of inaccessible regions when no evidence is present in accessible areas that would indicate potential degradation.
No justification for the augmentation covered in this item could be found, either in NUREG-1611 or in the working draft standard review plan for license renewal, which
~ implements the NUREG-1611 conclusion. Therefore, it is our view that the guic'ance in the Industry Reports is adequate and inspections are not required unless specified in the Industry Report, or in 10 CFR 50.55a.
License Renewal Issue 98-0050. "lWEllWL to include Basemat" Priority 3 Backaround Although Section 3.4 of the draft standard review plan for license renewal indicates that it does not address certain basemats, it addresses the potential aging effects of the sub-foundation layers below the basemats. In accordance with the BWR Containment Industry Report, SRP-LR page 3.4-1 identifies concrete containment as part of the pressure boundary. Section 3.3 indicates that the affected PWR containment structure components include the concrete basemat. NtEl agreed to delete this item.
NEl Comments During the telephone conference on March 16,1999, NEl agreed to delete this item since the concrete basemat is part of the pressure boundary for the Mark lli containments. NEl also agreed that the draft standard review plan for license renewal is consistent with the BWR Containment industry Report.
i
m a 3-
~
License Renewal issue 98-0051. "lWE/lWF: Jurisdiction"
. Priority 3 Backaround Page 46 of NUREG-1611 indicates that wear of BWR containment structures and supports would be adequately managed by Subsections IWE and IWF of Section XI of the ASME Code. Pages 33 and 34 of NUREG-1611 indicates that corrosion of BWR containment structures and supports would be adequately managed by Subsection IWE only. This information is incorporated into the draft standard review plan for license renewal.
NEl Comments Sections 3.4.ll.C.6 and 3.4.lli.C.6 of the draft standard review plan for license renewal call c,ut Examination Category F-A as a program adequate to manage the effects of lockup on certain BWR containment support structures, such as vent system supports for Mark i steel containments. Examination Category F-A should also be cited as an adequate program for managing the effects of loss of material (corrosion) in Sections 3.4.ll.C.10 and 3.4.lll.C.10.
License Renewal Issue 98-0052. "lWE/lWL Operatina Experience Reauirements" Priority 1 Backaround Sections 3.3 and 3.4 of the draft standard review plan for license renewal indicate that an applicant should discuss operating experience of their IWE/lWL containment inservice inspection program.
NEl Comments-l The draft standard review plan for license renewal includes a requirement for the applicant to discuss the operating experience with various Subsection IWEllWL inspections, including any corrective actions from these inspections that may have provided feedback to the plant owner, and that may have led to either inspection program enhancements or new inspections programs. Such an additi.onal requirement, over and above a determination by
- the applicant that Subsection IWE/lWL inspections manage the effects of aging through the license renewal term, seems unwarranted.
]
. License RenewalIssue 98-0046 " Inspection of Containment Welds and Bellows" Priority 2 Backoround Page 3 of NUREG-1611 indicates that an applicant should perform inspections of containment pressure retaining welds in accordance with Examination Categories E-B and E-F of Section XI of the ASME Code. In addition, NUREG-1611 indicates that an applicant should perform augmented inspections of containment penetration bellows to manage potential cracking. This information is incorporated into the draft standard review plan for license renewal.
NEl Comments This item derives from information Notice 92-20, " Inadequate Local Leak Rate Testing,"
dated March 23,1992, based on an incident at the Quad Cities Station, Unit 1, during the conduct of a Type B local leak rate test (LLRT) on a containment penetration bellows.
Excessive air leakage from the penetration was attributed to " cracks identified by the tests,"
but the cracking was not characterized. The Containment Industry Report recognized some potential for stress corrosion cracking of the dissimilar metal weld between the stainless steel bellows assembly and the carbon steel containment penetration sleeve, which represents a creviced geometry. However, the attachment design minimizes operational stresses from thermal cycling or pressure testing, and the environment is not aggressive with respect to chlorides, sulfates, or acidity. For those few cases where plant-specific conditions are such that this finding does not apply, the surface examinations of Exwication Cat _egory E-F (pressure-retaining dissimilar metal welds) and the visual ery,, nations JExamination Category E-P (Appendix J leak tests), are more than adequate to detect the crac!ng, and no augmented inspections are necessary. This item also represents an area of disagreement by the NRC staff in the License Renewal Project Directorate with the 10 CFR 50.55a rulemaking, where examinations of dissimilar metal welds, as required by Subsection IWE, were determined to have little safety benefit and were therefore determined to be optional. It is our position that the current requirements of 10 CFR 50.55a are adequate for both the current and renewal terms, including exam categories E-B and E-F."
License Renewal Issue 98-0040. " Freeze-Thaw Damaae in Concrete Containment Structures"
' Priority 3 Backoround NUREG-1611 and the draft standard review plan for license renewal indicate that acceptable methods for managing the effects of freeze-thaw in concrete containment structures are Subsection IWL of Section XI of the ASME Code.
FT
- p....<
o
, NEl Comments-
~
The industry technical position and its justification for freeze-thaw are provided in EPRI TR-107521, " Generic License Renewal Technical issue Summary," April 1998. This position provides criteria for determination of potential significance of freeze-thaw damage.
License Renewal issue 98-0041. " Alkali-Aooreaate Reactions in Concrete Containme_nt Structures" 1
Prioritv_3 Backaround NUREG-1611 and the draft standard review plan for license renewal indicate that acceptable methods for managing the effects of reaction with aggregates are Subsection IWL of Section XI of the ASME Code and the requirements in 10 CFR 50.55a.
L NEl Comments i
The industry technical position and its justification for alkali-aggregate reactions in concrete containment structures are provided in EPRI TR-107521. This position provides criteria for determination of potential significance of alkali-aggregate reactions in concrete containment structures.
License Renewal issue 98-0042. " Differential Settlement of PWR Containments and Class l Structures" Priority 3 Backaround NUREG-1611 and the draft standard review plan for license renewal indicate that acceptable methods for managing the effects of settlement are settlement monitoring during construction and continued monitoring during operation for sites with soft soil and/or significant changes in ground water conditions. This position provides criteria for determination of potential significance of differential settlement of PWR containments and class 1 structures.
NEl Comments
- The industry technical position and its justification for differential settlement of PWR l
containment and Class I structures are provided in EPRI TR-107521.
L-i l
4
o h
.,. License Renewal issue 98-0043. " Reinforcement Corrosion in PWR Containments" I'
Priority 3 I
Backaround NUREG-1611 and the draft standard review plan for license renewal indicate that l-acceptable methods for managing the effects of reinforcement corrosion in PWR
- containments are Subsection IWL of Section XI of the ASME Code and the requirements of 10 CFR 50.55a.' In addition, the management of inaccessible areas of embedded steel
. should be evaluated.
j NEl Comments '
i The industry technical position and its justification for reinforcement corrosion in PWR containments are provided in EPRI TR-107521. This position provides criteria for determination of potential significance of reinforcement corrosion in PWR conts.rments.
License Renewal issue 98-0084. "Aoina Review of Airlocks and Eauipment Hatches" Priority 3 packaround NUREG-1611 indicates that mechanical wear of pressure retaining components, including airlocks and equipment hatches, should be adequately managed by containment inservice inspections in accordance with Subsection IWE of Section XI of the ASME Code. This
- information is incorporated into the draft standard review plan for license renewal. This item is deleted since the issue will be addressed by the resolution of License Renewal issue 98-0012 on consumables.
NEl Comments NEl and EPRI agreed that the item should be deleted since the issue will be addressed by the resolution of License Renewal Issue 98-0012 on consumables.
License Renewal issue 99-106. " Ultrasonic Inspection Qualifications for Containments" 1 Priority 3
' Backaround p
.Page 2 of NUREG-1611 indicates that ASME Section XI, Appendices Vll and Vill should be j
implemented when ultrasonic examinations are utilized for the inspection of containments.
This information is incorporated into the draft standard review plan for license renewal on pages 3.3 56 and 3.4-71,
I-l NEl Comments.
This additional requirement does not seem to be applicable, and it is not clear why such a requirement is being imposed. The only volumetric examinations required by either l
~ Subsection IWE or IWL are those of Examination Category E-C for determining the wall
]
thickness of steel containment structures. The purpose of the personnel qualifications of j
' Appendix Vil and the performance demonstrations of Appendix Vill is to detect and size J
flaws in components for which standard flat-bottom hole calibration blocks are not representative, especially in situations where straight-beam methods could be unreliable.
i Measuring ultrasonic wave propagation across wall thickness is completely unrelated to these issues. The supplemental examinations of IWE-3200 could trigger surface examinations to size flaw lengths, but are not likely to trigger volumetric examination. Even if supplemental volumetric examination were to be considered, in order to justify continued j
operation without repair, the attemative of a containment shell repair in accordance with IWA-4150 would be preferable, especially since no calibration specimen satisfying Appendix j
Vill for containment shells is available.
License Renewal issue 99-107. " Erosion of Porous Containment Sub-foundation"
' Priority 3 Backaround i
NUREG-1611 indicates that an applicant should evaluate erosion of cement for porous concrete if sub-foundation layers of porous concrete are used in the construction of containment concrete basemat with the presence of underground water. This information is incorporated into the draft standard review plan for license renewal.
NEl Comments l
The Containment Industry Reports found that the effects from leaching of calcium hydroxide were not significant for concrete containment structures unless such structures were exposed to flowing water and, if exposed to flowing water, could not be shown to be constructed of dense, well-cured concrete. Porous concrete exposed to flowing water would I
fall into the category such that the effects of leaching could be potentially significant.
. Information Notices No.96-145, dated October 17,1996; 97-11, dated March 21,1997; and 98 26, dated July 24,1998; have addressed this concern for nine U. S. nuclear plants
, (Beaver Valley Units 1 and 2, Fitzpatrick, Haddam Neck, North Anna Units 1 and 2, Surry Units 1 and 2, and Perry), following the incidence of reported erosion of the containment foundation basemat at the Millstone Unit 3 Nuclear Power Plant. Information Notice 96-145 stated that there was no safety concem, since the Millstone Unit 3 containment foundation,
. Is founded on rock. Also, the amount of concrete that may have eroded since plant j
construction is only a small fraction of the total cement weight, and no adverse consequences have either been observed or predicted.
\\
l It is not clear why this relatively minor current plant issue is being called out in
' NUREG-1611, which will have been addressed by plants looking for evidence of slurry in j
f.
the drainage below their containment basemats, and for any settlement of the containment basemats. NRC also asked these plants if they have noticed any unusual conditions which may be related to the porous concrete sub foundation layers and if they are monitoring anything related to the drainage from the porous concrete sub-foundation layers below their containment basemats. The most recent Information Notice 98-26 has indicated that the NRC staff may inspect licensee sub-foundation monitoring and preventive maintenance programs as part of NRC Inspection Procedures 62002 and 62003. It is our position that this issue only applies to a few plants, and that the current practices that the subject plants have implemented are adequate to manage this aging effect for the renewal term.
i l
l
^
- DESCRIPTION OF CHANGES TO PRIORITY 1 LICENSE RENEWAL ISSUES License Renewal issue 98-0057. "Creditina Maintenance Rule Proaram for Monitorina Structural Aoina" Priority 1 Issue Clarification NEl agreed to submit information by March 25,1999, to clarify this issue.
License RenewalIssue 98-0068 "The Use of Code Proaram and Code Additions" Priority 1 Issue Clarification NEl agreed to submit information by March 25,1999, to clarify on this issue.
License Renewal issue 98-0082. "Cascadina for Nonsafety-Related Systems Within the Scope of 10 CFR Part 54" Priority 1 issue Clariification Staff guidance is required to identify the bounding scope of systems, structures, and components (SSCs) required to be included within the scope of the rule under 10 CFR 54._4(a)(3). More specifically this guidance will specify whether the applicant needs to consider those SCCs whose failure could prevent satisfactory accomplishment of any of the functions identified by the applicable Commission regulations, and the need to consider those SSCs that " cascade" from a review of second, third, and fourth-level support
- systems, 3
License Renewal issue 98-0086. " Inspection of Pressurizer Heater Bundle Partial Penetrations" Priority 1 issue Clarification i
This issue relates to the need for inspection of the pressurizer heater bundle partial
' penetration welds. NEl agreed to delete this item because Duke Energy has proposed a program for these pressurizer inspections for Oconee.
4
p 1
\\
4
.. License Renewal Issue 98-0085. " Vessel Fluence"
- Priority 1
-issue Clarification This issue addresses the reactor vessel material surveillance program for license renewal.
The title of the issue will be changed to " Reactor Vessel Surveillance Program." This issue has been previously clarified with NEl on a February 5,1999, telephone call.
- License Renewal issue 98-0087. " Containment Temperature Proaram" Priority 1 Issue Clarification
- NEl agreed to submit information by April 2,-1999, to clarify this issue.
License Renewal issue 98-0103. " Reactor Vessel Internals Embrittlement"
,P_rioritti Issue Clarification -
The NRC will proceed to develop guidance on this issue without additional input from NEl.
l L
c