Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI) Requirements (Generic Letter 89-21)
NOTE: ENCLOSURE 1 OF GENERIC LETTER 89-21 "REQUEST FOR INFORMATION
CONCERNING STATUS OF IMPLEMENTATION OF UNRESOLVED SAFETY ISSUE
REQUIREMENTS" CONTAINS 6 COLUMNS (USI/MPA NUMBER, TITLE,
REF. DOCUMENT APPLICABILITY, STATUS/DATE*, AND REMARKS). THE
STATUS/DATE* AND REMARKS COLUMNS ARE (AS DESCRIBED IN THE GENERIC
LETTER) BLANK AND ARE TO BE FILLED IN BY THE LICENSEE. IN ORDER
TO TRANSMIT THIS GENERIC LETTER ELECTRONICALLY, THE STATUS/DATE*
AND REMARKS COLUMN HAVE BEEN REMOVED. HOWEVER, THESE COLUMNS ARE
CONTAINED IN THE COPY OF THE GENERIC LETTER LICENSEE'S WILL
RECEIVE IN THE MAIL.
.
October 19, 1989
TO: ALL HOLDERS OF OPERATING LICENSES AND CONSTRUCTION PERMITS FOR
NUCLEAR POWER REACTORS
SUBJECT: REQUEST FOR INFORMATION CONCERNING STATUS OF IMPLEMENTATION OF
UNRESOLVED SAFETY ISSUE (USI) REQUIREMENTS (GENERIC LETTER 89-21)
As part of our continuing effort to validate staff understanding regarding
implementation of significant regulatory issues, the staff is conducting a
comprehensive review of the implementation status of unresolved safety issues
(USIs). An important aspect of this effort is to ensure that the licensee and
NRC agree on the status of USI resolution implementation at each facility.
The purpose of this letter is to request your review and reporting of the
status of implementation of USIs for which a final technical resolution has
been achieved and which are applicable to your facility.
To assist you in this effort, I have enclosed a table of USIs for which a
final technical resolution has been achieved (Enclosure 1). This table
indicates other information, such as multiplant action (MPA) number, generic
letter number, applicability, and reference NUREG number. For your
facility, determination of requirements for a particular USI may necessitate
review of applicable generic letters, NUREG documents, or plant-specific
correspondence. For your information, a summary of the resolution of each
USI is provided in Enclosure 2.
As in the case of our earlier correspondence related to the status of imple-
mentation of TMI Task Action Plan items, implementation should be considered
complete when activities have been performed necessary to satisfy the require-
ments (or assumptions) made in the staff's technical resolution of the
particular USI. If you have not fully completed an item, we ask you to mark
up the enclosure to reflect your projected implementation date. You should
add a short note identifying remaining work (e.g., hardware, procedures,
training, technical specifications). More explicit instructions are
provided as part of Enclosure 1.
Your NRC Project Manager is developing data sheets that identify significant
plant-specific correspondence between each licensee and the staff relating
to a particular USI. Once we have researched agency files we will provide
this information to your staff. This will ensure we both have a clear
record of major actions regarding the USI. The Project Manager can provide
additional clarification which may be of assistance to you and will work
with your staff to identify plant-specific references.
We request that this information be provided within 30 days of receipt of this
letter. The information we are requesting will be utilized to validate and
update our existing databases so that we will have an accurate and complete
understanding of the status of USI implementation at each nuclear power
plant.
8910160250
.
Generic Letter 89-21 - 2 - October 19, 1989
This request is covered by Office of Management and Budget Clearance Number
3150-0011, which expires December 31, 1989. The estimated average burden
hours is 80 person hours per plant, including searching data sources,
gathering and analyzing the data, and preparing the required letter. Send
comments regarding this burden estimate or any other aspect of this
collection of information, including suggestions for reducing this burden,
to the Records and Reports Management Branch, Division of Information
Support Services, Office of Information Resources Management, U.S. Nuclear
Regulatory Commission, Washington, D.C. 20555; and to the Paperwork
Reduction Project (3150-0011), Office of Management and Budget,
Washington, D.C. 20503.
Sincerely,
James G. Partlow
Associate Director for Projects
Office of Nuclear Reactor Regulation
Enclosures:
1. USI Table
2. USI Issues Summary
3. List of Most Recently Issued NRC Generic Letters
.
ENCLOSURE 1
UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL TECHNICAL RESOLUTION
HAS BEEN ACHIEVED
USI/MPA
NUMBER TITLE REF. DOCUMENT APPLICABILITY
A-1 Water Hammer SECY 84-119 All
NUREG-0927, Rev. 1
NUREG-0993, Rev. 1
NUREG-0737 Item
I.A.2.3
SRP revisions
A-2/ Asymmetric Blowdown NUREG-0609 PWR
MPA D-10 Loads on Reactor Primary GL 84-04, GDC-4
Coolant Systems
A-3 Westinghouse Steam NUREG-0844 W-PWR
Generator Tube Integrity SECY 86-97
SECY 88-272
GL 85-02
(No requirements)
A-4 CE Steam Generator Tube NUREG-0844, SECY 86-97 CE-PWR
Integrity SECY 88-272
GL 85-02
(No requirements)
A-5 B&W Steam Generator NUREG-0844, SECY 86-97 B&W-PWR
Tube Integrity SECY 88-272
GL 85-02
(No Requirements)
A-6 Mark I Containment NUREG-0408 Mark I-BWR
Short-Term Program
* C - COMPLETE
NC - NO CHANGES NECESSARY
NA - NOT APPLICABLE
I - INCOMPLETE
E - EVALUATING ACTIONS REQUIRED
.
- 2 -
USI/MPA
NUMBER TITLE REF. DOCUMENT APPLICABILITY
A-7/ Mark I Long-Term NUREG-0661 Mark I-BWR
D-01 Program NUREG-0661 Suppl. 1
GL 79-57
A-8 Mark II Containment NUREG-0808 Mark II-BWR
Pool Dynamic Loads NUREG-0487, Suppl. 1/2
NUREG-0802
SRP 6.2.1.1C
GDC 16
A-9 Anticipated Transients NUREG-0460, Vol. 4 All
Without Scram 10 CFR 50.62
A-10/ BWR Feedwater Nozzle NUREG-0619 BWR
MPA B-25 Cracking Letter from DG Eisenhut
dated 11/13/80
GL 81-11
A-11 Reactor Vessel Material NUREG-0744, Rev. 1 All
Toughness 10 CFR 50.60/
82-26
A-12 Fracture Toughness of NUREG-0577, Rev. 1 PWR
Steam Generator and SRP Revision
Reactor Coolant Pump 5.3.4
Supports
A-17 Systems Interactions Ltr: DeYoung to All
licensees - 9/72
NUREG-1174, NUREG-
1229, NUREG/CR-3922,
NUREG/CR-4261, NUREG/
CR-4470, GL 89-18
(No requirements)
A-24/ Qualification of Class NUREG-0588, Rev. 1 All
MPA B-60 1E Safety-Related SRP 3.11
Equipment 10 CFR 50.49
GL 82-09, GL 84-24
GL 85-15
.
- 3 -
USI/MPA
NUMBER TITLE REF. DOCUMENT APPLICABILITY
A-26/ Reactor Vessel Pressure DOR Letters to PWR
MPA B-04 Transient Protection Licensees 8/76
NUREG-0224
NUREG-0371
SRP 5.2
GL 88-11
A-31 Residual Heat Removal NUREG-0606 All OLs After
Shutdown Requirements RG 1.113, 01/79.
RG 1.139
SRP 5.4.7
A-36/ Control of Heavy Loads NUREG-0612 All
C-10, Near Spent Fuel SRP 9.1.5
C-15 GL 81-07, GL 83-42,
GL 85-11
Letter from DG
Eisenhut dated
12/22/80
A-39 Determination of SRV NUREG-0802 BWR
Pool Dynamic Loads NUREGs-0763,0783,0802
and Pressure Transients NUREG-0661
SRP 6.2.1.1.C
A-40 Seismic Design SRP Revisions, NUREG/ All
Criteria CR-4776, NUREG/CR-0054,
NUREG/CR-3480, NUREG/
CR-1582, NUREG/CR-1161,
NUREG-1233, NUREG-4776
NUREG/CR-3805
NUREG/CR-5347
NUREG/CR-3509
A-42/ Pipe Cracks in Boiling NUREG-0313, Rev. 1 BWR
MPA B-05 Water Reactors NUREG-0313, Rev. 2
GL 81-03, GL 88-01
.
- 4 -
USI/MPA
NUMBER TITLE REF. DOCUMENT APPLICABILITY
A-43 Containment Emergency NUREG-0510, All
Sump Performance NUREG-0869, Rev. 1
NUREG-0897, R.G.1.82
(Rev. 0), SRP 6.2.2
GL 85-22
No Requirements
A-44 Station Blackout RG 1.155 All
NUREG-1032
NUREG-1109
10 CFR 50.63
A-45 Shutdown Decay Heat SECY 88-260 All
Removal Requirements NUREG-1289
NUREG/CR-5230
SECY 88-260
(No requirements)
A-46 Seismic Qualification NUREG-1030 All
of Equipment in NUREG-1211/
Operating Plants GL 87-02, GL 87-03
A-47 Safety Implication NUREG-1217, NUREG- All
of Control Systems 1218
GL 89-19
A-48 Hydrogen Control 10 CFR 50.44 All, except
Measures and Effects SECY 89-122 PWRs with
of Hydrogen Burns large dry
on Safety Equipment containments
A-49 Pressurized Thermal RGs 1.154, 1.99 PWR
Shock SECY 82-465
SECY 83-288
SECY 81-687
10 CFR 50.61/
GL 88-11
.
GUIDE FOR UPDATING USI STATUS
(1) Enclosure 1 lists all unresolved safety issues (USIs) for which a final
technical resolution has been achieved. Please review the entire
listing for each licensed reactor unit. Where an item is not
applicable for your facility, mark "NA" in the status column.
(2) Where an item is applicable to your facility, but no changes were
necessary, mark "NC" in the status column.
(3) Where an item is applicable to your facility and changes are complete,
mark "C" in the status column and indicate month and year
implementation was complete, including reference to any supporting
documentation.
(4) Where an item is applicable to your facility and is not fully
implemented, provide your projected implementation date (month and
year) and a short note identifying the outstanding item (e.g.,
hardware, procedures, training, Technical Specifications). Mark "I"
for incomplete.
(5) Where a USI resolution was only recently issued, such as A-40 and A-47,
and you are evaluating your response, identify expected response date
and indicate "E" in the status column.
.
ENCLOSURE 2
ISSUES SUMMARIES FOR UNRESOLVED SAFETY ISSUES
1. USI NO. A-1 TITLE: Water Hammer
This Unresolved Safety Issue (USI) was resolved in March 1984, with the
publication of NUREG-0927, "Evaluation of Water Hammer in Nuclear Power
Plants- Technical Findings Relevant to Unresolved Safety Issue A-1." Also
on March 15, 1984, the EDO sent the Commissioners SECY 84-119 titled,
"Resolution of Unresolved Safety Issue A-1, Water Hammer."
In SECY 84-119, the staff concluded that the frequency and severity of water
hammer occurrences had been significantly reduced through (a) incorporation
of design features such as keep-full systems, vacuum breakers, J-tubes, void
detection systems, and improved venting procedures; (b) proper design of
feedwater valves and control systems; and (c) increased operator awareness
and training. Therefore, the resolution of USI A-1 did not involve any
hardware or design changes on existing plants. It did involve Standard
Review Plan (SRP) changes (forward fits) and a comprehensive set of
guidelines and criteria to evaluate and upgrade utility training programs
(per TMI Task Action Plan Item I.A.2.3). In addition, the assumption was
made that for BWRs with isolation condensers (ICs) a reactor-vessel high
water-level feedwater pump trip was in place or being installed. This was
necessary because calculated values had postulated an IC failure by water
hammer that opened a direct pathway to the environment.
2. USI NO. A-2 TITLE: Asymmetric Blowdown Loads in Reactor
Coolant System
This USI was resolved in January 1981 with the publication of NUREG-0609,
"Asymmetric Blowdown Loads on PWR Primary Systems."
In October 1975, the NRC notified each operating PWR licensee of a potential
safety problem concerning the fact that asymmetric LOCA loads had not been
considered in the design of any PWR piping system. In June 1976 the NRC
informed each PWR licensee that it was required to reassess the reactor
vessel support design of its facility. The staff expanded the scope of the
problem in January 1978 with a request for additional information to all PWR
licensees. NUREG-0609 provided guidance for these analyses. For operating
PWRs, Multi-Plant Action (MPA) Item D-10 was established by NRC's Division
of Licensing for implementation purposes.
During the course of the work on USI A-2, it was demonstrated that there
were only a very limited number of break locations which could give rise to
significant loads. Subsequently, after substantial new technical work, it
was demonstrated that pipes would leak before break and that new fracture
mechanics techniques for the analyzing of piping failures assured adequate
protection against failures in primary system piping in PWRs (Generic Letter
84-04). This was reflected in a revision of General Design Criteria (GDC)-4
(Appendix A to 10 CFR Part 50) published in the Federal Register in final
form on April 11, 1986, and in a subsequent revision to GDC-4 published in
the Federal Register
.
- 2 -
on July 23, 1986. In addition, it has also been satisfactorily demonstrated
in the course of the A-2 effort that there is a very low likelihood of
simultaneous pipe loading with both LOCA and safety shutdown earthquake
(SSE) loads. Therefore, the last revision of GDC-4 represented the final
technical action of NRC regarding the issue of asymetric blowdown loads
issue in PWRs primary coolant main loop piping.
3. USI NO. A-3,4,5 TITLE: Steam Generator Tube Integrity
USIs A-3, 4, and 5, were resolved in September 1988 with the publication of
NUREG-0844 "NRC Integrated Program for the Resolution of Unresolved Safety
Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." USIs
A-3, A-4, and A-5 did not result in new generic requirements for industry in
view of the small potential for reducing risk.
Steam generator tube integrity was designated an unresolved safety issue in
1978 after it became apparent that steam generator tubes were subject to
widespread degradation, frequent leaks, and occasional ruptures (i.e., gross
failures). USI Task Action Plans A-3, A-4, and A-5 were established to
evaluate the safety significance of these problems for Westinghouse,
Combustion Engineering, and Babcock & Wilcox steam generators, respectively.
These studies were later combined into a single effort because PWR vendors
were all experiencing many of the same problems.
NUREG-0844 provides a generic risk assessment that indicates that risk from
steam generator tube rupture (SGTR) events is not a significant contributor
to the total risk at a given site, nor to the total risk to which the
general public is routinely exposed. This finding is considered indicative
of the effectiveness of licensee programs and regulatory requirements for
ensuring steam generator tube integrity in accordance with 10 CFR Part 50,
Appendices A and B.
NUREG-0844 also identifies a number of staff-recommended actions that can
further improve the effectiveness of licensee programs in ensuring the
integrity of steam generator tubes and in mitigating the consequences of a
SGTR. As part of the integrated program, the staff issued Generic Letter
85-02 encouraging licensees of PWRs to upgrade their programs, as necessary,
to meet the intent of the staff-recommended actions; however, such
recommended actions do not constitute NRC requirements. The staff's
assessment of licensee responses to Generic Letter 85-02 was provided to the
Commission in SECY 86-97.
4. USI NO. A-6 TITLE: Mark I Containment Short-Term Program
This USI was resolved in December 1977 with the publication of NUREG-0408,
"Mark I Containment Short-Term Program Safety Evaluation Report."
The objectives of the Mark I short-term program were: (a) to examine the
containment system of each BWR facility with a Mark I containment design to
verify that it would maintain its integrity and functional capability when
subjected to the most probable hydrodynamic loads induced by a postulated
.
- 3 -
design-basis LOCA, and (b) to verify that licensed Mark I BWR facilities
could continue to operate safely, without undue risk to the public health
and safety until such time as a methodical, comprehensive long-term program
is conducted.
The NRC staff used a safety factor of at least two to failure for the
weakest structural or mechanical component in the Mark I containment system
in judging that containment integrity and functions would be assured under
most probable design-basis LOCA-induced hydrodynamic loads.
As indicated in NUREG-0408, the staff required full implementation of the
calculation of the hydrodynamic loads and structural analysis as an interim
measure until complete implementation of the long-term program had been
achieved. In NUREG-0408 the staff concluded that the objectives of the
Short-Term Program had been satisfied, thus documenting the basis for
resolving this safety issue. This issue is considered complete for all
affected BWRs.
5. USI NO. A-7 TITLE: Mark I Long-Term Program
This USI was resolved in August 1982 with the publication of Supplement 1 to
NUREG-0661, "Safety Evaluation Report, Mark I Containment Long-Term Program"
and Standard Review Plan Section 6.2.1.1.C. For operating BWRs, MPA D-01
was established for implementation purposes.
The focus of this USI was the suppression pool hydrodynamic loads,
associated with a postulated LOCA, which had not explicitly been included in
the original Mark I containment design. The issue was identified during
large-scale testing of a Mark III containment design. The staff addressed
this issue in NUREG-0661, pub-lished in July 1980, and in Supplement 1 to
NUREG-0661, published in August 1982.
The objective of the long-term program (LTP) was to establish the
design-basis loads that are appropriate for the anticipated life of each
Mark I BWR facility and to restore the originally intended design-safety
margins for each Mark I containment system. The principal thrust of the LTP
was the development of generic methods for defining suppression pool
hydrodynamic loadings and the associated structural assessment techniques
for the Mark I configuration. On the basis of experimental and analytical
programs conducted by the Mark I Owners Group, it was determined that the
hydrodynamic load definition procedures, with some modifications defined in
NUREG-0661, provided a conservative estimate of these loading conditions.
Thus, the requirements associated with this USI were concerned with the
structural assessment of Mark I containments and related structures to the
hydrodynamic loads defined by the staff in the LTP.
In January 1981, the staff issued "Orders For Modification of License and
Grant of Extension of Exemptions" to each licensee of a Mark I plant. The
orders required the licensees to assess the suppression pool hydrodynamic
loads in accordance with General Electric documents and NUREG-0661 on a
defined schedule. For some plants, the implementation schedule was extended
by a subsequent order.
.
- 4 -
6. USI NO. A-8 TITLE: Mark II Containment Pool Dynamic Loads
This USI was resolved in August 1981 with the publication of NUREG-0808,
"Mark II Containment Program Load Evaluation and Acceptance Criteria," and
Standard Review Plan (SRP) Section 6.2.1.1C. The requirement is that the 11
BWRs having the Mark II containment shall meet the requirements of GDC 16.
As stated in NUREG-0808, the original design of the Mark II containment
system considered only those loads normally associated with design-basis
accidents that were known at the time. These included pressure and
temperature loads associated with a LOCA, seismic loads, dead loads, jet
impingement loads, hydrostatic loads due to water in the suppression
chamber, overload pressure test loads, and construction loads. However,
since the establishment of the original design criteria, additional loading
conditions were identified that must be considered for the
pressure-suppression containment-system design.
In the course of performing large-scale testing of an advanced design
pressure-suppression containment (Mark III), and during inplant testing of
Mark I containments, new suppression-pool hydrodynamic loads were identified
that had not been included explicitly in the original Mark II
containment-design basis. These additional loads result from dynamic
effects of drywell air and steam being rapidly forced into the suppression
pool during a postulated LOCA and from suppression-pool response to
safety/relief valve (SRV) operation; these are generally associated with
plant transient operating conditions. Because these new hydrodynamic loads
had not been considered, the NRC staff determined that a detailed
reevaluation of the Mark II containment system was required.
The issuance of NUREG-0808, NUREG-0802, "Safety Relief Valve Quencher Loads:
Evaluation for BWR Mark II and III Containments," and NUREG-0487, "Mark II
Containment Lead Plant Program Load Evaluation and Acceptance Criteria,"
documented acceptable methods for calculating the hydrodymanic loads
associated with plant transient conditions. Specifically, the loads
referenced in these NRC staff reports, as modified by the acceptance
criteria, constituted the resolution of USI A-8. SRP Section 6.2.1 has been
modified to reflect the applicability of these reports to Mark II
containment evaluations.
Implementation is believed to be complete for all Mark II BWRS. As part of
the licensing process, the staff required that the applicants utilize the
new calculation methodology defined in the reference documents before a full
power license was issued.
7. USI NO. A-9 TITLE: Anticipated Transient Without Scram (ATWS)
per 10 CFR 50.62
ISSUES SUMMARY:
This USI was resolved in June 1984 with the publication of a final rule
(10 CFR 50.62) to require improvements in plants to reduce the likelihood of
failure of the reactor protection system (RPS) to shut down the reactor
following anticipated transients and to mitigate the consequences of an
anticipated transient without scram (ATWS) event.
.
- 5 -
The rule includes the following design-related requirements: 50.62(C)(1),
diverse and independent auxiliary feedwater initiation and turbine trip for
all PWRs; 50.62(C)(2), diverse scram systems for CE and B&W reactors;
50.62(C)(3) alternate rod injection (ARI) for BWRs; 50.62(C)(4); standby
liquid control system (SLCS) for BWRs; and 50.62(C)(5), automatic trip of
recirculation pumps under conditions indicative of an ATWS for BWRs.
Information requirements and an implementation schedule are also specified.
8. USI NO. A-10 TITLE: BWR Feedwater Nozzle Cracking
This issue was resolved in November 1980 with the publication of NUREG-0619,
"BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking."
MPA B-25 was established by NRC's Division of Licensing for implementation
purposes.
Inspections of operating BWRs conducted up to April 1978 revealed cracks in
the feedwater nozzles of 20 reactor vessels. It was determined that
cracking was due to high-cycle fatigue caused by fluctuations in water
temperature within the vessel in the nozzle region.
By letter dated November 13, 1980, Darrell G. Eisenhut provided licensees
with a copy of NUREG-0619. The letter stated that NUREG-0619 provided the
resolution of the staff's generic technical activity USI A-10, which
resulted from the inservice discovery of cracking in feedwater nozzles and
control rod drive return line nozzles. NUREG-0619 describes the technical
issues, General Electric and staff studies and analyses, and the staff's
positions and requirements. Licensees were required to respond, pursuant to
10 CFR 50.54(f), that they would meet implementation dates indicated in
NUREG-0619.
Generic Letter 81-11 was subsequently issued to provide technical
clarification to the November 13, 1980 letter, to clarify that it had been
sent to PWR licensees for information only, and that no response was
required from PWR licensees.
9. USI NO. A-11 TITLE: Reactor Vessel Materials Toughness
This USI was resolved in October 1982 with the publication of NUREG-0744,
"Pressure Vessel Material Fracture Toughness.". NUREG-0744 was issued by
Generic Letter 82-26 and provided only a methodology to satisfy the require-
ments of 10 CFR Part 50, Appendix G. No licensee response to Generic Letter
82-26 was required.
Because of the remote possibility that nuclear reactor pressure vessels
designed to the ASME Boiler and Pressure Vessel Code would fail, the design
of nuclear facilities does not provide protection against reactor vessel
failure. Prevention of reactor vessel failure depends primarily on
maintaining the reactor vessel material fracture toughness at levels that
will resist brittle fracture during plant operation. At service times and
operating conditions typical of current operating plants, reactor vessel
fracture toughness properties provide adequate margins of safety against
vessel failure; however,
.
- 6 -
as plants accumulate more and more service time, neutron irradiation reduces
the material fracture toughness and initial safety margins.
Appendix G to 10 CFR Part 50 requires that the Charpy upper shelf energy
throughout the life of the vessel be no less than 50 ft-lb unless it is
demonstrated that lower values will provide margins of safety against
failure equivalent to those provided by Appendix G of the ASME code. USI
A-11 was initiated to address the staff's concern that some vessels were
projected to have beltline materials with Charpy upper shelf energy less
than 50 ft-lb.
NUREG-0744 provides a method for evaluating reactor vessel materials when
their Charpy upper shelf energy is predicted to fall below 50 ft-lb. Plants
will use the prescribed method when analysis of irradiation damage predicts
that the charpy upper shelf energy is below 50 ft-lb.
10. USI NO. A-12 TITLE: Potential of Low Fracture Toughness and
Lamellar Tearing in PWR Steam Generator and
Reactor Coolant Pump Supports
This USI was resolved in October 1983 with the publication of NUREG-0577,
"Potential of Low Fracture Toughness and Lamellar Tearing in PWR Steam
Generator and Reactor Coolant Pump Supports." The resolution contained no
backfit requirements; it only applied to plants with a new construction
permit issued after October 1983. Standard Review Plan Section 5.3.4 was
issued at the same time this USI was resolved.
The concern in this USI, as the title indicates, was the potential of low
fracture toughness of some materials selected for fabrication of steam
generator (SG) and reactor coolant pump (RCP) supports in operating PWRs.
Lamellar tearing was also of concern. Fracture toughness is a measure of a
material's resistance to fracture in the presence of a previously existing
crack. Generally, a material is considered to have adequate fracture
toughness if it can withstand loading to its design limit in the presence of
detectable flaws under stated conditions of stress and temperature.
The modifications to address this USI could involve maintaining minimum
temperature around the supports above its fracture transition temperature,
or total replacement of existing SG and RCP supports with supports
fabricated of material grade which has a higher Charpy upper shelf energy
and a lower transition temperature. Analysis performed for the resolution
of this USI determined that, even with the failure of the SG and RCP
supports, the amount of incremental release of radioactivity would not be
sufficiently high enough to justify any modification in terms of increasing
the toughness of these supports. This conclusion is based on a value-impact
analysis documented in Appendix C of NUREG-0577.
11. USI NO. A-17 TITLE: Systems Interactions in Nuclear Power
Plants
Generic Letter (GL) 89-18, dated September 6, 1989, was sent to all power
reactor licensees and constitutes the resolution of USI A-17. The generic
letter did not require any licensee actions.
.
- 7 -
GL 89-18 had two enclosures which (a) outlined the bases for the resolution
of USI A-17, and (b) provided five general lessons learned from the review
of the overall systems interaction issue. The staff anticipated that
licensees would review this information in other programs, such as the
Individual Plant Examination (IPE) for Severe Accident Vulnerabilities.
Specifically, the staff expected that insights concerning water intrusion
and flooding from internal sources, as described in the appendix to
NUREG-1174, would be considered in the IPE program. Also considered in the
resolution of this USI was the expectation that licensees would continue to
review information on events at operating nuclear power plants in accordance
with the requirements of TMI Task Action Plan Item I.C.5 (NUREG-0737).
12. USI NO. A-24 TITLE: Qualification of Class 1E Equipment
This USI was resolved in July 1981 with the publication of NUREG-0588,
Revision 1, "Interim Staff Position on Environmental Qualification of
Safety-Related Electrical Equipment." Part I of the report is the original
NUREG-0588 that was issued for comment; that report, in conjunction with the
Division of Operating Reactor (DOR) Guidelines, was endorsed by a Commission
Memorandum and Order as the interim position on this subject until "final"
positions were established in rule making. On January 21, 1983 the
Commission amended 10 CFR 50.49 (the rule), effective February 22, 1983, to
codify existing qualification methods in national standards, regulatory
guides, and certain NRC publications, including NUREG-0588.
The rule is based on the DOR Guidelines and NUREG-0588. These provide
guidance on (a) how to establish environmental service conditions, (b) how
to select methods which are considered appropriate for qualifying the
equipment in different areas of the plant, and (c) such other areas as
margin, aging, and documentation. NUREG-0588 does not address all areas of
qualification; it does supplement, in selected areas, the provisions of the
1971 and 1974 versions of IEEE Standard 323. The rule recognizes previous
qualification efforts completed as a result of Commission Memorandum and
Order CLI-80-21 and also reflects different versions IEEE 323, dependent on
the date of the construction permit Safety Evaluation Report (SER).
Therefore, plant-specific requirements may vary in accordance with the rule.
In summary, the resolution of A-24 is embodied in 10 CFR 50.49. A measure
of whether each licensee has implemented the resolution of A-24 may
therefore be found in the determination of compliance with 10 CFR 50.49.
This was addressed by 72 SERs for operating plants issued shortly after
publication of the rule and subsequently in operating license reviews
pursuant to Standard Review Plan Section 3.11. This was further addressed
by the first-round environmental qualification inspections conducted by the
NRC.
13. USI NO. A-26 TITLE: Reactor Vessel Pressure Transient
Protection
This USI was resolved in September 1978 with the publication of NUREG-0224,
"Reactor Vessel Pressure Transient Protection for PWRs," and Standard Review
Plan Section 5.2. The licensees of all operating PWRs were requested to
.
- 8 -
provide an overpressure prevention system that could be used whenever the
plants were in startup or shutdown conditions. The issue affected all
operating and future plants, and the staff established MPA B-04 for
implementing the solution at operating PWRs.
Since 1972, there have been numerous reported incidents of pressure
transients in PWRs where technical specification pressure and temperature
limits have been exceeded. The majority of these events occurred while the
reactors were in a solid-water condition during startup or shutdown and at
relatively low reactor vessel temperatures. Since the reactor vessels have
less toughness at lower temperatures, they are more susceptible to brittle
fracture under these conditions than at normal operating temperatures. In
light of the frequency of the reported transients and the associated
potential for vessel damage, the NRC staff concluded that measures should be
taken to minimize the number of future transients and reduce their severity.
Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor
Vessel Materials and its Impact on Plant Operations," was published July 12,
1988. This generic letter provides guidance regarding review of pressure-
temperature limits and indicates that licensees may have to revise low-
temperature-overpressure protection setpoints.
14. USI NO. A-31 TITLE: Residual Heat Removal Shutdown Requirements
This USI was resolved in May 1978 with the publication of Standard Review
Plan (SRP) Section 5.4.7. Only those plants expected to receive an
operating license after January 1, 1979 were affected by this resolution.
The USI involved establishment of criteria for the design and operation of
systems necessary to take a power reactor from normal operating conditions
to cold shutdown.
SRP Section 5.4.7 stated that, for purposes of implementation, plants would
be divided into three classes: Class 1 would require full compliance for
Construction Permit (CP) or Preliminary Design Approval (PDA) applications
which were docketed on or after January 1, 1978. Class 2 required a partial
implementation for all plants for which CP or PDA applications were docketed
before January 1, 1978, and for which an Operating License (OL) issuance was
expected on or after January 1, 1979. Class 3 affected all operating
reactors and all other plants for which issuance of the OL was expected
before January 1, 1979. The extent to which Class 3 plants would require
implementation was based on the combined staff review of related plant
features. In general, the outcome of these evaluations were that only
plants receiving an OL after January 1, 1979 were affected by this USI
resolution, and there were no backfits to operating plants that had received
an operating license before January 1, 1979.
15. USI NO. A-36 TITLE: Control of Heavy Loads, Phases I & II
This USI was resolved in July 1980 with the publication of NUREG-0612,
"Control of Heavy Loads at Nuclear Power Plants," and Standard Review Plan
(SRP) Section 9.1.5. The staff established MPAs C-10 and C-15 for the
implementation of Phases I and II, respectively, of the resolution of this
issue at operating plants.
.
- 9 -
In nuclear power plants, heavy loads may be handled in several plant areas.
If these loads were to drop in certain locations in the plant, they may
impact spent fuel, fuel in the core, or equipment that may be required to
achieve safe shutdown and continue decay heat removal. USI A-36 was
established to systematically examine staff licensing criteria and the
adequacy of measures in effect at operating plants, and to recommend
necessary changes to ensure the safe handling of heavy loads. The
guidelines proposed in NUREG-0612 include definition of safe load paths, use
of load handling procedures, training of crane operators, guidelines on
slings and special lifting devices, periodic inspection and maintenance for
the crane, as well as various alternatives.
By Generic Letters dated December 22, 1980, and February 3, 1981 (Generic
Letter 81-07), all utilities were requested to evaluate their plants against
the guidance of NUREG-0612 and to provide their submittals in two parts:
Phase I (six month response) and Phase II (nine month response). Phase I
responses were to address Section 5.1.1 of NUREG-0612 which covered the
following areas:
1. Definition of safe load paths
2. Development of load handling procedures
3. Periodic inspection and testing of cranes
4. Qualifications, training and specified conduct of operators
5. Special lifting devices should satisfy the guidelines of ANSI
N14.6.6.
6. Lifting devices that are not specially designed should be
installed and used in accordance with the guidelines of ANSI B30.9
7. Design of cranes to ANSI B30.2 or CMAA-70
Phase II responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612
which covered the need for electrical interlocks/mechanical stops, or
alternatively, single-failure-proof cranes or load drop analyses in the
spent fuel pool area (PWR), containment building (PWR), reactor building
(BWR), other areas and the specific guidelines for single-failure-proof
handling systems.
As stated in Generic Letter 85-11, "Completion of Phase II of 'Control of
Heavy Loads at Nuclear Power Plants' - NUREG-0612," all licensees have
completed the requirement to perform a review and submit a Phase I and a
Phase II report. Based on the improvements in heavy loads handling obtained
from implementation of NUREG-0612 (Phase I), further action was not required
to reduce the risks associated with the handling of heavy loads. Therefore,
a detailed Phase II review of heavy loads was not necessary and Phase II was
considered completed.
While not a requirement, NRC encouraged the implementation of any actions
identified in Phase II regarding the handling of heavy loads that were
considered appropriate.
16. USI NO. A-39 TITLE: Determination of Safety Relief Valve Pool
Dynamic Loads and Temperature Limits
This USI was resolved with the publication of Standard Review Plan (SRP)
Section 6.2.1.1.C, in October 1982. In addition, NUREGs 0763, 0783 and 0802
were issued for Mark I, Mark II, and Mark III containments, respectively.
BWR plants are equipped with safety/relief valves (SRVs) to protect the
reactor
.
- 10 -
from overpressurization. Plant operational transients, such as turbine
trips, will actuate the SRV. Once the SRV opens, the air column within the
partially submerged discharge line is compressed by the high-pressure steam
released from the reactor. The compressed air discharged into the
suppression pool produces high-pressure bubbles. Oscillatory expansion and
contraction of these bubbles create hydrodynamic loads on the containment
structures, piping, and equipment inside containment.
NUREG-0802 presents the results of the staff's evaluation of SRV loads. The
evaluation, however, is limited to the quencher devices used in Mark II and
III containments. With respect to Mark I containments, the SRV acceptance
criteria are presented in NUREG-0661, "Safety Evaluation Report, Mark I
Containment and Long-Term Program," and are dealt with as part of USI A-7.
SRP Section 6.2.1.1.C addresses the applicable review criteria, since all
Mark II and III containment designs are understood to have completed their
operating license (OL) reviews subsequent to resolution of this USI and
reflection of the resolution in the SRP.
17. USI NO. A-40 TITLE: Seismic Design Criteria
The staff has resolved USI A-40 as documented in NUREG/CR-5347, "Recommenda-
tions for Resolution of Public Comments on USI A-40," issued in June 1989,
and NUREG-1233, "Regulatory Analysis for USI A-40," issued in September
1989.
For plants not covered under the scope of USI A-46, "Seismic Qualification
of Equipment in Operating Plants," the staff concluded that tanks in plants
that were subject to licensing review by the staff after 1984 had been
reviewed to current requirements and found acceptable. For tanks in plants
reviewed during 1980-1984, the staff identified four plant sites (six units)
that were not explicitly reviewed to current requirements. The four plants
(Callaway 1/2, Wolf Creek, Shearon Harris 1, and Watts Bar 1/2) are being
handled on a plant-specific basis.
USI A-40 originated in 1977. The basic objectives were (a) to study the
seismic design criteria, (b) to quantify the conservatism associated with
the criteria, and (c) to recommend modifications to the Standard Review Plan
(SRP) if changes are justified. Lawrence Livermore National Laboratory
(LLNL) completed the study and published its findings in NUREG/CR-1161,
"Recommended Revisions to USNRC - Seismic Design Criteria," dated May 1980.
The report recommended specific changes to the Standard Review Plan (SRP).
NRC staff reviewed the report and developed some other changes that would
reflect the present state of seismic design practices. The resulting SRP
changes were issued for public comment in June 1988, and the final SRP
changes are to be published in October 1989.
The major SRP changes consist of (a) clarification of development of site
specific spectra, (b) justification for use of single synthetic time-history
by power spectral density function, (c) location and reductions of input
ground motion for soil structure interaction, and (d) design of above-ground
vertical tanks. Except for item (d), these items do not constitute any
additional requirements for current licenses and applications, and thus, no
backfitting is being required for these items. However, the revised
provisions could be used for margin studies and reevaluations or individual
plant examination for external events (IPEEE).
.
- 11 -
The participant utilities in the Seismic Qualification Utility Group (SQUG)
agreed to implement the changed criteria for flexible vertical tanks for
their plants. For the four plants where this issue has to be resolved on an
individual basis a 10 CFR 50.54(f) request-for-information letter has been
sent to the affected utilities. If the information received indicates that
large above-ground vertical tanks do not meet the new criteria,
plant-specific backfits will be considered.
18. USI NO. A-42 TITLE: Pipe Cracks in Boiling Water Reactors
This USI was resolved in February 1981 with the publication of NUREG-0313,
Revision 1, "Technical Report on Material Selection and Processing
Guidelines for BWR Coolant Pressure Boundary Piping." That NUREG document
was issued to all holders of BWR operating licenses or construction permits
and to all applicants for BWR operating licenses. The staff established MPA
B-05 for implementation of the resolution at operating plants.
Pipes have cracked in the heat-affected zones of welds in primary system
piping in BWRs since mid-1960. These cracks have occurred mainly in Type
304 stainless steel, which is the type used in most operating BWRs. The
major problem is recognized to be intergranular stress corrosion cracking
(IGSCC) of austenitic stainless steel components that have been made
susceptible to this failure by being "sensitized," either by post-weld heat
treatment or by sensitization of a narrow heat affected zone near welds.
"Safe ends" that have been highly sensitized by furnace heat treatment while
attached to vessels during fabrication were found to be susceptible to IGSCC
in the late 1960s. Most of the furnace-sensitized safe ends in older plants
have been removed or clad with a protective material, and only a few BWRs
still have furnace-sensitized safe ends in use. Most of these, however, are
in smaller diameter lines.
Cracks reported before 1975 occurred primarily in 4-inch-diameter
recirculation loop bypass lines and in 10-inch-diameter core spray lines.
Cracking is most often detected during inservice inspections using
ultrasonic test techniques. Some piping cracks have been discovered as a
result of primary coolant leaks.
NUREG-0313, Revision 1 provided the NRC staff's revised acceptable methods
for reducing the IGSCC susceptibility of BWR code class 1, 2, and 3 pressure
boundary piping of sizes identified above and safe ends. In addition, it
provided the requirements for augmented inservice inspection of piping with
nonconforming materials.
As a result of further IGSCC degradations in larger piping, the staff
provided licensees with additional requirements in several NRC
communications (i.e., Bulletins 82-03, 83-2, and 84-11). The long-term
resolution of IGSCC in BWR piping (including the scope of A-42) was provided
in NUREG-0313, Revision 2 which was transmitted to all holders of BWR
operating licenses via Generic Letter 88-01.
.
- 12 -
19. USI NO. A-43 TITLE: Containment Emergency Sump Performance
The resolution of this USI was presented to the Commission in October 1985
in SECY-85-349. NUREG-0897, Revision 1, "Containment Emergency Sump
Performance," presents the results of the staff's technical findings. These
findings established a need to revise current licensing guidance on these
matters. RG 1.82 Revision 0 and Standard Review Plan Section 6.2.2,
"Containment Heat Removal Systems" were revised to reflect this new
guidance. No licensee actions were required.
Initially, an issue existed concerning the availability of adequate
recirculation cooling water following a loss-of-coolant accident (LOCA) when
long-term recirculation of cooling water from the PWR containment sump, or
the BWR residual heat removal system (RHR) suction intake, must be initiated
and maintained to prevent core melt.
The technical concerns evaluated under USI A-43 were: (a) post-LOCA adverse
conditions resulting from potential vortex formation and air ingestion and
subsequent pump failure, (b) blockage of sump screens with LOCA generated
insulation debris causing inadequate net positive suction head (NPSH) on
pumps, and (c) RHR and containment spray pumps inoperability due to possible
air, debris, or particulate ingestion on pump seal and bearing systems.
This revised guidance applies only to future construction permits,
preliminary design approvals, final design approvals, standardized designs,
and applications for licenses to manufacture. The staff performed a
regulatory analysis to determine if this new guidance should be applied to
operating plants. The results of this analysis were reported in NUREG-0869
Revision 1, "USI A-43 Regulatory Analysis," issued in October 1985. The
staff concluded that the regulatory analysis does not support any new
generic requirements for present licensees to perform debris assessments.
20. USI NO. A-44 TITLE: Station Blackout
This USI was resolved in June 1988 with the publication of a new rule (10
CFR 50.63) and Regulatory Guide 1.155.
Station blackout means the loss of offsite ac power to the essential and
nonessential electrical buses concurrent with turbine trip and the
unavailability of the redundant onsite emergency ac power systems.
WASH-1400 showed that station blackout could be an important risk
contributor, and operating experience has indicated that the reliability of
ac power systems might be less than originally anticipated. For these
reasons station blackout was designated as a USI in 1980. A proposed rule
was published for comment on March 21, 1986. A final rule, 10 CFR 50.63,
was published on June 21, 1988 and became effective on July 21, 1988.
Regulatory Guide 1.155 was issued at the same time as the rule and
references an industry guidance document, NUMARC-8700. In order to comply
with the A-44 resolution, licensees will be required to:
.
- 13 -
o maintain onsite emergency ac power supply reliability above a minimum
level
o develop procedures and training for recovery from a station blackout
o determine the duration of a station blackout that the plant should be
able to withstand
o use an alternate qualified ac power source, if available, to cope with
a station blackout
o evaluate the plant's actual capability to withstand and recover from a
station blackout
o backfit hardware modifications if necessary to improve coping ability
Section 50.63(c)(1) of the rule required each licensee to submit a response
including the results of a coping analysis within 270 days from issuance of
an operating license or the effective date of the rule, whichever is later.
21. USI NO. A-45 TITLE: Shutdown Decay Heat Removal Requirements
USI A-45 was resolved by SECY 88-260, "Shutdown Decay Heat Removal Require-
ments (USI-A-45)," issued September 13, 1988, without imposing any new
licensing requirements other than the Individual Plant Examination (IPE), as
described below. At the same time the staff issued NUREG-1289, "Regulatory
and Backfit Analysis: USI A-45." Since all of the significant USI A-45
results have been found to be highly plant specific, the Commission decided
it was not appropriate to propose a single generic corrective action to be
applied uniformly to all plants.
The Commission is currently implementing the Severe Accident Policy (50 FR
32138) and will require all plants presently operating or under construction
to undergo a systematic examination termed the IPE. The reason for this
examination is to identify any plant-specific vulnerabilities to severe
accidents. The IPE analysis intends to examine and understand the plant
emergency procedures, design, operations, maintenance, and surveillance, in
order to identify vulnerabilities. The analysis will examine both the decay
heat removal systems and those systems used for other related functions.
This includes CE plants without power-operated relief valves.
NRC has decided to subsume A-45 into the IPE program as the most effective
way of achieving resolution of specific plant concerns associated with A-45.
22. USI NO. A-46 TITLE: Seismic Qualification of Equipment in
Operating Plants
USI A-46 was resolved with the issuance of GL 87-02 on February 19, 1987,
which endorsed the approach of using the seismic and test experience data
proposed by the Seismic Qualification Utility Group (SQUG) and Electric
Power Research Institute (EPRI). This approach was endorsed by the Senior
Seismic Review and Advisory Panel (SSRAP) and approved by the NRC staff.
.
- 14 -
The scope of the review was narrowed to equipment required to bring each
affected plant to hot shutdown and maintain it there for a minimum of 72
hours. The review includes a walkthrough of each plant which is required to
inspect equipment. Evaluation of equipment will include: (a) adequacy of
equipment anchorage; (b) functional capability of essential relays;
(c) outliers and deficiencies (i.e., equipment with non-standard
configurations); and (d) seismic systems interation.
As an outgrowth of the Systematic Evaluation Program (SEP), the need was
identified for reassessing design criteria and methods for the seismic
qualification of mechanical equipment and electrical equipment. Therefore,
the seismic qualification of the equipment in operating plants must be
reassessed to ensure the ability to bring the plant to a safe shutdown
condition when subject to a seismic event. The objective of this issue was
to establish an explicit set of guidelines that could be used to judge the
adequacy of the seismic qualification of mechanical and electrical equipment
at operating plants in lieu of attempting to backfit current design criteria
for new plants.
Generic Letter 87-02 with associated guidance, required all affected
utilities to evaluate the seismic adequacy of their plants. The specific
requirements and approach for implementation are being developed jointly by
SQUG and the staff on a generic basis before individual member utilities
proceed with plant-specific implementation.
23. USI NO. A-47 TITLE: Safety Implication of Control Systems in
LWR Nuclear Power Plants
USI A-47 was resolved September 20, 1989, with the publication of Generic
Letter (GL) 88-19.
The generic letter states:
"The staff has concluded that all PWR plants should provide
automatic steam generator overfill protection, all BWR plants
should provide automatic reactor vessel overfill protection,
and that plant procedures and technical specifications for all
plants should include provisions to verify periodically the
operability of the overfill protection and to assure that
automatic overfill protection is available to mitigate main
feedwater overfeed events during reactor power operation.
Also, the system design and setpoints should be selected with
the objective of minimizing inadvertent trips of the main
feedwater system during plant startup, normal operation, and
protection system surveillance. The Technical Specifications
recommendations are consistent with the criteria and the risk
considerations of the Commission Interim Policy Statement on
Technical Specification Improvement. In addition, the staff
recommends that all BWR recipients reassess and modify, if
needed, their
.
- 15 -
operating procedures and operator training to assure that the
operators can mitigate reactor vessel overfill events that may
occur via the condensate booster pumps during reduced system
pressure operation."
Also, page 2 of the generic letter provides for additional actions for CE and
B&W plants. The generic letter provides amplifying guidance for licensees.
The generic letter requires that licensees provide NRC with their schedule and
commitments within 180 days of the letter's date. The implementation schedule
for actions on which commitments are made should be prior to startup after the
first refueling outage, but no later than the second refueling outage,
beginning 9 months after receipt of the letter.
24. USI NO. A-48 TITLE: Hydrogen Control Measures and Effects of
Hydrogen Burns on Safety Equipment
The NRC staff concluded April 19, 1989, that USI A-48 is resolved, as stated
in SECY 89-122.
USI A-48 was initiated as a result of the large amount of hydrogen generated
and burned within containment during the Three Mile Island (TMI) accident.
This issue covers hydrogen control measures for recoverable degraded core
accidents for all BWRs and those PWRs with ice condenser containments.
Extensive research in this area has led to significant revision of the Com-
mission's hydrogen control regulations, given in 10 CFR 50.44, published
December 2, 1981.
10 CFR 50.44 requires inerting of BWR Mark I and Mark II containments as a
method for hydrogen control. The BWR Mark I and Mark II reactor containments
have operated for a number of years with an inerted atmosphere (by addition of
an inert gas, such as nitrogen) which effectively precludes combustion of any
hydrogen generated. USI A-48 with respect to BWR Mark I and II containments
is not only resolved but understood to be fully implemented in the affected
plants.
The rule for BWRs with Mark III containments and PWRs with ice condenser
containments was published on January 25, 1985. The rule required that these
plants be provided with a means for controlling the quantity of hydrogen
produced, but did not specify the control method. In addition, the task
action plan for USI A-48 provided for plant-specific reviews of lead plants
for reactors with Mark III and ice condenser containments. Sequoyah was
chosen as the lead plant for ice condenser containments and Grand Gulf for
Mark III containments. Both of the lead plant licensees chose to install
igniter-type systems which would burn the hydrogen before it reached
threatening concentrations within the containment. Final design igniter
systems have been installed not only in both lead plants, Sequoyah and Grand
Gulf, but in all other ice condenser and Mark III plants as well. The staff's
safety evaluations of the final analyses required to be submitted by these
licensees by the rule are scheduled for completion in 1989.
.
- 16 -
Large dry PWR containments were excluded from USI A-48 because they have a
greater ability to accommodate the large quantities of hydrogen associated
with a recoverable degraded core accident than the smaller Mark I, II, III and
ice condenser containments. However, this issue has continued to be
considered and, in 1989, hydrogen control for large dry PWR containments was
identified as a high-priority Generic Issue (GI) 121. The resolution of GI
121 is being actively pursued in close coordination with more recent research
findings.
25. USI NO. A-49 TITLE: Pressurized Thermal Shock
The final rule (10 CFR 50.61) on pressurized thermal shock (PTS) was approved
by the Commission in July 1985. Regulatory Guide 1.154, "Format and Content
of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs,"
was later published in February 1987. Thus, this issue was resolved and new
requirements were established, applicable to PWRs only. The rule required
that each operating reactor meet the screening criteria provided in the rule
or provide supplemental analysis to demonstrate that PTS is not a concern for
the facility.
Neutron irradiation of reactor pressure vessel weld and plate materials
decreases the fracture toughness of the materials. The fracture toughness
sensitivity to radiation-induced change is increased by the presence of
certain materials such as copper. Decreased fracture toughness makes it more
likely that, if a severe overcooling event occurs followed by or concurrent
with high vessel pressure, and if a small crack is present on the vessel's
inner surface, that crack could grow to a size that might threaten vessel
integrity.
Severe pressurized overcooling events are improbable since they require
multiple failures and improper operator performance. However, certain
precursor events have happened that could have potentially threatened vessel
integrity if additional failures had occurred and/or if the vessel had been
more highly irradiated. Therefore, the possibility of vessel failure due to a
severe pressurized overcooling event cannot be ruled out.
Page Last Reviewed/Updated Tuesday, March 09, 2021