ML20137W114
| ML20137W114 | |
| Person / Time | |
|---|---|
| Issue date: | 04/17/1997 |
| From: | Shapaker J NRC (Affiliation Not Assigned) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137W117 | List: |
| References | |
| GL-97-01, GL-97-1, TAC-M95280, NUDOCS 9704180036 | |
| Download: ML20137W114 (1) | |
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION j
WASHINGTON, D.C. 20666-0001 gs,
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April 17, 1997 MEMORANDUM TO:
Document Control Desk Document Management Branch Division of Information Support Services Office of Information, Resources Management FROM:
JamesW.Shapaker// M $t/
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Events Assessment 1and Generic Comm0nications Branch Division of Reactor' Program Management Office of Nuclear Reactor Regulation
SUBJECT:
DOCUMENTS ASSOCIATED WITH NRC GENERIC LETTER 97-01.
DEGRADATION OF CONTROL ROD DRIVE MECHANISM AND OTHER VESSEL CLOSURE HEAD PENETRATIONS (TAC No. M95280)
The Materials and Chemical Engineering Branch (EMCB) in the Divisison of Engineering (DE) prepared the subject generic letter, which was issued on April 1, 1997, and given accession number 9703260336.
There is material related to the subject generic letter that should be placed in the NRC Public Document Room and made available to the public.
Therefore, by copy of this memorandum. I am providing the following documents to the NRC Public Document Room:
(1) a copy of the published version of the subject generic letter, (2) a copy of the information paper (SECY-97-063) that was sent to the Commission, (3) a-copy of each letter received in response to the notice of oportunity for public comment on the proposed generic letter that was published in the Federal Register on August 1. 1996, (4) a copy of the summary and resolution of public comments that were received. (S) a copy of the CRGR review package.
I request that you provide me with the Nuclear Documents System accession number for this memorandum. This information may be provided by telephone (415-1151) or by e-mail (JWS).
In addition..please modify the ap3ropriate -
NUDOCS entries to reflect the fact that the documents identified lerein are related to Generic Letter 97-01.
Attachments:
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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001
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April 1,1997
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i NRC GENERIC LETTER 97-01: DEGRADATION OF CONTROL ROD DRIVE MECHANISM NOZZLE AND OTHER VESSEL CLOSURE HEAD PENETRATIONS i
Addressees
' Ali holders of operating licenses for pressurized water reactors (PWRs), except those who.
have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
)
Purpose 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 1
addressees provide to the NRC a written response to the requested information. The information requested 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 whether an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required.
j Backaround 1
i Primary Water Stress Corrosion Cracking of Vessel Closure Head Penetrations l
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
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CRDM is screwed and seal-welded onto the top of the nozzle penetration, as shown in Figure 1, (Figure 1 is for illustrative purposes only and is not intended to be indicative of every nuclear steam supply system (NSSS) vendor's CRDM design.) The weld between the nozzle top and bottom pieces is a dissimilar metal weld, which is also called a bimetallic l
weld. The nozzles protrude below the vessel head, thus exposing the inside surface of the nozzles to reactor coolant. The CRDM nozzle and other VHPs are basically the same for all PWRs worldwide, which use a U.S. design (except in Germany and Russia). The areas of interest for potential cracking are the weld between the nozzle and reactor vessel head, and the portion of the nozzle inside the reactor vessel head above the nozzle-to-vessel weld.
Generally, there are 36 to 78 nozzles distributed over the low-alloy steel head. The vessel head is semi-spherical and the hwd panatrations are vertical so that the CRDM nozzle and other VHPs cre not perpendicular to the 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.
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GL 97-01 April 1,1997 e
Page 2 of 10 Beginning in 1986, leaks have been reported in several Alloy 600 pressurizer instrument 1
nozzles at both domestic and foreign reactors from several different NSSS vendors. The NRC staff identified primary water stress corrosion cracking (PWSCC) as an emerging technical issue to the Commission in 1989, after cracking was noted in Alloy 600 pressurizer heater sleeve penetrations at a domestic PWR facility. The NRC staff reviewed th7 safety significance of the cracking that occurred, as well as the repair and replacement a 'ivities 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 opportunity to take corrective action before a penetration would fail. Further, with the exception of the leak found at Bugey 3 during hydrostatic testing, the NRC staff is not aware of any failure of an Alloy 600 vessel closure head penetration during plant operation. The NRC staff issued Information Notice (IN) 90-10, " Primary Water Stress Corrosion Cracking (PWSCC) of inconel 600," dated February 23,1990, to inform the nuclear industry of the issue.
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in September 1991, cracks were found in an Alloy 600 VHP in the reactor head at Bugey 3, a French PWR. Examinations in PWRs in France, Belgium, Sweden, Switzerland, Spain, and Japan were performed, and additional VHPs with axial cracks were detected in several
,i European plants. About 2 percent of the VHPs examined to date contain short, axial cracks.
Close examination of the VHP that leaked at Bugey 3 revealed very minor incipient secondary circumferential cracking of the VHP. European and Japanese utilities have taken steps to detect and mitigate the PWSCC damage and to detect the leakage at an early stage.
European and Japanese utilities have inspected most of the CRDM nozzles and repaired the nozzles or replaced the vessel heads as appropriate. In Japan, the three most susceptible vessel heads are being replaced, even though no cracks were found in the nozzles of these heads. In France, Glecricit6 de France (EdF) is planning on replacing all vessel heads as a preventative measure. Inservice inspection of the upper head is now required in Sweden.
Removable insulation on the vessel head and leakage monitoring systems are installed at French and Swedish plants for early detection of leakage.
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 (CEOG) on March 25,1992, and the Babcock & Wilcox Owners Group (B&WOG) 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
GL 97-01 April 1,1997 Page 3 of 10 were held with NUMARC/NEl 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 avai;able 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 concem. 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 significant damage to the reactor vessel closure head would occur. In addition, the NRC staff had concerns related to unnecessary occupational radiation exposures associated with eddy current or other forms of nondestructive 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 remotely 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.
Intergranular Attack of CRDM Penetration Nozzle at Zorita In 1994, circumferential intergranu!ar attack (IGA) associated with the weld between the inner surface of the reactor closure head and the CRDM penetration (usually referred to as the J-grove 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 bead intrusions, which resulted in high concentrations of sulfates. Zorita has 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 and 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 [10.57 U.S. gallons) of cation resin entered the reactor coolant system (RCS). In September 1981, a mixed bed demineralizer screen failed and between 200 to 320 liters (52.83 to 84.54 U.S.
gallons) of resin entered the RCS. The coolant conductivity remained high for at least 4 months after the ingress. The increcse in conductivity was attributed to locally high
J GL 97-01 April 1,1997 Page 4 of 10
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concentrations of sulfates. 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 IN 96-11, " Ingress of Demineralizer Resins increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 'l996, 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.
Westinghouse notified the WOG plants, the B&WOG plants, and the CEOG plants of the l
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.
i Westinghouse further reported that U.S. plants monitor RCS conductivity on a routine basis, follow the EPRI guidelines on primary water chemistry, and monitor for sulfate three times a week. Westinghouse concluded that no immediate safety issue is involved and that the conclusions in its CRDM safety evaluation remain valid. Westinghouse 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 i
NRC staff does not have sufficient information to ascertain whether any significant primary system resin bead 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 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 ca.7not accurately size defects that are less than one mit deep (0.03 mm). These indications may be associated with the original fabrication and may not grow; however, they will be reexamined during the next refueling outage. A limited examination of eight in-core instrumentation penetrations conducted at the Palisades plant found no cracking. An examination of the CRDM penetrations at the D. C. Cook plant in the fall of 1994 revealed three clustered indications in one penetration.
The indications were 46 mm [1.81 in.],16 mm [0.63 in.), and 6 to 8 mm (0.24 to 0.31 in.] in i
length, and the deepest flaw was 6.8 mm [0.27 in ] deep. The tip of the 46-mm [1.81 in.] flaw l
was just below the J-groove weld.
Virginia Electric and Power Company inspected North Anna Unit 1 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.
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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 l
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j GL 97-01 I
Apnli,1997 4
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l the VHP, and the location of the VHP on the head. Each time a plant's VHPs are inspected, j
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 3
staff was informed that Framatome Technologies, Inc. [FTl, 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 j
PWRs. At present, none of the PWR Owners Groups (i.e., WOG, B&WOG, or CEOG) has i
submitted its models and assessments to'the NRC staff for review.
1 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 m
'j PWR VHPs and describes how the indtstry is managing the issue. The program outlined in j
the NEl white paper is based on the assumption that the issue is primarily an economic 1
s rather than a safety issue, and describes an economic decisior' tool to be used by PWR licensees to evaluate the probability of a VHP developmg a cmck 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 NEl in the i
j several meetings listed above that it did not agree with NEl that the issue was primarily i
economic.
1 Discussion s
j The results of domestic VHP inspections are consistent with the February 1993 analyses by j
the PWR Owners Groups, the NRC staff safety evaluation report dated November 19,1993, t
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' r.nalyses, and the
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European experience, the NRC staff has determined that it is probable that VHPs at other plants contain similar axial cracks. Further, if any significant resin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses are sufficient to cause j'
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 concem. Further, the NRC staff recognizes that the scope and timing of inspections may vary for different plants depending on their individual i
susceptibility to this form of degradation. In the long term, however, degradation of the j
CRDM and other VHPs is an important safety consideration that warrants further evhluation.
The vessel closure head provides the vital function of maintaining reactor pressure boundary,
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j 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 concem for the long term based on the i
possibility of (1) exceeding the American Society of Mechanical Engineers (ASME) Code for j
margins if the cracks are sufficiently deep and continue to propagate during subsequent
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. operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, 4
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i GL 97-01 April 1,1997 Page 6 of 10 i
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 of VHPs, is necessary. This was the conclusion of the staff's November 19,1993, safety evaluation, which stated, in part,
" the staff recommends that you consider enhanced leakage detection by visually examining j
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 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.
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, CEOG, WOG, or some subset thereof), to take 3dvantage of inspection results from other plants that have similar susceptibilities. The NRC staff does not discourage such group actions but notes that such an integrated industry inspection program must have a well-founded technical basis that justifies tne relationship between the plants and the planned implementation schedule.
Reauested Information The information requested in item 1 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 whether an augmented inspection program of the weld between the penetration nozzle and reactor vessel head as well as the portion of the nozzle above the weld is required, pursuant to 10 CFR 50.55a(g)(6)(ii), while the information requested in item 2 relates to the occurrence of resin bead intrusion in PWRs, such as occurred at Zorita.
Within 120 days of the date of this generic letter, each addressee is requested to provide a written report that includes the following information for its facility:
GL 97-01 April 1,1997 Page 7 of 10 1.
Regarding inspection activities:
1.1 A description of allinspections of CRDM nozzle and other VHPs performed to the date of this generic letter, including the results of these inspections'.
1.2 If a plan has been developed to periodically inspect the CRDM nozzle and other VHPs:
a.
Provide the schedule for first, and subsequent, inspections of the CRDM nozzle and other VHPs, including the technical basis for this schedule.
b.
Provide the scope for the CRDM nozzle and other VHP inspections, including the total number of penetrations (and how many will be inspected), which penetrations have thermal sleeves, which are spares, and which are 1
instrument or other penetrations.
1.3 If a plan has not been developed to periodically inspect the CRDM nozzle and other VHPs, provide the analysis that supports why no augmented inspection is necessary.
1.4 In light of the degradation of CRDM nozzle and other VHPs described above, provide the analysis that supports the selected course of action as listed in either 1.2 or 1.3, above. In particular, provide a description of all relevant data and/or tests used to develop crack initiation and crack growth models, the methods and data used to validate these models, the plant-specific inputs to these models, and how these models substantiate the susceptibility evaluation. Also, if an integrated industry inspection program is being relied on, provide a detailed description of t"is program.
2.
Provide a description of any resin bead intrusions, 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:
2.1 Were the intrusions cation, anion, or mixed bed?
2.2 What were the durations of these intrusions?
2.3 Does the plant's RCS water chemistry Technical Specifications follow the EPRI guidelines?
1 Those licensees that have previously submitted the requested information need not resubmit it, but may instei.d reference the appropriate correspondence in their response to this Generic Letter.
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1 GL 97-01 Apnl 1,1997 Page 8 of 10 i
ll 2.4 Identify any RCS chemistry excursions that exceed the plant administrative limits for the following species: sulfates, chlorides or fluorides, oxygen, boron, and lithium.
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2.5 identify any conductivity excursions which may be indicative of resin intrusions.
Provide a technical assessment of each excursion and any followup actions.
I i j 2.6 Provide an 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 i
inspections.
i Reauired Response i !
l Within 30 days of the date of this generic letter, each addressee is required to submit a
,a written response indicating: (1) whether or not the requested information will be submitted
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and (2) whether er not the requested information will be submitted within the requested time
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period. Addressees who choose not to submit the requested information, or are unable to j
satisfy the requested completion date, must describe in ineir response any alternative course
,l of action that is proposed to be taken, including the basis for the acceptability of the t
proposed alternative course of action.
NRC staff will review the responses to this generic letter and if concems are identified,
- j affected addressees will be notified.
Address the required written reports to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20555, under oath or affirmation under the provisions of Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).
In addition, submit a copy to the appropriate regional administrator.
1 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 CRDM nozzles and other VHPs. For this 3
reason, the above time periods are allowed for the responses.
I 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.
(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.
GL 97-01 April 1,1997 Page 9 of 10 Backfit Discussion Under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), this generic letter transmits an information request for the purpose of verifying compliance w'th applicable existing regulatory requirements. Specifically, the requested information would enable the NRC staff to determine whether or not the licensees'
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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 low. The requested information is also needed to determine whether an augmented inspect-tion program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required.to maintain public health and safety.
Additionally, no backfit is either intended or approved in the context of issuance of this generic letter. Therefore, the staff has not performed a backfit analysis.
Federal Reaister Notification A notice of opportunity for public comment was published in the Federal Register l
(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.
Paperwork 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 collections were approved b' the Office of Management and Budget, approval number 3150-0011, which j
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 potentialimpact 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 information to be collected?
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GL 97-01 April 1,1997 Page 10 of 10 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, NEOB-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 d.isplays a currently valid OMB control number.
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.
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Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: Keith R. Wichman (301) 415-2757 E-mail: krw@nrc. gov Jaines Medoff (30i) 415-2715 E-mail: jxm@nrc. gov Lead Project Manager: C. E. Carpenter, Jr.
(301) 415-2169 E-mail: cec @nrc. gov Attachments:
- 1. Figure 1. Typical Control Rod Drive Mechanism Nozzle
- 2. List of Recently issued NRC Generic Letters i
i GL 97-01 April 1, 1997 4
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Page 1 of 1 3
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1 CRDM HOUSING
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TYPE 304 STAINLESS STEEL SEAL WELD l
CROM NOZZLE l
I ALLOY 000 l
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REACTOR VESSEL HEAD
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y STAINLESS STEEL CLADDING NOZZLE 3
ALLOY 182 BUTTERING 4
STAINLESS STEEL y
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FUNNEL GUIDE O
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Figure 1. Typical control rod drive mechanism nozzle.
Copyright the Minerals, Metals & Materials Society; reprinted with permission.
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Attachm:nt 2 GL 97-01 April 1,1997 Page 1 of 1 LIST OF RECENTLY ISSUED GENERIC LETTERS G:n:ric Date of Letter Subiect issuance issued To 95-06, CHANGES IN THE OPERATOR 02/31/97 ALL HOLDERS OF OLs SUPP.1 LICENSING PROGRAM (EXCEPT THOSE LICENSEES OF PERMANENTLY SHUTDOWN REACTORS WHO ARE NO LONGER REQUIRED TO UTILIZE LICENSED REACTOR OPERATORS) FOR NPRs j
96-07 INTERIM GUIDANCE ON 12/05/96 ALL HOLDERS OF OLs TRANSPORTATION OF AND DECOMMISSIONING STEAM GENERATORS FACILITIES WITH POSSESSION-ONLY j
LICENSES FOR PRESSURIZED-WATER NPRs 96-06 ASSURANCE OF EQUIPMENT 11/13/96 ALL HOLDERS OF OLs OPERABILITY Ab.D CONTAIN-FOR NPRs, EXCEPT FOR MENT INTEGRIT / DURING THOSE LICENSES THAT DESIGN-BASIS ACCIDENT HAVE BEEN AMENDED TO CONDITIONS POSSESSION-ON'.Y STATUS 96-05 PERIODIC VERIFICATION OF 09/18/96 ALL HOLDERS OF OLs DESIGN-BASIS CAPABILITY (EXCEPT THOSE LICENSES OF SAFETY-RELATED MOTOR-THAT HAVE BEEN AMENDED OPERATED VALVES TO POSSESSION-ONLY STATUS) OR cps FOR NPRs 96-04 BORAFLEX DEGRADATION IN 06/26/96 ALL HOLDERS OF OLs SPENT FUEL POOL STORAGE FOR NPRs RACKS 95-09, MONITORING AND TRAINING OF 04/05/96 ALL U.S. NUCLEAR SUPP.1 SHIPPERS AND CARRIERS OF REGULATORY COMMISSION RADIOACTIVE MATERIALS LICENSEES OL = OPERATING LICENSE CP = CONSTRUCTION PERMIT NPR = NUCLEAR POWER REACTORS e
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pol.,1CY ISSUE (Information)
March 18, 1997 SECY-97-063 f.QB:
The Commissioners ff.QM:
L. Joseph Callan Executive Director for Operations
SUBJECT:
PROPOSED NRC GENERIC LETTER: " DEGRADATION OF CONTROL R0D DRIVE MECHANISM AND OTHER VESSEL CLOSURE HEAD PENETRATIONS" PURPOSE:
To inform the Commission, in accordance with the guidance in the December 20, 1991, memorandum from Samuel J. Chilk to James M. Taylor regarding SECY 172, " Regulatory Impact Survey Report - Final," of the staff's intent to issue the subject generic letter. The generic letter requests holders of operating licenses for pressurized water reactors (PWRs), except those who have certified to a permanent cessation of operations, to describe their program for ensuring the timely inspection of PWR control rod drive mechanism (CRDM) and other vessel closure head penetrations (VHPs) and require that all addressees provide to the NRC a written response to the requested information.
The information requested in the generic letter 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 whether an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(11), is required. The generic letter also requests information related to the potential for resin bead intrusions, such as occurred at the Zorita plant in Spain.
A copy of the proposed generic letter is attached.
SUMMARY
In September 1991, cracks were found in an Alloy 600 VHP in the reactor head of Bugey 3, a French PWR. Subsequently, most foreign countries have implemented VHP inspection programs, and VHP cracking was detected in several European plants. About 2 percent of the VHPs examined to date contain short, axial cracks.
European and Japanese utilities have taken steps to detect and mitigate the PWSCC damage and to detect the leakage at an early stage at most of their plants.
European and Japanese utilities have inspected most of the CRDM nozzles and repaired the nozzles or replaced the vessel heads as CONTACT:
C. E. Carpenter, Jr.
NOTE:
TO BE MADE PUBLICLY AVAILABLE IN 5 (301) 415-2169 WORKING DAYS FROM THE DATE OF THIS PAPER It.
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The Commissioners '
i appropriate.
In Japan, the three most susceptible vessel heads are being replaced, even though no cracks were found in the nozzles of these heads.
]
flecricits de France (EdF) is planning on replacing.all vessel heads as a preventative measure.
Inservice inspection of the upper head is now required in Sweden, and replacement of the Ringhals 2 vessel head was planned.
4 Removable insulation on the vessel head and leakage monitoring systems are i
insta11ed'at French and Swedish plants for early detection of leakage.
l Since the discovery of cracking at Bugey 3, the NRC staff has met with the PWR Owners Groups [ Westinghouse Owners Group (WOG), the Combustion Engineering i
Owners Group (CEOG), and the Babcock & Wilcox Owners Group (B&WOG)] numerous times to discuss their respective programs for investigating PWSCC of Alloy i
600 and to assess the possibility of cracking of VHPs in their respective 1
plants.
In February 1993, the PWR Owners Groups submitted safety assessments' l
to.the NRC on this issue. After reviewing the industry's safety assessments, i
the safety significance of the cracking that occurred, as well as the repair j
and replacement activities at the affected facilities, 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 l
concern.
This conclusion was based on the cracks being predominately axial in i
i orientation and that circumferential cracking, which could lead to VHP i
severance, was not expected.
Frther, the industry had committed to perform j
inspections at three U.S. plants.
i During 1994, four U.S. plants conducted VHP examinations and two found indications.
For one of the units, it could not be shown that the cracking 4
would not exceed ASME Code margins of safety and repairs had to be made.
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.
The inspection at the Oconee Nuclear Generating Station in the fall of 1994 l
revealed 20 indications in one penetration. A limited examination of eight in-core instrumentation penetrations conducted at the Palisades plant found no cracking. An examination of the CRDM penetrations at the D. C. Cook plant in the fall of 1994 revealed three clustered indications.in one penetration, and i
repairs were needed.
Some high-stress areas were examined on each outer ring l
CRDM penetration at North Anna Unit I during its spring 1996 refueling outage, j
and no indications were observed.
j i
The NRC staff was informed that the PWR Owners Groups (i.e., WOG, B&WOG, CE0G) have developed a susceptibility model for VHPs and that all domestic PWRs have been modeled and the ranking has been given to each licensee. At present, none of the.PWR.0wners Groups has submitted its models'and assessments to the j-NRC staff for review.
l In 1994,. circumferential intergranular attack (IGA) was discovered in the CRDM j
penetrations at Zorita, a Spanish reactor. The IGA, a different degradation i
mechanism than the PWSCC described above, is believed to have resulted from i
two cation resin bead ingress events, which resulted in high concentrations of t
sulfates.
Four of-the active and 16 of the 17 spare CRDM penetrations at Zorita showed axial and circumferential stress corrosion cracking and IGA.
Westinghouse notified the PWROGs of the Zorita incident by issuing NSAL 028, which suggested that U.S. PWR plants review their RCS chemistry and other
?
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s l.
1-The Commissioners !
operating records pertaining to sulfur ingress events. The results of any 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 bead intrusions have occurred at any U.S. PWR.
i DISCUSSION:
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 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 i
circumferential intergranular stress corrosion cracking (IGSCC). Appropriate monitoring for degradation of the CRDM and oder VHPs is important in the long term based on the possibility of (1) exceedi,y the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code margins if the i
' cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant
{
- safety.
}
The NRC Staff is not establishing a new position in this generic letter. The proposed generic letter is a request for information from licensees.
It does not require that licensees perform inspections but it does request licensees 4
j to inform the NRC staff whether or not they plan to inspect and the basis for 4
these plans. The information requested in the generic letter will allow the staff to determine if planned industry actions provide reasonable assurance of 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. Additionally, the information requested l
related to the potential for resin bead intrusions, such as occurred at Zorita, is needed to ascertain whether the potential exists for U.S. plants to i
have more severe cracking than observed to date.
i A notice of opportunity for public coment was published in the federal i
Register (61 FR 40253) on August 1, 1996, and extended on August 22, 1996 (61 FR 43393). Coments were received from seven licensees, two industry l
organizations, and one Code Comittee.
Copies of the staff evaluation of these coments have been made available in the public document room. The coments on the proposed generic letter focused on (1) the need for the generic letter, (2) editorial comments, and (3) backfit justification. The staff has revised the generic letter to clarify certain areas.
i' The proposed generic letter had formal review waived by the CRGR prior to its issuance. for public coment.
Following consideration of public coments, the staff sent the proposed generic letter to the CRGR with a description of the l
- staff's response to public coments and of changes made to the proposed generic letter as a result of those comments. The revised proposed generic l
' letter was reviewed by the CRGR during Meeting 299 on January 28, 1997, and i
endorsed by the CRGR during Meeting No. 300 on February 4, 1997.
The staff incorporated coments made by the CRGR in those meetings.
j i
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The Commissioners The Office of the General Counsel has reviewed the proposed GL and has no legal objections.
The proposed generic letter is a " rule" for purposes of the Small Business Regulatory Enforcement Act (5 U.S.C., Chapter 8). The staff has received confirmation from the Office of Management and Budget that the generic letter is a non-major rule.
l The staff intends to issue this generic letter five working days after the j
date of this information paper.
g L. J eph Callan 4
Executive Director for Operations
Attachment:
Proposed NRC Generic Letter: " Degradation of Control Rod Drive Mechanism and Other Vessel Closure Head Penetrations" J
DISTRIBUTION:
Commissioners 4
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August 6,1996 4
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, l
August 1,1996). We respectively request that the comment period be extended from j
September 3,1996, to October 3,1996.
i The Nuclear Energy Institute, on behalf of the nuclear utility industry, will be developing comments on this draft generic letter for submittal to the NRC. We are
]
requesting this comment period extension to permit sufficient time for industry l
representatives to assemble and develop comments. This extension will also allow sufficient time to transmit our draft comments to all nuclear utilities for consideration y
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 at this time as long as surveillance walkdowns [ visual inspection] in accordance with GIc88-05 continue." Utilities'are continuing the walkdowns. Furthermore, recent utility volumetric inspections results confirm that the PWR Owners Groups safety evaluations considered in the NRC staff a SER remain valid.
We 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, i
cv Ralph. Beedle w
- REB /KOClead Mr. C.E. (Gene) Carpenter, Jr. (NRR/DRPE/PDI 1) c:
Mr. Brian W. Sheron (NRR/DE) 7,e e,n. sa m, se
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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-60-69 Washington, D.C.
20555-0001
References:
1.
Docket No. 50-285 2.
Federal Register Volume 61, No.149, dated August 1, 1996 (61 FR 40253)
Subject:
Coments on Proposed Generic Communication Regarding Primary Water 4
i Stress corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations The Omaha Public 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 l
of the initial CEOG susceptibility assessments for the CEDM nozzles at CEOG 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 comments 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 slow, result in 4
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.
4 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 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.
i 45 512s Empioyment wnh Equal opportunny
U. S. Nuclear Regulatory Commission LIC-96-0122 Page 2 Please contact me if you have any questions.
Sincerely,
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f'o'r T. L. Patterson Division Manager
. Nuclear Operations TCM/ tem c:
Winston & Strawn L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager W. C. Walker, NRC Senior Resident Inspector i
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' Ul fd* m m 27 2392 USA September 30,1996 Mr. David Meyer Chief, Rules Review and Directives Branch
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US Nuclear Regulatory Conunission
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Washington, DC 20555-0001 Subj: Proposed Generic Conununication: Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanisms and Other Head Penetrations C (61 Fed. Reg. 40253)
Dear Mr. Meyer:
Enclosed are conunents resulting from review by individual members of the ASME BPVC Subconunittee on Nuclear Inservice Inspection. This review is not to be construed as a position or opinion on the subject docunwnt by ASMEt rather, the enclosed conunents are submitted as a constructive public service, and represent the opinion of individual conunittee members.
Yours truly, George F ter, Secretar; ASME BPVC Subconunittee on 4
l Nuclear Inservice Inspection (212)705 8018 cc with encl.
J. Perry D. Landers D. Canonico T. Mawson G. Eisenberg l.
The American Society of Mechanical E n g in e e's
l Section XI Subcomrninee commmee:
address wnter care of: ABB Combustion Engineering 2000 Day Hill Rd.
C'=naan on draft PWSCC Generic Letter Windsor, CT 06095-1521 subject:
i 9
September 27,1996 Westinghouse Energy Center l
date:
copy to: P.O. Box 355
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Pittsburgh, PA 15230 0355 j
i Mr. George Fechter to:
ASME Codes and Standards We have reviewed the NRC " Proposed Generic Communication: Pnmary Water Stress Corrosion Crackmg of Control Rod Drive Mechanisms and Other Head Penetrations *(61 Fed. Reg. 40253) and l
have a number of comments to bring to their anention.
Members of the ASME Section XI subcommittee are concerned that additional inspections are being imposed on the industry for what appears to be a non-safety issue. It is our concern that these -
j additional inspection requirements are being justified based on the lack of action by Section XI on this subject. This is not the case at all.
1 i
j The Section XI Subcommittee has been monitoring this issue closely since the condition was first identified in France, and it has been kept up to date by regular briefings of the Executive Committee at each meeting (four times per year). Their conclusion to date on this issue is that there are no safety concerns that cannot be addressed by the regular inspections for boron deposits already required. To date we have received no requests to add additional inspection requirements for the head penetrations l
from any of our members, which include representatives of industry, utilities, national laboratories, j
and the NRC.
The primary concern of the Section XI Subcommittee, as with all other committees in the ASME Boiler and Pressure Vessel Cod:, is safety. The principle focus of the Section XI Subcommittee is 4
j that the integrity of the reactor coolant pressure boundary is maintained. Industry experience has indicated that cracking that has occurred in the control rod drive penetrations is not a safery concern.
j After more than 5000 penetrations have been inspected worldwide, only one penetration has been j
found to have a through-wall crack from PWSCC.
f i
The proposed generic letter comes as a surprise to the Subcommittee, and we respectfully request that i
this letter be forwarded to the NRC with our request that they remove any implications from the letter j
that Section XI has been ignoring this issue. Please address the NRC care of Mr David L. Meyer, Chief, Rules Review and Directives Branch, USNRC, Washington, DC 20555-0001.
Sincerely, 3
ANAds- ~
hL Owen Hedden Chairman Warren Bamford, Chairman Subcommittee on Inservice Inspection Subgroup on Evaluation Standards
@E S The American Society of m
Mechanical Engineers
'f' Keep ASME Codes and Standards Department Informed
i cc: James A. Perry, BNCS Gery M. Eisenberg, ASME i
Domenic A. Canonico, ABB-CE l
Gil Millman, USNRC Tom Mawson, Northeast Utilities Don Landers Teledyne 1
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Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Comminion j
Mail Stop T-6D49 l
Washington, DC 20555 0001 i
Subject:
Proposed Generic Communication; Pnmary Water Stress Corrosion Cracking 1
of Control Rod Drive Mechanism and Other Vessel Head Penetrations 1
(61 FR 40253, dated August 1,1996)
Notice of Oppormairy for Public Comment d
l On August 1,1996, the Nuclear Regulatory Comnussion published for public comment.
j.
" Proposed Generic Communication: Pnmary Water Stress Conosion Cracking of Control Rod i
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 conool rod drive mechamam and other vessel head penetrations and require that all addressees provide a written response to the NRC regarding this generic letter. These comments l
are submitted on behalf of Florida Power & Light (FPL), a licensed operator of two nuclear power plant units in Dade County, Flanda 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, 0*
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m gW. H. Bohlke Vice President Nuclear Engmeeting
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AlvADM 114e5 SAUDs ARAgia FAcsmats acri 315 5850 43 nuc ou amour 1204 GENEVA. SwT2E RLAND October 3,1996 VIA MESSENGER U.S. Nuclear Regulatory Commission Rules, Review and Directives Branch Two White Flint
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11545 Rockville Pike Rockville, MD 20852-2738 1
Re:
Co.nments on Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations; Proposed Generic Co:nmunication: 61 Fed. Reg.'40,253 (August 1,1996); 61 Fed.
Ree. 43.393 (Aueust 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 Backritting and Reform Group (NUBARG).1' These comments concem the i
backfming 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. { 50.109) in 1985, and which has closely monitored the NRC's application of the rule since that time.
-qfrt&O80lVr' wpg-
.WINSTON & STRAWN U.S. Nuclear Regulatory Commission October 3,1996 i
Page 2 i
ASME Code continue to be satisfied, and to ensure that the safety significance of VHP cracking 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 analysis because the proposed 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)." M. at 40,256.
In this regard, the Staff claims that it is "not establishing a new position for such compliance in this
- generic letter." M.
- We believe the Staff is required to perform a backfit analysis on the proposed imposition of the development of a plan to it'spect the CRDM and other vessel head penetrations.
Under 10 C.F.R. s 50.109, a backfit includes "the modification of or addition to... the procedures
... 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 change the... operation of a facility...." 49 Fed. Reg. 47,034,47,035 (1985). Imposition of this 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 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 where it is not clear whether a backfit will follow an information request. Those cases should be i
resolved in favor of analysis."). In this case, a backfitting analysis is required under 10 C.F.R. 50.109.
1 A backfitting analysis not only is legally requimi but also will ensure the protection.
of the public health and safety, as well as provide practical benefits to both licensees and tne NRC Staff. 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 downtime and adverse schedule impacts are avoided by licensees and that any resulting radiation exposure is
- assessed and minimized.
Sincerely, r-J Daniel F. Stenger Kathryn M. Sunon
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,,a 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 Medsnirai 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 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 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 NEl 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 technicalissues. NEI's members include all utilities licensed to operate commer :ial nuclear power plants in the United States, nuclear plant designers, major architect / engineering firms. fuel fabrication facihties, materials licensees, and other organizations and individuals involved in the nuclear energy industry.
~-
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Mr. D2,vid L. Meyer Pcge 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
.i 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 signi5cance of the vessel head 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 i
e Groups' safety evaluations, which considered the possibility of through v.all j
cracks.
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 -esin 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 new requesting licensees to submit information on this topic. provides detailed comments on the speciSc text of the proposed generic letter.
If you have questions concerning these comments, please contact Alex Marion (202-739 8080) or me.
Sincerely, u
Ralph E. Beedle TET/AM/ead Enclosures c:
C. E. (Gene) Carpenter, NRC/NRR Brian Sheron, NRC/NRR Jack Strosnider, NRC/NRR
i i
ENCLOSURE 1 j
GENERAL COMMENT
S ON THE DR/ FT GENERIC LETTER
\\
- 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 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 currently performing augmented inspections,
- 2. On Page 10 of NUREG/CR 6245, Assessment of Pressurized Water Reactor j
Control Rod Drive Mechanism Nozzle Cracking, it states,"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 cau.: boric acid corrosion of the low-alloy steel base meta!." In the NRC staff's saf y 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 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 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 specified in f 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 satisfied,...." These goals are unique and separate from the stated purpose of 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 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 identined in the Discussion section based on the following:
_.. - - -. - -. -... ~
- [
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 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
.MM9 imervice inspections. However, should inservice inspection identify indications, licensees will dispositiois them per the ASME Code.
I
' GDC 14 states, "The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of I
abnormal leakage, of rapidly propagating failure, and of gross rupture."
l Licensees no meeting GDC 14 because:
t 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 abnormalleakage. These i
evaluations determined that PWSCC would eitl.er be arrestec' ar would grow very slowly requiring 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-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 accordance with j
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 i
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 (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 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 i
C
- 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 i
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 i
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) 9611, " 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 hTC 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 9611 was issued. The extra burden on licensees to respond to Item 3 of the Required Information section is not justified.
1 i
i 3
l DETAILEp COMimEWis ON THE DRArr GENERIC LETTER ENCLOSUDE 2 -
CMT#
SEcrtoN PARAGRAPH Ce"M #r CORRECTIONS It is unclear why the draft generic letter (GL) did not reference 1
I-C'"*I
~
nor discuss in detail the evaluations and conclusions contained it would be beneficial to contrast the in NUREG/CR-6245. This document provides a balanced conclusions of NUREGICR-6245 to the draft evaluation of the safety evaluations performed by the PWR GL's definition of long-term safety concerns."
Owners Groups.
2.
Generag The phrase "other vessel head penetrations' used throughout Additionalclarity may be gained if the the draft generic letter should be clarified to read ~other reactor vessel closure head penetrations".
phrase *other vessel head penetrations" is altered to read "other reactor vessel closure head penetrations.'
3 BacWnd 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 bener ial.
c 4-BackEround 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 confusmg.
5-
Background
4th The Bugey-3 cracking was discovered in September 1991.
I%per dates should be used.
- 6. -
Background
4th in the *cend sentence. it states that the Japanese have Provide a reference or delete Japan from
" uncovered" VHPs with cracks. We are unaware of any the list of c6untries which have identified available reference stating that the Japanese have detected PWSCC in'their VHPs.
PWSCC cracks in their VHPs. A source reference should be provided.
Improved clarity would be achieved if the 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 NUREGICR.6245 would detected ~long" before significant damage to the reactor vessel be consistent if the word "long' was added t
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 frem 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 inspection system's ability to detect and size defects.
described if the term " qualification' was modified.
10.
Background
7th This paragraph appears tojustify the draft generic letter based Delete this paragraph
- on advances m m, opection techniques rather than assess the safety significance of PWSCC. This implies that inspections
CMT#
SECTION PA2ACRAPil COMMENT CORTECTIONS should be required because industry has voluntarily developed improved inspection methods. The paragraph should focus on Ra ety concerns.
I 1.
Itackground 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), visual inspection 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.
Ilackground 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.
Itackground 8th First sentence. Inspections at the Zorita Plant did not identify These changes would provide factual c reumferential cracks in the J-groove weld, but found a clarity.
i through-wall crack at or near the bimetallic weld.
In the third sentence," resin bed" should be " resin head."
The text would be better understood if the measurements were provided in English as well as metric units, i.e., " liters' and
" gallons.'
1d'
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 asaeas 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. Ilowever, inquiries made to the PWR Owners Groups have not supported this. 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 paragraphs should be switched.
Chronological order of these paragraphs would be beneficial.
16.
Background
10th The draft generic letter does not discuss the recent VilPs re-nspections performed at Oconee and D C. Cook, nor the VilP It would be beneficial to document the most recent inspection activities and results.
2
CMT#
SECTION Pr?ACRAPfl COMMENT CORRECTIONS rep"ir et D.C. Cook.
17*
Background
loth The NRC states that they have not been provided with the WOG resin intrusion review. IN 96-1I does not require any specific The statement in this paragraph should action by licensees. Furthermore. Westinghouse NSAI,94-028 reflect the comment.
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 Comhustion Engineering are incorrect. The citations should be Use the correct citations.
the PWR Owners Groups;i.e., WOG, B&WOG, and the CEOG.
19.
Background
14 This par graph states that "(t)he program outlined in the NEl w hite paper is based on the assumption that the issue is an An appropriate statement would reflect f.he economic one rather than a safety issue.. " and that the NRC Section VII white paper text.
staff did not agree that the issue was only economic. This is not a correct interpretation of the NEI white paper. The white it is the NRC staffs prerogative to disagree paper documents the extensive safety evaluations developed by with positions taken in the white paper.
the PWR Owners Groups which addressed all the safety llowever, the NRC staff should identify I
concerns identified by the NRC staff. The method discussed in those safety concerns have not been the white paper to manage RPV head penetration cracking addressed by the NRC approved PWR acknowledges that the issue is not an immediate safety concern Owners Groups safety evaluations.
1 and that leak.before-break will occur. Using this knowledge, the management methodology discussed provides a four step approach; of which one step evaluates ths - anomic considerations.
20-Discussion let The sentence starting. " Further, if any significant. " is an absolute statement which has not been te-hnically justified inn This change provides clarity.
this document nor the references. It would be technically correct if the sentence was revised to read, "Further, if any sigmficant 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.
Di8C"88i 2nd The sentence which stacts," Cracking in the VilPs. "is potentially misleading. While cracking has occurred in 116 of A more precise statement would be the 5146 penetrations inspected, it has not been observed in the
" Cracking occurred in a few VilPs and could large majority of V]IPs. PWSCC is an age related degradation occur in others at some future time. An mechanism which could occur some time in the future, many existing crack may continue to grow, but 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 VIIPs to be a safety concern for the long-term based on the possibility of(1) exceeding the American Society of Mechanical addressed the safety concerns identified by 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
CMTS SECTION PAMORAPH COGGREENT CO2RECTIONS for plant safety.
These safety concerns are addressed by the PWR Owners Groups safety evaluations. These were summarised 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 r 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 GI, 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 basis" for a schedule is unclear.
q 24' Required 1.2.b The required information is unnecessarily prescriptive (e.g., the l Delete as this level of detail is not Information direction ofinspection (top or bottom) will not affect the quality neassary.
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,usceptibility ranking Delete the phrase "... include the Information ofyour plant and the factors used to determine this ranking."
This phrase is redundant with the first part of the sentence susceptibility ranking of your plant and the -
which states,"A description of the evaluation methods and factors used to determine this ranking."
results used to mesees 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 Information Owners Groups safety evaluations that were submitted and Since it is not possible to make a safety approved by the NRC staff. The susceptibility models and determination with the susceptibility -
subsequent rankings may be used by licensees to make economic rankings, this paragraph should be deleted.
4
....c
.~.=c.
.-w.
CMTC SECTION PARAG"APH COMMENT CO?fECTIONS evaluations, but are not sufficiently precise to be ad in a safety assessment that may be submitted to the NRC staff. In addition, it is unclear how the NRC staff will use such models to evaluate a selety concern.
27.
iM 2.
This requested information implies that the GL 88 05 visual Information inspection is inadequate to detect boric acid deposita and which Boric acid deposita will be identified by the could be caused by PWSCC. This implication is not supported visual inspections recommended in Generic Letter 88-05.
by operating history and safety evaluatio..s:
' The only through. wall VHPs cracks (Hugey and Zorita) were 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 wall leak.
Boric acid deposited over this time period would be readily observed using the GL 88 05 visual inspections.
28.
g d
3.
The intergranular stress corrosion cracking resulting from a g
stion Zorita type resin intrusion is a di& rent mechanism than the -
The resin induced intergranular stress primary water stress corrosion cracking (PWSCC). The resin corrosion cracking is different than the intrusion cracking is a degradation mechanism caused by an stated scope.and should be deleted.
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.
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
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Southern Nudear Operating Company m. u.<.y ve. pr. o.n Fartey Pro @ct the sOG!ne'n ei&C!!C sisfe'"
Septenber 30, 1996 Docket Nos.
50-348 50-364 Mr. David L. Meyer Chief, Rules Review and Directives Branch U. S. Nuclear Regulatory Commission Washington, D. C. 20555 c.
Comments on Proposed Generic Communication
[
" Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Othe Vessel Head Penetrations" (61 Federal Register 40253 dated August 1.1996) j
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,
! Y b i e -<
Dave Morey DNM/TWS J?lpicibDMd' SPP
-=
e U. S. Nuclear Regulatory Commission Page Two J
cc:
Southern Nuclear Operatine Comoany R. D. Hill, Plant Manager U. S. Nuclear Regulatory Commission. Washington. DC
- 11. Zimmerman, Licensing Project Manager, NRR j
U. S. Nuclear Regulatory Commission. Region II j
S. D. Ebneter, ReFional Administrator T. M. Ross, Senior Resident Inspector j
40 in$e< ness csnty Psenway dM[
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c.x.uccoy october 1, 1996 Georgia Power V.Ce Pre $' dea! Nuclea' Vogtse Protect l'Y SOUfheffl e'6CIf'C SrSle" Docket Nos.
50-321 50-424 HL-5247 50-366 50-425 LCV-0885 Mr. David L Meyer
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Chief, Rules Review and Directives Branch S
g U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Comments on Proposed Generic Communication
~
" Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Othere Vessel Head Penc: rations" (61 Federal Register 40253 dated August 1.1996)
Dear Sir.
Georgia 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.
Should you have any questions, please advise.
Respectfully submitted, e.41f '
C. K. McCoy CKM/TWS
GeorgiaPbwer1 october 1, 1996
]
l U. S. Nuclear Regulatory Commission Page Two cc:
Georgia Power Comoiny J. T. Beckham, Jr., Vice President - Plant Hatch 1
J. B. Beasley, General Manager - Vogtle Electric Generating Plant H. L. Sumner, Jr., General Manager - Plant Hatch U. S Nuclear Renulato_rv Commission. Washington. DC K. N. Jabbour, Licensing Project Manager - Hatch i
L. L Wheeler, Licensing Project Manager, Vogtle U S Nuclear Regulatory Commission. Renion II S. D. Ebneter, Regional Administrator B. L. Holbrook, Senior Resident Inspector - Hatch C. R. Ogle, Senior Resident Inspector - Vogtle a
I HL-5247 LCV-0885 REES File: G.03.19
//.)Of Z [/Jpfe r/t Flor da Power & Light CompaIy P.0 Box 14000, Jrro Beach, FL 33408 0420 6Mx Vasj FPL OCT 0 21996 g j gf
,g L-96-249 p
e..i.k L 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 Commission 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, h.
. cm. b c m gW. H. Bohlke Vice President Nuclear Engineering nn FPL Group company d&f07l0011i^1l'
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9 Log # TXX 96484
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File # 883
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7UELECTR/C October 3. 1996 C. Lunce Terry Greap Vke treaknr Chief. Rules Review and Directives Branch U. 5. Nuclear Regulatory Commission Washington. DC 20555 0001
SUBJECT:
COMANCHE PEAX STEAM ELECTRIC STATION (CPSES)
ENDORSEHENT OF NEI COMMENT LETTER ON PROPOSED NRC GENERIC COHHUNICATION. PRIMARY WATER STP.ESS CORROSION CRACKING OF CONTROL R00 DRIVE HECHANISH ANC 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 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
" Prim'ary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessal Head Penetrations."
TV Electric has reviewed and endorses the NEI letter (Reference 2).
TV Electric agrees with the NEI discussed issues, recommendations and rationale. TU 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 c-R. E. Beedle NEI P.O. Box 1002 Glen Rose. Texas 7600 4/aleibo4 W P
y soav v.mey Powtr Staten
/
SNppingport, PA 15077 0004
{'{
P october 4, 1996 y,d'2ly3g en, Nuclear Power Divtsion Mr. David L. Meyer Chief, Rules Review and Directives Branch U. S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555-0001 4
d
Subject:
Beaver Valley Power Station, Unit No. I and No. 2 l
BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Proposec' Generic Communication, " Primary Water Stress Corrosion Crackin;; 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 Ihsd Penetrations," which was published in the Federal Register on August 1, 1996 (61 FR 40253). CLC hereby submits the following comments.
DLC endorses the comments provided by the Nuclear Energy Institute (NEI). The NEI 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 Department, (412) 393-5210.
Sincerely, Sushil C. Jain DELIVERING GlJAL1iY ENERGV 4
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mm wrn October 4, !996 4
Chief, Rules Review and Directives Branch Serial No.96-076 U. S. Nuclear Regulatory Commission Washington,DC 20555 4
Gentlemen:
CONINiENTS ON PROPOSED GENERIC LETTER:
PRINf ARY WATER STRF5sS CORROSION CRACKING OF CONTROL ROD DRIVE MECHANISN1S 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 review:d the NRC proposed generic letter and fully endorse the Nuclear Energy Ins 5te 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 truly yours, M.
s1, Vice President vincering & Services Attachment cc: Mr. Thomas E. Tipton Vice President, Operations and Engineering Nuclear Energy Institute 1776 Eye Street Suite 300 Washington, DC 20006-3706
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