Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
April 1, 1997
NRC GENERIC LETTER 97-01: DEGRADATION OF CONTROL ROD DRIVE MECHANISM
NOZZLE AND OTHER VESSEL CLOSURE HEAD
PENETRATIONS
Addressees
All 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 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.
Background
Primary Water Stress Corrosion Cracking of Vessel Closure 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. (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 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 head penetrations are
vertical so that the CRDM nozzle and other VHPs are 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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Beginning in 1986, leaks have been reported in several Alloy 600 pressurizer
instrument 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 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 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.
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 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,
?lecricit‚ 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
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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
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 intergranular 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
increase in conductivity was attributed to locally high
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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, 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.
Westinghouse 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. 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 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 cannot accurately size defects that
are less than one mil 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 length, and the deepest flaw was 6.8 mm [0.27 in.] deep. The tip
of the 46-mm [1.81 in.] flaw 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.
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
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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.
By letter dated March 5, 1996, NEI 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 primarily an economic rather than a safety issue,
and 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 primarily economic.
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 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 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 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,
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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 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 advantage 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 the relationship between the plants and the planned
implementation schedule.
Requested 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:
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1. Regarding inspection activities:
1.1 A description of all inspections 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 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 this 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?
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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.
2.5 Identify any conductivity excursions which may be indicative of resin
intrusions. Provide a technical assessment of each excursion and any
followup actions.
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 inspections.
Required Response
Within 30 days of the date of this generic letter, each addressee is required
to submit a written response indicating: (1) whether or not the requested
information will be submitted and (2) whether or not the requested information
will be submitted within the requested time period. Addressees who choose not
to submit the requested information, or are unable to satisfy the requested
completion date, must describe in their response any alternative course of
action that is proposed to be taken, including the basis for the acceptability
of the proposed alternative course of action.
NRC staff will review the responses to this generic letter and if concerns are
identified, 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.
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 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.
(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.
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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 with applicable existing
regulatory requirements. Specifically, the requested information would enable
the NRC staff to determine whether or not the licensees' 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 Register 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.
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 by the Office of Management and Budget, approval
number 3150-0011, which expires July 31, 1997.
The public reporting burden for this collection of information is estimated to
average 80 hours 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
information to be collected?
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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 displays 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.
signed by
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
James Medoff
(301) 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
