Fact Sheet on Reactor Pressure Vessel Issues
Embrittlement
Reactor pressure vessels, which
contain the nuclear fuel in nuclear power plants, are made of thick steel plates
that are welded together. Neutrons from the fuel in the reactor irradiate
the vessel as the reactor is operated. This can embrittle the steel, or make it
less tough, and less capable of withstanding flaws which may be present.
Embrittlement usually occurs at a vessel’s “beltline,” that section of the
vessel wall closest to the reactor fuel.
Pressurized water reactors (PWRs) are more susceptible to embrittlement than
are boiling water reactors (BWRs). BWR vessels generally experience less neutron
irradiation and therefore less embrittlement. Many utilities owning PWRs use
core designs that reduce the number of neutrons that reach the vessel wall.
These design features therefore reduce the rate of embrittlement in the reactor
vessels. Other factors also contribute to the degree to which a particular
vessel material becomes embrittled. Steels with a higher proportion of copper
and nickel will tend to be more susceptible to embrittlement, than are steels
with lower proportions of these two elements. [Pertinent regulations that govern
this phenomenon include 10 CFR Part 50, Appendix G “Fracture Toughness
Requirements” and Appendix H, “Reactor Vessel Material Surveillance Program
Requirements.”]
Another reason reactor vessel embrittlement is more of a concern for PWRs is
because PWRs may experience pressurized thermal shock (PTS). PTS can occur under
some accident scenarios that introduce cold water into the reactor vessel while
the vessel is pressurized. Introduction of cold water in this manner can cause
the vessel to cool rapidly, resulting in large thermal stresses in the steel.
These thermal stresses, along with the high internal pressure and an embrittled
vessel, could lead to cracking and even failure of the vessel. The NRC has
established a regulation (10 CFR part 50.61 – the “PTS rule”) to address the
potential for the reactor vessels of PWRs to be affected by PTS events. The PTS
rule includes criteria that limit the amount of vessel embrittlement the NRC
will permit without requiring additional evaluations or corrective actions.
BWRs are not susceptible to PTS events. If cold water is pumped into a BWR
vessel, the steam in the vessel will condense and reduce the internal pressure.
BWRs may however be susceptible to overpressurization of the reactor pressure
vessel at low temperatures under certain conditions.
NUREG‑1511, “Reactor Pressure Vessel Status Report” discusses the issue of
vessel structural integrity and provides a plant‑by‑plant status. Updates were
published in October 1996 (Supplement 1) and in October 2000 (Supplement 2). All
plants are expected to maintain adequate toughness throughout their operating
lives.
Primary Water Stress Corrosion Cracking of Upper Reactor Vessel Head
Penetration Nozzles in PWRs
Control rod drive mechanism nozzles and other vessel head penetration
nozzles welded to the upper reactor vessel head are subject to
another phenomenon – primary water stress corrosion cracking. The issue is a
potential safety concern because a nozzle with sufficient cracking could break
off during operation. This would compromise the integrity of the reactor coolant
system pressure boundary – one of three primary barriers that protect the public
from exposure to radiation. The break may also result in the ejection of a
control rod, which could damage nearby components.
On August 3, 2001, the NRC issued Bulletin 2001-01, “Circumferential Cracking
of Reactor Pressure Vessel Head Penetration Nozzles,” to licensed holders of
U.S. pressured water reactors (PWR) following the discovery of cracked and
leaking nozzles in 2000 and 2001. In the bulletin, the staff requested
information from PWR licensees about the structural integrity of these nozzles
at their facilities. In response to the bulletin, licensees provided their plans
for inspecting their nozzles and the outside surfaces of their upper reactor
vessel heads to determine whether any nozzles were leaking. Inspections by
licensees during the fall of 2001 revealed vessel head penetration nozzle cracks
at Three Mile Island Unit 1, Crystal River Unit 3, North Anna Unit 1, and Oconee
Unit 3.
On August 9, 2002, the NRC issued Bulletin 2002-02, “Reactor Pressure Vessel
Head and Vessel Head Penetration Nozzle Inspection Programs.” The bulletin
suggested that visual inspections of upper reactor vessel heads and their
nozzles may need to be supplemented with non-visual non-destructive examinations
to assure that the structural integrity and leakage integrity of the nozzles is
maintained. Bulletin 2002-02 requested that PWR licensees provide information
about their inspection programs and plans to supplement existing visual
inspections with volumetric and surface examinations. Licensees responded with
descriptions of inspection plans for at least the first refueling outage
following the issuance of the bulletin. Many did not offer any long-term
inspection plans, but instead opted to follow guidance being developed by the
Materials Reliability Program, which is an industry-sponsored research
organization.
Inspections performed at several PWRs in 2002 including those performed at
the Davis-Besse Nuclear Plant, found leakage and cracks in vessel head
penetration nozzles or J-groove welds that have required repairs or prompted the
replacement of the vessel head. As a result of continuing concerns regarding
licensee inspection programs in this area, the NRC issued an Order on February
11, 2003, to all PWR licensees in the U.S. The Order requires specific
inspections of the vessel head and associated penetration nozzles based on their
susceptibility to primary water stress corrosion cracking. The Order may be
accessed at the following address on NRC's website: http://www.nrc.gov/reactors/operating/ops-experience/vessel-h...
.
Electric Power Research Institute's Materials Reliability Program as having a
high susceptibility to nozzle cracking. Inspections by licensees performed after
issuance of the latest bulletin and order, revealed nozzle or J-groove weld
cracks and/or leaks at Oconee Unit 2, North Anna 2, Arkansas Nuclear One Unit 1,
St. Lucie Unit 2, Milestone Unit 2, and Beaver Valley Unit 1. The utilities
owning the Oconee, Surry, Davis-Besse and North Anna nuclear stations have
replaced or are in the process of replacing their upper reactor vessel heads.
Approximately twenty other units have announced plans to have their upper
reactor vessel heads replaced within the next few years.
Reactor Vessel Damage at Davis-Besse
In early March of 2002, during an inspection prompted by Bulletin 2001-01,
Davis-Besse Nuclear Power Station identified a football-sized cavity in the
units reactor vessel head. The cavity was next to a leaking nozzle with a
through-wall crack and was in an area of the vessel head that had been covered
with boric acid deposits for several years. Inspections at Oconee Unit 1 and
Millstone Unit 2 also identified nozzle cracking. The discovery of leaks and
nozzle cracking at Davis-Besse and other PWR plants called for more effective
inspections of reactor pressure vessel heads and associated penetration nozzles.
On March 13, 2002, the NRC issued a Confirmatory Action Letter to First
Energy Nuclear Corporation confirming the company's commitments to evaluate and
resolve damage to the reactor pressure vessel head at the Davis-Besse Nuclear
Power Station, which is located in Oak Harbor, Ohio. Inspections revealed a
cavity in the top of the reactor pressure vessel head that may have been caused
by corrosion from boric acid deposits. Two meetings were held – one with
industry and one with the public – to discuss the generic implications of the
problem.
Head Degradation and Reactor Coolant Pressure Boundary Integrity,” to address
the generic implications of the degradation at Davis-Besse on the safe operation
of PWRs in the U.S. and on the health and safety of the public. NRC Bulletin
2002-01 and the NRC's review of the industry's responses to Bulletin 2002-01 may
be accessed at the following addresses on NRC's website: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/bullet...
and http://www.nrc.gov/reactors/operating/ops-experience/vessel-h....
Primary Water Stress Corrosion Cracking of Lower Reactor Vessel Head
Penetration Nozzles in PWRs
On April 12, 2003, the licensee for South Texas Project Unit 1 (STP 1)
discovered small boron deposits around two of the unit's bottom mounted
instrumentation penetration nozzles during a bare metal visual examination of
the reactor pressure vessel (RPV) bottom head. Subsequent nondestructive
examination of all 58 nozzles at the South Texas nuclear power plant confirmed
the existence of leaking, axially-oriented flaws in the two nozzles. No flaws
were found in any of the other 56 nozzles at the plant. The licensee repaired
the two cracked nozzles using a method known as a “half-nozzle” repair and
returned the unit to power operations in August 2003.
Based on the licensee's examination results and the information gained from a
material sample obtained from one of the leaking nozzles, the licensee for South
Texas 1 concluded that the observed cracking was due to primary water stress
corrosion of the Inconel Alloy 600 nozzle material. The licensee also concluded
that the most likely root cause of this cracking involved fabrication-related
defects which may have created conditions that lead to initiation of this nozzle
cracking.
Additional information on this experience may be accessed on NRC's web site
at: http://www.nrc.gov/reactors/operating/ops-experience/bottom-h....
On August 13, 2003, the NRC issued NRC Information Notice 2003-11, “Leakage
Found on Bottom-Mounted Instrumentation Nozzles,” to inform members of the U.S.
nuclear power industry and members of the public of this event. Information
Notice 2003-11 may be accessed on NRC's web site http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-n...
.
On August 21, 2003, the NRC issued NRC Bulletin 2003-02, “Leakage from
Reactor Pressure Vessel Lower Head Penetrations and Reactor Coolant Pressure
Boundary Integrity,” in order to address the generic safety implications of the
South Texas cracking experience on operations of pressurized water reactors in
the U.S. and specifically as the operations relate to maintaining the health and
safety of the public. The staff is currently in the progress of reviewing the
industry’s responses to NRC Bulletin 2003-02. NRC Bulletin 2003-02 may be
accessed on NRC's web site at: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/bullet...
.
December 2003
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