[Federal Register Volume 61, Number 149 (Thursday, August 1, 1996)]
[Notices]
[Pages 40253-40257]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-19588]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Proposed Generic Communication; Primary Water Stress Corrosion
Cracking of Control Rod Drive Mechanism and Other Vessel Head
Penetrations
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of opportunity for public comment.
-----------------------------------------------------------------------
SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to issue
a generic letter concerning primary water stress corrosion cracking in
control rod drive mechanisms and other vessel head penetrations of
nuclear power reactors. The purpose of the proposed generic letter is
to (1) request that addressees describe their program for ensuring the
timely inspection of PWR control rod drive mechanism (CRDM) and other
vessel head penetrations and (2) require that all addressees provide to
the NRC a written response to this generic letter. The NRC is seeking
comment from interested parties regarding both the technical and
regulatory aspects of the proposed generic letter presented under the
Supplementary Information heading.
The proposed generic letter was endorsed by the Committee to Review
Generic Requirements (CRGR) on July 25, 1996. The relevant information
that was sent to the CRGR will be placed in the NRC Public Document
Room. The NRC will consider comments received from interested parties
in the final evaluation of the proposed generic letter. The NRC's final
evaluation will include a review of the technical position and, as
appropriate, an analysis of the value/impact on licensees. Should this
generic letter be issued by the NRC, it will become available for
public inspection in the NRC Public Document Room.
DATES: Comment period expires September 3, 1996. Comments submitted
after this date will be considered if it is practical to do so, but
assurance of consideration cannot be given except for comments received
on or before this date.
ADDRESSES: Submit written comments to Chief, Rules Review and
Directives Branch, U.S. Nuclear Regulatory Commission, Mail Stop T-6D-
69, Washington, DC 20555-0001. Written comments may also be delivered
to 11545 Rockville Pike, Rockville, Maryland, from 7:30 am to 4:15 pm,
Federal workdays. Copies of written comments received may be examined
at the NRC Public Document Room, 2120 L Street, N.W. (Lower Level),
Washington, D.C.
FOR FURTHER INFORMATION CONTACT: C. E. (Gene) Carpenter (301) 415-2169.
SUPPLEMENTARY INFORMATION:
Generic Letter 96-##: Primary Water Stress Corrosion Cracking of
Control Rod Drive Mechanism and Other Vessel Head Penetrations (TACS
No. M95280)
Addressees
All holders of operating licenses for pressurized water reactors
(PWRs), except those licenses that have been amended to possession-only
status.
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 head penetrations and (2) require that all
addressees provide to the NRC a written response to this generic letter
relating to the requested information.
Background
Most PWRs have Alloy 600 CRDM nozzle and other vessel head
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. The weld
between the nozzle and the housing 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 control rod drive (CRD) nozzles and other VHPs are
basically the same for all PWRs worldwide, which use a U.S. design
(except in Germany and Russia).
Generally, there are 36 to 78 nozzles distributed over the low-
alloy steel head. The vessel head is semi-spherical and the head
penetrations are vertical so that the CRD nozzles and other VHPs are
not perpendicular to the 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)
between the nozzle and the sleeve.
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. Other leaks have occurred since 1986 in
several Alloy 600 pressurizer instrument nozzles at both domestic and
foreign reactors from several different nuclear steam supply system
vendors. 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. The NRC staff issued
Information Notice 90-10, ``Primary Water Stress Corrosion Cracking
(PWSCC) of Inconel 600,'' dated February 23, 1990, to inform the
nuclear industry of the issue.
In December 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, Switzerland, Sweden, Spain, and Japan have uncovered
additional VHPs with axial cracks. 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.
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
[[Page 40254]]
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 were held withNUMARC/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 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.
In 1994, circumferential intergranular attack (IGA) associated with
the J-groove weld in one of the CRDM penetrations was discovered at
Zorita, a Spanish reactor. This IGA is a different degradation
mechanism than the PWSCC described above. It is believed to have
resulted from the combination of ion exchange resin bed intrusions,
which resulted in high concentrations of sulfates. 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 of cation
resin entered the reactor coolant system (RCS). In September 1981, a
mixed bed demineralizer screen failed and between 200 to 320 liters 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 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 Information Notice (IN) 96-11, ``Ingress of
Demineralizer Resins Increases Potential for Stress Corrosion Cracking
of Control Rod Drive Mechanism Penetrations,'' dated February 14, 1996,
to alert addressees to the increased likelihood of sulfate-driven
stress corrosion cracking of PWR CRDMs and other VHPs if demineralizer
resins contaminate the RCS.
The Westinghouse staff notified the WOG plants, the B&WOG plants,
and the CEOG plants of the Zorita incident by issuing NSAL-94-028.
Westinghouse reported that no other plant had been found worldwide that
had experienced cracking similar to that at the Zorita plant. The
Westinghouse staff further reported that U.S. plants monitor RCS
conductivity on a routine basis, follow the EPRI guidelines on primary
water chemistry, and monitor for sulfate three times a week. The
Westinghouse staff concluded that no immediate safety issue is involved
and that the conclusions in its CRDM safety evaluation remain valid.
The Westinghouse staff suggested that U.S. PWR plants review their RCS
chemistry and other operating records pertaining to sulfur ingress
events. The results of this review have not been reported to the NRC
staff, and the NRC staff does not have sufficient information to
ascertain whether any significant primary system resin bed intrusions
have occurred at any U.S. PWR.
The first U.S. inspection of VHPs took place in the spring of 1994
at the Point Beach Nuclear Generating Station, and no indications were
uncovered 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, 16 mm, and 6 to 8 mm in
length, and the deepest flaw was 6.8 mm deep. The tip of the 46-mm 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 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.
[[Page 40255]]
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, neither
Westinghouse, FTI, nor CE 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 an economic
one 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 only economic. Inspections have shown that cracking
has initiated in some U.S. plants, and the industry has not provided
sufficient technical justification regarding susceptibility of the CRDM
and other VHPs to PWSCC to justify an inspection plan based on economic
considerations alone.
Discussion
The results of domestic VHP inspections are consistent with the
February 1993 analyses by the PWR Owners Groups, the NRC staff safety
evaluation report dated November 19, 1993, and the PWSCC found in the
CRDMs in European reactors. On the basis of the results of the first
five inspections of U.S. PWRs, the PWR Owner's Groups' analyses, and
the European experience, the NRC staff has determined that there is a
high probability that VHPs at other plants may contain similar axial
cracks caused by PWSCC. Further, if any significant resin intrusions
have occurred at U.S. PWRs such as occurred at Zorita, residual
stresses are sufficient to cause circumferential intergranular stress
corrosion cracking (IGSCC).
After considering this information, the NRC staff has concluded
that VHP cracking does not pose an immediate or near term safety
concern. Further, the NRC staff recognizes that the scope and timing of
inspections may vary for different plants depending on their individual
suceptibility 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 head
provides the vital function of maintaining a reactor pressure boundary.
Cracking in the VHPs has occurred and is expected to continue to occur
as plants age. The NRC staff considers cracking of VHPs to be a safety
concern for the long term based on the possibility of (1) exceeding the
American Society of Mechanical Engineers (ASME) Code for margins if the
cracks are sufficiently deep and continue to propagate during
subsequent operating cycles, and (2) eliminating a layer of defense in
depth for plant safety. Therefore, in order to verify that the margins
required by the ASME Code, as specified in Section 50.55a of Title 10
of the Code of Federal Regulations (10 CFR 50.55a) are met, that the
guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50
(10 CFR Part 50, Appendix A, GDC 14) is continued to be satisfied, and
to ensure that the safety significance of VHP cracking remains low, the
NRC staff believes that an integrated, long-term program, which
includes periodic inspections and monitoring, is necessary. In
addition, the NRC staff finds that the requested information is also
needed to determine if the imposition 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 wish to discourage
such group actions but notes that such an integrated industry
inspection program must have a well-founded technical basis that
justifies the relationship between the plants and the planned
implementation schedule.
Required Information
The information required in items 1 and 2, below, is required by
the NRC staff to determine if the imposition of an augmented inspection
program is required, while the information required in item 3 relates
to the potential for domestic resin intrusions, such as occurred at
Zorita.
Addressees are required to provide the following information:
1. Regarding inspection activities:
1.1 A description of all inspections of CRDMs and other vessel
head penetrations performed to the date of this generic letter,
including the results of these inspections.
1.2 If you have developed a plan to periodically inspect the CRDM
and other vessel head penetrations:
a. Your schedule for first, and subsequent, inspections of the CRDM
and other vessel head penetrations, including the technical basis for
your schedule.
b. Your scope for the CRDM and other vessel head penetration
inspections, including whether you plan to inspect from the top or
bottom of the head, the total number of penetrations (and how many will
be inspected), and which penetrations have thermal sleeves, which are
spares, and which are instrument or other penetrations.
1.3 If you have not developed a plan to periodically inspect the
CRDM and other vessel 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.
2. A description of the evaluation methods and results used to
assess the susceptibility of the CRDM and other VHPs in your plant to
PWSCC, including the susceptibility ranking of your plant and the
factors used to determine this ranking. Other than or in addition to
the boric acid visual examination (see Generic Letter 88-05, ``Boric
Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in
PWR Plants,'' dated March 17, 1988), include a description of all
relevant data and/or tests used to develop crack initiation and crack
growth models, and the methods and data used to validate these models.
Include a statement explaining the applicability of these models to the
VHP cracking issue. Also, if you are relying on any integrated industry
inspection program, provide a detailed description of this program.
3. A description of any resin intrusions in your plant, as
described in IN 96-11, that have exceeded the current EPRI PWR Primary
Water Chemistry Guidelines recommendations for primary water sulfate
levels, including the following information:
3.1 Were the intrusions cation, anion, or mixed bed?
3.2 What were the durations of these intrusions?
[[Page 40256]]
3.3 Do your RCS water chemistry Technical Specifications follow
the EPRI guidelines?
3.4 Identify any RCS chemistry excursions that exceed your plant
administrative limits for the following species: sulfates, chlorides or
fluorides, oxygen, boron, and lithium.
3.5 Identify any conductivity excursions which may be indicative
of resin intrusions, provide your technical assessment of each
excursion and your followup actions.
3.6 Provide your assessment of the potential for any of these
intrusions to result in a significant increase in the probability for
IGA of VHPs and any associated plan for inspections.
Required Response
All addressees shall submit in writing the information identified
above within 90 days from the date of this letter.
Any inspection results that do not satisfy the acceptance criteria
identified in the NRC staff's safety assessment dated November 16,
1993, should be reported to the NRC staff prior to plant restart.
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 CRDMs and other vessel head
penetrations. For this reason, the above time periods are allowed for
the responses.
Related Generic Communications
(1) Information Notice 90-10, ``Primary Water Stress Corrosion
Cracking (PWSCC) of Inconel 600,'' dated February 23, 1990.
(2) NUREG/CR-6245, ``Assessment of Pressurized Water Reactor
Control Rod Drive Mechanism Nozzle Cracking,'' dated October 1994.
(3) Information Notice 96-11, ``Ingress of Demineralizer Resins
Increases Potential for Stress Corrosion Cracking of Control Rod Drive
Mechanism Penetrations,'' dated February 14, 1996.
Backfit Discussion
This generic letter only requires information from the addressees
under the provisions of Section 182a of the Atomic Energy Act of 1954,
as amended, and 10 CFR 50.54(f). Therefore, the staff has not performed
a backfit analysis. The information collected will enable the staff to
verify that the margins required by the ASME Code, as specified in
Section 50.55a of Title 10 of the Code of Federal Regulations (10 CFR
50.55a) are met, that the guidance of General Design Criterion 14 of
Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14)
continues to be satisfied, and to ensure that the safety significance
of VHP cracking remains low, the NRC staff requires licensees to submit
information to assess compliance with the above stated requirements.
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 staff is not establishing a new position for
such compliance in this generic letter. Therefore, this generic letter
does not constitute a backfit and no documented evaluation or backfit
analysis need be prepared.
Dated at Rockville, Maryland, this 26th day of July, 1996.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Program Management, Office of
Nuclear Reactor Regulation.
BILLING CODE 7590-01-P
[[Page 40257]]
[GRAPHIC] [TIFF OMITTED] TN01AU96.000
[FR Doc. 96-19588 Filed 7-31-96; 8:45 am]
BILLING CODE 7590-01-C