ML20133K764
| ML20133K764 | |
| Person / Time | |
|---|---|
| Issue date: | 12/26/1996 |
| From: | Thadani A NRC (Affiliation Not Assigned) |
| To: | Jordan E Committee To Review Generic Requirements |
| References | |
| TAC-M95280, NUDOCS 9701210336 | |
| Download: ML20133K764 (27) | |
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UNITED STATES f
., j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON D.C. 20555-0001 o
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Deceter 26,1996 MEMORANDUM TO:
Edward L. Jordan, Chairman Committee to Review Generic Requirements h
A ok C. Thadani, Acting Deputy Director FROM:
h0fficeofNuclearReactorRegulation
~
SUBJECT:
REQUEST FOR ENDORSEMENT OF THE PROPOSED GENERIC LETTER ENTITLED " DEGRADATION OF CONTROL R00 DRIVE MECHANISM AND OTHER VESSEL HEAD PENETRATIONS" (TAC NO. M95280)
The Office of Nuclear Reactor Regulation (NRR) requests that the Committee to Review Generic Requirements (CRGR) endorse the subject proposed Generic Letter (GL).
Following endorsement, the GL will be issued and a notice will be published in the federal Register.
The CRGR previously waived review of the proposed GL, and the proposed GL was published in the federal Register (61 FR 40253) on August 1, 1996, for public comment, the comment period was extended on August 22, 1996 (61 FR 43393).
The Staff received comments from seven licensees, two industry organizations, and one Code Committee. The resolution of the comments is provided in.
NRR believes that the GL involves no new or revised regulatory requirements.
The purpose of the GL is to request addressees to describe their program for ensuring the timely inspection of pressurized water reactor (PWR) control rod drive mechanism (CRDM) and other upper vessel head penetrations. The availability of the information derived from these monitoring and inspection programs will aid NRC Staff in assessing whether addressees are in compliance
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with existing rules and regulations.
' is the GL as proposed by the Staff. The NRC is issuing this GL l
to (1) request addressees to describe their program for assuring timel,y w
inspection of PWR CRDM and other vessel head penetrations, and (2) re@ ire I that all addressees provide to the NRC a written response to the requested m information. The Staff considers this GL to be Category 2.
g g is the response to the questions contained in Section IVN of 5 tie CRGR Charter.
The responses to these questions document the justificgion for the required responses regulated by 10 CFR 50.54(f).
=
q Q
J is the compilation of comment letters received by the Staff.
The 3
comments are mostly editorial in nature, with the majority of the comment letters being endorsements of the comments provided by the Nuclear Energy
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Institute (NEI).
210031
Contact:
C. E. Carpenter, NRR 415-2169 7 h k' N
9701210336 961226 PDR ORG NRRA PDR l
4 I
i Edward L. Jordan.
i The PWR Owners Groups are aware of the NRC Staff's concern regarding this issue, and have informed the NRC Staff that they are taking appropriate actions to preclude a safety issue from developing. However, cracking in the vessel head penetrations (VHPs) has occurred and is expected to continue to j
occur as plants age. Therefore, while the NRC Staff has concluded that VHP cracking does not pose a safety concern'in the near term, the NRC Staff considers degradation of the CRDM and other VHPs to be a potential long term safety issu.e that warrants further evaluation. The vessel head provides the vital function of maintaining a reactor pressure boundary. The NRC Staff l
considers cracking of VHPs to be a potential safety concern for the long ters based on the possibility of (1) exceeding the American Society of Mechanical 4
Engineers (ASME) Code margins if the cracks are sufficiently deep and continue j
to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, in order to verify that the i
margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of Federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) continues to be satisfied, and to ensure i
that the sa(ety significance of VHP cracking remains low, the NRC Staff l
believes that an integrated, long-term program, which includes periodic inspections and monitoring, is necessary.
In addition, the NRC Staff finds
)
that the requested information is also needed to determine if the imposition 3
of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is j
required to maintain public health and safety.
The NRC Staff is not establishing a new position in this generic letter.
The Office of the General Counsel (OGC) reviewed this GL and has no legal objections.
Furthermore, OGC has determined that the proposed GL is not a
" Rule" under the provisions of the Small Business Regulatory Enforcement a
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Fairness (SBREF) Act (see 5 U.S.C., Chapter 8) enacted on March 29, 1996.
The GL is sponsored by Gus C. Lainas, Acting Director for Engineering.
Attachments:
1.
Proposed Generic Letter, titled " Degradation of Control Rod Drive Mechanism and Other Vessel Head Penetrations" 2.
Response to CRGR Charter Questions 3.
Comments received by the Staff 4.
Resolution of the Comments Distribution: see next page
4
/
Edward L. Jordan The PWR Owners Groups are aware of the NRC Staff's concern regarding this issue, and have informed the NRC Staff that they are taking appropriate However, cracking in the actions to preclude a safety issue from developing.
vessel head penetrations (VHPs) has occurred and is expected to continue to Therefore, while the NRC Staff has concluded that VHP occur as plants age.
cracking does not pose a safety concern in the near term, the NRC Staff considers degradation of the CRDM and other VHPs to be a potential long term safety issue that warrants further evaluation. The vessel head provides the vital function of maintaining a reactor pressure boundary. The NRC Staff considers cracking of VHPs to be a potential safety concern for,the long term based on the possibility of (1) exceeding the American Society of Mechanical Engineers (ASME) Code margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, in order to verify that the margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of Federal Regulatfons (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) continues to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the NRC Staff believes that an integrated, long-term program, which includes periodic i
inspections and monitoring, is necessary.
In addition, the NRC Staff finds that the requested information is also needed to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety.
The NRC Staff is not establishing a new position in this generic letter.
The Office of the General Counsel (0GC) reviewed this GL and has no legal objections. Furthermore, OGC has determined that the proposed GL is not a
" Rule" under the provisions of the Small Business Regulatory Enforcement Fairness (SBREF) Act (see 5 U.S.C., Chapter 8) enacted on March 29, 1996.
The GL is sponsored by Gus C. Lainas, Acting Director for Engineering.
Attachments:
1.
Proposed Generic Letter, titled " Degradation of Control Rod Drive Mechanism and Other Vessel Head Penetrations" 2.
Response to CRGR Charter Questions 3.
Comments received by the Staff 4.
Resolution of the Comments Distribution: see next page DOCUMENT NAME: G:\\CARPENTR\\CRDHCRGR.MEM
- See Previous Concurrence To receive a copy of this documerst, Iraficate in the bes:
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o UNITED STATES NUCLEAR REGULATORY COMMISSION Off!CE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 pscessE18X199.6 GENERICLETTER96!!#:
DEGRADATION OF CONTROL ROD DRIVE MECHANISM AND OTHER VESSEL CLOSURE HEAD PENETRATIONS (TAC NO. M95280).
Addressees All holders of operating licenses for pressurized water reactors (PWRs),
l except those licenses that have been amended to possession-only status.
Puroose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) request addressees to describe their program for ensuring the timely inspection of PWR control rod drive mechanism (CRDM) and other vessel closure head penetrations and (2) require that all addressees provide to the NRC a written response to the requested information.
Backaround Primary Water Stress Corrosion Crackina of Vessel Clo'sure Head Penetrations Most PWRs have Alloy 600 CRDM nozzle and other vessel head closure penetrations (VHPs) that extend above the reactor pressure vessel head. The stainless steel housing of the CRDM is screwed and seal-welded onto the top of the nozzle penetration, as shown in Figure 1.
(Figur'ellfis.-forfillustrative
"^Uiy6sei?ohlyjahdilidiot? Intendeditoibe findicative. of.every<' metal " weld, whichven p
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Thiseld'bitween the"noisle' and the housing is a dissim'ilar is also called a bimetallic weld. The nozzles protrude below the vessel head, thus exposing the inside surface of the nozzles to reactor coolant. The control rod drive (CRD) nozzles and other VHPs are basically the same for all PWRs worldwide, which use a U.S. design (except in Germany and Russia).
Generally, there are 36 to 78 nozzles distributed over the low-alloy steel head. The vessel head is semi-spherical and the head penetrations are vertical so that the CRD nozzles and other VHPs are not perpendicular to the j
vessel surface except at the center. The uphill side (toward the center of the head) is called the 180-degree location and the downhill side (toward the outer periphery of the head) is called the 0-degree location. Most nozzles have a thermal sleeve with a conical guide at the bottom end and a small gap (3-to 4-mm) [0.12 to 0.16 in.] between the nozzle and the sleeve.
[siksifWivTiiEEsFFid si1ReiU 986IiiiiiEEfil5 alt 6)T6005pii'siifiiiFIt' nit'F6ijiE6i i
hteamisupp1Lsys tem! vend 6rsj.7Th~s~ gs re act ors ; from iseve ral i d i f fe re hozzleMatibothidomesticiandJf6rei NR'C~itaff'idshtified jiri$aff wateFstress~
corrosion cracking (PWSCC) as an emerging technical issue to the Commission in 1989, after cracking was noted in Alloy 600 pressurizer heater sleeve Attachment I l
i 1
GL~ 96-##..
December 18,1996 Page:2Loffl0 penetrations at a domestic PWR facility. The NRC staff reviewed the safety significance of the cracking that occurred, as well as the repair and replacement activities at the affected facilities. The NRC staff determined that the cracking was not of immediate safety significance because the cracks were axial, had a low growth rate, were in a material with an extremely high flaw tolerance (high fracture toughness), and, accordingly, were unlikely to propagate very far. These factors also demonstrated that any cracking would result in detectable leakage and the o)portunity to take corrective action before a penetration would fail. Furtier,:the~ NRC staff'islnot^ aware ofTsrip
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Ta11ure:sff an1 Alloy 1600ivesselfclosure headipenetration duringsplant
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bperationf The'NRC~ stiff issued ~Information Notice 90-10, " Primary Water
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Stress Corrosion Cracking (PWSCC) of Inconel 600," dated February 23, 1990, to inform the nuclear industry of the issue.
In Se~ptsbiF1991, cracks were found in an Alloy. 600 VHP in the reactor head at B~gcy 3; i French PWR. Examf hatis'nsliniPWRsiin France,7 Belgium,lSwedeni u
SsitzerlaridRSpiihWasdfJipan were perfonned,' and additionalt.VHPs;withfaxial bracksiwe.re"det.ectedl. inns.everalEEuro ean plants. About 2 percent'of'the VHPs
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~Close examination of the VHP that leaked at Bugey 3 revealed very minor incipient secondary circumferential cracking of the VHP.
An action plan was implemented by the NRC staff in 1991 to address PWSCC of Alloy 600 VHPs at all U.S. PWRs. As explained more fully below, this action plan included a review of the safety assessments by the PWR Owners Groups, the development of VHP mock-ups by the Electric Power Research Institute (EPRI),
the qualification of inspectors on the VHP mock-ups by EPRI, the review of proposed generic acceptance criteria from the Nuclear Utility Management and Resource Council (NUMARC) (now the Nuclear Energy Institute (NEI)], and VHP inspections. As part of this action plan, the NRC staff met with the Westinghouse Owners Group (WOG) on January 7, 1992, the Combustion Engineering Owners Group (CE0G) on March 25, 1992, and the Babcock & Wilcox Owners Group (B&W0G) on May 12, 1992, to discuss their respective programs for investigating PWSCC of Alloy 600 and to assess the possibility of cracking of VHPs in their respective plants since all of the plants have Alloy 600 VHPs.
Subsequently, the NRC staff asked NUMARC to coordinate future industry actions because the issue was applicable to all PWRs. Meetings were held with NUMARC/
NEI and the PWR Owner's Groups on the issue on August 18 and November 20, 1992, March 3, 1993, December 1, 1994, and August 24, 1995. Summaries of these meetings are available in the Commission's Public Document Room, 2120 L Street, N.W., Washington, D.C. 20555.
Each of the PWR Owners Groups submitted safety assessments, dated February 1993, through NUMARC to the NRC on this issue. After reviewing the industry's safety assessments and examining the overseas inspection findings, the NRC staff concluded in a safety evaluation dated November 19, 1993, that VHP cracking was not an immediate safety concern. The bases for this conclusion were that if PWSCC occurred at VHPs (1) the cracks would be predominately axial in orientation, (2) the cracks would result in detectable leakage before catastrophic failure, and (3) the leakage would be detected during visual examinations performed as part of surveillance walkdown inspections before
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GL 96-##
December 18, 1996 Page 3 of 10 significant damage to the reactor vessel clos'ure head would occur.
In addition, the NRC staff had concerns related to unnecessary occupational radiation exposures associated with eddy current or other forms of non-destructive examinations (NDEs), if performed manually.
Field experience in foreign countries has shown that occupational radiation exposures can be significantly reduced by using remotely. controlled or automatic equipment to conduct the inspections.
In 1993, the nuclear industry developed renotely operated inservice inspection equipment and repair tools that reduced radiation exposure. Techniques and procedures developed by two vendors were successfully demonstrated in a blind qualification protocol developed and administered by the EPRI NDE Center.
In the demonstrations, examinations by rotating and saber eddy current and ultrasonics showed a high probability of detection of the flaws which were also sized within reasonable uncertainty bounds. The qualification testing also demonstrated that personnel qualified through the EPRI program can reliably detect PWSCC in CRDM nozzles.
Intearated Attack of CRDM at Zorita In 1994, circumferential intergranular attack (IGA) associated with the J-groove weld in one of the CRDM penetrations was discovered at Zorita, a Spanish reactor. This IGA is a different degradation mechanism than the PWSCC described above..It is believed to have resulted from the combination of ion exchange resin bed intrusions, which resulted in high concentrations of sulfates. Zoritahas 37 CRDM penetrations, of which 20 are active penetrations and 17 are spare penetrations. Sixteen of the 17 spare penetrations showed stress corrosion cracking and IGA. The cracks were both axial ar.d circumferential.
Four of the active CRDM penetrations had significant cracking with axial and circumferential cracks. Two cation resin ingress events occurred at Zorita, in August 1980, 40 liters [10:5710.S; sallons) of cation resin entered the reactor coolant system (RCS).
In Siptimber 1981,, a mixe.d bed demineralizer screen failed and between 200 to 320 liters [52;B3?to 84s54 U.Sr gallons] of resin entered the RCS. The coolant conduct ivity7kninined high"for at 1 east 4 months after the ingress. The
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increase in conductivity was attributed to locally high concentrations of sulf ates. Sulfates were found around the crack areas and on the fracture surfaces.
It is important to note that sulfate cracking can occur in regions that are not subject to significant applied or residual stresses.
The NRC staff issued Information Notice (IN) 96-11, " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 1996, to alert addressees to the increased likelihood of sulfate-driven stress corrosion cracking of PWR CRDMs and other VHPs if demineralizer resins contaminate the RCS.
The Westinghouse staff notified the WOG plants, the B&WOG plants, and the CEOG plants of the Zorita incident by issuing NSAL-94-028. Westinghouse reported that no other plant had been found worldwide that had experienced cracking similar to that at the Zorita plant. The Westinghouse staff further reported that U.S. plants monitor RCS conductivity on a routine basis, follow the EPRI
e GK96;H i
DecembeB1821996 Page14;of 10 guidelines on primary water chemistry, and monitor for sulfate three times a week. The Westinghouse staff concluded that no immediate safety issue is involved and that the conclusions in its CRDM safety evaluation remain valid.
The Westinghouse staff suggested that U.S. PWR plants review their RCS chemistry and other operating records pertaining to sulfur ingress events.
The results of this review have not been. reported to the NRC staff, and the NRC staff does not have sufficient information to ascertain whether'any significant primary system resin bed intrusions have occurred at any U.S. PWR.
The first U.S. inspection of VHPs took place in the spring of 1994 at the Point Beach Nuclear Generating Station, and no indications were detected in any of its 49 CRDM penetrations. The eddy current inspection at the Oconee i
Nuclear Generating Station in the fall of 1994 revealed 20 indications in one penetration. Ultrasonic testing (UT) did not reveal the depth of these indications because they were shallow. UT cannot accurately size defects that are less than one mil deep (0.03 mm). These indications may be associated 3
with the original fabrication and may not grow; however, they will be 4
reexamined during the next refueling outage. A limited examination of eight i
in-core instrumentation penetrations conducted at the Palisades plant found no
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cracking. An examination of the CRDM penetrations at the D. C. Cook plant in i
the fall of 1994 revealed three clustered indications in one penetration.
0
~ The indications were 46 mm [liBITiri?),16 mm [sa.6371n)), and 6 to 8 mm [0.24't6 and~the~ deepest flaw 0'.'31Eih7] in length,$] flaw was just below the J-groove weld.] deep.
~~ 6.'8 mm '[0;27Ein.*
The~tip s
bf"thi'46-mm{L81$i i
Virginia Electric and Power Company inspected North Anna Unit I during its spring 1996 refueling outage. Some high-stress areas (e.g., upper and lower hillsides) were examined on each outer ring CRDM penetrations and no indications were observed using eddy current testing.
i The NRC staff was informed during a meeting on August 24, 1995, that Westinghouse had developed a susceptibility model for VHPs based on a number of factors, including operating temperature, years of power operation, method of fabrication of the VHP, microstructure of the VHP, and the location of the VHP on the head. Each time a plant's VHPs are inspected, the inspection results are incorporated into the model. All domestic Westinghouse PWRs have been modeled and the ranking has been given to each licensee.
In addition, the NRC staff was informed that Framatome Technologies, Inc. [FTI, formerly Babcock & Wilcox (B&W)], also developed a susceptibility model for CRDM penetration nozzles and other VHPs in B&W reactor vessel designs. All domestic B&W PWRs have been modeled and the ranking has been given to each B&W licensee. The NRC staff was further informed that Combustion Engineering (CE) had performed an initial susceptibility assessment for the CE PWRs. At prssen@h6n~efofithe7PWR?0wners4Groupst(i.e.,: WOGW B&WOGMoriCEOG)t as
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h pbmit}e[itsimodelsynd?assessmentsttotthefNRC;stafffforkeview; By letter dated March 5,1996, NEl submitted a white paper entitled " Alloy 600 RPV Head Penetration Primary Stress Corrosion Cracking," which reviews the significance of PWSCC in PWR VHPs and describes how the industry is managing the issue. The program outlined in the NEI white paper is based on the assumption that the issue is an economic one rather than a safety issue, and e
GL 96-#
December 18,.1996 Page 5'of.10 describes an economic decision tool to be used by PWR licensees to evaluate the probability of a VHP developing a crack or a through-wall leak during a plant's lifetime. This information would then be used by a PWR licensee to evaluate the need to conduct a VHP inspection at their plant. The NRC staff informed NEI in the several meetings listed above that it did not agree with NEI that the issue was only economic..I.nspections have shown that cracking has initiated in some U.S. plants, and the industry has not provided sufficient technical justification regarding susceptibility of the CRDM and other VHPs to PWSCC to justify an inspection plan based on economic considerations alone.
Discussion The results of domestic VHP inspections are consistent with the February 1993 analyses by the PWR Owners Groups, the NRC staff safety evaluation report dated November 19, 1993, and the PWSCC found in the CRDMs in European reactors. On the basis of the results of the first five inspections of U.S.
PWRs, the PWR Owner's Groups' analyses, and the European experience, the NRC staff has determined that there is a high probability that VHPs at other plants may contain similar axial cracks caused by PWSCC.
Further, if any significant resin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses are sufficient to cause circumferential intergranular stress corrosion cracking (IGSCC).
After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern.
Further, the NRC staff recognizes that the scope and timing of inspections may vary for different plants depending on their individual susceptibility to this form of degradation.
In the long term, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation.
The vessel ' closure head provides the vital function of maintaining a reactor pressureboundari. Cracking in the VHPs has occurred and is expected to continue to occur as plants age. The NRC staff considers cracking of VHPs to be a safety concern for the long term based on the possibility of (1) exceeding the American Society of Mechanical Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, in order to verify that the margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of Federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) is continued to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the NRC staff Eontihues?to believe that an integrated, long-term program,s wash includes perio'dic' whic inspections and monitoring Thi the" conclusiontofith's staff'iEN5Vsmbe P1931993 M, is necessary.afety evaluation',"which stated?in part W.ths t
staffJrecommends}thatiyouiconsider enhanc~ed leakage'detectionLby-visually" hxaminihgitheTreact'or3v'es'sePhead until<eithereinspections have been complieted khowingrabsencefoficrackingborf on;linel leakage detectionf s: installed:intthe~
i hsad?areaw. nondestructiveiexaminations1should1be performed to1ensurerthere 1sjfhnexpectedicrackingjinLdomestic[PWRs([These~examinationsidolnotfhave?to i
GL 961##
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December l18; 1996 Page,6,of,10 tisW6Hd6Etid'iiinedlitelys.".7As'thsisurveillance walkdowns'iiroposed[by1UHARC
~
hreenottintendedifor! detecting small;le'aks~, it is conceivableithat 'some sffected!PWRsEcouldjotentially' operate with,small undetected'leakagecat CRDM/CEON penetrationsE lIn?thiscregird,cthe staff believes 1that?itiis. prudent
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for?NUMARCitoicorisiderL the1 implementation ofJan enhanced leakage ' detection inethodsforTdetecting'?sinall leaks'during' plant operation'" In addition, the NRC stiff 71iidi~ thit"the Fiqu$ted~information' is' also needed to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety.
The NRC staff recognizes that individual PWR licensees may wish to determine their inspection activities based on an integrated industry inspection program (i.e., B&WOG, CE0G, WOG, or some subset thereof), to take advantage of 4
inspection results from other plants that have similar susceptibilities. The 3
NRC staff does not wish to discourage such group actions but notes that such an integrated industry inspection program must have a well-founded technical basis that justifies the relationship between the plants and the planned imp'ementation schedule.
RiBUFilid Information The information Fs5ikitid in items 1 and 2, below, is needed by the NRC staff isWehiff*t5mpliance;withT107CFR 50.55a and"10"CFR Part:50,i AppendisA',
GDC(14Dand~to~deterinine~if'the~ imp}osition of 'an augmented inspection program,-
retjiiested Tn?107CFRiS0iS5i(g)(6)(li) purhantito is required, while the information
'itsm"37blates to the potential for domestic resin intrusions, such as occurred at Zorita.
Withihil207dijsTfrbini he date ofithip: generic. letter, addressees are requested t
t6W6~vids~the foll6idincj information 1.
Regarding inspection activities:
1.1 A description of all inspections of CRDMs and other vessel clostire head penetrations performed to the date of this generic letter, j
including the results of these inspections.
1.2 If you have developed a plan to periodically inspect the CRDM and other vessel closure head penetrations:
i a.
Your schedule for first, and subsequent, inspections of the CRDM and other vessel closure head penetrations, including the t
technical basis for your schedule.
i b.
Your scope for the CRDM and other vessel closdre head penetration inspections, including the total number of penetrations (and how 1
i Th65elli&nsee^s^ith^at hive: previously" submitted theErequ'ested information needbriottresubmitt itT but:.should instead reference;' the - appropriate correspondenceDin'their response to this Generic Letter.
l GL 96-##
December;18,.1996 Page 7 of 10 s
many will be inspected), which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
1.3 If you have not developed a plan to periodically inspect the CRDM and other vessel closuFe head penetrations, provide your technical or i
safety basis for"not periodically inspecting your VHPs; or, your-schedule for developing such a plan and the basis for that schedule.
2.
A description of the evaluation methods and results used to assess the susceptibility of the CRDM and other VHPs in your plant to PWSCC, including the susceptibility ranking of your plant and the factors used to determine this ranking. Other than or in addition to the boric acid visual examination (see Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," dated March 17, 1988), include a description of all relevant data and/or. tests used to develop crack initiation and crack growth models, et the methods and data used to validate these models
!ncise o 3teic.xntcglW Sg i
the 'pplic'hi14+y d there unie te t,y g""P creams mue.
Also, if you
.c arerelyingonanyintegratedindustrydspectionprogram,providea detailed description of this program.
ig4 M ( yy fSh.L qM y
3.
A description of any resin intrusions in your plant, as described in N
IN 96-11, that have exceeded the current EPRI PWR' Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
3.1 Were the intrusions cation, anion, or mixed bed?
3.2 What were the durations of these intrusions?
3.3 Do your RCS water chemistry Technical Specifications follow the EPRI guidelines?
3.4 Identify any RCS chemistry excursions that exceed your plant administrative limits for the following species:
sulfates, chlorides or fluorides, oxygen, boron, and lithium, l
l 3.5 Identify any conductivity excursions which may be indicative of resin intrusions, provide your technical assessment of each excursion and your followup actions.
3.6 Provide your assessment of the potential for any of these intrusions to result in a significant increase in the probability for IGA of VHPs and any associated plan for inspections.
Reouired Response All addressees shall submit in writing the information identified above within
}20 days from the date of this letter.
i
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1 Gl. 9_6-##
December 18, 1996 Page' 8 of_10 Any inspection results that do nol satisfy the acceptance criteria identified o
in the NRC staff's safety assessment dated November 16, 1993, should be reported to the NRC staff prior to plant restart.
1 Address the required written reports to the U.S. Nuclear Regulatory l
Commission, ATTN: Document Control Desk, Washington, D.C. 20555, under oath or affirmation under the provisions of S'ection 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).
In addition, submit a copy to the appropriate regional administrator.
The NRC recognizes the potential difficulties (number and types of sources, age of records, proprietary data, etc.) that licensees may encounter while ascertaining whether they have all of the data pertinent to the evaluation of their CRDMs and other vessel closure head penetrations.
For this reason, the above time periods are allowed ~for the responses.
Related Generic Communications (1)
Information Notice 90-10, " Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600," dated February 23, 1990.
(2)
NUREG/CR-6245, " Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking," dated October 1994.
1 (3)
Information Notice 96-11. " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 1996.
Backfit Discussion This generic letter only requires information from the addressees under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). Therefore, the staff has not performed a backfit analysis.
The information collected will enable the staff to verify that the margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) continues to be satisfied, and to ensure that the safety significance of VHP cracking remains lowkJSince the NRC staff requires licensees to submit information to assess compliance with the above stated re qu i reme nt s7nii!j dsti fibitibn) f6 rlthi s Tre qu e s ted l i n f6rmst ion ineed ino t t be p}epiited"ihfosati6n~is'Also~neided to determine if the imposition of an WedliF accordance withsSection?50;54(f). The NRC st'aff findsthat the
^
requ augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety. The staff is not establishing a new
{
position for such compliance in this generic letter. Therefore, this generic letter does not constitute a backfit and no documented evaluation or backfit analysis need be prepared.
)
J
I GL 96-##
December ~18;91996 Page 9 ofs10 Federal Reaister Notification A notice of opportunity for public comment was published in the federal Register (61 FR 40253) on August 1, 1996, and extended on August 22, 1996 (61 FR 43393). Comments were received from seven licensees, two industry organizations, and one Code Committee. Copies of the staff evaluation of-these comments have been made available in the public document room.
Paoerwork Reduction Act Statement This generic letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information 4
i collections were approved by the Office of Management and Budget, approval i
number 3150-0011, which expires July 31, 1997.
The public reporting burden for this collection of information is estimated to average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the collection of information contained in the generic letter and on the following issues:
1.
Is the proposed collection of information necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?
2.
Is the estimate of burden accurate?
3.
Is there a way to enhance the quality, utility, and clarity of the 4
information to be collected?
4.
How can the burden of the collection of information be minimized,
~
including the use of automated collection techniques?
Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch, T-6 F33, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs, NE0B-10202 (3150-0011), Office of Management and Budget, Washington, DC 20503.
The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
t
GLT96 ##
December $187T1996 Pagel10;of;10' If you have any questions about this matter, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Thomas T. Martin, Director -
Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts:
Keith R. Wichman (301) 415-2757 e-mail: krwenrc. gov 1
James Medoff (301) 415-2715 i
e-mail: jxm@nrc. gov Lead Project Manager:
C. E. Carpenter, Jr.
(301) 415-2169 e-mail: cec @nrc. gov Attachments:
1.
References 2.
List of Recently Issued NRC Generic Letters I
i l
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Figure 1. Head Penetration and Vessel Assembly
CRGR REVIEW PACKAGE PROPOSED ACTION:
Issue a generic letter on the primary water stress corrosion cracking of control rod drive mechanism and other vessel head penetrations.
CATEGORY:
2 RESPONSE TO RE0VIREMENTS FOR CONTENT'0F PACKAGE SUBMITTED FOR CRGR REVIEW (1)
The proposed generic requirement or staff position as it ls proposed to be sent out to licensees. Where the objective or intended result of a proposed generic requirement or staff position can be achieved by setting a readily quantifiable standard that has an unambiguous relationship to a readily measurable quantity and is enforceable, the proposed requirement should merely specify the objective or result to be attained, rather than prescribing to the licensee how the objective or result is to be attained.
The information requested in items 1 and 2, below, is needed by the NRC staff to verify compliance with 10 CFR 50.55a and 10 CFR Part 50, Appenqix A, GDC 14, and to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required, while the information requested in item 3 relates to the potential for domestic resin intrusions, such as occurred at Zorita.
Within 120 days from the date of this generic letter, addressees are requested to provide the following information':
1.
Regarding inspection activities:
~
1.1 A description of all inspections of CRDMs and other vessel closure head penetrations performed to the date of this generic letter, including the results of these inspections.
1.2 If you have developed a plan to periodically inspect the CRDM and other vessel closure head penetrations:
a.
Your schedule for first, and subsequent, inspections of the CRDM and other vessel closure head penetrations, including the technical basis for your schedule, b.
Your scope for the CRDM and other vessel closure head penetration inspections, including the total number of penetrations (and how many will be inspected), which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
Those licensees that have previously submitted the requested information need not resubmit it but should instead reference the appropriate correspondence in their response to this Generic Letter.
~
CRGR REVIEW PACKAGE !
1.3 If you have D21 developed a plan to periodically inspect the CRDM and other vessel closure head penetrations, provide your technical or safety basis for not periodically inspecting your l
VHPs; or, your schedule for developing such a plan and the basis for that schedule.
2.
A description of the evaluation methods and results used to assess the susceptibility of the CRDM and other VHPs in your plant to PWSCC, including the susceptibility ranking of your phant and the factors used to determine this ranking. Other than or in addition to the boric acid visual examination (see Generic Letter 88-05,
" Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," dated March 17,1988), include a description of all relevant data and/or tests used to develop crack initiation and crack growth models, and the methods and data used to validate these models.
Include a statement explaining the applicability of these models to the VHP cracking issue. Also, if you are relying on any integrated industry inspection program, provide a detailed description of this program.
3.
A. description of any resin intrusions in your plant, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
3.1 Were the intrusions cation, anion, or mixed bed?
3.2 What were the durations of these intrusions?
3.3 Do your RCS water chemistry Technical Specifications follow the EPRI guidelines?
3.4 Identify any RCS chemistry excursions that exceed your plant administrative limits for the following species: sulfates, chlorides or fluorides, oxygen, boron, and lithium.
3.5 Identify any conductivity excursions which may be indicative of resin intrusions, provide your technical assessment of each excursion and your followup actions.
3.6 Provide your assessment of the potential for any of these intrusions to result in a significant increase in the probability for IGA of VHPs and any associated plan for inspections.
(ii) Draft staff papers or other underlying staff documents supporting the requirements or staff positions.
(A copy of all materials referenced in the document shall be made available upon request to the CRGR staff.
Any Committee member may request CRGR staff to obtain a copy of any reference material for his or her use.)
(1)
Information Notice 90-10, " Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600," dated February 23, 1990.
I CRGR REVIEW PACKAGE (2)
NRC staff safety evaluation, " Potential Reactor Vessel Head l
Adaptor Tube Cracking," dated November 19, 1993 (3)
NUREG/CR-6245, " Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking," dated October 1994.
(4)
Information Notice 96-11, "Jngress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 1996.
(iii) Each proposed requirement or staff position shall contain the sponsoring office's position as to whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, cr would relax or reduce existing requirements or staff positions.
The results of domestic VHP inspections are consistent with the February 1993 analyses by the PWR Owners Groups, the NRC staff safety evaluation report dated November 19, 1993, and the PWSCC found in the CRDMs in European reactors. On the basis of the results of the first five inspections of U.S. PWRs, the PWR Owner's Groups' analyses, and the European experience, the NRC staff has determined that there is a high probability that VHPs at other plants may contain similar axial cracks caused by PWSCC. Further, if any significant r.esin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses are sufficient to cause circumferential intergranular stress corrosion cracking (IGSCC).
After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern.
Further, the NRC staff recognizes that the scope and timing of inspections may vary for different plants depending on their individual susceptibility to this form of degradation.
In the long term, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation. The vessel closure head provides the vital function of maintaining a reactor pressure boundary.
Cracking in the VHPs has occurred and is expected to continue to occur as plants age. The NRC staff considers cracking of VHPs to be a safety concern for the long term based on the possibility of (1) exceeding the American Society of Mechanical Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, in order to verify that the margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of Federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) is continued to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the NRC staff continues to believe that an integrated, long-term program, which includes periodic inspections and monitoring, is necessary. This l
was the conclusion of the staff's November 19, 1993, safety evaluation, l
which stated, in part, "...the staff recommends that you consider j
enhanced leakage detection by visually examining the reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area...
l
.CRGR REVIEW PACKAGE nondestructive examinations should be performed to ensure there is no unexpected cracking in domestic PWRs. These examinations do not have to be conducted immediately... As the surveillance walkdowns proposed by NUMARC are not intended for detecting small leaks, it is conceivable that some affected PWRs could potentially operate with small undetected leakage at CRDM/CEDM penetrations.
In this regard, the staff believes that it is prudent for NUMARC to consider the implementation of an enhanced leakage detection method for detecting small leaks during plant operation." In addition, the NRC staff finds that the requested information is also needed to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety. Therefore, this request does not increase nor reduce existing requirements.
It is a request to obtain information to confirm compliance with existing requirements.
The NRC staff recognizes that individual PWR licensees may wish to determine their inspection activities based on an integrated industry inspection program (i.e., B&WOG, CE0G, WOG, or some subset thereof), to take advantage of inspection results from other plants that have similar susceptibilities. The NRC staff does not wish to discourage such group actions but notes that such an integrated industry inspection program must have a well-founded technical basis that justifies the relationship between the plants and the planned implementation schedule.
(iv) The proposed method of implementation with the concurrence (and any coments) of OGC on the method proposed. The concurrence of affected program offices or an explanation of any nonconcurrences.
See attached concurrence page.
(v)
Regulatory analyses conforming to the directives and guidance of NUREG/BR-0058 and NUREG/CR-3568.
(This does not apply for backfits that ensure compliance or ensure, define, or redefine adequate protection.
In these cases a documented evaluation is required as discussed in IV.B.(ix).)
Not applicable (vi) Identification of the category of reactor plants to which the generic requirement or staff position is to apply (that is, whether it is to apply to new plants only, new OLs only, OLs after a certain date, OLs before a certain date, all OLs, all plants under construction, all plants, all water reactors, all PWRs only, some vendor types, some vintage types such as BWR 6 and 4, jet pump and nonjet pump plants, etc.).
All holders of operating licenses for pressurized water reactors (PWRs),
except those licenses that have been amended to possession-only status.
(vii) For backfits other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.109. The backfit analysis shall include, for each category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light
.1
]
.CRGR REVIEW PACKAGE of other ongoing regulatory activities. The backfit analysis shall document for consideration information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed action:
(a)
Statement of the specific objectives that the proposed action is designed to achieve; Not applicable.
(b)
General description of the activity that would be required by the licensee or applicant in order to complete the action; Not applicable.
(c)
Potential change in the risk to the public from the accidental release of radioactive material; Not applicable.
(d). Potential impact on radiological exposure of facility employees and other onsite workers; Not applicable.
(e)
Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction delay; Not applicable.
(f)
The potential safety impact of changes in plant or operational complexity, including the relationship of proposed and existing regulatory requirements and staff positions; Not applicable.
(g)
The estimated resource burden on the NRC associated with the proposed action and the availability of resources; Not applicable.
(h)
The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed action; Not applicable.
(i)
Whether the proposed action is interim or final, nd if interim, the justification for imposing the proposed action on an interim basis; Not applicable.
9 l
CRGR REVIEW PACKAGE 4 l
l (j)
How the action should be prioritized and scheduled in light of I
other ongoing regulatory activities. The following information I
may be appropriate in this regard:
1.
The proposed priority or schedule, 2.
A summary of the current backlog of existing requirements awaiting implementation, 3.
An assessment of whether implementation of existing requirements should be deferred as a result, and 4.
Any other information that may be considered appropriate with regard to priority, schedule, or cumulative impact.
For example, could implementation be delayed pending public comment?
Not applicable.
(viii)
For each backfit analyzed pursuant to 10 CFR 50.109(a)(2) (i.e.,
not adequate protection backfits and not compliance backfits), the proposing Office Director's determination, together with the
. rational for the determination based on the consideration of paragraph (i) and (vii) above, that:
(a)
There is a substantial increase in the overall protection of public health and safety or the common defense and security to be derived from the proposal; and (b)
The direct and indirect costs of implementation, for the facilities affected, are justified in view of this increased protection.
Not applicable.
(ix) For adequate protection or compliance backfits evaluated pursuant to 10 CFR 50.109(a)(4)
(a) a documented evaluation consisting of:
(1) the objectives of the modification (2) the reasons for the modification (3) the basis for invoking the compliance or adequate protection exemption.
(b) in addition, for actions that were immediately effective (and therefore issued without prior CRGR review as discussed in III.C) the evaluation shall document the safety significance and appropriateness of the action taken and (if applicable) consideration of how costs contributed to selecting the solution among various acceptable alternatives.
t Not applicable. The proposed generic letter is a request for information only. The NRC staff is not requesting any new actions from I
the PWR licensees; rather, the proposed generic letter is requesting the PWR licensees to provide to the NRC information that the PWROGs has l
l i
.CRGR REVIEW PACKAGE already told the NRC staff it has gathered, but has not shared with the NRC to date.
1 (x)
For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing Office Director's determination, together with the rationale for the determination based on the considerations or paragraphs (i) through (vii) above, that:
(a)
Public health and safety and the common defense and security would be adequately protected if the proposed reduction in requirements or positions were implemented, and (b)
The cost savings attributed to the action would be substantial enough to justify taking the action.
Not applicable.
(xi) For each request for information under 10 CFR 50.54(f) (which is not subject to exception as discussed in III. A) an evaluation that includes at least the following elements:
(a)
A problem statement that describes the nsed for the information in terms of potential safety benefit.
The NRC staff was informed during a meeting on August 24, 1995, that Westinghouse had developed a susceptibility model for VHPs based on a number of factors, including operating temperature, years of power operation, method of fabrication of the VHP, microstructure of the VHP,.
and the location of the VHP on the head.
Each time a plant's VHPs are inspected, the inspection results are incorporated into the model. All domestic Westinghouse PWRs have been modeled and the ranking has been given to each licensee.
In addition, the NRC staff was informed that Framatome Technologies, Inc. [FTI, formerly Babcock & Wilcox (B&W)],
also developed a susceptibility model for CRDM penetration nozzles and other VHPs in B&W reactor vessel designs. All domestic B&W PWRs have been modeled and the ranking has been given to each B&W licensee. The NRC staff was further informed that Combustion Engineering (CE) had performed an initial susceptibility assessment for the CE PWRs. At present, none of the PWR Owners Groups (i.e., WOG, B&WOG, or CEOG) has submitted its models and assessments to the NRC staff for review.
The results of domestic VHP inspections are consistent with the February 1993 analyses by the PWR Owners Groups, the NRC staff safety evaluation report dated November 19, 1993, and the PWSCC found in the CRDMs in European reactors. On the basis of the results of the first five inspections of U.S. PWRs, the PWR Owner's Groups' analyses, and the European experience, the NRC staff has determined that there is a high probability that VHPs at other plants may contain similar axial cracks caused by PWSCC.
Further, if any significant resin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses are sufficient to cause circumferential intergranular stress corrosion cracking (IGSCC).
.m
,CRGR REVIEW PACKAGE After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern.
Further, the NRC staff recognizes that the scope and timing of inspections may vary for different plants depending on their individual susceptibility to this form of degradation.
In the long term, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation. The vessel closure. head i
provides the vital function of mai'ntaining a reactor pressure boundary.
Cracking in the VHPs has occurred and is expected to continue to occur as plants age. The NRC staff considers cracking of VHPs to be a safety concern for the long term based on the possibility of (1) exceeding the American Society of Mechanical Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, in order to verify that the margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) is continued to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the NRC staff continues to believe that an integrated, long-term program, which includes periodic inspections and monitoring, is necessary. This was the conclusion of the staff's November 19, 1993, safety evaluation, l
which stated, in part, "...the staff recommends that you consider enhanced leakage detection by visually examining the reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area...
nondestructive examinations should be performed to ensure there is no unexpected cracking in domestic PWRs. These examinations do not have to be conducted immediately... As the surveillance walkdowns proposed by.
NUMARC are not intended for detecting small leaks, it is conceivable i
that some affected PWRs could potentially operate with small undetected leakage at CRDM/CEDM penetrations.
In this regard, the staff believes that it is prudent for NUMARC to consider the implementation of an enhanced leakage detection method for detecting small leaks during plant operation." In addition, the NRC staff finds that the requ sted information is also needed to determine if the impo.tition i ' an augmented inspection program, pursuant to 10 CFR 50.55a(g)h ;(ii), is required to maintain public health and safety. Therefore, ;his request does not increase nor reduce existing requirements.
It is a request to obtain information to confirm compliance with existing requirements.
The NRC staff recognizes that individual PWR licensees may wish to 3
determine their inspection activities based on an integrated industry inspection program (i.e., B&WOG, CE0G, WOG, or some subset thereof), to take advantage of inspection results from other plants that have similar susceptibilities. The NRC staff does not wish to discourage such group actions but notes tnat such an integrated industry inspection program must have a well-founded technical basis that justifies the relationship l
between the plants and the planned implementation schedule.
I i
(b)
The licensee actions required and the cost to develop a response to the information request.
l
.CRGR REVIEW PACKAGE i The information requested in items 1 and 2, below, is needed by the NRC staff to verify compliance with 10 CFR 50.55a and 10 CFR Part 50, Appendix A, GDC 14, and to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required, while the information requested in item 3 relates to the potential for i
domestic resin intrusions, such as occurred at Zorita.
4 Within 120 days from the date of this generic letter, addressees are requested to provide the following information :
r
]
1.
Regarding inspection activities:
1.1 A description of all inspections of CRDMs and other vessel closure head penetrations performed to the date of this generic letter, including the results of these inspections.
1.2 If you have developed a plan to periodically inspect the CRDM and other vessel closure head penetrations:
a.
Your schedule for first, and subsequent, inspections of the CRDM and other vessel closure head penetrations, including l
the technical basis for your schedule, i
b.
Your scope for the CRDM and other ' vessel closure head j
penetration inspections, including the total number of penetrations (and how many will be inspected), which penetrations have thermal sleeves, which are spares, and i
which are instrument or other penetrations.
1.3 If you have no.1 developed a plan to periodically inspect the CRDM and other vessel closure head penetrations, provide your 4
technical or safety basis for not periodically inspecting your i
VHPs; or, your schedule for developing such a plan and the
]
basis for that schedule.
2.
A description of the evaluation methods and results used to assess the susceptibility of the CRDM and other VHPs in your plant to PWSCC, including the susceptibility ranking of your plant and the factors used to determine this ranking. Other than or in addition to the boric acid visual examination (see Generic Letter 88-05,
" Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary i
Components in PWR Plants," dated March 17,1988), include a description of all relevant data and/or tests used to develop crack initiation and crack growth models, and the methods and data used to i
validate these models.
Include a statement explaining the applicability of these models to the VHP cracking issue. Also, if you are relying on any integrated industry inspection program, 1
provide a detailed description of this program.
l 2
l Those licensees that have previously submitted the requested 1
information need not resubmit it but should instead reference the appropriate correspondence in their response to this Generic Letter.
l
,CRGR REVIEW PACKAGE 3.
A description of any resin intrusions in your plant, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate l
levels, including the following information:
3.1 Were the intrusions cation, anion, or mixed bed?
3.2 What were the durations o'f these intrusions?
3.3 Do your RCS water chemistry Technical !ipecifications follow tiie EPRI guidelines?
3.4 Identify any RCS chemistry excursions that exceed your plant administrative limits for the following species:
3.5 Identify any conductivity excursions which may be indicative of resin intrusions, provide your technical assessment of each excursion and your followup actions.
3.6 Provide your assessment of the potential for any of these intrusions to result in a significant increase in the probability for IGA of VHPs and any associated plan for inspections.
All addressees are required to submit a written response with the information requested above within 120 days from the date of this letter.
Any inspection results that do ng1 satisfy the acceptance criteria identified in the NRC staff's safety assessment dated November 16, 1993, should be reported to the NRC staff prior to plant restart.
The public reporting burden for this collection of information is estimated to average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
The cost estimated for the collection of information is estimated to average $8000.00 ($100/ hour expended).
(c)
An anticipated schedule for NRC use of the information.
The NRC staff plans to make immediate use of the requested information to verify that licensees are monitoring vessel head penetration cracking l
so as to provide reasonable assurance that existing regulations are being satisfied and to determine if augmented inspection rules need to I
l be developed.
l l
l
.CRGR REVIEW PACKAGE (d)
A statement affirming that the request does not impose new requirements on the licensee, other than for the requested information.
Because the proposed generic letter only requests information from the PWR licensees, and the requested information has already been collected by.the licensees (as stated by the PWR Owners Groups to the NRC during the meeting on August 24,1995), the proposed generic letter does not impose new requirements on the licensees, other than submission of the requested information.,
(xii) An assessment of how the proposed action relates to the Commission's Safety Goal Policy Statement.
The NRC staff feels that the proposed Generic Letter has no impact on the Commission's Safety Goal Policy Statement since it is only requesting information.
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September 26, 1996 LIC-96-0122 U. S. Nuclear Regulatory Comission Attn: Chief, Rules Review and Directives Branch Mail Stop T-6D-69 Washington, D.C.
20555-0001
References:
1.
Docket No. 50-285 2.
Federal Register Volume 61, No.149, dated August 1, 1996 (61 FR40253)
Subject:
Coments on Proposed Generic Communication Regarding Primary Water Stress Corrosion Cracking of Control Rod' Drive Mechanism and Other Vessel Head Penetrations The Omaha Pubite Power District (OPPD) has reviewed the proposed generic comunications regarding primary water stress corrosion cracking of control rod drive mechanisms and other vessel head penetrations. OPPD, as the licensee for Fort Calhoun Station and a member of the Combustion Engineering Owner's Group (CEOG), has been monitoring this issue and has been involved in the development of the initial CE0G susceptibility assessments for the CEDM nozzles at CE0G plants. OPPD plans to participate in an updated assessment of the CEDM nozzles, which will incorporate the results of the CEDM testing performed at the Palisades nuclear plant.
OPPD has the following specific coments on the proposed Generic Letter:
1.
It should be noted in the background information that no CEDM nozzles in any plants worldwide have failed during plant operation.
Evidence of cracking has been revealed during planned inspections. As alluded to in the proposal, any through-wall cracking would be sl ow. result in detectable leakage, and provide an opportunity to take corrective action, because the leak rates of primary systems are tracked during the operation of all nuclear plants.
2.
The proposed response period of 90 days may be insufficient given the recognized potential data collection difficulties and the fact that the owners groups may still be completing cr updating their susceptibility We suggest that the)tter with the owners groups.
assessments.
iRC staff coordinate the issuance and response timing of the Generic
/'
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U. S. Nuclear Regulatory Comission l
LIC-96-0122 Page 2 l
Please contact me if you have any questions.
Sincerely,
- b h
O F'a' T. L. Patterson Division Manager
~
j Nuclear Operations TCWtes c:
Winston & Strawn L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager W. C. Walker, NRC Senior Resident Inspector 1
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August 6,1996 Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 The purpose of this letter is to request an extension of the comment period for the
" Proposed Generic Communication; Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations," (61 Federal Register 40253, August 1,1996). We respectively request that the comment period be extended from September 3,1996, to October 3,1996.
The Nuclear Energy Institute, on behalf of the nuclear utility industry, will be d:,veloping comments on this draft generic letter for submittal to the NRC. We are requesting this comment period extension to permit sufficient time for industry j
representatives to assemble and develo'p comments. This extension will also allow suficient time to transmit our draft comments to all nuclear utilities for consideration in developing and submitting plant-specific comments. We believe that a one month extension is necessary to accommodate industrywide review of the proposed generic letter.
The appropriateness of this extension request is supported by the NRC staffs November 19,1993, safety evaluation report (SER) conclusion that states primary water stress corrosion cracking of Alloy 600 head penetrations "is not a significant safety issue ct this time as long as surveillance walkdowns [visualinspection) in accordance with GIr88-05 continue." Utilities are continuing the walkdowns. Furthermore, recent utility v:lumetric inspections results confirm that the PWR Owners Groups safety evaluations considered in the NRC staffs SER remain valid.
Wo also request that the NRC staff notify either Alex Marion (202-739-8080) or Kurt Cozens (202-739 8085) of the NEI staff of the disposition of this extension request.
Sincerely, P
n Ralph E. Beedle C
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Mr. C.E. (Gene) Carpenter, Jr. (NRR/DRPE/PDI-1)
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Mr. Brian W. Sheron (NRR/DE)
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Tel 212 705-8500 345 Easi 47m Sueet fax 212 705 8501 i
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l Mr. David Meyer Chief, Rules Review and Directives Branch US Nuclear Regulatory Conunission
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Washington, DC 20555-0001 Subj: Proposed Generic Conununication: Priinary Water Simss Corrosion
[
Cracking of Control Rod Drive Mechanisms and Other Head Penetrations C
Dear Mr. Meyer:
Enclosed are,conunents resulting from review by individual meinbers of the ASME BPVC Subconunittee on Nuclear Inservice Inspection. This myiew is not to be construed as a position or opinion on the subject document by ASME; rather, the enclosed conunents am submitted as a constructive public service, and repmsent the opinion ofindividual coinmittee members.
J Yours truly, k
George F ter, Secretar;-
ASME BPVC Subconunittee on Nuclear Inservice Inspection (212)705-8018 cc with encl.
J. Perry D. Landers D. Canonico T. Mawson G. Eisenberg 0
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The America n Society of Meenanical Engineers
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COMMITTEE CORRNSPONDENCE i
commrttee:
ess g Combustion Engineering a
Comments on draft PWSCC Generic Letter 2000 Day Hill Rd.
Windsor, CT 060951521 September 27,1996 date:
Westinghouse Energy Center copy to: P.O. Box 355 Mr. George Fechter Pittsburgh, PA 15230 4355 to:
ASME Codes and Standards We have reviewed the NRC " Proposed Generic Cnr==maication: Pnmary Water Stres have a number of comments to bring to their attention.Crackmg of Ci Members of the ASME Section XI subcommittee are concerned that additiona imposed on the industry for what appears to be a non-safety issue. It is our concem th additional inspection requirements are being justified based on the lack of action by i subject. This is not the case at all.
i The Section XI Subcommittee has been monitoring this issue closely since the conditio identified in France, and it has been kept up to date by regular briefmgs of the Exec at each meeting (four times per year). Their conclusion to date on this issue is th concerns that cannot be addressed by the regular inspections for boron deposits alrea date we have received no requests to add additional inspection requirements for from any of our members, which include representatives of industry, utilities, nationa and the NRC.
The primary concern of the Section XI Subcommittee, as with all other committees in Boiler and Pressure Vessel Cod:, is safety. The principle focus of the Section XI Subc that the integrity of the reactor coolant pressure boundary is maintained. Industry expi indicated that crackmg that has occurred in the control rod drive penetrations is After more than 5000 penetrations have been inspected worldwide, only one penetratio found to have a through-wall crack from PWSCC.
The proposed generic letter comes as a smprise to the Subcommittee, and we res this letter be forwarded to the NRC with our request that they remove any implication that Section XI has been ignoring this issue. Please address the NRC care of Mr Da Chief, Rules Review and Directives Branch, USNRC, Washington, DC 20555-0001.
Sincerely.
A0db-.
)L Owen Hedden, Chairman Subcommittee on Inservice inspection Warren Bamford, Chairman Subgroup on Evaluation Standards The American Society of t
j Mechanical Engineers 346 East 47tn Street.
New York NY 10017 Keep ASME Codes and Standards Department informed i
cc: James A. Perry, BNCS Gery M. Eisenberg, ASME Domenic A. Canonico ABB-CE Gil Millman, USNRC Tom Mawson, Northeast Utilities Don Landers, Teledyne e>
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ENCLOSUREO DETAILED COMMENTS ON TIIE DRAFr GENERIC LETTER CMT#
SECTION PARAGRAPII COMMENT CORRECTIONS 1'
General It is unclear why the draft generic letter (GL) did not reference It would be beneficial to contrast the nor discuss in detail the evaluations and conclusions contained conclusions of NUREGICR-6245 to the draft in NUREG/CR-6245. This document provides a balanced Gils definition of"long-term safety evaluation of the safety evaluations performed by the PWR concerns."
Owners Groups.
2~
General The phrase "other vessel head penetrations" used throughout Additional clarity may be gained if the the draft generic letter should be clarified to read "other reactor phrase "other vessel head penetrations" is vessel closure head penetrations".
altered to read "other reactor vessel closure head penetrations."
3'
Background
let Figure I and text appear to only discuss CRDMs designed by A description of the CE and B&W Westinghouse.
penetration design features would be beneficial.
4'
Background
3rd The first sentence states that in 1989 the emerging issue was A chronologically sequenced paragraph identified, then the second sentence states leakage has occurred would be easier to understand.
since 1986. This is not chronologically correct and is confusing.
===5.
Background===
4th The Bugey.3 cracking was discovered in September 1991.
Proper dates should be used.
6~
Background
4th in the second sentence,it states that the Japanese have Provide a reference or delete Japan from
" uncovered" VitPs with cracks. We are unaware of any the list of countries which have identified available reference stating that the Japanese have detected PWSCC in'their VHPs.
PWSCC cracks in their VilPs. A source reference should be provided.
Improved clarity would be achieved if the word " uncovered" is changed to " detected."
It would be more precise to state that cracks were " detected" rather than " uncovered" in this paragraph and throughout the document's text.
7'
Background
6th Sub-item (3). NUREG/CR-6245 states that the leakage would be The draft GL and NUREG/CR-6245 would detected "long" before significant damage to the reactor vessel be consistent if the word "long" was added head would occur.
before the word "before."
8'
Background
6th The last two sentences discuss manual NDE and do not relate to Delete the last two sentences from the the remainder of the paragraph. The merits of manual NDE paragraph.
and automatic tooling are not the subject of the draft generic letter.
9'
Background
7th The purpose of the EPRI NDE demonstration was not to qualify The EPRI activities would be better tooling or operators, but was limited to the demonstration of an described if the term " qualification" was inspection system's ability to detect and size defects.
modified.
10.
Background
7th This paragraph appears tojustify the draft generic letter based Delete this paragraph.
on advances in inspection techniques rather than assess the safety significance of PWSCC. This implies that inspections
(!
C5IT C SECTION PARAGRAPil COMMENT CORRECTIONS should be required because industry has voluntarily developed improved inspection methods. The paragraph should focus on safety concerns.
I 1.
Background
8th The description of the Zorita event could be more precise.
A more precise statement would be:
"During the 1994 outage at Zorita (a Spanish reactor), visualinspection of the reactor vessel head discovered boron deposits on a single vessel head penetration.
A more thorough inspection of this penetration detected a crack approximately two inches below the bimetallic weld. An extensive investigation and root cause evaluation were performed it was determined that the indications were caused by intergranular stress corrosion cracking initiated by cation resin intrusion."
12'
Background
8th The Zorita concern was primarily with the response of sensitized it would be more precise to refer to " attack material attacked by reduced sulfur species.
by sulfur species on sensitized materials."
13.
Background
8th First sentence. Inspections at the Zorita Plant did not identify These changes would provide factual circumferential cracks in the J-groove weld, but found a clarity.
through wall crack at or near the bimetallic weld.
In the third sentence," resin bed" should be " resin bead."
The text would be better understood if the measurements were provided in English as well as metric units, i.e.," liters" and
" gallons."
14'
Background
9th it is our understanding that the NRC staff has Zorita resin Related reports and data should be made intrusion reports and data that are not publicly available. It is available.
difficult to assess the significance of the Zorita resin intrusion without all available information. In previous communications with the NRC staff, we have been told that these reports have been provided to all PWR Owners Groups. However, inquiries made to the PWR Owners Groups have not supported thia. We request the NRC staff to place allinforma'ior on the Zorita resin intrusion into the Public Document Room, and provide the opportunity for industry to evaluate.
15~
Background
9th and 10th To maintain the chronological order of events, the 9th and loth Chronological order of these paragraphs I
paragraphs should be switched.
would be beneficial.
I 16'
Background
10th The draft generic letter does not discuss the recent ViiPs re-It would be beneficial to A-ant the most inspections performed at Oconee and D.C. Cook, nor the VHP recent inspection activita m.d results.
2
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SECTION PARAGRAPil COMMENT COR RECTIONS t
repair at D.C. Cook.
17.
Background
10th The NRC states that they have not been provided with the WOG The statement in this paragraph should resin intrusion review. IN 96-11 does not require any specific reflect the comment.
l action by licensees. Furthermore, Westinghouse NSAlr94 028 did not request licensees to provide a response back to Westinghouse and no WOG report has been prepared.
18.
Background-13th The citation of Westinghouse, Framatome Technologies, and Use the correct citations.
(
Combustion Engineering are incorrect. The citations should be
^
the PWR Owners Groups;i.e., WOG, B&WOG, and the CEOG.
i 19'
Background
14 -
This paragraph states that "(t)he program outlined in the NEI An appropriate statement would reflect the white paper is based on the assumption that the issue is an Section VII white paper text.
econoraic one rather than a safety issue..." and that the NRC i
staff did not agree that the issue was only economic. This is not It is the NRC staffs prerogative to disagree l
a correct interpretation of the NEl white paper. The white with positions taken in the white paper.
paper documents the extensive safety evaluations developed by llowever, the NRC staff should identify the PWR Owners Groups which addressed all the safety those safety concerns have not been i
concerns identified by the NRC staff. The method discussed in addressed by the NRC approved PWR the white paper to manage RPV head penetration cracking Owners Groups safety evaluations.
acknowledges that the issue is not an immediate safety concern and that leak-before-break will occur. Using this knowledge, the L
management methodology discussed provides a four step approach; of which one step evaluates the economic j
considerations.
20' Discussion let The sentence starting, " Further, if any significant.." is an This change provides clarity.
t absolute statement which has not been technically justified in this document nor the references. It would be technically correct if the sentence was revised to read,"Further,if any significant resin intrusions have occurred at U.S. plants such as occurred at Zorita, the resultant chemistry condition in combination with stress may be significant."
r 21' Discussion 2nd The sentence which starts, Cracking in the VHPs.. "is A more precise statement would be potentially misleading. While cracking has occurred in 116 of
" Cracking occurred in a few VIIPs and could the 5146 penetrations inspected, it has not been observed in the occur in others at some future time. An large majority of VilPs. PWSCC is an age related degradation existing crack may continue to grow, but mechanism which could occur some time in the future, many could stop."
years beyond the initial or renewed license or never.
22-Discussion 2nd The paragraph states that the NRC staff cone'ders the cracking The PWR Owners Group safety evaluations of VIIPc to be a safety concern for the long-term based on the addressed the safety concerns identified by possibility of(1) exceeding the American Society of Mechanical NRC staff.
Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth 3
1
CMTr SECTION PARAGRAPH COMMENT CORRECTIONS for plant safety.
These safety concerns are addressed by the PWR Owners Groups safety evaluations. These were summarized on Page 10 of NUREG/CR 6245," Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking," which states that "There are two major safety concerns associated with CRDM nozzle cracking. First, a crack could eventually lead to a rupture of the nozzle and, if the nozzle is severed, to ejection of the connected CRDM housing. Second, a through-wall crack would allow the borated reactor coolant to come in contact with the vessel head and cause boric acid corrosion of the low-alloy steel base metal.. " In addition, the NRC taf"s safety evaluation dated November 19,1993, states that "The primary safety concern associated with stress corrosion cracking in Alloy 600 is the potential for circumferential cracks. Extensive circumferential cracking could lead to ejection of a CRDM..."
Since the PWR Owners Groups safety evaluations evaluated a through-wall crack and ejection of the connected CRDM housing, it appears that the two long term concerns identified by the draft GL are less severe than those already evaluated.
23.
Required 1.2.a The concept of scheduling augmented inspections is inconsistent Provide clarity.
Information with the concept of"long term safety concerns." Given that technical safety concerns have been addressed, requesting a
" technical basis" for a schedule is unclear.
24' Required 1.2.b The required information is unnecessarily prescriptive (e.g., the Delete as this level of detail is r.ot Information direction of inspection (top or bottom) will not affect the quality necessary.
of an inspection which a licensee may choose to perform, the presence of thermal sleeves, etc.)
25' Required 2.
The first sentence states,".. include the susceptibility ranking Delete the phrase ". include the information of your plant and the factors used to determine this ranking."
susceptibility ranking of your plant and the This phrase is redundant with the first part of the sentence factors used to determine this ranking."
which states,"A description of the evaluation methods and results used to assess the susceptibility of the CRDM and other VHPs in your (>lant to PWSCC, 26' Required 2.
The susceptibility models were not used as input to the PWR Since it is not possible to make a safety information Owners Groups safety evaluations that were submitted and determination with the susceptibility approved by the NRC staff. The susceptibility models and rankings, this paragraph should be deleted.
subsequent rankings may be used by licensees to make economic 4
CMTC SECTION PARAGRAPH COMMENT CORRECTIONS evaluations, but are not sufficiently precise to be used in a safety assessment that may be submitt.ed to the NRC staff. In addition, it is unclear how the NRC staff will use such models to
~
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evaluate a safety concern.
~
27.
Required 2.
This requested information implies that the GL 88-05 visual Boric acid deposits will be identified by the Information inspection is inadequate to detect boric acid deposits and which visualinspections recommended in Generic could be caused by PWSCC. This implication is not supported Letter 88-05.
by operating history and safety evaluations:
The only through-wall VHPs cracks (Bugey and Zorita) were e
detected by visual inspections.
GL 88-05 visual inspections are considered acceptable for detecting PWSCC in the remainder of the reactor coolant system.
A conservative definition for "long term safety concern" implied by NUREG-CR-6245 would infer a minimum of nine years after the initiation of a PWSCC through wallleak.
Boric acid deposited over this time period would be readily observed using the GL 88 05 visualinspections.
28' Required 3.
The intergranular stress corrosion cracking resulting from a The resin induced intergranular stress information Zorita type resin intrusion is a different inechanism than the corrosion cracking is different than the primary water stress corrosion cracking (PWSCC). The resin stated scope.and should be deleted.
intrusion cracking is a degradation mechanism caused by an abnormal operating event and is not a age-related degradation mechanism like PWSCC. Furthermore, the predictive tools for PWSCC are not capable of predicting resin intrusion. It is noted that the VIIP inspections performed on over 5200 penetrations at 87 plants worldwide did not identify any other plant that exhibited intergranular stress corrosion cracking similar to that exhibited at Zorita.
29.
Required 3.4 The draft generic letter has not provided a basis for supplying Delete.
Information information on chlorides, fluorides, oxygen, boron, or lithium.
The Zorita experience has been linked to the sulfates, but to our knowledge the other chemistry species have not been linked.
5
80/03/96 18:38 LICENSING & SPECIAL PROJECTS o 8130141552721078689 NO.313 P032 fh k
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.(((/ // Q Fiori4. Power s ught Company. P.O. Box 14000. Juno Beach. FL 334084420 j
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OCT 0 21996 L-96-249 Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Comminion Mail Stop T-6D-69 Washington, DC 20555-0001
Subject:
Proposed Generic Communication; Pnmary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations (61 FR 40253, dated August 1,1996)
Notice of Opponunity for Public Comment i
On August 1,1996, the Nuclear Regulatory Comminion published for public comment,
Proposed Generic Communication: Pnmary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations." The issuance of the proposed generic letter would request that addressees describe their program for ensuring the timely inspection of PWR control rod drive mechamsm and other vessel head penetrations and require that all addressees provide a written response to the NRC regarding this generic letter. These comments are submitted on behalf of Florida Power & Light (FPL), a licensed operator of two nuclear power plant units in Dade County, Florida and two units in St. Lucie County, Florida.
The Nuclear Energy Institute (NEI) is providing comments on the proposed generic letter (GL) on behalf of the industry. FPL endorses the NEI comments. Additionally, the Nuclear Utility Backfitting and Reform Group (NUBARG) is providing comments on the proposed GL. FPL endorses the NUBARG comments.
FPL appreciates the opponunity to comment on the proposed GL Very truly yours,
&.N.
-m gW. H. Bohlke Vice President Nuclear Engic.eering c$
ene e.
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NEW YORK, Ny 30166-ass 3 M,,(
RlYADH 11495. SAUDI ARABIA FACSIMILE (202) 37t 6950 43 Ryg OU RHONE 1204 GENEVA. SWITZERLAND October 3,1996 1
VIA MESSENGER U.S. Nuclear Regulatory Commission Rules, Review and Directives Branch Two White Flint i
11545 Rockville Pike j
l Rockville, MD 20852-2738 l
' Re:
Coenments on Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations; Proposed Generic Communiestion: 61 Fed. Reg. 40,253 (August 1,1996); 61 Fed.
Rec. 43.393 (August 22.1996)
ATTN: Rules. Review and Directives Branch On August 1,1996, the Nuclear Regulatory Commission (NRC) issued the above-captioned proposed generic communication for public comment. Provided below are the comments of the Nuclear Utility Backr,tting and Reform Group (NUBARG).1' These comments concern the back5tting implications of the proposed generic communication. We support NEl's comments on the substantive aspects of the proposed Generic Letter.
The generic letter would require licensees to provide written responses, including a description of their programs for ensuring the timely inspection of PWR control rod drive mechanisms (CRDMs) and other vessel head penetrations (VHPs). The proposed generic letter states that "[i]f you have not developed a plan to periodically inspect the CRDM and other vessel head penetrations, provide your technical or safety basis for not periodically inspecting your VHPs; or, your schedule for developing such a plan and the basis for that schedule." 61 Fed. Reg. 40,253, 40,255 (1986). The Staff indicates that the integrated, long term program that would include periodic inspections and monitoring is necessary in order to verify that the margins required by the l'
NUBARG is a consortium of 16 nuclear utilities formed in the early 1980s, which participated actively in the development of the NRC's backfitting rule (10 C.F.R. Q 50.109) in 1985, and which has closely monitored the NRC's application of the rule since that time.
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l ' WINSTON & STRAWN
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J U.S.' Nuclear Regulatory Commission 4
i October 3,1996 l
Page 2 i
i ASME Code continue to be satisfied, and to ensure that the safety significance of VHP cracking l
remains low. M. In addition, the Staff believes that the program is needed to determine if the imposition of an augmented inspection program is required. M.
The Staff asserts it is not going to perform a backfitting a5alysis because the proposed generic letter "only requires information from the addressees under the provisions of i
Section 182a of the Atomic Energy Act of 1954, as ammlad and 10 CFR 50.54(f)." M. at 40,256.
}
In this regard, the Staff claims that it is "not establishing a new position for such compliance in this 1
generic letter." M.
1 i
We believe the Staff is required to perform a backfit analysis on the proposed l
imposition of the development of a plan to inspect the CRDM and other vessel head penetrations.
l Under 10 C.F.R.
50.109, a backfit includes "the modification of or addition to... the procedures j
... required to... operate a facility; any of which may result from a new or amended provision in the Commission rules or the imposition of a [new] regulatory staff position...." The backfit rule includes within its scope any means used by the NRC "to create an obligation upon licensees to a
change the... operation of a facility...." 49 Fed. Reg. 47,034,47,035 (1985). Imposition of this l
new inspection requirement goes beyond simply asking licensees to provide an information response. Instead, the new requirement to develop an inspection program is a modification or j
addition to the operational procedures resulting from the imposition of a new regulatory staff j
position. The Commission has indicated that a backfitting analysis should be performed in close j
cases. 50 Fed. Reg. 38,097,38,102 (1985)("The Commission recognizes that there may be instances l
where it is not clear whether a backfit will follow an information request. Those cases should be i
i resolved in favor of analysis."). In this case, a backfitting analysis is required under 10 C.F.R. s 50.109.
A backfitting analysis not only is legally required but also will ensure the protection j
of the public health and safety, as well as provide practical benefits to both licensees and the NRC l
Staff. Specifically, by performing a backfitting analysis, the Staff can ensure that the requested j
periodic inspections and monitoring activities are effective from a safety perspective and are cost beneficial. Moreover, by performing the analysis, the Staff can ensure that unnecessary downtime and adverse schedule impacts are avoided by licensees and that any resulting radiation exposure is assessed and minimized.
i Sincerely, 1
r~
J Daniel F. Stenger
]
Kathryn M. Sutton i
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September 30,1996 Mr. David L. Meyer Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Mr. Meyer:
Enclosed are. Nuclear Energy Institute (NEI) comments on the " Proposed Generic Communication; Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations," (61 Fed. Reg. 40253, August 1, 1996). These comments were developed by an NEI task force comprised of representatives from utilities, PWR Owners Groups and EPRI. Additionally, these comments were forwarded to the industry for consideration by individual utility licensees in developing plant-specific comments.
NEI will continue to coordinate industry activities in managing primary water stress corrosion cracking (PWSCC) of vessel head penetrations. This coordination will involve EPRI and the PWR Owners Groups to ensure that necessary information is evaluated and communicated to utilities to support their decisions to conduct inspections. NEI continues to believe that the decision to conduct inspections.ese.s with individual utility management after due consideration of susceptibility, udence of boric acid deposition and economic risk. As in the past, NEI will continue to meet with NRC staff to discuss inspection results as they relate to the PWR Owners Group safety evaluations and inspection criteria, and the NRC's safety evaluation report.
NEI believes this approach in managing this issue is appropriate and sufficient given the low safety concern. Therefore, NEI concludes that there is no technical or regulatory basis for this generic letter.
1 NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States. nuclear plant designers, major architect / engineering firms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.
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- Mr.' David L. Meyer
, Page 2 eneral comments relating to the draft generic letter are provided in Enclosure 1 and are summarized as follows:
The stated purpose of the draft generic letter is to determine if augmented inspections are warranted. However, the draft generic letter essentially requests licensees to define and commit to an augmented inspection program.
If augmented inspections are determined by NRC to be necessary, then such inspections should be based on the safety significance of the vessel head i
penetrations experiencing primary water stress corrosion cracking, not whether or not licensees are performing augmented inspections.
The NRC staff safety concerns have been addressed by the PWR Owners Groups' safety evaluations, which considered the possibility of through-wall
- cracks, i
The stated scope of the draft generic letter is primary water stress corrosion cracking. The resin intrusion at the Zorita Plant resulted in intergranular stress corrosion which is a different degradation mechanism. Since the Zorita' resin intrusion was communicated to utility licensees by Information Notice 9611, and new concerns have not been identified, it is not clear why the NRC staffis now requesting licensees to submit information on this topic. provides detailed comments on the specific text of the proposed generic letter.
Ifyou have questions concerning these comments, please contact Alex Marion (202-739 8080) or me.
Sincerely, On t
c Ralph E. Beedle TET/AM/ead Enclosures c:
C. E. (Gene) Carpenter, NRC/NRR Brian Sheron, NRC/NRR Jack Strosnider, NRC/NRR
i k
ENCLOSURE 1 1
GENERAL COMMENT
S ON THE DRAFT GENERICIETTER
- 1. Items 1 and 2 in the Required Information section essentially requests licensees to define and commit to an augmented inspection program. The stated purpose
)
of the draft regulatory guide is to evaluate whether or not an augmented j
inspection program is necessary. The justification for the augmented inspection should be based on the safety significance of the vessel head penetration's (VHP) experiencing primary water stress corrosion cracking, not iflicensees are l
currently performing augmented inspections.
- 2. On Page 10 of NUREG/CR-6245, Assessment of Pressurized Water Reactor i
Control Rod Drive Mechanism Nozzle Cracking, it states, "There are two major i
safety concerns associated with CRDM nozzle cracking. First, a crack could i
eventually lead to a rupture of the nozzle and, if the nozzle is severed, to ejection of the connected CRDM housing. Second, a through wall crack would allow the borated reactor coolant to come in contact with the vessel head and cau.; boric
)
acid corrosion of the low alloy steel base metal." In the NRC staff's safc'y 2
evaluation dated November 19,1993, it states, "The primary safety concern
)
associated with stress corrosion cracking in Alloy 600 is the potential for circumferential cracks. Extensive circumferential cracking could lead to ejection j
of a CRDM." These safety concerns were considered by the PWR Owners Group safety evaluations submitted to and accepted by the NRC staff. The draft generic letter has not identified any safety concerns that were not previously evaluated and dispositioned. Summaries of these safety evaluations are I
contained in NUREG/CR 6245 and the NEI's white paper titled, " Alloy 600 RPV l
Head Penetration Primary Water Stress Corrosion Cracking."
l
- 3. The second paragraph of the Discussion section states that the goal of the draft i
generic letter is to "... verify that the margins required by the ASME Code as specified in 5 50.55a of Title 10 of the Code of Federal Regulations (10 CFR j
50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A i
to 10 CFR Part 50 (10 CFR Par 50, Appendix A, GDC 14)is continued to be satisfied,...." These goals are unique and separate from the stated purpose of j
the first paragraph in the Required Information section which states; "The information requested in Items 1 and 2, below, is required to determine if the imposition of an augmented inspection program is required...." Although not stated as such, the Discussions section appears to raise a question of compliance i
rather than determining if new regulatory requirements (augmented inspections per 10 CFR 50.55a(g)(6)(ii)) should be imposed. Utility licensees are presently in compliance with the requirements identified in the Discussion section based on the following:
i a
I
The design and fabrication of the reactor vessel heads satisfy all applicable ASME requirements.
Only the welds that attach VHPs to the reactor head are within the scope of the inservice inspection requirements (ASME Section XI, Table IWB-2500-1, Examination Category B-E). As noted in NUREG/CR 6245, the VHP surface which could experience PWSCC is not expected to be within the scope of ASME inservice inspections. However, should inservice inspection identify indications, licensees will disposition them per the ASME Code.
GDC-14 states, "The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."
Licensees are meeting GDC-14 because:
- The reactor vessel head was designed, fabricated, and erected to the ASME Code or other requirements apprcved by the NRC.
- The PWR Owners Group safety evaluations, accepted by the NRC staff (dated November 19,1993), addressed the potential for rapid crack propagation, gross rupture and abnormalleakage. These evaluations determined that PWSCC would either be arrested ar would grow very slowly requiring years to obtain a critical length.
. Axial cracks require many years to obtain criticallength.
Circumferential cracking requires through-wall leakage and will take signSicantly more time than the 40-year licensed operating period.
One conservative circumferential cracking evaluation estimated that it would take in excess of 90 years before gross failure would occur.
- Licensees are presently performing inspections in accordance with NRC Generic Letter 88 05 to detect leakage that could occur during.
operation. Ifleakage is detected, repairs and corrective action will be performed. In addition, corrective action is required ifleakage exceeds the Technical Specification criteria.
- This approach to GDC compliance is consistent with the leak-before break criteria applied to other primary piping systems.
- 4. The Required Information section asks licensees to summarize the inspections they have performed, define the inspections they plan to perform or justify why inspections are not being performed. The NRC staff witnessed the VHP inspections performed by licensees (Sve plants) and has received written reports on the results. Hence, this is a redundant request for those licensees who have already performed inspections and requests submittal ofinformation that the NRC already has in its possession. In addition, the NEI white paper discussed the method by which licensees are managing this issue, i.e., future inspections will be performed based on information sharing, predictive methodologies and tools, inspection results, and development of mitigation and repair technologies.
2
- 5. The topic of the draft generic letter is " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations." The inclusion of a different form of degradation (intergranular stress corrosion cracking due to resin intrusion)is not warranted. PWSCC of Alloy 600 is an time dependent degradation mechanism. Intergranular stress corrosion cracking due to resin intrusion is an abnormal operating event. Furthermore, of the over 5200 penetrations inspected worldwide, no evidence has been observed that suggests resin induced intergranular stress corrosion cracking has occurred in any reactor vessel other than Zorita. This is strong evidence that resin induced intergranular stress corrosion cracking is an outlier event that is not generic.
- 6. The NRC staffissued Information Notice (IN) 96-11, " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," that advised licensees of tne Zorita resin intrusion and potential intergranular stress corrosion cracking. It is unclear why the NRC has revised their position concerning a request for submitted information (Required Informati:n, Item 3), since no additional resin intrusions concerns have occurred since IN 96-11 was issued. The extra burden on licensees to respond to Item 3 of the Required Information section is not justified.
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50-348 50-364 Mr. David L. Meyer Chief, Rules Review and Directives Branch
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l U. S. Nuclear Regulatory Commission Washington, D. C. 20555 c.
Comments on Proposed Generic Communication
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" Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Othet Vessel Head Penetrations"
.o (61 Federal Register 40253 dated August 1.1996)
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Dear Sir:
Southern Nuclear Operating Company has reviewed the proposed rule " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations," published in the Federal Register on August 1,1996. In accordance with request for comments, Southern Nuclear Operating Company is in total agreement with the NEI comments which are to be provided to the NRC.
Respectfully submitted,
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Dave Morey DNM/TWS 1
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U. S. Nuclear Regulatory Commission Page Two cc:
Southem Nuclear Ooeratine Comoany R. D. Hill, Plant Manager U. S. Nuclear Regulatory Commission. Washington. DC J. I. Zimmerman, Licensing Project Manager, NRR U. S. Nuclear Regulatory Commission. Region II S. D. Ebneter, Regional Administrator T. M. Ross, Senior Resident Inspector P
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P C.K.McCoY October 1, 1996 Georgia Power V.ce Prescent Nuctear Vogne Protect the southern electrc system Docket Nos.
50-321 50-424 HL-5247 50-366 50-425 LCV-0885 Mr. David L. Meyer Chief, Rules Review and Directives Branch R
3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 e
Comments on Proposed Generic Communication
" Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Othey i
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Vessel Head Penetrations"
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(61 Federal Register 40253 dated August 1.1996)
- l1
Dear Sir:
G'eorgia Power Company has reviewed the proposed rule " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations," published in the Federal Register on August 1,1996. In accordance with request for comments, Georgia Power Company is in total agreement with the NEI comments which are to be provided to the NRC.
i Should you have any questions, please advise.
Respectfully submitted, a. *f '
C. K. McCoy CKM/TWS O
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GeorgiaPbwer d october 1. 1996 U. S. Nuclear Regulatory Commission Page Two cc:
Georgia Power Company J. T. Beckham, Jr., Vice President - Plant Hatch J. B. Beasley, General Manager - Vogtle Electric Generating Plant H. L. Sumner, Jr., General Manager - Plant Hatch U. S Nuclear Reoidatory Commission. Washington. DC K. N. Jabbour, Licensing Project Manager - Hatch L. L. Wheeler, Licensing Project Manager, Vogtle U. S. Nuclear Regulatory Commission. Region II l
S. D. Ebneter, Regional Administrator B. L. Holbrook, Senior Resident Inspector - Hatch C. R. Ogle, Senior Resident Inspector - Vogtle HL-5247 LCV-0885 REES File: G.03.19
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7, hfg /4 Florida PowIr & Light Cempiny, P O. Bsz 14000. Juno Buch, FL 33408 0420 om razy ppt OCT 0 21996 g f, p7f
,g L-96-249 p
Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555-0001
Subject:
Proposed Generic Communication; Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations (61 FR 40253, dated August 1,1996)
Notice of Opportunity for Public Comment On August 1,1996, the Nuclear Regulatory Conunission published for public comment,
" Proposed Generic Communication; Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations." The issuance of the proposed generic letter'would request that addressees describe their program for ensuring the timely inspection of PWR control rod drive mechanism and other vessel head penetrations and require that all addressees provide a written response to the NRC regarding this generic letter. These comments are submitted on behalf of Florida Power & Light (FPL), a licensed operator of two nuclear power plant units in Dade County, Florida and two units in St. Lucie County, Florida.
The Nuclear Energy Institute (NEI) is providing comments on the proposed generic letter (GL) on behalf of the industry. FPL endorses the NEI comments. Additionally, the Nuclear Utility Backfitting and Reform Group (NUBARG) is providing comments on the proposed GL. FPL endorses the NUBARG comments.
FPL appreciates the opportunity to comment on the proposed GL.
Very truly yours, N k.cS w gW. H. Bohlke Vice President Nuclear Engineering k
M nn FPL Group company
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e u:Ma M1 ELECTRIC October 3, 1996 C. Imce Terry Groep he Prendent Chief, Rules Review and Directives Branch U. S. Nuclear Regulatory Commission Washington, DC 20555 0001
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
END0RSEMEi!T OF NEI COMMENT LETTER ON PROPOSED NRC GENERIC COMMUNICATION, PRIMARY WATER STRESS CORROSION CRACKING OF CONTROL R0D CRIVE MECHANISH AND OTHER VESSEL HEAD PENETRATIONS REF:
1) 61 Federal Register 40253, Proposed Generic Communication, Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations, dated August 1, 1996 2)
Nuclear Energy Institute (NEI) letter, addressed to i
Chief, Rules Review and Directives Branch, USNRC, dated October 3, 1996 Gentlemen:
In response to the Federal Register notice of August 1,1996 (Reference 1)
TU Electric is providing comments on the proposed NRC Generic Communication
Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations."
TV Electric has reviewed and endorses the NEI letter (Reference 2).
TU Electric agrees with the NEI discussed issues, recommendations and rationale. TV Electric further agrees that no technical or regulatory basis exists for this generic communication and recommends that NRC provide due consideration of the NEI comments in the final evaluation of the proposed generic communication.
Sincerely, C. L. Terry By:
- S. Marshall Generic Licensing Manager RTB/grp 3
c-R. E. Beedle, NEI Q
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4 P. O. Box 1002 Glen Rose. Texas 76043
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october 4, 1996 yi N.22'*Lan Mr. David L. Meyer Chief, Rules Review and Directives Branch U. S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555-0001
Subject:
Beaver Valley Power Station, Unit No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Proposef Generic Communication, " Primary Water Stress Corrosion
, Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations"
Dear Mr. Meyer:
Duquesne Light Company (DLC) is responsible for the operation of Beaver Valley Power Station Units 1 and 2. DLC has reviewed the proposed generic communication,
" Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations," which was published in the Federal Recister on August 1, 1996 (61 FR 40253). ELC hereby submits the following comments.
DLC endorses the comments provided by the Nuclear Energy Institute (NEI). The NEl comments identify the key issues which need to be considered. DLC concurs with NEI that there is no technical or regulatory basis for this generic letter.
Thank you for the opportunity to comment on this issue. If you have any questions on this submittal, please contact Mr. Roy K. Brosi, Manager, Nuclear Safety Departnent, (412) 393-5210.
Sincerely, 9
Sushil C. Jain
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October 4,1996 Chief, Rules Review and Directives Branch Serial No.'96-076 U. S. Nuclear Regulatory Commission Washington,DC 20555 Gentlemen:
COMMENTS ON PROPOSED GENERIC LETTER:
PRIMARY WATER STRESS CORROSION CRACKING OF CONTROL ROD DRIVE MECHANISMS AND OTHER VESSEL HEAD PENETRATIONS On August 1,1996. the NRC requested comments on the " Proposed Generic Communication; Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanisms and Other Vessel Head Penetrations,"(61 Fed. Reg. 40253, August 1,1996).
We have reviewed the NRC proposed generic letter and fully endorse the Nuclear Energy Institute review comments provided in their letter dated September 30,1996.
We appreciate the opportunity to make comments on this proposed generic letter. Should you have any additional questions, please feel free to contact us.
Very tmly yours, M.
1, Vice President ineering & Services Attachment cc: Mr. Thomas E.Tipton Vice President, Operations and Engineering Nuclear Energy Institute 1776 Eye Street Suite 300 lq Washington, DC 20006-3706 y
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u October 3,1996 Mr. David L. Meyer Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Mr. Meyer:
Enclosed are Nuclear Energy Institute (NEI)2 comments on the " Proposed Generic Communication; Primary Water Stress Corrosion Cracking of Control R '. Drive 1
Mechanism and Other Vessel Head Penetrations," (61 Fed. Reg. 40253,.^.ugust 1, 1996). These comments were developed by an NEI task force comprised of representatives from utilities, PWR Owners Groups and EPRI. Additionally, these comments were forwarded to the industry for consideration by individual utility i
licensees in developing plant speci5c comments.
j NEI will continue to coordinate industry activities in managing primary water stress corrosion cracking (PWSCC) of vessel head penetrations. This coordination will involve EPRI and the PWR Owners Groups to ensure that necessary information is evaluated and communicated to utilities to support their decisions to conduct inspections. NEI continues to believe that the decision to conduct inspections rests with individual utility management after due consideration of susceptibility, evidence of boric acid deposition and economic risk. As in the past, NEI will continue to meet with NRC staff to discuss inspection results as they relate to the Owners Group safety evaluations and inspection criteria, and the NRC's safety evaluation report. NEI believes this approach in managing this issue is appropriate and sufficient given the low safety concern. Therefore, NEI concludes that there is no technical or regulatory basis for this generic letter.
1 NEI is the organization responsible for establialung umned nuclear industry policy on matters affectag the nuclear energy industry, includmg the regulatory aspects of generic operational and -
i techrucal issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect /engineermg nrms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.
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Mr. David L. Meyer, Chief L
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- October 3,1996 i
l Page 2 4
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Comments relating to the general thrust of the draft generic letter are provided in
- and are summarized as follows:
1 The draft generic letter essentially requests licensees to define and commit to an augmented inspection program. The stated purpose of the draft generic
]
letter is to determine if augmented inspections are warranted. If augmented i
inspections are determined by NR. to be necessary, then such inspections C
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should be based on the safety signi6cance of the vessel head penetrations experiencing primary water stress corrosion cracking, not whether or not l
licensees are currently performing-inspections.
The NRC staff safety concerns have been addressed by the PWR Owners 1
i Groups' safety evaluations, which considered the possibility of through wall cracks.
n The stated scope of the draft generic letter is primary water stress corrosion l
cracking. The resin intrusion at the Zorita Plant resulted in intergranular j
stress corrosion which is a different degradation mechanism. Sinct the Zorita 3
resin p1 trusion was communicated to utility licensees by Information Notice 96-I 11, and new concerns have not been identi6ed, it is not clear why the NRC staff is now requesting licensees to submit information on this topic. provides detailed comments on the speci6c text of the proposed generic letter.
i l
If you have questions concerning these comments, please contact Alex Marion (202-739 8080) or me.
Sincerely, Ralph E. Beedle TET/AM/ead Enclosures c:
C. E. (Gene) Carpenter, NRC/NRR Brian Sheron, NRC/NRR Jack Strosnider, NRC/NRR O
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ENCLOSURE 1
GENERAL COMMENT
S ON THE DRAFT GENERIC LETTER
- 1. Items 1 and 2. in the Information Requested section, essentially requests 4
licensees to define and commit to an augmented inspection program. The stated i
purpose of the draft regulatory guide is to evaluate whether or not an augmented inspection program is necessary. The justification for the augmented inspection should be based on the safety signi6cance of the vessel head 3
i penetration's (VHP) experiencing primary water stress corrosion cracking..not if j
licensees are currently performing augmented inspections.
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- 2. On Page 10 of NUREG/CR-6245, Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking, it states, "There are two major safety concerns associated with CRDM nozzle cracking. First, a crack could i
eventually lead to a rupture of the nozzle and, if the nozzle is severed, to ejection of the connected CRDM housing. Second, a through wall crack would allow the borated reactor coolant to come in contact with the vessel head and cause boric I
i acid corrosion of the low alloy steel base metal." In the NRC staffs safety evaluation dated November 19,1993, it states, "The primary safety concern associated with stress corrosion cracking in Alloy 600 is the potential for i
circumferential cracks. Extensive circumferential cracking could lead to ejection of a CRDM." These safety concerns were considered by the PWR Owners Group i
safety evaluations submitted to and accepted by the NRC staff. The draft generic letter has not identi6ed any safety concerns that were not previously evaluated and dispositioned. Summaries of these safety evaluations are contained in NUREG/CR-6245 and the NEI's white paper titled, " Alloy 600 RPV Head Penetration Primary Water Stress Corrosion Cracking."
- 3. The second paragraph of the Discussion section states that the goal of the draft generic letter is to "... verify that the margins required by the ASME Code as
.speci6ed in 5 50.55a of Title 10 of the Code of Federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Par 50, Appendix A, GDC 14)is continued to be satis 6ed,...." These goals are unique and separate from the stated purpose of the Erst paragraph in the Required Information section which states;"The information requested in Items 1 and 2, below, is required to determine if the i
imposition of an augmented inspection program is required,..." Although not stated as such, the Discussions section appears to raise a question of compliance
. rather than determining if new regulatory requirements (augmented in'spections per 10 CFR 50.55a(g)(6)(ii)) should be imposed. Utility licensees are presently in compliance with the requirements identined in the Discussion section based on the following:
4 l
l*_.. -. _.. _ __ _ _ _ _ _ _._. _ _ _ _
The design and fabrication of the reactor vessel heads satisfy all applicable l
ASME requirements.
Only the welds that attach VHPs to the reactor head are within the scope of l
the inservice inspection requirements (ASME Section XI, Table IWB 25001, Examination Category B E). As noted in NUREG/CR 6245, the VHP surface which could experience PWSCC is not expected to be within the scope of i
ACME inservice inspections. However, should inservice inspection identify indications, licensees will disposition them per the ASME Code.
GDC-14 states, "The reactor coolant pressure boundary shall be design'ed, j
fabricated, erected, and tested so as to have an extremely low probability of abnormalleakage, of rapidly propagating failure, and of gross rupture."
i Licensees are meeting GDC-14 because:
The reactor vessel head was designed, fabricated, and erected to the ASME Code or other requirements approved by the NRC.
j The PWR Owners Group safety evaluations, accepted by the NRC staff (dated November 19,1993), addressed the potential for rapid crack propagation, gross rupture and abnormalleakage. These evaluations determined that PWSCC would either be arrested or would grow very slowly reqmring years to obtain a critical length.
l Axial cracks require many years to obtain critical length.
)
Circumferential cracking requires through-wall leakage and will take f
signi6cantly more time than the 40-year licensed operating period.
i One conservative circumferential cracking evaluation estimated that j
it would take in excess of 90 years before gross failure would occur.
- Licensees are presently performing GL 88 05 inspections to detect leakage that could occur during operation. Ifleakage is detected, repairs and corrective action will be performed. In addition, corrective action is required ifleakage exceeds the Technical Specification criteria.
- This approach to GDC compliance is consistent with the leak-before-break criteria applied to other primary piping systems.
- 4. The Requested Information section asks licensees to summarize the inspections they have performed, define the inspections they plan to perform or justify why inspections are not being performed. The NRC staffwitnessed the VHP inspections performed by licensees (Sve plants) and has received written reports on the results. Hence, this is a redundant request for those licensees who have already performed inspections and requests submittal ofinformation that the NRC already has in its possession. In addition, the NEI white paper discussed
. the method by which licensees are managing this issue, i.e., future insp'ections will be performed based on information sharing, predictive methodologies and tools, inspection results, and development of mitigation and repair technologies.
2
1 J. -
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- 5. The topic of the draft generic letter is " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations.' The inclusion of a different form of degradation (intergranular stress corrosion cracking due to resin intrusion) is not warranted. PWSCC of Alloy 600 is an time dependent degradation mechanism. Intergranular stress corrosion cracking due to resin intrusion is an abnormal operating event. Furthermore, of the over 5200 penetrations inspected worldwide, no evidence has been observed that suggests resin induced intergranular stress corrosion cracking has occurred in any reactor vessel other than Zorita. This is strong evidence that resin' induced intergranular stress corrosion cracking is an outlier event that is not generic.
- 6. The NRC staffissued Information Notice (IN) 96-11, " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," that advised licensees of the Zorita resin intrusion and potentialintergranular stress corrosion cracking. It is unclear why the NRC has revised their position concerning a request for submitted information (Required Information, Item 3), since no additional resin intrusions concerns have occurred since IN 96-11 was issued. The extra burden on lice sees to respond to ltem 3 of the Required Information section is justified.
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6 3
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Comment Resolution Requested extension of comment period to October 3, Extended Comment Period to October 3, 1996.
1996.
2 It should be noted in the background information that Draft CL will be revised to reflect comment no CEDM nozzles in any plant worldwide have failed during plant operations. Evidence of cracking has been revealed during planned inspections. As alluded to in the proposal, any through-wall cracking would be slow, result in detectable leakage, and provide an opportunity to take corrective action, because the leak rates of primary systems are tracked during the operation of all nuclear plants.
The proposed response period of 90 days may be Response period will be extended to 120 days.
insufficient given the recognized potential data collection difficulties and the fact that the owners groups may still be completing or updating their susceptibility assessments. We suggest that the NRC Staff coordinate the issuance and response timing of the generic Letter with the owners groups.
3 Members of the ASME Section XI subcommittee are The NRC Staff neither stated nor implied that concerned that additional inspections are being the lack of action of the ASME,Section XI imposed on the industry for what appears to be a non-subcommittee on this issue was a part of the safety issue.
It is our concern that these additional decision reached to issue the generic letter.
inspection requirements are being justified based on The Staff's concern is that this issue is a long the lack of action by section XI on this subject.
term safety concern that the industry needs to address.
The Nuclear Energy Institute (NEI) is providing See responses to NEI comments.
comrnents on the proposed generic letter (GL) on behalf of the industry.
FPL endorses the NEI comments.
4 By letter dated iugust 6,1996, comments from Ralph Beedle, Senior Vice President; Nuclear Energy Institute; 1776 I Street; washington, DC By letter dated September 26,1996, comments from T. L. Patterson, Division Manager; Omaha Public Power District; 444 South 16th Street Mall; Omaha, NR 68102-2247 2
3 By letter dated September 30,1996 comments from George Fecter, Secretary; ASME BPVC Subcommittee on Nuclear Inservice Inspection; 345 East 47th Street; New York, NY 10017 By letter dated October 2,1996, comments from W. H. Bohlke, Vice President; Florida Power & Light Co.; P.O. Box 14000; Juno Beach, FL 33408 GL 96-XX, Degradation of CRDMs & Other VHPs (61 FR 40253) Comment Resolution
-l-
Comment Resolution Additionally, the Nuclear Utility Backfitting and See Response to NUBARG comments.
Reform Group (NUBARG) is providing comments on the proposed GL.
FPL endorses the NUBARG comments.
8 We believe the Staff is required to perform a backfit The Staff has reviewed the requirements, and has analysis on the proposed imposition of the development determined that requesting information to of a plan to inspect the CRDM and other vessel head determine if additional inspection requirements penetrations.
Imposition of this new inspection are necessary is not a backfit. Therefore, no requirement goes beyond simply asking licensees to backfit analysis needs be performed.
If the provide an information response.
Instead, the new staff determines, after reviewing the requested requirement to develop an inspection program is a information, that imposing additional inspection modification or addition to the operating procedures requirements is necessary, a backfit analysis resulting from the imposition of a new regulatory will be performed.
Staff position.
Specifically, by performing a backfitting analysis, the Staff can ensure that the requested periodic inspections and monitoring activities are effective from a safety perspective and are cost beneficial. Moreover, by performing the analysis, the Staff can ensure that unnecessary downtirne and adverse schedule impacts are avoided by licensees and that any resulting radiation exposure is assessed and minimized.
Items 1 and 2 in the Required Information section The proposed CL is requesting information from essentially requests licensees to define and commit to the licensees. The GL does not require a an augmented inspection program. The stated purpose commitment to an integrated, long-term of the draft regulatory guide is to evaluate whether inspection program, but rather asks what, if i
or not an augmented inspection program is necessary.
eny, periodic inspections licensees are The justification for the augmented inspection should performing, and the bases for concluding be based on the safety significance of the vessel head acceptability of these plans to (not) perform penetration's (VHP) experiencing primary water stress inspections.
corrosion cracking, not if licensees are currently performing augmented inspections.
8 By leust dated October 3,1996, comments frorn Kathryn M. Sunun; Winston & Strawn; 1400 L Stact NW: Waaingwn, DC 2OlMB By lener dated Septernber 30,1996, conunents from Ralph Beedle, Senior Vice President; Nuclear Energy Institute; 1776 i Street; Washington, DC Gl. 96-XX, Degradation of CRDMs & Other VHPs (61 FR 40253) Comment Resolution -
1 Comment Resolution On Page 10 of NUREG/CR-6245, Assessment of Pressurized Since (1) the industry has not provided Water Reactor Control Rod Drive Mechanism Nozzle
.information to the Staff regarding possible Cracking, it states, "There are two major safety primary water contamination that could increase concerns associated with CRDM nozzle cracking. First, the potential for circumferential cracking, and a crack could eventually lead to a rupture of the (2) the intent of the letter is to collect nozzle and, if the nozzle is severed, to ejection of information to understand licensee's plans for 1
the connected CRDM housing. Second, a through-wall inspection and monitoring to assure that the crack would allow the borated reactor coolant to come assumptions in their analytic safety evaluations in contact with the vessel head and cause boric acid are maintained over the long term, the Staff has corrosion of the low-ahoy steel base metal." In the determined that the requested information is l
NRC Staff's safety evaluation dated November 19, 1993, necessary to determine if the imposition of an it states, "The primary safety concern associated with augmented inspection program, pursuant to stress corrosion cracking in Alloy 600 is the 10 CFR 50.55a(g)(6)(ii), is required.
4 potential for circumferential cracks. Extensive circumferential cracking could lead to ejection of a CRDM."
These safety concerns were considered by the i
PWR Owners Group safety evaluations submitted to and accepted by the NRC Staff.
The draft generic letter has not identified any safety concerns that were not previously evaluated and dispositioned. Summaries of these safety evaluations are contained in NUREG/CR-6245 and the NEI's white paper titled, " Alloy 600 RPV Head Penetration Primary Water Stress corrosion Cracking."
GL 96-XX, DegHar= of CRDMs & 06er VHPs (6I FR 40253) Commess Resoludon. -.
Comment Resolution The second parasraph of the Discussion section states The commentor misunderstood the second paragraph that the goal of the draft generic lotter is to "...
of the Discussion section, which states, in verify that the margins required by the ASitE Code as parts specified JLn 550.55a of Title 10 of the Code of "The NRC Staff considers cracking of VHPs to Federal Regulations (10 CFR 50.55a) are met, that the be a safety concern for the long term....
guidance of General Design Criterion 14 of Appendix A Therefore, in order to verify that the to 10 CFR Part 50 (10 CFR Par 50, Appendix A, GDC 14),
margins required by the ASME Code, as is continued to be satisfied,
" These goals are specified in [10 CFR 50.55a] are met..."
unique and separate from the stated purpose of the first paragraph in the Required Information section which states;'"The information requested in Items I and 2, below, is required to determine if the imposition of an augmented inspection program is required....... Although not stated as such, the Discussions section appears to raise a question of compliance rather than determining if new regulatory requirements (augmented inspections per 10 CFR 50.55a(g)(6)(ll)) should be imposed. Utility licensees are presently in compliance with the requirements identified in the Discussion section based on the following:
The design and fabrication of the reactor vessel The quality of the design and construction is heads satisfy all applicable ASME requirements, not at issue.
The construction code does not address the operating environment nor the fact that IGSCC is an age-related degradation mechanism; it only contains a general corrosion statement.
Only the welds that attach VHPs to the reactor head The Regulations require that ASME Section XI be are within the scope of the inservice inspection met for the life of the plant.
If no specific requirements (ASME Section XI, Table 10-2500-1, requirements are included in Section XI for Examination Category B-E).
As noted in NUREG/CR-inspection and evaluation, an augmented 6245, the VHP surface which could experience PWSCC inspection program, pursuant to is not expected to be within the scope of ASME 10 CFR 50.55a(g)(6)(ii), may be required. The inservice inspections.
However, shculd inservice proposed GL is requesting information from the inspection identify indications, licensees will licensees to provide adequate assurance that disposition them per the ASME Code.
margins and defense-in-depth are being maintained for the long term.
GDC-14 states, "The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture." Licensees are meeting GDC-14 because:
GL 96-XX, Degradanon of CRDMs & Other VilPs (61 FR 40253) Commens Resoludon Pa===nt Resolution The reactor vessel head was designed, fabricated, and erected to the ASME Code or other requirements approved by the NRC.
The PWR Owners Group safety evaluations, accepted by the NRC Staff (dated November 19, 1993),
addressed the potential for rapid crack propagation, gross rupture and abnormal leakage.
These evaluations determined that PWSCC would
.either be arrested or would grow very slowly requiring years to obtain a critical length.
Axial cracks require many years to obtain critical length. Circumferential cracking requires through-wall leakage and will take significantly more time than the 40-year licensed operating period. One conservative circumferential cracking evaluation estimated that it would take in excess of 90 years before gross failure would occur.
Licensees are presently performing inspections in As was reported in IN 86-108, Supplement 1, accordance with NRC Generic Letter 88-05 to detect rapid and severe corrosion caused by boric acid leakage that could occur during operation.
If leakage can occur before it is noticed. While leakage is detected, repairs and corrective action the GL 88-05 program will find. gross leakage will be performed.
In addition, corrective action some period of time after it began, the concern is required if leakage exceeds the Technical of this GL is whether licensees are able to Specification criteria.
maintain required margins and defense in depth, as discussed above.
This approach to GDC compliance is consistent with This concern is directed towards the reactor the leak-before-break criteria applied to other vessel, not primary piping. Leak before break primary piping systems.
in the context of the Regulations is only applied to piping that has Do active degracation mechanisms. Therefore, the approach is not applicable.
=
GL 96-XX. Degradation of CRDMs & Other VHPs (6I FR 40253) Commens Resolution I
I l
Comment Resolution
)
The Required Information section asks licensees to Those utilities who have previously submitted summarize the inspections they have performed, define the requested information will not need to i
the inspections they plan to perform or justify why resubmit; it will be sufficient that they inspections are not being performed. The NRC Staff reference their previous submittals in their witnessed'the VHP inspections performed by licensees response letter.
(five plants) and has received written reports on the results.
Hence, this is a redundant request for those licensees who have already performed inspections and requests submittal of information that the NRC already has in its possession. In addition, the NEI white paper discussed the method by which licensees are managing this issue, i.e.,
future inspections will be performed based on information sharing, predictive methodologies and tools, inspection results, and development of mitigation and repair technologies.
The topic of the draft generic letter is " Primary The GL addresses the potential degradation of Water Stress Corrosion Cracking of Control Rod Drive CRDHs and other VHPs.
There are various Mechanism and Other Vessel Head Penetrations." The mechanisms (e.g.,
IGSCC, PWSCC, etc.) by which inclusion of a different form of degradation this degradation does occur, all of which have (intergranular stress corrosion cracking due to resin been seen at least once.
The Staff's position intrusion) is not warranted. PWSCC of Alloy 600 is an has been that there is an increased likelihood time dependent degradation mechanism.
Intergranular of stress corrosion cracking of PWR VHP if stress corrosion cracking due to resin intrusion is an demineralizer resins contamina'te the RCS, and abnormal operating event.
Furthermore, of the over that licensees will and consider actions, as 5200 penetrations inspected worldwide, no evidence has appropriate, to avoid similar problems.
been observed that suggests resin induced Therefore, to clarify the intent of the GL, the intergranular stress corrosion cracking has occurred title will be changed to reflect more clearly in any reactor vessel other than Zorita. This is the degradation concerns of the GL.
strong evidence that resin induced intergranular stress corrosion cracking is an outlier event that is not generic.
The NRC Staff issued Information Notice (IN) 96-11, IN 96-11 did not request specific action nor "Inoress of Demineralizer Resins Increases Potential written response from licensees, for Stress Corrosion Cracking of Control Rod Drive Since the PWROGs have already told the Staff Mechanism Penetrations," that advised licensees of the during the various meetings referenced in the Zorita resin intrusion and potential intergranular Draft GL that they have evaluated their plants stress corrosion cracking.
It is unclear why the NRC for susceptibility to a Zorita-type incident, has revised their position concerning a request for providing that information to the Staff should submitted information (Required Information, Item 3),
not be burdensome.
since no additional resin intrusions concerns have occurred since IN 96-11 was issued. The extra burden on licensees to respond to Item 3 of the Required Information section is not justified.
i GL %-XX. Degradation of CRDMs & Other VHPs (61 FR 40253) Commera Resolution 4-l
1 Comment Resolution It is unclear why the draft generic letter (GL) did NUREG/CR-6245 is listed as a reference under not reference nor discuss in detail the evaluations Related Generic Communications. The NRC staff's and conclusions contained in NUREG/CR-6245. This
' review of the PWR Owners Groups' safety document provides a balanced evaluation of the safety evaluation was documented in the etaff's SER evaluations performed by the PWR Owners Groups.
dated November 19, 1993. The safety evaluation stated, in part, the staff recommends that you consider enhanced leakage detection by visually examining the reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area i
nondestructive examinations shculd be performed to ensure there is no unexpected cracking in domestic PWRs.
These examinations l
do not have to be conducted immediately... As the surveillance walkdowns proposed by NUMARC are not intended for detecting small leaks, it is conceivable that some affected PWRs could i
potentially operate with small undetected i
leakage at CRDM/CEDM penetrations.
In this regard, the staff believes that it is prudent for NUMARC to consider the implementation of an enhanced leakage detection method for detecting small leaks during plant operation."
)
The phrase "other vessel head penetrations" used Corrected in final draft of GL.
1 throughout the draft generic letter should be clarified to read "other reactor vessel closure head penetrations".
Figure 1 and text appear to only discuss CRDMs The figure is only for illustrative purposes, it designed by Westinghouse.
is not considered necessary to show a type of every vendor's CRDMs.
The first sentence (in the third paragraph of the Corrected in final draft of GL.
Background section) states that in 1989 the emerging issue was identified, then the second sentence states leakage has occurred since 1986. This is not i
chronologically correct and is confusing.
The Bugey-3 cracking was discovered in September 1991.
Corrected in final draft of GL.
4-GL %-XX. Degradation of CRDMs & Other VHPs (61 FR 40253) Commens Resolution
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Comment Resolution l
In the second sentence [in the fourth paragraph of the The sentence is poorly worded.
It was not the l
Background section], it states that the Japanese have Staff's intent to imply that cracking was found i
" uncovered" VHPs with cracks. We are unaware of any at Japanese plants (the Staff is not aware of available reference stating that the Japanese have any such cracking), but rather that, like the detected PWSCC cracks in their VHPS. A source other countries, the Japanese inspected for i
reference should be provided.
cracking. Further, in Japan, the three most susceptible vessel heads were scheduled to be replaced because of safety considerations.
The sentence will be reworded.
It would be more precise to state that cracks were Corrected in final draft of GL.
" detected" rather than " uncovered" in [the fourth paragraph of the Background section) and throughout l
the document's text.
NUREG/CR-6245 states that the leakage would be This statement is true; however, as was stated detected "long" before significant damage to the in the staff's November 19, 1993, safety reactor vessel head would occur.
evaluation:
"...NUMARC should consider methods for detecting smaller (than 1 gpm) leaks to provide defense-in-depth to account for any potential uncertainty in its analyses."
Relying solely upon detection of leakage, when the Standard Technical Specifications (TSs) and most facilities' TSs state that no pressure boundary leakage is acceptable, is not consistent with maintaining defense-in-depth.
Further, as was reported in IN 86-108, supplement 1, rapid and severe corrosion caused by boric acid leakage can occur before it is noticed.
The last two sentences [in the sixth paragraph of the This information was provided to help indicate Background section) discuss manual NDE and do not why the Staff did not pursue immediate action in relate to the remainder of the paragraph. The merits 1991.
of manual NDE and automatic tooling are not the subject of the draft generic letter.
The purpose of the EPRI NDE demonstration was not to Per the staff's discussions with EPRI, the qualify tooling or operators, but was limited to the demonstration was not intended to qualify each demonstration of an inspection system's ability to operator, but rather to qualify the system, i
detect and size defects.
which included the operator and equipment.
- s.
GL %-XX, Degradation of CRDMs & Other VHPs (61 FR 40253) Comment Resolution s
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Comment Resolution
[The seventh paragraph of the Background section]
NUREG/CR-6245 described the basis for not appears to justi,fy the draft generic letter based on requiring inspections until such time as methods advances in inspection techniques rather than assess and equipment could be developed to minimize the safety significance of PWSCC.
This implies that radiation exposure.
inspections should be required because industry has voluntarily developed improved inspection methods.
The paragraph should focus on safety concerns.
The description of the Zorita event could be more The Zorita event was described in IN 96-11 and precise.
in Westinghouse NSAL-94-028, and further detail The Zorita concern was primarily with the response of is beyond the scope of this GL.
sensitized material attacked by reduced sulfur species.
Inspections at the Zorita plant did not identify circumferential cracks in the J-groove weld, but found a through-wall crack at or near the bimetallic weld.
In the third sentence (of the eight paragraph in the Corrected in final draft of GL.
Background section), " resin bed" should be " resin bead."
The text would be better understood if the corrected in final draft of GL.
measurements were provided in English as well as metric units, i.e.,
" liters" and " gallons."
It is our understanding that the NRC Staff has Zorita The Westinghouse staff notified the WOG plants, resin intrusion reports and data that are not publicly the B&WOG plants, and the CEOG plants of the available. it is difficult to assess the significance Zorita incident by issuing NSAL-94-028.
The of the Zorita resin intrusion without all available Staff does not have any information beyond that information.
In previous communications with the NRC described in the draft GL and in IN 96-11.
Staff, we have been told that these reports have been Specifically, as was stated in the December 1, provided to all PWR Owners Groups.
However, inquiries 1994, m'eting summary (dated December 20, 1994):
made to the PWR Owners Groups have not supported this.
"3. The Zorita data has been sent to the We request the NRC Staff to place all information on owners groups. The industry should the Zorita resin intrusion into the Public Document provide an assessment of the Zorita Room, and provide the opportunity for industry to experience and its implications with evaluate.
regard to their previously submitted stress analysis and safety evaluations."
To maintain the chronological order of events, the 9th These paragraphs are grouped by subject matter, and loth paragraphs (in the Background section] should be switched.
9-GL 96-XX. Degradation of CRDMs & Other VHPs (61 FR 40253) Commens Resolution
Comment Resolution l
The draft generic letter does not discuss the recent The GL was drafted before the staff had been VHPs reinspections performed at Oconee and D.C. Cook, informed of the results of the recent nor the VHP repair at D.C. Cook.
reinspections; relevant information regarding
]
the results of the re-inspections and the repair i
at D.C. Cook will be included in the final GL.
The NRC states that they have not been provided with Hence the need for the information requested by y
the WOG resin intrusion review.
IN 96-11 does not the draft GL.
The NRC staff had previously require any specific action by licensees.
requested information from the industry during Furthermore, Westinghouse NSAL-94-028 did not request the several meetings referenced in the craft GL.
j licensees to provide a response back to Westinghouse specifically, as was stated in the December 1, and no WOG report has been prepared.
1994, meeting summary (dated December 20, 1994):
"2. The NRC considers [PWSCC) in CRDM's a generic issue.
The NRC staff requests industry to provide an integrated inspection plan.
This should include
]
schedules for [PWRs] that have not
~
completed the CRDM penetration inspections, criteria for reinspections, and scope and method of inspection. The voluntary ad hoc approach suggested by NEI is undesirable in that plant specific submittals and reviews of this generic problem would be resource intensive both for the NRC and industry."
i The citation of Westinghouse, Framatome Technologies, corrected in final draft of GL.
and Combustion Engineering are incorrect. The citations should be the PWR Owners Groups; i.e.,
- WOG, j
B&WOG, and the CEOG.
l i
i i
.I i
~
GL 96-XX. Degradation of CRDMs A Other VHPs (61 FR 40253) Comment Resolution
. t o.
Comment Resolution (The fourteenth paragraph in the Background section]
This interpretation is based on both the content states that "(t)he program outlined in the NEI white of the White Paper and on comments made by NEI paper is based on the assumption that the issue is an during the various referenced meetings with the economic one'rather than a safety issue,..." and that Staff. The Staff disagrees with NEI that this the NRC Staff did not agree that the issue was only is not a safety issue, and continues to believe economic. This is not a correct interpretation of the that an integrated, long-term program, which NEI white paper. The white paper documents the includes periodic inspections and monitoring, is extensive safety evaluations developed by the PWR necessary.
Owners Groups which addressed all the safety concerns Further, it is difficult to come to a judgement identified by the NRC Staff.
The method discussed in on this matter since NEI has not provided the the white papdr to manage RPV head penetration various PWR Owners Groups ranking models to the cracking acknowledges that the issue is not an Staff.
immediate safety concern and that leak-before-break will occur. Using this knowledge, the management methodology discussed provides a four step approach; of which one step evaluates the economic considerations.
The sentence [in the first paragraph of the Discussion It is the Staff's technical opinion that, given section) starting, "Further, if any significant..."
similar Zorita-type resin intrusions, residual is an absolute statement which has not been stresses would be sufficient to cause technically justified in his document nor the circumferential cracking. This is consistent references. It would be technically correct if the with the industry's explanation of the Zorita sentence was revised to read, "Further, if any experience in meetings with the Staff.
significant resin intrusions have occurred at U.S.
plants such as occurred at Zorita, the resultant chemistry condition in combination with stress may be significant."
The sentence [in the second paragraph of the The statement is factual and needs no revision.
Discussion section) which starts, " Cracking in the As the comment agrees, cracking in VHPs has VHPs
" is potentially misleading. While cracking occurred and, since PWSCC is an age related has occurred in 116 of the 5146 penetrations degradation mechanism, is expected to continue inspected, it has not been observed in the large to occur as plants age, majority of VHPS.
PWSCC is an age related degradation mechanism which could occur some time in the future, many years beyond the initial or renewed license or never.
. GL 96-XX Degradation of CRDMs & Other VHPs (63 FR 40253) Commera Resolution
Comment Resolution The (second paragraph of the Discussion section]
The industry's comment is that compliance with states that the NRC Staff considers the cracking of ASME Code margins of safety, as required by 10 VHPs to be a safety concern for the long-term based on CFR 50.55a, and elimination of defense-in-depth, the possib,ility of (1) exceeding the American Society are less severe than the issue elevated in the of Mechanical Engineers (ASME) Code for margins if the owners groups safety evaluations; nonetheless, cracks are sufficiently deep and continue to propagate they represent compliance with Federal during subsequent operating cycles, and (2)
Regulations.
eliminating a layer of defense in depth for plant safety.
These safety goncerns are addressed by the PWR Owners Groups safety evaluations. These were summarized on Page 10 of NUREG/CR-6245, " Assessment of Pressurized I
Water Reactor control Rod Drive Mechanism Nozzle 1
Cracking," which states that "There are two major safety concerns associated with CRDM nozzle cracking.
First, a crack could eventually lead to a rupture of the nozzle and, if the nozzle is severed, to ejection of the connected CRDM housing. Second, a through-wall crack would allow the borated reactor coolant to come in contact with the vessel head and cause boric acid corrosion of the low-alloy steel base metal..." In addition, the NRC Staff's safety evaluation, dated November 19, 1993, states that "The primary safety concern aseociated with stress corrosion cracking in Alloy 600 is the potential for circumferential cracks.
Extensive circumferential cracking could lead to 4
ejection of a CRDM..."
Since the EWR Owners Groups safety evaluations evaluated a through-wall crack and ejection of the connected CRDM housing, it appears that the two long term concerns identified by the draft GL are less severe than those already evaluated.
The concept of scheduling augmented inspections is The Staff is collecting information to determine inconsistent with the concept of "long term safety if augmented inspections need to be required concerns." Given that technical safety concerns have pursuant to 10 CFR 50.55a(g)(6)(ii).
been addressed, requesting a " technical basis" for a schedule is unclear.
The required information is unnecessarily prescriptive Corrected in final draft of GL.
(e.g.,
the direction of inspection (top or bottom) will not affect the quality of an inspection which a licensee may choose to perform, the presence of thermal sleeves, etc.)
i
. GL 96-XX, Degradation of C*.DMs & Other VHPs (61 FR 40253) Comment Resolution
i comment Resolution The first sentence [in paragraph 2 of the Required The industry has indicated that a model has been Information section] states, "... include the developed and is being applied to determine if susceptibility ranking of your plant and the factors and when inspections are appropriate. The Staff used to determine this ranking." This phrase is is requesting the method and results of the redundant with the first part of the sentence which licensee's evaluation. Requesting the
- states, "A description of the evaluation methods and susceptibility ranking and the factors used to results used to assess the susceptibility of the CRDM determine this ranking is a very different and other VHPs in your plant to PWSCC, request.
The susceptibility models were not used as input to The susceptibility models are being used by the the PWR Owners Groups' safety evaluations that were industry to justify establishing vessel head submitted and approved by the NRC Staff. The penetration inspection schedules.
If they are susceptibility models and subsequent rankings may be "not sufficiently precise to be used in a safety used by licensees to make economic evaluations, but assessment," then how can they be precise enough are not sufficiently precise to be used in a safety to make a determination of when to inspect? The assessment that may be submitted to the NRC Staff.
In Staff wants to review the models and monitor addition, it is unclear how the NRC Staff will use their performance in order to assess their such models to evaluate a safety concern, precision and reliability.
i i
4 3
4 GL 96-XX Degr=hma of CRDMs & Other VHPs (61 FR 40253) Comment Resolution m+--*.
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s Comment Resolution This requested information implies that the GL 88-05 It is the Staff's understanding that the e
visual inspection is inadequate to detect boric acid cracking at Bugey-3 was detected during a deposits and which could be caused by PWSCC.
This hydrotest, but not during normal operations, implication is not supported by operating history and and the one at Zorita was detected during a safety evaluations:
refueling outage because of a buildup of boric The only through-wall VHPs cracks (Bugey and Zorita) acid crystals, e
As was reported in IN 86-108, Supplement 1, were detected by visual inspections.
e GL 88-05 visual inspections are considered rapid and severe corrosion caused by boric e
acceptable for detecting PWSCC in the remainder of acid leakage can occur for before it is the reactor coolant system.
noticed. While the GL 88-05 program will find A conservative definition for "long term safety gross leakage some period of time after it e
concern" implied by NUREG-CR-6245 would infer a began, the concern of this GL is whether minimum of nine years after the initiation of a licensees are maintaining defense in depth and i
Boric acid deposited over Code margins and are able to discover cracking this time period would be readily observed using the before it becomes severe enough to be noticed GL 88-05 visual inspections, under a GL 88-05 program. Further, standard Technical Specifications do not allow pressure boundary leakage, As stated in the NRC Staff's safety e
]
evaluation, dated November 19, 1993:
i "Once a leak start, B&WOG concluded that it would take 6 years before enough corrosion a
would occur to reduce the wall thickness of the reactor vessel head td below ASME code j
minimums..."
The Staff agreed that "an immediate safety [is not present) as long as the surveillance walkdowns required continue and as long as any leakage is corrected." The staff did not make a determination as to how long it would take for a leak to start, nor how long it would take to take to reduce the wall thickness of the reactor vessel head to below ASME code minimums. However, it should be noted that relying solely upon detection of leakage, when I
most facilities
- technical specifications i
state that no pressure boundary leakage is acceptable, is not consistent with maintaining defense-in-depth.
OL 96-XX, Degradataon of CRDMs & Other VHPs (61 FR 40253) Comment Resolution l l
o Comment Resolution The intergranular stress corrosion cracking resulting The Staff agrees that resin intrusion can only from a Zorita type resin intrusion is a different be identified through review of water chemistry mechanism than the primary water stress corrosion history. Further, the vast majority of cracking PWSCC). The resin intrusion cracking is a inspections that might identify problems degradatio(n mechanism caused by an abnormal operating resulting from the resin intrusions were not in event and is not a age-related degradation mechanism U.S. plants. The Staff requires the requested like PWSCC.
Furthermore, the predictive tools for information to determine if an augmented j
PWSCC are not capable of predicting resin intrusion.
inspection program is required.
i It is noted that the VHP inspections performed on over 5200 penetrations at 87 plants worldwide did not identify any other plant that exhibited intergranular stress corrosion cracking similar to that exhibited at Zorita.
The draft generic letter has not provided a basis for supplying information on chlorides, fluorides, oxygen, boron, or lithium. The Zorita experience has been linked to the sulfates, but to our knowledge the other chemistry species have not been linked.
Southern Nuclear Operating Company is in total See responses to NEI comments, agreement with the NEI comments which are to be provided to the NRC.
Georgia Power Company is in total agreement with See responses to NEI comments.
the NEI comments which are to be provided to the NRC.
See Comment #4 See responses to NEI comments.
By leuer dated September 30,1996, comments from Dave blorey, Vice President, Farley; Southern Nuclear Operating Co; PO Box 1295; Birmingham, AL 35201 7
By leuer dated October I,19%. comments from C. K. McCoy, Vice President, Vogtle; Georgia Power Company; 40 Iverness Parkway; PO Box 1295; Birmingham, AL 35201 8
By letter daico October 2,1996, comments from W. H. Bohlke, Vice President; Florida Power & Light Co.; P.O. Box 14000; Juno Beach, FL 33408
. GL 96-XX, Degradation of CRDMs & Other VHPs (61 FR 40253) Comment Resolution M
j s
Comment Resolution a
TU Electric has reviewed and endorses the NEI letter See responses to NEI comments.
[ dated October 3, 1996]. TU Electric agrees with the NEI discussed issues, recommendations and rationale.
TU Electric further agrees that no technical or regulatory' basis exists for this generic comunication and recommends that NRC provide due consideration of the NEI comments in the final evaluation of the proposed generic ~ communication.
i DLC endorses the comments provided by the Nuclear See responses to NEI comments.
Energy Institute (NEI).
The NEI comments identify the i
key issues which need to be considered. DLC concurs with NEI that there is no technical or regulatory basis for this generic letter.
'2 We have reviewed the NRC proposed generic letter and See responses to NEI comments.
fully endorse the comments provided by the Nuclear Energy Institute in their letter dated September 30, 1996.
i 4
5 i
By letter dated October 3,1996, comments from J. S. Marshall, General Licensing Manager: TU Electric; PO Box 1002; Glen Rose. TX 76043 By leuer dated October 4,1996, comments from Sushil C. Jain, Division Vice President; Duquesne Light Conrany; Beaver Valley Power Station; PO Box 4; Shippingport, PA 15077 12 By leuer dated Ocsober 4,1996, comments from M. R. Kansler, Vice President; Virginia Power,5000 Dominion Bouleverd; Glen Allen. VA 23060 GL 96-XX Degradation of CRDMs & Odur VHPs (6i FR 40253) Comment Resolution,
ENCLOSURE 2 DETAII ED COMMENTS ON THE DRAFT GENERIC Li.re ER CMTO' SMmON PARAGSAPH COMMENT CORRECTIONS 1.
General It is unclear why the draft generic letter (GL) did not reference It would be beneficial to contrast the nor discuss in detail the evaluations and conclusions contained conclusions of NUREG/CR-6245 to the draft in NUREGICR-6245. This document provides a balanced GL*s definition of"long-term safety evaluation of the s ifesy evaluations performed by the PWR concerns."
Owners Groups.
2.
General The phrase "other vessel head penetrations" used throughout Additional clarity may be gained if the the draft generic letter should be clarified to read "other reactor phrase "other vessel head penetrations"is vessel closure head penetrations" altered to read "other reactor vessel closure head penetrations."
===3.
Background===
lat Figure 1 and text appear to only discuss CRDMs designed by A description of the CE and B&W Westinghouse.
penetration design features would be beneficial.
===4.
Background===
3rd The first sentence states that in 1989 the emerging issue was A chronologically sequenced paragraph identified, then the second sentence states leakage has occurred would be easier to understand.
since 1986. This is not chronologically correct and is confusing.
===5.
Background===
4th The Bugey-3 cracking was discovered in September 1991.
Proper dates should be used.
===6.
Background===
4th In the second sentence, it states that the Japanese have Provide a reference or delete Japan from
" uncovered" VIIPs with cracks. We are unaware of any the list of countries which have identified available reference stating that the Japanese have detected PWSCC in their VHPa.
PWSCC cracks in their VHPs. A source reference should be provided.
Improved clarity would be achieved if the word " uncovered" is changed to " detected."
It would be more precise to state that cracks were " detected
- rather than " uncovered"in this paragraph and throughout the document's text.
===7.
Background===
6th Sub-item (3). NUREG/CR-6245 states that the leakage would be The draft GL and NUREG/CR.6245 would detected "long" befbre significant damage to the reactor vessel be consistent if the word "long" was added head would occur.
before the word "before."
===8.
Background===
6th The last two sentences discuss manual NDE and do not relate to Delete the last two sentences from the the remainder of the paragraph. The merits of manual NDE paragraph.
and automatic tooling are not the subject of the draft generic letter.
===9.
Background===
7th The purpose of the EPRI NDE demonstration was not to qualify The EPRI activities would be better tooling or operators, but was limited to the demonstration of an described if the term " qualification" was inspection system's ability to detect and size defects.
modified.
10.
Background
7th This paragraph appears to justify the draft generic letter based Delete this paragraph.
on advances in inspection techniques rather than assess the safety significance of PWSCC. This implies that inspections
-__-_ _______=___ ___ _-. _ _ _ _ _ _ _
w a.
'~
Co?aECTIONS should be rzquired beccuse indu:try h s vnluntarily davsloped improved inspection mzthods. % pirigraph chould focus on safety cuncarna.
I 1.
llackground 8th
% description of the Zorita event could be more precise.
A more precise statem'ent would be:
"During the 1994 outage at Zorita (a Spanish reactor), visual inspection of the reactor versel head discovered boron deposits on a single vessel head penetration.
A more thorough inspection of this penetration detected a crack approximately two inches below the bimetallic weld. An extensive investigation and root cause evaluation were performed. It was determined that the indications were caused by intergranular stress corrodion cracking initiated by cation resin intrusion."
12.
Background
8th h Zorita concern was primarily with the response of sensitized It would be more precise to refer to " attack material attacked by reduced sulfur species.
by sulfur species on sensitized materiale?
13.
Background
8th First sentence. Inspections at the Zorita Plant did not identify Nee changes would provide factual circumferential cracks in the J. groove weld, but found a clarity.
through. wall crack at or near the bimetallic weld.
In the third sentence. " resin bed" should be " resin bead."
% text would be better understood if the measurements were provided in English as well as metric unita, i.e.," liters" and "sallona?
14.
Background
9th It is our understanding that the NRC staff hae Zorita resin Related reports and data should be made intrusion reporta and data that are not publicly available. It is available.
difficult to assess il e significance of the Zorita resin intrusion without all available information. In previous communications with the NRC staff, we have been told that these reports have been provided to all PWR Owners Groups. However, inquiries made to the PWR Ownere Groups have not supported this. We request the NRC staff to place all information on the Zorita resin intrusion into the Public Document Room, and provide the ertunity for industry to evaluate.
15.
Background
9th and 10th To maintain the chronological order of events. the 9th and 10th Chronological order of these paragraphs paragraphe should be switched.
would be beneficial.
16.
Background
10th
% draft generic letter does not discuss the recent VH Ps re-It would be beneficial to document the most inspections performed at Oconee and D C. Cook, nor the VilP recent inspection activ; ties and results.
2 MI M
CO" ECTIONS rep-ir rt D C. Cook.
17 llackground loth The NRC(tites th;t they hiva not been provided with the WOG The statem:nt in thi2 pir grrph should r; sin intrusion review. IN 96-11 doe 3 not require rny specific refl:ct the comm:nt.
action by licensees. Furthermore, Westinghouse NSAl,94 028 did not request licensees to provide a response back to Westinghouse and no WOG report has been prepared.
18.
Itackground 13th The citation of Westinghouse, Framatome Technologies, and Use the correct citations.
Combustion Engineering are incorrect. The citations should be the PWR Owners Groups;i.e., WOG, B&WOG, and the CEOG.
19.
Background
11 This paragraph states that "(t)he program outlined in the NEI An appropriate statement would reflect the white paper is based on the assumption that the issue is an Section Vll white paper text.
economic one rather than a safety issue,.." and that the NRC staff did not agree that the issue waa only economic. This is not It is the NRC staffs prerogative to disat,ree a correct interpretation of the NEl white paper. The white with positions taken in the white paper.
paper documents the extensive safety evaluations developed by llowever, the NRC staff should identify the PWR Owners Groups which addressed all the safety those safety concerns have not been concerne identified by the NRC staff. The method discussed in addressed by the NRC approved PWR the white paper to manage RPV head penetration cracking Owners Groups safety evaluations.
acknowledges that the issue is not an immediate safety concern and that leak-before-break will occur. Using this knowledge, the management methodology discussed provides a four step approach; of which one step evaluates the economic considerations.
20.
Discussion lat The sentence starting," Further,if any significant. "is an This change provides clarity.
absolute statement which has not been technically justified in this document nor the references. It would be technically correct if the centence was revised to read,"Further, if any significant resin intrusions have occurred at U.S. plants such as occurred at Zorita, the resultant chemistry condition in combination with stress may be significant."
21.
Discussion 2nd The sentence which starts," Cracking in the VIIPs.." is A more precise statement would be potentially misleading. While cracking has occurred in i16 of
" Cracking occurred in a few VilPs and could the 5146 penetratione inspected, it has not been observed in the occur in others at some future time. An large majority of VIIPs. PWSCC is an age related degradation existing crack may continue to grow, but mechanism which could occur some time in the future, many could stop."
years beyond the initial or renewed license or never.
22.
Discussion 2nd The paragraph states that the NRC staff considers the cracking The PWR Owners Group safety evaluations of VilPs to be a safety concern for the long-term based on the addressed the safety concerns identified by possibility of(1) exceeding the American Society of Mechanical NRC staff.
Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth s
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Coansenous 7dl in l"\\ If
.Il for piznt s fety.
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These safety concernura eddr: seed by the PWR Owners Groups safety evaluations. These were summarized on Page 10 of NUREGICR-6245," Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking." which states that"There are two major s afety concerna associated with CRDM nozzle cracking. First, a crack could eventually lead to a rupture cf the nozt.e a.nd, if the nozzle is severed, to ejection of the connected CRDM housing. Second, a through. wall crack would allow the borated reactor coolant to come in contact with the vessel head and cause boric acid corrosion of the low-alloy eteel base metal..." In addition, the NRC staff's safety i
evaluation dated November 19,1993, states that "The primary safety concern sesociated with strees corrosion cracking in Alloy 600 is the potential for circumferential cracks. Extensive circumferential cracking could lead to ejection of a CRDM..."
Since the PWR Owners Groups safety evaluatione evaluated a through. wall crack and ejection of the connected CRDM housing, it appears that the two long term concerna identified by the draft GL are less severe than those already evaluated.
23.
Required 1.2.a The concept of scheduling augmented inspections is inconsistent Provide clarity.
Information with the concept of"long term safety concerns." Given that technical safety concerne have been addressed, requesting a
" technical basia" for a schedule is unclear.
24.
Required 1.2.h The required information is unnecessarily prescriptive (e.g., the Delete as this level of detail is not Information direction ofinspection (top or bottom) will not affect the quality necessary.
of an inspection which a licensee may choose to pertarm, the presence of thermal sleeves, etc.)
25.
Required 2.
The first sentence states "... include the susceptibility ranking Delete the phrase ".. include the information of your plant and the factore used to determine this ranking."
ausceptibility ranking of your plant and the This phrase is redundant with the first part of the sentence factore used to determine this ranking."
which states,"A description of the evaluation methods and resulta used to assess the susceptibility of the CRDM and other VHPs in your plant to PWSCC, 26.
Required 2.
The susceptibility models were not used as input to the PWR Since it is not possible to make a safety Information Owners Groups safety evaluations that were submitted and determination.with the susceptibility approved by the NRC staff. The susceptibility models and rankings, this paragraph should be deleted.
subsequent rankings may be used by licensees to make economic 4
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evtluati:ni, but cre not cufficiently precise to be used in a i
safety casessmint that m:y be submitted to the NRCct:tff. In addition, it is unclear how the NRC staff will use such models to evaluate a safety concern.
j 27.
Required 2.
information This requested information implies that the GL 88-05 visual j
inspection is inadequate to detect boric acid deposits and which Boric aci'd deposits will be identified by the visual inspections recommended in Generic could be caused by "WSCC. This implication is not supported Letter 88-05.
3 by operating history and safety evaluations:
The only through wall VHPs cracks (Bugey and Zorita) were i
detected by visual inspections.
GL 88-05 visual inspections are considered acceptable for detecting PWSCC in the remainder of the reactor coolant system.
t A conservative definition for "long term safety concern" implied by NUREG CR 6245 would infer a minimum of nine years after the initiation of a PWSCC through wall leak-Boric acid deposited over this time period would be readily obse.md using the GL 88-05 visual i :::tions.
28.
Required 3.
The intergranular stress corresson cracking resulting from a The resin induced intergranular stress Information Zorita type resin intrusion is a different mechanism than the corrosion cracking is different than the primary water stress corrosion cracking (PWSCC). The resin stated scope and should be deleted.
intrusion cracking is a degradation mechanism caused by an abnormal operating event and is not a age-related degradation mechanism like PWSCC. Furthermore, the predictive tools for PWSCC are not capable of predicting resin intrusion. It is noted that the VHP inspections performed on over 5200 penetrations at 87 plants worldwide did not identify any other plant that exhibited intergranular stress corrosion cracking similar to that exhibited at Zorita.
29.
Required 3.4 The draft generic letter has not provided a basis for supplying Delete.
i Information information on chlorides, fluorides, oxygen, boron, or lithium.
The Zorita experience has been linked to the sulfates, but to our knowledge the other chemistry species have not been linked.
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